Approved by Order of the Federal Environmental, Industrial and Nuclear Supervision Service dated February 13, 2017 No. 53 FEDERAL RULES AND REGULATIONS IN THE FIELD OF ATOMIC ENERGY USE "REQUIREMENTS FOR THE CONTENT OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANT UNITS WITH VVER REACTORS" (NP-006-16) I. Purpose and scope 1. These federal rules and regulations in the field of atomic energy use "Requirements for the content of safety analysis reports for nuclear power plant units with VVER reactors" (NP-006-16) (hereinafter - the Requirements) have been developed in accordance with Federal Law dated November 21, 1995 No. 170-FZ "On atomic energy use", Decree of the Government of the Russian Federation dated December 1, 1997 No. 1511 “On approval of the Regulation on development and approval of Federal rules and regulations in the field of atomic energy use” (Collected legislation of the Russian Federation, 1997, No. 49, Art. 5600; 1999, No. 27, Art. 3380; 2000, No. 28, Art. 2981; 2002, No. 4, Art. 325; No. 44, Art. 4392; 2003, No. 40, Art. 3899; 2005, No. 23, Art. 2278; 2006, No. 50, Art. 5346; 2007, No. 14, Art. 1692; No. 46, Art. 5583; 2008, No. 15, Art. 1549; 2012, No. 51, Art. 7203) and establish the requirements for the content of the safety analysis report for a nuclear power plant unit with a VVER reactor as well as for the procedure for its development and maintenance in compliance with the actual state of the nuclear power plant. 2. These Requirements shall be applicable to safety analysis reports for nuclear power plant units with VVER reactors. 3. For nuclear power plant units with the construction licenses issued prior to entry of these Requirements into effect as well as for power units in operation the procedure, time limits and scope of bringing the nuclear power plant unit safety analysis report into compliance with these Requirements shall be defined by the competent authority for state regulation of safety in atomic energy use under the conditions of the issued licenses with due regard for the proposals of the operating organization. 4. The list of abbreviations is given in Appendix 1 hereto. II. Content of the NPP SAR
Transcript
Approved byOrder of the Federal
Environmental, Industrialand Nuclear Supervision Servicedated February 13, 2017 No. 53
FEDERAL RULES AND REGULATIONSIN THE FIELD OF ATOMIC ENERGY USE
"REQUIREMENTS FOR THE CONTENT OF SAFETY ANALYSIS REPORTSFOR NUCLEAR POWER PLANT UNITS WITH VVER REACTORS"
(NP-006-16)
I. Purpose and scope
1. These federal rules and regulations in the field of atomic energy use "Requirements for the content of safety analysis reports for nuclear power plant units with VVER reactors" (NP-006-16) (hereinafter - the Requirements) have been developed in accordance with Federal Law dated November 21, 1995 No. 170-FZ "On atomic energy use", Decree of the Government of the Russian Federation dated December 1, 1997 No. 1511 “On approval of the Regulation on development and approval of Federal rules and regulations in the field of atomic energy use” (Collected legislation of the Russian Federation, 1997, No. 49, Art. 5600; 1999, No. 27, Art. 3380; 2000, No. 28, Art. 2981; 2002, No. 4, Art. 325; No. 44, Art. 4392; 2003, No. 40, Art. 3899; 2005, No. 23, Art. 2278; 2006, No. 50, Art. 5346; 2007, No. 14, Art. 1692; No. 46, Art. 5583; 2008, No. 15, Art. 1549; 2012, No. 51, Art. 7203) and establish the requirements for the content of the safety analysis report for a nuclear power plant unit with a VVER reactor as well as for the procedure for its development and maintenance in compliance with the actual state of the nuclear power plant.
2. These Requirements shall be applicable to safety analysis reports for nuclear power plant units with VVER reactors.
3. For nuclear power plant units with the construction licenses issued prior to entry of these Requirements into effect as well as for power units in operation the procedure, time limits and scope of bringing the nuclear power plant unit safety analysis report into compliance with these Requirements shall be defined by the competent authority for state regulation of safety in atomic energy use under the conditions of the issued licenses with due regard for the proposals of the operating organization.
4. The list of abbreviations is given in Appendix 1 hereto.
II. Content of the NPP SAR
5. Information contained in the NPP SAR shall confirm compliance of the NPP power unit with the requirements of federal rules and regulations in the field of atomic energy use as well as any NPP safety assurance criteria and principles established in the NPP design documentation (hereinafter - the design).
6. In case references to any documents containing the missing information are given in the NPP SAR instead of the information presentation according to these Requirements, such documents shall be provided together with the NPP SAR. Any other documents referenced to in the NPP SAR shall be provided upon the request of the competent authority for state regulation of safety in atomic energy use.
7. The NPP SAR shall consist of the "Introduction" section and 18 chapters namely:
Chapter 1. "General description of the NPP";
Chapter 2. "Characteristics of the NPP region and site";
Chapter 3. "General provisions for design of the NPP buildings, structures, systems, and components";
Chapter 4. "Reactor";
Chapter 5. "Primary circuit and interfaced systems";
Chapter 6. "Steam turbine plant";
Chapter 7. "Control and monitoring";
Chapter 8. "Power supply, communications and warning";
Chapter 9. "Supporting systems of the NPP power unit";
Chapter 13. "Commissioning of the NPP power unit";
Chapter 14. "Operation";
Chapter 15. "Analysis of abnormal operation including design basis and beyond design basis accidents";
Chapter 16. "Safe operation limits and conditions. Operation limits and conditions";
Chapter 17. "Quality assurance";
Chapter 18. "Decommissioning".
The content of the "Introduction" section is presented in Appendix 2 hereto.
The content of Chapters 1-18 of the NPP SAR is presented in Appendix 3 hereto.
8. In case degree of materials completion at the stage of development of the NPP SAR submitted within the package of documents substantiating safety assurance in the course of the NPP construction does not comply with the provisions of these Requirements information contained in the NPP SAR shall reflect the actual state of the NPP SAR development as well as any design documentation used to develop it. In this case the following shall be additionally provided:
information on any planned justifications of the technical solutions used in the design absent as of the NPP SAR submittal in order to give sufficient insight into their technical content;
substantiation of the necessity to apply new technical solutions in the design with indication of their impact on the NPP safety;
schedule for submittal of the missing information.
9. Information on compliance with the requirements of federal rules and regulations in the field of atomic energy use presented in the NPP SAR shall contain documented substantiation of such compliance.
10. In case it is necessary to provide similar information in several chapters of the NPP SAR (or sections within the same chapter) such information shall be presented in one of the NPP SAR chapters (or chapter sections) and references to this information shall be given in any other chapters (or chapter sections).
11. Information on the performed calculation analyses shall confirm sufficiency and completeness of the performed calculation analyses and consideration for all factors affecting the results. Description of software tools mentioned in the NPP SAR shall be provided. Information on validation of these ST shall be presented and in case no ST validation has been performed results of the ST verification through the use of analytical and experimental methods shall be provided. The information on the use of SW in the areas of application specified in validation certificates shall be presented. Data sufficient to perform repeated calculation analysis in case of necessity (diagrams, assumptions, source data) shall be submitted upon the request of the competent authority for state regulation of safety in atomic energy use.
12. The NPP SAR shall be developed by the operating organization in cooperation with the RF and NPP designers and submitted to the competent authority for state regulation of safety in atomic energy use within the package of documents substantiating the NPP safety in the course of siting, design, construction and operation. The NPP SAR shall be agreed with the RF and NPP designers and approved by the operating organization.
13. An individual NPP SAR shall be developed for each power unit of a multi-unit NPP.
14. The NPP SAR consisting of the "Introduction" section, Chapter 1 "General description of the NPP" and Chapter 2 "Characteristics of the NPP region and site" shall be submitted to the competent authority for state regulation of safety in atomic energy use within the package of documents substantiating the NPP safety in the course of siting.
15. The NPP SAR within the package of documents substantiating the NPP safety in the course of construction or operation shall be submitted to the competent authority for state regulation of safety in atomic energy use in full as specified in item 7 hereof.
16. Subsequent to completion of the pilot operation stage of the NPP power unit commissioning the NPP SAR shall be corrected with due regard for the results obtained in the course of precommissioning works, physical start-up, power start-up and pilot operation of the NPP power unit. In this case any information presented in the NPP SAR shall correspond to the actual state of the NPP power unit subsequent to the results of construction, manufacturing, installation and commissioning of the NPP power unit.
17. The NPP SAR shall be formed by separate chapters. In case of a great amount of information in any chapter of the NPP SAR the NPP SAR may be formed by sections and subsections organized as separate books within the chapter.
Complete table of contents of the entire NPP SAR shall be given in the beginning of each individual chapter, section, or subsection.
The NPP (NPP power unit) name, complete name of the NPP SAR and the relevant chapter, section or subsection shall be indicated on the cover of each individual chapter, section and subsection.
III. Maintenance of the NPP SAR in compliance with the actualstate of the NPP
18. The NPP SAR shall be maintained in compliance with the actual NPP state.
19. Any amendments shall be introduced to the NPP SAR by replacement of pages. Introduction of
amendment by corrections in the text of the NPP SAR is not permitted.
When individual pages are replaced in the NPP SAR the reference number of the revision and the date (month, year) of replacement shall be specified on each page in the top right corner on the margins.
A revision sheet shall be placed in the end of each chapter or section and subsection of the NPP SAR.
20. Any amendments introduced to the NPP SAR shall be agreed with the organizations engaged in its development and approved by the operating organization.
Appendix 1to federal rules and regulations
in the field of atomic energy use"Requirements for the content of
safety analysis reportsfor nuclear power plant units with VVER reactors" approved by Order
of the Federal Environmental,Industrial and Nuclear Supervision Service
dated February 13, 2017 No. 53
ABBREVIATIONS
AE - absorbing element
ALA - accident localization area
APCS - automated process control system
ARMS - automated radiation monitoring system
ARSMS - automated radiation situation monitoring system
ASB - automatic switch-over to backup
ASW - air shock wave
BDBA - beyond design basis accident
BRU-A - steam dump valve to atmosphere
BRU-K - steam dump valve into condenser
CD - control device
CPS - control and protection system
CPSS - containment pre-stressing system
CR - control rod
CSS - control safety systems
CW - commissioning works
DGP - dangerous geological processes
ECCS - emergency core cooling system
ECR - emergency control room
EFPP - engineered features of physical protection
EP - emergency protection
EPSS - emergency power supply system
FA - fuel assembly
FE - fuel element
FFS - fresh fuel storage
GFE - uranium-gadolinium fuel element
HPH - high pressure heater
IC - ionization chamber
IE - initiating event
LRW - liquid radioactive waste
LSS - localizing safety system
M&R - maintenance and repair
MBA - material balance area
MCL - minimum controllable level
MCR - main control room
NF - nuclear fuel
NFME - neutron flux monitoring equipment
NI - nuclear installation
NM - nuclear materials
NOCS - normal operation control systems
NPC - neutron and physical characteristics
NPP - nuclear power plant
NPP QAP - nuclear power plant quality assurance program
NPP SAR - nuclear power plant unit safety analysis report
OBE - operating basis earthquake
OSTP - on-site transport packing
PDVCM - present-day crustal motion
PERCP - protected emergency response control post
PHRS - passive heat removal system
PORV - pilot-operated relief valve
PPE - physical protection equipment
PPH - physical protection hardware
PPS - physical protection system
PRZR - pressurizer
PSA - probabilistic safety assessment
PSS - protective safety systems
QAP - quality assurance program
R&D - research and development
RCC - reactor coolant circuit
RCP - reactor coolant pump
RCPS - reactor coolant pump set
RF - reactor facility
RI - reactor Internals
RS - radioactive substances
RW - radioactive waste
SA - supervised area
SC - short circuit
SCC - short-circuit current
SDGS - standby diesel generator station
SFP - Spent Fuel Pool
SG - steam generator
SM - structural materials
SNF - spent nuclear fuel
SNFS - spent nuclear fuel storage
SPA - sanitary-protective area
SR - scientific research
SRD - scientific research and development
SS - safety system
SSE - safe shutdown earthquake
SSS - stress-strain state
ST - software tools
TOR - terms of reference
TP - transport packing
TS - technical specifications
VLLW - very low level waste
VVER - pressurized water reactor
WCR - water chemistry regime
Appendix 2to federal rules and regulations
in the field of atomic energy use"Requirements for the content of
safety analysis reportsfor nuclear power plant units with VVER reactors" approved by Order
of the Federal Environmental,Industrial and Nuclear Supervision Service
dated February 13, 2017 No. 53
REQUIREMENTS FOR THE CONTENT OF THE "INTRODUCTION" SECTION OF THE NPP SAR
General information on the NPP and its design, data on the NPP and RF designers as well as any organizations engaged in the NPP SAR development, on the development stage for the entire NPP design and general characteristics of the NPP SAR shall be given in the "Introduction" section of the NPP SAR.
1. The basis for the NPP design development, siting, construction and operation of the NPP.
Information on resolutions of the Government of the Russian Federation and federal executive bodies serving as the basis for siting, design, construction or operation of the NPP shall be provided.
In this case the following data shall be presented:
availability of positive opinions of the state environmental expert review on the objects of state environmental expert review;
availability of permits for the construction and commissioning of a nuclear facility within the NPP power unit issued by the state atomic energy use controlling authority.
2. General characteristics of the NPP.
General description of the NPP (planned capacity, the number of units, operation modes, RF type, turbine plant and turbine generator types, RF containment type, brief information on the NPP power supply and service water supply systems) shall be provided.
3. Development stage.
Information on the actual stage of design and operational documentation development as of the moment of the NPP SAR formation shall be provided.
4. Information on the NPP SAR developers.
Information on the operating organization, developers of individual independent chapters or sections of the NPP SAR, information on their experience in the area under consideration, availability of licenses issued by the authority for state regulation of safety in atomic energy use shall be presented.
5. Characteristics of the NPP SAR.
Information on compliance of the data presented in the NPP SAR with these Requirements shall be given.
If the NPP design development is at one of the initial stages and thus the information provided in
the NPP SAR does not comply with these Requirements it shall be indicated in this section of the NPP SAR. In this case the schedule of the work completion with indication of the time limits for submittal of the required information shall be additionally presented.
Appendix 3to federal rules and regulations
in the field of atomic energy use"Requirements for the content of
safety analysis reportsfor nuclear power plant units with VVER reactors" approved by Order
of the Federal Environmental,Industrial and Nuclear Supervision Service
dated February 13, 2017 No. 53
REQUIREMENTS FOR THE CONTENT OF NPP SAR CHAPTERS
I. Requirements for the content of Chapter 1 "General description of the NPP"
Information on the NPP representing in brief the content of Chapters 2-18 of the NPP SAR shall be given.
Information presented in Chapter 1 of the NPP SAR shall provide the possibility for the governmental authorities, non-governmental organizations and the public to familiarize with the concept and basic technical solutions for the NPP safety assurance in general without any need to refer to other chapters of the NPP SAR.
1.1. NPP siting conditions.
The following information on the NPP site and its location area shall be provided:
the main hydrological, climatic, meteorological and aerological characteristics of the NPP location area;
average monthly, extreme for the entire observation period, the maximal of the average monthly and average decade temperatures of the air and water in the service water source as well as characteristics of extreme natural impacts;
geological, hydrogeological and seismotectonic characteristics;
seismic intensity in the NPP site location region for the SSE and OBE levels, boundaries of the solid block where no seismic deformations will take place at various earthquakes levels;
characteristics of soils to the depth of not less than 100 m with indication of the distribution of compressible (clayey, sandy) and non-compressible (rocky and semi-rocky) soils;
depth of the first aquifer from the surface and its connections to the surface waters;
information on the population density within the zone with the radius of 30 km around the NPP with due regard for the NPP personnel and the engaged workers;
size of the sanitary-protective area and the supervised area, the protective action planning zone at the initial stage of the accident and the compulsory population evacuation action planning zone, information on the populated localities subject to relocation prior to the NPP commissioning.
1.2. Location plan.
Brief characteristic of the NPP location area and information on the location of enterprises, water
conduits, pumping stations, water reservoirs, irrigating channels, hydroelectric dams, airfields, motor roads and railways with reference to the sanitary-protective area, supervised area, protective action planning zone, and compulsory population evacuation action planning zone shall be given.
Information on the site terrain characteristics and inclinations towards water bodies and brief information on land use shall be provided.
Information on high-voltage power transmission lines of the NPP, access motorways and railways and expected position of residential areas shall be presented.
Information on any enterprises with extreme explosion hazard, fire hazard and releases of toxic and radioactive substances into the environment shall be provided. The NPP location plan on a scale of 1:25000 shall be provided.
1.3. Description of the NPP flow diagram.
The NPP flow diagram with information on the following systems and components shall be provided:
primary circuit;
reactor;
RCP;
SG;
PRZR;
coolant purification systems;
SS;
special-purpose hardware for BDBA management;
fuel pool and its cooling systems;
primary circuit makeup and blowdown system;
steam pipelines;
steam turbine plant;
feedwater system;
systems for heat removal to the ultimate heat sink;
service water supply system;
NPP auxiliary power supply system from external and internal sources.
Systems and components located inside the containment shall be specified at this flow diagram. The list of systems and components indicated on the flow diagram with specification of their basic characteristics shall be given.
Brief description of interaction between systems and components shall be attached to the flow diagram.
1.4. Basic technical characteristics of the NPP.
The following basic technical characteristics of the NPP shall be presented:
number of the NPP units;
service life of the RF, the steam turbine plant, the RF containment;
power and thermal capacity of the NPP;
heating capacity;
capacity factor;
auxiliary power consumption;
fueling;
main parameters of the primary and secondary circuit coolant.
1.5. Grid characteristics.
The principal scheme of the grid where the NPP is to be operated as well as the following information on the grid shall be provided:
voltage and frequency in the grid mains;
state of the grid as of the NPP start-up with indication of the type and capacity of power plants in the grid;
general levels of power consumption and maximal loads in the grid (daily, weekly, seasonal and annual), capacity margins in relation to the maximal loads;
operation modes of the grid automation and protection devices affecting the NPP operation mode;
NPP operation modes associated with disturbances in the grid operation resulting in load reductions up to the auxiliary level.
Information on the expected number of the NPP power supply disturbance cycles with due regard for external impacts shall be provided.
Information on the time for recovery of the auxiliary NPP power supply from the external sources shall be provided for any expected NPP power supply disturbances.
Requirements for the grid reliability indicators shall be specified.
1.6. NPP operation modes.
Information on the main typical NPP operation modes shall be presented.
Information on the acceptable number of normal operation and abnormal operation modes including design basis accidents shall be provided.
1.7. NPP safety assurance concept.
1.7.1. Basic principles and criteria of the NPP safety assurance.
The following information shall be provided in this section:
the basic safety criteria and design limits for various operational states established in the NPP design;
the list of regulatory documents containing the requirements compliance with which is to be confirmed by the NPP safety analysis;
information on the use of inherent self-protection properties in the NPP design with indication of the means to implement them;
information on the NPP safety assurance due to systematic implementation of the defense-in-depth principle;
information on the measures taken in the NPP design in order to provide independence of various defense-in-depth levels;
configuration of safety systems, confirmation of compliance with the SS arrangement principles: single failure, redundancy, diversity, independence (information shall be supplemented with the principal structural diagrams characterizing arrangement of safety systems);
information on the scope of safety functions and the ways of their performance;
configuration of the special-purpose hardware for BDBA management;
substantiation of protection of safety systems and special-purpose hardware for BDBA management against common cause failures (fires, flooding, external natural and human-induced impacts);
evidence of protection of safety systems and special-purpose hardware for BDBA management against human errors;
information on the experience in design, construction, installation, operation, testing and surveys used in technical and administrative solutions adopted in order to ensure the NPP safety;
information on design basis and beyond design basis accidents considered in the NPP design, the measures aimed to prevent BDBA development and to mitigate their consequences and the measures for severe accident management.
1.7.2. Nuclear safety assurance.
Information on the nuclear safety objectives and the systems providing achievement thereof shall be presented.
1.7.2.1. Maintenance of control over chain nuclear reaction in the reactor core.
Information on the ways to assure nuclear safety due to the use of inherent self-protection properties of the reactor shall be provided.
Information on the reactivity balance for all possible states of the normal NPP operation and any abnormal operation including design basis accidents shall be provided; the possibility for any positive reactivity effects in case of accidents shall be analyzed and their potential consequences shall be assessed.
The structure of the provided technical reactivity control means, functions of individual systems and subsystems and their reliability shall be described.
Information on the emergency protection efficiency, reliability and fast response shall be presented.
1.7.2.2. Prevention of local criticality during refueling, transportation and storage of NF.
Brief information on the methods to prevent local criticality in the course of the above-mentioned works shall be provided.
1.7.3. Radiation safety assurance.
Information on the engineering features and administrative measures aimed to ensure protection of the personnel, the public and the environment against ionizing radiation shall be specified.
Information on the efficiency of protection against potential radiation sources in order to ensure non-exceedance of individual risk under normal NPP operation conditions shall be provided; it shall be specified that the risk of potential radiation exposure does not exceed the overall risk limits for the personnel and the public within the period established in the regulations.
Information on adherence to the requirements for releases and discharges into the environment (with due regard for all NPP power units) established by the regulations of the Russian Federation shall be provided.
1.7.4. Fire safety assurance.
Information on the regulatory documents used as the basis to substantiate the NPP fire safety shall be presented.
Information on any fire safety assurance measures provided in the NPP design shall be presented.
The following information on consideration of the fire safety provisions and criteria in the NPP design shall be specified:
classification of the main NPP buildings according to their fire and explosion-and-fire hazard and fire resistance rating;
assurance of the design fire safety level by compliance with the general safety criteria in all NPP normal operation modes as well as in case of any accidents at the NPP;
consideration of a fire among initiating events for analysis of design basis accidents with assessment of a fire breakout probability; analysis of the fire impact on the NPP safety;
performance of the probabilistic analysis of the possibility for a fire coincidence with any other independent NPP abnormal operation and analysis of consequences of the above-mentioned fires from the viewpoint of the NPP safety assurance in such cases;
external and internal impacts on the fire detection and extinguishing means as well as on the fire confinement means;
consideration of a fire as a consequence of any NPP abnormal operation and the NPP safety analysis for such scenarios;
assessment of the fire consequences with due regard for any potential failures in the fire extinguishing plants;
substantiation of the principle for arrangement of active fire extinguishing systems classified as safety systems, their reliability level, analysis of the capability of these systems to withstand single failures of equipment (in case fire extinguishing systems in the rooms where components of safety-related systems are located are classified as non-safety-related systems substantiation of such classification shall be given instead of the above-mentioned information);
basic principles of the fire protection system: multi-barrier, redundancy of SS channels, their physical separation;
the NPP operation procedure in case of any fire in the rooms where safety-related equipment is located and in the rooms where a fire breakout results in the necessity of the RF shutdown;
substantiation of impossibility of simultaneous loss of control from the MCR and the ECR in case of a fire;
data on the fact that in case of any spurious actuation of fire extinguishing units the impact of fire extinguishing means on safety-related components does not result in any dangerous consequences from the NPP safety viewpoint;
determination of the design number of simultaneous fires at the NPP site;
adherence to the building zoning principle (division into fire zones and compartments) and approach to confinement of a fire within an individual compartment or zone;
the list of fire-attackable safety-related systems (components) detected during the analysis of fire impacts on the NPP safety and fire protection measures for these systems (components), regulatory or scientific and technical justification of the adopted solutions.
Information on the impact of any fires at the NPP site (outside the NPP buildings) on the personnel work as well as on civil structures of the buildings located in the vicinity of the fire and safety-related systems and components shall be specified.
1.7.5. Assurance of the NPP protection against natural and human-induced impacts.
The following information shall be provided:
the list of external natural and human-induced impacts and their combinations considered in the NPP design with indication of the impact intensity characteristics and frequency of occurrence;
regulatory basis for calculation of protection against external impacts, provided measures for protection against external impacts considered in the design basis;
information on protection against external impacts exceeding the intensity of impacts considered in the design basis;
information on the methods and calculation programs for assessment of external impacts and the required protective measures.
1.7.6. Action plans for the personnel and the public protection in case of any radiation accident at the NPP
The basic provisions of the action plans for the personnel and the public protection in case of any radiation accident at the NPP, sizes and borders of the protective action planning zone and the compulsory population evacuation action planning zone shall be specified; the expected scope of protective measures shall be also indicated.
Information on protected emergency response control posts located at the NPP site as well as outside the NPP site shall be provided.
1.8. Results of the quantitative safety analysis.
1.8.1. Reliability of safety-related systems and components.
The following information on reliability of safety-related systems and components shall be presented:
the list (range) of reliability indicators for the NPP systems and components subject to reliability analysis in accordance with the federal rules and regulations in the field of atomic energy use;
results of justifying calculations and experimental substantiation of the reliability indicators;
conclusions on compliance of the reliability indicators with the requirements of regulations and the design criteria;
results of the qualitative reliability analysis;
assessment of uncertainty for the reliability analysis results;
references to applied calculation methods and programs;
characteristics of input data regarding reliability.
1.8.2. Deterministic safety analysis.
Brief information on the performed analyses of abnormal operation including design basis accidents at the NPP as well as beyond design basis accidents described in Chapter 15 of the NPP SAR shall be given.
Information shall be provided for all groups of the considered emergency modes and contain the following data for each group:
the number of the considered modes;
justification of selection of modes and purpose of the analysis;
characteristics of the obtained results and assessment of their conservatism.
Information on substantiation of the completeness and representativeness of the final BDBA list (including severe accidents) as well as on consideration of accidents taking place simultaneously at several power units of a multi-unit NPP shall be specified for beyond design basis accidents with reference to the information given in Chapter 15 of the NPP SAR.
1.8.3. Probabilistic safety assessment.
The following information on the PSA results shall be provided:
types of PSA performed (PSA of Level 1, PSA of Level 2);
information on consideration of all locations of NM, RS and RW at the NPP;
information on consideration of all initial states of the NPP normal operation (operation at full and partial power level; refueling, warming and cooldown states);
description of the applied reliability database;
information on the list of considered initiating events (internal IEs, fires and floods as well as external impacts of natural and human-induced origin);
information on the quantitative reliability analysis performed for the systems and consideration of interrelations between the systems;
information on the applied accident sequence models, success criteria for performance of the required functions of the main systems;
information on consideration of common cause failures;
Information on consideration of the personnel actions and errors;
information on sensitivity and uncertainty assessments;
the final PSA results with information on compliance with the requirements of the federal rules and regulations governing the basic requirements for probabilistic safety assessment of the NPP power unit.
Information on the main contributors to probability of a severe accident and probability of a large-scale emergency release and distribution of their relative contribution shares shall be specified.
1.9. Basic design solutions.
1.9.1. Reactor, primary circuit and associated systems.
The following information shall be provided:
purpose of the reactor and the primary circuit as well as their components and associated systems, information on the functions performed by the above-mentioned systems and components;
information on the reactor installation in the cavity, biological and radiation protection;
classification of the systems and components constituting the reactor, the primary circuit and associated systems;
basic operational characteristics of the systems and components;
principles and criteria provided in the RF design.
The description shall be accompanied with:
process flow diagrams;
drawings of the reactor installation in the cavity, the reactor as an assembly, section across the core, main components of the core, the reactor pressure vessel, RCP, SG, the pressurizer, hydroaccumulators, kinematic diagram of the CPS drive.
1.9.2. Steam turbine plant.
Information on the steam turbine plant and associated systems with the process flow diagram of the steam turbine plant and arrangement drawings (plans and sections) shall be provided. The above-mentioned information shall reflect the steam turbine plant configuration and boundaries. Information on the steam turbine plant impact on the RF shall be given; in this case interaction of the steam turbine plant and the RF both through the process parameters and via the control and protection system shall
be described.
Information on the possibility (in case of any NPP abnormal operation) of leakage or RS accumulation shall be provided.
The possibility of any missile generation by the steam turbine plant (the turbine set, pipelines, high-pressure vessels) that can destroy or damage safety systems or cable routes shall be described. Substantiation of protection against the above-mentioned impacts shall be provided.
Substantiation of strength, stability and operability of the steam turbine plant and the associated systems under any external natural and human-induced impacts shall be presented.
The earthquake level at which the plant shall maintain its operability shall be specified.
1.9.3. Circulation and service water supply systems.
The following information on the circulation and service water supply systems shall be provided:
information on the service water sources (water reservoirs, rivers, lakes, seas);
description of the circulation water supply system;
description of the service water supply system.
The provided information shall contain: the list of systems, buildings, structures, main thermal, hydraulic and structural characteristics of systems and equipment (supply and discharge channels, water intake structures, pump stations, cooling towers, makeup systems and sources for circulating systems), fundamental principles and criteria established in the NPP design, description of the system functioning under normal operation conditions and in case of any abnormal operation including design basis accidents and external impacts.
The process flow diagrams of the above-mentioned systems shall be provided.
1.9.4. Electrical systems.
The following information shall be provided for electrical systems:
purpose of the systems;
components of the systems;
scheme of power distribution, number of lines, voltages;
provision of the NPP auxiliary power supply from external and internal sources;
the list of protection, automation and control systems;
information on fire protection of the electric devices;
information on operation of power supply systems in case of any abnormal operation, accidents and external natural and human-induced impacts;
equipment selection criteria.
The following principal schemes shall be provided:
the scheme of the NPP connection to the grid;
the main electric circuit diagram;
the principal scheme of auxiliary power supply for the NPP, EPSS and the power supply system included into the BDBA management hardware;
structural scheme of protections;
structural scheme of control and automation;
structural scheme of communication.
1.9.5. Water chemistry regime of the NPP.
Concept of the WCR selection for the RF coolant and working media of the safety-related systems shall be presented; the adopted technical solutions and administrative measures aimed to assure quality of coolant and working media in the safety-related systems shall be described; information on validation of the applied methods and certification of the measurement means shall be provided.
In shall be specified what controlled coolant quality parameters are regulated and diagnostic; it shall be demonstrated what operation limits and safe operation limits (if any) are established for the regulated parameters and what actions shall be performed in case values of the parameters are beyond the established limits. Frequency of measurement shall be specified for regulated and diagnostic quality parameters.
It shall be demonstrated that the adopted water chemistry regime is established and maintained in such a way so that to provide integrity of physical barriers (claddings, the coolant circuit boundaries).
Information shall be presented to confirm that the adopted WCR ensures the following:
corrosion resistance of the structural materials of equipment and pipelines within the entire power unit service life by minimization of corrosion processes in the structural materials under all operation modes;
reduction of deposition growth intensity on the heat transfer surfaces;
physical and chemical conditions of the process media and the NPP equipment surfaces enabling to maintain radiation exposure for the personnel as low as reasonably achievable with due regard for social and economic factors;
limitation of radiation accident consequences for the NPP personnel and the public due to release of iodine radionuclides into the atmosphere;
arrangement of conditions to reduce the amount of generated radioactive wastes and RS discharges beyond the NPP boundaries established in the design.
Information on quality control for the primary and secondary circuit coolant and working media of the safety-related systems shall be provided for all NPP operation modes: hydraulic testing, circulation flushing, hot trial, start-up of the power unit, power operation, shutdown of the power unit and outage mode.
1.9.6. Fuel management system.
1.9.6.1. The following information shall be presented for the set of NF storage and handling systems (outside the reactor):
the list of all NF storage facilities for both fresh and spent NF;
characteristics of fresh NF used at the NPP as well as the fuel removed from the core with indication of the methods for burn-up determination;
the maximum design capacity (volume) of each storage facility and the number of places reserved for emergency core unloading and storage of rejected nuclear fuel (both fresh and spent);
brief characteristic of the NF storage techniques both in FFS and SNFS; presence of any absorbing additives in the storage facility materials or in the coolant shall be specified;
the way of NF delivery to the NPP and SNF removal from the NPP, information on the proposed transportation frequency and the TP types used;
information on in-plant transportation (means of transport and types of transport packages);
information on the rejected NF handling (for both fresh and spent NF) beginning with the rejection method;
the list of initiating events the set of NF (SNF) storage and handling systems is designed for and analysis of any abnormal operation including design basis accidents.
1.9.6.2. Provision of heat removal from the reactor core and the spent fuel pool.
Information of heat removal from the reactor core and also from the fuel pool (SNF storage facilities) to the ultimate heat sink under normal operation conditions and in case of any abnormal operation including accidents shall be provided. The flow diagram of the heat removal systems shall be presented.
Brief characteristics of the liquid RW handling system, the main objectives, criteria and principles of its design as well as information on the means to achieve these objectives shall be provided.
Brief characteristics of the solid RW handling system, the main objectives, criteria and principles of its design as well as information on the means to achieve these objectives shall be provided.
Brief characteristics of the gaseous RW handling system, the main objectives, criteria and principles of its design as well as information on the means to achieve these objectives shall be provided. Information on special gas treatment systems, exhaust ventilation systems of the controlled access area used to reduce releases of radioactive aerosols, various forms of iodine (aerosol, molecular and organic) and inter radioactive gases into the atmosphere and the NPP rooms shall be provided. Information on the decontamination factors shall be presented separately for each of the above-mentioned systems.
1.9.8. NPP process control system.
Brief information on the following aspects shall be provided:
APCS of the NPP power unit and its structure, classification of functional groups for control of the NPP power unit processes, location of the APCS rooms in the NPP buildings, control rooms of the NPP power unit, the system of warning and emergency annunciation signals for the NPP power unit
personnel;
NPP NOCS;
reactivity control systems as well as emergency protection of the reactor;
CSS;
the system for provision of safety-related information to the operator;
The lists of protective, localizing, supporting, control safety systems as well as the systems classified as special-purpose hardware for BDBA management shall be given. The following information shall be provided for each of the listed systems (hardware):
purpose and configuration of the system (hardware); accidents when operation of the relevant system (hardware) is required;
compliance with the safety principles and criteria established in the regulations and the NPP design;
criteria for performance of the prescribed functions by the system;
description of the system: flow diagram, arrangement, protection against internal and external impacts, monitoring and control;
state of the system under normal NPP operation conditions; integrated testing of the system, procedure for the system monitoring and control;
the system operation mode in case of accidents.
1.9.10. General layout and configuration of the NPP
1.9.10.1. General layout.
The general layout drawing with the list of the main NPP buildings and structures shall be presented.
The following information shall be provided:
substantiation of location of the main NPP buildings and structures, hydraulic engineering structures, outdoor switchgear, auxiliary NPP buildings and structures on the general layout;
information on the process interrelations, natural terrain of the area, the prevailing wind directions, geological and hydrological conditions at the site, inclination of the NPP site terrain, grading elevations of the NPP site, priority of construction for the power units;
information on orientation of the main NPP buildings in relation to the prevailing wind directions (wind rose);
distance between the main buildings and structures and their justification;
motor roads and railways, conditions for entrance into the main buildings and structures;
protection of the site against surface water inflow;
utility networks, transport, process and electrical connections between the main NPP buildings and structures, between the controlled and uncontrolled access areas.
1.9.10.2. Arrangement principles for the main structures and equipment.
The following information shall be provided:
location of seismic-resistant systems and equipment of seismic categories I and II established in accordance with federal rules and regulations in the field of atomic energy use;
division of the main building structures into the controlled and uncontrolled access areas.
1.9.10.3. List of the main buildings and structures and their purpose.
The following information shall be provided:
basic layout solutions;
the list of safety-related systems and components located in the building.
Layout drawings (schemes and sections) of the main buildings and structures with indication of the main equipment shall be provided in order to give a general idea of the adopted layout solutions.
1.9.11. Ventilation systems.
1.9.11.1. Design criteria.
The following information shall be provided:
maintenance of the preset air temperature inside the rooms in the design NPP operation modes;
provision of the acceptable working conditions for the operating personnel in all design operation modes according to the sanitary regulations;
arrangement of the conditions for repair and refueling works.
The list of the following main ventilation systems with indication of their purpose shall be given:
plenum and exhaust;
recirculation;
air conditioning.
1.9.11.2. Description of ventilation systems.
Brief information on the following ventilation systems shall be provided:
normal operation, safety-related;
belonging to supporting safety systems.
The information shall include: purpose and configuration of each ventilation system, design criteria, operation modes.
Schemes of the systems with indication of the main equipment and its characteristics shall be provided.
1.9.12. Radiation protection and radiological monitoring.
Classification of the NPP zones and rooms adopted in the NPP project and used as the basis for design of the biological protection against penetrating radiation and prevention of contamination of the equipment surfaces, building structures and air inside attended rooms with radioactive substances shall be presented.
General information on biological protection shall be given for the main radiation sources.
The adopted dose criteria and the main technical and organizational solutions aimed to maintain the minimum achievable level of radiation shall be described.
Criteria for selection of the radiological monitoring hardware, arrangement of the scheme of sampling points and location of the equipment (instrumentation) shall be presented; general description of the radiological monitoring hardware provided in the NPP design and the ARSMS system shall be given.
1.9.13. Physical protection system.
General information on the PPS arrangement concept and the main requirements for the PPS shall be presented.
1.9.14. Fire safety arrangements.
The following brief information on implementation of fire safety measures in the NPP design shall be specified:
information on division of the main NPP buildings into fire zones with determination of fire resistance ratings for their boundaries;
the short list of the rooms with high fire loads in the main NPP buildings with indication of the following: fire and explosion-and-fire hazard category of the room, the specified fire resistance rating of fire barriers, the main fire protection measures in the construction section, ventilation, equipment of the rooms with fire alarm systems, automatic and self-contained fire extinguishing units, the extinguishing agent;
the list of the main fire protection measures in the architectural and construction and process engineering sections of the NPP design in order to ensure operation of the main equipment classified as safety-related components in case of any fire at the NPP;
information on physical separation of different SS channels through the use of fire partitions (barriers) and safe distances;
information on the SS channel separation for the rooms containing any safety-related equipment;
information on arrangement of any means for prevention or limitation of any liquid spillage and spreading from oil-filled equipment and pipelines in case of a fire;
information on equipment of the NPP rooms with fire extinguishing units;
information on assurance of hydrogen explosion protection under normal NPP operation conditions and in case of any abnormal operation;
information on the means for protection of cable routes against fire spreading;
information on any fires extinguished in any way capable of direct or indirect impact on the safety-related components;
information on types of fires determinant for calculations of the fire protection system;
information on the determinant provisions for analysis of fire loads on the rooms of the main NPP structures with characteristics of fire and explosion hazard of substances and materials;
information on the personnel evacuation routes and systems for fire warning and the personnel evacuation management in the buildings;
information on firewater supply for the NPP, the main structures of the NPP, equipment of the buildings with the internal firewater pipeline, arrangement of water intake from various sources through the use of mobile fire-fighting equipment;
information on the fire detection and alarm systems;
the list of the main automatic and self-contained fire extinguishing units and their purpose;
analysis of fire hazard in the main NPP structures, analysis of fire consequences for the NPP safety;
analysis of fires caused by destruction of the NPP buildings and structures due to external impacts.
1.10. Brief description of the NPP operation.
1.10.1. Preparation of the NPP power unit for start-up in the course of operation.
Information on any differences in the NPP power unit preparation for start-up in the course of the NPP power unit commissioning and during the NPP operation shall be specified.
Brief information on the RF preparation for start-up shall be provided:
state of individual RF elements and components;
the primary circuit filling (duration of filling shall be specified);
information on the RCP start-up;
information on leak-tightness and strength tests for the primary and secondary circuit;
information on testing of the systems and equipment prescribed in the step-by-step testing program.
Boundary characteristics of pressure and temperatures in the primary and secondary circuit characterizing the relevant stage of the NPP power unit preparation for start-up and duration of this stage shall be specified.
1.10.2. Cold start-up of the NPP power unit up to the rated power of the NPP unit.
Brief information on cold start-up of the NPP power unit up to the rated RF power shall be provided:
the method for the reactor core warming after refueling;
information on the core condition monitoring;
information on the primary circuit checks for leak-tightness and strength;
information on the SG checks for leak-tightness and strength via the secondary circuit;
information on checks of protections and interlocks in accordance with the requirements of the process regulations;
information on the integrated testing of CPS;
information on measurement of neutron and physical characteristics of the reactor core after achievement of the reactor MCL;
information on testing with the use of commissioning measurement systems;
information on the coolant warming, methods of warming, preparation of turbine generators for start-up, warming of the main pipelines;
information on the position of EP control device groups;
information on boric acid discharge;
the RF rising to power.
In case of necessity the list of measurements and tests shall be supplemented with due regard for experience in commissioning of the first-of-a-kind power units.
The following information shall be provided:
boundary parameters of the primary circuit coolant (pressure, temperature);
pressure in the secondary circuit;
warming rate;
conditions for the RF warming completion;
conditions for the reactor power rising to the minimum controlled level;
the reactor power when connection of the turbine is possible;
coolant parameters at the rated RF power.
The RF warming schedule shall be provided.
1.10.3. Power operation.
The following information shall be provided:
the range of "power operation" with due regard for accuracy of power maintenance by the control system;
the main RF parameters at the rated power;
the main parameters of the steam turbine plant;
the main parameters of the turbine generator and electrical systems of the NPP unit;
conditions for functioning of the main process systems in the primary and secondary circuit during
"power operation" of the unit;
compensation of reactivity wander, maintenance of the critical state of the reactor in load reduction modes and transient modes;
conditions for occurrence and characteristics of xenon oscillations and algorithm for their suppression;
the main characteristics of the primary circuit makeup/blowdown system;
the main characteristics of the SG blowdown system.
1.10.4. The NPP unit power control.
Brief information on operation of the main RF and turbine plant controllers shall be presented.
1.10.5. Transient modes.
Brief characteristics of the initial RF state shall be given for each transient mode. Besides information on typical events in the following modes shall be provided:
scheduled switch-off of the RCP;
connection of a previously non-operating loop of the reactor coolant pipeline;
scheduled switch-off of the feed pump;
disconnection of the turbine generator from the grid;
accelerated RF unloading;
HPH disconnection (connection);
turbine generator load reduction up to the auxiliary level.
1.10.6. Switching of the NPP power unit from "power operation" state into "hot state".
Characteristics of the "hot state" defined in the RF design shall be specified.
The following information shall be provided:
sequence of functioning of the primary and secondary circuit systems;
cooldown rate;
the method for cooldown and residual heat removal;
the reactor sub-criticality and the methods to achieve it;
information on the turbine generator unloading, the RF power reduction and the main controlled parameters;
information on the RF cooldown after the turbine generator unloading to 10-15%, control of level in the SG, BRU-K actuation;
information on the reactor bringing into the "hot state" with prior assurance of its sub-criticality;
information on the SG PORV testing and the method of testing;
information on the pressurizer PORV testing and the method of testing.
Information on the boundary parameters for each cooldown stage shall be specified for the primary and secondary circuit; the cooldown schedule shall be provided.
1.10.7. Operation of the NPP power unit in "hot state" and permissible maintenance works.
The following information shall be provided:
coolant temperature and pressure with due regard for brittle strength conditions assurance;
the short list of malfunctions resulting in the RF transition to the "hot state";
the possibility for elimination of defects and the RF maintenance in the "hot state".
1.10.8. The NPP power unit cooldown to the "cold state".
Characteristics of the "cold state" defined in the RF design shall be specified.
The following information shall be provided:
sequence of functioning of the primary and secondary circuit systems;
the RF cooldown rate;
the method for the RF cooldown and residual heat removal;
the reactor sub-criticality and the methods to achieve it;
information on the turbine generator unloading, the RF power reduction and the main controlled parameters;
information on the RF cooldown after the turbine generator unloading to 10-15% of the rated power, control of level in the SG, BRU-K actuation;
information on the reactor bringing into the "hot state" with prior assurance of its sub-criticality;
information on the SG PORV testing and the method of testing;
information on the pressurizer PORV testing and the method of testing;
information on the RF cooldown, feedwater temperature decrease, nitrogen supply to the pressurizer, switch-off of the RCP, nitrogen discharge from the pressurizer;
information on the RF cooldown completion.
Information on the boundary parameters for each RF cooldown stage shall be specified for the primary and secondary circuit; the RF cooldown schedule shall be provided.
1.10.9. The NPP power unit operation in the "cold state" without opening the primary circuit.
The following information shall be provided:
sub-criticality conditions for the reactor;
brittle strength assurance conditions for the reactor;
the list of the main emergency modes resulting in the necessity of "cold state".
1.10.10. Refueling.
The following brief information on the refueling procedure shall be provided:
operations for the reactor decompression;
operations for SNF unloading from the reactor to the fuel pool; rearrangement of fuel inside the core and fresh fuel loading; tightness control for fuel elements;
scope of control in the course of refueling;
information on residual heat removal assurance in the course of refueling.
Time schedule of a typical refueling procedure shall be provided.
The list of maintenance and repair works in the course of refueling shall be presented.
1.11. The NPP impact on the environment and the public.
Information on assessment of the NPP impact on the environment and the public (chemical impact, radiation exposure, thermal pollution, electromagnetic and acoustic impacts) shall be provided.
Information on the measures aimed to prevent and (or) reduce any potential adverse impact of the NPP on the environment shall be specified.
1.12. Comparison with similar designs of national and foreign NPPs.
Information on the NPP counterparts shall be provided.
Any NPP with the same RF type where the same or similar NPP safety assurance principles are implemented may be the NPP counterpart.
It shall be substantiated in the course of comparison that the new NPP design is comparable or advantageous with regard to its concept and the adopted technical solutions and also complies with state-of-the-art approaches to NPP safety assurance.
The proposed NPP design shall be compared with the counterpart with regard to all safety-related normal operation systems, safety systems and special-purpose hardware for BDBA management.
1.13. The NPP construction schedule, contracting parties and contractors.
The network NPP construction schedule, names and addresses of the main participants - the NPP designers and the NPP construction contractors - shall be provided.
Information on the operating organization, the main contracting parties, developers and their battery limits shall be presented (information on the NPP and RF designers, APCS developers, developers of the main RF and steam turbine plant equipment shall be specified).
1.14. Principle provisions on the arrangement of the NPP operation.
1.14.1. Commissioning of the NPP power units.
Brief information on the commissioning work program including testing of the structures, systems and components in the course of the NPP power unit commissioning shall be provided; in this case the basic process restrictions and instructions, conditions and measures for safe performance of works and tests shall be specified.
The main stages of commissioning testing at the NPP shall be listed with description of their plan enabling to assess the possibility for successful performance of commissioning works and success criteria for completion of all items mentioned in the plan. The objective to be achieved in the course of these commissioning tests shall be specified for each stage.
Procedures and methods used to analyze the obtained results and to define achievement of the objectives shall be specified, and brief information on assessment of the results of initial criticality achievement, step-by-step power increase and the most important characteristics of the RF and NPP SS equipment shall be provided.
The procedure for issuance, submittal and storage of reporting documentation with indication of conditions for access thereof shall be described.
1.14.2. The NPP operation management.
Information on preparation and arrangement of the NPP operation shall be provided including brief description of the operating organization structure specifying responsibilities of individual persons and departments for the NPP operation. Description of the operating organization shall contain the basic issues on training of the NPP personnel with the required qualification (availability of training centers, training programs, timely training, the procedure for competence assessment and admittance to unsupervised works).
Information on maintenance activities and monitoring of the operational (current) NPP state shall be provided. Information on consideration of the inspection and testing results in the programs for the NPP operational safety assessment, the ways to consider the NPP operation experience in development of maintenance schedules and the procedure for preparation and submittal of regular information on the current safety level shall be presented.
1.14.3. Safe operation limits and conditions.
General information on safe operation limits and conditions, approaches to their substantiation and determination shall be presented without any particular numerical values. Reference shall be given to sections of Chapter 16 of the NPP SAR where the safe operation limits and conditions established in the NPP design are specified.
1.14.4. Decommissioning of the NPP power unit.
Basic provisions of the NPP power unit decommissioning concept shall be stated.
Information on the expected sequence of actions for decommissioning of the NPP power unit and radiation safety assurance in the course of these actions shall be provided.
Information on the planned measures aimed to ensure radiation safety of the personnel, the public and protection of the environment at the preservation stage (supervised storage), the disposal stages (restricted use of the NPP site) and liquidation of the NPP unit (unrestricted use of the NPP site) shall be presented.
Information on the measures aimed to provide the following at all stages of the NPP power unit decommissioning shall be given: generation of the minimum quantities (amounts) of radioactive wastes and reduction of radiation exposures for the personnel and the public, reduction of RS releases into the environment to the minimal achievable level.
1.15. Quality assurance.
Brief information on the activities of the participants of works for the NPP siting, design,
construction, operation and decommissioning as well as for development and manufacturing of safety-related systems (components) shall be provided in order to confirm capability of these organizations to ensure quality of the performed works and rendered services affecting the NPP safety.
The general quality system arrangement scheme in the course of the NPP siting, design, construction, operation and decommissioning as well as for development and manufacturing of safety-related systems (components) shall be described in order to demonstrate interaction between the operating organization and any organizations performing works or rendering any services to the operating organization, distribution of works and responsibility among them.
Responsibility of each organization performing works or rendering any services to the operating organization for assurance of the NPP quality, reliability and safety shall be specified.
Presence of independent quality assurance control for all works, products or services affecting the NPP safety in the operating organization shall be reflected.
Information on the state of the quality system development, implementation and functioning in the operating organization and in any organizations performing works or rendering services to the operating organization shall be provided.
Information on the state of the NPP QAP development and implementation as of the SAR submittal in the operating organization and in any organizations performing works or rendering services to the operating organization shall be provided.
II. Requirements for the content of Chapter 2 "Characteristics of the NPP region and site"
Information on geographical, topographic, hydrological, meteorological, geological, hydrogeological, seismotectonic, geoengineering and engineering-geological conditions of the NPP location, human-induced external impacts, the existing and future population distribution and land use for development shall be presented in Chapter 2 of the NPP SAR.
Completeness and sufficiency of the performed surveys and studies in the NPP location area and at the NPP site in order to reveal and obtain reliable characteristics of the area that shall be taken into account in the design basis at all stages of the NPP life cycle and in emergency planning for evacuation of the personnel and the public from the NPP location area shall be substantiated.
The following shall be determined:
the list of parameters and characteristics of the external natural and human-induced impacts on the NPP in accordance with the range of processes, phenomena and factors of natural and human-induced origin established in the federal rules and regulations in the field of atomic energy use for consideration of external natural and human-induced impacts on nuclear facilities;
the list of parameters and characteristics of the NPP impact on the environment within the NPP location area;
critical values of the controlled parameters of external impacts with hazard classes I and II established in accordance with the federal rules and regulations in the field of atomic energy use that require to make a decision on the necessity to implement administrative and technical safety assurance measures subsequent to the results of local monitoring and control in the course of the NPP power unit construction and operation.
Besides quantitative values of the parameters adopted as reference intensity levels for external impacts of natural and human-induced origin with dynamic nature of manifestation (earthquakes, external explosions and airplane crashes) that require automatic or manual shutdown of the NPP power
unit in case of their exceedance (during occurrence of the above-mentioned external impact at the NPP site) shall be specified.
The following information shall be presented in Chapter 2 of the NPP SAR submitted within the package of documents to substantiate the NPP safety in the course of siting:
the list of external natural and human-induced processes, phenomena and factors in the NPP location area and at the NPP site capable of affecting the NPP power unit safety;
maximum parameters (intensity, frequency) of external impacts to be considered in the NPP power unit safety assurance;
hazard class of external impacts and the site class;
information on absence of any external factors preventing from the NPP location at the site and the possibility to develop administrative and technical measures for safety assurance in case any adverse external natural and human-induced phenomena and factors are manifested at the site.
Brief information on the sites considered as alternative in relation to the selected one shall be also provided. Information for the selected and approved NPP site shall be provided in Chapter 2 of the NPP SAR submitted within the package of documents to substantiate the NPP safety in the course construction or operation.
2.1. Description of the NPP location area.
The following territory pick-up radii shall be applied (the main building (the reactor building) shall be assumed as the NPP site center):
region - at least 300 km;
the neighboring region (location) - at least 30 km;
the NPP site - at least 3 km.
The NPP location shall be recorded with indication of the latitude, longitude and altitude in the unified coordinate and elevation system.
2.1.1. Geographical position.
The following information shall be provided:
administrative location of the NPP site (republic, territory, region);
name of the administrative center;
distance to the administrative center;
distance to the nearest administrative boundaries;
distance to the national boundaries and names of the nearest states;
position of the NPP site in relation to natural and artificial benchmarks (populated localities, rivers, seas, airports, railway stations, sea and river ports);
hazardous industrial facilities (plants, factories, chemical works, food industry facilities, power engineering facilities), hydraulic engineering structures that can affect the NPP safety;
the nearest transportation facilities (gas and oil pipelines, railway roads, motor roads, airfields, sea and river ports);
the nearest military facilities;
territories where location of any NPPs is prohibited by the environmental legislation of the Russian Federation;
engineering protection facilities of the NPP site (dams, dikes and drainage systems) and any soil modifications at the NPP site (soil replacement or levelling changes).
2.1.2. Topographic conditions.
The list of documentation containing the results of topographical surveys and studies as well as analysis of these results shall be given.
The terrain of the NPP region and location site shall be characterized. In this case the following data shall be specified:
maximum and minimum absolute elevations of the NPP location area;
surface slope and its direction;
presence of any terrain peculiarities (creeks, clifts, depressions, karst manifestations);
presence of wetlands;
presence of forests and farming lands.
The following documents shall be presented for the neighboring region of the NPP:
a topographic map on the scale of 1:25000 - 1:10000;
a topographic and bathymetric plan and map on the scale of 1:10000 for the shelf area with the bottom contour section by isohypses with the interval of 5-2.5 m combined with topographic plans of the onshore area of the location;
materials of observation over present-day crustal motion (observation scheme);
a topographical map (plan) of the NPP site on the scale of 1:10000 (1:5000);
topographic and bathymetric plans and maps of the shelf area for the NPP site on the scale of 1:10000 - 1:5000.
The topographic materials shall be obtained not later than five years prior to their presentation. Topographic materials for the terrain unaltered prior to the NPP construction and information on all terrain modifications as of the NPP SAR development shall be additionally provided.
2.1.3. Demographics.
The provided data shall be based on the results of the latest census of population with due regard for the population migration and growth, the need for efficient evacuation of the public in the NPP construction area as well as any population travelling along transportation lines. The following information shall be presented in this section:
population density in the zone with the radius of 30 km from the NPP site boundary: prior to construction commencement, during the construction period and during the NPP operation period;
distance to the cities with the population size exceeding 100 000 people for the zone with the radius of 100 km from the NPP site;
borders of the sanitary-protective area, supervised area, the protective action planning zone and the compulsory population evacuation action planning zone;
population distribution (by size and density) on the map by sectors (rings) around the NPP bounded with the radii of 10, 10 - 15, 15 - 20 and 20 - 30 km and divided into 16 rhumbs;
information on specific groups of population: permanent and temporary residents, age groups (children, the elderly), hard-to-evacuate groups (patients, prisoners);
population food ration, share of imported and local food products;
domestic water demand, sources of water supply;
information on daily and seasonal migration of the population;
duration of the population stay in open spaces;
characteristics of transportation means and transportation lines to be used in case of emergency.
2.2. Human-induced conditions of the NPP location.
2.2.1. Basic materials for determination of the frequency of occurrence and parameters for human-induced external impacts.
Data sufficient to substantiate probability assessment for occurrence of external impacts and to predict their intensity, parameters and characteristics both for the purpose of consideration in the NPP design basis and for assessment of the NPP compliance with the benchmark for probability of a large-scale emergency release established in the federal rules and regulations in the field of atomic energy use shall be provided.
The consolidated list of processes and factors of human-induced external impacts shall be compiled based on the construction area and the NPP site survey.
Data shall be presented in the form of text information, maps, diagrams and tables.
2.2.1.1. Airplane crash or other missiles
The following information shall be provided:
information on location of airports, air corridors, intersections of air routes in the NPP location region (on a schematic map);
information on air traffic types, types and characteristics of aircraft, frequency of flights;
schemes of aircraft take-off, landing and parking;
presence of any military facilities or air envelopes used as bombing ground at the distance of up to 30 km from the NPP site and information on the types of any potential missiles, their characteristics and hazard occurrence probability;
archive data on airplane crashes.
2.2.1.2. Fire due to external reasons.
Information on presence of any external fire hazard sources in the NPP location area and at the NPP site within the radius of 5 km shall be provided:
forested areas;
warehouses for explosive materials (solid, liquid and gaseous);
product pipelines and oil and gas main pipelines;
railways and motorways, river and sea routes;
airfields, air routes and flight lines;
residential areas;
coal and peat extraction facilities;
areas with peat deposits. Archive and statistic data on fires and their causes in the NPP location area and at the NPP site shall be provided for at least 5 recent years.
Information on the stock of flammable materials shall be presented.
2.2.1.3. Explosions at facilities with due regard for explosions in the course of scheduled works.
The term "explosions at facilities" means any explosions at the NPP site and beyond it, in the NPP location area; the scope of requirements for consideration of these explosions is defined in the federal rules and regulations in the field of atomic energy use.
Information on presence of any potential stationary and mobile explosion sources in the NPP location area and at the NPP site shall be provided:
warehouses and storage facilities for explosive substances within the radius of 10 km;
any enterprises where hazardous technologies are applied, process-related explosions are possible and also any pressurized vessels or high-pressure plants with gases, vapors and other liquids are installed within the radius of 5 km;
motorways and railways, water transport with indication of transported explosive substances and transportation vehicles within the radius of 5 km;
oil and gas main pipelines, product pipelines, process equipment or pipelines for flammable gases or highly flammable liquids that can become a source of leakage resulting in formation of clouds of explosion- and fire-hazardous mixtures within the radius of 7 km;
military facilities within the radius of 30 km.
Information on the stock of explosive materials shall be presented.
Archive and statistic data on any explosions in the NPP location area shall be provided.
The map of external sources (in relation to the NPP site) shall be provided.
2.2.1.4. Discharges of explosive, flammable and toxic vapors, gases and aerosols into the atmosphere, drifting cloud explosions.
The following shall be provided:
information on any releases of explosive and flammable vapors, gases and aerosols into the atmosphere from chemical plants and any fire sources in the NPP location area and at the NPP site within the radius of 7 km;
transportation schemes for mobile explosion sources;
meteorological conditions;
statistic data on any incidents in the NPP location area.
2.2.1.5. Discharges of toxic vapors, gases and aerosols into the atmosphere.
The following shall be provided:
information on any sources of toxic releases from chemical plants and any fire sources in the NPP location area and at the NPP site within the radius of 7 km;
transportation schemes for mobile toxic hazard sources; dispersion of impurities in the atmosphere;
information on potential volumes of toxic substances;
meteorological and aerological conditions;
statistic data on any incidents in the NPP location area.
2.2.1.6. Corrosive and toxic liquid discharges into surface and ground waters.
Results of chemical analysis of water and soil samples within the NPP location area shall be provided.
Information on hydrogeological properties of the NPP site, brief characteristics of aquifers, chemical composition of underground waters, its variations over time, any possible flooding of underground NPP structures, conditions for perched water formation (temporary aquifers above the groundwater level) shall be specified. Degree of aggressive impact of soils shall be defined below the underground water level.
The list of any potential sources of liquid corrosive and toxic discharges at the NPP site and in the NPP location area with description of their parameters, distance from the NPP site, statistic data on discharges of corrosive and toxic substances stored, produced or transported at the NPP site and in the NPP location area shall be provided.
2.2.1.7. Breakage of natural and man-made water reservoirs.
The following information shall be provided:
the plan of water reservoirs location in relation to the NPP site;
reliability characteristics of hydraulic engineering structures with due regard for external impacts of natural and human-induced origin;
statistical data obtained through processing of hydro-meteorological information for a long-time period (not less than 50 years) containing sets of annual parameter values and peaking data;
data on annual water level measurement in the upstream reach;
statistical estimates of the maximum water storage in the upstream reach;
measurement data obtained through standard hydro-meteorological observation programs with hourly measurements at the NPP site.
2.2.1.8. Spillage of oils and petroleum products on coastal surfaces of water bodies.
Maps containing the information on any facilities capable of becoming a source of oil and petroleum product spillages on the coasts of water bodies in the NPP location area and at the NPP site, on passing of any ship routes, motor roads and railway roads shall be provided.
Information on any potential volumes of oil and petroleum product spillages shall be provided; the size of possible contamination plumes on the coastal surfaces of water bodies (archive and statistic data) shall be specified.
2.2.1.9. Accident at a radiation-hazardous facility.
The following information shall be provided:
distance to the facilities;
parameters of the potential RS release in case of an accident at a radiation-hazardous facility;
statistic data on any incidents in the NPP location area.
2.2.1.10. Electromagnetic pulses and emissions (including pulses and emissions caused by a storm discharge).
Information on intensity of electric and magnetic fields shall be provided.
2.2.1.11. The list of organizations that have legally confirmed the data related to hazard sources of human-induced nature.
The list of organizations that have legally confirmed the data related to hazard sources of human-induced nature with indication of the details for documents confirming this information shall be provided.
2.2.2. Methods for prediction of parameters and characteristics of human-induced external impacts.
Detailed description of the techniques and calculation methods for the basic parameters and characteristics of human-induced external impacts shall be provided, information on the applied mathematical tools, assumptions and limitations, results of experimental substantiations shall be given. For validated software tools presentation of information on their validation is sufficient.
2.2.3. Parameters and characteristics of human-induced external impacts.
The following parameters and characteristics of external impacts shall be determined:
2.2.3.1. Aircraft crash or other missiles.
Information on the type of considered aircraft and other missiles shall be presented.
Parameters of the falling aircraft and other missiles shall be specified:
stiffness properties of the hitting objects;
weight of the objects;
fuel weight;
impact velocity;
angle of impact on the NPP structures;
impact direction;
impact area;
point of impact.
The following design characteristics shall be provided:
fuel spillage and subsequent fire;
ASW;
discharges of explosive, flammable and toxic vapors, gases and aerosols into the atmosphere, drifting cloud explosions.
Probability of an airplane crash and falling of any other missiles on the NPP site, buildings and structures shall be assessed.
2.2.3.2. Fire due to external reasons.
The following information shall be provided:
probability of a fire breakout;
equivalent area of the territory affected by the fire;
heat flux in the source of fire and its changes towards the NPP site;
fire propagation rate towards the NPP.
2.2.3.3. Explosions at facilities.
Types of the considered explosion loads (detonation, deflagration) shall be specified.
Probability of the event (explosion) occurrence shall be assessed. It is permitted not to assess probability of occurrence for the particular type of external impact if it is demonstrated that no NPP abnormal operation will appear in case of any impact of this type physically feasible in the NPP location area.
The following information shall be provided:
overpressure at the ASW front;
TNT equivalent;
design concentration and toxicity of gas near the NPP;
probability of an explosive cloud drift towards the NPP, probability of the cloud conflagration;
ignition source power.
2.2.3.4. Discharges of explosive, flammable and toxic vapors, gases and aerosols into the atmosphere, drifting cloud explosion.
Probability of the event (discharge, explosion) occurrence shall be assessed. It is permitted not to assess probability of occurrence for the particular type of external impact if it is demonstrated that no NPP abnormal operation will appear in case of any impact of this type physically feasible in the NPP location area.
The following information shall be provided:
amount of vapors, gases and aerosols that can be involved in the event;
initial concentration of vapors, gases and aerosols at the point of discharge;
dispersion of discharges in the atmosphere;
concentration of vapors, gases and aerosols from the primary sources and secondary effects;
duration of exposure;
ignition source presence and power;
concentration of vapors, gases and aerosols upon the drifting cloud approaching to the NPP.
2.2.3.5. Discharges of toxic vapors, gases and aerosols into the atmosphere.
Probability of the event (discharge) occurrence shall be assessed.
It is permitted not to assess probability of occurrence for the particular type of external impact if it is demonstrated that no NPP abnormal operation will appear in case of any impact of this type physically feasible in the NPP location area.
The following information shall be provided:
amount of toxic vapors, gases and aerosols that can be involved in the event;
initial concentration of toxic vapors, gases and aerosols at the point of discharge;
dispersion of discharges in the atmosphere;
concentration of toxic vapors, gases and aerosols from the primary sources and secondary effects;
duration of exposure.
2.2.3.6. Breakage of man-made water reservoirs.
Probability of the event (breakage or dewatering) occurrence shall be assessed. It is permitted not to assess probability of occurrence for the particular type of external impact if it is demonstrated that no NPP abnormal operation will appear in case of any impact of this type physically feasible in the NPP location area.
The following information shall be provided:
wave height;
wave speed;
absolute level elevation and duration of the territory flooding in case of combination of adverse factors;
extreme levels of spring tide or rainfall flood in the watercourse within the NPP location area with due regard for the reservoir breakage wave height.
In this case analysis of the ultimate heat sink loss that can be caused by this event shall be considered in Chapter 15 of the NPP SAR.
2.2.3.7. Corrosive and toxic liquid discharges into surface and ground waters.
The following information shall be provided:
initial concentration of corrosive and toxic liquid discharges;
concentration of corrosive media interacting with the NPP systems as a function of time and distance;
distance from the discharge source to the NPP;
possible concentration of corrosive media near the NPP systems;
duration of exposure;
assessment of the impact of corrosive liquid discharges on the NPP safety.
2.2.3.8. Electromagnetic pulses and emissions (including pulses and emissions caused by a storm discharge).
Probability of the event (electromagnetic pulses and emissions) occurrence shall be specified. It is permitted not to assess probability of occurrence for the particular type of external impact if it is demonstrated that no NPP abnormal operation will appear in case of any impact of this type physically feasible in the NPP location area.
The following information shall be provided:
distance to the source;
intensity of electrical and magnetic fields.
2.2.3.9. Spillage of oils and petroleum products on coastal surfaces of water bodies.
The following information shall be provided:
spill area and film thickness;
chemical composition;
distance to the NPP;
distance to the NPP water intake point;
heat flux in the source of fire and its changes towards the NPP.
2.2.3.10. Accidents at radiation-hazardous facilities.
Information on the RS amounts released into the environment shall be provided.
2.2.3.11. Other human-induced external impacts.
Dependence of the impact intensity on its probability shall be specified for other human-induced external impacts.
2.3. Hydro-meteorological conditions.
Meteorological, aerological and hydrological conditions for the NPP location area shall be provided.
The following analysis results for meteorological, aerological and hydrological conditions in the NPP location area shall be presented:
the list of meteorological, aerological and hydrological processes and phenomena characteristic for the NPP location area;
substantiated opinion on presence or absence of certain meteorological, aerological and hydrological processes and phenomena at the NPP site.
Information shall be provided separately for each type of meteorological, aerological and hydrological processes and phenomena. Conclusions on intensity and frequency of occurrence for processes and phenomena shall be accompanied with evidence in the form of description of the results of special-purpose observations, calculations, analysis of statistic data.
2.3.1. Meteorological characteristics.
The following data shall be provided:
average monthly and average annual wind speed, design maximum wind speeds up to the exceedance probability of 1, 0.1 and 0.01% (repeatability of once per 100, 1000 and 10000 years); repeatability of wind directions (wind rose) within the observation period;
average and extreme values of air saturation with water vapors (absolute and relative humidity), daily variations of humidity;
average, extreme for the entire observation period and design maximum amount of precipitation (liquid, solid) up to the exceedance probability of 1, 0.1 and 0.01% (repeatability of once per 100, 1000 and 10000 years), the daily maximum; duration of precipitation; their intensity distribution: monthly and annual wind roses bringing precipitation;
average and maximum repeatability and duration of fogs, smogs, thunderstorms, blizzards, hail, glaze, dust and sand storms;
average, extreme observed monthly and annual and design maximum air temperature values up to the exceedance probability of 1, 0.1 and 0.01% (repeatability of once per 100, 1000 and 10000 years);
average and extreme temperature of soil on the surface and at standard depths;
average and extreme atmospheric pressure;
pollution, dust content and corrosive activity of the atmosphere;
annual probability assessment for hazardous meteorological phenomena (tornadoes, hurricanes, cyclones, snow avalanches, black frost, ice phenomena in the watercourses (ice jams, ice gorges), dust storms, lightning strikes, tsunami, floods, precipitation, extreme snowfall and snowdrift);
2.3.2. Aerological characteristics.
Repeatability of zero wind conditions, wind directions and average wind speeds in 16 rhumbs at the height of 10, 100, 200 and 300 m, repeatability and average intensity of ground and raised inversions in the bottom 1000-meter layer of the atmosphere, repeatability of the atmospheric stability categories, height of the mixing layer in various atmospheric stability categories, average values of the vertical temperature gradient in the layers of 0 - 300, 0 - 600 and 0 - 900 m, joint repeatability of wind speed and direction in 16 rhumbs for various atmospheric stability categories, long-term and short-term atmospheric dispersion of impurities, probabilistic distribution of the atmospheric dispersion parameters for the least favorable (with high probability of exceedance) meteorological conditions for the impurity dispersal in the atmosphere typical for the NPP location area under normal NPP operation conditions (long-term atmospheric dispersion) and under the worst conditions in case of any accident at the NPP (short-term atmospheric dispersion) shall be specified.
2.3.3. Hydrological characteristics.
The following aspects shall be assessed on the basis of historical materials, data of the state and departmental networks of hydrological observations:
average and extreme observed water levels in water bodies and water flows in rivers by months and per year, design maximum and minimum values of the parameters up to the exceedance probability of 1, 0.1 and 0.01%, 99, 99.9 and 99.99% respectively (repeatability of once per 100, 1000 and 10000 years);
dependence between the water levels and flows up to the flow value with the exceedance probability of 0.01% (curve Q = f(H)) for rivers in the NPP location cross-sections;
annual distribution of river stream flows by seasons and months for indicative years (exceedance probability of 50, 95 and 97%);
characteristics of ice phenomena registered within the observation period;
characteristics of tidal phenomena, waves, surges and seiches for seas;
assessment of tsunami hazard in the water bodies and the limits of the territory flooding with a design tsunami wave;
activity of the deformation processes in the coastal areas and beds of water bodies in the NPP location areas;
chemical composition of the surface water sources, description of the capability of surface layers to disperse, dilute or concentrate wastes (hydrological dispersion);
characteristics of water turbidity, suspended and settled sediment flows, alongshore transport, sediments at the NPP water intake structures;
annual probability assessment for hazardous hydrological phenomena.
2.3.4. Basic materials for determination of quantitative and probabilistic parameters and characteristics of hydro-meteorological processes and phenomena.
The list of materials in compliance with which quantitative and probabilistic parameters and characteristics of hydro-meteorological impacts on the NPP are defined hereinafter referred to as basic and obtained in the course of surveys, studies and observations for detection and collection of statistic data on hydro-meteorological processes and phenomena considered in order to determine the comprehensive list of external impacts expected in the NPP construction area due to hydro-
meteorological processes and phenomena shall be provided:
measurement data obtained through standard hydro-meteorological observation programs with hourly measurements at the NPP site;
initial information used to determine design parameters with probabilistic nature of distribution in the long-term context (up to 50 years) containing sets of annual parameter values as well as peaking data obtained from the above-mentioned information sources;
calculated values of impact probabilities and parameters.
2.3.5. Calculation methods for characteristics and parameters of meteorological, aerological and hydrological processes and phenomena.
Those of all analyzed events that are considered in the NPP design shall be listed, and characteristics of their impact on the NPP structures and systems shall be given.
Source data sufficient to calculate the loads on the NPP structures caused by these impacts shall be provided.
Information on the calculation methods for the main parameters and characteristics required to calculate the loads on the structures, assemblies and systems from the following processes and phenomena shall be presented.
2.3.5.1. Calculation of meteorological parameters.
2.3.5.1.1. Wind.
Calculation of wind speed, its repeatability intervals, vertical speed profiles and gust factors shall be specified.
2.3.5.1.2. Tornado.
Input data for calculation of tornado loads shall be provided:
design tornado intensity class;
forward speed;
maximum horizontal velocity of the tornado wall rotation (tangential velocity);
pressure difference between the periphery and the funnel rotation center and design pressure drop rate;
length and width of the tornado path;
characteristics of debris and missiles caused by the tornado.
2.3.5.1.3. Extreme snowfalls and snowpacks.
Extreme height of snow blanket on the horizontal surface shall be substantiated.
2.3.5.1.4. Glaze.
The following shall be provided:
calculation of the rated linear ice load for components with circular cross-section;
calculation of the rated surface ice load for other components.
2.3.5.1.5. Air temperature.
The following shall be provided:
calculation of average temperature changes over time and temperature differential across the component section in warm and cold seasons;
calculation of average daily external air temperatures in warm and cold seasons;
temperature increment calculation;
calculation of the initial temperature corresponding to integration of the structure or any part thereof into the complete system in warm and cold seasons.
2.3.5.1.6. Snow avalanches.
Assessment of probability of a snow avalanche in the NPP location area and its potential characteristics shall be provided.
2.3.5.2. Calculation of aerological parameters.
The following calculations shall be provided:
repeatability of zero wind conditions;
repeatability of wind directions and average wind speeds in 16 rhumbs at the height of 10, 100, 200 and 300 m;
repeatability and average intensity of ground and raised inversions in the bottom 1000-meter layer of the atmosphere;
repeatability of the atmospheric stability categories;
height of the mixing layer in various atmospheric stability categories;
joint repeatability of wind speed and direction in 16 rhumbs for various atmospheric stability categories;
long-term and short-term atmospheric dispersion of impurities, probabilistic distribution of the atmospheric dispersion parameters for the least favorable (with high probability of exceedance) meteorological conditions for the impurity dispersal in the atmosphere typical for the NPP location area under normal NPP operation conditions (long-term atmospheric dispersion) and under the worst conditions in case of any accident at the NPP (short-term atmospheric dispersion).
2.3.5.3. Calculation of hydrological parameters.
The following processes and phenomena shall be analyzed from the viewpoint of the water level rise or decrease at the NPP site:
flood;
coastal area regime of water bodies (positive and negative surges, storm waves);
tsunami;
seiches;
extreme precipitation;
tides;
watercourse icing (ice jams, ice gorges);
changes in water resources (extremely low flow, abnormal decrease in water level);
tropical cyclones.
In this case information on the possibility of flooding shall be provided based on the water level calculations for flood rise and (or) the groundwater level increase; calculations of the maximum level, the maximum water flow due to precipitation, flood rise, seiches, tsunami, waves, ice jams, tides, breakage of natural or man-made water reservoirs shall be presented; calculations of any potential water level decrease due to severe drought, seiches, tsunami, waves, ice jams, negative surges, tides and other phenomena shall be also presented.
2.4. Geological, hydrogeological, seismotectonic and geoengineering conditions.
Results of the engineering surveys (geological with topographic basis) sufficient to substantiate the NPP safety as well as to study seismotectonic conditions in the NPP construction area, other dangerous geological processes (landslides, rockfalls, karst, depressions, mud streams, avalanches, erosion of banks, slopes and stream beds, underground scouring, cryogenic processes, crevasses, subsidence, territory flooding, mud volcanoes and volcanic eruptions) and their combinations shall be provided. Besides predictions of any unfavorable changes in geological, hydrogeological and seismic conditions that can activate any DGP in the course of the NPP construction, operation and decommissioning or preservation shall be provided.
Information on the soil properties and stability shall be given. The list of dangerous geological processes and phenomena and also the calculation methods for the main parameters of geological and seismic processes and phenomena shall be provided.
Information on the chemical composition of underground water sources, description of the capability of surface layers to disperse, dilute or concentrate radioactive wastes shall be presented.
2.4.1. Basic materials for analysis of geological, hydrogeological, seismotectonic and geoengineering conditions at the NPP site.
The list of materials (hereinafter referred to as basic) developed subsequent to the results of surveys and studies in the region in order to detect geological, hydrogeological, seismotectonic and geoengineering conditions at the NPP site shall be presented.
2.4.2. Results of analysis of geological, hydrogeological, seismotectonic and geoengineering conditions.
Analysis results for the basic materials specified in Chapter 2 of the NPP SAR in accordance with item 2.4.1 of this Appendix shall be provided with substantiated opinions on absence or presence of any dangerous geological processes at the NPP site, their quantitative and probabilistic characteristics and parameters to be considered in the course of the NPP design shall be determined.
Information shall be presented separately for each type of processes and phenomena according to the following sequence.
2.4.2.1. Fissure seismotectonic displacements, seismic dislocations, seismic and tectonic upheavals and settling of crustal blocks.
The following shall be specified for the territories with the seismic intensity of VIII or more on the MSK-64 scale (a 12-grade earthquake intensity scale developed by Medvedev, Sponheuer and Karnik) within the radius of 150-300 km from the NPP:
location of the seismogenic subsurface fault, type of the fault;
length of the fault;
amplitude of displacement along the fault (vertical and (or) horizontal);
shares of creep and seismogenic motions in the displacement amplitude;
rocks of the fault banks (sides) in the fissure zone;
location, length and width of the seismically active fault zone including motion parameters (velocities and amplitudes of vertical and horizontal displacement, inclinations) on the sides and in the fault zone before and after a severe earthquake;
parameters of tectonic soil deformations;
thickness of the seismogenic layer.
The same parameters as for tectonic creep and also geological seismicity criteria shall be used for predictable seismotectonic fissure displacements.
2.4.2.2. Neotectonic, quaternary, modern differentiated crustal motions, tectonic creep.
The following information shall be provided:
location of tectonically active faults, regional and other fissures with due regard for buried fissures manifested as geodynamical zones on the surface.
length and width of these fault and fissure zones as well as geodynamical zones including buries active faults and fissures;
structure of tectonically active faults, their disruptive zones and sub-zones;
rising and sinking rate for tectonic blocks and wedges;
tectonic creep velocity in various motion modes (stable, variable, before and after an earthquake);
displacement (rising and sinking, shift, inclination) of tectonic blocks and wedges;
presence of creep;
irregular motion gradient - the ratio between the displacement amplitude and the deformation zone (geodynamical zone) width and a unit of time;
age and displacement amplitude in the youngest tectonic creep and nature of their manifestation in the terrain;
background values of the PDVCM velocity gradient vector at the NPP site, its magnitude and direction.
2.4.2.3. Residual seismic deformations of crust.
The following information shall be provided:
location of tectonically active faults, regional and other fissures with due regard for buried fissures;
length and width of these fault and fissure zones;
structure of tectonically active faults, their disruptive zones and sub-zones;
rising and sinking rate for tectonic blocks and wedges;
tectonic creep velocity in various motion modes (stable, variable, before and after an earthquake);
displacement (rising and sinking, shift, inclination) of tectonic blocks and wedges;
presence of creep;
irregular motion gradient - the ratio between the displacement amplitude and the deformation zone width and a unit of time;
age and displacement amplitude in the youngest tectonic creep and nature of their manifestation in the terrain;
background values of the PDVCM velocity gradient vector at the NPP site, its magnitude and direction.
2.4.2.4. Earthquake
Earthquakes shall be analyzed irrespective of their origin.
The following information shall be provided for each zone of potential earthquake sources within the earth radius from the NPP:
the maximum magnitude, effective depth of the source, seismic intensity in the epicenter (in grades on the MSK-64 scale);
seismic dislocations, seismo-gravitational processes and phenomena, breakage of waterfronts;
seismic intensity and consequences of any dangerous geological and hydrogeological phenomena in the NPP location area;
ground motion parameters on the surface of the NPP site and at the foundation bed level of the structures (calculated or analog accelerograms and consolidated response spectra, frequency characteristics of soils, dynamic factors, maximum acceleration amplitudes, velocities and displacements of horizontal and vertical vibration components, the corresponding periods and number of cycles).
2.4.2.5. Volcanic eruption.
The following information shall be provided:
activity of the volcano (active, dormant, extinct);
characteristics of dangerous phenomena accompanying eruption of an active volcano (lava stream, mud streams, floods, hot cloud, toxic gases);
height and inclination of the volcanic neck;
type of the volcano according to eruption nature.
2.4.2.6. Mud volcanism.
The following information shall be provided:
mud flooding rate;
flooded area increase per a year;
mud level rise rate;
mud flooding area at the specified mud level;
mud temperature in the flooded area and at the blowout point;
parameters of air contamination with gases.
2.4.2.7. Landslides.
The following information shall be provided for active landslides as well as for potentially seismo-gravitational landslides:
location scheme and contours;
slope length and area;
slope terrain (configuration, height, steepness);
development history, origin and age of the slope;
mode of occurrence of weakness zones and surfaces in the slope block and physical and mechanical properties of rocks in these surfaces and zones;
faulting of the slope rocks with assessment of impact on the landslide activity;
assessment of impact of modern tectonic motion and seismic activity on landslide dislocations;
level and pressure regime of the aquifers and conditions for their discharge at the slope with assessment of impact of underground water on the landslide activity;
degrees of weathering, erosion, scouring of the slope, erosion of banks with assessment of impact on landslide development;
character of movement: continuous, regular with long-term and geological time intervals (in new forms);
rate of movement along the slope in various motion modes (stable, variable, before and after an earthquake);
displacements along the slope in various time intervals;
type, humidity and volume of the landslide rocks.
2.4.2.8. Rockfalls and earth slip-falls.
The following information shall be provided for rockfalls on dangerous slopes:
location scheme for the existing and expected rockfalls with the volume exceeding 10 m3;
height and steepness of rockfall-hazardous slopes;
slope surface form;
weathering degree for the slope rocks, presence of any weakened zones, layers of plastic or suffosion-unstable rocks, tectonic faults;
shearing resistance, bulk weight, humidity and stress-strain modulus of rocks in weakened zones and inter-layers, in fissure fillers;
size and volume of the expected rockfall;
preconditioning signs of a rockfall or an earth slip-fall: breakout and falling of separate boulders, expansion of the existing fissures and appearance of new ones, narrowing of displacement rents, periodical crackling, small-scale movements of rock blocks.
2.4.2.9. Mudflows.
The following shall be indicated on the mudflow hazard map of the territory within the radius of up to 50 km from the NPP:
mudflow basin boundaries;
hydrographic network with characteristics of the bed inclinations, zones of mudflow formation, movement and accumulation;
glaciers, moraines, water reservoirs, hydraulic engineering structures, mudflow protection facilities and other objects.
The following shall be specified on the mudflow basin map:
mudflow sources and volume of material in them;
erosional features of the catchment basin terrain and soil and vegetation cover;
mudflow beds and points of any potential jams, volume and activity of rockfalls, screes, landslides in the mudflow bed area;
volume, area, depth, length and width of mudflow deposits in the mudflow accumulation zones.
The following shall be indicated on the potential mudflow movement scheme:
maximum velocity, depth, width and flow;
mudflow flooded areas (with catastrophic damage, with mudflow deposit drift);
mudflow affected areas;
zones of potential loss of slope stability due to scouring;
safe areas and evacuation routes;
contours of the designed and existing facilities.
The section shall specify the following:
origin, conditions of occurrences, formation mechanisms, types and frequency of mudflows;
maximum volumes of non-recurrent mudflow mass transport and dynamic parameters of mudflows;
physical and mechanical properties of soils in the mudflow sources and deposition areas.
2.4.2.10. Snow and stone and crushed gravel-block avalanches.
The following information shall be provided for avalanche-prone mountain slopes:
location scheme for avalanche faults, their morphology, avalanche routes;
height, steepness, surface form, degree of weathering;
speed-up path length along the slope, the section (channel) depth and shape, location of benches in the channel;
material of the sliding surface (rock, soil, snow);
maximum distance of throw and volume of the avalanche, maximum travelling velocity, height and width of the avalanche front in the NPP location area;
effective density of the avalanche material;
maximum avalanche pressure (dynamic, static).
The following shall be specified in order to assess average avalanche hazard at the location site or along the route:
number of sources per 1 km2 of the location site or per 1 km of the valley bottom length;
share of avalanche-hazardous area in the total one;
ratio between the avalanche-affected valley bottom length and the total length at this section;
share of channel avalanche sources in the total area of avalanche-hazardous slopes;
average width of the channel avalanche discharge area.
2.4.2.11. Scouring of banks, slopes, stream beds.
The following information shall be specified for wave abrasion of banks:
annual amount of processing per a unit of the shore length;
active scouring zone length;
displacement of the encroachment line and the bench edge per a year.
For erosion of slopes and stream beds - increase of erosive roughness, length and volume of creeks, displacement of the river bed per a year or any other period of time.
2.4.2.12. Subsidence and crevasses.
Information on the NPP location territory with regard to crevasses of any origin (karst, thermokarst, suffosion, man-made geological workings and pumping of water, oil and gas) shall be provided; intensity of crevasse formation (by number of crevasses per a year at a unit of area) and average diameter of crevasses or average width of longitudinal crevasses shall be determined. Information on any negative landforms (weathering crust, sink holes, craters, hollows, karstic depressions, valleys, subsidence troughs), their contours and plan dimensions (area, length, width) shall be provided.
Average and maximum depths and the earth surface subsidence rates shall be specified for individual typical forms.
2.4.2.13. Underground erosion including underground karst formation.
The following information shall be provided for the territories with any underground erosion (karst, suffosion, leaching) manifestation on the earth surface:
mode of occurrence of the rocks prone to scouring with underground water;
hydrogeological conditions of scouring;
boundaries of the areas with different degree of underground erosion.
The following data shall be indicated on the underground site erosion map:
decompaction and destruction zones;
fissures expanded by dissolution, suffosion, cavity leaching;
channels, galleries, caves and other cavities, their sizes;
dislocation of rocks due to displacement and collapses above cavities, destroyed and decompacted zones;
cavity filling degree and composition;
tectonically weakened zones;
other manifestations of underground erosion.
Karst activity shall be characterized by the ratio between the volume of dissoluble rocks and the volume of the assessed element or the entire block in percent per 1000 years.
Suffosion rate shall be characterized by the bulk volume carried away through suffosion per a year.
depth, thickness, lithological composition, filtration properties, temperature, heat capacity and
heat conductivity of frozen and unfrozen block;
active layer thickness;
amount of heat emitted by the facility to the block;
cryogenic processes and formations (solifluction, moulds, heaves, ice-break crack formations, thermokarst, ice blisters), shapes and dimensions of cryogenic formations (diameter and height of molds, depth, length, width and area of thermokarst crevasses and depressions, thermokarst development depth, area, volume and thickness of ice blisters, sizes of ice-break cracks);
rate of cryogenic processes (rate of heaving, accumulation of ice blisters, solifluction movement, deepening of crevasses and depressions).
2.4.2.15. Deformations of specific soils due to the development of natural and human-induced processes (liquefaction, solifluction and suffosion processes).
The following basic parameters shall be specified for subsiding soils:
stress-and-strain modulus, specific cohesion and angle of internal friction with natural humidity and in water-saturated state, degree of their variation in plan view and in depth;
type of soil conditions in terms of subsidence, thickness of the subsidental stratum and its layers, their variations;
relative subsidence;
initial subsidence pressure.
Any potential interacting and interdependent processes and phenomena of natural and human-induced origin shall be considered separately.
2.4.2.16. Micro-deformations of soils at the foundations of essential structures of the NPP power unit.
Information on micro-deformations of soils at the foundations of essential structures of the NPP power unit shall be provided.
2.4.2.17. Conclusions on classification of processes and phenomena.
Conclusions on classification of processes and phenomena according to their hazard degree, their intensity and frequency of occurrence shall be provided together with substantiations in the form of descriptions, graphical materials (profiles, plans, sections, borehole logs, maps, photographs), results of analysis thereof as well as information on any special-purpose field or laboratory investigations and laboratory analyses. The following information shall be provided.
2.4.2.17.1. Information on the NPP location area.
The following shall be provided for the NPP location area:
analysis of archive and library materials;
cartographic schemes and profiles on the scale of 1:100000 - 1:500000 for geological, tectonic, recent and modern motions with the use of aero-, photo- and space images;
a seismotectonic map or a map of geological seismicity criteria, a detailed seismic zoning map, a schematic map of any potential earthquake source zones with indication of expected maximum
magnitude, its repeatability, effective depth of the source in each zone; historical data on earthquakes and other geological and geoengineering events and processes;
description of lithology and stratigraphy in the region, composition and thickness of quaternary deposits, structure and occurrence depth of the basement rock;
a schematic map of zoning in accordance with the hazard of exogenous geological processes;
data on: freezing depth and thickness of the active layer, landslides, rockfalls, subsidence and crevasses, karst and creek formation; erosion of banks; potential soil movements due to extraction of gas, liquid and solid mineral resources and human-induced loads on the Earth surface (water reservoirs, high-density multi-storey urban development, seismic loads from explosions in quarries); observed settlement and tilt of foundations for the buildings and structures; results of geodetic observations over modern crustal motions;
data on hydrogeological conditions: depth and level variations of groundwater; connections of aquifers with each other and with surface water; recharge and discharge areas of the aquifers; assessment of hydrogeological dispersion in underground water; data on the groundwater level depth with the exceedance probability of 10% and seasonal level variations, flow directions and velocities and also soil permeability in various layers of the section shall be presented on hydrogeological maps;
results of macro-seismic and instrumental seismological studies in the region;
description of soil types, their location at the NPP site;
geological and geophysical profiles and block schemes of the main key strata up to the depth of the first hundreds of meters on the scale of: horizontal 1:100000 - 1:500000, vertical 1:5000 - 1:20000 (at the constructed facility the horizontal scale is 1:20000 - 1:50000, vertical - 1:1000 - 1:5000);
interpreted aero-, photo- and space images;
results of high-accurance repeated geodetic measurements of modern crustal motions.
2.4.2.17.2. Information on the NPP location site.
Maps of geoengineering site zoning and seismic site micro-zoning with indication of geological profiles, test holes and the main structures from the general layout (horizontal scale 1:2000 - 1:10000, vertical - 1:200 - 1:1000), as well as geoengineering profiles, logs of exploratory holes drilled at the site and in the points of essential structure locations and additional profiles along the axes of essential structures (horizontal scale - 1:500 - 1:2000, vertical - 1:50 - 1:200) shall be provided for the NPP location site. All layers (geoengineering elements) shall be distinguished and represented on the profiles, the standard, physical and mechanical and dynamic properties of soils in their natural and water-saturated state (and in natural and unfrozen state for permafrost soils) under dynamic impacts and static impact from the weight of structures shall be specified. Information on any unstable soils with non-steady coherency and properties present in the profile shall be specified.
Recommendations for improvement of soil properties shall be provided.
The following data shall be provided to characterize seismotectonic conditions of the NPP site:
intensity on the MSK-64 scale for the medium grade of soils;
SSE and OBE for particular points of the site with due regard for human-induced changes and soil conditions at the site;
calculated accelerograms and consolidated soil response spectra in graphical and digital format
with the specified probability;
geodynamic characteristics of the site.
2.4.3. Methods, procedures, hardware and testing equipment used to detect geological and geoengineering processes and phenomena and to define characteristics of soils and underground water.
Methods, procedures, hardware and testing equipment used for the following purposes shall be described:
seismic survey, electric prospecting and other geological and geophysical studies at the NPP site prescribed by the regulations in order to detect geoengineering and geological processes, phenomena and factors;
determination of the physical and mechanical properties of soils, specific properties of subsiding, swelling, soft and high-plasticity, loose and permafrost soils in each layer of the investigated stratum in the top section of the geological profile up to the depth of at least 120 m, chemical composition of underground water.
Accuracy characteristics for the equipment, units and methods applied in the course of geological, geophysical and laboratory investigations of the region, location and site in order to supplement, adjust and specify the data on geoengineering and seismological micro-zoning of the site selected for the NPP location shall be provided so that to confirm reliability of the obtained information.
2.4.4. Methods for prediction of parameters and characteristics of factors and processes.
Information on the methods used to predict parameters and characteristics of factors and processes shall be presented, and reliability of the applied methods shall be substantiated.
2.5. NPP impact on the environment and the public.
The main results of assessment of the NPP impact on the environment and the public shall be presented. The following data shall be provided:
background state of the environmental components in the NPP location area and at the NPP site: natural radioactivity, contamination with man-made radionuclides and (chemical) pollutants;
the main migration paths of (chemical) pollutants and radionuclides in natural media;
the main migration paths of (chemical) pollutants and radionuclides via agricultural products;
results of the NPP environmental impact assessment at the construction and operation stages with regard to radiation (potential consequences for the public and ecosystems in case of any radionuclide releases into the atmosphere, discharges into surface and ground water) and non-radiation factors (releases and discharges of chemical substances, acoustic and thermal impacts);
critical paths of ingress of radioactive and chemical substances into the human organism.
2.6. Monitoring programs.
2.6.1. List of programs.
Information on the monitoring programs for the following natural and human-induced factors within the period of the NPP design, construction and operation shall be provided:
present-day crustal motion: vertical and horizontal displacements of the earth surface in the NPP
location area and at the NPP site; detected geodynamical zones, zones of the potential earthquake sources and dangerous tectonic creep, unstable slopes - geodetic monitoring;
seismic phenomena (natural and induced by explosion seismicity and seismic loads) - seismic monitoring;
underground water regime - hydrogeological monitoring;
surface water regime - hydrological monitoring;
meteorological observations - hydro-meteorological and aerological monitoring;
dangerous variations of the groundwater level, humidity, density, load-bearing capacity of the soils at the foundations of essential structures, settlement and deformations of essential structures - geoengineering control in the course of construction and operation;
any other natural phenomena in the NPP location area (landslide phenomena, development of karst potholes).
Information on the environmental monitoring programs shall be provided.
Information on the programs with indication of the types of observations shall be specified for the above-mentioned observations.
2.6.2. Content of the monitoring observation programs.
2.6.2.1. Monitoring observation programs at the NPP site within the periods of design, construction and operation.
The following information shall be provided for programs from the list given in Section 2.6.1 of this Appendix:
lists of observed processes, phenomena and factors as well as types of monitoring;
safety criteria (adopted on the basis of the design requirements as well as established in the NPP design);
locations and elevations of the observation and measurement points;
performed observations and measurements;
brief description of the observation and measurement methods and characteristics of the equipment and test facilities;
recording systems and their location;
the procedure for collection, storage, analysis and transmission of information;
forms of reporting.
2.6.2.2. Information on geodetic monitoring of settlement and deformations of the NPP buildings and structures.
Information on geodetic monitoring of settlement and deformations of the NPP buildings and structures shall be provided. The following information shall be provided:
acceptance of the structures referred to categories I and II according to their importance for
radiation and nuclear safety as established in compliance with federal rules and regulations in the field of atomic energy use after completion of their construction;
materials on the database arrangement for geodetic observations (geodetic monitoring) over settlement and deformations of the buildings and structures.
The following information shall be specified in the description of the geodetic observation (geodetic monitoring) database:
description of the foundation arrangements;
calculated values of design settlement and tilt of buildings and structures;
values of the physical and mechanical properties of soils at the foundations of buildings and structures to be controlled in the course of monitoring;
arrangement schemes for settlement points;
elevation records for the settlement points starting from the foundation arrangement completion (the first cycle of geodetic observations).
2.6.3. Application of the monitoring results.
Results of the monitoring (control) of any processes, phenomena and factors of natural and human-induced origin performed by various methods shall be provided for the entire observation period starting from the NPP site selection with the frequency prescribed by the monitoring programs specified in the NPP design:
geodetic;
seismic;
aero-meteorological;
hydrological;
hydrogeological;
geoengineering;
as well as monitoring of the processes, phenomena and factors related to development of the demographic situation, industry and transport options.
Information on any interruptions in the observations and (or) data losses with indication of the causes thereof shall be provided. Information on recovery and (or) compensation of the lost data with indication of the relevant procedure shall be presented.
Results of environmental monitoring shall be provided.
Information on interdependent processes, phenomena and factors of natural and human-induced origin detected in the course of engineering surveys and studies shall be specified. Presence or absence of such processes, phenomena and factors shall be substantiated. In case of any interacting or interdependent processes and phenomena of natural and human-induced origin information on special programs for monitoring and (or) integrated application of the results of the performed monitoring types as well as analysis of their impact on the NPP safety shall be provided.
2.7. Life support for the personnel and population in the NPP location area and evacuation of the
personnel and population in case of emergency.
Results of analysis of any accidents at the NPP and in the NPP location area caused by intensive earthquakes and other extreme external impacts as well as emergency planning shall be provided. Administrative and technical measures aimed to protect the evacuation routes shall be described.
Any cases with damage of transportation lines, airfields, bridges, tunnels caused by fissures, crevasses, overthrust faults and other surface deformations (gravitational phenomena), screes, rockfalls, landslides shall be analyzed.
Information on the possibility to use the existing access roads in case of emergency, the necessity for relocation or refurbishment of roads, bridges, ports, construction of new transportation routes shall be provided.
2.8. Summary table with a list of external impacts at the NPP site.
The summary table with the list of external impacts at the NPP site considered in the NPP design shall be provided including:
characteristics and parameters of human-induced impacts obtained through calculations and analyses specified in Chapter 2 of the NPP SAR in accordance with item 2.2 of this Appendix;
characteristics and parameters of hydro-meteorological processes and phenomena of natural origin obtained through calculations and analyses specified in Chapter 2 of the NPP SAR in accordance with item 2.3 of this Appendix;
characteristics and parameters of geological, hydrogeological, seismotectonic and geoengineering factors and processes as well as any physical and mechanical properties of soils determined and expected in the course of operation with due regard for the impact of any potential hazardous processes and phenomena;
characteristics of the combinations of natural and human-induced external impacts (both for dependent and independent impacts).
The following information shall be presented for each external impact at the NPP site:
name of a process, phenomenon, factor;
source of hazard, genesis of a process, phenomenon, factor;
degree of hazard;
probability of occurrence;
quantitative values of the impact characteristics and parameters;
additional information.
III. Requirements for the content of Chapter 3 "General provisions for design of the NPP buildings, structures, systems and components'
Information on consideration of the requirements of federal rules and regulations in the field of atomic energy use as well as any other regulatory documents in the design of the NPP buildings, structures, systems and components shall be provided in Chapter 3 of the NPP SAR. Approaches and methods applied in the NPP design shall be described.
Besides information on any administrative and technical measures provided in the NPP design in
order to ensure safety with due regard for adverse external natural and human-induced processes, phenomena and factors in the NPP power unit location area and at the NPP site as determined in Chapter 2 of the NPP SAR shall be presented in Chapter 3 of the NPP SAR.
3.1. Main regulatory criteria and principles for design of the NPP buildings, structures, systems and components.
3.1.1. Assessment of compliance with the requirements of federal rules and regulations in the field of atomic energy use and any other regulatory documents.
3.1.1.1. The list of federal rules and regulations in the field of atomic energy use and any other regulatory documents.
The list of federal rules and regulations in the field of atomic energy use and also any other regulatory documents specifying the requirements considered in the NPP design shall be given.
3.1.1.2. Assessment of compliance with the NPP safety assurance requirements.
Information on compliance with the main requirements for the NPP safety assurance shall be provided:
implementation of defense-in-depth, application of the system of physical barriers in the way of ionizing radiation and RS releases into the environment, implementation of the system of technical and administrative measures for protection of the physical barriers, the public and the environment, accident management measures;
validation of technical and administrative solutions for the NPP safety assurance through the previous experience, testing, studies, prototype operation experience;
quality assurance measures at all stages of the complete NPP life cycle;
consideration of the human factor aimed to prevent errors and mitigate the consequences of any erroneous actions of the NPP personnel;
measures to reduce RS releases and discharges to the environment with due regard for all radioactivity sources;
fire protection measures.
3.1.2. Non-compliances with the requirements of federal rules and regulations in the field of atomic energy use and implemented compensating measures.
The list of any non-compliances with the requirements of federal rules and regulations in the field of atomic energy use existing at the NPP (NPP power unit), the content of these non-compliances, assessment of their impact on the NPP safety and the implemented compensating measures shall be provided.
3.2. Classification of the NPP systems and components.
Information on justification of the presented classification shall be provided.
Classification of the NPP power unit systems and components, buildings and structures in accordance with federal rules and regulations in the field of atomic energy use shall be presented in tabular format.
The table "Classification of the NPP systems and components" shall contain the following
information on the NPP systems and components:
name;
in-plant designation (code);
information on safety classification (in compliance with federal rules and regulations in the field of atomic energy use "General provisions for safety assurance at nuclear power plants" classification according to the purpose, safety importance and performed functions shall be provided for the systems and components and safety classes shall be additionally specified for the components);
seismic category of the NPP components in compliance with federal rules and regulations in the field of atomic energy use;
group for equipment and pipelines covered by federal rules and regulations in the field of atomic energy use governing the requirements for design and safe operation of NPP equipment and pipelines;
other classifications (these data shall be specified in cases when a system or a component is subject to classification in accordance with federal rules and regulations in the field of atomic energy use establishing the requirements for safety-related control systems or the requirements for arrangement and safe operation of hoisting cranes for nuclear facilities or the requirements for NPP pipeline valves and building structures of NPPs).
3.3. Description and justification of layout solutions at the NPP site.
The general layout of the NPP with explication of buildings and structures shall be provided. The explication of buildings and structures shall contain safety classes and seismic categories. Description of the NPP general layout and substantiation of the spatial location of the NPP buildings and structures from the viewpoint of performance of all safety functions by the NPP in all design modes and under all external natural and human-induced impacts considered in the NPP design (from the sources in the NPP location area and at the NPP site) shall be provided. For multi-unit NPPs distance between the units shall be substantiated with due regard for any potential accidents.
Dimensions and engineering and technical solutions for the buildings and structures representing safety-related components shall be described and justified. It shall be specified what safety-related systems are located in these buildings and structures.
Location of the water supply lines, communication lines and any other safety-related networks, access ways, water intake units, switchgear, surface and underground storage facilities for diesel fuel and oil, the transformer yard, storage facilities for fire- and explosion-hazardous substances, pressurized vessels shall be indicated on the general layout of the NPP.
Information on the implemented fire protection measures with regard to location of buildings and structures on the general layout of the NPP, classification of the NPP buildings in accordance to fire and explosion safety and fire resistance rating shall be presented and their necessity and sufficiency for fire protection assurance shall be substantiated.
3.4. Potential scenarios of consequences in case of any external natural or human-induced impacts in the NPP location area, at the NPP site and outside the NPP site.
Results of qualitative and quantitative analysis for potential development scenarios of NPP abnormal operation caused by any of the following impacts shall be presented:
external impacts of natural and human-induced origin in the NPP location area and at the NPP site;
impacts caused by accidents at the NPP site.
It shall be substantiated that analysis of the NPP safety under external impacts is performed in accordance with the requirements of federal rules and regulations in the field of atomic energy use establishing the requirements for consideration of external natural and human-induced impacts on nuclear facilities. This scheme of analysis is also applicable for qualitative and quantitative analysis of consequences in case of any internal impacts on the NPP.
Qualitative analysis results for potential scenarios of consequences in case of any natural and human-induced initiating events in the NPP location area and at the NPP site shall be presented in the form of a table as given in Appendix 5 hereto.
3.5. Parameters of impacts caused by human-induced and natural factors on the NPP buildings, structures, systems and components.
Parameters of external natural and human-induced impacts considered in the NPP design, their combinations as well as internal impacts on the NPP power unit buildings, structures, systems and components shall be presented and substantiated.
The entire range of external impacts on the NPP power unit buildings, structures, systems and components established in the federal rules and regulations in the field of atomic energy use specifying the requirements for consideration of external natural and human-induced impacts on nuclear facilities shall be analyzed.
Information on the loads from several types of impacts as well as in cases when any impact results from another one shall be also specified.
The following material presentation order shall be applied for each type of the considered impacts on the NPP power unit buildings, structures, systems and components:
description of the impact on the NPP power unit buildings, structures, systems and components;
substantiation of characteristics of this impact specified in the NPP design;
description and substantiation of the methods used to convert characteristics of impacts on the NPP power unit buildings, structures, systems and components into the parameters considered in the NPP design;
description of the parameters of this impact on the NPP power unit buildings, structures, systems and components considered in the NPP design and reference to the NPP SAR section where results of analysis of this impact consequences for the NPP safety are presented;
the list of design materials used to develop this section of the NPP SAR.
3.5.1. External impacts exterior in relation to the NPP buildings and structures.
3.5.1.1. Air shock waves.
Analysis of any potential sources and causes of explosions due to breakage of pressurized vessels, tanks with liquefied or compressed gas, fires and explosions in the storage facilities for petroleum, oil and lubricants, accidents on any railway and motor roads situated in the NPP location area and at the NPP site as well as on water transport shall be provided. Parameters used as input data for calculation of ASW impacts shall be specified. Methods used to calculate ASW parameters, to convert ASW parameters into effective loads on the NPP buildings and structures and to calculate dynamic loads from missiles caused by ASW shall be described.
3.5.1.2. Missiles
Analysis of the possibility for missile formation that can be caused by breakage of pressurized equipment with rotating parts in case of any rotation speed exceedance or an accident on pressurized equipment and pipelines shall be provided. Missiles that can be formed in case of destruction of buildings, structures, warehouses with materials, storage facilities for liquefied or compressed gas, pipelines and any other equipment located at the NPP site.
Selection of particular missiles shall be substantiated. The following data shall be specified for the selected missiles: dimensions, weights, energy, speed, ejection angle limits and other parameters required to determine their penetrating power. The points of possible missile hitting (target areas) on the safety-related system components shall be indicated on plans and elevations. Substantiation of the mathematical models used to analyze formation of missiles and to determine their characteristics and flight paths shall be presented.
Probability of missile hitting shall be analyzed for safety-related systems and brief description of the calculation methods shall be provided.
All assumptions used in the probability analysis shall be specified and input data for these assumptions shall be substantiated.
3.5.1.3. Dynamic impacts caused by pipeline breakages.
Impacts on safety-related NPP systems and components, buildings and structures caused by pipeline breakages shall be described and their classification shall be provided. Routing diagrams shall be provided for these pipelines with indication of the safety-related systems and components located within the reach of the broken pipeline fragments and jets in case of the working medium release.
Information on the pipeline rupture points shall be presented and points of application of the resulting loads to safety-related systems, components and building structures shall be determined.
Analysis of the possibility for missile formation in case of a pipeline rupture and analysis of impact of these missiles on safety-related building structures, systems and components shall be provided with due regard for the information presented in Chapter 3 of the NPP SAR in accordance with the requirements stated in item 3.5.1.2 of this Appendix.
Methods used for the following purposes shall be described and substantiated:
dynamic analysis of consequences in case of complete or partial breakage of pipelines;
assessment of shock impacts on safety-related systems and components (jet - support, jet - the neighboring pipeline, jet - equipment, pipeline - pipeline, pipeline - equipment, pipeline - building structures, equipment internals) caused by a pipeline rupture or appearance of a hole.
Description of the applied software tools and information on their validation shall be provided. It shall be substantiated that these ST were used in the areas of application specified in the validation certificates.
In case pipelines are not equipped with limit stops the list of compensating measures and substantiation of their sufficiency shall be provided.
3.5.1.4. Chemical and corrosive impact.
The following information shall be provided:
the list of sources of potentially hazardous chemical and corrosive impacts and information on
their location;
analysis of reactions in the course of interaction between these sources and metal of the equipment, concrete, plastic and insulation coatings, paints;
analysis of these reaction products from the viewpoint of their toxicity, flammability, explosion hazard, chemical and corrosive activity;
levels of corrosion damage for safety-related systems (components) and substantiation of the fact that they do not exceed the permissible limits.
3.5.1.5. Impact of toxic gases and aerosols.
Results of the probability analysis for releases of toxic gases and aerosols into the atmosphere due to integrity loss of any vessels containing toxic substances and accidents with vehicles transporting toxic substances shall be provided.
Description of the assessment methods and values of toxicity parameters shall be specified for these situations.
Probability of gas and aerosol ingress into rooms and their impact on the personnel safety shall be presented.
3.5.1.6. Radiation impacts.
In case any damages of buildings and (or) structures containing radioactive substances are possible due to external natural or human-induced impacts radiation intensity as well as parameters of radionuclide dispersal processes in the atmosphere, surface and ground water shall be defined.
The following analysis results shall be provided:
resistance to radiation exposure for the systems and components that can be subjected to such exposure;
impact of the natural environment radiation.
3.5.1.7. Damage effects of fires.
The probability of fire breakouts shall be analyzed. The following information shall be provided in the analysis:
how damage effects of fires are formed in case of any fires in fire-hazardous rooms with oils, in cable and other rooms, and what load combinations they can contribute to;
for what structures safety factors shall be substantiated in consideration of the damage effects of fires.
Results of review and analysis shall be presented in the relevant sections of the NPP SAR.
3.5.1.8. Meteorological parameters.
3.5.1.8.1. Wind.
The following information shall be provided: description of the methods used to convert wind speed into effective pressure on the wind-facing surfaces of structures; results of wind loads calculations; the applied vibration mode factors for the structures; wind pressure distributions along the
height of structures.
3.5.1.8.2. Tornado.
Description of the following methods shall be provided:
determination of pressure on the surface of structures;
determination of pressure difference between the tornado funnel and periphery;
determination of dynamic loads from debris caused by a tornado.
Information on pressure distribution on flat surfaces and round structures (like the NPP containment) and combinations of the above-mentioned loads with specification of those that can result in the most adverse aggregate tornado impact on the structures shall be provided.
3.5.1.8.3. Extreme snowfalls and snowpacks.
Distribution diagrams for snow loads and factors of snow blanket mass conversion into snow load on the surface shall be provided.
3.5.1.8.4. Snow avalanches.
The following shall be provided:
calculation of static and dynamic pressure of sliding snow on the snow-retaining structures;
calculation of the avalanche impact strength per 1m2 of a fixed stiff obstacle surface located perpendicular to the avalanche movement direction;
calculation of the avalanche load on the damping obstacle during avalanche flow around it;
calculation of pressure in case of a diagonal avalanche impact;
calculation of loads on the facility roof;
calculation of avalanche pressure on the concave surface;
calculation of excessive air shock wave pressure.
3.5.2. Impacts interior in relation to the NPP buildings and structures.
The following information shall be presented in this section in addition to the data similar to those specified in the NPP SAR section compiled in accordance with the requirements of item 3.5.1 of this Appendix:
description of the devices aimed to prevent any impermissible movements due to reactive forces caused by ruptures of the primary circuit pipelines and substantiation of their efficiency in case of design basis accidents;
values of maximum impacts on the containment considered in the design with due regard for ambient pressure, temperature and humidity in the RF containment rooms in case of any design basis and beyond design basis accidents;
description and substantiation of methods and software tools used for strength analysis;
results of the performed strength analysis;
description and substantiation of the maximum corium impact on other systems and support structures considered in the design as well as strategy and method of the corium confinement.
3.6. Impacts occurring under normal NPP operation conditions including transient modes, their parameters.
The list of impacts shall be given, and all operation modes of facilities, buildings, building structures under the following conditions shall be analyzed:
under normal operation of the NPP with due regard for transient modes in case of the power level changes;
in the course of the NPP commissioning with due regard for commissioning works and testing;
in the course of the NPP decommissioning, as well as in other modes leading to additional loads on the building structures to be considered in the design.
Information on the number of cycles and load change values expected within the NPP service life for each mode shall be provided with substantiation of the specified values. Chapters and sections of the NPP SAR where the calculation results for determination of the transient mode parameters for systems and components are contained shall be indicated. Impacts on building, facilities and structures, their quantitative characteristics and parameters shall be specified in this section in the form which is further to be used for the purpose of analysis.
3.7. Design combinations of loads on the NPP facilities, buildings, building structures and equipment.
Information on the approaches used in the NPP design to determine design combinations of the following loads shall be provided:
from natural and human-induced impacts exterior in relation to the NPP buildings and structures;
from impacts with sources located inside the NPP buildings and structures occurring under normal NPP operation conditions.
Impacts on safety-related buildings and structures caused by destruction of non-safety-related structures, systems and components shall be analyzed.
It shall be substantiated that combinations of loads on the NPP buildings and structures selected for consideration in the NPP design comply with the requirements of federal rules and regulations in the field of atomic energy use.
Information on all types of loads on the NPP buildings and structures and their combinations considered in the NPP design shall be provided. Combination factors for each load shall be specified in the table.
The list of design materials used as the basis to develop this section of the NPP SAR shall be given.
3.8. Territory protection against hazardous hydro-meteorological, geological and geoengineering processes
Information and substantiation for the set of measures provided in the NPP design in order to prevent or reduce negative impact of any hazardous hydro-meteorological, geological and geoengineering processes and factors on the NPP territory shall be presented. It shall be substantiated that the measures provided in the NPP design for engineering protection of the NPP territory against any hazardous hydro-meteorological, geological and geoengineering processes comply with the criteria
and requirements specified in federal rules and regulations in the field of atomic energy use establishing the requirements for consideration of natural and human-induced external impacts on nuclear facilities.
The following data shall be provided:
the list and characteristics of hazardous hydro-meteorological, geological and geoengineering phenomena, processes and factors considered in the NPP design;
the list of design materials used as the basis to develop this section of the NPP SAR;
schematic map of design arrangements for the NPP territory protection against DGPs, measures for protection against flooding (run-off control, drainage of surface and ground water), arrangement of anti-mudflow barriers and dams, reinforcement of landslide-hazardous and scoured slopes;
substantiation of efficiency, reliability and sufficiency of the protective measures with indication of any characteristics of external impacts on the NPP site modified due to implementation of these protective measures.
3.9. Protection against flooding.
Information on any technical and administrative measures provided in the design for engineering protection of the NPP against flooding and substantiation of efficiency and sufficiency of these measures shall be presented. It shall be substantiated that the measures provided in the NPP design for engineering protection of the NPP against flooding comply with the criteria and requirements specified in federal rules and regulations in the field of atomic energy use establishing the requirements for consideration of natural and human-induced external impacts on nuclear facilities.
The following information shall be provided:
the list and characteristics of floods considered in the NPP design;
the list of systems that shall maintain their capability to perform safety-related functions in case of any flooding and description of the buildings where these systems are located.
The following information shall be provided for the above-mentioned buildings:
schemes of the facilities with indication of the rooms where systems that shall maintain their capability to perform safety-related functions in case of any flooding are located; location of any entrance apertures and passages in the buildings and structures situated below the maximum flooding level considered in the NPP design shall be indicated;
the list of systems that shall maintain their capability to perform safety-related functions in case of any flooding. The following information shall be provided for each of the above-mentioned systems:
the system code and name, codes of the building and room where the system is located;
flooding elevation upon reaching of which the system cannot perform its safety-related functions;
the maximum flooding elevation considered in the NPP design;
the time for reaching of the flooding level when the system cannot perform its safety-related functions;
description and substantiation of methods used to determine impact of flooding on the NPP buildings, structures, systems and components;
description of the means for protection against flooding provided in the NPP design;
description of protection against water ingress due to cracks in the walls of structures, elimination of water leakages and protection against splashing;
preventive actions of the personnel in case of any threat of flooding and time required to perform these actions;
actions of the personnel in case of flooding and time required to perform these actions;
the list of design materials used as the basis to develop this section of the NPP SAR.
3.10. Justification methods and criteria for ensuring mechanical strength of the NPP buildings and structures
3.10.1. Safety-related buildings, facilities, building structures and foundations.
The following shall be specified:
description and substantiation of the applied methods for design analysis of mechanical strength and stability of buildings, facilities, building structures and foundations in relation to external and internal impacts considered in the NPP design with due regard for peculiarities of the buildings, facilities, hydraulic engineering and geoengineering structures, units and channels as well as their components (leak-tight rooms, foundations, building structures, the containment);
mechanical strength criteria (strength, stability, leak-tightness, fire resistance, seismic resistance and any other criteria established in the NPP design).
It shall be substantiated that the applied methods for justification of mechanical strength of buildings, facilities, building structures and foundations under external impacts comply with state of the art in science, technology and production.
3.10.2. Software used.
The list of software tools used to justify mechanical strength of buildings and structures shall be presented.
The following information shall be provided for each ST:
name and purpose;
implemented calculation method;
basic restrictions and assumptions;
validation data;
results of the program verification by analytical and experimental methods (in case no ST validation has been performed).
3.10.3. Methods for testing and full-scale investigations of the NPP buildings, facilities and structures.
The following information shall be presented in description of testing and full-scale investigations of the NPP buildings, facilities and structures:
techniques and methods for determination of dynamic characteristics of the NPP buildings, facilities and structures;
mechanical strength determination criteria for the NPP buildings, facilities and structures;
procedures for setting of impacts and methods for determination of loads;
methods for assessment of test errors and reliability of the obtained results.
The following information shall be also provided for model-based testing methods:
description of the methods applied to test the models of the NPP buildings, facilities and structures;
description of test stands and testing equipment.
The following information shall be provided in relation to full-scale investigation methods for the NPP buildings, facilities and structures:
description of the applied full-scale investigation methods and programs for the NPP buildings, facilities and structures;
criteria for selection of measurement points.
3.10.4. Mechanical strength criteria for the NPP buildings, facilities and building structures.
The list of safety-related buildings, facilities and structures shall be presented in tabular format. Limits states and criteria of their occurrence established in the NPP design shall be specified in tabular format for the above-mentioned buildings, facilities and structures.
3.11. Determination of loads from external and internal dynamic impacts transferred via building structures to the NPP systems and components.
Methods applied to determine loads on the NPP systems and components in order to analyze their mechanical strength under external and internal dynamic impacts shall be described.
In this case it shall be substantiated that these methods comply with federal rules and regulations in the field of atomic energy use establishing the requirements for design of seismic-resistant NPPs and any other federal rules and regulations in the field of atomic energy use.
3.11.1. Input data for dynamic calculations.
The list of the NPP building and structures and the relevant elevations subject to determination of floor accelerograms and response spectra for further analysis of mechanical strength of the NPP equipment, pipelines and other systems and components under external impacts shall be presented.
Information on analysis of the approach to arrangement of the NPP structures subject to dynamic calculations and the possibility to divide the structures into independent sub-systems shall be provided. The following information shall be presented for each structure:
basic characteristics (dimensions, total weight, weight distribution among sub-systems);
description of the foundation slab arrangements (structures with common foundation slabs shall be specified);
mutual arrangement of individual foundations in order to consider their impact on stress state of the bases.
3.11.1.1. Accelerograms (seismic analysis).
The set of applied OBE and SSE accelerograms for horizontal and vertical ground motion shall be provided. The following shall be determined: the maximum acceleration, the basic frequency, effective accelerogram duration, rise and fall time of the accelerogram amplitude.
All design accelerograms selected from among the available records of past earthquakes or obtained through the use of well-known methods for synthesis of accelerograms on the basis of response spectra shall be accompanied with applicability substantiation for the methods used to select accelerograms for calculations.
The maximum residual displacement shall be specified for accelerograms.
The corresponding response spectra for various damping values used in the design of structures, systems and components shall be provided for accelerograms selected for impact analysis. Frequency ranges for calculation of the spectral values shall be specified.
Comparison of the response spectra obtained in the free field on the ground surface and at the foundation level of safety-related structures with the design spectra shall be performed for each damping value used in the design of structures. It shall be substantiated that the design accelerograms are compatible with the design response spectra.
The method for usage of the selected set of accelerograms for systems and components shall be described.
3.11.1.2. Response spectra (seismic analysis).
Response spectra obtained at the ground surface level and at the foundation level and used to justify seismic resistance of buildings, facilities and structures shall be provided. Response spectra shall be provided for various damping coefficients of horizontal and vertical ground motions.
Sources used as the basis for selection of the design response spectra shall be specified and the selection shall be justified.
The method for usage of the design response spectra in dynamic analysis shall be described.
3.11.1.3. Soil modelling.
Soils at the base of each facility referred to seismic category I as established in accordance with federal rules and regulations in the field of atomic energy use shall be described. The description shall contain the following:
depth of the foundation;
basic dimensions of the foundation;
thickness of sedimentary soils above the rock bottom;
characteristics of soil strata.
Information on the mathematical soil model used in the dynamic calculations shall be provided. If the model of a multi-layer base with underlying semi-space is used the following information shall be specified for each soil layer: shear wave speed, specific gravity, thickness of layers, Poisson's ratio and damping.
3.11.1.4. Damping coefficients for soils.
Damping coefficients for soils shall be specified and the methods used to determine damping
coefficients shall be described or reference shall be given to the sources used as the basis for selection of these coefficients.
3.11.2. Methods for analysis of dynamic behavior of structures.
3.11.2.1. Analysis methods.
Selection of the mathematical models used to calculate vibration parameters of the facilities and building structures referred to seismic category I established in accordance with federal rules and regulations in the field of atomic energy use shall be described and substantiated with indication of the peculiarities used in the course of modelling.
The method used in the seismic analysis for determination of the maximum relative shift of supports shall be specified.
If the modal analysis method was used criteria for selection of the number of natural modes sufficient for analysis shall be specified. The way to consider hydrodynamic effects in the dynamic work analysis for the building structures (tanks) with liquid as well as non-linear effects of dynamic work of the building structures shall be demonstrated.
3.11.2.2. Modelling methods.
Criteria and methods applied in the calculation schemes within the framework of the selected model shall be provided.
Selection of the calculation schemes used for determination of dynamic characteristics shall be described and substantiated for all facilities referred to seismic category I established in accordance with federal rules and regulations in the field of atomic energy use. The main dynamic calculation results shall be provided for each facility. In case modal analysis was used in the calculations frequency, modal mass and modal damping shall be specified for each vibration mode. Uncertainty of the results introduced by reduction of the number of modes used in the calculations shall be assessed.
Dynamic characteristics of the structures obtained for the schemes with due regard for the ground and with fixed base shall be provided. Impact of interaction between the soil and the structure on the main dynamic characteristics shall be assessed.
Information on the peculiarities of structure modelling for separate calculation of their dynamic characteristics per each dynamic impact shall be presented.
Criteria for distinguishing of individual assemblies or parts of the system under consideration as an independent sub-system shall be specified.
3.11.2.3. Interaction of soil with structures.
Description and applicability substantiation shall be provided for the applied methods for analysis of interaction between soil and the structures.
In case the equivalent elasticity method is used the way to obtain the parameters used for analysis shall be described. The methods used in the analysis to take into account physical and mechanical properties of soils, attitude of strata and changes of the soil properties shall be also described. Applicability of the equivalent elasticity method for the conditions of this particular NPP site shall be justified.
Criteria and methods used to take into account the impact of the adjacent structures on the response of the structure under consideration shall be provided in the analysis of interaction between soil and the structures.
3.11.2.4. Interaction of structures.
Information on consideration of interaction between the structures located on the common foundation or on separate foundation shall be provided. Criteria used to take into account joint seismic vibrations of the structures or their parts shall be specified.
3.11.2.5. Impact of earthquake in three mutually perpendicular directions.
The applied method for consideration of the impact of earthquake in three mutually perpendicular directions for determination of seismic response of structures, systems and components shall be described and its compliance with the requirements of federal rules and regulations in the field of atomic energy use establishing the requirements for design of seismic-resistant NPPs shall be substantiated.
3.11.2.6. Method used to consider the torsional impact of earthquakes.
In case any approximation method is used for calculations of any facilities referred to seismic category I established in accordance with the federal rules and regulations in the field of atomic energy use instead of consolidated dynamic analysis of these facilities under vertical, horizontal and torsional impacts the possibility to apply this method shall be justified. Information on the method used to consider torsional impact in the seismic resistance analysis for facilities referred to seismic category I established in accordance with the federal rules and regulations in the field of atomic energy use shall be provided.
3.11.2.7. Combination of natural vibration modes.
In case the response spectrum method is applied information on the procedure used to summarize the relevant vibration modes and to determine force factors and displacement factors (shifts, moments, stresses, deflections and accelerations).
3.11.2.8. Input data and basic results of dynamic calculations.
The following shall be provided:
dynamic characteristics of the structures with due regard for interaction between the structure and the bottom soil;
data on the impact of effects of interaction between soil and the structures on the basic dynamic characteristics;
vibration parameters for the facilities and building structures;
dependence of maximum displacements from the elevation;
dependence of maximum accelerations from the elevation;
The foundation depth, thickness of soil strata above the rock bottom, dimensions of the foundations, total weight of the structure as well as soil characteristics (shear wave velocity, shear modulus, density) shall be specified for the facilities referred to seismic category I established in accordance with the federal rules and regulations in the field of atomic energy use.
3.11.2.9. Floor accelerograms and response spectra.
Description of the methods used to obtain floor accelerograms and response spectra with due regard for three ground vibration components shall be provided. In case the response spectrum method is used to determine floor response spectra conservatism of this method in relation to the over-time direct integration method shall be substantiated. Information on the methods used to obtain design floor response spectra (criteria for definition and smoothing of the envelopes, peak expansion) and the methods for determination of design floor accelerograms corresponding to the design response spectra shall be provided.
Criteria for selection of loads obtained under various external impacts for their further use in the mechanical strength analysis for the NPP systems and components shall be specified and substantiated. The methods used to consider the impact of uncertainties in structural and physical and mechanical soil properties on interaction between the soil and structures, floor response spectra or floor accelerograms shall be described.
The obtained sets of floor accelerograms and response spectra under dynamic impacts considered in the design and defined with due regard for interaction between the structure and the base shall be provided for all facilities referred to seismic category I established in accordance with the federal rules and regulations in the field of atomic energy use.
3.11.2.10. Seismic isolation and other methods for correction of vibration parameters.
Seismic isolation of the reactor building used to reduce dynamic seismic, shock and vibration impacts on the systems and components located in it, substantiations of its reliability as well as rules for acceptance into operation and control in the course of operation shall be described. Efficiency of seismic isolation shall be assessed. Absence of seismic isolation shall be justified in the NPP design.
Information on the methods for protection of the buildings and structures referred to seismic category I established in accordance with the federal rules and regulations in the field of atomic energy use against seismic and other dynamic impacts shall be provided.
The applied engineering features (seismic isolators, hydraulic shock-absorbers) shall be described.
3.11.3. Dynamic loads from the impacts of non-seismic origin.
Methods for determination of the dependence between the resulting loads and the time shall be described or references to the relevant sources shall be provided for dynamic loads of non-seismic origin.
Methods for development of floor response spectra under dynamic loads of non-seismic origin shall be described.
3.11.3.1. Airplane crash.
Information on the methods used to describe interaction of the structure with structural elements of an airplane, both deformable (the hull and wings) and stiff (airplane engines) shall be provided.
Selection of the buildings and structures for analysis of stability in case of an airplane crash, selection of impact directions and impact force application points shall be substantiated. Permissible damage of building structures shall be substantiated as the limit state criteria in case of an airplane crash.
3.11.3.2. Air shock wave.
Information on the methods used to determine the ASW front pressure, loads on the envelope structures of the buildings and facilities, values of dynamic factors (in case of quasi-static calculations)
shall be provided. Permissible damage of building structures shall be substantiated as the limit state criteria in case of air shock wave impact. Methods for consideration of missiles and their fragments accompanying the air shock wave shall be substantiated.
3.11.3.3. Tornado.
Criteria and substantiation for selection of the buildings and structures for stability analysis under the impact of a tornado shall be specified for tornado impacts. The methods used to analyze stability of these buildings and structures under active pressure of the tornado and also under aspiration in the tornado funnel shall be described. Stability of buildings and structures under the impact of missiles accompanying the tornado shall be substantiated.
3.12. Buildings, facilities, building structures, bases and foundations.
3.12.1. General requirements for presentation of information on safety-related buildings and structures.
The following information shall be provided for each building and structure under consideration:
classification of buildings and structures in accordance with federal rules and regulations in the field of atomic energy use;
description of structural solutions for the buildings, facilities, building structures and foundation bases and substantiation of their compliance with the requirements of regulations;
diagrams (drawings) with numbered structural elements specified in the description of buildings, facilities, building structures and foundation bases;
results of substantiation of their strength, leak-tightness, fire resistance and mechanical strength under any external and internal impacts;
the list and substantiation of the arrangements for reinforcement of bases under the foundations of safety-related buildings, facilities and structures;
the list of applied design documents containing substantiation of the structural solutions for buildings, facilities, building structures, foundations, bases, seismic isolation;
description of testing programs and operational suitability control for the structures.
Durability of the building structures for safety-related buildings and facilities shall be substantiated with due regard for the design service life of the NPP.
3.12.2. Reactor building.
3.12.2.1. Summary table of impacts and their combinations considered in the NPP design for the reactor building.
Summary table of impacts (loads) and their combinations considered in the NPP design for the reactor building shall be provided.
3.12.2.2. Bases and foundations.
3.12.2.2.1. General information on the foundations.
Information on the foundation design shall be provided including the following data:
layout of the foundation section of the building;
any other adjacent foundations that can influence stress conditions of the bases;
dimensions, prefabricability, structural arrangement of the junction assemblies, the applied materials (types, grades and classes of concrete and reinforcement);
basic reinforcement, floor lining with the anchorage system;
the anchorage system for fixing of the internal structures on the foundation slab as well as anchorage through the lining (in case any is provided in the NPP design);
shear strain of the foundation under horizontal loads caused by seismic impacts;
seismic isolation of the foundations (if any is provided in the design), the way to transfer horizontal loads (caused by seismic impacts) from the building to seismic isolation devices.
3.12.2.2.2. Stability assurance for bases and foundations.
The following information shall be provided:
information and substantiation of the engineering measures provided in the NPP design in order to ensure stability of the reactor building base and foundations;
information on any arrangement provided in the NPP design in order to prevent unacceptable deformations of the bases due to potential groundwater level increase, under static and dynamic loads, in case of soil liquefaction and also under the impact of any other dangerous geological processes and phenomena considered in the NPP design;
information on transmission of loads and forces to the base surface of the foundations;
information on interaction between the foundations and the soil;
assessment of impact of any adjacent foundations and structures on stress conditions of the base;
assessment of the reactor building foundation capability to take up shear forces in presence of waterproofing.
3.12.2.2.3. Assessment of interaction between structures and bases.
The following information shall be provided:
description of methods for calculation of settlement, tilt, stability of the reactor building;
prediction of settlement within the reactor building construction and operation period with due regard for load increase over time;
calculation results for deformations and load-bearing capacity of the bases and foundations;
design analysis results for interaction of the foundation bearing surface with soils;
design limits of the parameters characterizing stability of the structure and its foundation;
differential settlement and tilting and shifting margins;
substantiation of compliance with the design and regulatory requirements for tilt, settlement and displacements of the reactor building by the NPP commissioning commencement; prediction of
settlement and tilt development for the entire NPP service life; it shall be demonstrated that the tilt of structures does not exceed the limits established by the regulations;
measures aimed to ensure integrity of pipelines and other networks of safety systems coming to the reactor building in case of any considerable settlement and horizontal displacements of the building.
3.12.2.2.4. Monitoring and inspections of foundations.
The following information shall be provided:
description of the program for monitoring of foundation settlements and the reactor building tilt during the NPP construction and operation period;
results of the reactor building foundation inspections during the NPP construction and operation period;
information on the controlled parameters;
information on the applied hardware;
requirements for tests in order to monitor stress conditions of the bottom soils;
requirements for observations over the structure settlement and tilt;
the diagram showing growth of loads on the foundation base over time and prediction of the foundation settlement.
Results of the foundation inspections and observations for the entire observation period specified in the foundation observation program established in the NPP design shall be presented. It shall be substantiated that settlement and tilt values for the structure do not exceed the limits specified in the NPP design.
3.12.2.3. Containment.
3.12.2.3.1. General requirements for the information.
Results of substantiation of the containment compliance with the regulatory requirements as well as the design requirements for strength. leak-tightness and mechanical strength of the containment under external and internal impacts shall be presented.
The following information on the containment shall be provided:
purpose, description and structural peculiarities;
applied structural materials;
considered loads (impacts) and their combinations;
calculation methods;
efficiency assessment for the selected structural solutions;
description of the quality control programs for materials, testing and in-service inspection programs for the containment building structures.
3.12.2.3.2. Leak-tight steel lining.
The following data shall be provided:
a) description of the lining design including:
the list of elements constituting the lining;
substantiation of the lining thickness selection;
information on any structures ensuring leak-tightness, information on weld joints of the lining manufactured in factory conditions, at the pre-assembly site and in the course of installation; weather strips arranged above weld joints; methods for fastening of embedded parts supporting the equipment and pipelines to the lining sheets, support elements, brackets penetrating through the lining sheets and fastened to the reinforced concrete wall; design of anchorage to the concrete blocks of the containment bottom, cylinder and dome; any other structural elements;
information on any structures ensuring leak-tightness of the bottom in the output area of anchor rods intended for fastening of internal structures and equipment supports to the bottom;
description of the structure for fastening of the metal lining to the concrete blocks of the containment bottom, cylinder and dome;
drawings and diagrams of the lining structure;
b) information on the methods for the lining behavior analysis including:
description of the calculation methods, information on the applied software tools; any assumptions used and validation and verification information shall be specified for software tools;
substantiation of reliability and representativeness of experimental studies (in case any results of experimental studies were used);
substantiation of the lining stability under compression and increased temperature;
values of shearing and tearing forces at the junction points between dowels and the lining;
critical forces and their comparison with the effective forces (with the specified interval of anchor rods or angles) under all loads (impacts) and their combinations considered in the design as well as under loads from simultaneous thermal impact and compression;
design strain and shear resistance of the metal lining material in the area of anchorage devices;
substantiation of the weld joint leak-tightness maintenance in case of loss of the metal lining stability;
safety margins for loss of the lining stability under any thermal loads and compression loads considered in the NPP design as well as any other loads (impacts) and their combinations considered in the NPP design;
relative deformations of the lining caused by compression, compression strain in the lining under the impact of simultaneous forces for different containment zones;
c) information on quality control for the materials;
physical and mechanical properties of steels used for the lining, anchors, embedded items;
requirements for the lining quality control in the course of manufacturing, assembly and installation;
methods for assessment of technical condition of the lining subsequent to the testing results;
information on any measures enabling to maintain the design quality level for the lining in the course of operation.
3.12.2.3.3. Reinforced concrete structure of the containment.
The following data shall be provided:
a) description of the reinforced concrete containment design, its dimensions and the ,ost critical structural elements; general description of the containment shall include:
description of the outer and inner containment (configuration, basic dimensions);
arrangement of the cylindrical part of the containment, information on concrete and reinforcement rods, presence of stiffeners;
arrangement of the dome and the supporting ring (if any), information on concrete, reinforcement, metalwork and pre-stressing of the dome; brief description of the dome installation;
arrangement of the foundation part, information on concrete and reinforcement, description of support structures for anchors of the stressed reinforcement elements, description of large apertures and their reinforcement (manholes and locks for equipment, operating personnel, apertures for the main pipelines);
the main structural fasteners penetrating through the metal lining sheets and fixed to the reinforced concrete wall;
requirements for the pre-stressing system;
design loads (inpacts) and their combinations;
characteristics of the applied materials (concrete; reinforcement steel, its mating and welding; anchorage of the structural elements) and prediction of any changes in their properties in the course of operation;
b) information on the analytical methods used to design the reinforced concrete containment:
description of the calculation methods, information on the applied software tools; any assumptions used and validation and verification information shall be specified for software tools;
substantiation of reliability and representativeness of experimental studies (in case any results of experimental studies were used).
Description and substantiation of the following shall be presented:
methods for consideration of loads;
methods for consideration of creep deformation, concrete settlement, crack formation in concrete and plastic deformations caused by crack opening in the calculations;
methods for analysis of SSS with information on any assumptions used in selection of the limit conditions and the reinforcement selection methods;
methods for calculation of SSS at the areas of the largest apertures in the containment; the basic results of the obtained stress and strain condition analysis and references to the performed calculations shall be provided;
calculation uncertainties and sensitivity of the obtained results to any potential changes of the applied assumptions and material characteristics.
The following information shall be provided for the containment pre-stressing system:
substantiation of the current CPSS condition compliance with the design requirements and requirements of federal rules and regulations in the field of atomic energy use establishing the requirements for arrangement and operation of the NPP LSSs;
limit states of the containment and the pre-stressed tendons established in the NPP design;
assessment of safety margin up to the limit state of the containment for the most important containment sections: apertures, manholes, areas of fastening assemblies, the joint between the cylindrical part of the containment and the supporting plate and the dome;
substantiation of compliance with the requirements of federal rules and regulations in the field of atomic energy use establishing the requirements for arrangement and operation of the NPP LSSs;
description of the installation method, stressing of the CPSS tendons and also tension control methods.
The following information on the programs for material quality control, testing and in-service inspections of the containment shall be provided:
information in the quality control program in the course of the containment manufacturing and installation;
mechanical properties of the materials and physical and mechanical properties of the structural materials (for concrete components; reinforcement rods and their weld joints; the pre-stressing system; embedded parts; anti-corrosion compounds used to protect the CPSS tendons);
monitoring methods for the pre-stressing system (if any);
concrete pouring control methods with indication of the assembly tolerances for reinforcement elements;
information on the diagnostics systems for the containment building structures, observations over tilt, settlement, SSS; information on the containment equipment with grade pegs, benchmarks, instruments shall be provided and the method for data registration and processing shall be described;
information on any measures enabling to maintain the design level of parameters characterizing operability of the containment.
3.12.2.3.4. PHRS building structures (if any provided in the NPP design).
Description of the PHRS building structures with the required drawings and diagrams shall be provided. Information on the purpose of various PHRS rooms and the requirements for the PHRS building structures shall be specified. Information on any considered loads and load combinations on the PHRS structures as well as their limits states with indication of the criteria shall be provided.
References shall be given to any materials where strength and stability of the internal structures under process loads and impacts in various NPP operation modes (including accidents) as well as in case of any natural and human-induced external impacts are substantiated. Conclusions subsequent to calculation results shall be specified. Compliance with the limit state criteria shall be substantiated in these conclusions.
3.12.2.3.5. Internal rooms and building structures of the containment.
The following information shall be provided:
the list of internal rooms and building structures of the containment, the applied loads and combinations of loads, limit states;
the list of rooms where fire breakout is possible with indication of potential causes of fire hazard, information on compliance with the requirements for fire resistance of the internal structures;
description of layout and structural solutions with drawings of the internal structures;
information on the materials, reinforcement, loads from the equipment installed in the containment rooms;
justifying calculations of the internal containment structure strength with information on the design diagrams of the internal building structures, substantiation of the applied assumptions and conclusions on the calculation results for these structures;
description of the quality control programs for materials, testing and in-service inspection programs for the building structures. The final objective of testing and the sdopted criteria for assessment of the results shall be specified.
3.12.2.4. Auxiliary structures of the containment.
The following shall be provided:
description of the foundations and building structures for the auxiliary structures of the containment, their plans and basic elevations;
description of the purposes of the auxiliary rooms and design requirements for them;
any loads (impacts) on the structural elements of the auxiliary structures and their combinations considered in the design, definitions of the limit states for the structures and their criteria;
information on consideration of the mutual impacts of individual structural elements of the auxiliary structures via junction assemblies with indication of forces and loads transmitted to the foundations;
characteristics of the applied materials (concrete; reinforcement steel, welded and mechanical joints of reinforcement elements, their mating and welding; anchorage of the structural elements) and prediction of any changes in their properties in the course of operation;
description and substantiation of the applied calculation methods and calculation models for the auxiliary structures, basic results of the calculations with references to the reports where they are taken from, information on validation of the calculation programs;
description of the applied calculation models for the auxiliary structures with substantiation of any adopted assumptions;
calculation results and their comparison with the regulatory and design criteria used as the basis for the conclusions on strength, deformability, crack resistance of separate elements and the entire structure; applied safety margins for stresses and forces in the reinforcement elements and concrete, for deformations and crack resistance;
description of the quality control programs for materials, testing and in-service inspection programs for the building structures of the auxiliary containment structures.
When describing arrangement of the auxiliary structures it is necessary to comply with the number of SS channels; layout solutions eliminating simultaneous damage of different SS channels in case of an airplane crash and any other external impact shall be described.
The conclusion on efficiency of the adopted structural solutions shall be provided.
3.12.3. Other NPP buildings and structures.
The following information shall be provided:
the list of safety-related buildings and structures;
summary table of loads (impacts) and their combinations considered in the NPP design for the NPP buildings and structures specified in the above-mentioned list;
description of the calculation methods, information on the applied software tools; any assumptions used and validation and verification information shall be specified for software tools.
The following information shall be provided for each NPP building and structure specified in the above-mentioned list:
detailed description of layout and structural solutions;
basic diagrams and drawings;
substantiation of strength, leak-tightness, fire resistance and stability of the building structures of buildings and facilities under any external impacts as well as stability of their bases and foundations;
substantiation of compliance with the requirements of federal rules and regulations in the field of atomic energy use governing design of NPP buildings and structures and design of seismic-resistant NPPs;
description of the quality control programs for materials, testing and in-service inspection programs for the building structures.
The conclusion on strength and stability of all considered buildings, facilities and building structures as well as stability of the NPP bases and foundations shall be made subsequent to the results of the performed calculations.
3.12.4. Diagnostics of building structures.
Information on the diagnostic system for facilities and building structures, the system for monitoring of tilt, settlement, stress and strain conditions of the building structures, forces in the pre-stressing tendons (for pre-stressed containments), vibrations of buildings and state of their foundations shall be provided.
Particular facilities and building structures subject to mandatory diagnostics in order to ensure safety of the NPP power unit shall be listed. Information on the measured parameters, quantity and type of measuring instruments, settlement points, the measurement procedures, recording and storage of the results as well as on use of the results for life management of the NPP buildings and structures as well as references to the monitoring programs shall be provided.
Facilities and structures subject to mandatory diagnostics shall be specified. Information on compliance with the regulatory requirements for equipment of the NPP buildings and structures with benchmarks, systems for monitoring of tilt, settlement, vibration of buildings and structures, state of the foundations and also their SSS shall be provided. Information on the observation program shall be
presented for the above-mentioned observations.
Subsequent to installation of the equipment results of the following observations shall be provided:
settlement and tilt of buildings and structures;
stresses in the structures and foundations;
deformations of the containment;
forces in the pre-stressing tendons (for pre-stressed containments).
3.12.5. Investigation program and action plans for inspection of safety-related NPP buildings and structures.
Information on any studies and observations over the state of foundations, buildings, facilities, building structures, soils, groundwater, description of monitoring of the general conditions of the facilities and radioactive leaks in the wells provided in the NPP design shall be presented.
3.13. Methods of justification of strength and operability of NPP equipment, pipelines, systems and components taking into account the loads caused by natural and human induced impacts and transferred via structural elements of buildings and structures.
Information containing the basic calculations for determination of the capability of mechanical, instrumentation and control and electrical systems to perform their functions under combined impact of external conditions, internal emergency impacts and normal operation impacts shall be presented.
3.13.1. Consideration of external conditions in analysis of mechanical and electrical equipment.
3.13.1.1. Equipment identification and external conditions.
Information on location of the safety-ensuring equipment inside the containment or in any other places that shall function in the course of and after any design basis accident shall be provided. Both normal and accident conditions shall be defined for each type of equipment. For external accident conditions these parameters shall be specified depending on time with indication of any causes of such external conditions.
Potential duration of operation under external accident conditions shall be specified for each mechanism.
3.13.1.2. Tests and studies.
Information on any tests and studies performed or to be performed for each mechanism and equipment in order to check its operability under the combination of such impacts as temperature, pressure. humidity, chemical composition and radiation shall be provided. Their particular values shall be indicated.
Testing results shall be specified for each type of equipment.
3.13.2. Mechanical systems, equipment and pipelines.
3.13.2.1. Individual components of mechanical systems and equipment.
Information on the methods for strength and stability analysis of the components of mechanical systems, equipment and pipelines shall be provided.
3.13.2.1.1. Transient analysis.
The list of transient modes to be used in the cyclic strength calculations for all mechanical systems, equipment, pipelines and support structures shall be given. For each transient mode information on the number of transient modes of this type considered in the NPP design as well as the number of load variation cycles within the transient mode with correctness substantiation for the presented values shall be provided. Information on the sources containing all calculations for determination of transient mode parameters shall be presented.
3.13.2.1.2. Software tools used in calculations.
The list of software tools used for static and dynamic analyses performed in order to determine structural and functional integrity of all systems, assemblies, equipment and support structures referred to seismic category I established in accordance with the requirements of federal rules and regulations in the field of atomic energy use shall be presented. Brief description of software tools, area of application as well as information on the program validation or verification shall be provided in the list.
3.13.2.1.3. Experimental studies of stresses.
Information confirming feasibility of experimental stress analysis methods shall be provided if these methods are used instead of or in addition to analytical calculation methods for the equipment referred to seismic category I established in accordance with the requirements of federal rules and regulations in the field of atomic energy use.
3.13.2.1.4. Assessment of accident conditions.
Analytical and (or) experimental methods used to assess stresses in the equipment referred to seismic category I established in accordance with the requirements of federal rules and regulations in the field of atomic energy use under accident conditions shall be described. Description shall include substantiation of compatibility of these methods with the applied type of dynamic system analysis.
Information and substantiation of dependence between stresses and deformations used for the equipment strength analysis shall be provided.
3.13.2.2. Dynamic tests and analysis.
Criteria, testing and dynamic analysis methods used to confirm structural and functional integrity of systems, pipelines, mechanical equipment subjected to dynamic loads with due regard for the loads caused by the coolant flow and seismic impacts shall be provided.
3.13.2.2.1. Pre-operational measurements of the equipment and pipeline vibrations.
Information on the results of vibration measurements for all equipment units and pipelines belonging to safety-related NPP components and subjected to vibration loads in the course of functional tests during the commissioning period shall be provided.
Information on any measurements of amplitude and frequency characteristics of equipment and pipeline vibrations during modelling of different operation modes as provided in the course of the commissioning works, guidelines on the list of measuring regimes, selection of control and measurement points, criteria for assessment of the measurement results as well as information on the commissioning works programs regulating performance of these measurements shall be presented.
3.13.2.2.2. Seismic resistance checking tests for safety-related mechanical equipment.
Seismic resistance tests for mechanical equipment required to confirm structural integrity and operability in the course of and after any seismic impacts shall be described. The following information shall be provided in the NPP SAR:
seismic resistance criteria, testing methods and the main parameters of test modes, the method to consider impact of the equipment location elevation on the parameters of selected test modes, as well as sufficiency substantiation for the seismic characteristics determination program; information on consideration of broad-bandness in seismic excitation, random direction of seismic impacts and dynamic interaction between seismic loads in different directions in development of the seismic resistance checking programs shall be also specified;
methods used to check operability of mechanical equipment referred to seismic category I established in accordance with federal rules and regulations in the field of atomic energy use during and after SSE impact and to confirm structural and functional integrity of the equipment subsequent to impact of several OBEs in combination with normal operational loads; this provision shall be applied to such mechanical equipment as fans, pump drives, actuators of the reactivity control devices, tube bundles of heat exchangers, valve drives, racks for batteries and tools, control panels, control rooms and cable routes;
techniques and methods for analysis and testing of the supports for mechanical equipment referred to seismic category I established in accordance with federal rules and regulations in the field of atomic energy use as well as verification methods used to consider any potential increase of design loads (amplitude and frequency) in case of seismic vibrations.
Results of tests and analyses shall be provided in order to confirm correctness of compliance with the requirements of federal rules and regulations in the field of atomic energy use and design criteria and to substantiate sufficiency of the performed tests.
3.13.3. Electrical equipment.
Methods used to substantiate operability of electrical equipment shall be described, and information demonstrating compliance of the technical specifications and testing methods with the regulatory requirements and design criteria shall be provided.
3.13.3.1. Criteria for electrical equipment operability check under dynamic loads.
The range of electrical equipment referred to seismic category I as established in accordance with federal rules and regulations in the field of atomic energy use shall be provided.
Seismic resistance verification criteria including the criteria for selection of special tests or analytical methods, determination of the input vibration parameters as well as sufficiency substantiation for the program of stability verification under dynamic loads shall be provided.
The list of loads applied to the equipment in order to check its operability shall be presented.
3.13.3.2. Methods and procedures for testing of equipment stability and operability under dynamic loads.
Information on the methods and procedures used to check seismic resistance of electrical equipment referred to seismic category I established in accordance with the federal rules and regulations in the field of atomic energy use shall be provided.
In shall be substantiated that the above-mentioned electrical equipment performs any safety functions prescribed in the NPP design during and after a SSE and maintains its operability after a OBE.
3.13.3.3. Methods and procedures for analysis or testing in order to verify stability of support structures.
Information on the methods and procedures for analysis or testing in order to verify stability of support structures for electrical equipment referred to seismic category I established in accordance with the federal rules and regulations in the field of atomic energy use under dynamic loads and verification methods used to consider any potential increase of design loads (amplitude and frequency) in case of seismic vibrations shall be provided.
3.13.4. Power-generating equipment.
The list of power-generating equipment referred to safety-related components shall be given. Criteria used in the course of testing or analytical studies in order to justify operability of the power-generating equipment shall be defined. Information on any peculiarities of the testing programs and calculation methods, applied combinations of loads shall be provided.
The main results of strength calculations confirming operability of the power-generating equipment shall be presented. Methods and procedures used to check stability of support structures for the power-generating equipment under any selected combinations of effective loads with due regard for loads from external impacts shall be presented.
3.13.5. Pump sets and valves.
The list of all pump sets and valves referred to safety-related components shall be provided. Criteria used in the course of testing or analytical studies in order to justify operability of the pump sets and valves shall be defined. Information on any peculiarities of the testing programs and calculation methods, applied combinations of loads shall be provided. Information on any maximum stress and deformation levels obtained in the course of testing programs or analytical studies as well as the operability check results for pump sets and valves for the entire planned service life shall be specified.
3.13.6. Instrumentation and controls and the APCS equipment.
The list of all instrumentation and controls, the APCS equipment and their support structures referred to seismic category I shall be presented. Criteria for verification of seismic resistance and stability under external impacts shall be specified. Parameters used as input data for verification of seismic resistance and stability under external impacts shall be presented. Information on the methods and procedures used to verify stability of instrumentation, controls and equipment under external impacts shall be provided. It shall be substantiated that this instrumentation and equipment performs the safety-related functions prescribed in the NPP design during and after any external impacts considered in the design. Methods and procedures used to verify stability of support structures for I&C and APCS equipment under any external impacts as well as verification methods used to consider potential increase of design loads under any external impact shall be provided.
3.13.7. Ventilation equipment and air ducts, equipment of filtration systems.
Stability of ventilation equipment and air ducts as well as the equipment of filtration systems under any loads specified in the NPP SAR in accordance with items 3.4, 3.5 of this Appendix shall be substantiated.
The range of equipment, the list of safety-related air ducts and filtration systems shall be provided.
References to the sources containing complete analysis of strength and stability under the impacts of internal origin and natural and human-induced external impacts shall be given.
The following information shall be provided:
data on design loads and their combinations;
calculation and analysis methods, modelling methods, methods for dymanic analysis of air systems under loads;
testing methods, test stands and testing equipment;
criteria for stability and strength of ventilation equipment, air ducts, filtration systems;
ways of fastening on the structures, strength of support assemblies, explanatory diagrams and drawings.
3.13.8. Hoisting, transport and handling equipment.
Strength, durability and stability of hoisting transport and handling equipment with due regard for the complete range of impacts specified in the NPP SAR in accordance with items 3.4, 3.5 of this Appendix shall be substantiated. In this case information on acceptability of the methods selected for substantiation and reliability of the results shall be provided.
3.13.9. Seismic instrumentation.
3.13.9.1. Measurement program.
The measurement program for the parameters of seismic impacts shall be provided and substantiated.
3.13.9.2. Description of instrumentation and controls and their location.
Information on seismic instrumentation and controls to be installed on the selected blocks of systems and in the selected facilities referred to seismic category I established in accordance with federal rules and regulations in the field of atomic energy use shall be provided. Besides selection of these facilities, blocks and locations of I&C shall be substantiated and information on the procedure for usage of readings of these instruments after earthquakes in order to verify seismic resistance calculations shall be provided.
3.13.9.3. Notification of the MCR operator.
Information on the measures to be implemented within the shortest possible time to notify the MCR operator on the acceleration type rate and response spectra values shall be provided. Besides the particular preset values that shall start taking readings of seismic instrumentation in order to submit them to the operator shall be substantiated.
3.13.10. Software used.
The list of software tools used to substantiate stability of the NPP equipment, pipelines, systems and components under external impacts shall be given. The following information shall be provided for each software tool:
purpose of the program;
calculation method implemented in the program;
basic restrictions and assumptions imposed by the software tool on the class of problems under consideration;
validation and verification information.
3.13.11. Testing methods for systems and components.
Criteria, testing and dynamic analysis methods used to confirm integrity and operability of pipeline systems, mechanical equipment subjected to dynamic loads with due regard for the loads caused by the coolant flow and seismic impacts shall be described and substantiated.
IV. Requirements for the content of Chapter 4 "Reactor"
Information and analysis results required to substantiate safety of the reactor operation within the design service life under normal operation conditions and in case of any abnormal operation including accidents as well as any information necessary to analyze abnormal operation in order to obtain results specified in Chapter 15 of the NPP SAR shall be provided in Chapter 4 of the NPP SAR.
4.1. Purpose of the reactor.
4.1.1. Purpose and functions.
Information on the configuration, purpose and functions of the reactor and its components shall be provided.
Information on any regulatory documents used to develop the RF design with regard to the reactor shall be specified.
4.1.2. Design basis.
The following information shall be provided:
design characteristics of heat energy generation;
nuclear fuel used;
design of the reactor and its components;
the mode of nuclear fuel use;
NF burn-up;
duration of the RF power operation within a year;
design service life of the reactor components;
reliability indicators for the reactor components and systems.
4.2. Design of the reactor.
4.2.1. Description of the reactor.
Description of the reactor, its components and systems with references to the relevant sections of the NPP project shall be presented.
Information on the reactor and brief information on the building where the reactor is located, protection of the reactor building against any external and internal impacts of natural and human-induced origin (specified in Chapter 2 of the NPP SAR) and any events at the NPP site exterior in relation
to the reactor building shall be provided. Orientation of the reactor in relation to the reactor building, mutual arrangement and interaction of the described equipment and systems and their mutual impacts shall be understandable from the provided data.
The list of the reactor systems (components) shall be given. The list shall include:
nuclear core;
CPS control rods;
CPS actuating mechanisms;
reactor pressure vessel;
reactor internals;
in-core instrumentation system;
other systems and components included into the reactor in accordance with the RF design.
4.2.1.1. Nuclear core.
4.2.1.1.1. Purpose and design basis.
Information on the purpose and design basis for the nuclear core, fuel assemblies, their safety class and seismic category shall be provided.
The list of regulatory documents containing the requirements to be considered in the core design as well as the design criteria and safety principles, basic requirements for the core configuration and structure of fuel assemblies shall be presented.
In case of any reactor core refurbishment the project materials for such refurbishment and safety analysis for such refurbishment as well as sufficiency substantiation for the performed bench-scale and reactor research works shall be provided.
4.2.1.1.2. Description of the core configuration.
The core configuration and structure of fuel assemblies shall be described, their general outlines showing mutual arrangement, basic dimensions, fastening methods and orientation in relation to the reactor axes, diagrams of coolant distribution for fuel assemblies in the core shall be provided.
Fuel load patterns shall be provided for the first loading of the core, intermediate loadings and steady reactor operation mode, and information on the NF quantity shall be presented. Reference to the relevant drawing from the technical project list for the nuclear core and fuel assemblies shall be given for each figure presented.
Information on the nuclear core and fuel assemblies shall be accompanied by the list of their basic technical characteristics.
4.2.1.1.3. Materials, nuclear fuel, coolant.
Selection of the materials for fuel assemblies in the nuclear core shall be substantiated and nuclear fuel and coolant shall be described; in this case the following information shall be provided:
a) on structural materials:
mechanical, thermal and physical properties of the structural materials depending on the
exposure dose and temperature;
time of irradiation of structural materials during their operation in the reactor and the dose accumulated within the design service life;
corrosive interaction with fission products and coolant depending on the NF burn-up, temperature and irradiation period for the structural materials;
cyclic strength depending on the exposure dose, temperature, load and number of cycles;
b) on welding:
applied types of welding with the list of regulatory documents governing the requirements for welding;
operation experience for weld joints or their testing under similar conditions;
differences of mechanical and corrosion properties of weld joints in comparison with base metal under normal operation conditions, in case of any abnormal operation and accidents;
c) on nuclear fuel:
chemical composition, enrichment, density, loading, irregularities in distribution of density and fissionable isotopes, methods for their control, validation of the control methods;
NF creep flow and swelling depending on temperature, exposure dose and load;
mechanical, thermal and physical properties depending on burn-up level, temperature, content of fissionable isotopes (melting temperature, thermal capacity, thermal conductivity, heat expansion, breaking point);
compatibility with the cladding material, mass transfer depending on burn-up, temperature and time;
behavior in case of any abnormal operation and accidents (criteria for loss of a fuel element leak-tightness, contact with coolant, temperature increase);
possibility and feasibility of NF processing;
in case of any refurbishment of the reactor core with usage of different fuel type - results of calculations and experimental studies for this fuel as well as predictive estimates of permissible burn-up depth;
d) on absorbing materials:
chemical composition, dimensions, NF enrichment with regard to absorbing materials, density, control methods, validation of control methods;
compatibility with the cladding materials;
behavior in case of accidents (loss of leak-tightness, contact with coolant, temperature increase);
behavior under radiation and changes of properties;
in case of any refurbishment of the reactor core - results of justifying calculations and experimental substantiation of AE behavior under radiation and predictive estimates of permissible poison nuclide burn-up in the AE;
e) on the coolant:
thermal and physical properties;
any admissible impurities.
4.2.1.2. Reactor internals.
Information on the reactor internals shall be provided.
Information on the following components shall be presented:
core baffle;
core barrel;
protective tube unit;
surveillance specimens;
in-core I&C devices with connectors and fasteners;
other reactor internals.
The following data shall be provided for structural materials of the reactor internals:
mechanical, thermal and physical properties depending on the exposure dose (in the units of "displacements per atom") and temperature (yield point and breaking point, residual plasticity, thermal conductivity, thermal capacity);
time of irradiation of structural materials in the reactor and the dose accumulated within the design service life;
cyclic and static strength depending on the exposure dose, temperature, load and number of cycles.
4.2.2. Control and monitoring.
The list of the controlled parameters for the core and fuel assemblies, frequency of control, range of parameter measurements, permissible measurement errors, configuration and location of sensors shall be provided and substantiated.
Information on the core condition monitoring and the RF power control shall be provided with due regard for the data specified in Chapters 4 and 7 of the NPP SAR:
on protections and interlocks, controllers, diagnostic systems, automated control algorithms;
on emergency protection;
on the neutron flux monitoring system;
on the drive control system;
on the in-core monitoring system;
on the tightness control system for fuel element claddings;
on the RF power control system;
on preventive protection and interlocks.
In case of any refurbishment of the reactor core associated with usage of new fuel type applicability of the systems engaged in the core monitoring and control or refurbishment of the above-mentioned systems shall be substantiated.
Information on any engineering features and methods provided in the NPP design for tightness control of fuel element claddings in the shut-down and operating reactor in order to ensure reliable and timely detection of any leaky fuel elements shall be presented. Procedures used to control tightness of fuel element claddings and to detect causes of tightness loss in the shut-down and operating reactor shall be provided and substantiated.
4.2.3. Tests and inspections.
Information on any control programs and methods, tests and studies performed in the course of the design development, manufacturing, commissioning and operation and confirming the design characteristics of the core and fuel assemblies shall be presented; the list of regulatory documents defining the requirements for the scope and methods of control and testing shall be given. Incoming control programs provided at the NPP for the core assemblies shall be presented.
In case of any refurbishment of the reactor core the testing procedures and programs for refurbished fuel assemblies shall be specified.
4.2.4. Design analysis.
4.2.4.1. Normal operation.
Functioning of the nuclear core and fuel assemblies under normal RF operation conditions, in the course of power rising to the MCL, in transient modes in case of scheduled start-ups and shutdowns shall be described. Characteristics describing the core state in the above-mentioned modes and interaction with other reactor systems shall be specified.
4.2.4.2. Design limits and conditions.
Design limits and conditions as well as safe operation limits and conditions related to the nuclear core (in case these are established in the NPP design) shall be specified and substantiated with due regard for the information provided in Chapter 16 of the NPP SAR. In this case if no above-mentioned limits are established in the NPP design the relevant information shall be specified in this section of the NPP SAR.
The following information shall be provided:
limits with regard to fuel (fuel temperature or absence of melting, mean radial cross-sectional enthalpy);
critical heat flux ratio;
limits with regard to fuel element claddings (temperature, equivalent oxidation rate);
limits with regard to reactivity coefficients;
limits with regard to power and power change period;
limits with regard to the primary circuit coolant activity;
presence of any dependence between the primary circuit coolant activity limit and the safe
operation limit with regard to FE damage;
any other limits established in the NPP design.
Values of the EP setpoints shall be specified and availability of sufficient margin from the setpoint to the safe operation limit shall be substantiated. The margin shall be determined as the sum of the following components: calculation error plus accuracy grade of the setpoint control instrument.
In case of any refurbishment of the reactor core any changes of safe operation limits shall be substantiated.
4.2.4.3. Substantiation of the core and FA design.
Information on any works performed in order to substantiate the core and FA design shall be provided; this information shall be divided into the following groups:
neutron and physical substantiation (performed in accordance with the provisions of item 4.2.6 of this Appendix);
substantiation of heating performance reliability (performed in accordance with the provisions of item 4.2.7 of this Appendix);
strength substantiation as well as substantiation of mechanical strength and absence of any impermissible deformations and vibrations.
Information on any scientific research and development works performed in order to substantiate the core design shall be presented in accordance with the following scheme:
the list of experimental SRD works, experiments and tests performed on test stands, research reactors and operating NPPs;
description of the experimental methods;
analysis of the experiment results.
In case of any refurbishment of the reactor core substantiated scope of additional bench-scale and reactor tests shall be specified.
4.2.4.4. Functioning in case of any failures.
Analysis of failures of the RF components as well as human errors shall be provided and their impact on the reactor operability and the NPP safety at all levels of defense-in-depth shall be assessed. Analysis of consequences of common cause failures shall be provided and their impact on the reactor operability and the NPP safety shall be assessed. Information on any systems and components required to mitigate and (or) to eliminate consequences of these failures shall be provided. The list of failures subject to analysis provided in Chapter 15 of the NPP SAR shall be defined.
The final list of initiating events for design basis accidents and the final list of beyond design basis accidents shall be provided in this section with due regard for the information specified in Chapter 15 of the NPP SAR.
In case of any refurbishment of the reactor core associated with usage of new fuel type the reviewed final list of initiating events for design basis accidents and the final list of beyond design basis accidents shall be presented with due regard for peculiarities of the new fuel type that shall be analyzed in Chapter 15 of the NPP SAR.
4.2.5. CPS control rods.
4.2.5.1. Purpose and functions.
Safety class and seismic category of the CPS CR shall be specified.
Information on the regulatory requirements and safety criteria considered in the design basis shall be provided.
4.2.5.2. Design basis.
Information on the design basis for the CPS CR under normal operation conditions and in case of any abnormal operation including accidents shall be provided. The requirements for their efficiency and fast response shall be specified.
4.2.5.3. Description of the design of CPS CR.
Description of the design of CPS CR with indication of the purpose of the main components and information on belonging of the CPS CR to CPS CR groups shall be provided.
Information on the design, configuration, structure, characteristics and operation procedure for the CPS shall be presented. The number, efficiency, location, configuration of groups, working positions, movement sequence, and velocities for the CPS CR as well as the number of drives shall be substantiated.
The main design characteristics of the CPS CR as well as information on the reactor shutdown systems in case of any failures of the CPS CR shall be provided.
4.2.5.4. Materials.
Information on the materials within the scope similar to that specified in item 4.2.1.1 of this Appendix shall be provided. Information on the sources confirming operability of the materials for the CPS CR and the CPS guide channels shall be presented.
4.2.5.5. Quality assurance.
Information on the QA in the course of the development (engineering), manufacturing, acceptance, and installation of the CPS CR shall be provided. The main requirements specified in the QAP and the regulatory documents governing the requirements for quality assurance shall be listed.
Information on compliance of the testing types and results for the actuators of the core reactivity control devices with the requirements of federal rules and regulations in the field of atomic energy use governing arrangement and operation of the actuators for reactivity control devices shall be provided.
4.2.5.6. Tests and inspections.
Frequency of control and the list of controlled parameters for the CPS CR used to determine the criteria of operability loss (decrease of physical efficiency below the certain level, absence of any rod movements) shall be provided and substantiated.
Methods and conditions for the CPS CR tests, inspections, and replacement shall be specified and substantiated.
4.2.5.7. Control and monitoring.
Information on the CPS CR monitoring and control with due regard for the data specified in Chapter 7 of the NPP SAR shall be provided.
4.2.5.8. Safe operation limits and conditions, operation limits and conditions.
Safe operation limits and conditions as well as operation limits and conditions for the state of the CPS CR system shall be specified with due regard for the information provided in Chapter 16 of the NPP SAR.
4.2.5.9. Design analysis.
4.2.5.9.1. Normal functioning.
Functioning of the CPS CR under normal operation conditions and in case of any abnormal operation including design basis accidents shall be described, and information on the state of the CPS CR in these modes specifying the ways to determine and ensure their operability shall be provided.
4.2.5.9.2. Functioning in case of any failures.
Analysis of failures and damages of the CPS CR and their consequences for the NPP safety shall be provided.
Information on any measures for prevention of failures or mitigation of their consequences adopted in the design of control devices and the CPS guide channels and in the course of their operation shall be presented. Any possible failures of the equipment in the course of the CPS CR loading and unloading, in the refueling mode and in case of unsuccessful withdrawal from the cell shall be analyzed.
4.2.5.9.3 Reliability analysis.
Reliability analysis substantiating compliance of the CPS reliability indicators with the requirements of the regulations (with due regard for the information specified in Chapter 7 of the NPP SAR) shall be provided.
4.2.5.9.4. Design substantiation.
Justifying calculations and experimental substantiations performed in order to substantiate design and operability of the CPS CR shall be provided.
Information on manufacturing and physical weighing of mock-ups, manufacturing and hydraulic testing of mock-ups shall be presented.
Information on the operation results for CPS CR of similar design as well as the results of bench-scale tests and calculations shall be provided; information on any works performed in order to substantiate the CPS CR design shall be presented:
substantiation of thermal and hydraulic characteristics;
substantiation of operability (strength and reliability).
Information of each group of works shall include two parts - calculation and experimental.
The calculation part shall consist of:
the list of calculations;
applied measuring methods and programs with information on their validation;
results of the calculations and their analysis.
The experimental part shall consist of:
the list of performed SRD works;
description of the applied measuring methods;
analysis of the experimental results.
The following shall be provided:
design efficiency value of the CPS CR with the relevant poison loading, efficiency decrease, burn-up, AE, and CPS CR fluence within the specified service life;
basic thermal and hydraulic characteristics of the CPS CR; distribution of the coolant flow, temperature of the poison, AE claddings, parts of the CPS rods and shroud tubes, pressure difference in the rods and buoyancy force applied to them;
basic strength characteristics of the CPS CR and CPS sleeves determining their reliability: SSS of the CPS CR claddings and components, any changes of the AE dimensions and shape due to swelling, creep, temperature, interaction between poison and the cladding, interaction between the AE bundle and the shroud tube, interaction between the CPS CR parts and the CPS shroud tube;
the specified life limit, the specified service life and the specified shelf life of the CPS rods;
operability loss criteria for the CPS CR.
In case of any refurbishment of the reactor core associated with usage of new fuel type sufficiency of the existing reactor shutdown systems particularly the ones performing the functions of emergency protection with regard to their efficiency and fast response shall be substantiated or the design materials for the refurbished reactor shutdown systems shall be provided.
4.2.5.9.5. Design assessment.
Compliance of the design with the requirements of the regulations and the design criteria shall be assessed.
4.2.6. Neutron and physical section of the reactor design.
Information presented in this subsection shall be based on the design materials for the RF, the nuclear core, fuel assemblies, the RF systems and equipment, the SRD results as well as on the available experience.
4.2.6.1. Design basis.
Information on the requirements of the federal rules and regulations in the area of atomic energy use as well as any other requirements considered in the reactor core design shall be provided.
The main requirements for neutron and physical characteristics of the reactor core shall be specified and substantiated.
The following shall be provided:
limit values of average fuel burn-up per a fuel assembly, a fuel element and a fuel pellet;
planned duration of the fuel kernel operation in the established refueling pattern;
types of fuel used;
maximum enrichment of the fuel with 235U isotope, maximum content of fissionable plutonium isotopes in MOX (mixed uranium and plutonium oxide fuel) and REMIX (mixed regenerated uranium and plutonium fuel) fuel in the loaded fuel assemblies;
the limit of burnable poison content in the loaded fuel, the maximum number of gadolinium fuel elements in fuel assemblies;
the limit concentration of boric acid in the primary circuit coolant at the minimum controlled power level and in safety systems;
refueling conditions in the established mode - planned duration of refueling, general characteristics of the applied refueling pattern;
restrictions for distribution of energy emissions in the nuclear core including the maximum linear heat generation rate of a fuel element, the maximum relative capacity of a fuel element, permissible axial offset of energy emissions and the burnout ratio under normal operation conditions;
criteria of stability and control of energy emission distribution in the course of transient processes;
the maximum reactivity margin;
the maximum re-criticality temperature of the shut-down reactor in the end of the fuel kernel operation;
in-process adjustment - mechanical and chemical (boron);
the maximum permissible EP actuation time;
assurance of reliable and fast shutdown of the reactor in case of any pre-accident situations and accidents.
4.2.6.2. Calculation results for neutron and physical characteristics of the RF required in safety analyses particularly as input data for the thermohydraulic section of the design and thermal and mechanical calculations aimed to substantiate operability of fuel elements and fuel assemblies.
Information on the input data and models used to perform neutron and physical calculations, results of these calculations as well as analysis of the obtained results for compliance with the design criteria shall be provided.
4.2.6.2.1. Description of calculation models and input data.
The following information shall be provided:
general description of the nuclear core; information on the number of loaded fuel assemblies, the interval between fuel assemblies, the number of control devices, height of the core, thermal power of the reactor, coolant flow through the core, coolant pressure above the core, share of heat generated in the fuel elements, input temperature of the coolant at full-rated power and depending on the reactor power, coolant temperature at the minimum controlled power level and in the course of refueling, information on the coolant flow and temperature with partial quantity of the operating RCPs;
the map of the CPS CR location in the core, their segregation into groups with indication of the functional purpose, sequence and modes of their movement under normal operation conditions and in case of any abnormal operation;
the main characteristics of fuel assemblies; flat-to-flat dimensions, the number of fuel elements
(both ordinary and gadolinium fuel elements), the interval between fuel elements in the grid, the number of guide channels, the number of measuring channel, thickness and material of the shroud (if any), dimensions and material of stiffening angles (if any), the map of fuel elements, guide and measuring channels location inside the fuel assembly;
the main characteristics of fuel elements and gadolinium fuel elements; outer and inner diameter of the cladding, material of the cladding and plugs, the nominal fuel weight in a fuel element and a gadolinium fuel element including burnable poison integrated into the fuel matrix;
the main characteristics of fuel pellets in fuel elements and gadolinium fuel elements; material, outer diameter, diameter of the central aperture, height and density of a fuel pellet with and without burnable poison, characteristics of the guide channel including outer and inner diameter (particularly in the area of hydraulic brake), material and characteristics of the measuring channel including outer and inner diameter, material of the channel and configuration of the measuring device;
the main characteristics of absorbing elements within the CPS CR: the number of AEs in the CPS absorbing rods, outer and inner diameter of the cladding, material of the cladding, dimensions and material of the termination, number and type of absorbing materials, their density, total length of AEs and section lengths for each absorbing material;
the considered fuel loading patterns with indication of the FA relocation diagram in the course of refueling;
description of the FA types engaged in the formation of fuel kernels with diagrams and maps of structural elements location inside them (elevations and plans) and characteristics of the applied fuel elements and gadolinium fuel elements;
for each type of fuel assemblies - information on the number, material, weight, location of spacer and enhancer grids, information on flow resistance coefficients for these grids as well as any other elements determining pressure differential across the core;
information on the composition (volume parts of the coolant and structural materials), dimensions and location of the radial and axial reflectors provided in the design, on the applied limit conditions.
4.2.6.2.2. Description of software tools and methods.
The following information shall be provided:
information on any software methods and tools used in neutron, thermal and hydraulic and neutron and physical calculations of the reactor facility, information on their verification and validation, information on accuracy of the obtained calculation results for neutron and physical characteristics with due regard for uncertainty analysis;
brief characteristics of the neutron and physical model; the method for preparation of constants, the list of state parameters for consideration of inverse relationships, approximations in solutions of the neutron transfer equation, the number of energy groups, boundaries of the groups, parameters of the computational grid, the method for consideration of gamma-radiation energy emission, the method for consideration of spacer grids, position and movement of the CPS CR, poison burn-up in the CPS CR absorbing elements and 10B in boric acid, methods for calculation of reactivity coefficients, determination and application of adjustment factors, approaches to calculations of energy emissions per a fuel element, consideration of process and calculation errors in the analysis of the obtained results;
the main thermal and physical calculation approximations for consideration of inverse relationships, characteristics of the thermohydraulic model of the core and the thermal and physical (thermal and mechanical) model of a fuel element: data on consideration of local flow resistance values,
thermal conductivity of the gap and the fuel kernel depending on the linear load and burn-up.
4.2.6.2.3. Neutron and physical characteristics of fuel assemblies.
Basic parameters of cell calculations shall be specified in preparation of constants: temperature of fuel and coolant, coolant density, boric acid concentration, specific energy release.
Information on dependencies K between the content of uranium and plutonium isotopes and fuel burn-up (energy yield) calculated with the above-mentioned parameters shall be specified in tabular format for all types of fuel assemblies used in the kernels; information on dependence of the ratio between total content of burnable gadolinium isotopes 155Gd and 157Gd and their initial content from the fuel burn-up value shall be specified in tabular format for fuel assemblies with gadolinium fuel elements.
4.2.6.2.4. Main neutron and physical characteristics of the nuclear core.
The following shall be provided:
characteristics of fuel assemblies loaded into the reactor and removed from the reactor; the number of loaded fresh fuel assemblies with breakdown by types and indication of average enrichment and (or) content of fissionable plutonium isotopes; the number of burnable poisons; the number of removed fuel assemblies with breakdown by types; average burn-up for each type of removed fuel assemblies and for all unloaded fuel; maximum burn-up per a fuel assembly, a fuel element and a fuel pellet (local burn-up) with indication of the maximum burn-up points;
neutron and physical characteristics of fuel kernels: complete duration of the fuel campaign; duration of operation on power reactivity effect; reactivity margin as of the campaign beginning; critical concentration of boric acid in the beginning of the campaign at the minimum controlled power level, at full-rated power with and without Xe poisoning; maximum coefficients of energy emission irregularity at full-rated power - Kq, Kr (for an ordinary fuel element and a gadolinium fuel element); average linear heat generation rate of fuel elements; maximum linear heat generation rate for ordinary and gadolinium fuel elements during the campaign with due regard for engineering safety margins; reactivity coefficients by the coolant temperature and density, by fuel temperature, by boric acid concentration in the beginning and in the end of the campaign at full-rated power and at the MCL; power reactivity coefficient in the beginning and in the end of the campaign at full-rated power; maximum efficiency of a single AR from a working group at full-rated power; efficiency of control groups and EP groups (without overlapping) in the beginning and in the end of the campaign at full-rated power; EP efficiency in the absence of the most efficient CPS CR in the beginning and in the end of the campaign at the MCL and at full-rated power; re-criticality temperature in the end of the campaign in Xe-poisoned state; boric acid concentration providing the required sub-criticality level in the course of refueling (2%) and the reactor start-up after lifting of the CPS CR performing emergency protection (1%) in the state with maximum reactivity margin; efficiency of the liquid reactor shutdown system in case of the boric acid concentration change from the critical value to the outage one at the minimum controlled power level in the beginning and in the end of the fuel campaign; effective fraction of delayed neutrons and lifetime of prompt neutrons in the beginning and in the end of the fuel campaign; maximum full neutron flux in the internal area of the core, in peripheral fuel assemblies and in the areas adjacent to the top and bottom end reflectors in the beginning and in the end of the fuel campaign with the rated power of the RF; maximum values of fast fluence (En > 0.1 MeV) and the damaging dose (dpa) on the claddings of fuel elements and gadolinium fuel elements, on the guide channel, the central tube and the shroud (angle) of the fuel assembly (if any) with indication of the maximum points in the end of the fuel kernel operation.
Changes of the main parameters in the course of the fuel kernel burn-up shall be demonstrated in tabular and graphical format: the reactor power, flow and input temperature of the coolant, position of
the working group, critical concentration of boric acid, axial energy emission offset, Kq, Kr, maximum linear heat generation rate of a fuel element with indication of the maximum points.
Information on the mass of uranium, plutonium and minor actinide isotopes in the unloaded fuel under the established refueling mode shall be provided in tabular format.
4.2.6.2.5. Power distribution in the nuclear core:
The following information shall be provided:
the map of relative FA capacity distribution in the beginning of the campaign at the minimum controlled power level;
maps of distributions of relative FA capacity. maximum relative capacities of fuel elements in the FA, maximum relative linear heat generation rate values of fuel elements in the FA;
maps of average fuel burn-up in fuel assemblies in the beginning, in the middle and in the end of the boric campaign as well as in the end of the operation on power reactivity effect;
distribution of relative capacities and burn-ups of fuel elements in the fuel assembly with the maximum Kr value in the beginning, in the middle and in the end of the boric campaign;
radial energy emission distributions in a fuel element and a gadolinium fuel element and their change in the course of burn-up used in thermohydraulic, thermal and mechanical calculations;
distribution of relative capacity of the design layers along the core height (axial energy emission profile) in the beginning of the campaign at the minimum controlled power level;
distributions of relative capacity of the design layers along the core height (axial profiles) in the beginning, in the middle and in the end of the boric campaign at full-rated power as well as in the end of the fuel kernel operation on power reactivity effect;
elevational distributions (axial profiles) of average fuel burn-up for the design layers in the nuclear core in the beginning and in the end of the campaign;
elevational distributions (axial profiles) of average coolant temperature and density values, fuel temperature in the design layers of the nuclear core in the beginning, in the middle and in the end of the boric campaign as well as in the end of the operation on power reactivity effect;
permissible values of linear heat generation rates of ordinary and gadolinium fuel elements depending on the distance from the core bottom in the course of the RF operation with all RCPSs at full-rated power and in the course of operation with lesser number of RCPSs;
values of engineering safety margins for coefficients of irregularity, characteristics of energy emission and burn-up distribution in the core considered in the safety analysis;
distributions of linear heat generation rates and (or) relative linear heat generation rate along the fuel element (gadolinium fuel element) height assumed in the safety analyses in Chapter 15 of the NPP SAR.
4.2.6.2.6. Nuclear core operation modes in the course of the campaign.
Brief information on the nuclear core operation modes in the course of the campaign and the results of calculation modelling for the following processes shall be provided:
the reactor start-up after refueling, after short-term and long-term (maximum Xe poisoning) shutdown in the course of the fuel kernel operation;
operation at reduced power levels;
operation with reduced number of RCPSs;
operation in the daily load following mode and any other load-following modes provided in the NPP design;
operation on power reactivity effect and variable parameters;
scheduled shutdown of the reactor;
4.2.6.2.7. Power (energy emission) distribution control in the nuclear core.
Brief description of the applied in-core neutron detectors, their characteristics and location for the reactor power measurement shall be provided. Information on the software tools used to reconstruct the power density field in the core based on the readings of detectors, information on their verification and validation, assessment of error in reconstruction of the FA capacity, linear heat generation rate of fuel elements and any other parameters used to control power distribution shall be provided.
4.2.6.2.8. Reactivity effects and coefficients.
4.2.6.2.8.1. Information on the calculation methods and values of reactivity effects related to the following processes shall be provided:
the reactor warm-up from the cold state to the minimum controlled power level;
power increase up to the rated level;
equilibrium xenon poisoning at the rated power level;
samarium poisoning,
as well as on the reactivity value compensated with boric acid solution in the coolant in the beginning, in the middle and in the end of the campaign.
Maximum reactivity margins for the campaign shall be specified:
in the cold state;
at the minimum controlled power level;
at the rated power level without xenon and samarium, with equilibrium samarium, with equilibrium samarium and xenon.
4.2.6.2.8.2. Reactivity coefficients in the beginning and in the end of the campaign in the course of the reactor operation at the rated power level (with and without Xe poisoning) and at the minimum controlled power level shall be specified: reactivity coefficient by fuel temperature, coolant temperature, coolant density, boric acid concentration in the coolant, coolant pressure in the primary circuit, the reactor power. Exact definitions or assumptions used in calculation of the above-mentioned reactivity coefficients shall be established namely: by effective Doppler temperature, by the parameters assumed to be constant in determination of the power reactivity coefficient, by spatially-heterogeneous variations of the parameters.
4.2.6.2.8.3. Information on the dependence of reactivity coefficients by fuel temperature, coolant density and power from the reactor power shall be provided for the beginning and the end of the fuel campaign.
4.2.6.2.8.4. Limit values of reactivity coefficients intended for application in safety analyses and covering any possible design values of reactivity coefficients in various operational states with due regard for design margins, calculation errors, deviations of power, coolant temperature and boric acid concentration shall be specified.
4.2.6.2.9. Reactivity balance and control efficiency.
4.2.6.2.9.1. Reactivity balance analysis shall be provided. Compliance of the reactivity characteristics with the requirements of federal rules and regulations in the field of atomic energy use shall be confirmed. Reactivity balance shall be arranged with due regard for any possible errors in determination of reactivity effects and efficiency of the control devices. The core reactivity balance shall be determined for the beginning and the end of the campaign as well as for the least favorable (from the viewpoint of compliance with the regulatory requirements) intermediate moments of the fuel kernel burn-up. The following factors affecting the reactivity and depending on various operational states shall be taken into account:
purpose of the CPS CR groups, their expected and minimal permissible efficiency;
efficiency of burnable poison;
concentration and efficiency of boric solution;
disturbances of the moderator and fuel temperature as well as any possible void disturbances;
fuel burn-up (fission products);
xenon and samarium poisoning;
permissible depths of the CPS CR insertion into the core and their acceptable mismatch.
The minimum required sub-criticality margin for the promptly shut-down reactor shall be substantiated for different moments of the campaign with due regard for uncertainties of such margin and experimental verification in operating reactors.
4.2.6.2.9.2. The following information on the reactivity control methods in the course of adjustment under normal operation conditions and in case of any abnormal operation shall be presented:
use of liquid poison;
movements of the CPS control devices including rods affecting the axial energy emission profile;
any possible changes of the coolant flow or temperature.
Impact on volumetric energy emission distributions (in case of xenon redistribution and xenon oscillations) as well as impact (if any) on fuel-burn-up through spectral regulation and (or) modification of the coolant parameters shall be described.
4.2.6.2.9.3. Information on the location of the CPS control devices planned for use during the fuel cycle shall be provided. Detailed information on their segregation into groups or sets, sequence and degrees of their withdrawal from the nuclear core, any substantiated restrictions for their positions depending on the power level, the campaign moment or any other parameters shall be specified. Expected positions of rods or groups in critical states at the MCL and the rated power level both in the beginning and in the end of the campaign; permissible and recommended position of the CPS CR working group in steady states at various power levels shall be described.
4.2.6.2.9.4. The following design efficiency values for individual CPS control devices and CPS CR
groups shall be specified for the beginning, the middle and the end of the fuel campaign at various power levels with due regard for the calculation error and permissible deviation of the initial and final position of the CPS control devices:
maximum efficiency values for individual CPS control devices in case of accidents with ejection or falling of absorbing rods;
maximum efficiency of the CPS CR working group as the function of insertion into the core in case of any accident with uncontrolled withdrawal of the working group; reactivity change in case of design withdrawal and insertion of the CPS CR regulating groups;
maximum positive reactivity insertion velocity upon withdrawal of the regulating groups; efficiency values for individual groups of the CPS control devices;
values of full efficiency of the emergency protection and the EP efficiency without the most efficient CPS CR rod;
efficiency and reactivity change upon insertion of the CPS control devices performing emergency protection into the core.
4.2.6.2.9.5. Dependencies between coefficients of irregularity and maximum linear energy emission rate in the core (Kq, Kr and Ql) and insertion of the working group within the permissible range of movement at the full-rated power shall be specified for the beginning, the middle (or any other intermediate moment of the campaign characterized by the maximum coefficients of energy emission irregularity and linear heat generation rate of fuel elements) and the end of the boric campaign.
4.2.6.2.10. Analysis of the reactor sub-criticality in the course of refueling.
The following information shall be provided:
design sub-criticality estimates in the course of refueling;
description of the reactor sub-critical state monitoring;
neutron background of the nuclear core depending from the isotope composition of fuel and its burn-up degree;
requirements for the refueling control and compliance with these requirements in the NPP design under consideration.
4.2.6.2.11. Stability of the core power distribution.
Information on xenon stability of the nuclear core both in axial and radial directions shall be provided. General description of the means for xenon instability detection and control actions for its suppression shall be provided, the algorithm and the procedure for the NPP personnel's actions for suppression of xenon oscillations shall be substantiated. Sufficiency of the measures for detection and suppression of xenon oscillations in order to prevent any exceedance of the design limits shall be substantiated.
4.2.6.2.12. Afterheat release and radiation characteristics of fuel.
The curves of the residual nuclear core power caused by radioactivity of fission products depending on the time after the reactor transition into sub-critical state shall be provided for the beginning, the middle and the end of the boric campaign as well as for the end of operation on power reactivity effect. Information on the data and approximations used as the basis for the presented curves shall be provided. Information on radiation characteristics of the fuel used to assess radiological
consequences of design basis and beyond design basis accidents in Chapter 15 of the NPP SAR shall be specified.
4.2.6.3. Calculations of fluence, radiation damage and radiation energy emissions on the reactor internals and the reactor pressure vessel.
Information on the calculation model, methods, software tools and constants used to calculate fast fluence, radiation damage and radiation energy emissions on the reactor internals and the reactor pressure vessel as well as information on the ST validation and verification shall be provided.
Distributions of the fluence accumulation rate, neutron fluence value, the damaging dose in displacements per atom at the nuclear core boundaries on the reactor internals, the inner surface of the reactor pressure vessel and within the reactor pressure vessel during the design service life shall be provided.
Information and substantiation for the scope of neutron fluence monitoring on the reactor pressure vessel in the course of operation shall be provided.
Calculation results for radiation energy emission in the reactor internals and pressure vessel in the course of the reactor operation at full-rated power shall be presented.
4.2.6.4. Calculations of neutron flux in the detectors of the neutron flux monitoring equipment (NFME).
The allocation scheme and brief characteristics of the detectors (ionizing chambers) shall be provided. The calculation model, method, software tools and constants used to calculate neutron flux in the ionizing chambers shall be described. Information on validation of the applied software tools shall be presented.
Total neutron flux values at the location points of ionizing chambers at the minimum controlled power level and the full-rated power in the beginning, in the middle and in the end of the boric campaign as well as in the end of operation on power reactivity effect shall be specified.
4.2.6.5. Reactivity assessment for a beyond design basis accident with the core melting.
The maximum possible value of Keff for the least favorable combination of composition, configuration and external conditions for the corium in the reactor pressure vessel; in case of any melt-through of the vessel bottom as well as in the core catcher (if any) shall be provided and substantiated.
4.2.6.6. Limits of neutron and physical characteristics for the purpose of safety analysis.
The summary table of the limit values of neutron and physical characteristics the least favorable for the RF safety as adopted in safety analyses in Chapter 15 of the NPP SAR shall be provided.
4.2.7. Thermohydraulic section of the project.
4.2.7.1. Design restrictions.
The following information on the design restrictions affecting thermal and hydraulic characteristics of the RF shall be provided:
maximum temperature of fuel element claddings;
maximum temperature of the coolant;
coolant temperature change rate;
maximum linear load of fuel elements;
maximum coolant flow velocity in the nuclear core;
net positive suction head and other operational restrictions for RCPs;
minimum critical heat flux ratio;
pressure drop in fuel assemblies and the CPS ARs;
hydraulic stability criteria for the coolant flow;
temperature margin to the fuel melting under the rated conditions.
4.2.7.2. Thermohydraulic calculation of the nuclear core.
The following information shall be provided:
a) distribution of the coolant flow and linear energy emission; in this case the following data shall be presented:
distribution of the coolant flow and enthalpy through the reactor and through the nuclear core;
average and maximum values of linear energy emission;
coolant temperature at the outlet of the fuel assembly, the core and the reactor with due regard for distribution of coolant flow rates in fuel assemblies;
distribution of coolant flow rates in fuel assemblies;
maximum temperature of fuel element claddings;
burnout ratio;
pressure differentials in the nuclear core and flow resistance values; in this case the coolant flow arrangement diagram at the reactor inlet, pressure differential values in the core and the corresponding distributions of flow resistance in the core elements shall be provided;
b) methods and calculation programs, in this case the following data shall be presented:
information on any methods and software tools used in thermohydraulic calculations of the nuclear core, information on their verification or reliability substantiation for the obtained results;
information on the accuracy of the obtained results of thermohydraulic calculations with due regard for uncertainty analysis;
information on validation of software tools.
4.2.7.3. Thermohydraulic calculations of the RF.
Information on thermohydraulic calculations of the primary circuit and the emergency heat removal system shall be provided.
The following data shall be specified in the description:
information on the layout of equipment and pipelines of the primary RF circuit;
thermohydraulic diagram of the RF:
the number of coolant circulation circuits and their purpose;
type of the coolant movement activator (forced circulation, natural circulation);
the list of equipment and pipelines in each circulation circuit, design values of coolant flow for each circuit components and pressure differentials at the corresponding flow rates;
coolant circulation diagrams for each circuit, elevational allocation of the loop elements (equipment, pipelines) for different circuits, their geometric characteristics;
coolant levels and pressure values in the RF primary circuit components;
design RF operation modes under normal operation conditions:
the list of design normal operation modes (with the reference to the relevant subsection of Chapter 4 of the NPP SAR);
thermal and hydraulic peculiarities of each design normal operation mode;
coolant parameters and their change rates in various design modes;
coolant temperature distribution in the design modes;
methods and calculation programs:
information on any methods and calculation programs used in thermohydraulic calculations for the RF;
information on their validation;
information on the accuracy of the obtained results of thermohydraulic calculations with due regard for uncertainty analysis.
4.2.7.4. Tests and inspections.
Programs and methods for tests and inspections to be used in order to confirm design thermal and hydraulic characteristics of the nuclear core and the RF circulation circuits shall be described.
4.2.8. The CPS actuating mechanisms.
4.2.8.1. Purpose and design basis.
The following information shall be provided:
information on the configuration, purpose and functions of the actuating mechanisms;
safety class and seismic category of the actuating mechanisms;
design limits for the actuating mechanisms under normal operation conditions, in case of any abnormal operation and design basis accidents;
permissible limits for the main mechanical and strength characteristics and permissible values of reliability parameters for the actuating mechanisms.
4.2.8.2. Description of the design.
The following information shall be provided:
information on design of the actuating mechanisms with indication of individual devices (components) performing independent functions;
information on the controls, fasteners and sealing devices;
drawings and diagrams demonstrating the design, kinematic schemes of action and location of the actuating mechanisms;
basic technical characteristics of the actuating mechanisms;
the list of systems and equipment affecting the functioning of the actuating mechanisms.
4.2.8.3. Materials.
Information on the grades and properties of steels and materials used in the actuating mechanisms as well as substantiation of their operability within the required time period in water medium under the design temperatures and radiation exposures corresponding to normal operation of the RF and abnormal operation including design basis accidents shall be specified.
4.2.8.4. Quality assurance.
Information on the quality assurance program in the course of the development (design), manufacturing, acceptance and installation of the CPS actuating mechanisms shall be provided. The main requirements specified in the QAP and the regulatory documents governing the requirements for quality assurance shall be listed.
4.2.8.5. Control, monitoring and testing.
The following information shall be provided:
principles for control of the actuating mechanisms and monitoring of their state;
characteristics of the control signals for the actuating mechanisms;
analysis of any potential controlling actions of automation devices and operators on the actuating mechanisms;
methods, means, scope and frequency of state control and testing for the actuating mechanisms in order to ensure their operability in the course of operation;
information on the commissioning works for the actuating mechanisms, the list of testing programs demonstrating sufficiency of pre-operational tests for the actuating mechanisms in order to substantiate safety of the RF operation and the list of arrangements aimed to prevent accidents in the course of testing.
It shall be specified that the actuating mechanisms have intermediate position indicators for their control devices, end position detectors, and limit switches actuated directly by the CD.
It shall be substantiated that the design of the reactor and the CPS actuating mechanisms ensures disengaged position of the CPS control devices in case of the upper unit removal and the diagnostic means provided in the design record the disengaged position.
4.2.8.6. The CPS actuating mechanisms.
4.2.8.6.1. Normal functioning.
The following information shall be provided:
description of functioning of the actuating mechanisms under normal RF operation conditions with due regard for transient modes during scheduled start-ups, power changes and shutdowns;
requirements for reliability of the safety-related systems and components interacting with the actuating mechanisms.
4.2.8.6.2. Functioning in case of any failures.
The following shall be provided:
analysis of consequences in case of any failures of the actuating mechanisms with due regard for any failures caused by human errors;
description and sufficiency substantiation for the measures aimed to prevent common cause failures of the actuating mechanisms with due regard for internal and external impacts and failures of systems and equipment;
analysis of failure consequences and characteristics of any changes in the main safety-related RF parameters caused by failures;
the list of failures of the actuating mechanisms representing initiating events for abnormal operation including design basis accidents that require additional analysis in the relevant section of the RF safety analysis report.
4.2.8.6.3. Design substantiation.
It shall be substantiated that the actuating mechanisms comply with the requirements of federal rules and regulations in the field of atomic energy use as well as the design criteria, and that they were tried out in operation of VVER reactors or tested under the conditions similar to the required ones and justified by SRD works.
4.2.9. Reactor vessel.
Information on the purpose and functions of the reactor vessel shall be provided. Detailed description and substantiation of the reactor vessel operability shall be given in the section of Chapter 5 of the NPP SAR performed in accordance with item 5.3 of this Appendix.
4.2.10 Conclusions.
Subsequent to the analysis results for the reactor and its components conclusion on their compliance with the requirements of the regulations, design principles and criteria shall be made.
4.2.11. The list of applied documentation.
The list of the design and engineering documentation used to develop the design of the reactor and its components shall be given.
V. Requirements for the content of Chapter 5 "Primary circuit and interfaced systems"
Information on the primary circuit components and any systems related to the primary circuit
shall be provided in Chapter 5 of the NPP SAR.
The primary circuit components and any interfaced systems shall be described in accordance with the standard system description framework given in Appendix 4 hereto.
The following systems and components shall be considered:
the reactor coolant circuit (the primary circuit) intended for the coolant circulation through the core which usually includes:
the reactor vessel with the upper unit cover and the main joint sealing parts;
RCP;
SG;
reactor coolant pipeline (pipelines connecting the above-mentioned components);
PRZR;
systems (or parts of systems) related to the RCC within the primary circuit pressure boundary:
systems supporting normal functioning of the RCC;
SS;
auxiliary systems;
primary circuit valves;
fastening elements:
as the configuration of the primary circuit components and the systems related to this circuit may differ for various RF types the complete set of these systems and components shall be defined by the NPP SAR developers depending on the NPP design peculiarities;
supports, shock absorbers, travel stops and other spacing elements between the primary circuit components and the building structures shall be considered in the description of the primary circuit components.
5.1. General description.
General description of the primary circuit and interfaced systems shall be provided.
5.1.1. The primary circuit and interfaced systems.
Brief information on the design and performed justifying calculations and experimental substantiations for the primary circuit systems and components shall be provided.
All components included into the primary circuit in compliance with the NPP design shall be listed.
The complete list of systems related to the primary circuit with references to the sections of other NPP SAR chapters containing their detailed descriptions shall be given.
Description and purpose of the primary circuit, its main components, and interfaced systems shall be provided. Components performing independent functions as well as safety functions of each system and component shall be defined. Tables of the design and working (operational) characteristics shall be
presented.
Information on the design principles and criteria taken as the basis for the design of the primary circuit systems and components shall be provided.
Information on performance of the main primary circuit function - assurance of heat removal from the core by the sufficient amount of coolant with the adequate quality under normal operation conditions and in case of any abnormal operation including design basis accidents in compliance with the design limits as well as fuel damage limits shall be specified.
References to other sections of the NPP SAR containing more detailed requirements for individual systems and components of the primary circuit shall be given.
It shall be specified that the NPP design provides for monitoring of the coolant temperature, pressure and chemical composition in the primary circuit under normal operation conditions and in case of any abnormal operation including design basis accidents.
It shall be specified that all systems and components of the primary circuit were designed with due regard for the capability to withstand the ambient conditions (pressure, temperature, humidity, radiation) appearing in the course of normal operation and in case of any abnormal operation including design basis accidents as well as in post-accident modes and the containment testing modes within the entire service life.
Information on consideration of the ambient conditions occurring in case of a beyond design basis accident in the design of the primary circuit systems and components shall be provided.
All components for accommodation of seismic loads installed on the pipelines and equipment shall be described, and it shall be demonstrated that any failures of the components with a lower seismic category will not result in failures or breakage of any components with a higher seismic category.
It shall be specified that the NPP design provides for submittal of the following information to the operator:
any abnormal operation in the primary circuit;
reaching of the operation limits and (or) safe operation limits for the operating parameters.
The following information on compliance with the federal rules and regulations in the field of atomic energy use establishing the requirements for arrangement and safe operation of equipment and pipelines of nuclear power installations shall be provided:
the possibility for the radioactive coolant drainage, decontamination of surfaces and disposal of decontamination solutions; information on absence (presence) of any areas where removal of contamination products together with detergents and decontamination solutions is impossible shall be provided;
the possibility for removal of air in the course of filling with the medium as well as the working medium and condensate formed in the course of the circuit warm-up or cooldown;
compliance with the requirements for heat insulation of the external surfaces of the equipment and pipelines;
access to the primary circuit components for technical examination, maintenance and repair works; information on any places inaccessible for technical examination as well as information on any methods and means for their technical examination provided in the design documentation shall be
specified.
The following information shall be provided: information on the performed calculations, the list of experimental works and analysis of the experimental results.
5.1.2. Process flow diagram.
The process flow diagram of the primary circuit with indication of the primary circuit boundaries and all main components, the operating pressure, temperatures, flow rates and coolant volume in the steady plant operation mode at full-rated power shall be provided. All systems connected to the primary circuit and the way of their isolation from the primary circuit shall be indicated on the flow diagram.
The routes of pipelines within the reactor building shall be presented in isometric view.
5.1.3. Instrumentation and control diagram.
The instrumentation and control diagram of the primary circuit and interfaced non-isolable systems within the pressure boundary of the primary circuit shall be provided. Measurement point labelling adopted in the design shall be specified. Information on any instrumentation devices for measurement of pressure, temperature, flow, level, chemical composition of water and gas, concentration of the liquid poison solution as well as for control of movements and leak-tightness with indication of the instrumentation accuracy grade shall be provided. Information on redundancy of sensors and communication channels and the place for display of the information received from the relevant instrumentation and control devices shall be presented.
5.1.4. General assembly drawings.
General assembly drawings with indication of the equipment elevations and the basic primary circuit dimensions in relation to the supports and the surrounding concrete structures showing the possibility for maintenance and technical examination as well as compliance with the requirements for arrangement of the conditions for natural circulation established in the federal rules and regulations in the field of atomic energy use governing nuclear safety of the NPP RF shall be provided. Information on the biological protection provided in the NPP design shall be specified on the drawings.
5.1.5. Compliance with the requirements of federal rules and regulations in the field of atomic energy use.
The table containing information on compliance of the primary circuit components and interfaced systems with the requirements of federal rules and regulations in the field of atomic energy use shall be provided.
5.2. Integrity (strength and tightness) of the primary circuit pressure boundaries.
Any measures provided in the NPP design in order to ensure strength and leak-tightness of the primary circuit equipment and pipelines shall be substantiated.
Detailed confirmation of compliance with the requirements for the reactor coolant circuit stated in the federal rules and regulations in the field of atomic energy use shall be presented, and it shall be specified that all equipment and pipelines of the primary circuit withstand static and dynamic loads and thermal impacts caused by any considered initiating events without any breakage within the entire NPP service life.
5.2.1. Design limits for pressure and temperature.
The adopted design limits for pressure and temperature in normal operation modes and in case of
any abnormal operation and accidents shall be substantiated with due regard for hydraulic testing modes in the course of normal operation and during the commissioning works.
Assurance of the primary circuit integrity in the most severe modes shall be substantiated or references to the relevant sections of the NPP SAR shall be given.
Pressure and temperature limits shall be specified for the following conditions:
preliminary factory hydraulic tests of the primary circuit components;
operational (in the course of operation and the NPP power unit commissioning) tests for leak-tightness and strength;
normal operation with due regard for the warming and cooldown modes.
Design data on the temperature and pressure limits shall be provided. References to the testing programs used to perform hydraulic tests of the primary circuit components shall be given.
Subsequent to completion of the NPP power unit commissioning the results of leak-tightness and strength tests for the primary circuit as well as temperature and pressure limits based on the obtained characteristics shall be provided. Information on any documents (certificates) approved in accordance with the established procedure and containing the above-mentioned results shall be given.
Information on the main stages of sealing and unsealing of the main reactor vessel joint (with due regard for the data specified in Chapter 4 of the NPP SAR) and other pressurized detachable joints with indication of the measures to ensure strength and leak-tightness of the joints shall be presented.
5.2.2. Primary circuit overpressure protection.
All arrangements and methods provided in the NPP design for protection of the primary circuit systems against overpressure exceeding the design limits under normal operation conditions, in case of any abnormal operation and accidents as well as during hydraulic testing shall be listed. The list of components performing the overpressure protection functions in the primary circuit shall be provided.
Reference shall be given to the relevant section of other chapters of the NPP SAR where descriptions and substantiations for the primary circuit pressure maintenance systems and the primary circuit overpressure protection systems are presented.
Minimum pressure setpoints for safety devices of the primary circuit, their number and throughput capacity shall be specified and substantiated with reference to the relevant sections of the NPP SAR.
5.2.3. Protection of the primary circuit against pipeline ruptures and their consequences.
Information on protection of the primary circuit against pipeline ruptures and their consequences shall be provided with due regard for the data specified in Chapter 3 of the NPP SAR.
The section shall contain information on the ways to minimize probability of a pipeline rupture and equipment breakage with parts tearing off.
5.2.3.1. Pipeline breakage criteria.
Information on the points of potential pipeline ruptures (connections to the equipment, areas with maximum stresses) as well as the areas with potential risk of any damage to the adjacent safety-related equipment shall be specified.
For low-temperature conditions the design data shall be provided to confirm that the pressure in the primary circuit components at low temperatures (below the operating temperature) is limited to the values eliminating brittle failure or the pressure corresponds to the level of stresses permissible for this temperature range.
5.2.3.2. Analysis of pipeline rupture consequences.
Results of the analysis of consequences in case of pipeline breakages shall be provided with regard to the following impacts on the adjacent equipment:
temperature;
pressure impact;
loads on the adjacent equipment and pipelines from jets caused by water and steam releases;
humidity and radiation impact;
reactive loads causing vibration and whipping in the damaged pipes;
damages caused by missiles;
flooding of safety-related components.
In case the "leak before break" concept is applied it shall be specified what pipelines it is applied to, and information on the document justifying its application shall be also provided.
5.2.3.3. Protection against the consequences of pipeline breakages.
Information on the methods used in the NPP design for physical separation of the pipelines and limitation of their movements shall be presented.
The following shall be substantiated:
that rupture of any pipeline does not result in rupture of any other pipeline required for mitigation of the accident consequences;
that rupture of any pipeline not belonging to the primary circuit does not result in any accidents with loss of coolant;
that rupture of any primary circuit pipeline does not result in the containment destruction;
that coolant discharge and dynamic impacts do not prevent any works in the control rooms and do not impair any systems required to eliminate the accident consequences.
5.2.4. Primary circuit materials.
Information confirming that the materials, manufacturing and control methods for the primary circuit pressure zone comply with the requirements of federal rules and regulations in the field of atomic energy use shall be provided.
5.2.4.1. Standards and technical specifications.
The list of standards and technical specifications for all materials used to manufacture the primary circuit components and fasteners as well as for welding and facing materials shall be provided.
Information on the ways to consider the material properties listed below and significantly
affecting the pressure boundary integrity assurance in selection of the primary circuit materials shall be specified:
chemical compatibility with the coolant;
compatibility with the materials of heat insulation, supports, coatings of sealing units and also any other materials contacting with the primary circuit components;
cyclic strength, creep-rupture strength and yield;
corrosion (with due regard for stress corrosion), corrosive and cyclic and erosion characteristics;
radiation damage (for steels subjected to neutron exposure);
crack resistance;
brittle failure resistance;
fabricability;
activation under irradiation;
behavior in case of any abnormal operation including accidents.
Information on control of chemical elements (cobalt content in nickel-containing steels; copper, nickel and phosphorus in reactor vessel steel; carbon, sulfur, phosphorus and silicium in carbon steels) with adverse effect on operational characteristics of the materials as well as any measures to eliminate these chemical elements (impurities) shall be provided.
5.2.4.2. Compatibility of structural materials with the primary circuit coolant.
The following information related to compatibility of the primary circuit coolant with structural materials and external insulation of the pressure zone shall be provided:
chemical composition of the primary circuit coolant with reference to the relevant design documentation and regulatory documents defining the requirements for water chemistry regime of the primary circuit;
chemical composition changes in various modes if any additives are used; maximum permissible content of chlorides, fluoride compounds, oxygen, hydrogen and corrosion products;
compatibility of structural materials with the primary circuit coolant;
the list of structural materials contacting with the primary circuit coolant and description of the material compatibility with the coolant, impurities and radiolysis products that can contact with it; in case any non-metal materials come in contact with the primary circuit coolant compatibility of these materials with the coolant shall be described;
compatibility of structural materials with external heat insulation of the primary circuit.
The list of structural materials of the primary circuit having heat insulation and information on their compatibility with external heat insulation shall be provided. Their compatibility with external heat insulation in case of any coolant leakage shall be demonstrated. Information on non-metal heat insulation of austenitic stainless steel shall be provided in order to demonstrate whether concentration of chlorides, fluoride compounds, sodium and silicates in the heat insulation will be within the acceptable limits; these limits shall be substantiated.
5.2.4.3. Manufacturing and processing of carbon steels.
Information on manufacturing and processing of carbon and low-alloyed steels shall be specified.
The following information shall be provided:
peculiarities of the production process for semi-finished products and items;
description of non-destructive control operations for all primary circuit pressure zone components in order to confirm their compliance with the requirements of federal rules and regulations in the field of atomic energy use governing arrangement and safe operation of equipment and pipelines of nuclear power installations and the requirements of federal rules and regulations in the field of atomic energy use regulating control of weld joints and hard-facing for equipment and pipelines of nuclear power installations shall be provided; reference to the quality control program shall be given.
5.2.4.4. Manufacturing and processing of austenitic stainless steels.
The following information on manufacturing and processing of austenitic stainless steels used in the primary circuit components shall be specified:
peculiarities of the production process (forging, welding, heat treatment) preventing formation of cracks due to stress corrosion as well as ferritic phase limitations; in this case control methods applied in the course of manufacturing and eliminating stress corrosion shall be specified;
control of production processes in order to reduce any contacts with media capable of causing stress corrosion; in this case measures for protection of the component surfaces against any contamination and damage promoting stress-corrosion cracking (from the manufacturing stage to the installation completion) shall be specified;
characteristics and mechanical properties of cold-deformed austenitic stainless steels for the primary circuit components and the permissible deformation degree;
measures to prevent hot cracking in the course of welding and assembly; in this case requirements for the welding materials shall be specified, and compliance of the welding and control techniques with the federal rules and regulations in the field of atomic energy use regulating welding and hard-facing for equipment and pipelines of nuclear power installations and federal rules and regulations in the field of atomic energy use governing control of weld joints and hard-facing for equipment and pipelines of nuclear power installations shall be demonstrated;
description of non-destructive control operations for all primary circuit pressure zone components in order to confirm their compliance with the requirements of federal rules and regulations in the field of atomic energy use establishing the requirements for arrangement and safe operation of equipment and pipelines of nuclear power installations and the requirements of federal rules and regulations in the field of atomic energy use regulating control of weld joints and hard-facing for equipment and pipelines of nuclear power installations; reference to the quality control program shall be given.
5.2.5. Metal control and the primary circuit testing in the course of operation.
The following information on in-service inspection and testing programs for the primary circuit components referred to groups A and B in accordance with the classification established in the federal rules and regulations in the field of atomic energy use specifying the requirements for arrangement and safe operation of equipment and pipelines for nuclear power installations shall be provided:
names and details of the programs;
boundaries of the systems subject to control;
information on control of supports and fasteners;
location of systems and components taking into account access for control purposes;
control methods and procedures ensuring compliance with the requirements established in the federal rules and regulations in the field of atomic energy use governing arrangement and safe operation of equipment and pipelines for nuclear power installations;
control frequency;
requirements of the in-service inspection program;
methods to assess the control results;
frequency and procedure for hydraulic testing (for strength and leak-tightness): compliance with the requirements of federal rules and regulations in the field of atomic energy use governing arrangement and safe operation of equipment and pipelines for nuclear power installations shall be substantiated.
Information on any peculiarities of control and testing for individual primary circuit components shall be provided in this section, and references to the relevant NPP design documents shall be given.
5.2.6. Detection of leakages through the primary circuit pressure boundaries.
The leakage detection system shall be described in accordance with the standard system description framework given in Appendix 4 hereto.
The permissible leakage value specified in the NPP design for the primary circuit coolant as well as accuracy of the leakage location and intensity determination substantiated in the design shall be provided.
The applied leakage detection methods, sensibility, response time and also reliability of the instruments and equipment shall be described; the minimum leakage value that can be detected through the use of the applied methods shall be specified.
Besides information on the systems (methods) used for alarm purposes and representing indirect indicators of leakages shall be provided.
Information on the combination of methods (systems) applied in the NPP design in order to detect location of any leakage and the accuracy of such detection shall be provided.
The program for processing of sensor signals enabling the operator to obtain reliable information on location and intensity of any leakages shall be described.
Information on the testing procedures for leakage detection systems shall be provided.
Information on any automated coolant radioactivity monitoring and control of RS discharges and releases as well as monitoring of the radiation situation in the NPP rooms provided in the NPP design shall be specified with references to other chapters.
5.2.7. Interface with the secondary circuit.
The following information shall be presented in the form of a table:
amount of coolant flowing to the secondary circuit in case of the SG tube breakage;
time required for equalization of pressure between the emergency SG and the primary circuit;
minimum volume of water and maximum volume of steam in the SG under normal operation conditions.
Criteria defining permissible primary-to-secondary leakages under normal operation conditions and criteria defining inoperability of the primary circuit as a physical barrier shall be specified. The selected primary circuit pipeline rupture diameter considered in the analysis of design basis accidents shall be substantiated.
Minimum pressure setpoints for safety devices of the secondary circuit, their number and throughput capacity shall be specified and substantiated with reference to the Chapter 12 of the NPP SAR.
Measures for the SG isolation from the main steam line in any emergency modes related to loss of the primary circuit integrity shall be specified with reference to Chapter 12 of the NPP SAR.
Measures provided in the NPP design in order to reduce primary-to-secondary leakages shall be described.
5.3. Reactor pressure vessel and its components.
5.3.1. Purpose and design basis.
The following information shall be provided:
information on the purpose and functions of the pressure vessel and its components (the cover and parts of the main reactor joint, fasteners and support structures);
classification of the pressure vessel;
regulatory documents taken into account in the design development, criteria and principles used as the basis for the design of the pressure vessel and its components;
input data for the design defining the required characteristics and parameters of the reactor vessel and cover as well as external conditions and operation factors when these characteristics are to be ensured; load limits for the above-mentioned components under normal operation conditions and in case of any abnormal operation including accidents as well as under any external impacts (typical for the NPP location site) shall be specified;
layout requirements;
requirements for reliability indicators;
the list of the reactor vessel failures considered in the NPP safety analysis.
5.3.2. Description of the design.
The following information shall be provided:
information on the design of the reactor vessel and cover with indication of individual components performing independent functions;
information on the controls, fasteners and sealing devices;
drawings and diagrams demonstrating the design with indication of the constituent parts and materials;
the main technical characteristics of the reactor vessel and cover;
information on the fasteners, supports and their characteristics.
Information on compliance of the design of the reactor pressure vessel and its components with the requirements of federal rules and regulations in the field of atomic energy use governing arrangement and safe operation of equipment and pipelines for nuclear power installations shall be provided.
5.3.3. Materials.
5.3.3.1. Compliance with the requirements of regulatory documents.
It shall be substantiated that materials, manufacturing and control techniques for the reactor pressure vessel comply with the requirements of federal rules and regulations in the field of atomic energy use:
establishing the requirements for arrangement and safe operation of equipment and pipelines for nuclear power installations;
governing the rules for control of weld joints and hard-facing for equipment and pipelines of nuclear power installations;
governing general provisions in the area of welding and hard-facing for equipment and pipelines of nuclear power installations.
5.3.3.2. Standards and technical specifications.
5.3.3.2.1. Materials used to manufacture the reactor vessel and cover as well as materials of the equipment contacting with the reactor vessel shall be listed. Standards and technical specifications for the materials shall be indicated.
5.3.3.2.2. Information on the material selection criteria and substantiation of compliance with these criteria shall be provided.
5.3.3.2.3. Any implemented measures for improvement of the material properties and quality (restrictions with regard to impurities, smelting peculiarities) shall be specified.
5.3.3.2.4. Capability of the materials to operate within the design RF service life in water and steam-gas media under the design temperatures, temperature variations and radiation exposure corresponding to normal operation of the RF and any abnormal operation including design basis accidents shall be substantiated.
5.3.3.2.5. Information on the materials of fasteners and support structures shall be provided in accordance with items 5.3.3.2.2, 5.3.3.2.4 of this Appendix.
5.3.3.3. Manufacturing process.
5.3.3.3.1. The applied manufacturing techniques shall be listed, and compliance with the federal rules and regulations in the field of atomic energy use establishing the requirements for arrangement and safe operation of equipment and pipelines of nuclear power installations shall be substantiated.
5.3.3.3.2. Information on the principal process for manufacturing of the reactor vessel components and its assembly with indication of heat treatment regimes and welding types shall be provided.
5.3.3.3.3. Information on any testing of the blanks for the reactor vessel in the course of manufacturing shall be specified.
5.3.3.3.4. Information on any non-standard or special-purpose processing techniques shall be provided (if any is used), and it shall be substantiated that their application will not affect the reactor vessel integrity.
5.3.3.3.5. Information on operation experience for the pressure vessels manufactured in accordance with the applied methods shall be presented.
5.3.3.3.6. Information on the manufacturing process for fasteners and support structures shall be provided similar to the requirements specified in items 5.3.3.3.1 - 5.3.3.3.5 of this Appendix.
5.3.3.4. Non-destructive control methods.
5.3.3.4.1. Methods for detection of surface and internal defects shall be described in detail, and references to the methodologies shall be given. Information on the quality control program shall be specified.
5.3.3.5. Special control methods for carbon and austenitic stainless steels.
5.3.3.5.1. Information on the control methods for welding, hard-facing, heat treatment and other process operations prescribed for manufacturing of the reactor vessel shall be provided. Compliance with the requirements and recommendations of the following federal rules and regulations in the field of atomic energy use shall be substantiated:
governing the rules for control of weld joints and hard-facing for equipment and pipelines of nuclear power installations;
governing general provisions in the area of welding and hard-facing for equipment and pipelines of nuclear power installations.
5.3.3.5.2. Information on the relevant quality control programs shall be specified.
5.3.3.6. Brittle failure.
5.3.3.6.1. Information on any testing aimed to determine brittle failure resistance characteristics shall be provided, acceptance criteria of the tests shall be specified and information on compliance with these criteria for all reactor vessel components as well as fasteners and support structures shall be given.
5.3.3.6.2. Information on the critical components and points of the reactor vessel and support structures where the permissible brittle failure resistance limits are reached shall be provided.
5.3.3.6.3. The permissible brittle failure resistance limits (fluence, brittle-to-ductile transition temperature, stress intensity factor) shall be specified and substantiated.
5.3.3.7. Material state monitoring in the course of operation.
5.3.3.7.1. Methods, means, scope and frequency of material state monitoring for the reactor vessel shall be provided in order to confirm their operability in the course of operation and compliance with the regulatory requirements of federal rules and regulations in the field of atomic energy use.
5.3.3.7.2. Detailed description of the state monitoring program for the materials of the pressure vessel and weld joints in the course of operation shall be presented.
5.3.3.7.3. It shall be substantiated that the program complies with the requirements of federal rules and regulations in the field of atomic energy use establishing the requirements for arrangement and safe operation of equipment and pipelines of nuclear power installations.
5.3.3.7.4. Surveillance specimen.
5.3.3.7.4.1. Information on the control program with the use of surveillance specimen shall be provided, characteristics of the specimen, their set and planned schedule of withdrawal shall be specified.
5.3.3.7.4.2. It shall be substantiated that the number of specimen complies with the requirements of federal rules and regulations in the field of atomic energy use establishing the requirements for arrangement and safe operation of equipment and pipelines of nuclear power installations.
5.3.3.7.4.3. The scheme for location of specimen in the container and containers in the reactor, information on the container fastening techniques shall be provided; representativeness of the specimen location shall be substantiated with due regard for neutron fluence and irradiation temperature. Information on expected impact of irradiation on the material characteristics (shift of the brittle-to-ductile transition temperature) shall be provided based on the certification tests of the material.
5.3.3.8. Fasteners of the reactor vessel.
Materials and design of the reactor vessel fasteners shall be described. Their compliance with the requirements specified in the federal rules and regulations in the field of atomic energy use governing arrangement and safe operation of equipment and pipelines for nuclear power installations as well as the requirements of standards for the materials permitted for use in manufacturing of equipment and pipelines shall be substantiated.
The following information shall be provided:
non-destructive control operations in the course of manufacturing with the reference to the quality control program;
type, scope and frequency of control in the course of operation.
5.3.4. Control and monitoring.
5.3.4.1. Information shall be provided within the scope in accordance with the requirements of item 4.2.2 of this Appendix.
5.3.4.2. The list of measurement points and information on the diagnostic systems shall be provided.
5.3.5. Tests and inspections.
5.3.5.1. Information on the procedure and scope of technical examination for the pressure vessel and its components shall be specified. Information on compliance of the technical examination procedure and scope with the requirements of federal rules and regulations in the field of atomic energy use governing arrangement and safe operation of equipment and pipelines for nuclear power installations and the requirements of the standard control program shall be provided.
5.3.5.2. Information on the design requirements for the pressure vessel integrity control (based on the requirements of federal rules and regulations in the field of atomic energy use governing arrangement and safe operation of equipment and pipelines for nuclear power installations and federal
rules and regulations in the field of atomic energy use governing the rules for control of weld joints and hard-facing for equipment and pipelines of nuclear power installations) shall be provided; in case the design requirements are assigned by the design developer substantiation of their assignment shall be presented. Information on any control methods adopted by the NPP design developer in addition to the prescribed regulatory documents with indication of the way to document the results of the initial vessel condition inspection shall be provided.
5.3.5.3. Information on the applied control means, their characteristics and experience of application at similar facilities confirming their acceptability shall be provided.
5.3.5.4. Measures to ensure comparability of the control results in various periods of operation as well as in the course of commissioning shall be provided.
5.3.5.5. The following information shall be provided:
incoming control of the reactor vessel and its components prior to installation;
control in the course of installation;
strength, leak-tightness and stability testing after installation.
5.3.6. NPP design analysis.
5.3.6.1. Normal operation and abnormal operation.
The following information shall be provided:
description of the reactor vessel functioning under normal operation conditions in all modes prescribed in the operating regulations for any potential combination of loads (thermal, cyclic, seismic, vibration, radiation and corrosion impact) as well as in case of any abnormal operation;
compliance of the mechanical and strength characteristics and reliability indicators of the reactor vessel and its components with the prescribed requirements in all normal operation modes and in case of any abnormal operation.
5.3.6.2. Functioning in case of any failures.
The following information shall be provided:
analysis of consequences in case of any failures of the reactor vessel with assessment of their consequences;
the list of the reactor vessel failures representing initiating events for abnormal operation, design basis and beyond design basis accidents and requiring additional analysis in the relevant section describing the RF safety analysis.
It shall be substantiated that probability of the reactor vessel breakage does not exceed the value specified in the federal rules and regulations in the field of atomic energy use.
5.3.6.3. Design substantiation.
5.3.6.3.1. Compliance of the design of the pressure vessel and its components with the requirements of regulatory documents shall be substantiated.
5.3.6.3.2. Information on any application of experience in manufacturing, installation, testing and operation of pressure vessels and their components at similar operating plants in the main structural
solutions shall be provided.
5.3.6.3.3. Information on studies and experiments, SRD works performed for the design justification shall be provided.
5.3.6.3.4. The reactor vessel integrity assurance shall be substantiated.
5.3.6.3.5. Calculation results shall be provided proving that the pressure vessel, its components and support structures can accommodate any loads appearing in the course of normal operation, in case of any abnormal operation (including accidents) and also under any natural and human-induced external impacts considered in the NPP design without any loss of operability within the entire service life.
5.3.6.4. Design limits.
5.3.6.4.1. Design limits for normal operation and any abnormal operation including accidents ensuring integrity of the reactor vessel and strength of the reactor support structures shall be specified. Assurance of the pressure vessel integrity in the most severe modes shall be substantiated or references to the relevant sections of the NPP SAR shall be given.
5.3.6.4.2. The following design limits shall be specified for the pressure vessel and the reactor support structures:
pressure;
temperature;
force loads (support structures);
radiation exposure (damaging dose in displacements per atom and fast fluence value).
5.3.6.5. Transportation and installation.
5.3.6.5.1. Information on the measures for protection of the pressure vessel against corrosion and damage in the course of transportation and transportation peculiarities of the permissible means of transport shall be specified.
5.3.6.5.2. Information on the loading and unloading techniques and the installation diagram with indication of the basic operations shall be provided.
Information on the techniques for the pressure vessel installation on supports shall be presented.
5.3.6.6. Maintenance and maintainability.
5.3.6.6.1. Information on maintenance and repair of the pressure vessel and support structures of the reactor and brief description of the repair processes shall be provided.
5.3.6.6.2. Information on the main stages of sealing and unsealing of the main reactor vessel joint and other pressurized detachable joints with indication of the measures to ensure strength and leak-tightness of the joints shall be presented. Information on the assembly procedure, torque values and control methods in the course of the above-mentioned operations shall be specified.
5.3.6.6.3. Measures aimed to eliminate any possibility for occurrence of impermissible stresses in the fastening elements during sealing of detachable joints shall be provided.
5.3.6.7. Reliability analysis for the reactor pressure vessel and support structures.
5.3.6.7.1. Information on reliability analysis and design probability of any failure (breakage) of the
reactor vessel and support structures shall be specified.
5.3.6.7.2. In case of any refurbishment of the reactor core distribution of neutron fluence accumulation rate, damaging dose in displacements per atom at the nuclear core boundaries on the reactor internals, the inner surface of the reactor vessel and within the reactor vessel within the design service life shall be additionally provided; radiation resistance of the reactor vessel and support structures as well as the reactor internals shall be substantiated.
5.3.6.8. Design assessment.
Compliance of the reactor vessel design with the requirements of federal rules and regulations in the field of atomic energy use as well as with the design requirements shall be assessed.
5.4. Primary circuit components.
Information on the components included into the primary circuit as well as individual interfaced systems not described in other chapters of the NPP SAR shall be provided.
Each component or system shall be described in accordance with the standard system description framework given in Appendix 4 hereto.
The component (system) counterpart with well-known operation experience shall be specified; any variations from the counterpart and the reasons to introduce them shall be indicated.
Information on the impact of any damages and failures of the components on the RF safety shall be provided, and any failures with the consequences requiring special analysis shall be indicated.
The final set of the primary circuit components and systems shall be defined in the course of the NPP SAR development for a particular type of RF; in this case the place where safety analysis for each component and system shall be provided (in Chapter 5 of the NPP SAR or in other chapters of the NPP SAR) shall be also defined depending on their peculiarities. In this case they shall be anyway described in accordance with the Standard framework given in Appendix 4 hereto.
Requirements for specific information on individual components of the primary circuit and any interfaced systems which shall be provided in the NPP SAR in addition to the information necessary in accordance with the Standard framework given in Appendix 4 hereto are presented below. This information reflects the peculiarities of individual primary circuit components or interfaced systems.
5.4.1. Main circulation pumps.
Description of the auxiliary RCP systems shall be included into the scope of provided information. Their description shall be given in individual subsections in accordance with the Standard framework given in Appendix 4 hereto.
The following information shall be provided as well:
it shall be substantiated that in case of any loss of the RCP power supply and actuation of the reactor EP at any reactor power level the RCP has sufficient inertia to ensure forced flow of the primary circuit coolant up to the moment when natural circulation assures residual heat removal without any exceedance of operation limits for fuel element damage;
measures to ensure integrity of the RCP flywheel in case of its rotation speed increase during accidents with large-scale coolant leaks or measures to prevent increase of its rotation speed; references to the relevant calculations shall be given.
5.4.2. Steam generators.
The SG characteristics shall include design limits of radioactivity level in the SG secondary circuit under normal operation modes and substantiation of these limits.
Radiation consequences of any ruptures of heat exchange tubes of the SG header and other design basis accidents with primary-to-secondary leakages shall be analyzed or references shall be given to the relevant sections of the NPP SAR where these situations are considered.
Design criteria for prevention of any impermissible damage of the SG heat exchange tubes shall be specified and compliance of the NPP design with these criteria shall be substantiated.
The following information shall be included into the data on performed justifying calculations:
design conditions and assumptions, the list of analyzed modes (from among normal operation modes, abnormal operation and emergency modes) determinant for strength assessment of the heat exchange tubes and their attachment points in the headers;
results of calculations and experiments confirming that the assumed stress intensity level ensures reliable SG operation in accordance with the requirements of federal rules and regulations in the field of atomic energy use;
proofs of integrity maintenance for the heat exchange tubes, the tube sheet and the SG headers in case of design basis accidents with large-scale leaks (ruptures) of the primary and secondary circuit pipelines;
heat exchange surface margin in order to compensate for deterioration of its heat transfer characteristics in the course of operation.
Information on compliance with the requirements of federal rules and regulations in the field of atomic energy use establishing the requirements for arrangement and safe operation of equipment and pipelines of nuclear power installations with regard to the SG equipment with instrumentation and control devices shall be provided.
Information on the SG equipment with the wall metal temperature monitoring devices and the coolant level indicators shall be provided.
5.4.2.1. SG materials.
Information on selection of materials with due regard for specific peculiarities of the SG and its manufacturing process affecting the requirements for the materials shall be provided; it shall be demonstrated how these peculiarities are taken into account in selection of materials.
Information on the SG design peculiarities that can affect any changes of material properties in the course of operation shall be presented.
Compatibility of the SG materials with the primary and secondary circuit coolant shall be substantiated. Brief information on the manufacturing process for the main SG assemblies and detailed information on the manufacturing process for the headers, welding of complex weld joints and the heat exchange tube attachment technique shall be provided, selection of the applied technique shall be substantiated with indication of the measures aimed to prevent crack formation in the perforated zone of the header, information on the heat exchange tube expansion degree shall be specified. Techniques for the heat exchange surface cleaning in the course of manufacturing and cleanness control methods shall be provided. Material selection of the heat exchange tubes shall be substantiated, and requirements for the surface condition, heat treatment and other parameters important to ensure operability of the tubes shall be specified.
Information on the SG transportation methods, measures provided in the design in order to
prevent damage of any SG components in the course of transportation and installation, heat exchange surface preservation necessity and technique, preservation and cleanness control for the inner surface in the course of storage, installation and final assembly at the NPP shall be presented. Brief information on the SG installation procedure shall be provided.
5.4.2.2. SG control and maintenance in the course of operation.
Information on the measures provided in the SG design in order to monitor the state of all its components in the course of operation as well as to ensure the possibility to control each heat exchange tube, the state monitoring program for the SG components and the SG metal condition monitoring program shall be specified. Information on the way to ensure comparability of the control methods prior to the commissioning and in the course of operation shall be provided. It shall be substantiated that the program complies with the requirements of federal rules and regulations in the field of atomic energy use.
Detailed information on the monitoring methods and techniques for the heat exchange tubes, their attachment points, the perforated zone of the tube sheet or the header, phase interface areas, weld joints, detachable joints and the internals shall be provided. Labor intensity of the monitoring and the related dose intensity as well as the work automation level shall be assessed.
Information shall include description of the equipment used for monitoring, operations, control accuracy, recording methods, assessment criteria, intervals of monitoring, measures taken in case of any defect detection and the methods for elimination of any defects in the heat exchange tubes.
Information on the most important SG maintenance activities in the course of operation, cleaning techniques for the heat exchange tubes in order to recover their heat transfer capacity, the procedure for sludge removal from the SG shell shall be specified; characteristics of the water chemistry regime in the secondary circuit and any measures provided in the design for assurance thereof shall be described. WCR restrictions upon breach of which the SG operation is not permitted shall be specified.
In case of necessity references to other sections of the NPP SAR or the relevant materials of the NPP design shall be given.
5.4.3. Pipelines containing the primary circuit coolant.
Information on the set of pipelines operating under the primary circuit pressure (non-isolable section of the primary circuit) shall be provided.
Information on the non-isolable section of the primary circuit shall include the following data:
reactor coolant pipeline;
connection lines for the adjacent systems within the primary circuit pressure boundary.
Description of the pipeline shall include the relevant references to detailed information on the criteria, methods and materials presented in Chapters 3 and 5 of the NPP SAR.
It shall be specified that layout and dimensions of the primary circuit pipelines provide the conditions for natural coolant circulation in the primary circuit in case of any loss or absence of forced circulation under normal operation conditions and any abnormal operation including design basis accidents.
Information on any devices installed of the primary circuit pipelines for control and prevention of impermissible movements under the impact of reactive forces caused by ruptures shall be provided. Strength and efficiency of these devices in case of design basis accidents shall be substantiated.
Information on the measures taken in order to monitor any factors promoting stainless steel cracking due to stress corrosion shall be provided.
WCR characteristics of the primary circuit and any measures provided in the design to assure thereof shall be specified. WCR restrictions upon breach of which operation of the primary circuit components is not permitted shall be specified.
In case the "leak before break" concept is applied in the NPP design reference shall be given to the NPP SAR section where application of this concept is substantiated.
5.4.4. Nuclear core cooling system.
In case the core cooling system used in the NPP design combines safety functions with normal operation functions its description and safety analysis shall be presented in Chapter 12 of the NPP SAR. Only the reference to the relevant section of Chapter 12 of the NPP SAR shall be given in this section.
5.4.5. Residual heat removal system.
All residual heat removal methods (systems) used in the design shall be listed with indication of their functions. Brief information containing description of the residual heat removal process shall be provided.
The section shall contain references to sections of other NPP SAR chapters where residual heat removal systems and (or) systems and components engaged in the residual heat removal process are described.
5.4.6. Pressurizer.
Description of the pressurizer shall contain references to the sections of the NPP SAR chapters where the primary circuit overpressure protection system and the pressure maintenance system for the primary circuit are described.
In presentation of the information this system shall be divided into components (subsystems for pressure reduction in case of its increase and pressure increase in case of its drop). Information on the functions of each subsystem, criteria for performance of the functions prescribed for the subsystems, states of each subsystem components in typical operation modes shall be provided.
Values of the main parameters when each subsystem is actuated and their efficiency characteristics (pressure reduction or increase rate) under normal operation conditions and in case of any abnormal operation including accidents as well as enabling and disabling signals shall be specified. Redundancy degree for the subsystem components and state of the subsystems during design basis accidents shall be indicated.
5.4.8. Valves.
Information on shutoff, check, isolation and control valves included into the primary circuit and (or) the RF shall be provided.
The following information confirming compliance with the requirements of federal rules and regulations in the field of atomic energy use shall be presented:
flywheel rotation direction for closing (opening) of power-operated valves;
presence of gate position indicators on the valves;
force required to close (open) hand-operated valves;
it shall be confirmed that no control valves are used as shutoff valves in the NPP design and no shutoff valves are used as control ones;
information on usage of valves to segregate high and low pressure sections; technical and administrative measures provided in the design in order to eliminate any possibility to change the state of the above-mentioned valves due to erroneous actions of the operating personnel;
information on locking devices and position indication for shutoff valves.
Information confirming compliance with the requirements of federal rules and regulations in the field of atomic energy use establishing general technical requirements for NPP pipeline valves shall be provided.
The main technical data, characteristics of the valves and the list of regulatory documents used as the basis for design, manufacturing and operation of the NPP valves shall be provided; their compliance with the TS shall be confirmed, and references to TORs and TSs shall be given.
Requirements for the valves specified in TORs and TSs and different from the requirements of federal rules and regulations in the field of atomic energy use establishing general technical requirements for NPP pipeline valves shall be listed and substantiated.
The following information shall be provided as well:
working medium velocities in the pipeline at the valve inlet ensuring its operability;
information on leak-tightness of the valves (information on any tests performed to confirm the above-mentioned valve characteristic);
change modes for the working medium parameters;
the required valve closing (opening) time (information on any tests performed to confirm this value);
ambient parameters during and after emergency impacts when the valves maintain their operability;
information on justifying calculations and experimental substantiation of seismic resistance and seismic strength of the valves;
reliability indicators for the applied valves.
5.4.9. Safety and relief devices.
If any safety or relief device is included into the systems described in any other sections of the NPP SAR reference shall be given to the information specified in these sections.
If a safety device performs any safety function or combines performance of a safety function with normal operation functions in the design its description and safety analysis shall be presented in Chapter 12 of the NPP SAR, and reference shall be given in this section to the relevant section of Chapter 12 of the NPP SAR.
5.4.10. Support structures of the main components.
Sketches and brief description of the primary circuit support structures with indication of the loads they are designed for shall be provided.
VI. Requirements for the content of Chapter 6"Steam turbine plant"
Information on the steam turbine plant within the boundaries of the secondary circuit systems shall be provided.
Information on any aspects of the steam turbine plant design and operation affecting the NPP safety shall be specified in Chapter 6.
The framework given in Appendix 4 hereto shall be observed in analysis of the turbine set design performed in Chapter 6 of the NPP SAR in accordance with item 6.1 of this Appendix and the designs of systems within the steam turbine plant performed in accordance with item 6.2 of this Appendix.
6.1. Turbine set.
6.1.1. Design basis.
The following information shall be provided:
type of the turbine set;
requirements for the turbine set cyclic load capability with indication of the permissible number of start-ups within the service life (cold start-up, hot start-up, scheduled and unscheduled shutdowns, load reduction to idle run, load reduction to the lowest limit of the adjustment range with subsequent loading; design duration of start-ups in various thermal conditions from steam supply to the turbine set up to the rated load; adjustment range of automatic power variation; deviation of the rotor rotation speed within the adjustment range and under emergency conditions);
requirements for the turbine set protection against missiles, the turbine set overspeed and short circuit in the generator with indication of regulatory documents and manuals;
requirements for the turbine set structural design, reliability, time between failures, inter-maintenance period, complete service life, seismic resistance;
requirements for the turbine set layout and orientation;
requirements for location of explosion-hazardous and flammable materials;
turbine set start-up and shutdown conditions;
parameters characterizing impermissible exceedance of the turbine set rotation speed.
6.1.2. Design of the turbine set.
Compliance with the requirements used as the basis for the turbine set design shall be substantiated.
Zones of potential ejection of missiles that can be formed due to mechanical breakage of the turbine set rotor or blades caused by overspeed or short circuit in the generator in the sector 25° in relation to the rims of the intermediate and low pressure cylinders for each turbine set in the turbine hall room as well as the area of potential missile hitting in relation to all safety-related systems (components) shall be indicated on the turbine set layout plan. It shall be substantiated that any possible damages caused by missiles due to mechanical breakage of the turbine set rotor or blades will
not result in any disturbances of the SS functions, damage of oil systems, systems containing flammable gases or high-pressure gases.
The following information on the turbine set components shall be provided:
information on brittle strength calculations for the rotor;
information on tearing resistance characteristics of the turbine set rotor;
strength characteristics of the turbine set disks and other most stressed devices;
substantiation of the selection of devices for protection of the turbine set and its equipment against impermissible overpressure;
descriptions of the turbine set overspeed protection system with indication of the redundancy methods, reliability assessment for the assemblies, the procedure for control and testing of this system.
Information on the materials used to manufacture the turbine set components as well as data on the manufacturing process for rotors, disks and rotating blades shall be provided.
Information on emergency modes for the turbine set shall be specified, in this case operation of BRU-A and BRU-K shall be reflected.
Initiating events associated with the turbine set failures that may result in accidents shall be listed. Reference shall be given to analysis of accidents caused by the turbine set failures provided in Chapter 15 of the NPP SAR.
6.1.3. Control and monitoring of the turbine set operation.
The following shall be described:
process parameters used to arrange the turbine set protection and affecting the reactor EP, the reactor power governor, preventive protection of the reactor and the turbine set overspeed;
systems for monitoring of the turbine set rotation overspeed with indication of information on redundancy of the control and monitoring devices and the applied overspeed control device type;
protection and interlocks affecting the reactor EP, the reactor load reduction and power limitation device, preventive protection of the reactor.
6.1.4. Tests and inspections.
Information on the programs for pre-operational commissioning testing and in-service inspection of the entire turbine set, its locking control devices and the turbine set overspeed governor shall be provided.
6.1.5. Design analysis.
Normal operation of the turbine set, modes with sudden load reduction and any potential transient processes shall be analyzed, in this case information on operation of the turbine set control system and its overspeed protection shall be provided.
Functioning of the turbine set in case of any abnormal operation, emergency situations and accidents shall be considered. Information on functioning of the turbine set and associated systems in case of any abnormal operation related directly to the turbine set or caused by deviations in the steam turbine plant systems shall be provided. Impact of the above-mentioned modes on the reactor power governor, preventive protection and emergency protection as well as on BRU-A and BRU-K operation
shall be analyzed. The list of initiating events in the turbine set leading to accidents shall be presented. It shall be substantiated that any initiating event in the turbine set will not result in an accident or emergency situation at the NPP.
Information on the turbine set functioning under external impacts shall include description of the turbine set state (operation or shutdown) under external impacts considered with regard to the turbine set. Information on the level of external impacts when the turbine set is to be stopped shall be specified.
The following information shall be provided:
any possible missiles caused by rupture of pipelines and pressurized vessels; analysis of impact of such ruptures on the SS functioning and heat removal from the RF;
results of calculations confirming strength, stability and operability of the turbine set components under external impacts of natural and human-induced origin in accordance with their classification;
impact of the turbine set failures on actuation of safety systems.
6.2. Systems within the steam turbine plant boundaries.
Designs of systems within the steam turbine plant boundaries shall be analyzed in accordance with the list provided below (in this case individual sections may be excluded from this list or new ones may be added depending on presence (absence) of any systems in the NPP design):
turbine set lubrication system;
hydrolift and barring gear system;
turbine set sealing system;
turbine set drainage system;
turbine set control oil supply system;
vacuuming system;
turbine set condenser system;
main condensate pipeline system;
high pressure regeneration system;
separation and reheating system;
fresh steam pipeline system;
feedwater system;
turbine set bypass system;
secondary circuit overpressure protection system;
secondary circuit makeup system;
water chemistry regime of the secondary circuit and systems for its maintenance;
condensate purification system;
secondary circuit process media sampling system.
Besides the following information shall be provided for the analyzed systems:
any possible missiles caused by rupture of pipelines and pressurized vessels; analysis of impact of such ruptures on the SS functioning and heat removal from the RF;
results of calculations confirming strength, stability and operability of the components of systems within the steam turbine plant boundaries under external impacts of natural and human-induced origin in accordance with their classification;
impact of any failures in the components of systems within the steam turbine plant boundaries on actuation of safety systems.
VII. Requirements for the content of Chapter 7 "Control and monitoring"
Each safety-related control system shall be described in accordance with the standard system description framework given in Appendix 4 hereto.
Besides additional information required in accordance with this chapter as well as information specific for a particular system shall be provided for each safety-related system under consideration.
7.1. Introduction.
7.1.1. The list of safety-related control systems.
Safety-related control systems as well as components of these systems (instrumentation and control devices; indicators; controls; sensors; transducers; programmable digital devices; software used to perform controlling and information functions of the safety-related control systems) shall be listed.
Information on availability of quality assurance programs for safety-related control systems developed with due regard for the federal rules and regulations in the field of atomic energy use governing the requirements for quality assurance programs at nuclear facilities shall be provided. Information on availability of software quality assurance procedures shall be provided for each stage of the software life cycle and the software life cycle report.
Information on the design names and designations of the systems shall be provided, and classification of the systems as NOCS, CSS or special-purpose hardware for BDBA management shall be described.
If any system performs NOCS and CSS function at the same time it shall be described in the section of Chapter 7 of the NPP SAR performed in accordance with item 7.4 of this Appendix. The section of Chapter 7 of the NPP SAR performed in accordance with item 7.2 of this Appendix shall contain only the name of the system, normal operation functions performed by it and reference to the relevant section of Chapter 7 of the NPP SAR with its complete description.
Functional groups of the control systems shall be specified and their classification in accordance with the provisions of federal rules and regulations in the field of atomic energy use establishing the requirements for safety-related control systems shall be provided.
Information on newly developed systems (functional groups, components, software) and any applied systems (components, software) proven by the previous operation experience shall be provided. Application of any commercial off-the-shelf and previously developed software in safety-related control systems shall be substantiated.
7.1.2. Basic safety principles and criteria.
Information on regulatory requirements, design criteria and any other requirements taken into account in the design of the systems (functional groups, components) listed in the section of Chapter 7 of the NPP SAR performed in accordance with item 7.1.1 of this Appendix shall be presented.
Compliance of the adopted design solutions with state of the art in science, technology and production shall be substantiated.
7.1.3. Measures for prevention or protection against common cause failures.
The list of considered common cause failures shall be presented. Measures for prevention or
protection against common cause failures shall be specified. Sufficiency of these measures for the NPP safety assurance shall be substantiated.
Results of the analysis of safety-related control systems susceptibility to common cause failures shall be provided.
7.1.4 Measures aimed to ensure computer security and integrity of the software used to perform controlling and information functions of safety-related control systems.
Information on the measures aimed to ensure computer security and integrity of the software shall be specified. Sufficiency of these measures for the NPP safety assurance shall be substantiated.
7.1.5. Test results.
Results of proving-ground trials of the systems or their individual components shall be provided.
7.2. Safety-related normal operation control systems.
7.2.1. Unit NOCS.
7.2.1.1. Purpose and design basis.
Information on the requirements used as the basis to design the safety-related unit control systems, substantiation of these requirements, purpose of systems, principles and design criteria taken as the basis for their design shall be provided.
7.2.1.2. Description of the system. Functioning under normal operation conditions.
Information containing description of the unit NOCS, data on the configuration, basic technical characteristics, description of the unit NOCS operation principle under normal operation conditions and in case of any abnormal operation including accidents with due regard for interaction with other systems shall be provided.
Information on the unit NOCS constituent parts and components ensuring the following functions shall be specified:
remote, automated and (or) automatic control of the NPP normal operation systems;
monitoring of the parameters characterizing the NPP functioning in all possible normal operation modes and presentation of any information thereof as well as information on any deviations from normal operation to the operator;
presentation of consolidated information on the state of NPP safety functions (the safety parameters presentation system) to the personnel in normal operation modes and in any abnormal operation modes including design basis and beyond design basis accidents;
group and individual communication between the MCR, the ECR and the NPP operating personnel performing any works outside the control rooms;
interface with the adjacent systems and data transmission;
state diagnostics for the unit NOCS equipment and software and hardware tools.
Information on the unit NOCS components (functional groups) shall also contain data on their configuration, basic technical characteristics, location, diagrams of systems and tools, description of the operation principle under normal operation conditions and in case of any abnormal operation including
accidents.
The following information shall be provided:
methods and results of reliability assessment for the system;
data on power supply, stability in case of any variations of power supply parameters and electrical impacts, electromagnetic compatibility and protection, resistance to the environmental impacts and operation conditions; systems supporting the required ambient parameters at the locations of the system equipment and the personnel;
stability substantiation for the automatic regulation circuits;
description of the adopted approaches to computer security;
results of conformity assessment and testing of the systems (system components);
figures, schemes, diagrams, curves, tables explaining the adopted technical solutions;
information on verification and validation (with indication of their scope substantiation and results) of the software used in NOCS.
7.2.1.3. Commissioning works.
The adopted scope of commissioning works, completeness of the scope of the planned administrative and technical measures, the list of potentially hazardous works and arrangements aimed to prevent accidents shall be substantiated.
Information on the methods to check operability of the unit NOCSs and their components, their integrated testing, integration within the APCS, diagnostics and documentation of their characteristics, acceptance criteria and their substantiation shall be provided.
7.2.1.4. Operation limits and conditions. Maintenance.
Operation limits and conditions related to the unit NOCS and ensuring prevention of any deviations from the NPP safe operation limits and conditions shall be substantiated.
Solutions for diagnostics and regular state control for the unit NOCSs, their regular inspections and testing for performance of the required functions, recording and documentation of any malfunctions and failures as well as the personnel training shall be substantiated.
Implemented measures and procedures aimed to eliminate any malfunctions and defects in the course of maintenance shall be substantiated.
Absence of any adverse impact of maintenance on the NPP safety shall be substantiated.
Ageing management measures shall be substantiated.
7.2.1.5. System functioning in case of failures and abnormal operation.
Results of analysis of the unit NOCS failure types and their impact on the NPP safety demonstrating compliance with the design criteria as well as regulatory requirements shall be provided.
Analysis of response of systems and components to external and internal impacts, response of systems to any possible failures and malfunctions, errors and erroneous decisions of the personnel shall be provided.
Absence of any impact of failures on the NPP safety shall be substantiated for non-safety-related components of the unit NOCS.
7.2.2. Main control room.
7.2.2.1. Purpose and design basis.
Information on the requirements used as the basis for the MCR design, substantiation of these requirements, MCR purpose, principles and design criteria taken as the basis for its design shall be provided.
7.2.2.2. Description. Functioning under normal operation conditions.
Description of the MCR and any instrumentation and control devices related to it shall be provided as well as:
general layout of the MCR;
configuration of the MCR in-process circuit panels with any automation devices located on them;
general layouts of the MCR consoles and boards with any automation devices located on them;
information on location of safety-related automation devices and information necessary to substantiate ergonomic requirements for their application, arrangement of information and body fields on the control room panels and boards of the control station (stations).
Substantiation of the following technical solutions shall be provided:
automatic provision of information on the state of process equipment and safety-related automation devices to the operator;
independent operability check for process equipment and safety-related automation devices performed by the operator in the course of functioning;
the list of functions performed automatically with submittal of the relevant information to the operator;
the list of functions performed by the operators.
Information on recording of the controlling personnel's actions in case of any abnormal operation shall be provided with reference to the information specified in the section of Chapter 8 of the NPP SAR performed in accordance with item 8.5 of this Appendix.
Information substantiating duplication of automatically implemented functions with functions performed with involvement of the operator shall be presented.
It shall be specified how the MCR ensures control and monitoring of the following items under normal operation conditions and in case of accidents: the RF; safety systems; any other systems of the NPP power unit as provided in the design.
Information on the operation principle of the MCR and its components in coordination with other systems and any equipment related to it under normal operation conditions as well as in case of any abnormal operation including accidents shall be provided.
Instrumentation and control devices making information suitable for the operator to perform any necessary actions aimed to ensure safety shall be described.
Solutions for human-machine interface adopted in the NPP design shall be substantiated.
Information on substantiation of workspace sufficiency for all operating personnel both under normal operation conditions and in case of any abnormal operation including accidents shall be provided.
Sufficiency of the implemented measures aimed to restrict access to the control rooms for the persons not included into the shifts both under normal operation conditions and in case of any abnormal operation including accidents shall be substantiated.
It shall be specified that adequate conditions for performance of all required functions are arranged at workplaces of the operators.
The following information shall be provided:
location of the information display means depending on its importance for the NPP safety on the MCR panels and the boards of the station (stations);
distinctive coloring of the information display means depending on its importance for the NPP safety;
the operator's convenience in observation over the information display (zone of vision, sizes of scales, figures and other symbols);
reliability of the applied lighting for scales, figures and other symbols on the display means;
location of the controls for operating elements of safety-related systems (components) on the fields of the control room panels and boards of the station (stations) with due regard for convenience of observation over the displayed information required for control through the use of these means;
distinctive coloring of the controls for operating elements of safety-related systems (components);
devices for authorized access to the controls for operating elements of safety-related systems (components) if such requirements are prescribed.
The following shall be substantiated:
illumination of the operators' workplaces;
color, sound, and other distinctive characteristics of alarms that shall be well identified by the operator and have uniform interpretation at all control rooms of the NPP power unit;
application of the communication means (with due regard for the information specified in the section of Chapter 8 of the NPP SAR performed in accordance with item 8.5 of this Appendix);
application of the closed circuit TV means (with due regard for the information specified in the section of Chapter 8 of the NPP SAR performed in accordance with item 8.5 of this Appendix);
application of the MCR information means intended for use by all operators in the shift;
ergonomics of the technical solutions for manual and automated information recording by the operator at the workplace;
structural solution for storage of any documentation necessary for in-process application at the operator's workplace;
techniques and means for organization of the operator's meals at the workplace in any regular
and contingency situations as well as in case of accidents.
7.2.2.3. Commissioning works.
The adopted scope of commissioning works, completeness of the scope of the planned administrative and technical measures, the list of potentially hazardous works and arrangements aimed to prevent accidents shall be substantiated.
7.2.2.4. Operation limits and conditions. Maintenance.
Operation limits and conditions related to the MCR and ensuring prevention of any deviations from the NPP safe operation limits and conditions shall be substantiated with due regard for the information specified in Chapter 16 of the NPP SAR.
7.2.2.5. Functioning in case of failures and abnormal operation.
Results of analysis of the MCR equipment failure types and their impact on the NPP safety demonstrating compliance with the design criteria as well as regulatory requirements shall be provided.
Analysis of response of systems and components to external and internal impacts, response of systems to any possible failures and malfunctions, errors and erroneous decisions of the personnel shall be provided.
Reliability analysis results for all components and constituent parts of the MCR shall be provided and selection of the parameters to be displayed for the operator under normal operation conditions and in case of any abnormal operation including accidents shall be justified; it shall be substantiated that the selected and displayed parameters ensure submittal of unambiguous information on compliance with the NPP safe operation limits and conditions to the operator as well as identification and diagnostics of the SS actuation and functioning.
Survivability and habitability of the MCR under normal operation conditions and in case of any abnormal operation including accidents shall be substantiated.
Analysis shall be provided to demonstrate that the operator has sufficient information to perform any manual operations required from the viewpoint of the NPP safety and sufficient time to make correct decisions and to perform any actions in case they are necessary.
It shall be substantiated that the operator can take the data and readings of the instruments in order to monitor conditions in the reactor, the primary circuit, the RF containment, the state of safety systems and BDBA management hardware in all normal operation modes and also in case of any abnormal operation including accidents.
Information shall include design criteria, types of reading units, the number of reading channels, the measurement range for the parameters in these channels, accuracy and location of the instruments as well as substantiation of the calculation sufficiency.
7.2.3. NOCS not referred to unit NOCS.
7.2.3.1. Purpose and design basis.
Information on the requirements used as the basis to design the safety-related control systems not referred to unit NOCS, substantiation of these requirements, purpose of systems, principles and design criteria taken as the basis for their design shall be provided.
7.2.3.2. Functioning under normal operation conditions.
Information containing description of the NOCSs not referred to the unit NOCS, data on the configuration, basic technical characteristics, description of each NOCS operation principle under normal operation conditions and in case of any abnormal operation including accidents with due regard for interaction with other systems shall be provided.
The following information shall be provided:
on the RF diagnostic means, the RF NOCS diagnostic means;
on the in-core monitoring system;
on the recording system;
on the means for monitoring of neutron poisoning isotopes content in the primary circuit;
on the means for monitoring of neutron poisoning isotopes content in the boric solution tanks;
on the operator information support systems;
on the closed circuit TV systems or means;
on the means for communications with the MCR, the ECR and local control stations;
on the means for transmission of signals to and from protected emergency response control posts;
on the devices for generation of emergency warning alarm signals, emergency warning, indicating, independent information recording and storage means.
For each NOCS information on the NOCS components and constituent parts performing the following functions shall be provided:
remote, automated and (or) automatic control of the NPP normal operation systems;
monitoring of the parameters characterizing the NPP functioning in all possible normal operation modes and presentation of any information thereof as well as information on any deviations from normal operation to the operator;
interface with the adjacent systems and data transmission;
state diagnostics for the NOCS equipment and software and hardware tools.
Information on the NOCS components (functional groups) shall also contain data on their configuration, basic technical characteristics, location, diagrams of systems and tools, description of the operation principle under normal operation conditions and in case of any abnormal operation including accidents.
The following information shall be provided:
methods and results of reliability assessment for the system;
data on power supply, stability in case of any variations of power supply parameters and electrical impacts, electromagnetic compatibility and protection, resistance to the environmental impacts and operation conditions, systems supporting the required ambient parameters at the locations of the system equipment and the personnel;
stability substantiation for the automatic regulation circuits;
description of the adopted approaches to computer security;
results of conformity assessment and testing of the systems (system components);
figures, schemes, diagrams, curves, tables explaining the adopted technical solutions;
information on verification and validation (with indication of their scope substantiation and results) of the software used in NOCS.
It shall be described how the RF NOCS, its constituent parts and components ensure monitoring of the RF technical state and safe control of the RF under normal operation conditions.
7.2.3.3. Commissioning works.
The adopted scope of commissioning works, completeness of the scope of the planned administrative and technical measures, the list of potentially hazardous works and arrangements aimed to prevent accidents shall be substantiated.
Information on the methods to check operability of the NOCSs and their components, their integrated testing, integration within the APCS, diagnostics and documentation of their characteristics, acceptance criteria for inspections and tests and their substantiation shall be provided.
7.2.3.4. Operational limits. Maintenance.
Operation limits and conditions related to the NOCSs and ensuring prevention of any deviations from the NPP safe operation limits and conditions shall be substantiated with due regard for the information specified in Chapter 16 of the NPP SAR.
Solutions for diagnostics and regular state control for the NOCSs, their regular inspections and testing for performance of the required functions, recording and documentation of any malfunctions and failures as well as the personnel training shall be substantiated.
Implemented measures and procedures aimed to eliminate any malfunctions and defects in the course of maintenance shall be substantiated.
Absence of any adverse impact of maintenance on the NPP safety shall be substantiated.
Ageing management measures shall be substantiated.
7.2.3.5. System functioning in case of failures and abnormal operation.
Results of analysis of the NOCS failure types and their impact on the NPP safety demonstrating compliance with the design criteria as well as regulatory requirements shall be provided.
Analysis of response of systems and components to external and internal impacts, as well as analysis of the response of systems to any possible failures and malfunctions, errors and erroneous decisions of the personnel shall be provided.
Absence of any impact of failures on the NPP safety shall be substantiated for non-safety-related NOCS components.
7.3. Control and protection system.
7.3.1. Purpose and design basis.
Information on the requirements used as the basis for the CPS design, substantiation of these requirements, purpose of the system, principles and design criteria taken as the basis for its design shall be provided.
7.3.2. Description of the control and protection system.
Information containing description of the CPS, data on its configuration, basic technical characteristics, description of the system operation principle under normal operation conditions and in case of any abnormal operation and accidents with due regard for interaction with other systems and any equipment related to it shall be provided.
Information on the CPS subsystems and components performing the following functions shall be presented:
remote, automated and (or) automatic control;
submittal of information on the RF and NPP parameters to the operator;
interface with the adjacent systems and data transmission;
state diagnostics for the CPS equipment and software and hardware tools.
The following information shall be provided:
methods and results of reliability assessment at different stages of the system life cycle;
data on power supply, stability in case of any variations of power supply parameters and electrical impacts, electromagnetic compatibility and protection, resistance to the environmental impacts and operation conditions, systems supporting the required ambient parameters at the locations of the system equipment and the personnel;
stability substantiation for the automatic regulation circuits;
description of the adopted approaches to computer security;
equipment conformity assessment results;
equipment testing results;
figures, schemes, diagrams, curves and tables necessary to substantiate the adopted technical solutions for performance of the required functions.
Description of the systems included into the CPS shall also contain:
the system structure;
information on the hardware;
functions performed by the system automatically;
functions performed by the operator;
description of the subsystem operation principle;
description of non-safety-related system components;
description of safety-related system components.
The following information shall be provided:
lists of the reactor EP actuation conditions;
description of the EP actuation condition formation logic for each parameter;
description of the redundant protection actuation techniques;
description of the conditions for authorized access to actuation of protections;
description of redundancy of the channels implementing the protection functions;
substantiation of the EP system compliance with the principle of diversity.
Besides the following information shall be provided for each CPS system:
operation algorithms;
configuration, structure and characteristics of channels (components);
power supply;
information on location of the hardware.
Neutron flux and reactivity control systems and the RF power control systems as well as their channels and components shall be described:
monitoring channels;
recording devices;
additional monitoring system (in case of necessity);
reactimeters;
automatic operability check means for the monitoring channels and malfunction warning alarm;
automatic reactor power controller;
reactor power governor;
preventive protection systems;
core sub-criticality control means;
monitoring of energy emission heterogeneity across the core;
in-process calculation of the burnout ratio as well as power density field variation monitoring and control means.
Information on all other systems included into the CPS shall be provided.
Initial design information on all parameters and characteristics of the CPS systems, their diagrams and location data shall be provided.
Scope and completeness of the metrological support for the system shall be substantiated.
Information on verification and validation (with indication of their scope substantiation and results) of the software used in CPS shall be provided. Besides information on metrological validation of
the software shall be specified. In case any RF parameter calculation algorithms through the use of numerical modelling of physical processes are implemented in the software (with the possibility to assess uncertainty of the calculation results) information on validation of such software components as individual software tools shall be provided.
7.3.3. Commissioning works.
The adopted scope of commissioning works, completeness of the scope of the planned administrative and technical measures, the list of potentially hazardous works and arrangements aimed to prevent accidents shall be substantiated.
Information on the methods to check operability of the CPS, its integrated testing, integration within the APCS, diagnostics and documentation of its characteristics, acceptance criteria and their substantiation shall be provided.
7.3.4. Maintenance.
Operation limits and conditions related to the CPS and ensuring prevention of any deviations from the NPP safe operation limits and conditions shall be substantiated.
Solutions for diagnostics and regular state control for the CPS, its regular inspections and testing for performance of the required functions, recording and documentation of any malfunctions and failures as well as the personnel training shall be substantiated.
Implemented measures and procedures aimed to eliminate any malfunctions and defects in the course of maintenance shall be substantiated.
Absence of any adverse impact of maintenance on the NPP safety shall be substantiated.
Ageing management measures shall be substantiated.
7.3.5. Functioning in case of failures and abnormal operation.
Results of analysis of the CPS failure types and their impact on the NPP safety demonstrating compliance with the design criteria as well as regulatory requirements shall be provided.
Analysis of response of systems and components to external and internal impacts, as well as analysis of the response of systems to any possible failures and malfunctions, errors and erroneous decisions of the personnel shall be provided.
Absence of any impact of failures on the NPP safety shall be substantiated for non-safety-related CPS components.
Information demonstrating that failures of any channel of the automatic power controller or its disabling will not cause any changes of the reactor power due to action of the automatic control system shall be also provided.
Analysis results shall substantiate that failures of the level and (or) neutron flux density change rate control channels are accompanied with an alarm for the operator and recording of the failure. Any measures implemented to prevent insertion of positive reactivity shall be also substantiated.
Analysis shall be presented in order to define provision of the following information to the operator in all RF operation modes:
parameters determining the reactor core state;
parameters of the primary circuit and state of the systems performing heat removal to the ultimate heat sink;
state of safety systems;
state of the automation devices;
parameters in the RF containment.
7.4. Control safety systems (except for the CPS).
7.4.1. Purpose and design basis.
Results of analysis of the CSS failure types and their impact on the NPP safety demonstrating the CSS compliance with the design criteria as well as regulatory requirements shall be provided.
Analysis of response of systems and components to external and internal impacts, as well as analysis of the response of systems to any possible failures and malfunctions, errors and erroneous decisions of the personnel shall be provided.
7.4.2. Description of the control safety systems.
Information containing description of the CSS, data on the configuration, basic technical characteristics, description of the CSS operation principle under normal operation conditions and in case of any abnormal operation including accidents with due regard for interaction with other systems shall be provided.
Information on the CSS constituent parts and components ensuring the following functions shall be specified:
monitoring of the parameters characterizing operation of the NPP and submittal of the relevant information to the operator;
interface with the adjacent systems and data transmission;
state diagnostics for the CSS equipment and software and hardware tools.
Information on the CSS components (functional groups) shall also contain data on their configuration, basic technical characteristics, location, diagrams of systems and tools, description of the operation principle under normal operation conditions and in case of any abnormal operation including accidents.
The following information shall be provided:
methods and results of reliability assessment for the system;
data on power supply, stability in case of any variations of power supply parameters and electrical impacts, electromagnetic compatibility and protection, resistance to the environmental impacts and operation conditions, systems supporting the required ambient parameters at the locations of the system equipment and the personnel;
stability substantiation for the automatic regulation circuits;
adopted approaches to computer security;
results of conformity assessment and testing of the systems (system components);
figures, schemes, diagrams, curves, tables explaining the adopted technical solutions;
information on verification and validation (with indication of their scope substantiation and results) of the software used in CSS.
Description of each CSS shall contain:
system structure;
functions performed by the system automatically;
description of non-safety-related system components;
system operation algorithms;
configuration, structure and characteristics of the system channels;
description of the system operation principles.
Implementation of the independence, redundancy and diversity principles in the systems shall be substantiated.
7.4.3. Commissioning works.
The adopted scope of commissioning works, completeness of the scope of the planned administrative and technical measures, the list of potentially hazardous works and arrangements aimed to prevent accidents shall be substantiated.
Information on the methods to check operability of the CSSs and their components, their integrated testing, integration within the APCS, diagnostics and documentation of their characteristics, acceptance criteria and their substantiation shall be provided.
7.4.4. Maintenance.
Implemented measures and procedures aimed to eliminate any malfunctions and defects in the course of maintenance shall be substantiated.
Absence of any adverse impact of maintenance on the NPP safety shall be substantiated.
7.4.5. Functioning in case of failures and abnormal operation.
Results of analysis of the CSS failure types and their impact on the NPP safety demonstrating compliance with the design criteria as well as regulatory requirements shall be provided.
Analysis of response of systems and components to external and internal impacts, as well as analysis of the response of systems to any possible failures and malfunctions, errors and erroneous decisions of the personnel shall be provided.
It shall be specified how the possibility of any safety system disabling by the operator within 10-30 minutes after its automatic activation is prevented in the CSS design.
7.5. Emergency control room.
7.5.1. Purpose and design basis.
Information on the requirements used as the basis for the ECR design, substantiation of these requirements, the ECR purpose, principles and design criteria taken as the basis for its design shall be
provided.
7.5.2. Description of the emergency control room.
Information on the ECR and any instrumentation and control devices related to its shall be provided as well as:
general layout of the ECR;
configuration of the ECR panels with any automation devices located on them;
general layouts of the ECR consoles and boards with any automation devices located on them;
information on location of safety-related automation devices and information necessary to substantiate ergonomic requirements for their application, arrangement of information and body fields on the control room panels and boards of the control station (stations).
Information on recording of the personnel's actions shall be provided with reference to the information specified in the section of Chapter 8 of the NPP SAR performed in accordance with item 8.5 of this Appendix.
Information substantiating duplication of automatically implemented functions with functions performed with involvement of the operator shall be presented.
It shall be specified how the ECR ensures monitoring and control of the RF and the safety systems under normal operation conditions and in case of any accidents.
Information on the operation principle of the ECR and its components in coordination with other systems under normal operation conditions as well as in case of any abnormal operation including accidents shall be provided.
Instrumentation and control devices making information suitable for the operator to perform any necessary actions aimed to ensure the NPP safety shall be described.
Adopted solutions for human-machine interface shall be substantiated.
Sufficiency of the workspace for the operating personnel shall be substantiated.
Sufficiency of the implemented measures aimed to restrict access to the control rooms for the persons not included into the shifts both under normal operation conditions and in case of any abnormal operation including accidents shall be substantiated.
It shall be specified that adequate conditions for performance of all required functions are arranged at workplaces of the operators.
The following information shall be provided:
location of the information display means depending on its importance for the NPP safety on the ECR panels and the boards of the station (stations);
distinctive coloring of the information display means depending on its importance for the NPP safety;
the operator's convenience in observation over the information display (zone of vision, sizes of scales, figures and other symbols);
reliability of the applied lighting for scales, figures and other symbols on the display means;
location of the controls for operating elements of safety-related systems (components) on the fields of the control room panels and boards of the station (stations) with due regard for convenience of observation over the displayed information required for control through the use of these means;
distinctive design of the controls for operating elements of safety-related systems (components);
devices for authorized access to the controls for operating elements of safety-related systems (components) if such requirements are prescribed.
The following shall be substantiated:
illumination of the operators' workplaces;
color, sound, and other distinctive characteristics of alarms that shall be well identified by the operator and have uniform interpretation at all control rooms of the NPP power unit;
application of the communication means with due regard for the information specified in the section of Chapter 8 of the NPP SAR developed in accordance with item 8.5 of this Appendix;
application of the closed circuit TV means with due regard for the information specified in the section of Chapter 8 of the NPP SAR developed in accordance with item 8.5 of this Appendix;
ergonomics of the technical solutions for manual and automated information recording by the operator at the workplace;
structural solution for storage of any documentation necessary for in-process application at the operator's workplace;
techniques and means for organization of the operator's meals at the workplace in any regular and contingency situations as well as in case of accidents.
It shall be specified that the adopted solutions ensure reliable bringing of the reactor into sub-critical state and long-term maintenance of this state, heat removal to the ultimate heat sink, actuation of safety systems and obtaining of the information on the reactor condition through the use of the ECR.
Independence of the ECR from the MCR shall be substantiated by detailed description of the adopted arrangements and technical solutions. It shall be substantiated that any common cause failure of the MCR and the ECR is prevented.
7.5.3. Commissioning works.
The adopted scope of commissioning works, completeness of the scope of the planned administrative and technical measures, the list of potentially hazardous works and arrangements aimed to prevent accidents shall be substantiated.
7.5.4. Operational limits. Maintenance.
Operation limits and conditions related to the ECR and ensuring prevention of any deviations from the NPP safe operation limits and conditions shall be substantiated.
The adopted solutions for maintenance of the ECR operability under normal NPP operation conditions shall be substantiated.
7.5.5. Functioning in case of failures and abnormal operation.
Results of analysis of the ECR equipment failure types and their impact on the NPP safety demonstrating compliance with the design criteria as well as regulatory requirements shall be provided.
Analysis of response of systems and components to external and internal impacts, response of systems to any possible failures and malfunctions, errors and erroneous decisions of the personnel shall be provided.
Reliability analysis results for all components and constituent parts of the ECR shall be provided and selection of the parameters to be displayed for the operator under normal operation conditions and in case of any abnormal operation including accidents shall be justified. It shall be substantiated that the selected and displayed parameters ensure submittal of unambiguous information on compliance with the NPP safe operation limits and conditions to the operator as well as identification and diagnostics of the SS actuation and functioning.
Survivability and habitability of the ECR under normal operation conditions and in case of any abnormal operation including accidents shall be substantiated.
It shall be substantiated that the operator has sufficient information to perform any operations required to ensure performance of safety functions and control over the conditions of the RF and SNF pools and sufficient time to make correct decisions and to perform any actions in case they are necessary.
It shall be substantiated that the operator can take the data and readings of the instruments in order to monitor conditions in the reactor, the primary circuit, the RF containment, the state of safety systems and BDBA management hardware in all normal operation modes and also in case of any abnormal operation including accidents.
Information shall include design criteria, types of reading units, the number of reading channels, the measurement range for the parameters in these channels, accuracy and location of the instruments as well as substantiation of the calculation sufficiency.
The list of safety functions implemented from the ECR and the list of safety systems and special-purpose BDBA management hardware controlled from the ECR shall be provided. Conditions for transfer of the MCR operating personnel to the ECR in case of the MCR failure shall be also described.
Solutions for assurance of the ECR habitability and survivability in case of design basis and beyond design basis accidents shall be analyzed.
7.6. Non-safety-related intrumentation and control systems.
7.6.1. Description.
The following information shall be provided:
the list of systems and components;
the list and substantiation of the design peculiarities of the systems not identical to the similar systems of the operating NPP power units.
7.6.2. Functioning in case of failures and abnormal operation.
Analysis shall be provided in order to demonstrate that the systems are not required for the NPP safety assurance.
VIII. Requirements for the content of Chapter 8 "Power supply, communications and warning"
Information substantiating functional development and reliability of the supporting power supply systems, capacity sufficiency, multi-channel arrangement, independence, resistance to external and
internal impacts, the possibility for maintenance, testing and repair, compliance with the requirements of federal rules and regulations in the field of atomic energy use on the basis of their functioning analysis under normal operation conditions and in case of any abnormal operation and failures of power supply systems with due regard for human errors as well as during design basis and beyond design basis accidents shall be provided in Chapter 8 of the NPP SAR. Besides qualitative and quantitative analyses of power supply reliability shall be presented in Chapter 8 of the NPP SAR.
The basic principles for design and operation of the NPP electrical systems shall be stated in Chapter 8 of the NPP SAR.
Each safety-related power supply system shall be described in accordance with the standard system description framework given in Appendix 4 hereto. Besides additional information required in accordance with this chapter as well as information specific for a particular system shall be provided for each system under consideration.
8.1. External power grid.
8.1.1. Power output diagram.
The following information shall be provided:
development of the power grid;
the NPP purpose and function in the power grid;
characteristics of the power output diagram and the main electric circuit diagram;
the possibility of power output to regional substations without construction of any switchgear at the NPP;
protection of the networks and substations against external impacts;
availability of emergency automation, its schematic diagram and quantitative reliability characteristics;
surge voltage protection;
voltage variations;
availability of the automated dispatch control system;
organization of operation for electric power grids;
requirements for the NPP cyclic load capability.
8.1.2. Characteristics of the power grid.
The following information shall be provided:
short circuit current in the NPP circuits;
reliability of auxiliary power supply for the NPP in case of any failures of in-house power supply sources;
sufficiency of demand management capacity of the system for operation in the basic mode, possibility to limit the capacity of any other generation sources except for the NPP; besides it shall be specified when the need for the NPP power limitation can occur in the power grid (with what rate and
for what time);
the possibility for manual and automatic frequency control in the system in case of any system accidents;
the possibility for automatic or manual isolation of the NPP from the power grid with switching to the auxiliary power supply mode;
permissible unit power of a single NPP power unit on condition of maintenance of the power grid stability in case of its automatic or manual outage;
the possibility for the NPP separation for balanced load in case of any system accidents;
types and intensity of disturbances in the power grid operation;
number of power transmission lines and the possibility for full NPP power output in case of any disturbances in the power grid;
sufficiency of the grid power to ensure self-starting of auxiliary mechanisms in case of complete NPP load rejection (if self-starting mode is provided in the NPP design);
type of the turbine generator excitation system on condition of the power grid stability maintenance;
the possibility to receive voltage from the NPP auxiliary power supply system in case of any external natural impacts and directly after them;
impact of the power grid on the NPP operation;
reliability indicators of the power grid in the form of dependence (in tabular or graphical format) between the frequency of power losses in the grid resulting in the NPP power supply loss and their duration (the power grid reliability indicators shall be determined with due regard for external natural and human-induced impacts typical for the NPP site location area and characterized in Chapter 2 of the NPP SAR);
comparison with permissible number of disturbances for the main NPP equipment (reactor, turbine, generator);
analysis of impact of various disturbances on the NPP safety (the following types of disturbances shall be considered: complete blackout with loss of connections to the external power grid; frequency deviations; three-, two- and single-phase short circuits; voltage variations; synchronous and asynchronous swings in the power grid; asynchronous swings in case of any failure of the asynchronous mode elimination automatics);
8.2. Main electric circuit diagram.
8.2.1. General description.
Information on compliance with the requirements of the regulatory documents and substantiation of the diagram for connection of turbine generators to the grid from the viewpoint of reliable NPP auxiliary power supply assurance shall be provided.
The primary switching diagram shall be presented.
Fire protection means shall be listed.
Protection circuits and setpoints shall be provided for the power transmission lines and other
equipment of the main circuit.
Design reliability indicators shall be specified for the main electric circuit with breakdown by types, frequency and duration of disturbances.
Disturbance associated with complete blackout of the switchgear (due to sudden outage of the NPP power unit) shall be considered.
8.2.2. Turbine generator, unit transformer and their auxiliary systems.
General information and technical characteristics of the main and auxiliary equipment shall be provided:
primary switchning electric and process diagrams;
fire and explosion protection;
secondary switching diagrams with protection setpoints.
8.2.3. Fire safety of the main circuit equipment.
Impact of fire hazard of the main circuit equipment on the NPP safety shall be analyzed. Description of the fire extinguishing system with the scheme description and calculations shall be provided.
8.2.4. Main circuit control rooms.
Description of the main circuit control rooms with measurement and alarm systems shall be presented. Their survivability and habitability shall be substantiated.
8.3. Auxiliary power supply system of the NPP.
8.3.1. Normal operation auxiliary power supply system.
8.3.1.1. AC/DC auxiliary electrical power supply of the NPP.
Information on the working and redundant power supply sources located at the NPP site and outside it as well as their quantitative reliability assessment shall be provided. Independence of the power supply sources for the consumers ensuring integrity of the main equipment, fire safety and the NPP power unit start-up and shutdown shall be substantiated.
Technical characteristics of the equipment, devices, cables and buses shall be specified. Their compliance with the requirements of regulatory documents shall be substantiated. The primary switching diagrams shall be presented.
8.3.1.2. Calculations of short circuit currents and line-to-earth faults in the networks with insulated neutral.
Results of calculations for selection of electrical equipment, devices, buses and cables, calculations of the protection parameters and parameters of automatic devices, the possibility for self-starting of the auxiliary NPP power unit consumers (if the self-starting mode is provided in the NPP design) as well as diagrams of protection, automation and other secondary switching circuits shall be presented.
8.3.1.3. Setpoint selection justification.
Selection of the setpoints for ASB activation and automatic devices for switching of the emergency
power supply network as well as the reliable normal operation power supply network (if any is provided in the NPP design) to independent sources as well as the possibility for safe operation of the turbine generators for BOP needs in the thermal and mechanical run-out mode with frequency and voltage parameters below the permissible values shall be substantiated.
8.3.1.4. Layout plans of the system components.
Layout plans shall be provided for the equipment, devices and cables.
8.3.1.5. Overvoltage protection.
Information on any potential overvoltage and the relevant protection shall be specified.
8.3.1.6. Fire safety assurance.
Any possible causes of fire breakouts in the electrical section of the NPP, fire propagation paths and their impacts on the NPP safety shall be analyzed.
Information on fire safety assurance with indication of the data on the automatic fire detection and extinguishing systems as well as the results of calculations substantiating fire safety assurance shall be provided.
8.3.1.7. Protection against erroneous actions of the personnel.
Protection of the electrical NPP section against any unintended erroneous actions of the personnel (impossibility to activate any equipment with disabled protections and interlocks; availability of automatic devices aimed to change the logic of protections and interlocks in case of any individual equipment withdrawal from services; automatic control of correct assembly of electric and process circuits; impossibility to disable protections and interlocks without the relevant automatic change of operation modes for the main and auxiliary equipment) shall be substantiated.
8.3.1.8. Monitoring and control.
Information on the control rooms, controlled parameters, alarm types, instrumentation classes, sensors, measuring transformers, metrological control and protection against external and internal interference shall be provided.
8.3.1.9. Reliability analysis.
Results of the quantitative reliability analysis for the NPP auxiliary power supply at all voltage levels shall be provided and its acceptability to ensure the design NPP safety level, survivability of the control rooms in case of any accidents and external impacts shall be confirmed.
8.3.2. Emergency power supply system.
The requirements of this section shall be applicable to the EPSS description and description of the reliable normal operation power supply system (if any is provided in the NPP design) as well as to description of emergency power supply sources included into special-purpose hardware for BDBA management.
8.3.2.1. Characteristics of the consumers.
The list of auxiliary consumers requiring power supply from independent sources in case of loss of power supply from normal operation sources shall be provided with indication of the following permissible values for each consumer:
duration of power supply interruption;
quantitative characteristics of power supply reliability;
voltage and current frequency decrease (increase) with indication of permissible duration;
current waveform changes and permissible duration of such changes;
period over which repeated voltage supply to the consumer is possible and any other requirements imposed by process and control systems.
Information on the certificate data for each consumer shall be provided with indication of the period within which it shall function without any power supply from normal operation sources.
Redundancy principle shall be described for the systems supplied from the EPSS.
Information on the requirements for fire safety, fire and explosion protection of the equipment and devices and fire resistance rating of the EPSS structures and electric equipment of safety systems shall be provided.
Information on the operation conditions for electrical equipment, devices and cables of safety systems and the EPSS in the course of normal operation and in case of any abnormal operation (including accidents) shall be provided with regard to temperature, humidity, pressure, radiation exposure and any other external impacts with indication of the exposure period.
8.3.2.2. Technical characteristics of the EPSS.
The following information shall be provided:
configuration of the system;
electric primary switching diagram of the system with substantiation of its selection;
boundaries of the system;
sufficiency substantiation for the selected number of EPSS channels;
sufficiency substantiation for the continuous functioning time of power supply sources;
current voltage and frequency setpoints for the SDGS start;
substantiation of setpoints and selected SDGS warming time for load acceptance from the moment of the relevant signal injection;
the SDGS switch-on and load increase technique and its substantiation;
information on any operator's intervention prohibition with indication and substantiation of the prohibition period;
description of the algorithm for switching of the EPSS consumers to independent power supply sources;
description of the SDGS starting algorithm based on the process parameters of the NPP and substantiation of selection of these parameters;
technical characteristics of the current sources: their rated and maximum power, permissible duration of continuous operation, current voltage and frequency stability, any possible deviations from
the sinusoidal current waveform;
certificate data or technical characteristics of the equipment, buses, cables, devices, leak-tight penetrations used in the EPSS;
calculation results for short circuit currents and line-to-earth fault currents in the networks with insulated neutral, selection of the electrical equipment, devices, buses and cables;
potential overvoltage levels and the relevant protection;
substantiation of the neutral mode selection (earthed or non-earthed);
confirmation of the system protection against human errors in the course of its actuation (impossibility of switch-on without activation of the relevant protections and automatic devices, automatic control of correct assembly of electric and process circuits);
layout plans for the EPSS equipment, devices and cables;
fire protection substantiation with the results of calculation of the maximum temperatures that the envelope, load-bearing and localizing structures can reach in case of complete combustion of flammable substances in a single cable room or separate equipment box; results of calculations confirming sufficient strength of these structures at such temperatures and impossibility of further fire propagation.
8.3.2.3. Protection against short circuit currents and faults.
Information on protection against short circuit currents and earth-faults in the networks with insulated neutral shall be provided. Information on automatics and process protections of diesels shall be provided, the protection types, their purposes and action zones, technical characteristics and protection redundancy rates shall be specified. Calculations for selection of protections and their setpoints shall be presented. Requirements for reliability of internal protections of electrical equipment, cables and diesels with indication of their activation priorities in relation to performance of the EPSS safety functions shall be specified. Diagrams of protections, automation and other secondary switching circuits shall be provided.
8.3.2.4. Monitoring, control and automation.
Information on the control rooms, their survivability in case of any abnormal operation including accidents and external impacts shall be provided.
Information on the controlled parameters, alarm types, classes of instrumentation, sensors and measuring transformers shall be specified. Information on metrological control shall be presented.
8.3.2.5. Possibility for testing and maintenance.
The following information shall be provided:
automatic diagnostic monitoring of the systems and components;
frequency of testing, testing methods and programs, controlled parameters;
the possibility to perform tests on the operating equipment or with shutdown;
types and time limits for maintenance of the switching equipment, protection and automation cables;
operability recovery techniques;
time limits for replacement of equipment and cables with exhausted lifetime;
accessibility for maintenance and testing with due regard for the radiation situation and ambient conditions.
8.3.2.6. Criteria for selection of the power supply source capacity.
The following information shall be provided:
calculations of loads on transformers, diesel-generators, power supply lines, invertors and batteries, charging devices;
coordination of the source capacity with design loads;
coordination of the load characteristics (active, reactive) with characteristics of sources;
permissible voltage and frequency variations, deviations from harmonicity, inrush currents and asynchronous ASB currents;
characteristics of batteries with confirmation of their compliance with the requirements of the consumers;
substantiation of the battery operation time in autonomous mode without any charging;
characteristics of the charging devices;
electromagnetic compatibility of the sources, consumers, protections and automatic devices;
substantiation of the continuous operation duration for any sources with limitations imposed by fuel stocks.
8.3.2.7. Location, protective grounding, lightning protection, fire protection.
Information on physical separation of the EPSS channels, switchgear rooms, sources and cable routes as well as their protection against external impacts of natural and human-induced origin shall be provided. Sufficiency of physical separation for the EPSS channels shall be assessed.
The following information shall be provided:
lightning protection and protection against secondary lightning effects;
protective grounding;
fire alarm and fire extinguishing;
provision of climatic conditions (temperature, humidity);
protection of equipment, cables and leak-tight penetrations against any missiles in case of any breakage of process equipment and pipelines and also against water jets;
accessibility for the equipment maintenance with due regard for radiation hazard conditions, permissible time of the operating personnel stay, control of access to the equipment, possibility for immediate access in case of necessity.
8.3.2.8. Criteria for selection of equipment, cables and leak-tight penetrations.
The following information shall be provided for the EPSS components:
ambient conditions;
seismic resistance;
equipment resistance to short circuit currents, heat resistance of cables, heat resistance of cables in case of short circuit current tripping with stand-by protections and after repeated voltage supply to non-eliminated short circuit;
dust and moisture protection;
start and self-start assurance;
thermal class of insulation;
insulation class with regard to contaminations;
service life, possibility for recovery and replacement;
resistance to external and internal impacts;
fire protection.
8.3.2.9. Compliance with the regulatory requirements.
Information on the solutions adopted to ensure compliance with the regulatory requirements shall be provided.
Information on compliance with the following requirements shall be specified:
single failure principle;
protection against external and internal impacts;
independence of switchgear and cable routes (high and low voltage, between cable routes of the EPSS channels, between the EPSS cable routes and normal operation cable routes);
protection against common cause failures;
independence of the connections;
possibility for testing and maintenance, lifetime monitoring;
distinctive labelling of equipment and cables;
completion of protective actions;
absence of any adverse impact on reliable performance of safety functions in case of multi-purpose use.
8.3.3. Fire protection of cable systems.
8.3.3.1. Applied types of cables.
The following information shall be provided:
conditions for flammability, fire resistance, flame retardance, smoke emission and toxicity;
conditions for flame retardance of separate cables and cable bundles.
8.3.3.2. Cable laying techniques in the areas with various hazard degrees.
The areas where cables are laid shall be characterized in terms of their explosion hazard and risk of fire and mechanical damages.
8.3.3.3. Passive protection.
The following information shall be provided:
fire-resistant protective structures;
fire partitions restricting fire propagation through walls and floors as well as at long cable routes;
measures to reduce fire hazard of cable routes laid in the same fire zone.
8.3.3.4. Active protection.
The following information shall be provided:
fire alarm system;
automatic fire extinguishing systems;
any other fire extinguishing systems.
8.3.3.5. Overheating protection in case of overloads.
Heat and fire resistance in case of any overloads shall be substantiated.
8.3.3.6. Protection against external and internal impacts.
Information on any technical solutions for protection against external and internal impacts shall be provided.
8.4. Operation.
8.4.1. Operation guidelines.
The following general provisions of the operation guidelines for power supply systems shall be specified:
the procedure for works and commutations in order to activate individual equipment and systems and to take them out of service for repair;
the procedure for testing of individual equipment and systems in general;
quality control for fuel and oil, time limits, criteria and procedure for their replacement;
frequency and procedure for testing and inspections of the equipment and rooms of the systems.
8.4.2. Maintenance and repair.
The following information shall be provided:
scope and frequency of the equipment repair, protection and automation checks;
time limits and procedure for replacement of the equipment with exhausted lifetime;
frequency and scope of verification for measuring tools.
8.4.3. Commissioning.
Information on the adjustment and testing programs for individual equipment, devices and systems in general shall be provided. Information on the scope of protection and automation verification in the course of commissioning shall be specified.
8.5. Communication systems.
Information on the communication systems (in-plant, with external facilities, with protected emergency response control posts, with the operating organization; the system for recording of the controlling personnel's actions in any pre-accident situations; the closed circuit TV system and any other communication systems provided in the NPP design) shall be presented. Information on the purpose and configuration of the communication systems, power supply diagram, the communication equipment layout plan, analysis of the communication system operation stability during design basis and beyond design basis accidents as well as under any external impacts shall be provided.
8.6. Regulatory requirements.
The list of regulatory documents containing the requirements that have been taken into account in the design of power supply systems shall be provided.
IX. Requirements for the content of Chapter 9 "Auxiliary systems of the NPP power unit"
Information on safety-related auxiliary systems of the NPP power unit not considered in any other chapters of the NPP SAR shall be provided in Chapter 9 of the NPP SAR.
9.1. Set of nuclear fuel storage and handling systems.
Configuration of the set of NF storage and handling systems shall be provided:
fresh (non-irradiated) nuclear fuel storage and management system;
nuclear core refueling system;
NF handling system consisting of:
near-reactor SNF storage systems;
systems for the SNF storage in the storage facilities outside the reactor hall (if any);
shielding chamber (if any);
in-plant NF transportation system.
Each system from the set of NF storage and handling systems shall be described in accordance with the standard system description framework given in Appendix 4 hereto.
Besides additional information required in accordance with this chapter as well as information specific for a particular system shall be provided for each system under consideration.
Issues related to NF transportation within the NPP territory from acceptance of a vehicle with fresh nuclear fuel to shipment of spent nuclear fuel shall be considered.
Information on organization of the NM accounting and control at the NPP shall be provided.
In case of any reactor core refurbishment associated with usage of new fuel type, new types of fuel assemblies or in case of any change of permissible burnout depth the revised list of design basis accidents and the list of beyond design basis accidents in the course of NF handling as considered in the design shall be presented with due regard for the peculiarities of the new fuel types that shall be analyzed in Chapter 15 of the NPP SAR.
Description of hoisting cranes shall contain the information on compliance with the requirements of federal rules and regulations in the field of atomic energy use governing arrangement and safe operation of hoisting cranes for nuclear facilities.
The following information shall be provided:
maintenance and (or) recovery of operability for the cranes and their components under any internal and external impacts;
stability of the cranes against tumbling, shifting (displacement) along and across the rails and separation from the rails under the impact of seismic loads;
substantiation (if necessary) of absence of any devices preventing the crane break-off;
substantiation of the crane classification as infrequently-used.
9.1.1. Fresh (non-irradiated) nuclear fuel storage and handling system.
9.1.1.1. System design.
The following information shall be presented for each fresh fuel storage facility:
maximum design capacity of the storage facility;
storage norms;
characteristics of the fresh fuel intended for storage;
distinctive signs characterizing fuel enrichment in fuel assemblies and techniques for their identification - visual and (or) through the use of refueling devices;
distinctive signs for fuel assemblies with burnable poison, mixed fuel (if any) and techniques for their identification.
Information on the interior layout of the storage facility shall be provided, class of the storage facility in accordance with federal rules and regulations in the field of atomic energy use governing the requirements for safe storage and transportation of nuclear fuel at nuclear facilities and the ambient storage parameters as well as compliance of the storage facility with the requirements of federal rules and regulations in the field of atomic energy use shall be specified.
Information on the initiating events of design basis accidents the system is designed for shall be presented. Combination of loads for calculations shall be specified.
Lists of procedures and programs used to substantiate safety of NF storage and transportation shall be presented, their application scope as well as information on verification and validation of the programs in accordance with the established procedures shall be specified.
The list of parameters, systems included into the fresh (non-irradiated) NF storage and handling system, and components of the system ensuring its safe functioning as defined in the design shall be provided.
In case of any reactor core refurbishment associated with usage of new fuel type, new types of fuel assemblies or in case of any change of permissible burnout depth and necessity for its storage the possibility to use the existing fresh NF storage facilities for this purpose shall be confirmed, or materials of the project for fresh fuel storage facilities refurbishment and also potential modification of individual sections of transportation corridors and individual components of transport and handling equipment shall be presented.
It shall be substantiated that layout of the rooms and design solutions eliminate any possibility of flooding with water and ingress of any other neutron-moderating materials into the fresh fuel storage areas; free evacuation of the personnel from the rooms is ensured in case of any accidents (accident type, evacuation routes, evacuation time calculations); no routes to any other operational rooms pass through the fresh fuel storage facilities (the system of access and access control shall be described).
The following information shall be provided:
techniques and methods to comply with the prohibition for transportation of any cargoes (except for the parts of hoisting and refueling devices) above the stored fuel in the course of refueling or location of any cargoes above the storage facilities covered with any structures; confirmation of the capability of these structures to withstand dynamic and static loads occurring in the course of cargo transportation or location;
information on division of the FFS buildings and rooms into the controlled access area and the uncontrolled access area;
information on classification of the FFS rooms into categories according to radiation and fire safety and information on the FFS rooms where the radiation situation can charge drastically in the course of process operations;
information on compliance with the principle for separate ventilation in the FFS rooms of the controlled access area and the uncontrolled access area as well as absence of any common ventilation air ducts in the rooms with different service categories;
information on installation of leak-tight doors at all emergency (fire) entrances and exits from the controlled access area;
it shall be substantiated that the design of the storage facility enables to decontaminate any surfaces easily in case of necessity and the surfaces in the controlled access area rooms are protected with materials with poor RS absorption that are easy to decontaminate.
The following information on the fresh NF storage system components shall be provided:
configuration of the fuel storage and handling system components with brief description of their design: components used for fuel storage, transportation, handling and canting operations, for depreservation, inspection (incoming control) and repair of fuel assemblies (if any);
description of the OSTP (TP) maintenance systems (if any are present in the FFS).
In case of any reactor core refurbishment associated with usage of new fuel type, new types of fuel assemblies or in case of any change of permissible burnout depth and necessity for its storage in OSTPs (TPs) in the FFS applicability of the existing OSTSs (TPs) for this purpose shall be confirmed, or materials of the new OSTP (TP) design ensuring non-exceedance of the standard radiation burden values on its surface shall be provided, and information on the radiological monitoring measures and
maintenance conditions for transport packings with such fuel shall be specified.
Information on any other equipment and materials located in the FFS shall be provided.
The following information shall be provided:
techniques and methods to comply with the prohibition for storage of flammable materials as well as any materials not included into packaging sets or having any other hazardous properties in case of a fire in the FFS;
the list of the nuclear core components different from the nuclear fuel (in case they are stored in the FFS) and regulation of their locations in the design;
techniques and methods to comply with the prohibition for storage of any efficient neutron-moderating materials between or inside casings, between the racks or packaging groups.
The following information on the systems and components related to functioning of the set of fresh fuel storage and handling systems as well as any associated systems and components performing independent functions shall be provided:
information on location of each system, its components, redundancy, specified service life, working media, parameters; information shall contain the parameters corresponding to the functional purpose of the described system; the parameter values shall be specified with indication of potential dispersion (tolerance);
confinement devices (if any) intended to prevent or limit propagation of RS and ionizing radiation generated in case of accidents inside the storage facility and their release to the environment;
information on the alarm system in case of self-sustaining chain reaction occurrence (if any);
information on the fire alarm system;
information on the operational and emergency lighting system;
information on the closed circuit TV system (if any);
information on the ventilation systems;
information on the drainage systems (if any);
information on the communication systems;
information on decontamination systems;
information on the storage facility heating system.
In case of any reactor core refurbishment associated with usage of new fuel type, new types of fuel assemblies or in case of any change of permissible burnout depth and necessity for its storage in the existing fresh fuel storage facilities sufficiency of the existing systems related to the fresh NF storage system functioning shall be confirmed, or refurbishment of such systems shall be substantiated.
9.1.1.2. Materials.
The following information shall be presented in substantiation of safety for selection of the materials applied in the FFS system components as well as any welding materials with due regard for normal operation conditions and any abnormal operation including accidents:
mechanical and process characteristics of the materials with references to technical specifications and standards;
permits for use of the materials if required in accordance with federal rules and regulations in the field of atomic energy use;
permits for use of non-metal materials (if any) if required in accordance with federal rules and regulations in the field of atomic energy use;
resistance of the materials to the conditions arising in the course of operation, decontamination and abnormal operation including accidents;
information on stability of absorbing additives in the FFS structural materials (if any) under any conditions arising in the course of operation, decontamination and abnormal operation including accidents;
compliance with the requirements for fire resistance or slow burning of lining, finishing, sound-absorbing, sound- and heat-insulating materials used for interior finishing of the FFS;
confirmation of the fact that the FFS envelope structures are made of non-combustible materials and have the fire resistance rating in accordance with the requirements;
confirmation of the fact that surfaces of the FFS rooms and the FFS equipment is protected with moisture-proof materials poorly absorbing radioactive substances and easy to decontaminate;
information on any hazardous properties of the materials used and stored in the FFS (if any) an potential manifestations of such properties in the course of normal operation and in case of any abnormal operation including accidents.
9.1.1.3. System reliability analysis.
Quantitative reliability indicators shall be provided for the FFS components referred to safety classes 1 and 2.
Qualitative reliability analysis shall be provided and quantitative values of the reliability parameters shall be determined for the transport and handling scheme of fresh fuel acceptance and supply.
9.1.1.4. Assessment of the fresh NF storage project.
Techniques and methods for determination of the permissible number of packages or casings in a group or stack shall be substantiated.
9.1.2. Core refueling system.
9.1.2.1. Design basis.
Information complying with the requirements specified in item 9.1.1.1 of this Appendix shall be presented for the core refueling system; information on description and substantiation of any operations with nuclear fuel in the reactor building shall be also provided.
In case of any reactor core refurbishment associated with usage of new fuel type, new types of fuel assemblies or in case of any change of permissible burnout depth the possibility to use the existing core refueling systems for refueling shall be confirmed, or refurbishment of the process equipment or individual components of the process equipment shall be substantiated.
9.1.2.2. Description of the refueling system.
Information on the flow diagram of refueling operations with indication of the equipment, devices and components performing independent functions shall be provided. Particular system equipment configuration shall be specified.
The design flow diagram of handling operations in case of unloading of the nuclear core and its components shall be provided, and its distinctions from the refueling pattern as well as any special-purpose equipment shall be specified.
Information on any measures aimed to eliminate errors in the course of the core loading (refueling) shall be presented.
The following information shall be provided:
techniques and methods used to identify unloaded fuel assemblies and (or) core components for compliance with the refueling plan;
the selected refueling technique and its substantiation;
the reloading box state in the course of refueling;
system and design of the assembly for the core components loading into the reactor;
frequency, scope and procedure for the refueling and substantiation thereof;
engineering features provided in the NPP design in order to prevent accidental ingress of any foreign objects into the reactor during refueling and repair works;
configuration of the refueling system with its sufficiency substantiation and also with indication of the requirements for this system ensuring safe handling of fuel assemblies under normal operation conditions as well as in case of any failures and damages;
engineering features ensuring heat removal from the reloaded fuel assemblies.
Besides the following information shall be provided:
measures to prevent damage, deformation, breakage or falling of fuel assemblies;
measures to prevent application of impermissible forces to the fuel assemblies in the course of their withdrawal or installation;
engineering features aimed to prevent falling of fuel assemblies in case of power supply loss;
provided protective features ensuring movement of refueling devices within the permissible limits;
equipment provided in the technical design for reliable transportation of fuel to safe locations in case of any failure or disturbances of safe operation conditions for the refueling devices;
consoles (panels) provided in the refueling devices in order to display information on the position (state) and orientation of the fuel assemblies and grips.
It shall be specified that all loads occurring under normal operation conditions as well as asymmetric loads and loads caused by acceleration are taken into account in the design of the refueling equipment; in this case it shall be substantiated that any stresses arising due to impact of these loads do not exceed the acceptable limits for various equipment components.
Operability of the refueling system shall be substantiated.
Information on any systems related to functioning of the core refueling system, brief information on location of each system, configuration of its equipment, redundancy, expected service life, working media and parameters shall be provided.
Information on the following systems shall be specified:
closed circuit television for the refueling monitoring with the list of refueling operations controlled through the use of closed circuit TV;
fuel cladding integrity monitoring,
operating and emergency lighting;
fire extinguishing;
ventilation and air purification;
communication and warning;
emergency alarm.
9.1.2.3. Materials.
Information on the materials used shall be provided. Description shall be given in accordance with item 9.1.1.2 of this Appendix.
9.1.3. Set of systems for handling of spent (irradiated) fuel.
9.1.3.1. Near-reactor SNF storage system.
9.1.3.1.1. System design.
Information complying with the requirements stated in item 9.1.1.1 of this Appendix with regard to the near-reactor SNF storage system shall be provided.
In case of any reactor core refurbishment associated with usage of new fuel type, new types of fuel assemblies or in case of any change of permissible burnout depth and necessity for storage of such SNF the possibility to use the existing SNF storage facilities for this purpose shall be confirmed, or refurbishment of SNF storage facilities and also potential modification of transport and handling equipment components shall be substantiated.
Information on the maximum design heat removal capacity in the spent fuel pool, parameters of the storage environment and the SNF storage norms shall be specified for the near-reactor SNF storage system. It shall be substantiated that the SFP capacity allows holding of nuclear fuel in order to reduce radioactivity and heat emissions and also arrangement of the conditions for unloading of one complete nuclear core at any moment of operation.
Characteristics of nuclear fuel planned for storage shall be specified.
Information on fresh fuel and any other elements on temporary or long-term storage in the near-reactor SNF storage facilities shall be provided with indication of the reasons, time limits and norms of storage as well as the properties of these elements.
It shall be substantiated that:
the possibility for detection of the cooling medium leaks and their locations is provided in the SFP design, and the design of the spent fuel pool enables to eliminate these leaks; systems for collection of radioactive water leakages in controlled water discharge sumps are arranged in the spent fuel pool;
the possibility for irradiated NF cooling in case of design basis and beyond design basis accidents is ensured;
Design of the equipment used for the SNF location and storage, for handling of leaky fuel assemblies, as well as the equipment for storage of any other core components (if any) shall be described.
In case of SNF storage in the existing near-reactor storage facilities administrative and technical measures for storage of damaged and leaky fuel assemblies with such fuel shall be provided. It shall be specified that the following is provided in the near-reactor SNF storage facilities:
the technique for handling of leaky fuel assemblies, conditions and limits for their safe storage;
leakiness criteria for fuel assemblies requiring use of special-purpose casings and other equipment and implementation of measures in order to prevent propagation of fission products to the cooling medium in exceedance of the acceptable values;
devices in the design of casings enabling to remove highly active cooling medium from the casings in case of necessity without any mixing with the cooling medium in the spent fuel pool;
the possibility for lighting of the interior spent fuel pool space, in this case materials of the equipment used for lighting shall be corrosion-resistant under the SFP medium impact and shall prevent contamination of the medium;
filtration equipment of the ventilation system designed and operated in such a way so that to limit any potential releases of radioactive substances;
the ventilation system which shall provide dilution and safe removal of hydrogen generated in the course of water radiolysis;
measures aimed to stop all regular transportation works in case of dropping of nuclear fuel, leak-tight bottles, casings to the SFP bottom prior to their withdrawal;
equipment for withdrawal of dropped nuclear fuel, leak-tight bottles or casings without the SFP drying and complete unloading of nuclear fuel.
Information on any systems related to functioning of the SNF storage and handling system shall be provided. Information on location of each system, configuration of its equipment, redundancy, expected service life, media, and parameters shall be provided. Parameters corresponding to the functional purpose of the described system shall be specified.
Information on the following systems shall be specified:
LSSs intended to prevent or limit propagation of radioactive substances and ionizing radiation generated in case of accidents inside the storage facility and their release to the environment;
coolant;
SFP filling and emptying;
SFP makeup;
imtermediate cooling circuit;
ventilation and air purification;
process control;
fire extinguishing;
communication and warning;
emergency alarm (if any).
References to other sections of the NPP SAR chapters where safety analysis for the above-mentioned systems is presented shall be given in description of the functions of these systems and substantiation of their operability.
In case of any reactor core refurbishment associated with usage of new fuel type, new types of fuel assemblies or in case of any change of permissible burnout depth and necessity for storage of such SNF sufficiency of the existing systems supporting the SNF storage system functioning shall be confirmed, or design materials for refurbishment of such systems shall be presented.
9.1.3.1.2. Materials.
Materials shall be described in accordance with item 9.1.1.2 of this Appendix.
9.1.3.1.3. System reliability analysis.
Requirements of item 9.1.1.3 of this Appendix shall be taken into account in presentation of the reliability analysis with regard to the near-reactor SNF storage system.
9.1.3.2. System for SNF storage in water or any other cooling medium in the spent fuel pool located outside the reactor hall in the SNFS specially built for this purpose (if any).
9.1.3.2.1. System design.
Information shall be provided in accordance with items 9.1.1.1 and 9.1.3.1 of this Appendix with regard to the spent fuel pool located outside the reactor hall in the storage facility (SNFS) specially built for this purpose.
Information on any systems related to the SNFS complex functioning shall be specified. Information on location of each system, configuration of its equipment, redundancy, expected service life, working media and parameters shall be provided. Parameters corresponding to the functional purpose of the described system shall be specified.
Information on the following systems shall be specified:
LSSs intended to prevent or limit propagation of radioactive substances and ionizing radiation generated in case of accidents inside the storage facility and their release to the environment;
water cooling (except for the cases when it is confirmed that exceedance of the design water temperature values in the storage facility is prevented even without any special-purpose cooling);
water treatment;
filling and emptying (drainage system) of the spent fuel pool;
SFP makeup;
water supply;
collection of radioactive water leaks in controlled water discharge sumps (leakage recycling);
emergency makeup of the spent fuel pool;
ventilation and air purification;
underwater lighting;
fire extinguishing;
communication and warning;
emergency alarm (if any);
power supply.
In case of any reactor core refurbishment associated with usage of new fuel type, new types of fuel assemblies or in case of any change of permissible burnout depth and necessity for storage of such SNF in the spent fuel pool located outside the reactor hall in the existing SNFS possibility for such storage shall be confirmed, or refurbishment of the SNFS SFP and the systems related to functioning of the SNF storage system as well as potential modification of the transportation and handling equipment components shall be substantiated.
9.1.3.2.2. Materials.
Materials shall be described in accordance with item 9.1.1.2 of this Appendix.
9.1.3.2.3. System reliability analysis.
Information complying with the requirements of item 9.1.1.3 of this Appendix shall be provided with regard to the system for SNF storage in water or any other cooling medium in the spent fuel pool located outside the reactor hall in the storage facility (SNFS) specially built for this purpose.
9.1.3.3. System for "dry" SNF storage in the storage facility located outside the reactor building in the building specially built for this purpose (if any).
9.1.3.3.1. System design.
Information shall be provided in accordance with items 9.1.1.1 and 9.1.1.3 of this Appendix with regard to "dry" storage.
Information on any systems related to the "dry" SNF storage complex functioning shall be specified. Information on location of each system, configuration of its equipment, redundancy, expected service life, media, and parameters shall be provided. Parameters corresponding to the functional purpose of the described system shall be specified.
Information on the following systems shall be specified:
localizing safety systems intended to control and limit accumulation of RS in the storage facility atmosphere and release of RS and ionizing radiation generated in case of accidents to the environment;
providing heat removal from OSTPs with due regard for non-exceedance of the design temperature on the exterior OSTP surface;
temperature control;
control of water ingress into OSTPs;
ventilation;
radiological monitoring;
fire extinguishing;
communication and warning;
emergency alarm;
power supply of the systems and supporting devices.
9.1.3.3.2. Materials.
Materials shall be described in accordance with item 9.1.1.2 of this Appendix.
9.1.3.3.3. System reliability analysis.
Information complying with the provisions of item 9.1.1.3 of this Appendix shall be specified with regard to the system for "dry" SNF storage in the storage facility located outside the reactor building in the building (SNFS) specially built for this purpose (if any).
9.1.3.4. Shielding chamber system.
9.1.3.4.1. System design.
Information complying with the provisions of item 9.1.1.1 of this Appendix shall be specified with regard to the shielding chamber system.
In case the shielding chamber is used to handle new SNF type the design materials shall be presented for the shielding chamber intended for this purpose, or refurbishment of the existing shielding chamber equipment shall be substantiated.
9.1.3.4.2. Description of process flow diagram.
Description of the process flow diagram as well as the following information shall be provided:
information on arrangement of access to the shielding chamber rooms;
substantiation of compliance with the sanitary requirements;
information on the SNF management zones in the shielding chamber system where the radiation situation can change in the course of process operations.
Information on any systems related to the shielding chamber system functioning shall be specified. Brief information on location of each system, configuration of its equipment, redundancy, expected service life, working media and parameters shall be provided. Parameters corresponding to the functional purpose of the described system shall be specified.
Information on the following systems shall be specified:
localizing safety systems intended to prevent or limit releases of RS and ionizing radiation generated in the course of process operations and (or) in case of accidents to the environment;
ventilation and air purification;
lighting (operating and emergency);
independent radioactive drain system;
decontamination of the complex;
gas supply;
vacuuming;
power supply of the systems and supporting devices;
fire extinguishing;
communication and warning;
emergency alarm;
collection and storage of nuclear materials in the form of spills;
temporary storage of spills.
9.1.3.4.3. Materials.
Materials shall be described in accordance with item 9.1.1.2 of this Appendix.
This item shall be arranged in accordance with item 9.1.1.1 of this Appendix with regard to the system of in-plant NF transportation within the NPP territory.
9.1.4.2. Description of the system.
Information on the parking area for the transportation vehicle and location of in-plant railway lines for NF transportation, methods and scope of incoming control for containers with nuclear fuel, methods for transfer of unloaded nuclear fuel from the train to the storage facility, the scheme of NF transportation within the NPP site, methods for NF transportation to the power units through the use of in-plant transport containers and special-purpose vehicles shall be presented.
Information on the transportation and handling equipment used for NF transportation shall be provided.
In case of any need for in-plant transportation of new fuel types and new types of fuel assemblies or in case of any change of permissible burnout depth the possibility for its transportation in the existing OSTPs shall be confirmed, or materials on improvement of the OSTP design for transportation of new fuel types shall be presented, and information on the radiological monitoring measures and transportation conditions for new fuel types by special-purpose vehicles shall be provided.
9.1.5. Arrangement of NM accounting and control at the NPP.
Information on arrangement of NM accounting and control at the NPP with indication of the data on NF identification, places of installation (stacking), registration of the time of receipt in the storage facility and transfer to the reactor building, maintenance of maps and other recording documentation as well as distribution of responsibility for accounting shall be presented.
It shall be specified that the accounting and control procedures for fissionable nuclear materials provide reliable information on the quantity and location of nuclear fuel, timely detection of any losses
and unauthorized use or theft.
The following information shall be provided:
description of the structure of NM balance areas and key points for measurement of the inventory and flows of nuclear materials with regard to the FFS;
classification of fissionable nuclear materials into categories;
description of the registration procedures for any changes of the fissionable NM inventory;
description of the procedure for delivery of fissionable NM into the MBA and removal from it with regard to the FFS;
maintenance of accounting records and operational records for the MBAs and key measurement points;
description of the NM physical inventory arrangement;
description of the procedure for development of reports on MBAs;
confirmation of compliance with the regulatory requirements for NM accounting and control.
9.2. Systems with process water medium.
Description of the systems with process water medium used at the power unit shall be presented. Each system shall be described in accordance with the standard system description framework given in Appendix 4 hereto.
Issues related to leaks and accumulation of radioactive substances as well as the possibility for flooding of the NPP site and safety-related systems (components) in case of any pipeline ruptures in this system shall be considered in the analysis of the circulation water supply system.
9.3. Fire protection systems.
Information on the fire protection systems (not described in Chapters 7 and 12 of the NPP SAR) of the rooms where safety-related systems are located shall be presented.
Each system shall be described in accordance with the standard system description framework given in Appendix 4 hereto.
9.4. Physical protection system.
9.4.1. General information on the PPS arrangement and functioning.
The following information shall be provided:
arrangement of the security service responsible for organization and assurance of physical protection at the nuclear facility where the NPP power units are located (hereinafter - the nuclear facility) in accordance with the provision on security services effective at the nuclear facility;
definition of threats for the nuclear facility and adversary models in the physical protection documentation existing at the nuclear facility in accordance with the requirements of Decree of the Government of the Russian Federation No. 456 dated July, 19, 2007 "On approval of the physical protection rules for nuclear materials, nuclear installations and nuclear material storage facilities" (Collected legislation of the Russian Federation, 2007, No. 31, Art. 4081; 2009, No. 18, Art. 2248; 2010, No. 38, Art. 4825; 2011, No. 7, Art. 979; No. 21, Art. 2961; 2012, No. 36, Art. 4914; 2013, No. 8, Art. 831;
2014, No. 8, Art. 820; No. 12, Art. 1288) and federal rules and regulations in the field of atomic energy use governing the requirements for physical protection systems of nuclear materials, nuclear installations and NM storage facilities;
definition of NM categories and categories of consequences of any unauthorized actions in relation to the physical protection objects (nuclear materials and vulnerable points of the NPP power unit) in the physical protection documentation existing at the nuclear facility;
definition of the protected area, interior area, critical areas, controlled access areas at the NPP as well as categories of buildings, structures and rooms where physical protection objects are located and the nuclear facility in general;
the nuclear facility vulnerability analysis with the relevant vulnerability analysis report;
assessment of the PPS efficiency for the physical protection objects at the NPP power unit with the efficiency assessment report;
establishment of the minimum permissible PPS efficiency level by the NPP management and confirmation on this level by the PPS efficiency assessment;
design of the EFPP set with the detailed documentation, equipment of the nuclear facility (NPP power units) by the EFPP set and acceptance of the EFPP set into operation;
organization of guarding for the nuclear facility and individual NPP power units in accordance with the act of the intra-departmental commission on security arrangement (for interior forces of the Ministry of Internal Affairs), the act of departmental security service and the provision on the departmental security service division (for departmental security services);
establishment of the system for arrangement of access to the NPP, its protected areas, buildings, structures and rooms and access to the physical protection objects and information on the PPS functioning in the physical protection documentation existing at the nuclear facility;
establishment of the access control at the nuclear facility in the physical protection guidelines existing at the nuclear facility;
establishment of the provision on the internal security policy at the NPP in the physical protection documentation existing at the nuclear facility;
establishment of the action plan for the physical security staff and the NPP personnel in any regular and emergency situations in the physical protection documentation existing at the nuclear facility;
establishment of the plan for interaction of the NPP management, military units (divisions) with the RF law enforcement agencies and the RF FSS bodies in any regular and emergency situations in the physical protection documentation existing at the nuclear facility;
availability of the anti-terrorism security certificate for the NPP;
provision of qualified physical security staff for the NPP including the guard force personnel and workers of the facility performing any assigned functions for physical protection assurance;
organization of the physical security personnel training and professional development in accordance with the plan existing at the nuclear facility;
availability of the permits of the competent authority for state regulation of safety in atomic energy use for the NPP management officers in order to perform any works in the field of atomic energy use with regard to physical protection assurance;
organization of the EFPP operation and compliance with the plan for the EFPP technical condition and operability verification;
availability of certificates for the physical protection equipment;
organization and performance of in-plant control of the physical protection condition;
availability of the PPS improvement plan with indication of the financing source for the works.
9.4.2. General information on the EFPP set.
General information on configuration and functioning of the EFPP set shall be provided in the following scope:
a) general information on equipment of the perimeters of protected NPP and NPP power unit areas with:
PPH;
security alarm means;
call alarm means;
optoelectronic surveillance and situation assessment means;
operational communication and notification means;
power supply and security lighting means;
information protection means;
exclusion area equipment at the perimeter of the NPP protected zone;
b) general information on equipment of the access control posts and (or) access points (passages) at the perimeters of protected areas, buildings and structures with:
security alarm means;
call alarm means;
optoelectronic surveillance and situation assessment means;
access monitoring and control means;
operational communication and notification means;
information protection means;
power supply and security lighting means;
means for protection of the guarding personnel in the access control posts and access points against any possible attack of the adversaries;
c) general information on arrangement and equipment of the central and local PPS control rooms with EFPP:
means for protection of the operators and personnel in the control rooms against any attacks of the adversaries;
information collection and processing means;
optoelectronic surveillance and situation assessment means;
security alarm means;
call alarm means;
operational communication and notification means;
power supply and security lighting means;
information protection means.
General conclusion on equipment of the NPP with the EFPP devices in order to ensure detection of adversaries, surveillance and situation assessment, detention of adversaries, access monitoring and control, physical protection control, protection of the physical security staff, response, seizure and neutralization of adversaries in accordance with the requirements for physical protection and the required PPS efficiency shall be presented.
PPS shall be described only in general without any disclosure of the location of control rooms, guard forces, protected areas and physical protection objects, without indication of locations and types of the EFPP devices, principles of their arrangement and functioning, without detailed description of the security organization, quantitative characteristics of security units and actions of the physical protection personnel, without any particular information on characteristics of the PPS in general and its individual functional systems and means.
X. Requirements for the content of Chapter 10 "Radioactive waste management"
Information on management of gaseous, liquid and solid radioactive wastes of the NPP shall be presented, any potential paths of RS releases into the environment shall be specified and the RW handling techniques shall be described in Chapter 10 of the NPP SAR.
RW management principles shall be stated and their compliance with the requirements of federal rules and regulations in the field of atomic energy use shall be substantiated.
Each RW management system shall be described in accordance with the standard system description framework given in Appendix 4 hereto.
Additional information in accordance with the requirements specified in items 10.1-10.5 of this Appendix as well as any information specific for the particular system shall be provided for each system under consideration.
10.1. RW generation sources.
Information on the RW generation sources with typical parameters used as the main input data for development of handling systems for all radioactive waste types generated both in the course of normal NPP operation and in case of accidents shall be provided.
Parameters used to determine activity of each radionuclide in the primary and secondary circuit coolant shall be specified and any existing assumptions shall be substantiated.
Quantitative characteristics of radioactive substances supplied into the coolant because of any fuel element cladding integrity losses shall be substantiated by calculated values with due regard for thermal loads on the fuel elements and other necessary parameters as well as the existing operation
experience for similar fuel assemblies, accident experience, temperature regimes and fuel burnout degree.
Information on concentration (activity) of fission and corrosion product radionuclides as well as actinides used in calculations of the energy radiation spectra from equipment and wastes shall be provided. In this case it shall be specified how activation of water and any impurities contained in it is taken into account. Information on the radionuclide composition of wastes, their generation mechanism and data on radionuclide concentration in the wastes shall be provided.
Mathematical models used to calculate the input data taken as the basis for design of the RW handling systems with due regard for the parameters of normal operation and transient modes shall be described.
Design values of controlled and uncontrolled leakages of the primary and secondary circuit coolant, auxiliary equipment circuits, equipment decontamination water being the sources of potential ingress of radioactive substances into the environment shall be systematized.
Information on the sources of leaks and flows, their parameters as well as their estimated contributions to the total radioactivity level shall be presented in tabular format; the above-mentioned values shall be compared with the data on operation of similar operating NPP power units.
Ingress of radioactive substances in the form of liquids, gases and aerosols into the rooms shall be assessed with regard to the main dose-generating radionuclides, the paths of their further propagation, releases and discharges to the environment shall be demonstrated. Information on the leakage measuring methods and any special-purpose means provided in the design to reduce leakage values shall be given.
Any systems that can become potential sources of RS releases (discharges) in the course of their operation or maintenance but are not referred to the RW handling systems shall be listed. Assessment of RS releases with description of the mechanism of their potential migration, propagation and ingress into the environment shall be provided for each of the above-mentioned sources both under normal operation conditions and in case of any possible failures. Information on the leakage flow rates, concentration of radionuclides and any other parameters sufficient to perform calculation estimates shall be specified. Information on the source confinement solutions adopted in the design shall be provided.
Mechanisms of RS generation and propagation at the NPP shall be described, and information on the global nuclides (such as 14C, 3H, 129I, 85Kr) generation rate shall be provided.
Analysis of the main design solutions aimed to reduce RS content in the primary circuit coolant as compared with the previous generation of NPP designs shall be provided. Comparison of the design data and data on operation of similar operating NPP power units shall be presented.
Information on all NPP systems being potential sources of RS releases into the environment in the form of gases or aerosols shall be provided in this section. The NPP design capabilities to handle gaseous wastes in all operation modes and in case of design basis accidents shall be described in the NPP SAR.
10.2.1. Design basis.
Information on the basic safety principles and criteria as well as regulatory requirements taken as the basis for the system design development shall be provided.
Purposes and criteria of the system calculations shall be specified with indication of expected annual RS releases and committed exposure doses for the personnel and the public due to their impact.
It shall be substantiated that the implemented principles and the corresponding techniques enhance efficiency of RS treatment. The above-mentioned substantiation shall confirm that the adopted systems contain all state-of-the-art technical achievements aimed to reduce radiation exposure for the personnel and the public.
All applied calculation methods and assumptions shall be specified. Information on the ways to consider peculiarities of the NPP site specified in Chapter 2 of the NPP SAR with regard to meteorological and hydrological conditions shall be provided.
Assessment shall be provided to confirm that the systems have sufficient capacity and the required redundancy to ensure RS treatment in all operation modes with the coolant activity corresponding to the safe operation limit.
It shall be substantiated that the systems have sufficient capacity, efficiency and the required redundancy to ensure the necessary RS treatment level and non-exceedance of the established permissible limits for the releases under normal operation conditions and also to limit RS releases in case of any accidents.
Peculiarities of the NPP design including the means to reduce the scope of maintenance, downtime of the equipment, the possibility for RS ingress into the rooms and means to enhance efficiency of the cleaning methods for equipment and rooms shall be described. The adopted design values of radionuclide activity in all units of the system shall be specified together with the input data for determination of these values. Layout and geometry of the system equipment shall be provided in order to perform calculations of biological protection in accordance with the requirements of item 11.3 of this Appendix for Chapter 11 of the NPP SAR.
Information on any measures provided in the NPP design in order to control RS ingress outside the gaseous RW management systems shall be presented. Information on any possible human errors and single failures that can cause uncontrolled releases to the environment shall be provided. Information on the release control means provided in the NPP design shall be specified. Efficiency of the preventive measures for process and radiological monitoring and control of the automatic release limitation system shall be substantiated.
All equipment of the systems where generation of explosive gas concentrations is possible shall be listed, and the design pressure and substantiation of the equipment adopted in the NPP design shall be provided. Information on the process instrumentation and control devices and any measures provided in the NPP design to prevent explosions and complete loss of tightness due to explosion shall be specified.
The radiological monitoring systems for processes and releases shall be described in the section of Chapter 10 of the NPP SAR developed in accordance with item 10.5 of this Appendix.
10.2.2. Description of the systems.
Information on each gaseous RW management system and diagrams of gas flows demonstrating the process equipment, gas movement paths in the system, capacity of the system and the corresponding equipment and redundant equipment shall be provided. For complex multi-functional systems the subsystems divided into independent parts shall be specified with the relevant equipment description. Maximum and normal input gas flow rates and RS concentrations for all operation modes shall be specified for each system in the tabular format or on the diagrams. The input data used to determine the above-mentioned values shall be provided. Information on the gas flow components and the technique for handling of hydrogen-containing flows shall be presented.
Interfaces of the systems and their boundaries with regard to the equipment of different classification groups shall be indicated on the process flow diagrams.
Information on the instrumentation and control devices and the system controls shall be provided.
Information on the existing bypass lines as well as the conditions affecting their use and the expected frequency of use of bypass lines due to downtime of the equipment shall be provided.
Information on location of hydraulic lock tanks (hydraulic locks) shall be provided, measures aimed to prevent their failure shall be described, and location on vents and secondary circulation paths shall be indicated for each system.
Information on both normal operation modes and any modes not related to normal operation shall be specified; information on ventilation systems for each building where appearance of radioactive substances may be expected shall be provided. Description shall include volumes of the buildings, expected flow rates in the ventilation systems of the buildings and their rooms, characteristics of filters and the calculation criteria taken as the basis for determination of these values. Information on the normal operation mode for each ventilation system and peculiarities of operation in different NPP operation modes including design basis accidents shall be presented.
The table with design concentrations of airborne radioactive substances with particle size distribution in the building rooms and corridors shall be provided for all operation modes including design basis accidents.
Information on any other NPP systems being potential sources of gaseous RS releases into the environment shall be provided; RS concentrations for all these systems and for all operation modes including design basis accidents shall be specified. The input data for determination of these concentrations shall be given.
Information on expected frequency and amount of released steam within the period of its discharge to the atmosphere upon any potential actuation of the primary and secondary circuit safety devices shall be summarized in a table with indication of the input data for determination of RS concentrations in the released steam.
10.2.3. RS releases.
Information on actual RS releases (qualitative and quantitative composition) to the environment within the last five years shall be provided.
Information on expected RS releases (qualitative and quantitative composition) in all operation modes including design basis accidents shall be specified for each subsystem and the system in general.
Design information for normal operation shall be provided for the operation limit with regard to the primary circuit coolant activity taking into account potential additional ingress of fission products from the fuel to the coolant in transient modes and in case of the NPP power unit shutdown. Any possible short-term increase of RS releases upon reaching of the safe operation limit with regard to the primary circuit coolant activity shall be also estimated.
Any assumptions taken into account, dilution coefficients, all sources of gaseous RS releases into the environment shall be indicated on the process gas flow diagrams and general layout drawings of the NPP.
Operational data on annual RS releases into the environment from the power units used as the prototype for the designed NPP shall be specified. Comparative analysis of the presented information shall be performed.
Geometric characteristics of the release sources, dimensions of the buildings, dispersion of the aerosol component, chemical and aggregate composition of releases as well as thermal and hydraulic
characteristics of gas-air mixture in which radioactive substances are released from the source (temperature, velocity and flow rate) shall be specified.
10.3. LRW management systems.
Information on all NPP systems for liquid RW management shall be presented in this section, the main characteristics of the liquid RW handling systems under normal operation conditions and in case of any abnormal operation including design basis accidents shall be specified.
10.3.1. Design basis.
Information on the basic safety principles and criteria as well as regulatory requirements taken as the basis for the system design development shall be provided.
Purposes and criteria of the system calculations shall be provided with indication of average expected amounts of generated LRW per year and per the entire NPP operation period and expected exposure doses for the personnel and the public due to their impacts.
Information on solidification techniques for liquid radioactive wastes shall be provided. It shall be substantiated that the adopted systems contain all state-of-the-art technical achievements in order to reduce radiation exposure for the personnel and the public.
All applied calculation methods and assumptions shall be specified. Information on the ways to consider peculiarities of the NPP site specified in Chapter 2 of the NPP SAR (with regard to meteorological and hydrological conditions) shall be provided.
Peculiarities of the NPP design including the means to reduce the scope of maintenance, downtime of the equipment and RS ingress into the rooms and means to enhance efficiency of the waste processing methods shall be described. The adopted design values of radionuclide activity in all units of the system shall be specified together with the input data for determination of these values. Layout and geometry of the system equipment shall be provided in order to perform calculations of biological protection in accordance with the requirements of item 11.3 of this Appendix.
Information on any measures provided in the NPP design in order to control RS ingress outside the LRW management systems shall be presented. Information on any possible human errors and single failures that can result in uncontrolled RS discharges to the environment and efficiency of the implemented precautions (both process-related and through the use of protections, interlocks, I&C devices) shall be provided. Information on any measures and controls provided in the design in order to prevent accidental and uncontrolled RS discharges into the environment shall be specified.
The health physics and radiological monitoring systems for processes and discharges shall be described in a section of Chapter 10 of the NPP SAR developed in accordance with item 10.5 of this Appendix.
10.3.2. Description of the systems.
The system shall be described in accordance with the standard system description framework given in Appendix 4 hereto.
Description of each system shall include process flow diagrams showing the equipment, normal direction of LRW flows, capacity of the system and the relevant equipment components, redundant equipment. For complex multi-functional systems the subsystems divided into independent parts shall be specified with the relevant equipment description. Information on the techniques for handling of all possible LRW shall be provided.
Maximum and normal liquid flow rates 273 (in m3/day per reactor) and radioactivity values for all
operation modes including design basis accidents shall be specified for each system in tables or on diagrams. The input data used to determine the above-mentioned values shall be provided.
Information on segregation of LRW flows and principles of their segregation shall be provided. All possible bypass lines as well as the conditions affecting their use and the expected frequency of use of bypass lines due to downtime of the equipment shall be specified.
Interfaces of the systems and their boundaries with regard to the equipment of different classification groups shall be indicated on the process flow diagrams. Information necessary to develop Chapter 11 of the NPP SAR shall be provided particularly the equipment components and blocks and pipelines containing increased concentration of radionuclides shall be indicated on the diagrams.
Information on all normal operation modes for each system and differences within the periods of various NPP operation modes including design basis accidents shall be specified.
10.3.3. RS discharges.
Information on the established standards and reference levels for RS discharges into the environment shall be specified for each RS discharge source and for the NPP in general.
Information on actual RS discharges (qualitative and quantitative composition) to the environment within the last five years of operation shall be provided, and it shall be substantiated that they do not exceed the established limits.
Information on expected RS discharges (qualitative and quantitative composition) in all operation modes including design basis accidents as well as for beyond design basis accidents with the most severe radiological consequences shall be specified for each subsystem and the system in general.
Estimate of the maximum possible short-term daily RS discharge from the NPP into the environment upon reaching of the safe operation limit with regard to the primary circuit coolant activity shall be provided.
Information on the applied assumptions, dilution coefficients, all sources of RS discharges into the environment shall be specified on the process flow diagrams and general layout drawings of the NPP.
Information shall be presented in the form of tables and diagrams. Operational data on annual RS discharges into the environment from the power units used as the prototype for the designed NPP shall be specified. Comparative analysis of the presented information shall be performed.
Design information for normal operation shall be provided for the operation limit with regard to the primary circuit coolant activity taking into account potential additional ingress of fission products from the fuel to the coolant in transient modes and in case of the NPP power unit shutdown. Any possible short-term increase of RS content in the effluents upon reaching of the safe operation limit with regard to the primary circuit coolant activity shall be estimated.
10.4. Solid radioactive waste management system.
Information on the capabilities of the NPP systems to handle solid radioactive wastes shall be provided in this section for all operation modes as well as abnormal operation including design basis accidents.
10.4.1. Design basis.
Information on the basic safety principles and criteria as well as regulatory requirements taken as the basis for the system design development shall be provided in this subsection.
Purposes and criteria for calculations of the solid RW management systems shall be specified with due regard for characteristics of the wastes, their maximum and expected amounts subject to processing, their radionuclide composition and activity of the wastes.
Information contained in this section of the NPP SAR shall be presented through the use of tables.
10.4.2. Description of the systems.
Information provided for each system shall include descriptions of the solid waste handling subsystem used to process ion exchange resins, sludge, concentrates, the solidification systems for liquid radioactive wastes; components of each subsystem shall be listed. Information on their design capacity and structural materials shall be provided.
Maximum and expected amounts of wastes, their physical state, composition, sources of wastes, radionuclide composition and specific activity shall be presented in the form of tables. Information on the input data used to obtain the above-mentioned values shall be provided, the methods to be used for processing of each waste type, the type of packaging container for wastes, the final form of conditioned wastes shall be described.
Process flow charts showing normal sequence of operations, flow rates in the system, duration of processing for each unit, expected isotope composition of each flow and capacity of the equipment shall be provided. Information on the process controls and instrumentation devices shall be specified, flow charts with indication of interfaces between the systems and boundaries between the equipment of different classification groups, I&C devices shall be provided.
Diagrams of the packaging, storage, handling and transportation sections for various categories of wastes at the NPP shall be presented.
Design information for normal operation shall be provided for the operation limit with regard to the primary circuit coolant activity taking into account potential additional ingress of fission products from the fuel to the coolant in transient modes and in case of the NPP power unit shutdown.
Any possible short-term increase of radioactive substance content in the solid wastes upon reaching of the safe operation limit with regard to the primary circuit coolant activity shall be estimated.
Information on any precautions provided in the NPP design in order to prevent RS ingress into the rooms and the environment shall be specified.
Efficiency of the implemented measures aimed to prevent RS ingress into the rooms and the environment and use of instrumentation and control devices for this purpose shall be substantiated. All possible human errors and single failures of the equipment that can result in RS ingress into the environment shall be listed and described.
Information on the solid waste management subsystem intended to process contaminated working clothes, equipment, tools, filters of the ventilation systems as well as other pressed and non-pressed radioactive wastes shall be provided. Maximum and expected input data for the above-mentioned wastes with indication of the waste sources, amounts, radionuclide composition and activity shall be presented in the form of tables. The input data used to determine the applied values shall be provided. Methods for conditioning and packaging of wastes as well as any equipment used for this purpose shall be described. Methods for processing and packing of large-scale wastes shall be described. Containers to be used for packaging of radioactive wastes shall be described. Compliance of the subsystem with the requirements of federal rules and regulations in the field of atomic energy use that govern handling of radioactive wastes shall be substantiated. Measures provided for sealing, decontamination and transportation of containers with wastes to the storage facilities shall be described together with the analysis of any possible abnormal operation including accidents. Information on the
arrangements for collection of wastes and decontamination techniques in case of any container integrity loss shall be provided. Precautions taken in the course of waste storage prior to loading and transportation, the expected time of RW storage at the NPP site, diagrams of the packaging sections, storage facilities, handling and transportation sections shall be provided. Information on the maximum possible and expected annual amounts, radionuclide composition and activity of each RW category subject to removal from the site shall be presented.
Information on the operational radioactive wastes and radioactive wastes generated in the course of the NPP power unit decommissioning shall be provided.
Conditions for temporary storage of the operational wastes at the plant and the planned location for their long-term storage shall be described.
Information on processing and removal of various RW categories generated in the course of the NPP decommissioning from the plant shall be provided.
10.5. Radiological monitoring and sampling system.
The system providing radiological monitoring with due regard for the information specified in Chapter 11 of the NPP SAR as well as sampling in the course of RW handling, RS releases and discharges under normal operation conditions as well as in case of any abnormal operation including accidents shall be described in this section.
10.5.1. Design basis.
Information on the basic principles and safety criteria taken as the basis for the system design development shall be provided in this subsection.
Information on the purposes, principles and criteria used in the radiological monitoring system design shall be provided; it shall be demonstrated how they were applied in the design of the system as a whole and its individual subsystems. Differences of the subsystems intended for functioning under normal NPP operation conditions as well as in case of any NPP abnormal operation including accidents shall be specified.
It shall be substantiated that the radiological monitoring system for RS releases and discharges can record all radionuclides subject to state accounting and regulation (quantitative and qualitative composition) supplied to the environment with the NPP releases and discharges both under normal operation conditions and in case of any accidents.
10.5.2. Description of the systems.
Information on the purpose of systems shall be specified, the principal structural diagrams and information on their operation principles shall be provided.
The following information shall be provided:
reliability and sufficiency of measurements for all operating conditions of the systems;
degree of protection against unauthorized access to the stored information;
sufficient redundancy of the system components under normal operation conditions and in case of their functioning under extreme conditions;
sufficiency of the primary detector locations;
correct selection of the sampling points and sufficiency of their number in order to provide correct
monitoring of the media composition;
sufficiency of warning means in case of any abnormal operation as well as in case of any disturbances requiring declaration of an emergency situation, substantiation of their correct location and selection of the alarm setpoints.
The following information shall be also provided for radiological monitoring of the waste management processes:
location of sensors;
types of sensors, characteristics, type of measurements;
instrumentation and control devices, redundancy, independence of performed measurements;
RS concentration measurement range and input data for determination of the provided range;
types and locations of the warning devices, radiation level alarms and controllers and their description;
stand-by power supply;
setpoint values of emergency alarm and actuation of protections, interlocks and controllers;
input data for determination of these values;
description of the measures provided for calibration, maintenance, verification, decontamination and replacement of control instruments.
The following information shall be presented for each sampling device:
basis for selection of sampling point locations;
expected flow, composition and concentration of radioactive and chemical substances in the samples;
frequency of sampling, type of the sampling equipment and methods used to obtain representative samples;
laboratory analysis methods and sensitivity of the instruments.
XI. Requirements for the content of Chapter 11 "Radiation protection"
Principles and criteria for assurance of radiation safety for the personnel and the public (with regard to dose limits, RS releases and discharges) under normal operation conditions and in case of any NPP abnormal operation including accidents shall be provided in Chapter 11 of the NPP SAR.
It shall be substantiated that individual exposure doses for the personnel and the public will not exceed the established limits under normal NPP operation conditions and ingress of radioactive substances into the environment in case of design basis accidents will not require implementation of any measures for protection of the public.
Information on the radiation situation monitoring in the rooms, radiological monitoring of the environment as well as personal monitoring shall be provided.
The following data shall be provided:
methods for protection against external exposure (gamma rays and neutrons from the nuclear core, structural materials of the reactor, reloaded fuel assemblies and equipment containing radionuclides);
methods for protection against internal exposure.
Information on compliance with the applicable requirements for radiation safety specified in the regulatory documents shall be provided in each section of Chapter 11 of the NPP SAR. References to the information presented in other sections of the NPP SAR shall be given.
Quantitative values of the criteria used to identify abnormal operation including accidents shall be provided and substantiated.
11.1. Assurance of the minimum achievable exposure level.
11.1.1. Radiation safety concept.
Information on the engineering features and administrative measures aimed to ensure protection of the personnel, the public and the environment against ionizing radiation shall be specified. It shall be substantiated that application of the proposed protection means and implementation of protective measures will not result in exceedance of the established dose limits, eliminate any unreasonable exposure and the existing radiation impact is maintained as low as possible with due regard for economic and social factors.
In this case information on the following design limitations shall be provided:
individual exposure doses for the personnel;
collective annual exposure dose for the personnel;
emergency exposure levels.
Efficiency of the protection systems with regard to non-exceedance of individual risk under normal NPP operation conditions shall be substantiated. It shall be substantiated that potential exposure risk does not exceed limit values of cumulative risk for the personnel and the public within a year established in the sanitary standards and rules for radiation safety.
11.1.2. Design basis.
Information on the design solutions ensuring reduction of the occupational exposure dose to the minimum achievable level with due regard for economic and social factors shall be provided.
It shall be substantiated that protection of the personnel against external exposure is designed with the margin of at least 2 for the entire design service life of the NPP. It shall be demonstrated through the use of conservative approach that the protection will ensure non-exceedance of the established annual effective exposure dose (with due regard for internal and external exposure) within the entire design service life of the NPP.
Information on the use of experience in NPP design and operation with regard to reduction of occupational exposure doses to the minimum achievable level taking into account economic and social factors as well as information on any changes in the NPP design (as compared with designs of similar NPPs) aimed to reduce occupational exposure doses shall be provided.
Additional costs in relation to these changes as well as economic benefits that may be obtained due to expected reduction of occupational exposure doses shall be assessed.
Information on any means provided in the NPP design in order to reduce dose rates in the controlled access area rooms and to reduce the time of the operating personnel stay in these rooms, to decrease the number of RS sources, to enhance the protection, to reduce the scope and time of maintenance, to facilitate access to the equipment, to simplify operational procedures and also to reduce and simplify any other actions required within the operation period shall be specified.
11.1.3. Arrangement of operation.
Information shall be provided to confirm that the requirements aimed to reduce occupational exposure doses to the minimum achievable level with due regard for economic and social factors are taken into account in arrangement of operation. It shall be described how the requirements for operation arrangement and experience in operation of similar power units are taken into account in development of the equipment, biological protection and design of the plant in accordance with the information provided in the sections of Chapter 11 of the NPP SAR performed in compliance with items 11.1.2 and 11.3.1 of this Appendix.
Radiation criteria used to develop guidelines and engineering features for performance of radiation-hazardous works in order to reduce occupational exposure doses shall be specified.
Information on RS content in the equipment (except for the equipment of RW management systems described in Chapter 10 of the NPP SAR) being the source of radiation considered in the biological protection calculations and design shall be provided.
Information on the reactor core as the source defining ionizing radiation levels in the course of the reactor power operation in the rooms beyond the biological protection where presence of the operating personnel can be necessary as well as the source of fission products supplied to the primary circuit shall be provided.
The following information shall be provided:
activity accumulated in the gas gap of gas-tight fuel elements under assumed power unit operation with 100% power by the end of the campaign in the course of the reactor power raising to steady-state load as the source of ionizing radiation in case of any accidents associated with loss of the reactor coolant pipeline tightness;
volumetric power of the ionizing radiation sources in spent fuel assemblies as the sources in the course of spent fuel handling;
average group neutron flux density values within the core volume;
activity of the reactor internals as the source in the course of transportation and handling operations with the reactor internals as well as their repair and maintenance;
the primary circuit as the source of activation products for the primary circuit coolant and activated corrosion products as well as fission products entering the coolant due to any defects of fuel element claddings;
the secondary circuit as the source of radioactive substances in case of any primary circuit coolant leakages;
other RF systems and components that may contain radioactive substances;
the system of refueling, SNF storage and transportation containing fission products in the irradiated fuel and structural material activation products;
other radiation sources: start-up neutron sources for calibration of instruments and devices, sources for gamma-radiography, nuclear reaction by-products and any other sources requiring radiation protection.
Description of the radiation sources (except for the nuclear core) shall contain the table of radionuclide composition and radiation energies, information on the activity, geometric parameters of the source as well as the input data for determination of the specified values. Information on radionuclide composition, amount and physical and chemical states of all sources with the activity exceeding 109 Bq shall be provided in the NPP SAR.
It shall be substantiated that ingress of fission products into the coolant in the course of power operation corresponds to the operation limit with regard to the primary circuit coolant activity. Increased ingress of fission products from the fuel to the coolant in any transient modes and in case of any abnormal operation including accidents shall be taken into account.
Information shall be presented in such a way so that it could be used as the input data for the biological protection calculations.
Information on the methods for calculation of activity in the process media and equipment and any software tools used for calculations shall be provided.
Location of all radiation sources as well as any possible and actual RS migration paths shall be indicated on the general layout drawings of the RF equipment and on the plans.
11.2.2. Gaseous RS sources.
Information on the sources of gaseous RS ingress into the atmosphere of the controlled access area rooms taken into account in development of protective measures and assessment of occupational exposure doses shall be specified. Apart from the sources existing under normal operation conditions information on the sources appearing due to failures of the main equipment as well as in the course of repair works shall be provided.
Description shall contain the calculation results for concentrations of radioactive gases and aerosols expected in normal operation modes, transient modes and in case of any abnormal operation in the controlled access area rooms.
Models, parameters and input data required to calculate concentration of radioactive gases and aerosols shall be provided. In case of input data absence operational data on similar NPPs may be used.
11.3. Consideration of the radiation protection design peculiarities.
11.3.1. Location plan and layout of the buildings, structures and equipment.
The plan of the complex of operating buildings, structures and rooms of the NPP with layout of the process equipment being the radiation source inside them as well as all radiation sources specified in the section of Chapter 11 of the NPP SAR developed in accordance with item 11.2 of this Appendix and Chapter 10 of the NPP SAR shall be presented.
Information on the planning and layout of the buildings, structures and equipment performed in order to provide radiation protection shall be specified.
The following information shall be presented on the plan:
boundaries of the controlled access area and division of its rooms into non-attended, periodically attended and attended as well as rooms of the uncontrolled access area;
location of the administrative and amenity building;
location of airlocks, stationary decontamination locks, active laundry, medical posts;
schemes of the personnel and transport movement, delivery of clean equipment and materials and removal of contaminated ones;
location of the sites for storage of contaminated equipment, decontamination sections, places for solid RW collection, control panels for the equipment and mechanisms of the RW processing systems;
location of sensors and control panels of the radiological monitoring system;
location of laboratories for analysis of the radioactive media samples (chemical, radiochemical, radiometric, spectrometric), laboratory of personal monitoring as well as monitoring of metals, repair and calibration room (workshop), storage facilities of ionizing radiation sources;
location of external health physics laboratories, observation stations, and checkpoints;
places for collection of non-radioactive wastes and arrangement of control in order to eliminate accidental ingress of radioactive substances into non-radioactive wastes.
Information on classification of the NPP zones and rooms adopted in the design and taken as the basis for design of biological protection against penetrating radiation and prevention of air contamination in attended and periodically attended rooms of the controlled access area with radioactive substances shall be provided.
11.3.2. Structural peculiarities of the systems and equipment components.
Design peculiarities of the equipment and facilities enabling to reduce exposure doses for the personnel shall be specified; it shall be illustrated through examples how these characteristics influence the main requirements for the operating regulations stated in the section of Chapter 11 of the NPP SAR developed in accordance with item 11.1.3 of this Appendix.
Information on the structural peculiarities reducing maintenance or other operations in radiation fields, decreasing intensity of sources and also providing quick entrance, easy access to the workplace, remote performance of the works or reduction of the personnel stay time or any other measures aimed to reduce exposure of the personnel shall be specified.
The section shall include description of the methods applied in the design in order to reduce generation, distribution and accumulation of active corrosion products. Illustrative examples shall be provided: equipment drawings and piping schemes for the components requiring access of the personnel during the NPP unit power operation (equipment of the active water treatment systems, vessels (tanks), coolers, deaerators, pumps, SG, sampling systems (devices)). Information on location of sampling points, instrumentation and control devices and control panels (boards) shall be provided.
11.3.3. Biological protection.
Information on biological protection shall be provided for each radiation source specified in Chapter 10 of the NPP SAR and the section of Chapter 11 of the NPP SAR developed in accordance with the requirements of item 11.2 of this Appendix.
Information on special-purpose protective devices and equipment including containers, casings, screens, hoisting equipment used to handle any types of radioactive materials shall be presented.
Information on the software tools with applied assumptions and also information on their verification and validation shall be provided; results of calculations shall be presented.
Information on the design radiation levels in attended and periodically attended rooms of the controlled access area, in the rooms of the uncontrolled access area and also in the administrative and amenity building under normal operation conditions, during design basis accidents and in the course of the NPP power unit decommissioning shall be provided.
11.3.4. Ventilation, filtration and conditioning systems.
Information on the basic parameters of the ventilation system design for the controlled access area from the viewpoint of the personnel protection as well as on any components aimed to ensure the personnel safety referred to ventilation systems but not included in Chapters 9 and 10 of the NPP SAR shall be provided. Removal of gas and aerosol fission products from the controlled access area rooms, process vents as well as the RS release control system shall be described in Chapter 10 of the NPP SAR.
Principle of separate ventilation for the rooms of the controlled access area and the uncontrolled access area shall be described.
Measures provided in the design for air purification from radioactive gases and aerosols shall be specified, information on the plan of rooms where purification is performed and purification devices (filtering stations) are installed and also piping schemes and filter valves shall be presented.
Information on the maintenance conditions as well as monitoring, testing and isolation means for the systems shall be provided. Information on the means for air treatment efficiency determination, replacement and transportation of the used filters (filtering elements) shall be provided. Characteristics of the applied air treatment means as well as the criteria established for replacement of filters (filtering elements) shall be specified. Decontamination factors applied in the radiation safety analysis shall be specified; due to significant dependence of these factors from filtration conditions for the purposes of radiation situation assessment they shall be assumed on the basis of the most severe operation conditions for the filtration systems.
11.3.5. Radiological monitoring system.
Criteria for selection of radiological monitoring hardware, development of the sampling point scheme and location of the devices (instruments) shall be specified. Information on the radiological monitoring hardware provided in the NPP design shall be presented:
continuous monitoring devices based on stationary automated systems and stationary instruments;
in-process monitoring devices based on portable, transportable and (or) mobile instruments and units;
laboratory analysis devices based on laboratory instruments, units, means for sampling and preparation of radioactive samples for analyses;
individual exposure monitoring devices for the personnel.
The list of radiological monitoring objects, classification of control types in accordance with the requirements of regulatory documents shall be presented.
Description of the radiological monitoring system shall include the basic technical characteristics (controlled parameters, types and number of sensors, measurement range, error), information on the metrological support, alarm units, recording devices and location of sensors, indicating (reading) and signaling devices (instruments). Diagrams of sampling lines with valves and flow activators shall be
provided.
Information on location of the air sampling points (places) for gas and aerosol activity monitoring, on the air sampling system and also criteria and methods aimed to obtain representative measurements of radioactive gas and aerosol concentrations shall be provided.
Radiological monitoring hardware for measurement of the radiation situation parameters and the personnel exposure doses as well as for measurement of intensive radiation in case of a radiation accident shall be described. Necessity for additional instrumentation and control devices for such measurements shall be substantiated.
Information on the software tools for information processing and presentation, software tools ensuring prediction of the radiological consequences in case of any events at the NPP, collection, storage and systematization of the data on radiation contamination of the environment and exposure doses for the personnel and the public as well as information on the ST verification and (or) validation shall be provided.
Information on the purpose and configuration of the environmental ARSMS shall be provided.
Configuration and equipment of the stationary external health physics laboratory and the mobile laboratory shall be described.
Information on location and equipment of the stationary observation stations and checkpoints for the radiation situation in the environment within the sanitary-protective area and the supervised area shall be provided.
It shall be substantiated that the environmental monitoring system ensures functioning and can provide reliable information both under normal operation conditions and in case of accidents.
ARMS and ARSMS shall be described in accordance with items 7.2.1.1 - 7.2.1.5 of this Appendix.
11.4. Assessment of dose commitments under normal operation conditions and in case of accidents
Duration (within a year) of the personnel stay in the controlled access area rooms shall be assessed with indication of the number of people and duration of their stay in these rooms under normal NPP operation conditions, in transient modes of the NPP and in the course of repair works. For the controlled access area rooms where gas and aerosol activity specified in the section of Chapter 11 of the NPP SAR developed in accordance with item 11.2.2 of this Appendix is expected duration of the personnel stay in man-hours and ingress of radioactive substances into the human organism due to inhalation shall be assessed.
Annual individual dose (aggregate and separate for external and internal exposure) and dose commitments of the personnel (collective dose) under normal operation conditions, during maintenance, in-service inspection and examination of weld joints, RW management, refueling of the reactor core, repair works and also in case of accidents shall be assessed.
It shall be specified that exposure doses and dose commitments are assessed dynamically depending on the NPP service life.
The input data, calculation methods and models and assumptions applied to determine the above-mentioned values shall be specified.
Information on exposure doses and dose commitments of the personnel obtained during operation of similar NPPs may be used to assess doses and dose commitments.
Annual dose at the boundaries of the controlled access area, the NPP site and the sanitary-protective area as well as in the locations of the main radioactivity sources within the NPP territory shall be assessed. Annual exposure dose for the workers of any organizations performing works or rendering services to the operating organization from these sources at operating NPPs in the course of the next power unit construction shall be assessed. Input data, calculation methods and models shall be provided.
Exposure doses for the public in case of design basis accidents shall be assessed.
11.5. Radiation safety assurance.
11.5.1. Organization.
Information on the organizational structure of the operating organization departments ensuring radiation safety shall be presented. Information on qualification and experience of the personnel, their authorities and responsibilities for radiation safety assurance as well as for supervision over handling of radioactive substances, nuclear materials and radiation sources shall be specified.
Information on any technical and administrative measures aimed to control the personnel stay in the controlled access area and compliance with the guidelines for performance of radiation-hazardous works shall be provided. Information on any provided mobile units equipped with the hardware enabling to obtain information on radioactive contamination both under normal operation conditions and in case of any abnormal operation including accidents shall be presented.
Information on the storage conditions for radiological monitoring instruments, their calibration and metrological validation shall be specified.
It shall be specified how the authorities for state regulation of safety in atomic energy use are informed on the results of any works for radiation safety assurance.
11.5.2. Radiological monitoring assurance.
Information on radiological monitoring with indication of the applied procedures and methods under normal NPP operation conditions and in case of accidents shall be provided. The scope of information to be provided for each type of control listed below is given in Appendix 6 hereto.
11.5.2.1. Radiological monitoring at the NPP power unit.
Information on arrangement of integrity and state monitoring for the physical barriers in the ways of RS and ionizing radiation propagation shall be provided.
It shall be specified how the following aspects are arranged at the NPP:
obtaining of information on integrity and state of the barriers;
alarm upon reaching of the regulated intervention levels (operation limits and safe operation limits for physical barriers at the NPP);
independent and prompt notification of the state administration and supervision authorities about integrity and state of the barriers.
Information on arrangement of the personnel exposure monitoring shall be provided.
It shall be specified how the following aspects are arranged at the NPP:
determination on exposure dose rates in attended, periodically attended and non-attended rooms
of the NPP (for the latter - within the period of repair with the NPP power unit shut down);
determination and assessment of the personnel exposure doses within the entire range of possible radiation impact levels occurring in the course of normal operation as well as in case of any design basis and beyond design basis accidents;
calculation and prediction of individual exposure doses for the personnel under normal NPP operation conditions and in case of accidents;
obtaining of information for emergency assessment of the radiation situation in the areas of the personnel stay for timely selection and implementation of optimal protection measures during development of design basis and beyond design basis accidents.
Information on arrangement of the RW management supervision shall be provided.
It shall be specified how the following aspects are arranged at the NPP:
obtaining of information on the radiation situation created by radioactive releases and discharges to the environment, determination of exposure doses for the personnel at the NPP, in the sanitary-protective area and the public in the supervised area;
determination of the amounts and radionuclide composition of the radioactive wastes generated and stored at the NPP;
obtaining of information on dose burden for the personnel occurring in the course of RW handling works;
detection and recording of any exceedance of the established values for radioactive releases and discharges to the environment, reference levels as well as unauthorized RW movement and accumulation at the NPP site.
Arrangement of supervision over non-dissemination of radioactive contaminations shall be described.
It shall be specified how the following aspects are arranged at the NPP:
determination of the RS contamination levels for surfaces of the process rooms and equipment, skin, footwear, working clothes, personal protection equipment of the personnel and any vehicles upon crossing of the controlled access area boundaries;
determination of the RS contamination levels for personal clothes and footwear of the personnel upon crossing of the NPP territory boundaries;
determination of the RS contamination levels for vehicles and transported cargoes upon crossing of the NPP territory boundaries.
11.5.2.2. Radiological monitoring of the environment in the sanitary-protective area and the supervised area.
Information on arrangement of radiological monitoring in the sanitary-protective area and the supervised area of the NPP in order to monitor radioactive contamination of the environmental media and exposure of the personnel and the public shall be provided.
It shall be specified how the following aspects are arranged in the sanitary-protective area and the supervised area of the NPP:
obtaining of information for assessment of exposure for the critical groups of the public and the
personnel;
obtaining of information for assessment of tendencies and changes of RS accumulation in the environmental media and human organism;
determination of correlation between the results of the radiological monitoring of the environment and the radiological monitoring data on RS releases and discharges;
obtaining of information for emergency assessment of the radiation situation in the territory subjected to radioactive contamination during a beyond design basis accident in order to establish the radiation accident zone boundaries and to implement the required measures for protection of people and the environment (intervention nature) with due regard for the fact that the proposed intervention shall do the society more good than harm.
11.5.2.3. Radiological monitoring in case of any abnormal operation including accidents.
Information on arrangement of radiological monitoring at the NPP in case of any abnormal operation including accidents (with due regard for potential accident development scenarios with RS releases into the environment) as well as control of the radiation situation in the radiation accident zone by the NPP workforce and means interacting with the radiation monitoring means arranged by the institutions and stations of the Russian National Automated Radiation Situation Monitoring System shall be provided. It shall be specified how the following aspects are arranged at the NPP (in case of any abnormal operation including accidents):
detection of any physical barrier integrity breaches;
determination of values of RS releases (discharges) to the environment (amounts and radionuclide composition of released (discharged) radioactive substances);
arrangement of steam-gas mixture sampling from the reactor building rooms after the accident beginning;
determination, assessment and prediction of the radiation situation in the NPP rooms, at the NPP site, in the sanitary-protective area and the supervised area;
determination, assessment and prediction of the external and internal exposure doses for the personnel and all persons present within the NPP site, in the sanitary-protective area, the critical population group in the supervised area;
determination of the boundaries of the protective action planning zone, the compulsory population evacuation action planning zone and the radioactive contamination zone based on prediction of the radiation situation;
prediction of possibility for reaching of the intervention levels and establishing of emergency preparedness levels;
guaranteed functioning of the radiological monitoring system part under the conditions formed by a considered beyond design basis accident with the most severe radiological consequences;
development and implementation of the optimal measures to protect the personnel and the public;
prediction of the radiation situation in the environment along the path of radioactive release propagation in the atmosphere during development of a beyond design basis accident for immediate protection of the public with due regard for the established criteria for implementation of the public protection measures in case of any radiation accident at the NPP;
timely notification of the state atomic energy use controlling authorities and the authorities for state regulation of safety on any necessity to implement measures for the public protection.
11.5.3. Medical care and health protection for the personnel.
11.5.3.1. Medical care organization.
Information on the organizational structure of medical support and health monitoring for the personnel shall be provided. Information on the program for assessment of internal exposure doses for the personnel (whole body and individual organs), selection criteria for the personnel to be examined within the program, frequency of radionuclide content assessment in the whole body and individual organs shall be provided.
11.5.3.2. Equipment, protective means and devices.
Information on location of the medical and sanitary facilities (first aid facilities, sanitary stations, active laundry, change rooms, shower rooms, rooms of duty dosimetricians and outgoing health physics monitoring stations) and types of the equipment (instruments, devices) for sanitary monitoring shall be provided.
Information on the personal protection equipment shall be provided, their characteristics and data on their usage and maintenance shall be specified.
Information on location of the equipment ensuring radiation safety of the personnel, laboratory facilities for radio- and spectrometric analysis, storage facilities for protective clothes, respiratory protection devices, decontamination equipment (for equipment and the personnel) shall be provided.
11.5.3.3. Methods of radiation protection.
Information on the methods of radiation protection specified in the guidelines used for refueling, monitoring of the state of metal and weld joints, handling of spent nuclear fuel, radioactive waste, in the course of normal operation and repair works as well as methods for handling and storage of sealed and unsealed by-products, sources, special nuclear materials shall be presented.
Methods of special-purpose air sampling as well as selection and use of special-purpose equipment and devices for respiratory protection shall be described.
Criteria and methods of radioactive contamination control for the personnel, equipment and surfaces shall be provided.
XII. Requirements for the content of Chapter 12"Safety systems. Special-purpose hardware
for beyond design basis accident management"
Chapter 12 of the NPP SAR shall contain information on any protective, localizing and supporting safety systems provided in the NPP design as well as on special-purpose hardware for BDBA management.
Description of the CPS shall be presented in Chapter 7 of the NPP SAR.
Assessment of the capability of safety systems as well as special-purpose hardware for BDBA management to perform any functions prescribed in the NPP design shall be provided and substantiated.
12.1. Protective safety systems.
12.1.1. Each protective safety system shall be described in accordance with the standard system description framework given in Appendix 4 hereto.
Besides additional information required in accordance with Section 12.1 of this Appendix shall be provided for each system under consideration.
12.1.2. Information on the following protective safety systems shall be provided:
emergency reactor tripping system (with due regard for the information on this system or part thereof specified in Chapter 7 of the NPP SAR);
active emergency core cooling system;
passive emergency core cooling system (hydroaccumulators);
primary circuit overpressure protection system;
secondary circuit overpressure protection system data;
primary circuit emergency gas removal system;
emergency boron injection system;
emergency SG water supply system;
active system of heat removal from steam generators;
system of passive heat removal from steam generators;
primary circuit emergency cooldown system;
main steam line isolation system;
any other protective safety systems provided in the NPP design.
12.1.3. The following characteristics of protective safety systems shall be specified and substantiated:
flow rate;
pressure;
temperature;
volume of the tanks;
boric acid concentration;
efficiency (in units эфф / с ) of emergency boron injection by the systems intended for boric acid supply to the primary circuit;
flow resistance of the lines;
characteristics of the valves (response time, principle of operation);
redundancy of power supply sources and active components of protective safety systems;
period from the signal transmission moment to the reactivity change recording by the systems supplying boric acid to the primary circuit;
redundancy of instrumentation and controls.
Information on any provided hardware for control of neutron poisoning nuclides content in the liquid poison solution in the emergency boric acid storage tanks in the course of the NPP operation shall be specified.
12.1.3.1. The following characteristics shall be presented and substantiated for overpressure protection systems:
quantity of safety devices (valves);
discharged medium and its total mass;
changes of the medium flow rate through each valve over time;
valve actuation pressure;
valve opening time;
need for power supply, belonging to the group of EPSS consumers;
reliability indicators in accordance with the requirements of federal rules and regulations in the field of atomic energy use establishing general technical requirements for NPP pipeline valves;
number of bubblers for steam receipt;
changes of steam flow rate over time and total amount of steam supplied to the bubblers;
initial and final water temperature in the bubbler (in case the madium is discharged to the bubbler).
Analysis results for any transient modes that can be accompanied with pressure increase in the primary and secondary circuits shall be provided with reference to Chapter 15 of the NPP SAR. Information on the design mode accompanied with maximum pressure increase shall be presented in order to determine throughput capacity of safety devices.
The basic characteristics of each component shall be specified: information on its design, orifice size, design throughput capacity and places of installation for the valves as well as diameter, length and routes of pipelines.
The list of design parameters shall be provided for each component under consideration, the number and type of work cycles shall be defined, and external conditions the components are designed for shall be specified.
Results of thermohydraulic calculations for the overpressure protection systems of the primary and secondary circuit and results of the analysis of the system capability to perform its functions shall be provided.
Results of the analysis demonstrating the influence of any changes in the operation modes, parameters and operating characteristics of the equipment on the system characteristics shall be provided.
Design substantiation for the selection of throughput capacity, the number of safety valves and their opening (closing) setpoints shall be presented with reference to the results of performed
calculations.
Information on the actions and tools for installation of pressure relief devices within the primary circuit boundaries and on the SG from the secondary circuit side shall be provided. The input data for calculations of the permissible loads on the components (axial thrust, bend and torsion) shall be provided. The list of these loads and the resulting stresses shall be given. Results of strength calculations shall be presented.
12.1.3.2. It shall be specified how the systems are protected against unauthorized intervention of the personnel.
12.1.4. Information on the requirements for the systems that support functioning of the protective safety systems shall be specified:
characteristics of power supply for the systems;
distribution of consumers by EPSS systems and groups;
algorithm for connection of the consumers during start-up of the systems with power supply from independent sources;
permissible deviations of connection time, frequency and voltage;
characteristics of the compressed air supply system for the PSS components (flow rate, air parameters and quality, functioning in case of any failures of the compressed air supply system);
system oil supply characteristics (flow rate, oil parameters and quality);
information on the capacity of the system drains and vents;
information on ventilation in the system rooms (characteristics of blowers and ventilation systems, heat emission intensity, air exchange rate).
12.1.5. Information on the methods, means and procedure for control of metal condition in the pipelines and equipment of the systems shall be provided.
12.1.6. Information on the methods and means for control of vibrations, noise and leakages shall be specified.
12.1.7. Information on heat removal from the systems shall be presented:
characteristics of heat emissions;
cooling media; characteristics of the media supply;
characteristics of mechanical impurities.
12.1.8. Information on purification of the working media from radioactive substances and mechanical impurities shall be provided:
means of treatment;
water exchange rate;
measures to prevent clogging of the system components and loss of their heat transfer and throughput capacity.
12.1.9. Information on gas removal and gas vents from the systems as well as on fire protection
means shall be provided.
12.1.10. Information on the possibility to use PSS systems and components for beyond design basis accident management shall be provided.
12.1.11. Additional information on the system components with due regard for peculiarities of these components shall be presented:
design of supports, fasteners, suspenders, penetrations and expansion joints;
drains and vents;
welding information;
information on structural and welding materials as well as on their compatibility with the process media;
permissible warm-up and cooldown rates; information on safety devices;
b) valves (except for BRU and safety valves):
regulatory documentation in compliance with which valves are manufactured and operated;
manufacturing plant;
design; information on structural materials and welding;
information on compatibility of the structural and welding materials with the process media;
characteristics (leak-tightness, flow resistance, opening pressure - for check valves; information on the drives - the drive parameters, response time, permissible pressure differential);
conditions of layout, location and external environment;
design of supports and fasteners;
permissible warm-up and cooldown rates;
labelling;
c) heat exchangers:
regulatory documentation in compliance with which heat exchangers are manufactured and operated;
manufacturing plant;
design;
information on structural and welding materials and their compatibility with the process media;
characteristics: flow rates and velocities of the media, parameters of the media (pressure, temperature), heat transfer coefficient, flow resistance values of the circuits, protections and interlocks;
layout and media allocation conditions;
requirements for the cooling water quality;
information on instrumentation and controls;
design of supports and fasteners;
permissible warm-up and cooldown rates;
the list of parameters controlled in the course of operation and scope of diagnostics (movements, vibration, leakages, media parameters, characteristics of mechanical impurities in the media, changes of heat transfer coefficients);
heat insulation design;
labelling, painting, corrosion protection;
repairability;
overpressure protection (diagram, design and characteristics of safety devices);
the technique for tube leakage detection and elimination of defects;
d) pump sets:
regulatory documentation in compliance with which pump sets are manufactured and operated;
manufacturing plant;
design;
information on structural and welding materials and their compatibility with the process media;
characteristics: capacity, pressure head, power, time of turn, net positive suction head, starting current of the motor, suction lift, information on suction vortex formation, requirements for the water purity from any mechanical impurities;
vibration characteristics, temperature of the pumped water, permissible number of starts per hour;
information on instrumentation and controls;
protections and interlocks;
layout and location conditions;
design of supports and fasteners;
lubrication system parameters;
the list of parameters controlled in the course of operation and scope of diagnostics (movement, vibration, leaks from sealing glands, water and oil parameters, pump characteristics);
labelling;
maintainability.
e) tanks:
regulatory documentation in compliance with which tanks are manufactured and operated;
design;
information on structural materials and their compatibility with the process media;
characteristics: volume, medium exchange rate;
design of drains and vents;
poison concentration uniformity assurance;
sludge removal technique;
assurance of the design process medium level and overflow prevention;
layout and location conditions;
design of supports and fasteners;
the list of parameters controlled in the course of operation (levels, permissible leakage value, media parameters, poison concentration);
f) bubblers:
regulatory documentation in compliance with which bubblers are manufactured and operated;
manufacturing plant;
design;
information on structural and welding materials and their compatibility with the process media;
thermal calculation data, substantiation of complete steam condensing by the bubbler water;
characteristics: changes of media flow rates over time, media velocities, media parameters (pressure, temperature, water volume and parameters, flow rate and amount of received steam, time period within which the bubbler can condense steam; protections and interlocks);
characteristics of the integrated heat exchanger: changes of media flow rates over time, media velocity, cooling water parameters, heat transfer coefficient, flow resistance, pressure differential;
information on instrumentation and controls;
layout and location information;
design of supports and fasteners;
requirements for the quality of condensing and cooling water; the list of parameters controlled in the course of operation and scope of diagnostics (movement, vibration, leakages, parameters of condensing and cooling water, characteristics of mechanical and chemical impurities, heat transfer coefficient changes);
overpressure protection (diagram, design and characteristics of safety devices);
technical measures to prevent underpressure in the steam line supplying steam below the water
level in the bubbler;
g) BRU and safety valves:
regulatory documentation in compliance with which BRU and safety valves are manufactured and operated;
manufacturing plant;
design and principle of operation;
information on structural materials and welding;
information on compatibility of the structural and welding materials with the process media;
characteristics (throughput capacity, flow characteristics, actuation pressure, opening time, information on leak-tightness, drive characteristics and parameters);
need for power supply;
repairability;
labelling.
12.2. Localizing safety systems.
12.2.1. Each localizing safety system shall be described in accordance with the standard system description framework given in Appendix 4 hereto.
Besides additional information required in accordance with items 12.2.2, 12.2.3 of this Appendix shall be provided for each system under consideration.
In case any other LSSs (not specified in this section) are present at the NPP power unit, information in accordance with the standard system description framework given in Appendix 4 hereto as well as in accordance with additional requirements stated in items 12.2.2, 12.2.3 of this Appendix shall be provided for these systems.
Individual aspects of LSS operation and testing shall be described in accordance with the requirements stated in items 12.2.4, 12.2.5 of this Appendix.
12.2.2. General requirements for LSS description.
12.2.2.1. Information on any technical and administrative measures provided in the NPP design in order to limit leakages from the containment in case of accidents shall be specified. It shall be described how these measures limit pressure and temperature of the medium within the ALA space, prevent detonation of explosion-hazardous mixtures, protect the containment against dynamic impacts of jets and missiles and limit RS releases into the environment.
12.2.2.2. It shall be substantiated that all LSSs and their components withstand the number of in-house tests provided in the NPP design and also the required number of cycles with excessive pressure and underpressure during the containment system strength and tightness testing in the course of commissioning works and operation without any loss of operability.
12.2.2.3. The time period from the beginning of a design basis accident with loss of coolant and up to the moment when the personnel access to the ALA is possible shall be substantiated. Similar information shall be provided for beyond design basis accidents considered in the NPP design.
12.2.2.4. Information on the measures implemented in order to maintain operability of the LSS components subjected to low ambient temperatures that can result in ice formation on their surfaces shall be provided.
12.2.2.5. Information on any sealings applied in the LSS components shall be presented. Information on frequency and methods of their replacement and requirements for their leak-tightness shall be specified.
12.2.3. Requirements for description of individual LSS systems and (or) components.
12.2.3.1. Containment system.
All components included into the containment system shall be listed.
It shall be substantiated that the building structures of the containment ensure performance of their functions in accordance with federal rules and regulations in the field of atomic energy use.
In case steel containments are used in the NPP design it shall be substantiated that steel containments of the NPP comply with the federal rules and regulations in the field of atomic energy use establishing the requirements to strength calculations for the steel NPP containments.
Information on the load limits for the containment components due to impacts occurring during design basis accidents and beyond design basis accidents included into the final BDBA list (shock waves, jets, missiles, forces from connected pipelines) as well as impacts of any external natural and human-induced factors shall be provided. Information on the medium pressure and temperature within the ALA space in case of design basis accidents and beyond design basis accidents included into the final BDBA list shall be specified.
The adopted design leakage value from the containment system shall be indicated.
Information on performance of the functions for biological protection against ionizing radiation by the containment system shall be provided.
It shall be specified how leak-tightness of the containment part covered with concrete is monitored and also how it is to be repaired in case of necessity.
In case double containments are used the following information shall be also provided:
safety functions performed by each containment;
the required underpressure to be maintained in the annulus under normal NPP operation conditions and in case of any abnormal operation including accidents;
techniques for purification and removal of any leaks to the annulus from the inner containment under normal operation conditions, in case of any abnormal operation (including accidents) accompanied with pressure increase inside the containments and also in post-accident modes;
information on monitoring of underpressure, medium temperature and RS concentration within the space between the inner and outer containment (information on the place for display of the controlled parameters shall be specified);
design leakage value to the environment from the outer containment;
measures for protection against any possible increase of pressure and (or) temperature in the annulus caused by ruptures of pipeline sections inside it; pressure and temperature limits in the annulus endured by the inner and outer containment as well as the safety-related system components located in
the annular space between containments;
dimensions of the annulus;
description of the control methods for technical condition and repair of the containment building structures and maintenance of the equipment located within this space.
12.2.3.1.1. Leak-tight steel lining.
Information on the ways to arrange joints between the leak-tight steel lining parts and between these parts and other containment components and to perform regular leak-tightness checks for these joints shall be specified. Information on the ways to arrange control of assembly weld joints of the leak-tight steel lining in the course of its acceptance and operation as well as in-process detection of defects shall be specified.
The following information shall be provided:
the list of regulatory documentation in accordance with which strength calculations were performed for the leak-tight steel lining;
conditions for selection of the anchorage type and interval;
steel type and grade for the leak-tight steel lining, substantiation of the material selection;
substantiation of thickness selection for the leak-tight steel lining, assumptions used in its strength calculations, algorithm of these calculations and the input data thereof.
Information on the technical solutions adopted in the NPP design for the rooms used as vessels for any working media with walls or floors forming part of the containment shall be provided.
Input data and calculation results substantiating maintenance of the lining leak-tightness with due regard for strength characteristics of the reinforced concrete containment structures and thermal stresses occurring during design basis accidents shall be provided. Information on the measures implemented in order to maintain structural integrity of the lining in case of any beyond design basis accident included into the final BDBA list shall be provided.
In case the leak-tight steel lining is used as external reinforcement and (or) formwork information on compliance of the adopted technical solutions with the requirements of federal rules and regulations in the field of atomic energy use governing arrangement and operation of NPP LSSs shall be provided.
the list of regulatory documents used as the basis for selection of loads and impacts on the reinforced concrete containment structures as well as their combinations for strength calculations of the containment system;
the list of regulatory documents used as the basis for design of the LSS building structures performing the function of biological protection against ionizing radiation of the medium within the containment system;
processes and factors taken into account for selection of metal reinforcement elements and pre-stressed components of the reinforced concrete containment structures;
it shall be demonstrated that reinforced concrete containment structures are designed with due regard for the possibility of their testing in accordance with the requirements of federal rules and
regulations in the field of atomic energy use governing arrangement and operation of the NPP LSSs;
the permissible number of the containment system loading cycles within the entire service life with due regard for the acceptance and in-process testing;
it shall be demonstrated how the possibility for regular tensioning of the pre-stressed components is provided in the reinforced concrete containment structures made of pre-stressed reinforced concrete (in cases prescribed in the NPP design);
possibility for in-service inspection and replacement of pre-stressed reinforcement elements shall be demonstrated.
Information on any means provided for monitoring and recording of stress and strain condition and temperature of the reinforced concrete containment structures shall be specified.
Criteria for the possibility to operate the NPP power unit in case of any breakage of individual pre-stressed elements shall be specified for the reinforced concrete containment structures made of pre-stressed reinforced concrete.
Input data, methods and results of strength calculations confirming functional operability of the reinforced concrete containment structures shall be provided.
12.2.3.1.3. Embedded parts.
The list of regulatory documents used as the basis for design and manufacturing of embedded parts for the containment system shall be presented.
Information on the materials of embedded parts (strips, plates) affecting the containment system leak-tightness degree and substantiation for selection of these materials shall be provided.
Materials for embedded parts (anchors, strips, plates and other profiles) as well as for anchorage elements of the leak-tight steel lining not affecting the containment system leak-tightness degree shall be specified and selection of these materials shall be substantiated.
Information on the techniques and points for the leak-tight steel lining fastening to the reinforced concrete containment structures shall be provided.
Information on the fastening devices and (or) points for scaffolds, cradles and other appliances shall be specified.
12.2.3.1.4. Manholes, locks, doors and their embedded parts.
The following information shall be provided:
substantiation for selection of a particular containment system component; conditions taken into account to select the number of locks, manholes or doors;
purpose of each manhole, lock or door and imposed requirements for their leak-tightness; relevant drawings;
technique for fastening of the embedded parts (framing of apertures, door frames, embedded parts for locks) to the lining and fastening of the lock casing to the embedded part;
method and frequency of leak-tightness checks for these joints in the course of operation as well as their accessibility;
the possibility to control tightness of manholes, locks and doors from the outside in relation to the
ALA, for manholes and doors - after each opening-closing cycle;
it shall be substantiated that design of locks, manholes and doors with their embedded parts provide for the prescribed design leakage value and ionizing radiation weakening rate both under normal operation conditions and in case of any abnormal operation including design basis accidents;
information on the measures implemented in order to maintain structural integrity of locks, manholes and doors in case of any beyond design basis accident included into the final BDBA list;
design leakage value through locks, manholes and doors;
the door opening direction (into the ALA or visa versa), information on availability of position indication for the manhole covers and door leaves (sealed - unsealed) in the MCR and ECR and mechanical or electrical locking to prevent simultaneous opening of both lock doors; whether lock doors are equipped with pressure equalizing valves with their position indicators;
the possibility to activate opening-closing mechanisms of the lock doors (manholes) manually by one person both from outside and from inside the ALA or the lock;
the possibility for prompt evacuation of the personnel from the ALA in case of accidents;
permissible closing time for doors, manholes, locks;
information on emergency lighting and two-way communication system between the locks and the MCR and ECR;
the list of regulatory documents taken as the basis for strength calculations of the structures for manholes, locks, doors and their embedded parts;
the list of regulatory documents taken as the basis to develop anchorage for embedded parts of locks, manholes and doors;
information on elevations in relation to the floors in the rooms where locks, manholes and doors used for evacuation are installed. Potential water level on the floor occurring in case of any accidents shall be also specified here.
12.2.3.1.5. Penetrations.
The following information shall be provided:
all types of penetrations, diagrams and (or) drawings of these penetrations;
techniques for connection of penetrations with the embedded parts and the embedded parts with leak-tight steel lining;
substantiation of operability for pipeline penetrations with due regard for impacts from the connected pipelines;
substantiation of operability for the containment structures with due regard for thermal impacts from penetrations;
methods to control leak-tightness of weld joints in the course of manufacturing, installation and operation;
design leakage value through each penetration;
the way to arrange group electrical penetrations taking into account the physical separation
principle for safety channels.
12.2.3.1.6. Isolating devices.
All pipelines crossing the containment and any isolating devices installed on them shall be listed. Principles for installation of isolating devices on any pipelines crossing the containment shall be stated. The relevant diagrams and tables showing the number of isolating devices installed on each pipeline, their places of installation, drive types, required response time, actuation principle, position of the shutoff device in different NPP operation modes shall be provided. Information on the pipelines they are installed on shall be provided: diameter, working medium, flow direction (out of the containment or into it).
The list of pipelines where isolating devices are not installed shall be given, in this case the relevant substantiation shall be provided.
Information on the calculations used as the basis for selection of isolating devices and their response time shall be presented. The input data for calculations, calculation methods and programs shall be provided.
The following information shall be provided:
the list of initiating events requiring shutoff of the main pipelines crossing the containment; dependence of discharge from time in case of any failure of isolating devices shall be specified for each main pipeline;
the list of regulatory documents containing the requirements that the pipeline valves used as isolating devices shall comply with;
design leakage value for all types of isolating devices;
frequency of testing for isolating devices;
means and measures aimed to prevent unauthorized opening of isolating devices both during an accident and in the post-accident period due to loss of the drive power supply or any other causes;
it shall be substantiated that testing of isolating devices during power operation of the reactor (individually or within a SS channel) will not have any adverse effect on the NPP safety.
12.2.3.1.7. Bypass and safety devices.
The following information shall be provided:
purpose of the applied bypass and safety devices, places of their installation, description of their functioning;
substantiation of leak-tightness assurance for the rooms by the safety devices under the design basis accident parameters;
substantiation for selection of the number of safety devices, their throughput capacity, opening (closing) pressure;
the procedure, means and frequency of testing for safety and bypass devices;
techniques and frequency for replacement of sealing elements, inspection and repair with the reactor shut down;
methods for detection of any malfunctions of safety devices and safety measures in case of their
failures.
12.2.3.2. Systems of pressure reduction, heat removal, hydrogen removal and gas-aerosol treatment.
12.2.3.2.1. Passive steam condensation system. System of passive heat removal from the accident localization area.
Results of justifying calculations and experimental substantiation of operability for the passive steam condensation system (the system of passive heat removal from the ALA) in all operation modes shall be presented; it shall be demonstrated that the system ensures reduction of the medium pressure and temperature in the ALA to the values established in the NPP design.
It shall be substantiated that passive condensers of steam generated during accidents with the primary circuit integrity loss have sufficient reserve of cooling medium in order to ensure reliable condensing of all generated steam. Time of independent operation shall be specified for the system.
The requirements adopted in the design of the passive steam condenser walls shall be specified in case these walls form part of the containment and also in case condensing devices are installed in the tanks.
It shall be substantiated that pipelines, equipment, their fastening elements and other structures are designed for the impact of steam-air mixture stream and any other potential dynamic impacts.
Requirements for chemical composition of the solution used in the system shall be specified. Information on the filling and emptying systems for water discharge sumps of the passive steam condensation system shall be provided. Information on any measures provided in the NPP design in order to eliminate any non-homogeneity of the solution in the water discharge sumps, means for the solution purification and maintenance of its chemical composition shall be presented.
It shall be substantiated that passive steam condensers maintain their operability under design tilt of the reactor building. In this case maximum permissible vertical deviation of the steam supply devices for the entire service life of the NPP power unit shall be specified. In case any exceedance of the permissible vertical deviation is possible the method to correct position of steam supply devices of the passive steam condenser shall be described.
Information on the arrangements provided in the NPP design in order to prevent any damage of the system components due to hydraulic shocks possible during steam condensation as well as potential ALA vacuuming shall be presented.
Information on accessibility of the system component surfaces for repair and inspection shall be provided.
12.2.3.2.2. Active spray system.
Results of justifying calculations and experimental substantiation of operability shall be presented for the active spray system in all modes of its operation. It shall be substantiated that the system ensures reduction of the medium pressure and temperature and RS concentration in the ALA to the values established in the NPP design.
Requirements for chemical composition of the solution sprayed by the spray system shall be specified. Means for the solution purification and maintenance of its chemical composition shall be described. Information on the measures aimed to limit corrosive impact of the solution on the materials inside the containment shall be provided.
The following information on functional tests of the active spray system shall be provided: when
they are performed and with what periodicity, what parameters and (or) characteristics are checked, criteria of successful testing, description of the procedure and requirements of the documentation taken as the basis to perform tests.
Information on any operability checks for active components of the spray system performed in the course of the NPP unit power operation shall be presented.
Absence of any adverse impacts on the equipment related to operation of the spray system in the course of testing shall be substantiated.
Information on elimination of any possibility for loss of the containment tightness through the spray system pipelines in case of the spray pump failure to start upon an emergency signal shall be provided.
Systems for monitoring of thermal and technical parameters of the active spray system (pressure, temperature, flow rate) with indication of the instruments and sensors as well as for monitoring of chemical parameters (concentration of chemical additives) of water sprayed inside the ALA shall be described.
It shall be specified how homogeneous composition of the solution in the sump tank is ensured.
12.2.3.2.3. Water discharge sumps of the spray system pumps.
The following information shall be provided:
substantiation for selection of the design and number of the water discharge sumps for the spray system pumps with the use of water discharge sumps of other safety systems; it shall be substantiated that combination of functions by the components of these systems will not result in any violation of the NPP safety requirements;
description of the arrangements provided in the water discharge sump design for treatment of water supplied to pumps in order to remove mechanical contaminations and heat insulation washed from pipelines in case of rupture and to eliminate water loss in any NPP power unit operation mode;
substantiation of simultaneous operation of all spray pumps connected to this water discharge sump and pumps of other safety systems without any interruptions in supply with due regard for delayed water return to the water discharge sump from the ALA rooms.
Experimental data confirming operability of the sump tank (sump tanks) or pools with due regard with heat insulation washed from the pipelines during accidents shall be provided. Information on substantiation of the amount of heat insulation assumed in the course of experiments shall be provided. It shall be specified how homogeneous composition of the solution in the sump tanks is ensured.
12.2.3.2.4. Ventilation and cooling systems.
Results of justifying calculations and experimental substantiation of operability shall be presented for ventilation and cooling systems in all modes of their operation. It shall be substantiated that the systems ensure: heat removal from the ALA, underpressure in the ALA, reduction of RS concentration in the ALA, the required underpressure in the annulus (in case of the double containment at the NPP).
It shall be substantiated that ventilation and cooling systems comply with the requirements of federal rules and regulations in the field of atomic energy use governing arrangement and operation of safety-related NPP ventilation systems.
Permissible values of temperature, underpressure and RS concentration adopted in the NPP design that shall be provided by ventilation and cooling systems shall be substantiated.
The following information shall be provided:
operating characteristics of the ventilation systems (system components);
capacity;
pressure;
power;
resistance;
permissible working medium temperature range;
permissible working medium humidity;
characteristics of filtering units (filters) included into the system:
type of filters;
class of filters;
required efficiency;
initial and final filter resistance;
information on the filtration materials used: quantity and type of the applied filtration materials, service life, requirements for filtered air;
design limits:
filter resistance and air flow through the filters;
-radiation dose rate from RS accumulated in the filter;
RS concentration upstream and downstream of the filters;
permissible delay time for automatic switch-on of the back-up ventilation system components after failure of the main ones;
for filters (filtering materials) - information on the methods and frequency of repair and replacement as well as the requirements for the scope, methods and frequency of operating characteristics verification;
information on radiological monitoring in the ventilation systems: the list of control types, types of radiometric and health physics instruments, sampling points, and frequency of control.
12.2.3.2.5. Hydrogen removal system.
Selection of the system (systems) adopted in the design in order to ensure hydrogen explosion protection at the NPP shall be substantiated.
Information on the engineering features used in the hydrogen removal system shall be provided.
Principles and criteria of hydrogen explosion protection at the NPP shall be specified.
It shall be substantiated that the hydrogen removal system ensures performance of its functions in accordance with the requirements of federal rules and regulations in the field of atomic energy use
regulating hydrogen explosion protection assurance at NPPs and arrangement and operation of the NPP LSSs. Results of the hydrogen explosion protection substantiation shall be presented.
Information on the requirements imposed on the hydrogen concentration monitoring system shall be provided, and any means used in this system shall be described (information on the hydrogen concentration monitoring system and its components shall be presented in Chapter 7 of the NPP SAR).
It shall be substantiated that the hydrogen concentration monitoring system ensures performance of its functions in accordance with the requirements of federal rules and regulations in the field of atomic energy use regulating hydrogen explosion protection assurance at NPPs and arrangement and operation of the NPP LSSs.
The list of processes and sources causing hydrogen generation in systems, components and rooms located within the containment space under normal operation conditions and in case of any abnormal operation including accidents shall be provided.
The following information shall be provided with indication of the results of the relevant justifying calculations and experimental substantiation:
safe operation limits and conditions for the NPP with regard to the hydrogen-containing mixture parameters;
composition of explosion-hazardous hydrogen-containing mixtures in the rooms (with indication of the design volume of the rooms) located within the containment under normal operation conditions and in case of any abnormal operation including accidents (within the entire period of the accident);
information on hydrogen generation, accumulation and distribution as well as explosion hazard indicators of hydrogen-containing mixtures in the systems (components) and rooms located within the containment under normal operation conditions and in case of any abnormal operation including accidents (within the entire period of the accident);
number and locations of the hydrogen removal system equipment;
number and locations of the concentration monitoring means for hydrogen and other components (steam and oxygen) as well as parameters of the hydrogen-containing mixture (pressure and temperature) in the room located within the containment;
it shall be demonstrated how monitoring and control is arranged in the hydrogen removal system; actuation setpoints of protections, interlocks and alarms for the hydrogen-containing mixture parameters shall be specified;
mechanical and thermal loads on the RF containment due to combustion or explosion of hydrogen-containing mixtures in the course of accidents as well as any potential consequences of mechanical and thermal impacts on the building structures, systems and components; in this case measures aimed to protect systems and components against the above-mentioned impacts shall be described;
implemented technical measures aimed to eliminate any sources of explosion initiation for hydrogen-containing mixtures.
Information on experimental substantiation of the technical characteristics of the equipment and individual technical solutions applied in the hydrogen removal system with due regard for all possible operation modes shall be provided.
Information on the materials used in the equipment, pipelines and building structures as well as in their heat insulation and anti-corrosion coatings located in the rooms within the containment that can
be engaged into chemical reactions with hydrogen generation under normal NPP operation conditions and in case of any abnormal operation (including accidents) shall be provided. Information on the measures aimed to minimize application of these materials shall be presented.
For any systems and components providing hydrogen explosion protection conditions of their outage for maintenance, operability verification and repair shall be specified. In this case the minimum required set of equipment to ensure the NPP safety shall be specified.
For systems and components providing hydrogen explosion protection information on frequency and methods of their verification for compliance with the design characteristics shall be presented.
Information on life management for the systems and system components providing hydrogen explosion protection shall be presented.
12.2.3.2.6. Emergency gas and aerosol treatment plants.
Results of justifying calculations and experimental substantiation of operability shall be provided for emergency gas and aerosol treatment plants in all operation modes; it shall be substantiated that they ensure the following: reducation of medium pressure in the ALA, reduction of RS concentration in the ALA and RS releases into the environment. It shall be specified that filtering elements of the emergency gas treatment plant are accessible for replacement under normal operation conditions and in the post-accident period; in this case the required degree of tightness and biological protection shall be ensured for these components; methods for replacement and (or) cleaning of the used filters shall be described.
Information similar to the information presented in accordance with the requirements for ventilation and cooling systems specified in item 12.2.3.2.4 of this Appendix shall be provided.
12.2.3.2.7. Other rooms where the ALA is established.
In case of any other rooms at the NPP apart from the RF containment where the ALA boundaries are established and LSSs are provided they shall be described in accordance with the requirements of the standard system description framework given in Appendix 4 hereto and also in accordance with the provisions specified in items 12.2.1, 12.2.2, 12.2.3 of this Appendix.
12.2.4. Testing of the containment and its components.
12.2.4.1. General.
It shall be specified how the containment system and its components shall undergo verification for compliance with the design characteristics after manufacturing, in the course of commissioning, after repair and regularly within the entire service life of the NPP power unit.
The following information shall be provided:
information on seismic resistance verification for the containment components;
types of testing of the containment system and its components for compliance with the design characteristics, information on verification of the test methods;
the list of documents used as the basis for testing of the LSS components after their manufacturing, installation and in the course of operation.
Devices and (or) systems necessary for strength and tightness testing of the containment system shall be listed. Information on the parameters of testing media provided by these systems shall be specified.
References to the testing programs shall be given.
Subsequent to the results of the NPP power unit commissioning the testing results for the containment system components as well as references to documents and certificates containing the above-mentioned data shall be provided.
Information on the procedure for appointment of the persons responsible for the testing and the personnel engaged in the tests; procedures for the personnel admittance to performance of tests; procedures for admittance to inspection of the structures in the course of pressure or load increase shall be provided; it shall be specified where the personnel participating in the tests shall be in the course of load increase or reduction; locations of the control instruments used in the course of testing, requirements for safety assurance in the course of testing and requirements for the personnel's actions in case of any defect detection shall be described.
12.2.4.2. Strength tests for the containment system.
Cases when strength tests of the containment system shall be carried out and also criteria used as the basis to make decisions on performance of repeated strength tests shall be listed. The organization making decision of repeated strength testing shall be named.
The following information shall be provided:
criteria for assessment of the testing results (criteria for strength assessment based on visual inspection (for reinforced concrete structures) as well as criteria for assessment of stress and strain conditions based on the measured parameters shall be specified);
substantiation of the following values adopted in the NPP design: overpressure of the test medium, underpressure of the test medium, design pressure and underpressure in the containment;
overpressure or underpressure change rate in the course of testing;
pressure value (stage) when hold-up in the course of testing is carried out and duration of hold-up;
parameters registered in the course of strength testing.
Information on the testing procedure shall be provided.
Information on the means for measurement of stress, deformation (displacement) and inclination values shall be presented.
Information on the types of sensors for measurement of the stress-strain condition parameters shall be provided, and their measurement errors shall be indicated with reference to the relevant technical specifications.
12.2.4.3. Leak-tightness tests for the containment system.
Information on the methods used to determine the leak-tightness degree of the containment system shall be presented. It shall be specified that it complies with the leakage value determination accuracy. Information on validation of the method shall be presented.
It shall be specified when leak-tightness tests for the containment system under design pressure and underpressure are to be performed; frequency of in-service leak-tightness testing of the containment system under reduced pressure and underpressure shall be also indicated.
The testing procedure shall be described, and safety measures implemented in the course of
testing shall be specified. Values of the design gauge pressure and reduced gauge pressure shall be specified. Values of the design underpressure and testing underpressure shall be indicated for the containments systems tested for leak-tightness with underpressure.
Substantiation of the reduced testing pressure value adopted in the NPP design shall be provided.
The following information shall be provided in the testing procedure description:
the list of systems to be disabled for the period of testing (it shall be specified at what testing stage they shall be disabled);
how (by what signals) the isolating devices on the pipelines crossing the containment shall be closed in the course of leak-tightness testing; requirements for assurance of closed position of the hand-operated isolating devices shall be specified; it shall be stated at what testing stage the isolating devices are to be closed;
description of the systems and components used to create air overpressure and (or) underpressure inside the ALA;
criteria to determine stabilization of parameters inside the ALA;
the list of recorded parameters;
recording frequency for the parameters;
duration of testing: time to reach the required testing pressure (underpressure) in the containment; hold-up time necessary for stabilization of the parameters; time for performance of measurements and time to reduce the parameters down to the normal operation values;
description of the requirements and procedure for recording of the detected containment defects;
the number of testing pressure and underpressure stages for the containment leak-tightness testing in the course of commissioning works;
criteria for assessment of the containment leak-tightness testing results both in the course of commissioning works under design and reduced pressure and underpressure and under reduced pressure and underpressure in the course of operation;
pressure or underpressure increase and decrease rate inside the ALA in the course of leak-tightness testing;
description of the testing medium supply and discharge lines to and from the containment, the relevant diagrams, information on any filters used to discharge the testing medium from the containment in the course of the containment system leak-tightness testing during operation;
calculation algorithm for leakage value in the course of the containment system leak-tightness testing.
12.2.4.4. Leak-tightness tests for the containment system components.
All containment system components subject to leak-tightness testing shall be listed.
The following shall be provided: drawings containing detailed information on the design of each containment system component subjected to testing; the testing procedure; criteria of successful testing completion both in the course of the power unit construction and commissioning works and in the course of operation.
The following information shall be provided:
the list of cases when testing is to be performed;
requirements for the containment components with regard to their accessibility for such testing;
scope of testing for the containment components in the course of commissioning works;
scope of incoming control and post-installation testing and also acceptance criteria for the components;
frequency of testing for the containment components in the course of operation and criteria for performance of unscheduled testing.
12.2.4.5. Hydraulic testing of rooms, water discharge sumps and tanks.
Information on the rooms and tanks being the components of localizing safety systems and subject to hydraulic testing shall be provided. Cases when such testing is to be performed shall be specified.
Information on the filling and drainage systems for the above-mentioned components shall be provided.
The hydraulic testing procedure and parameters of the media used for testing shall be specified.
Criteria of successful testing as well as criteria for early termination of the tests shall be presented.
12.2.4.6. Functional tests for the water discharge sumps of the spray system pumps.
The following information shall be provided:
frequency of functional tests for the water discharge sumps of the active spray system pumps;
parameters and (or) characteristics checked in the course of testing;
testing procedure;
successful testing criteria;
testing intervals;
the list of documents containing the requirements used as the basis to perform testing.
12.2.4.7. Testing of the reinforced concrete containment structures as biological protection.
The following information shall be provided:
time for testing of the reinforced concrete containment structures as biological protection;
sections of the reinforced concrete containment structures subject to testing;
design ionizing radiation dose rates;
testing procedure, information on its validation, criteria for suitability of the reinforced concrete containment structures for operation as biological protection;
successful testing criteria.
12.2.5. Operation and maintenance of the containment.
The list of documents containing the requirements which govern operation and maintenance of the containment system shall be given; information on the measures for safe servicing of the containment system, maintenance of its operability, the content of the NPP power unit process regulations with regard to the containment system shall be provided.
The following information shall be provided:
scope and frequency of technical condition monitoring (including corrosive wear), maintenance, repair, testing and operability verification for the containment system components, success criteria for inspections and tests;
requirements for recording of the inspection and testing results;
basic requirements for the containment system operation guideline;
persons responsible for development, preparation, coordination and approval of the containment system operation guidelines;
information on the containment system availability for performance of its functions at any power level;
conditions of the containment system when the reactor start-up is prohibited;
scope of inspections prior to the reactor start-up after any repair works in the containment system;
conditions when unsealing of the containment is permitted; safety measures implemented in these cases and substantiation of their sufficiency;
the list of the containment system components with forbidden personnel access during power operation of the reactor;
the list of the containment system components with permitted personnel access during power operation of the reactor (the permissible duration of access shall be specified);
the list of controlled parameters in the containment systems and components in the course of the NPP unit power operation;
limitations for deviations of the process parameters of the containment system and its components, requirements for the minimum set of operable containment system components and also conditions for outage of the containment system components for maintenance, repair and testing (safe operation conditions for the containment system);
time (with substantiation) required to recover operability of the containment system upon expiry of which the reactor shall be brought to sub-critical state if operability of the system is not recovered;
information on the documents issued upon completion of repair works and functional checks of the repaired LSS component (or the entire containment system in case of necessity);
information contained in the certificates of the containment system components.
12.3. Supporting safety systems.
12.3.1. Each supporting safety system shall be described in accordance with the standard system description framework given in Appendix 4 hereto. Besides additional information required in
accordance with Section 12.3.3 of this Appendix shall be provided for each system under consideration.
12.3.2. Information on the following supporting safety systems shall be provided:
emergency power supply (with due regard for the information on this system or part thereof specified in Chapter 8 of the NPP SAR);
nitrogen and compressed air used as energy source for safety systems (if any such supporting systems are present at the NPP);
service water supply, cooling;
fire extinguishing in the safety system rooms;
supporting ventilation systems.
In case any other supporting safety systems are present at the NPP, information in accordance with the standard system description framework given in Appendix 4 hereto as well as in accordance with additional requirements stated in item 12.3.3 of this Appendix shall be provided for these systems.
12.3.3. Additional requirements for description of supporting safety systems.
12.3.3.1. Emergency power supply system.
The following information shall be provided:
performance of safety functions by the EPSS in case of design basis accidents;
performance of safety functions by the EPSS in case of any accidents accompanied by the auxiliary NPP power supply loss;
substantiation for duration (limited or unlimited) of operation in the period of accident (the required duration of the EPSS operation under the power unit blackout conditions as well as the required duration of independent EPSS operation for supply of power to the first-group consumers in case of the power unit blackout accompanied by failures of independent emergency power supply sources (diesel-generators) shall be specified);
minimum permissible grid parameters (voltage, frequency) ensuring power supply to the EPSS consumers;
substantiation of the time for non-attended operation of the diesel-generator;
substantiation for the list of non-disableable internal protections of the supporting SS components provided in the design;
system control functions performed manually (with time-limited prohibition of the operator's intervention or without any time limitations);
time period from issuance of the starting command to readiness for load acceptance, sequence of switching for the system and its consumers under the NPP blackout conditions in accordance with the step-by-step start-up program;
substantiation of the EPSS compliance with the federal rules and regulations in the field of atomic energy use governing the requirements for the NPP EPSS;
process flow diagrams for the SDGS systems.
12.3.3.2. Service water supply system. Cooling systems.
Information on heat removal from the system consumers to the ultimate heat sink shall be provided. In this case the following data shall be specified:
characteristics of all system consumers (thermal power to be removed from the consumer in all service water supply system operation modes);
cooling medium parameters (flow rate, inlet temperature, pressure);
information on the ultimate heat sink used in the system;
information on water purification from radioactive substances and mechanical impurities (characteristics of mechanical impurities, means of treatment, water exchange rate, measures aimed to prevent clogging of the system components and loss of their heat transfer and throughput capacity);
information on the system filling and makeup (volumes, flow rates in the course of filling and makeup);
information on stability of the applied materials;
information on generation of toxic and explosion-hazardous secondary decomposition products when the system and its components are subjected to the conditions different from the design ones (freon decomposition processes in the refrigerating units in case of a fire and any other processes with adverse effect on the NPP safety);
information on the required stock of expendables - spare parts, lubricants, cooling agents;
substantiation for selection of locations and capacity of drains and vents.
12.3.3.3. Supporting ventilation systems.
The following information shall be provided:
a) operating characteristics of the ventilation systems (system components):
capacity;
pressure;
power;
resistance;
permissible working medium temperature range;
permissible working medium humidity;
b) characteristics of filtering units (filters) included into the system:
type of filters;
class of filters;
required efficiency;
initial and final filter resistance;
data on the applied filtration materials particularly: quantity and type of the applied filtration materials, service life, requirements for filtered air;
c) design limits:
filter resistance and air flow through the filters;
γ -radiation dose rate from RS accumulated in the filter;
RS concentration upstream and downstream of the filters;
permissible delay time for automatic switch-on of the back-up ventilation system components belonging to safety systems after failure of the main ones;
information on the design frequency of repair and replacement, requirements for the scope, methods and frequency of operating characteristics verification for filters (filtration materials);
information on radiological monitoring in the ventilation systems: the list of control types, types of radiometric and health physics instruments, sampling points, and frequency of control.
Compliance of ventilation and cooling systems with the requirements of federal rules and regulations in the field of atomic energy use governing arrangement and operation of safety-related NPP ventilation systems shall be substantiated.
12.4. Special-purpose hardware for beyond design basis accident management.
12.4.1. The list of special-purpose hardware for beyond design basis accident management provided in the design shall be presented.
Each system referred to special-purpose hardware for BDBA management shall be described in accordance with the standard system description framework given in Appendix 4 hereto. Besides additional information in accordance with item 12.4.2 of this Appendix shall be provided for each system under consideration:
12.4.2. The following information shall be provided:
assessment of impact of the consequences of failure of any system (hardware) and its components referred to special-purpose hardware for BDBA management on operability of other systems and the NPP safety in general;
arrangements aimed to ensure independence of special-purpose hardware for BDBA management from normal operation systems and safety systems;
substantiation of the time required for actuation of special-purpose hardware for BDBA management;
assessment of protection of special-purpose hardware for BDBA management against accident effects as well as against any external impacts of natural and human-induced origin;
assessment of availability of the required transportation routes, their protection against accident effects and external natural and human-induced impacts, availability of tractive vehicles (for mobile special-purpose hardware for BDBA management);
sequence of the personnel's actions for actuation and use of special-purpose hardware for BDBA management;
assessment of efficiency (sufficiency) of special-purpose hardware for BDBA management in case
of accidents at a single NPP power unit and simultaneous accidents at several power units of a multi-unit NPP and also in case of accidents accompanied by disturbances of transportation lines and other utilities outside the NPP site.
12.4.3. Information on the core catcher with substantiation of the corium sub-criticality assurance and measures for residual heat removal shall be presented (in case any technical arrangements are provided in the NPP design in order to confine the corium inside the reactor vessel they shall be described and their efficiency shall be substantiated).
XIII. Requirements for the content of Chapter 13 "Commissioning of the NPP power unit"
Chapter 13 of the NPP SAR shall be submitted to the competent authority for state regulation of safety within the package of documents substantiating the NPP safety in the course of the NPP power unit construction or operation. Chapter 13 of the NPP SAR shall be updated with due regard for the results of works performed at the stages of the NPP power unit commissioning: pre-commissioning works, physical start-up, power start-up and pilot operation.
Subsequent to completion of the NPP power unit commissioning information confirming compliance with the requirements for sequence and scope of works performed at the stages of the NPP power unit commissioning specified in Chapter 13 of the NPP SAR as well as the requirements of Chapter 13 of the NPP SAR for arrangement, performance, control, assessment of results and acceptance of the works at the stages of the NPP power unit commissioning shall be included into the NPP SAR.
13.1. Arrangement of works and workers.
13.1.1. The organizational structure of the operating organization with segregation of the NPP organizational structure within it (in case these organizational structures are identical to the organizational structures presented in Chapter 14 of the NPP SAR reference to the above-mentioned chapter of the NPP SAR shall be given) and the organizational structure established by the operating organization at the NPP for arrangement, performance, control, assessment of results and acceptance of the works at the stages of the NPP power unit commissioning shall be provided. It shall be specified that the organizational structure established by the operating organization for the NPP power unit commissioning is endued with all required authorities, provided with funds, materials, equipment and human resources, regulatory documents and scientific and technical support. Information on responsibility for the NPP power unit commissioning activities established by the operating organization and information on supervision over these activities shall be presented.
13.1.2. Information on distribution of tasks, functions, authorities and responsibilities between the operating organization, the NPP management and the organizational structure for the power unit commissioning shall be provided.
13.1.3. Information on the procedure for interaction (with regard to the works performed in the course of the NPP power unit commissioning) between the operating organization (and the NPP operating personnel within it) and scientific, design, engineering, construction, installation and commissioning organizations, suppliers and other organizations engaged in the NPP power unit commissioning shall be presented. It shall be confirmed that all workers have the required professional skills and admitted to works with due regard for the implemented administrative and technical measures for quality assurance (safety-related) specified in the individual quality assurance programs of the organizations participating in the NPP power unit commissioning.
13.1.4. Prior to delivery of the nuclear fuel for the commissioned NPP power unit to the NPP it shall be confirmed that the NPP power unit is staffed with the NPP operating personnel having the required professional skills and admitted to unsupervised works in accordance with the procedure
established by the operating organization.
13.1.5. The plan for engagement of workers at each stage of the NPP power unit commissioning shall be provided. It shall be substantiated that the number of workers participating in the NPP power unit commissioning with due regard for the number of the NPP operating personnel is sufficient for safe performance, control, processing, assessment of results and acceptance of the works.
13.2. Scope and sequence of works.
13.2.1. Requirements for sequence and scope of works established and substantiated in the NPP design and ensuring achievement of the NPP power unit construction aim in a safe manner shall be specified. The NPP safety in the course of works shall be substantiated; in this case it shall be justified that the NPP safety in the course of works is ensured regardless of operability of any unchecked systems (components) related to safety of the commissioned NPP power unit.
13.2.2. Approach to classification of works as nuclear-hazardous shall be described.
13.2.3. Distribution of the works by stages and sub-stages of the NPP power unit commissioning as well as objectives, sequence and time limits for completion of these stages and sub-stages, criteria of the NPP power unit readiness for commencement of stages and sub-stages and their completion criteria, administrative and technical measures aimed to ensure the NPP power unit safety at these stages and sub-stages shall be presented.
13.3. Work programs.
13.3.1. Information shall be provided to confirm that works at the stages and sub-stages of the NPP power unit commissioning are prescribed in the NPP power unit commissioning program and performed in accordance with the programs containing the NPP safety assurance measures based on the NPP safety analysis (substantiation) for performance of these works.
13.3.2. The list of programs for pre-commissioning works, physical and power start-up and pilot operation with indication of the programs and procedures for individual works shall be provided. Information on the usage of accumulated experience in commissioning of similar NPP power units shall be provided and the commissioning programs for this particular NPP power unit and other NPP power units shall be compared.
13.3.3. Administrative measures and engineering features for the NPP safety assurance during the first delivery of nuclear fuel to the NPP and physical start-up of the reactor at the commissioned NPP power unit shall be specified.
13.3.4. Information on the methods for the first nuclear fuel loading into the reactor and the first bringing of the reactor to critical state shall be provided.
13.3.5. It shall be specified at what stages and sub-stages of the NPP power unit commissioning, in what ways and in what scope works on safety-related systems as well as tests of the NPP power unit in various states and operation modes under normal operation conditions, operation with deviations, and abnormal operation are performed.
13.3.6. States and operation modes of the NPP power unit that are impossible to be checked experimentally shall be specified, and it shall be substantiated that in these states and operation modes of the NPP power unit it is permitted not to perform operability verification for safety-related systems and components.
Permissibility of verification non-performance shall be substantiated by specific results of experimental studies and tests carried out on experimental installations (test beds) simulating the states
and operation modes of the NPP power unit.
13.4. Work performance diagram.
13.4.1. The work performance diagram for the NPP power unit commissioning indicating the period of time required for organization and performance of the works and documenting of the work results shall be provided. Besides information on the time required to develop detailed instructions for the NPP operating personnel in order to perform operations in the course of works as well as the time for approval of the adjusted programs, work procedures and operation documentation containing such instructions shall be also presented on the diagram.
13.4.2. Names of the works and information on commencement and completion dates for each work, its duration, sequence and interdependence of works as well as the critical path and milestones indicating the beginning and the end of the NPP power unit commissioning stages shall be presented on the work performance diagram.
13.5. Reporting documentation and acceptance of works.
13.5.1. Information on the types of reporting documentation in the course of the NPP power unit commissioning (acceptance certificates, certificates of readiness, work performance protocols, certificates of completion for works and stages (sub-stages), records of performance, certificates of compliance of the work results with design values of parameters, characteristics and algorithms), the procedure for its development, approval and storage shall be provided; it shall be also specified what particular information is contained in this documentation.
13.5.2. Information on the procedure for acceptance of works at the NPP power unit commissioning stages and transition from one NPP power unit commissioning stage to another as well as to the NPP power unit lifecycle phase - operation - shall be provided.
13.6. General NPP safety measures at the NPP power unit commissioning stages.
13.6.1. The following information shall be provided:
the procedure for actions in case of any deviations from design characteristics, parameter values and algorithms of functioning for safety-related systems and components detected in the course of works, non-compliance with the criteria specified in programs and (or) documentation of the RF and NPP designers, requirements of regulatory and other documentation;
the established procedure for investigation of any abnormal operation in the course of the NPP power unit commissioning;
measures for protection of the personnel, workers of any organizations engaged in the NPP power unit commissioning and the public in case of any accident in the course of the NPP power unit commissioning works;
arrangements aimed to restrict access of the personnel and workers of any organizations engaged in the NPP power unit commissioning to the NPP power unit rooms;
fire safety arrangements in the course of the NPP power unit commissioning works.
13.6.2. Sufficiency of the measures for isolation of the commissioned NPP power unit from the operating NPP power units shall be substantiated.
13.7. Assessment and accounting of the work results.
13.7.1. Subsequent to completion of the pilot operation performance of the NPP power unit
commissioning program shall be assessed. Information based on the reporting documentation and confirming performance of the entire planned scope and sequence of works prescribed in the above-mentioned programs and also confirming compliance with the criteria established to assess results of these works shall be provided.
13.7.2. Causes of non-compliance and (or) deviation from the requirements of the above-mentioned programs (if any have taken place in the course of works) shall be specified; their impact on the NPP safety shall be analyzed and conclusions in relation to program modifications for similar NPP power units shall be provided.
13.7.3. Information on any adjustments of the RF and NPP designs and operation documentation performed subsequent to the results of the NPP power unit commissioning works shall be presented.
XIV. Requirements for the content of Chapter 14 "Operation"
This chapter shall contain information on the organizational structure of the operating organization and the NPP, selection and training of the NPP operating personnel, operation documentation, maintenance and repair in the course of operation, arrangement of control and submittal of information on the NPP and the NPP power unit safety assessment as well as on emergency planning.
14.1. Organizational structure of the operating organization.
14.1.1. The organizational structure scheme of the operating organization showing structural units of the operating organization intended to perform activities for safe operation of the NPP directly at the NPP site shall be presented. The procedure for interaction between structural units of the operating organization performing the activities for assurance of safe NPP operation shall be described.
14.1.2. Information on distribution of tasks, functions, authorities and responsibilities among structural units of the operating organization and also on professional skills of the operating organization employees shall be provided.
14.2. NPP organizational structure.
14.2.1. Information shall be presented to confirm that the operating organization has arranged organizational structures intended to perform activities for safe operation of the NPP directly at the NPP, endued with all required authorities and provided with funds, materials, equipment and human resources, regulatory documents and scientific and technical support. Information on responsibility of the above-mentioned organizational structures for safe NPP operation activities as well as information on supervision over these activities shall be provided.
14.2.2. The NPP organizational structure scheme shall be presented. The procedure for interaction of the NPP structural units performing the activities aimed to ensure safe operation of the NPP including the units and (or) persons carrying out in-process control of the NPP and the NPP power unit shall be described.
14.2.3. For multi-unit NPPs any planned amendments and supplements introduced to the NPP organizational structure subsequent to commissioning of new NPP power units shall be indicated on the NPP organizational structure scheme.
14.2.4. Information on the NPP organizational structure and distribution of tasks, functions, authorities and responsibilities among the NPP structural units and also on professional skills of the NPP workers shall be provided.
14.2.5. Information on distribution of authorities and personal responsibilities of the managers and executives, on succession of authorities, on transfer of the right for issuance of directives and orders, on responsibility of the NPP management for the safe NPP operation activities shall be specified.
14.2.6. Information shall be provided to confirm that the number and composition of the NPP operating personnel admitted to unsupervised works comply with the minimum requirements for the personnel size and composition substantiated in the NPP design.
14.3. Selection and training of the NPP operating personnel.
14.3.1. Requirements for the NPP personnel qualification level shall be specified.
14.3.2. Prior to delivery of the nuclear fuel for the commissioned NPP power unit to the NPP it shall be confirmed that the NPP (NPP power unit) is staffed with the NPP operating personnel having the required professional skills and admitted to unsupervised works in accordance with the procedure established by the operating organization.
14.3.3. It shall be specified that persons from the operating personnel perform certain activities in the field of atomic energy use against presentation of permits granted by the competent authority for state regulation of safety in atomic energy use. It shall be specified that qualification requirements are established by the operating organization for the persons not required to obtain permits from the competent authority for state regulation of safety in atomic energy use and control over compliance with these requirements is arranged.
14.3.4 Information shall be presented to confirm that the system of recruiting and training of the NPP operating personnel arranged by the operating organization is aimed at achievement, control and maintenance of their qualification level required for safe operation of the power unit in all states and modes and for performance of any activities aimed at mitigation of accident consequences in case of their occurrence.
14.3.5. Information shall be presented to confirm that the system of recruiting and training of the operating personnel arranged by the operating organization provides for regular exercises and drills to train actions in case of accidents.
14.3.6. Information shall be presented to confirm that the system of recruiting and training of the NPP operating personnel arranged by the operating organization provides for accounting of any modifications in the documentation of the RF and NPP designers and operation documentation related to the NPP power unit as well as consideration of the analysis results for previous errors and erroneous decisions of the operating personnel.
14.3.7. Information shall be presented to confirm that formation of safety culture is a constituent element of the NPP operating personnel training.
14.3.8. Information on the training department, any simulators and other engineering features provided in the NPP design for the operating personnel training shall be specified. Analysis shall be provided to confirm compliance of the training facility with the requirements of regulatory documents as well as with state of the art in science, technology and production.
14.3.9. Information on the established procedure for the NPP personnel health monitoring shall be provided. It shall be confirmed that the operating personnel are subject to regular medical examinations (the in-process staff shall be also subject to pre-shift medical examination) and health condition of the persons from among the operating personnel ensures performance of their professional duties for the power unit operation.
14.4. Operation documentation.
14.4.1. Development of operation documentation.
14.4.1.1. It shall be specified at what stages of the NPP power unit construction the particular operation documents are developed and put into effect.
14.4.1.2. Information on the established procedure for the operation documentation development shall be provided. It shall be confirmed that prior to pre-commissioning works the NPP management arranges development of operation guidelines for the systems and equipment based on the process regulations for operation of the NPP power unit and documentation of the RF and NPP designers.
14.4.2. Process regulations for the NPP power unit operation.
14.4.2.1. The requirements taken into account in development of the process regulations for the NPP power unit operation, adopted approaches to definition of their structure and content shall be specified. Information on the aims and objectives of the process regulations for the NPP power unit operation shall be provided. Operation limits and conditions to be included into the process regulations for the NPP power unit operation shall be specified.
14.4.3. Operation guidelines for systems and equipment.
14.4.3.1. Information on the requirements taken into account in development of operation guidelines for systems and equipment and the adopted approaches to definition of their structure and content shall be provided. It shall be substantiated that operation guidelines for systems and equipment are provided for all safety-related systems and equipment and adjusted with due regard for the results of the NPP power unit commissioning.
14.4.3.2. It shall be substantiated that operation guidelines for systems and equipment contain specific instructions for the personnel on the ways to perform works under normal operation conditions and in case of any abnormal operation.
14.4.4. Maintenance, repair, testing and inspection regulations.
14.4.4.1. Information on the requirements taken into account in development of maintenance and repair regulations and the adopted approaches to definition of their structure and content shall be provided. It shall be substantiated that maintenance, repair, testing and inspection regulations are developed for all safety-related systems and components in order to maintain operability of safety systems and to prevent failures of safety-related systems and components.
14.4.4.2. Frequency and scope of the works for maintenance, repair, testing and inspection of safety-related systems and components shall be substantiated.
14.4.4.3. It shall be specified that calculation and (or) experimental operability verification has been performed for safety-related systems and components under the conditions identical to the conditions in case of natural impacts, external human-induced impacts typical for the NPP site and (or) any potential hydraulic, mechanical, thermal, chemical and any other impacts occurring due to accidents when operation of the above-mentioned systems and components is necessary.
In case factory tests and tests on experimental plants have been performed instead of testing at the NPP it shall be confirmed that the results of factory testing (tests on experimental plants) are representative (correct) for the systems (components) installed at the NPP.
14.4.5. Programs of nuclear-hazardous works.
Information on the adopted approach to development of nuclear-hazardous work programs shall be specified.
14.4.6. Operation documentation.
It shall be confirmed that the procedure for maintenance and storage of operation documentation is established and supported at the NPP.
14.4.7. Accident-prevention manuals and guidelines.
The requirements taken into account in development of accident-prevention manuals and guidelines (including the design basis accident mitigation guidelines and the beyond design basis accident management guidelines (particularly for severe accidents)) and approaches to development of their structure and content shall be specified. Accident management strategies (sequences of actions) provided in the design basis accident mitigation guidelines and the beyond design basis accident management guidelines (particularly for severe accidents) shall be substantiated. It shall be substantiated on the basis of the justifying calculation results that all accidents included into the final BDBA list are taken into account in the BDBA management guidelines. The relevant substantiation shall be presented in Chapter 14 of the NPP SAR and (or) as an appendix to the SAR with reference to it in Chapter 14 of the NPP SAR.
14.5. Maintenance and repair.
14.5.1. Organization of maintenance and repair.
14.5.1.1. Information on the maintenance and repair system for the NPP systems and components as well as information on planning and performance of maintenance and repair works, post-repair tests and inspections and quality assurance in the course of maintenance and repair works shall be provided. Information on the procedure for interaction between the structural units of the NPP and organizations performing maintenance and repair shall be specified.
14.5.1.2. Information shall be presented to confirm that the organizational structures intended to perform maintenance and repair activities are endued with all required authorities and provided with funds, materials, equipment and human resources, regulatory documents and scientific and technical support and are responsible for these activities and also carry out supervision over these activities.
14.5.2. Periodic maintenance and repair plans.
14.5.2.1. Periodic maintenance and repair plans shall be provided for the NPP systems and components with indication of the main types and scopes of activities.
14.5.2.2. It shall be specified how the previous operation experience is taken into account in development of periodic maintenance and repair plans for the NPP systems and components.
14.5.3. Conditions for maintenance and repair.
14.5.3.1. It shall be specified that outage of safety-related systems for maintenance and repair as well as tests and inspections are performed at the NPP in compliance with the safe operation conditions substantiated in Chapter 16 of the NPP SAR.
14.5.3.2. Information shall be provided in order to confirm that subsequent to maintenance and repair safety-related components and systems are checked for operability and compliance with the design characteristics, and the results of such checks are documented.
14.6. NPP safety control organization. Submittal of information on the NPP abnormal operation and NPP safety assessment.
14.6.1. Information on arrangement of the NPP and NPP power unit safety control performed by the operating organization and description of the procedure for submittal of information on the NPP safety assessment with due regard for any NPP abnormal operation to the state authority for regulation of safety in atomic energy use shall be presented.
14.6.2. Information on the units and officers of the operating organization performing the NPP safety control, their number and professional skills shall be provided.
14.6.3. It shall be confirmed that collection, processing, analysis, systematization and storage of the information on any erroneous actions of the personnel and failures of safety-related system components is arranged at the NPP, and this information is submitted to all concerned organizations in a timely manner according to the procedure established by the federal rules and regulations in the field of atomic energy use governing the procedure for investigation and accounting of any NPP abnormal operation; it shall also be confirmed that the operation experience is taken into consideration at the NPP.
14.6.4. It shall be substantiated that development and implementation of corrective measures aimed to prevent recurrent abnormal operation at the NPP is ensured.
14.7. Emergency planning.
14.7.1. Protection of the NPP personnel and the public.
14.7.1.1. Information on any planned and implemented administrative and technical measures for protection of the NPP personnel and the public in case of any accident at the NPP (NPP power unit, several NPP power units, any other nuclear facilities located at the NPP site) that can be accompanied with disturbances of infrastructure at the NPP site and outside it shall be presented.
14.7.1.2. The scheme of emergency response organizational structure, distribution of responsibilities and the procedure for interaction between the employees (officers) of this structure shall be provided; persons responsible for coordination of actions with any exterior organizations, emergency warning and implementation of the action plans for protection of the personnel and the public in case of accidents shall be specified. Information shall be provided to confirm that the number, qualification, psycho-physiological training and health of the employees and officers of the above-mentioned organizational structure ensure efficient emergency response in the states of emergency preparedness and emergency situation as well as under the impact of any external natural and human-induced factors corresponding to the emergency situation criteria and in case of any act of terrorism or threat thereof.
14.7.1.3. Criteria for declaration of emergency preparedness and emergency situation and levels of intervention shall be substantiated, and methods and means for notification of the personnel and the public shall be also specified.
14.7.1.4. Accident scenarios used as the basis to develop action plans for protection of the personnel and the public shall be described and representativeness of these scenarios, their relation to the BDBA analysis presented in Chapter 15 of the NPP SAR shall be substantiated; types and amounts of radioactive substances supplied to the NPP rooms and the environment in the considered accident scenarios, time for access and stay of people in the NPP rooms and at the NPP site, radiation exposure paths for the personnel and the public, characteristics of radioactive releases and exposure for people shall be provided.
14.7.1.5. Information shall be provided to confirm that the MCR, the ECR, shelters for the NPP
personnel and PERCPs at the NPP territory are equipped with air regeneration means and iodine absorbers.
14.7.1.6. Information shall be provided to confirm that sufficient number of the personnel is present at the NPP in order to assess the emergency situation, to perform corrective and protective actions, to apply special-purpose accident management hardware, to organize communications and documentation maintenance and also to render aid to the injured.
14.7.1.7. Criteria for evacuation of the personnel and the public shall be specified, and information shall be provided to confirm availability of evacuation routes, places for assembly and first aid treatment, shelters for the operating personnel, staff of military units and fire brigades, seconded personnel and any other personnel performing activities at the NPP, special-purpose vehicles with leak-tight cabins equipped with removable filtration and ventilation units and intended for transportation of people and delivery of foodstuff, radiation shelters for the personnel and their family members in the near-plant populated localities as well as availability of ARSMS at the NPP territory, in the sanitary-protective area and the supervised area.
14.7.1.8. Information on the arrangements aimed to ensure preparedness of the main and back-up evacuation routes and places (provision of the required amount of medicines, personal protection equipment, the required quantity and quality of the transportation routes (roads), control of the exposure dose levels for people and radiation contamination of the environment and foodstuff) shall be provided.
14.7.2. Protected emergency response control posts.
14.7.2.1. It shall be confirmed that prior to delivery of nuclear fuel for the commissioned NPP power unit to the NPP protected emergency response control posts with the necessary equipment, instruments and communication means have been arranged and maintained in constant preparedness with due regard for the information provided in the section of Chapter 8 of the NPP SAR performed in accordance with item 8.5 of this Appendix.
14.7.2.2. Locations of PERCPs at the NPP territory and in other near-plant areas shall be indicated; in this case it shall be confirmed that the PERCP locations are selected with due regard for any possible infrastructure disturbances outside the NPP site.
XV. Requirements for the content of Chapter 15 "Analysis of abnormal operation including design basis and
beyond design basis accidents"
Results of the analysis of any abnormal operation including design basis and beyond design basis accidents shall be provided in Chapter 15 of the NPP SAR.
It shall be substantiated on the basis of the analysis results for abnormal operation including design basis accidents that compliance with the design limits and safety criteria established in the NPP design is ensured for all operational states of the NPP in case of any abnormal operation including design basis accidents.
Analysis of abnormal operation including design basis accidents shall be performed for all initiating events included into the final list of initiating events for analysis of design basis accidents.
Analysis of beyond design basis accidents shall be performed for all accidents included into the final BDBA list.
Information on the BDBA management strategy developed on the basis of analysis and enabling to bring the NPP power unit back to the controlled state when the basic safety functions are performed, measures for the containment protection are implemented and radiological consequences of the accident are limited shall be provided. Recommendations for development of BDBA management guidelines and action plans for the personnel protection shall be provided on the basis of the BDBA analysis.
15.1. Abnormal operation including design basis accidents.
15.1.1. Final list of initiating events for analysis of design basis accidents.
The final list of initiating events of design basis accidents, the approach to compilation of this list shall be provided and substantiated; compliance of the adopted approach to the regulatory requirements and state of the art in science and technology shall be substantiated.
Information shall be presented to confirm that the final list of initiating events provided in the NPP SAR for analysis of design basis accidents was developed with due regard for:
the approximate list of initiating events for design basis accidents provided in Appendix 7 hereto (exclusion of any initiating event specified in the above-mentioned approximate list from analysis shall be substantiated);
analysis of failure consequences for the NPP components selected for subsequent detailed analysis;
operation experience for this NPP power unit and operation experience for similar power units;
probabilistic assessments.
It shall be substantiated that the following is taken into account in the final list of initiating events for analysis of design basis accidents:
all types of events capable of causing an accident namely: equipment failures, human errors, external impacts of natural and human-induced origin (from the sources located at the NPP site and outside it);
all possible locations of nuclear materials, radioactive substances and radioactive wastes where an accident can occur: RF, SFP, NF storage facilities, locations of nuclear fuel in the course of transportation, tanks and pipelines of the RW handling system, etc.;
all possible NPP operational states (operation at rated power, operation at reduced power levels, operation at the minimum controlled power level, shutdown for repair and refueling, "cold" state, "hot" state and other states defined in the NPP design).
15.1.2. Classification of initiating events.
Classification of initiating events by the type of their impact on the RF and probability of their occurrence shall be presented.
15.1.3. Design limits and safety criteria.
Information on the design limits and safety criteria adopted in analysis of design basis accidents shall be provided.
It shall be substantiated that the adopted design limits and safety criteria are established with due
regard for probability of the IE occurrence, in this case more stringent design limits and safety criteria shall be specified for more frequent initiating events.
15.1.4. Conservative approach assurance.
It shall be substantiated that safety analysis for each initiating event included into the final list of initiating events for analysis of design basis accidents is based on conservative approach.
It shall be substantiated that obtaining of conservative results in relation to compliance with design limits and safety criteria is ensured by selection of values for the NPP parameters and characteristics and (or) any other methods.
15.1.5. Presentation of the analysis results.
The following data shall be presented for each analyzed initiating event:
definition of the initiating event (cause of occurrence, level of disturbance in case of any parameter deviations, leakage size and location in case of a rupture, flow rate in case of unauthorized switch-on of a pump and any other characteristics and (or) parameters depending on the IE nature);
initial state of the NPP (determinant initial parameters, state of the NPP systems and components);
assumptions applied in the course of analysis;
name of the software tool used for analysis and reference to the ST validation certificate;
design limits and safety criteria applied to assess the analysis results.
Results of the analysis shall contain the following information:
a) transient process development history:
sequence of actuation for mechanisms and systems with indication of setpoints determining their actuation;
time limits for the SS operation commencement and completion;
moments of time corresponding to reaching of the maximum (minimum) values of the parameters characterizing the margins to the established design limits and safety criteria;
significant events that determine changes in the accident process development (SG drainage, core uncovery, failure (recovery) of natural circulation), sub-criticality reaching, re-criticality occurrence, etc.);
actions of the operating personnel (if any are provided);
b) text description of the process development based on the analysis of any changes in the representative parameters of the RF and NPP;
c) time-based graphs of changes in the representative parameters characterizing the state of the basic safety functions, physical barriers and also compliance with the design limits and safety criteria; the minimal list of parameters characterizing processes in the RF is given in Appendix 8 hereto;
d) substantiation for consideration of the single failure principle, consideration of any dependent failures and failures undetectable under normal operation conditions;
e) substantiation for sufficient duration of the calculation period (it shall be substantiated that at the analysis completion moment the NPP is in steady and controlled state, and in this case steady and controlled state of the NPP cannot change significantly due to minor change of any parameter, and besides no unavoidable threats to lose the controlled safe state not related to accidental failures of the NPP equipment exist);
f) conclusion on compliance with the design limits and safety criteria established in the NPP design;
g) necessary references to any documents confirming feasibility of the presented results.
For initiating events causing leakages of the primary or secondary circuit coolant within the containment development of processes inside the containment shall be described, and curves of changes of the representative parameters in the containment rooms shall be provided. The minimal list of the parameters characterizing processes inside the containment is given in Appendix 8 hereto.
In case the accident development conditions result in any thermomechanical deformation of fuel element claddings affecting the core cooling conditions information on the ways to consider these phenomena in the analysis shall be provided.
The number of leaky fuel elements shall be assessed on the basis of the analysis results. Fuel element leakage criteria adopted in the analysis shall be specified with reference to the relevant experimental justification of the adopted criteria.
Results of the radiological consequences analysis and description of the processes determining RS releases into the containment as well as from the containment to the environment shall be provided for design basis accidents leading to damage of fuel elements exceeding the established safe operation limits or to RS releases into the environment. Exposure doses for the personnel and the public under accident conditions shall be assessed.
Dose burdens for the public related to the consequences of design basis accidents shall be assessed for the critical group under the least favorable meteorological conditions (with the probability of 99.5%) typical for the NPP site and location area.
The minimal list of the presented data on radiological consequences of design basis accidents is given in Appendix 8 hereto.
15.2. Beyond design basis accidents.
15.2.1. Final list of beyond design basis accidents.
The final list of beyond design basis accidents (including severe ones) adopted in the NPP design shall be provided. It shall be specified that for multi-unit NPPs the list includes scenarios of accidents taking place simultaneously at several power units of the multi-unit NPP. Information on the approach adopted for development of the BDBA list shall be presented. It shall be substantiated that the adopted BDBA list includes representative scenarios for definition of measures aimed to manage such accidents and that representativeness of the scenarios is ensured by consideration of the NPP state severity levels and potential states of operability or inoperability of safety systems and special-purpose hardware for BDBA management.
Information shall be presented to confirm that the final list of beyond design basis accidents provided in the NPP SAR was developed with due regard for:
the approximate list of beyond design basis accidents given in Appendix 9 hereto;
analysis of failure consequences for the NPP components selected for subsequent detailed
analysis;
operation experience for this NPP power unit and operation experience for similar power units;
probabilistic assessments.
It shall be substantiated that the following is taken into account in the final BDBA list:
all types of events capable of causing an accident namely: equipment failures, human errors, external impacts of natural and human-induced origin (from the sources located at the NPP site and outside it), fires and floods;
all possible locations of nuclear materials, radioactive substances and radioactive wastes where an accident can occur: RF, SFP, NF storage facilities, locations of nuclear fuel in the course of transportation, tanks and pipelines of the RW handling system, etc.;
all possible NPP operational states (operation at rated power, operation at reduced power levels, operation at the minimum controlled power level, shutdown for repair and refueling, "cold" state, "hot" state and other operational states defined in the NPP design).
Results of realistic (non-conservative) analysis of beyond design basis accidents included into the final BDBA list containing probability assessment for the BDBA development scenarios and radiological consequences shall be provided in the NPP SAR. Dose burdens for the public related to the consequences of a severe accident shall be assessed for the critical group under the least favorable meteorological conditions (with the probability of 95%) typical for the NPP location area.
Analysis of beyond design basis accidents presented in the NPP SAR serves as the basis for development of action plans in order to protect the personnel and the public in case of any accidents and also for development of BDBA management guidelines. Thus efficiency substantiation for any actions prescribed for BDBA management and assessment of their radiological consequences shall be the main objectives of deterministic BDBA analyses.
The scope of requirements for presentation of analysis results for beyond design basis accidents not classified as severe accidents is similar to the requirements for analysis of design basis accidents presented in item 15.1.5 of this Appendix.
The list of additional data to be presented subsequent to the analysis results for severe accidents is given in Appendix 8 hereto.
Results of BDBA analysis presented in the NPP SAR shall demonstrate efficiency of the BDBA management measures provided in the NPP design.
In case the NPP personnel intervention is needed to bring the NPP into controlled and safe state under the conditions of the considered beyond design basis accident both analysis results for the scenario disregarding the NPP personnel's actions (in order to determine the main BDBA development stages and available time reserves for the NPP personnel prior to any accident management actions) and results of the analysis with due regard for the NPP personnel's actions (in order to confirm efficiency of the accident management measures) shall be presented in the NPP SAR.
Information on any special-purpose hardware for BDBA management as well as any other engineering features existing at the NPP for BDBA management that have been taken into account in the accident analysis shall be provided. Priority for use of various systems and equipment for accident
management shall be established, and the time required for activation of the above-mentioned engineering features shall be assessed.
Requirements for the BDBA management actions shall be defined for each NPP state severity level.
The scope of information on the NPP state required to control the NPP state and to manage beyond design basis accidents shall be determined, engineering features and methods enabling to obtain this information shall be specified. If direct measurement of the required parameters is impossible the possibility for indirect assessment of the required parameters shall be taken into account, and the methods for such assessment shall be described.
It shall be substantiated based on the BDBA analysis results that the BDBA management measures provided in the NPP SAR ensure achievement of the BDBA management objectives: bringing the NPP back to safe and controlled state, implementation of achievable measures for the containment protection, mitigation of the radiological accident consequences.
15.3. Software used.
The list of applied software tools with indication of information on their validation shall be presented. The number of the ST certificate, date of issuance and the ST certificate validity period shall be specified.
In case any software tool or calculation method is not validated in accordance with the established procedure the scheduled time limits for validation shall be specified, and the following necessary information confirming the possibility to use the method or software tool for safety analysis within its application scope shall be presented:
description of physical processes and phenomena modelled through the use of the software tool;
description of the mathematical models with indication of the constitutive relations;
description of the calculation nodalization schemes used for safety analysis; applicability of the calculation schemes used for analysis shall be confirmed by the sensitivity analysis results;
basic results of the ST verification (may be presented in a separate report);
information on the ST application restrictions and maximum calculation errors determined on the basis of verification results;
results of uncertainty analysis for the calculations.
The following information shall be provided for all software tools used to perform safety analyses:
the ST purpose and scope of application, the ST application restrictions;
brief description of the procedures and programs (reference to the sources containing more detailed description of software tools shall be given);
nodalization schemes used for analysis (their applicability for the performed analysis shall be substantiated).
15.4. Input data for the analysis.
The input data for the analysis within the scope necessary to perform independent analysis shall be provided in an individual section.
The following data shall be presented in this section:
main parameters and characteristics characterizing initial state of the NPP (for all NPP states considered in the analyses);
geometric and topological input data;
physical input data;
technological input data.
The minimal list of input data to be presented in the NPP SAR is given in Appendix 10 hereto.
XVI. Requirements for the content of Chapter 16 "Safe operation limits and conditions. Operation limits and conditions"
Information on safe operation limits and conditions as well as operation limits and conditions established in the NPP design shall be provided in Chapter 16 of the NPP SAR. Information shall be presented for all safety-related systems and components. Operation limits and conditions and safe operation limits and conditions shall be specified for all NPP operational states (operation at rated power, operation at reduced power levels, operation at the minimum controlled power level, shutdown for repair and refueling, "cold" state, "hot" state and other operational states defined in the NPP design).
Safe operation limits and conditions as well as operation limits and conditions shall be substantiated.
16.1. Safe operation limits.
16.1.1. The list of controlled parameters and their limits.
The list of safe operation limits shall be given.
The list of controlled parameters for which the safe operation limits are established, the method and exact points of their measurements, substantiation of the adopted value and its measurement accuracy, ranges of the parameter change and measurement, accuracy of the performed justifying calculations and (or) experimental substantiation for the parameter, permissible interruption with information loss, redundancy of the measuring channels shall be presented.
The list of parameters for which the safe operation limits are established and the values of the above-mentioned limits shall be substantiated separately for each NPP operational state.
16.1.2. Safety system actuation setpoints.
Actuation setpoints shall be specified for safety systems. The adopted setpoint values shall be substantiated, any modes (processes) that determine reaching thereof as well as their measurement accuracy, measurement points, redundancy of the measuring channels and the principle for generation of the SS actuation command shall be provided. Actuation setpoints of the warning and emergency alarms shall be specified with substantiation of the interval to the SS actuation setpoint values.
Correspondence of the safe operation limits established in the NPP design and setpoints of safety systems shall be substantiated.
16.2. Operation limits and conditions.
16.2.1. Limit values of the process parameters.
Operation limits and conditions shall be specified for safety-related systems. The selected values of parameters for different operational states of the NPP shall be substantiated, their measurement accuracy, measurement points, redundancy of the measuring channels and permissible information loss period shall be specified.
16.2.2. Process protections, interlocks and automatic controllers with their actuation setpoints.
Values of the process parameters when actuation of the main process protections, interlocks and automatic controllers is prescribed in different NPP operational states shall be specified and substantiated. Information on the location of sensors, their redundancy and power supply shall be presented. Actuation setpoint values for the warning alarms shall be specified, the range between actuation setpoint values of the warning alarms, process protections and interlocks as well as actuation setpoint values of safety systems shall be substantiated.
16.3. Safe operating conditions.
16.3.1. Power levels and operational states of the NPP.
Information on the NPP operational states (operation at rated power, operation at reduced power levels, operation at the minimum controlled power level, shutdown for repair and refueling, "cold" state, "hot" state and other operational states defined in the NPP design) and the corresponding permissible RF power levels shall be provided. Definitions of the above-mentioned NPP operational states shall be presented.
Operation limits and conditions for the following main RF parameters shall be specified for operational states of the NPP and for each RF power level (with due regard for the information provided in the section of Chapter 16 of the NPP SAR developed in accordance with the requirements of item 16.2 of this Appendix): the reactor power, energy emission distribution, reactivity coefficients, the primary circuit coolant pressure, pressure change rate, coolant temperature, temperature change rate, chemical composition, leaks through the primary circuit pressure boundaries, radioactivity of the primary circuit coolant, radioactivity of the secondary circuit working medium.
It shall be specified that the above-mentioned operation limits and (or) conditions are expressed via values of the parameters controlled by the NPP personnel, otherwise information on the relation between the limit and (or) condition and directly controlled parameters shall be provided through the use of relevant tables, diagrams or calculation methods.
Any imposed restrictions for the permissible RF power levels and the NPP operational states shall be substantiated with references to the relevant sections of the NPP SAR.
16.3.2. Configuration and condition of operable NPP systems and components required for start-up and operation in the NPP operational states. Permissible loading cycles.
Information on the configuration and condition of safety-related systems operability or availability of which is required for the NPP (NPP power unit) start-up and operation in the NPP operational states specified in the NPP SAR in accordance with the requirements of item 16.3.1 of this Appendix shall be provided.
The following information shall be specified for each safety-related system: configuration and number of components (subsystems, channels) operability of which is required for the NPP (NPP power unit) start-up and operation in the permissible normal operation modes; requirements for their characteristics.
Conditions related to permissible loading cycles of the main RF equipment shall be specified with due reagrd for the design lifetime.
16.3.3. Conditions for maintenance, monitoring and testing of safety-related systems and components.
Requirements for the scope, frequency and other conditions for maintenance, monitoring and testing of safety-related systems and components shall be specified.
Information on the methods and means for performance of these works and any existing NPP operation restrictions (if any) shall be provided.
Requirements for metal condition control for the RF equipment and pipelines shall be presented.
16.3.4. Safe NPP states that the NPP power unit shall be brought to in case of any non-compliance with safe operation limits and conditions.
Characteristics and description of safe NPP states provided in the design that the NPP power unit shall be brought to in case of any non-compliance with the established limits and conditions in the course of the reactor operation shall be specified.
16.4. Documentation of the information on monitoring of safe operation limits and conditions.
Requirements for the procedures for recording of any deviations from safe operation limits and conditions and control of their observance shall be specified.
XVII. Requirements for the content of Chapter 17"Quality assurance"
Requirements for the information on quality assurance for all works and services affecting the NPP safety shall be presented in Chapter 17 of the NPP SAR.
17.1. General.
17.1.1. Information on any measures aimed to ensure that the NPP siting, design, construction, operation and decommissioning as well as development and manufacturing of safety-related systems (components) is carried out and planned to be carried out in accordance with the quality assurance requirements shall be provided.
17.1.2. Information on the areas of activities for quality assurance described in the section of Chapter 17 of the NPP SAR performed in accordance with item 17.2 of this Appendix shall be provided.
17.1.3. Chapter 17 shall be divided into sections with the titles corresponding to areas of activities for quality assurance in accordance with item 17.2 of this Appendix.
17.1.4. Information provided in Chapter 17 of the NPP SAR shall be prepared taking into account the analysis results for the general and individual NPP QAPs and their implementation as of the NPP SAR development moment.
17.2. Requirements for the information on quality assurance activity areas.
17.2.1. Quality policy.
Information on the quality policy adopted by the operating organization shall be provided.
It shall be specified that the quality policy is coordinated with other areas of the operating organization activities, brought to the knowledge of executives and establishes the following:
principles and aims defined for the purpose of safety assurance as top-priority in relation to other aims of the activities in the field of atomic energy use;
main quality objectives and the methods to achieve them;
liabilities of the operating organization management in the area of quality assurance.
17.2.2. Organizational activities.
17.2.2.1. Organization of works.
Information on the main functional duties, authorities and responsibilities of the officers (units) of the operating organization managing the QAP development and implementation and also controlling fulfillment and assessing efficiency of the QAP implementation as well as information on distribution of responsibility among officers (units) performing works and controlling their quality shall be presented.
The section shall contain the following:
the list of documents defining the form of incorporation for the operating organization and the NPP;
information on the structure of quality assurance departments in the operating organization and any organizations performing works and rendering services to the operating organization;
description of the procedure for interaction of the operating organization with superior organizations, contractors, any other organizations performing activities in the field of atomic energy use, executive authorities engaged in state control of atomic energy use and state regulation of safety in atomic energy use;
infrastructure of the operating organization formed by specialized enterprises and organizations to which it partially delegates its functional duties, authorities and responsibility retaining overall responsibility without any prejudice for liabilities and legal responsibility of contractors;
distribution of responsibility for quality assurance between organizations performing works and rendering services to the operating organization;
description of the procedure for distribution of works affecting the NPP safety assurance and interaction during performance of these works between production units of the operating organization or organizations performing the works and rendering services to the operating organization as well as registration of such interaction in the provisions on production units, job descriptions for the workers and (or) any other organizational and administrative documents.
17.2.2.2. Quality management system of the operating organization.
The following information shall be provided:
structure of the quality management system for the operating organization and its branches;
the list of the main quality management system documents of the operating organization;
reporting documentation containing analysis of the quality management system efficiency for the operating organization and its branches, results of its inspections and corrective measures.
17.2.2.3. Quality assurance programs.
The section shall contain the following:
information on development, issuance and results of the NPP QAP implementation verification;
information on implementation of the general and individual NPP QAPs as of the moment of the NPP SAR submittal by the Applicant;
information on the level of the NPP QAP compliance with the provisions established by federal rules and regulations in the field of atomic energy use;
the scope of the NPP QAPs developed and planned for development as of the NPP SAR submittal moment;
information confirming that any activity affecting safety-related systems (components) is subject to the relevant control within the NPP QAP framework;
description of the measures taken by the operating organization in order to ensure current implementation of the NPP QAP.
17.2.3. Human resources management.
Information on any procedures established in the operating organization for HR management with regard to the following shall be provided:
safety culture formation and maintenance;
training, knowledge and skill checks for the workers engaged in any works affecting the NPP safety assurance and supervision over these works as well as workers carrying out tests, inspections and verifications;
determination of the needs for training of workers and organization of training, retraining, development of competence and certification of workers;
analysis of programs for training, retraining, development of competence and certification of workers;
maintaining records on training, retraining, development of competence, and certification of workers.
This section shall contain information on any requirements for the personnel's responsibilities, qualification, scope of knowledge and skills contained in job descriptions for the personnel and (or) any other documents of the operating organization.
17.2.4. Documentation management.
This section shall contain description of any procedures effective in the operating organization for development, coordination, approval, entry into force, identification, modification, revision, distribution, storage and destruction of expired documents.
The following information shall be provided:
the list of regulatory and technical documents effective in the operating organization and applied to carry out any activities related to quality assurance;
information on analysis of regulatory and technical support performed by the operating organization;
description of the measures taken by the operating organization in order to ensure development of the missing regulatory documents as detected in the course of analysis.
The following information shall be provided:
description of the procedure for control of quality assurance information concerning the NPP quality assurance issues;
information on the procedures effective in the operating organization for issuance and maintenance of the quality assurance documentation (determination of the type of records depending on their importance, identification, collection, indexing, access, filing, storage, maintenance and disposal of the registered quality data: results of inspections, tests, verifications of processes, analysis of the supplied equipment, components and materials);
information on the procedure for recording, storage and issuance of documentation that shall be maintained in accordance with the written procedures;
description of the NPP QAP implementation reporting system.
17.2.5. Design (development) control.
The section shall contain the following:
description of the measures (procedures) planned and implemented by the operating organization for design (development) control that shall prescribe verification of correctness of substantiation for the applied solutions as well as their compliance with the design requirements;
description of the control methods for input and output data, the design (development) process, verification of the design and engineering documentation, compliance with the requirements of federal rules and regulations in the field of atomic energy use, terms of reference (or any other document containing the necessary and sufficient requirements for product development) and any other documents specified in the quality assurance program;
verification of compliance with the requirements for documenting of the results of the design and engineering documentation inspection so that to enable investigation or revision of the verification method after its completion;
verification of compliance with the requirements for time limits of the inspections that shall be completed after prototype or commercial prototype testing prior to issuance of any documentation for its manufacturing or the NPP construction;
check of compliance with the criteria of mandatory testing provided for verification of the design and engineering documentation, the necessity to ensure representativeness of tests and modelling of the least favorable conditions defined on the basis of safety analysis;
description of the measures aimed to define and control delimitation of works in the course of design (development);
information on availability and performance of the procedure for control of any amendments to the design and engineering documentation as well as modifications of the NPP buildings, structures and safety-related systems (components) (in the course of their design, manufacturing, installation, repair, refurbishment, retrofitting and replacement).
17.2.6. Procurement management with regard to equipment, components, materials, semi-finished products, documents and software, as well as services rendered.
The section shall contain description of the following procedures effective in the operating organization:
procurement arrangement with regard to equipment, components, materials, semi-finished products and software as well as provision of services, selection of organizations performing works and rendering services to the operating organization;
identification, assurance of control and testing completeness, traceability of the control and testing results for the purchased equipment, components, materials, semi-finished products, documents and software;
storage, transportation, preservation, packaging of equipment, components, materials and semi-finished products;
organization of control over compliance with the requirements for the rendered services.
17.2.7. Production activities.
This section shall contain information on any procedures effective in the operating organization for planning, performance and control of production activities and special processes:
designer supervision and maintenance service;
technical examination, maintenance, repair, refurbishment and retrofitting of the systems (components);
lifetime extension for safety-related systems (components);
validation of welding techniques, methods for the equipment monitoring and diagnostics.
This section shall contain description of any procedures effective in the operating organization in order to ensure performance of tests, technical control, in-service inspection of metal and checks of safety-related systems (components) in accordance with the established schedules and criteria.
The following information shall be provided in this section:
the list of tests, trials and checks for safety-related systems (components) performed in order to confirm their operability;
information on the ways to reflect the product operation model, requirements for metrological support, acceptability conditions for the testing results and representativeness of tests in the testing programs;
methods for registration and documentation of the testing results and assessment of their acceptability;
references to the testing reports and description of the testing results.
17.2.8. Metrological support.
The following information shall be provided in this section:
availability of the list of measuring instruments and their verification (calibration) procedures;
availability of the list of testing equipment and it validation programs and procedures;
availability of the provision on identification of the measuring instruments and testing equipment;
on the following procedures effective in the operating organization:
arrangement of accounting, verification (calibration), validation and identification of the measuring instruments and testing equipment and also accounting and validation of the measuring procedures (methods);
maintenance of the measuring instruments and testing equipment;
accounting and storage of verification certificates, calibration certificates for the measuring instruments, testing equipment validation reports.
17.2.9. Quality assurance for software tools and calculation methods.
The following information shall be provided in this section:
description of any quality assurance procedures for software and calculation methods effective in the operating organization;
the list of calculation methods and validated software tools used for the NPP safety analysis and (or) assurance and also for assessment of any characteristics of systems (components) as well as their strength, durability and failure-free performance;
information on usage of any validated databases for development of programs;
information on assimilation and implementation of alternative domestic and foreign programs;
description of the procedure aimed to train the executives in state-of-the-art methods for solving calculation tasks.
17.2.10. Reliability assurance.
This section shall contain the following information:
information on any procedures effective in the operating organization in order to ensure reliability and control compliance of reliability indicators for safety-related systems (components) with the established requirements within the entire service life (with due regard for extended service life);
information on the system for collection, registration, processing, accumulation, storage, analysis and distribution of information on reliability of safety-related systems (components) to the concerned organizations performing works and (or) rendering services to the operating organization.
17.2.11. Non-conformity management.
The section shall contain information on the following procedures effective in the operating organization and aimed to provide:
detection and registration of any deviations from the requirements for quality of works (services) and (or) equipment (design or manufacturing errors, defects and failures of equipment, abnormal operation modes, human errors);
determination and analysis of causes of the detected non-conformities (with due regard for impact of non-conformities on the NPP safety);
prevention of usage of any products that do not comply with the established requirements (procedure for segregation, disposal, documenting, identification of such products) or acceptance of any works and (or) services that do not comply with the established requirements;
arrangement of the system for collection and processing of data on any detected non-conformities, violations, defects, their causes and implemented corrective measures;
development, performance and control of implementation of the corrective measures aimed to prevent recurrent non-conformities and preventive actions and also analysis of their efficiency;
notification of the management at the relevant level and management of the concerned organizations on any detected nonconformities and implemented corrective and preventive actions.
Information on any cases of decisions on detected non-conformities registered as of the NPP SAR submittal moment with due regard for analysis of their causes, implemented corrective and preventive measures with account of their efficiency analysis as well as analysis results for any tendencies in changes of causes and nature of non-conformities shall be also provided.
17.2.12. Inspections (audits).
Information on any procedures effective in the operating organization in order to carry out and record the results of independent inspections (audits) for the general and individual NPP QAPs containing criteria for their efficiency assessment shall be presented.
Information on the results of inspections and implementation efficiency assessment for the general and individual NPP QAPs as per the NPP SAR submittal moment shall be provided with indication of the developed corrective and preventive measures.
XVIII. Requirements for the content of Chapter 18"Decommissioning"
The following information shall be provided in Chapter 18 of the NPP SAR:
concept of the NPP power unit decommissioning;
design solutions aimed to ensure low activation level for the NPP components within the entire service life of the NPP power unit and minimization of the generated RW amount;
radiation safety assurance for the NPP personnel, the public and protection of the environment (after NF removal) at all stages of the NPP power unit decommissioning;
on very low level wastes and materials for unrestricted use (reusable).
18.1. Decommissioning concept.
Information on the NPP power unit decommissioning concept, sequence of actions for decommissioning of the NPP power unit and the basic criteria of radiation safety assurance in the course of these actions shall be provided in the NPP SAR.
Information on any possible variants of the NPP power unit decommissioning and expected end states after their implementation shall be presented.
The selected variant of the NPP power unit decommissioning shall be substantiated based on the analysis of technical and economic comparison of possible decommissioning variants.
The decommissioning program shall be provided for the selected NPP power unit decommissioning variant and the time limits for the following arrangements shall be specified:
comprehensive engineering and radiation survey of the NPP unit;
development of the report subsequent to the final examination results for the decommissioned NPP unit.
It shall be specified how the following is ensured at all stages of the NPP power unit decommissioning:
reduction of dose burdens for the personnel and the public in accordance with the optimization principle;
limitation of RS releases and discharges into the environment to the levels complying with the requirements of regulations;
prevention of accidents and mitigation of consequences in case of their occurrence.
Procedure for collection and storage of information important for safety assurance in the course of the NPP power unit decommissioning in the NPP power unit decommissioning database shall be described.
Information on the ways to eliminate impact of the NPP power unit decommissioning works on safety of other NPP power units operated at the NPP site shall be provided.
It shall be specified that the NPP power unit operation experience is taken into account in the course of the works for its decommissioning.
18.2. Design solutions.
Information on selection of the materials for manufacturing of the NPP power unit equipment, systems and structures ensuring low level of their activation within the entire service life of the NPP power unit and minimization of RW amounts generated due to activation shall be presented.
Content of chemical elements (basic, additional and traces at the level of 10 -2 to 10-5 mass %) in the materials of the reactor internals, reactor vessel (carbon and special steels), concrete (ordinary or special) of the protections and other building structures shall be assessed in order to substantiate radiation safety and minimize RW amounts generated in the course of decommissioning. Information on radioactive wastes generated during dismantling of equipment and protective building structures and decontamination of equipment and rooms as well as on exposure doses for the personnel during dismantling of these components and handling of radioactive wastes shall be specified.
Analysis results for the possible variants of radionuclide amount reduction in steel structures due to neutron absorption by the RF materials shall be provided:
replacement of alloys with high cobalt content used in reactor building by alloys with low cobalt content or alloys without cobalt;
decrease of cobalt, silver, niobium and nickel content in the structural materials.
Analysis of lithium content in the materials of the near-reactor protective building structures shall be provided as it becomes the source of tritium after neutron absorption.
Information on the structural solutions ensuring minimization of surface contamination of the NPP power unit equipment, systems and structures with radioactive substances in the course of its operation shall be provided.
Information on the places reserved at the NPP site (in case of necessity) for potential installation of plants and storage facilities intended for processing and storage of radioactive (solid and liquid) wastes generated in the course of the NPP power unit decommissioning shall be specified.
Information on the technical solutions provided in the NPP design in order to facilitate dismantling of the protective building structures shall be specified, and efficiency analysis shall be presented for these solutions.
The adopted approaches for limitation or complete elimination of serpentinite, chromite and magnetite use in the protective building structures (radiation protections) due to high content of cobalt and iron shall be specified and their application shall be substantiated.
Information and substantiation for selection of the places for location of surveillance specimen enabling to ensure reliable assessment of radionuclide content in the structures and equipment due to neutron activation shall be provided.
It shall be specified how the minimum possible portland cement content in concretes is ensured in order to reduce their neutron-induced activity and exposure doses for the personnel.
18.3. Radiation sources.
Input data and calculation (assessment) results with regard to activity of the materials of equipment and protective building structures as well as assessment of radiation field intensity from these components, total amount of radioactive wastes and their isotope composition shall be provided; amounts of the materials of unrestricted use (reusable) subject to disposal shall be also assessed. It shall be specified that the calculations are performed for activating neutron energies within the entire reactor spectrum with its division into groups corresponding to the preliminary neutron flux density calculation groups. The obtained calculated data on neutron activation of equipment and protective building structures as well as on the relevant dose rates shall include dependencies of the calculated data from the time period after the NPP power unit reactor shutdown. It shall be specified that validated software tools have been used for calculations.
Approximate estimates of radionuclide contamination shall be provided for the equipment, protective building structures and the NPP power unit rooms based on the decommissioning experience for similar NPP power units and their radiation surveys.
Amount and particle size of radioactive aerosols generated in the course of works for dismantling of equipment and structures during the NPP power unit decommissioning shall be estimated based on expected metal and material cutting and breaking techniques and data on the particular equipment used for these purposes.
18.4. Safety assurance.
It shall be specified how radiation safety of the personnel and the public and protection of the environment (after NF removal) is to be ensured at all stages of the NPP power unit decommissioning.
Information on the scope of radiological and health physics monitoring based on the analysis of ionizing radiation sources and characteristics of radioactive aerosols shall be provided.
It shall be specified that the proposed radiological monitoring system complies with any requirements imposed on it and will retain its operability after the NPP power unit shutdown within the entire period of its decommissioning.
It shall be specified that the radiological monitoring system can ensure measurement of the following parameters characterizing the radiation situation in the course of the NPP power unit decommissioning:
RS activity of the buildings and facilities, structures, systems and equipment as well as gamma radiation dose rates in the rooms;
gamma radiation dose rates from individual reactor internals assemblies, the reactor vessel and their fragments in the course of dismantling, segregation and transportation;
level of radioactive contamination of the work room and equipment surfaces with beta-active nuclides;
level of radioactive contamination of the work room and equipment surfaces with alpha-active nuclides;
specific volumetric activity of aerosols in air;
specific volumetric activity of aerosols in the vent stack.
It shall be demonstrated that ingress of radionuclides from the list of radionuclides subject to state regulation measures in the area of environmental protection and released (discharged) due to decommissioning works into the environment is controlled in the course of decommissioning.
Assurance of lifetime, operability and maintainability of any systems and equipment required for the NPP power unit decommissioning within the design service life and within the NPP power unit decommissioning period or the possibility for their replacement upon expiry of their lifetime shall be substantiated.
The following information shall be provided:
usage of robotics and manipulators for works with highly active equipment and radioactive wastes;
organization of transportation routes for highly active equipment and radioactive wastes;
sufficient capacity of the existing NPP power unit ventilation systems in order to perform dismantling works in full and (or) necessity to apply additional ventilation systems and air purification devices.
Information on consideration of dependence between the particle size of radioactive aerosols and the applied dismantling techniques shall be specified in sufficiency substantiation for the provided ventilation systems and air purification devices as sizes of aerosol particles determine selection of filters and other protective barriers.
18.5. Very low level wastes and materials for unrestricted use (reusable).
The list of materials for unrestricted use generated in the course of the NPP power unit decommissioning shall be provided, and their amount shall be assessed. Content of radionuclides in these materials shall not exceed numerical values of permissible specific activity of the basic long-lived radionuclides for unrestricted use of materials and man-made radionuclides allowing unrestricted use of solid materials as established by the regulations.
Amounts of very low level wastes that can be generated in the course of the NPP power unit decommissioning shall be assessed, and measures aimed to ensure radiation safety of the personnel, the public and the environment during handling of very low level wastes and at their disposal stages shall be provided.
It shall be specified how the NPP wastes are to be segregated and classified as very low level ones.
It shall be specified how the works for handling of very low level wastes are to be performed depending on the type of wastes and radionuclide forms in the wastes.
Information on the ways to arrange collection, storage, processing (if necessary) and
transportation (from the places of generation and collection to the interim storage room (area), outside the NPP site) of very low level wastes generated in the course of the NPP power unit decommissioning shall be presented.
Information on the methods for final disposal of very low level wastes and location of the VLLW repository shall be provided.
It shall be specified how the requirements for the outgoing radiological monitoring of the materials returned to the national economy without any restrictions of their further use are implemented in the design documentation for the NPP power unit.
Information on the methods to carry out decontamination of buildings, facilities, structures, systems and equipment as well as to arrange special-purpose rooms used for decontamination of radioactive equipment, its cutting, waste conditioning and handling of materials for unrestricted use shall be presented.
It is necessary to provide analysis results for the measures aimed to ensure radiation safety in the course of the NPP power unit site reclamation (methods for soil decontamination or immobilization of contaminants: water flushing, skimming, soil scraping and piling).
Appendix 4to federal rules and regulations
in the field of atomic energy use"Requirements for the content of
safety analysis reportsfor nuclear power plant units with VVER reactors" approved by Order
of the Federal Environmental,Industrial and Nuclear Supervision Service
dated February 13, 2017 No. 53
STANDARD FORMFOR DESCRIPTION OF SYSTEMS IN THE SAFETY ANALYSIS REPORT
Information on any safety-related system shall be presented in the following order.
In the course of information presentation references to other section or chapters of the NPP SAR where this information is provided in more details are possible.
Particular content of each subsection may change depending on the system peculiarities.
Individual subsections may be omitted or supplemented with other ones if it is determined by the system peculiarities.
1. Design basis.
The following shall be provided in this subsection:
purpose of the system and description of the performed functions;
the list of regulatory documents containing the requirements the system shall comply with;
principles and criteria taken as the basis for the system design;
modes of normal operation, abnormal operation including accidents (lists of initiating events for accidents, failures, external impacts, human errors and combinations of the above-mentioned events) when operation of the system is required; in case of special-purpose hardware for BDBA management the list of beyond design basis accidents requiring operation of this hardware shall be provided;
input data for the design defining the required characteristics and parameters of the system as well as external conditions when these characteristics are to be ensured; load limits for the system components under normal operation conditions and in case of any abnormal operation including accidents as well as under any external impacts (typical for the NPP location site) requiring operation of this system shall be specified;
requirements for the system layout;
requirements for associated systems;
requirements for the system reliability indicators (if any);
information on consideration of the requirements for the NPP power unit decommissioning.
Information on the following principles taken as the basis for the system design shall be presented, and compliance with these principles shall be demonstrated:
compliance with the single failure principle for safety functions performed by the system under consideration (for safety systems);
redundancy assurance (information on redundancy of individual system components (performing the same function irrespective of the requirements related to compliance with the single failure criterion) adopted in the design in order to enhance reliability of the system shall be specified);
diversity assurance (it shall be specified how the diversity principle is implemented in the design of systems and components in order to prevent common cause failures);
independence assurance (it shall be specified how functional and (or) physical separation of the system channels (components) is arranged so that a failure of any channel (components) would not result in a failure of any other channel (components));
principle of actuation for safety systems (if actuation is not performed automatically the relevant substantiation shall be provided); algorithm for resetting of safety systems;
in case safety-related systems combine safety functions with normal operation functions it shall be substantiated that it does not result in any violation of the requirements for the NPP safety assurance and reduction of required reliability for systems (components) performing safety functions.
2. System design.
2.1. Structure and process flow diagram.
This subsection shall contain description of the structure and (or) process flow diagram (electrical diagram) of the entire system and any channels, components, facilities, supports and foundations included into it. The list of components included into the system with indication of their design designation, basic technical characteristics of the system and components shall be presented.
Descriptions of individual components shall be provided (if necessary) in independent subsections with the same structure as the description of the entire system.
Drawings and schemes illustrating the structure and operation of the system and its components, its spatial layout and interfaces with other NPP systems in sufficient detail shall be provided. The adopted coding of the system and its components shall be indicated on the drawings and schemes.
2.2. Description of components.
Technical characteristics of the system components shall be specified. Safety class assigned in the NPP design, classification designation in accordance with federal rules and regulations in the field of atomic energy use "General provisions for nuclear power plant safety assurance" (NP-001-15) approved by Order No. 522 of the Federal Environmental, Industrial and Nuclear Supervision Service dated December, 17, 2015 (registered by Ministry of Justice of the Russian Federation on February, 2, 2016, registration number 40939; Official Internet Portal of Legal Information http://www.pravo.gov.ru, February, 5, 2016, No. 0001201602050003), seismic category and also group for equipment and pipelines subject to federal rules and regulations in the field of atomic energy use governing the requirements for arrangement and safe operation of NPP equipment and pipelines and any other classification designations shall be specified and substantiated for each component (in case a component is subject to classification in accordance with federal rules and regulations in the field of atomic energy use establishing the requirements for safety-related control systems, or establishing the requirements for arrangement and safe operation of hoisting cranes for nuclear facilities, or establishing the requirements for NPP pipeline valves).
Information on all limit stops, supports and shock absorbers installed on the pipelines and
equipment shall be provided.
It shall be specified how the design of these components ensures compliance with the design requirements for equipment and pipelines with regard to consideration of gravity loads, seismic impacts, thermal expansion in all modes provided in the design and also from any external and internal impacts taken into account in the design.
It shall be substantiated that all system components are designed with due regard for their operating conditions. Information on the ambient conditions in the rooms where these components are located in different NPP power unit operation modes shall be presented. It shall be substantiated that safety-related system components located inside the RF containment retain their operability in the course of all containment tests prescribed in the design.
Measures provided in the design for the system protection against external impacts as well as against internal impacts in case of any accidents shall be described.
It shall be substantiated that any failure of systems and components not referred to seismic category I does not cause failures of systems and components referred to seismic category I established in accordance with federal rules and regulations in the field of atomic energy use.
Measures provided in the NPP design in order to prevent harmful impact of microorganisms on the system components contacting with solutions in the course of normal operation shall be described.
It shall be substantiated that access to the system components is provided in the course of the NPP operation in order to carry out maintenance and repair works and requirements for radiation safety of the personnel are observed in this case.
Information on diagnostics of the system components, control methods and means for metal of pipelines and equipment, condition of assemblies, vibration, noise, loss of tightness, electrical resistivity shall be specified.
Information on the possibility for decontamination of systems and their components shall be provided.
Information on anti-corrosion protection and heat insulation of the system components shall be presented.
2.3. Materials.
The list of materials used to manufacture the system components shall be provided. In this case it shall be substantiated that the materials are selected with due regard for normal operation conditions and any abnormal operation including accidents that require operation of the system.
2.4. Safe operation limits and conditions, operation limits and conditions.
Operation limits and conditions, safe operation limits and conditions related to the system shall be specified. If any of the above-mentioned limits or conditions are not established in the NPP design for this system the relevant information shall be also specified in the NPP SAR.
Requirements for the system chemistry (water chemistry regime) shall be presented.
3. Monitoring and control of the system.
Values of the controlled parameters, protections and interlocks of the system prescribed in the NPP design for all normal operation modes and also for abnormal operation (including accidents) requiring operation of the system shall be specified and substantiated; location of the measurement
points shall be indicated, control methods shall be described, information on metrological validation of the applied methods and the requirements for instrumentation and control devices shall be presented, accuracy of parameter determination shall be specified. The system interfaces with control systems, redundancy of sensors and communication channels shall be described.
The material shall be presented in the following order:
description of the measurement points;
description of emergency and warning alarms;
description of protections and interlocks;
description of the system control and algorithms of its operation.
Information and characteristics of the places where monitoring and control of the system and its components is arranged shall be provided, sufficiency of the implemented measures to ensure survivability and habitability of the control rooms shall be substantiated. Information on the scope of monitoring and control of the system and its components carried out from the MCR and the ECR shall be presented. Information on availability of any position indication for shutoff and safety devices (valves) in the MCR and the ECR shall be provided.
In case the system is not controlled automatically or any personnel intervention into its operation is possible information on the means provided in the design for elimination or human errors and mitigation of their consequences shall be presented.
Information on the means aimed to support the operator in the system controlling shall be provided.
Permissible duration of power supply interruption for the systems and their components shall be specified and substantiated. Information on belonging of the systems (system components) to the relevant group of the EPSS consumers shall be provided.
4. Tests and inspections.
The basic requirements related to quality assurance for the system and its components during its manufacture, erection (construction) and installation shall be provided.
Information on the regulations for maintenance and periodic testing of the system and (or) its individual components shall be presented.
Information on factory tests of the system components shall be provided.
The following shall be specified and substantiated:
incoming control methods and scope for the system components;
scope of the commissioning tests (sufficiency for the NPP safety assurance, sequence and procedure, acceptance criteria);
methods, scope and frequency of tests and inspections within the operation period and their metrological support;
scope and type of instrumentation used in the course of testing.
Information on the parameters of hydraulic (pneumatic) testing of the systems and components subject to federal rules and regulations in the field of atomic energy use establishing the requirements
for arrangement and safe operation of equipment and pipelines for nuclear power installations shall be provided.
Works which may cause any abnormal operation at the NPP shall be listed, and measures aimed to prevent accidents shall be defined.
5. System design analysis.
5.1. Methods and means for substantiation of design functions performance by the system.
Description of software tools used in the design to analyze strength and operability of the system and its components, basic input data for the calculations, assumptions and limitations of the calculation schemes, calculation results and conclusions shall be provided. Information on verification and validation of software tools shall be presented.
If any experiments have been performed in order to substantiate operability of the system conditions of these experiments shall be described, correspondence of these conditions to actual operating conditions of the system shall be analyzed, experimental base and metrological support of the experiments shall be described, and the main results of the experiments shall be presented.
5.2. System functioning description.
Functioning of the system under normal operation conditions and in case of any abnormal operation including accidents (when operation of the system is required), interfaces with other systems, information on the operator's actions to control the system shall be described.
The following information shall be provided:
basic characteristics of the system for all operation modes prescribed in the design; it shall be demonstrated on the basis of justifying calculations and experimental substantiations that these characteristics do not exceed their limits established in the design and (or) regulatory documents;
calculation results confirming that the system and its components are capable to accommodate any loads on the equipment and building structures without any loss of operability under normal operation conditions and in case of any abnormal operation (including accidents) requiring operation of this system and also under any natural and human-induced external impacts considered in the NPP design.
5.3. Analysis of the system failures. Quantitative and qualitative reliability analysis.
Failures of the system components with due regard for human errors and impacts of failure consequences on operability of the system under consideration, any interfaced systems and the NPP safety in general shall be analyzed. Failures requiring special consideration in Chapter 15 of the NPP SAR shall be listed.
Common cause failures of the system (particularly due to fires) shall be analyzed, and impact of these failure consequences on operability of the system under consideration, any interfaced systems and the NPP safety in general shall be assessed.
Results of qualitative and quantitative analysis of the system reliability and also reliability analysis for the system components shall be provided in any cases prescribed by the federal rules and regulations in the field of atomic energy use.
5.4. Compliance of the system design with the applied requirements, principles and safety criteria shall be analyzed.
5.5. Information on the system specified in item 5.1-5.4 of this Appendix shall be presented in the following order:
functioning of the system during normal operation;
functioning of the system in case of any failures of its components;
functioning of the system in case of any abnormal operation including accidents;
functioning of the system under external impacts;
the system reliability indicators with regard to performance of the functions;
practical approval of design solutions.
Each subsection shall be completed with assessment of compliance with the requirements, principles and criteria specified in the regulatory documents as well as the design requirements. In case of any deviations from the above-mentioned requirements, principles and criteria in the system design the following shall be provided:
substantiated assessment of impact of the above-mentioned deviations on operability of this system and the NPP safety in general;
measures aimed to eliminate or compensate any deviations and time limits for implementation of these measures.
6. Conclusions.
Subsequent to the results of the system analysis conclusion shall be made on its compliance with the requirements of federal rules and regulations in the field of atomic energy use, any other regulatory documents containing mandatory requirements and also design principles and criteria.
7. The list of applied documentation.
Information on the NPP design taken as the basis for the system description as well as the list of any materials used in the design development and related to the system or its components (explanatory notes; reports with descriptions of the applied software tools, calculation schemes, calculation results, experimental base, results of experiments; documentation confirming characteristics of the system components) shall be provided.
Appendix 5to federal rules and regulations
in the field of atomic energy use"Requirements for the content of
safety analysis reportsfor nuclear power plant units with VVER reactors" approved by Order
of the Federal Environmental,Industrial and Nuclear Supervision Service
dated February 13, 2017 No. 53
Resultsof qualitative analysis for potential scenarios of consequences
in case of any natural or human-induced initiating eventsin the NPP location area and at the NPP site
1. Vibration of buildings and structures2. Missiles3. Vibration of systems and components
All NPP systems and components
Yes
1.1.2.
1.2. Human-induced impacts
1.2.1.
1.2.2
2. Impacts from the sources at the NPP site outside the NPP buildings
2.1. Natural impacts
2.1.1.
2.1.2.
2.2. Human-induced impacts
2.2.1.
2.2.2.
3. Impacts from the sources inside the NPP buildings
3.1.
3.2.
Note. In case safety-related systems are specified in Column 5 "Yes" shall be written in Column 6. In this case quantitative assessment results for probability of the initiating event shall be provided and substantiated in the relevant sections of the NPP SAR. If probability of this event exceeds the value at which an external impact shall be considered in the design basis (in accordance with federal rules and regulations in the field of atomic energy use) parameters of effects on the systems and components subjected to this impact shall be specified in the NPP SAR, and conclusion on stability of these systems and components under this impact shall be made.
Appendix 6to federal rules and regulations
in the field of atomic energy use"Requirements for the content of
safety analysis reportsfor nuclear power plant units with VVER reactors" approved by Order
of the Federal Environmental,Industrial and Nuclear Supervision Service
dated February 13, 2017 No. 53
SCOPE OFINFORMATION SPECIFIED IN DESCRIPTION OF INDIVIDUAL
RADIOLOGICAL MONITORING TYPES
1. Objects of control.
2. Control means and their metrological support.
The following information shall be provided:
types of stationary, portable, and laboratory equipment and instruments used for health physics and radiometric monitoring of surface contamination levels; content of volatile and gaseous radioactive substances in atmosphere of the rooms; sampling, individual monitoring of the personnel under normal operation conditions, in the course of repair and accidents;
information on the ways to provide the possibility for redundancy (with regard to the number and locations in case of accidents) of the measuring channels, means for presentation and documentation of information on the radiation situation in the rooms and at the NPP site with display of information in the emergency response control center outside the sanitary-protective area;
information on the mobile radiological monitoring laboratory (laboratories) equipped and completed for radiological monitoring and control.
3. Software and mathematical support.
Information on the possibility to predict propagation of radioactive substances and the radiation situation in the NPP rooms, at the NPP site and in the environment as well as to predict the radiation situation within the entire radiation accident zone depth in accordance with the list of beyond design basis accidents considered in the NPP design shall be provided.
It shall be demonstrated how geographical conditions, meteorology and development of the adjacent territories is taken into account in the calculations.
It shall be specified how prognostic mathematical models are implemented through the use of the radiological monitoring system software tools (configuration of the ST package shall be substantiated).
4. Computer equipment and methods for processing, analysis, presentation and transmission of information.
Capabilities of the computing machine or the network of personal computers used in the radiological monitoring system shall be described. It shall be demonstrated that they are sufficient for prediction of RS propagation and the radiation situation on the scale of the entire radiation accident zone within the minimum time required to solve this task.
5. Scope and frequency of control for radiological and meteorological parameters.
Information on the scope and frequency of control for radiological and meteorological parameters shall be presented.
Appendix 7to federal rules and regulations
in the field of atomic energy use"Requirements for the content of
safety analysis reportsfor nuclear power plant units with VVER reactors" approved by Order
of the Federal Environmental,Industrial and Nuclear Supervision Service
dated February 13, 2017 No. 53
APPROXIMATE LIST OF INITIATING EVENTS FOR ANALYSIS OF DESIGN BASIS ACCIDENTS <1>
--------------------------------
<1> This approximate list of initiating events has been developed in accordance with the provisions of item 1.2.15 of federal rules and regulations in the field of atomic energy use "General provisions for nuclear power plant safety assurance" (NP-001-15) approved by Order No. 522 of the Federal Environmental, Industrial and Nuclear Supervision Service dated December 17, 2015 (registered by the Ministry of Justice of the Russian Federation on February 2, 2016, registration number 40939; Official Internet Portal of Legal Information http://www.pravo.gov.ru, February 5, 2016, No. 0001201602050003). In accordance with the provisions of the above-mentioned item of federal rules and regulations in the field of atomic energy use the final list of initiating events for design basis accidents shall be presented in the NPP SAR.
1. Initiating events with changes of reactivity and energy emission distribution:
ejection of CPS CR;
uncontrolled withdrawal of the working CPS CR group from the reactor core with the operating rate;
misalignment of the CPS CR (falling of CPS CR, erroneous withdrawal of CPS CR, static misalignment of CPS CR);
connection of a non-operating loop of the reactor coolant pipeline without prior power reduction;
Inadvertent dilution of boric acid in the primary circuit coolant;
error in the course of fuel assembly loading into the reactor;
human error in the course of xenon oscillations suppression (movement of CPS CR resulting in the maximum possible deformation of the power density field).
2. Initiating events resulting in reduction of heat removal from the primary circuit:
reduction of steam flow rate to the turbine (due to improper operation or failure of the steam pressure controller);
closing of the turbine stop valves or loss of external electrical load;
loss of flow from feedwater pumps;
feedwater pipeline break;
unauthorized closure of the main steam isolation valve;
loss of vacuum in the turbine condenser or any other failures resulting in the turbine trip;
loss of auxiliary power supply of the NPP.
3. Initiating events resulting in increase of heat removal from the primary circuit:
deviations in operation of the feedwater system causing decrease of feedwater temperature;
deviations in operation of the feedwater system causing increase of feedwater flow rate;
unintended opening of the secondary circuit steam relief valve with subsequent failure to seat;
increase of steam consumption per a turbine (due to improper operation or failure of the steam pressure controller);
a range of steam pipeline ruptures inside and outside the containment.
4. Initiating events resulting in decrease of coolant flow rate in the primary circuit:
disconnection of various number of RCPs for different initial states of the RF;
instant seizure of an RCP;
RCP shaft breakage;
emergency frequency deviation in the NPP auxiliary power supply network.
5. Initiating events resulting in decrease of coolant mass in the primary circuit:
unauthorized opening of the pressurizer safety valve with subsequent failure to seat;
compensated primary circuit leakage inside the containment;
a range of primary circuit coolant leaks including gross-section rupture of the reactor coolant pipeline;
coolant leakages from the reactor during shutdown with unsealed reactor and in the course of refueling;
ruptures of pipelines connected to the primary circuit and crossing the containment resulting in leakages outside the containment;
breakage of a steam generator heat exchange tube;
breakaway of the steam generator header cover;
SG header breakage.
6. Initiating events caused by unauthorized actuation of safety systems:
unauthorized actuation of the algorithm for management of a primary-to-secondary leakage accident;
unauthorized injection to the pressurizer;
unauthorized actuation of the ECCS;
unauthorized actuation of the SG emergency cooldown system;
unauthorized actuation of the passive heat removal system.
7. Initiating events caused by any occurrences in the course of SNF storage and handling:
falling of separate fuel assemblies, leak-tight bottles, casings with fuel assemblies in the course of transportation and handling operations;
falling of any objects (into the reactor or spent fuel pool) that can change the position and damage the integrity of fuel assemblies (in the reactor and spent fuel pool);
breakage of packaging fasteners in the course of FA transportation;
falling of transportation containers with spent fuel assemblies;
failures of any components of the NF storage and handling systems;
hang-up of a spent fuel assembly in the course of refueling operations;
leakage from the fuel pool or rupture of a pipeline resulting in water level decrease in the fuel pool;
failures in the SFP cooling system;
reduction of homogeneous poison concentration in the SFP water;
possible errors in the course of the core loading (refueling).
8. Initiating events caused by releases of radioactive media from auxiliary systems and equipment:
medium leaks through the equipment seals;
pipeline leakages in the RW transportation, storage and processing systems containing radioactive gas;
leakage or damage of systems containing liquid radioactive media;
medium leakage from any vessel containing radioactive substances.
9. Fires
10. Flooding
11. External impacts of natural and human-induced nature
Appendix 8to federal rules and regulations
in the field of atomic energy use"Requirements for the content of
safety analysis reportsfor nuclear power plant units with VVER reactors" approved by Order
of the Federal Environmental,Industrial and Nuclear Supervision Service
dated February 13, 2017 No. 53
MINIMAL LISTOF PARAMETERS PRESENTED SUBSEQUENT TO ANALYSIS OF OPERATIONAL
OCCURRENCES INCLUDING ACCIDENTS
1. Time-based changes of the RF parameters:
reactor power;
pressure in the primary circuit (at the core outlet, in the pressurizer);
coolant temperature (at the core outlet, at the reactor inlet, saturation temperature);
coolant mass in the primary circuit, in the reactor;
burnout ratios for fuel elements with different capacity;
flow rate and enthalpy of the coolant discharged from the circuit (for accidents with primary or secondary circuit leakages);
flow rate and enthalpy of the medium supplied from the safety system to the primary circuit and steam generators;
heat flux removed by the safety system from the primary circuit and steam generators;
fuel element cladding temperature for fuel elements with different capacity;
fuel temperature for fuel elements with different capacity;
fuel enthalpy (for reactivity accidents);
pressure under the fuel element cladding (pressure differential on the fuel element cladding);
steam content in the nuclear core (volumetric and mass);
coolant level in the reactor (mass and physical);
pressure in the secondary circuit (in the SG and the main steam header);
feedwater flow rate in the SG;
coolant flow rate at the reactor outlet;
coolant level in the SG (mass and physical);
heat flux through the SG;
feedwater temperature (enthalpy) at the SG inlet;
oxidized zirconium mass (percent) (in case of any steam-zirconium reaction);
aggregate reactivity of the nuclear core and its components (for reactivity accidents);
flow rate (steam, water) via steam relief devices of the primary and secondary circuit;
medium temperature and level in the SFP (in case the accident results in any disturbances in the cooling of fuel stored in the SFP).
Maps of parameters distribution for representative moments of time shall be provided for design analyses in quasi-three-dimensional approximation.
1.1. Additionally for severe accidents:
time-based changes of the aggregate composition of fuel, poison and structural elements;
accumulation of corium components in the reactor vessel bottom part;
energy emission due to oxidation reaction.
2. Changes of parameters inside the containment:
pressure in the containment design volume;
temperature in the containment design volume;
radionuclide releases into the containment;
water volume (level) in the containment sump;
hydrogen generation rate;
steam and gas medium composition in the design volume;
flow rate and enthalpy of water from the spray system;
heat flux removed from the containment;
leakage flow rate from the containment.
2.1. Additionally for severe accidents:
mass and composition of the corium released from the reactor vessel;
energy emission in the corium released from the reactor vessel;
rate of flammable gas generation in the course of corium interaction with concrete;
heat flux supplied to the containment from the corium;
concrete ablation rate;
concentration of radionuclides and aerosols in the containment compartments;
flow rate of water supplied for the corium cooling (in case any water is supplied for corium cooling).
3. Radiological consequences
Information on the fission product migration processes inside the containment with indication of the following data shall be provided:
value of fission product leakage into the environment (at all accident development stages with due regard for all potential leakage paths);
characteristics of the process room surfaces along the fission product migration routes;
thermal and physical characteristics of atmosphere required to calculate fission product migration.
Results of dose rate calculations in the NPP rooms and at the NPP site and also calculations of exposure doses for the public shall be provided.
Time-dependent radionuclide composition and activity of releases into the environment at different stages of the accident shall be specified.
Appendix 9to federal rules and regulations
in the field of atomic energy use"Requirements for the content of
safety analysis reportsfor nuclear power plant units with VVER reactors" approved by Order
of the Federal Environmental,Industrial and Nuclear Supervision Service
dated February 13, 2017 No. 53
APPROXIMATE LIST OF BEYOND DESIGN BASIS ACCIDENTS <2>
--------------------------------
<2> This approximate list of beyond design basis accidents has been developed in accordance with the provisions of item 1.2.16 of federal rules and regulations in the field of atomic energy use "General provisions for nuclear power plant safety assurance" (NP-001-15) approved by Order No. 522 of the Federal Environmental, Industrial and Nuclear Supervision Service dated December 17, 2015 (registered by the Ministry of Justice of the Russian Federation on February 2, 2016, registration number 40939; Official Internet Portal of Legal Information http://www.pravo.gov.ru, February 5, 2016, No. 0001201602050003). In accordance with the provisions of the above-mentioned item of federal rules and regulations in the field of atomic energy use the final list of beyond design basis accidents shall be presented in the NPP SAR.
1. RF states with power operation of the reactor (including MCL) and shutdown states with hot primary circuit.
1.1. The range of expected NPP abnormal operation with coincident failure of emergency protection.
1.2. The range of primary circuit leakages (within the containment, outside the containment to adjacent systems, primary-to-secondary leakages) with failures of active ECCS components (with a failure of the normal operation NPP auxiliary power supply system and the emergency power supply system).
1.3. The range of primary-to-secondary leakages with failure of the damaged SG isolation (scenarios not considered within design basis accidents).
1.4. Long-term complete blackout of the NPP (loss of the normal operation NPP auxiliary power supply system and the emergency power supply system).
1.5. Large-scale primary circuit leakages with additional failures resulting in the containment bypass.
1.6. Rupture of the secondary circuit pipeline with failure of the damaged SG isolation (scenarios not considered within design basis accidents).
1.7. Failure of normal operation systems and active safety systems removing heat from the RF and SFP to the ultimate heat sink.
1.8. External impacts of natural and human-induced nature with the intensity exceeding the values considered in the NPP design and also combinations of these impacts.
1.9. Fires in the NPP rooms and at the NPP site (scenarios not considered within design basis accidents).
1.10. Flooding in the NPP rooms and at the NPP site (scenarios not considered within design basis accidents).
1.11. The range of severe accidents (including in-vessel and ex-vessel stages).
1.11.1. Long-term complete blackout of the NPP resulting in the SG dewatering and boil-off of the primary circuit coolant (with implementation of measures for reduction of the primary circuit pressure and without such measures).
1.11.2. The range of primary circuit leakages inside the containment with failures of active ECCS components resulting in severe damage of fuel elements in the nuclear core.
1.11.3. The range of primary-to-secondary leakages with failures of active ECCS components resulting in severe damage of fuel elements in the nuclear core.
2. States with the reactor shut down and cold primary circuit.
2.1. Long-term complete blackout of the NPP (loss of the normal operation NPP auxiliary power supply system and the emergency power supply system).
2.2. The range of primary circuit leakages with coincident additional failures (scenarios not considered within design basis accidents).
2.3. Accidents in the course of fuel storage and transportation.
2.4. Accidents in the course of RS and RW handling.
2.5. Range of severe accidents.
Appendix 10to federal rules and regulations
in the field of atomic energy use"Requirements for the content of
safety analysis reportsfor nuclear power plant units with VVER reactors" approved by Order
of the Federal Environmental,Industrial and Nuclear Supervision Service
dated February 13, 2017 No. 53
INPUT DATAFOR ANALYSIS OF ABNORMAL OPERATION
INCLUDING DESIGN BASIS AND BEYOND DESIGN BASIS ACCIDENTS
1 Initial conditions
The following shall be provided:
reactor power;
pressure above the nuclear core;
coolant temperature at the reactor inlet;
coolant flow through the reactor;
level in the pressurizer;
level in the SG (mass (physical));
feedwater flow rate per a SG;
feedwater temperature in the nominal mode;
feedwater temperature in case of the HPH disconnection.
2. Geometric and topological input data
The main structural characteristics (volume, length, passage areas, elevation differences, heat exchange surfaces, weights, wall thickness, hydraulic diameters, local resistance) shall be specified for:
the reactor (lower and upper chambers, downtake annular channel, the space under the reactor cover);
the nuclear core (fuel element, cassette, inter-cassette space);
the reactor coolant pipeline (hot and cold legs, hydraulic gates, inlet and outlet branch locations);
SG (vessel, headers, tubes, nozzles);
the pressurizer;
steam pipelines;
ECCS hydroaccumulators;.
Elevations of the design volumes, connections, pipeline rupture points, coolant and feedwater supply points shall be provided for the applied calculation nodalization scheme.
3. Physical input data
The following shall be provided:
neutron and physical characteristics (power distribution irregularity coefficients, reactivity coefficients, integrated efficiency of the CPS CR (without one most efficient CPS CR), prompt neutron lifetime, fractions of delayed neutrons);
residual energy emission curve (table);
accumulation of fission products under fuel element claddings and in the fuel as per the moment of accident;
properties of the structural materials and fuel (thermal capacity, thermal conductivity and density);
physical and chemical properties of reagents and solutions generated in the course of accident, their radiation stability, partition constants and chemical reactions with basic iodine compounds;
initial mass of fuel and structural materials in the reactor (fuel, zirconium, steel, poison);
the following shall be provided in case of three-dimensional calculations: the nuclear core loading patterns, maps of fuel burnout and energy emissions for each fuel assembly.
4. Process input data
The following shall be provided:
the main design characteristics (power, pressure head, capacity, operation algorithms, actuation/ disabling setpoints, etc.) of safety systems and safety-related normal operation systems provided in the NPP design and considered in the safety analyses;
the list of EP actuation setpoints;
the list of process protections and interlocks with regard to the primary and secondary circuit parameters and the parameters inside the containment;
operation algorithms for the main controllers (controller type, operation modes, input signals, adjustment ranges);
