SUB-SECTION L.4 DOSE UPTAKE PREDICTION

SUB-CHAPTER: L.4 SECTION : -

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FUNDAMENTAL SAFETY OVERVIEW VOLUME 2: DESIGN AND SAFETY CHAPTER L: RADIATION PROTECTION

SUB-SECTION L.4 DOSE UPTAKE PREDICTION

1. GENERAL

The objectives of the EPR dose optimisation initiative are:

- to set radiological protection demands at the same level as those for safety, achieving an optimisation approach to radiological protection similar to that applied to safety,

- to establish the EPR reactor as an improved reactor design in relation to the best units currently operating in France, updating the EPR dose targets in line with the continuous performance improvements of these units,

- to reduce dose uptake to the most exposed groups by focussing optimizing actions on the plant worker groups with the highest individual dose uptake,

- to improve the unit availability by allowing operators to enter the reactor building during power operation, while still complying strictly with radiological protection and conventional safety rules.

In order to meet these objectives:

- optimisation studies were carried out mainly based on recent operational feedback from the best operating units (individual dose uptake aspects, collective dose uptake, and good practices),

- the designer was positioned at the centre of the optimisation initiative,

- the project was given an ambitious collective dose target: 0.35 man-Sv per year per unit, averaged over ten years,

- the EPR activities optimized first and foremost were those which concerned the most exposed groups.

Exchange of information (documents, on-site meetings) with operating power plants and support from the organizations responsible for radiological protection have likewise allowed the designer to adopt significant best practice and specific lessons learned from the former.

German OPEX based on the design of the Konvoi units was also used for the design of specific EPR operations.

2. EPR DOSE UPTAKE PREDICTION

The ALARA approach adopted for EPR design studies is aimed at giving maximum benefits to the most exposed worker groups. The approach gives confidence that the ambitious collective dose target can be met. It sets out to establish an improved design in comparison with the best units of the French operating fleet (see figure L.4, FIG. 1).


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2.1. METHOD

The method proposed for the detailed EPR dose prediction analysis consists of:

- collecting dose uptake statistics from 1300 MWe NPPs starting with the NCAD1 codes from the best units, supplemented with the UTO (EDF Corporate Technical Support Department) data for planned maintenance and for elective maintenance, and with the data from the German facilities, for the EPR operations similar to the Konvoi design (e.g. aeroballs),

- selecting high priority activities in shut-down and in operation for radiation protection optimisation, giving priority to high dose activities, and involving the designer in the optimisation initiative;

- achieving the EPR dose uptake prediction from the concatenation of available NCAD codes, taking into account the type of outage,

- deriving the annual collective EPR dose prediction over a ten years cycle.

A flowchart for the optimisation process is shown in figure L.4 FIG, 1. In accordance with an ALARA approach, it allows for iteration on the choice of activities to be optimized and on the EPR features which have an impact on radiological protection.

The method used to estimate dose uptakes in the design stage of the EPR project recognises that:

- the EPR is an industrial facility (to be differentiated from medical or laboratory installations)

- is in its design phase and as a result no operational feedback is available for this type of reactor.

Estimating dose uptake increases or reduction remains a complex and challenging exercise because:

- the number of exposure hours is very significant,

- the number of workers is high during an outage,

- the dose rate levels can vary widely depending on the water level in the various systems,

- operation depends on the chosen design, particularly with regard to radiological protection,

- plant outage durations are shorter for the EPR, and some shut-down preparation activities are carried out during power operation.

1 NCAD: New Framework for Dosimetry Analysis – the NCAD codes correspond to entry codes for a

controlled area, where each elementary activity performed by a worker is defined by a unique NCAD code.




The objectives of the exercise are:

- to meet regulatory requirements,

- to provide qualitative arguments when the studies are not sufficiently advanced to estimate quantitative benefits,

- to draw the design team (materials, installation, operation) into considering the cross-disciplinary subjects of radiological protection, human factors, and conventional safety.

2.2. ESTABLISHING THE REFERENCE DOSE

The reference dose is determined from recent statistical values for the best performing French units.

2.2.1. Assumptions

The following assumptions are made to establish the reference dose:

- use of the best dose statistics from French units2: recent data and units (2001 to 2003, respectively; P’43 and N44 series),

- 18-month EPR fuel cycle,

- dose to be an average over ten years (type considered: NRO-ROO-NRO-ROO-NRO-ISIO).

Dose values are classified according to the type of outage: ROO (Refuelling Only Outage), NRO (Normal Refuelling Outage), or ISIO (In Service Inspection Outage); dose statistics for units during power operation have also been incorporated.

2.2.2. Results

Application of the above assumptions and analyses give the following results:

1) The calculated reference dose is 448 man.mSv per year per unit. This value takes into account the frequency of EPR shut-downs. It is noted that this value is close to that achieved at the best operating unit of the French fleet, GOLFECH 2, which has achieved 440 man.mSv per year over a full cycle of ten years. The latter value was thus chosen as the reference dose. It was assumed that the percentage distribution of the dose amongst each of the basic activities was that obtained from data for the sixteen best units.

2) The dose percentage per basic activity was determined by the type of outage (ROO, NRO or ISIO) and for a unit in operation.

3) The radiological protection high-priority activities (and the criterion characterizing the issue) are listed below:

2 Statistics obtained from the DPN central services 3 Golfech (1,2), Belleville (1,2), Nogent (1,2), Penly (1,2), and Cattenom (1,2,3,4) 4 Chooz (1,2) and Civaux (1,2)




- removal and installation of thermal insulation (operation involving the largest exposed population)

- opening and closing the reactor pressure vessel (high collective dose operation)

- preparation and inspection of primary side SG (high dose-rate worksite)

- site logistics (operation involving highly exposed worker groups)

- RCP [RCS], RCV [CVCS] and RIS-RRA [SIS-RHR] valves and component maintenance (operations involving highly exposed worker groups)

- waste treatment (operation involving radiological cleanliness)

- spent fuel posting out (high collective dose operation).

These activities only represent about 50% of the total annual dose uptake on operating units, (depending on unit shut-downs performed during a year), but they also involve the most exposed worker groups. The first results of the optimisation studies for these activities are described in Section 2.3.2 in this Sub-chapter.

2.3. ESTABLISHING THE OPTIMIZED DOSE

The optimized dose is obtained by improving the parameters (source term, dose rate, and amount of exposed work) which contribute to the reference dose.

In this section we describe:

- the effect of developments in the EPR design on the dose uptake assessment

- a summary of dose uptake results for the optimisation study.

2.3.1. Dose uptake assessment for the design evolution

The EPR design developments which have an impact on radiological protection (source term and amount of exposed work) are presented below.

2.3.1.1. Source term and dose rate optimisation

The EPR design source term has been altered by the following design changes in particular:

- optimizing the use of StelliteTM in the reactor vessel internals and in the valves,

- pressuriser developments:

- installation of a floor separating the spray and discharge systems at the pressuriser-dome level. This feature should reduce the average dose rate in the safety valves area since in the NPP fleet, this dose rate comes mainly from the spray nozzles. Total dose uptake associated with the maintenance of the pressuriser safety valves should be reduced by a minimum factor of 4, depending on the type of the work




- monitoring the nominal pressure will be done remotely via a special pressurised line;

- lower part of the pressuriser: the expansion line / pressuriser nozzle thermal sleeve has been removed in order to avoid dead zones. Inspection of the pressuriser heaters is facilitated by removing the thermal blanket. The assembly flange between the heaters and their sleeves is easy to replace. The heater pitch is square to allow for automatic inspection.

- installation optimisation:

- separate routing of the RCV [CVCS] pipework from the valves and pumps

- RRA [RHR] operation provided by the RIS [SIS] in the Safeguard and Electrical Buildings.

- inclusion in the Reactor Building of an area entirely dedicated to storing the pressure vessel head (located 4.45 m above the operating floor, with appropriate shielding).

- measures to remove "hot spots" in the design:

- elimination of pipe connections using socket welds, on all pipework carrying radioactive fluids

- chemistry optimisation (high flow rate purification)

- reduction of the amount of antimony in the primary pumps and reduction of the amount of chromium.

2.3.1.2. Optimizing the Amount of Exposed Work

The amount of exposed work (subject to radiation) expected for the EPR project has been altered in particular by the following design choices:

- execution of operations with the unit powered up,

- choice of important equipment with bolted connections (pressuriser heaters, control rod drive mechanisms)

- increased diameter of primary and secondary manways

- SG waterbox (BAE) layout which includes a cylindrical section (better access to tubes at the periphery)

- design of the ARE [MFWS] and ASG [EFWS] systems (ARE [MFWS] line connection point in the tapered shell and installation of a thermal sleeve)

- reactor vessel head heat insulation removable as a single unit

- absence of a forced ventilation device for the CRDMs (removal of the opening and closing operations for the CDRM ventilation system air duct)

- improved measuring instrumentation of the reactor vessel level (eliminating of removal and installation operations for the reactor vessel level piping)




- routing of the ex-core instrumentation into the reactor-vessel pit through the pool concrete wall (eliminating opening and closing operations for the Nuclear Instrumentation system covers at the bottom of the pool)

- similar or even reduced number of large valves (ND > 50)

- optimisation of the fuel handling operation duration,

- installation of shielding protections around radiating equipment (shielding, full cavity decontamination, … )

- modular maintenance valves.

2.3.1.3. Design developments being considered

Other EPR options, currently being studied, impact radiological protection:

- the maintenance program

- electropolishing of the SG waterboxes

- removal of the thermal sleeves

- in-service inspection programs linked to break preclusion, to the inspection of the CRDMs pressurized shrouds.

Since the selection of these options has not been finalized, corresponding dose uptake increases or reductions have not been considered in the EPR dose uptake assessment.

2.3.1.4. Specific features of man entries during power operation

- One significant characteristic of the EPR is its reactor building accessibility during power operation, which is planned to occur in particular seven days before the reactor trip (to prepare the outage) and three days after re-start.

- The reactor building accessible areas are the annular spaces and the service floor level, including the polar crane. For work to be carried out in these areas during power operation, the shielding and design are planned such that the total dose rate (gamma and neutron) is kept below 25 µSv/hr, and the neutron dose rate will not exceed 2.5 µSv/hr.

- Airborne contamination in the reactor building is due to uncollected leakage of primary coolant into the containment during power operation. A general design requirement of the ventilation systems aims at protecting personnel against airborne radiological hazards, by reducing to acceptable levels the concentrations of volatile nuclides in the accessible areas. To reach this goal, radiological protection studies (carried out with the designers responsible for ventilation and civil engineering) have demonstrated the need to select a "two-zone" concept. A description of how the two-zone concept works is given in Chapter L.3.3. The choice of a two-zone design allows total individual exposure due to inhalation to be guaranteed to be zero.




2.3.2. Results of high priority optimisation activities

In parallel with the dose assessment of the design developments, optimisation studies were carried out for the high dose activities by the designers responsible for installation, materials and operation.

For each activity, the optimisation modifications have been listed under three categories:

- known and proven design modifications;

- modifications being studied

- modifications still to be studied.

The results of studies considering the proven modifications determine the Initial Predicted Dose Estimate (EDPI).

The results of studies considering the proven modifications and those being studied calculate the updated Optimized Predicted Dose Estimate (EDPOa).

The results of studies, considering the proven modifications, those being studied, and those still to be studied, define a dose target (the Optimized Predicted Dose Estimate (EDPOc)).

The dose estimates will be updated as the detailed design phase progresses.

2.3.2.1. Thermal Insulation Operations

The "removal and installation of thermal insulation" activity represents between 5 and 7% of the total annual dose based on ROO and NRO type outages, and 13% for a ISIO type outage. It is a high dose task in terms of collective dose as well as individual dose. Indeed, the laggers represent the most exposed worker group on the plant. Owing to the type of activities associated with removal and installation of thermal insulation, optimizing their individual dose is based on reducing the source term and exposure time.

For this activity, radiological protection modifications currently being studied are:

- identification of the thermal insulation and its associated pipework

- sufficient permanent or supplementary lighting

- use of fast assembly-disassembly thermal insulation throughout the primary circuit (CPP) [RCPB] up to the second isolation valve and throughout the entire main secondary circuit (CSP) [MSS],

- use of fast, independent assembly-disassembly thermal insulation for the SGs

- use of fast, independent assembly-disassembly thermal insulation for the pressurizer

- use of fast assembly-disassembly thermal insulation on the welds and sensitive tap points

- the possibility of storing thermal insulation at fall-back areas.




Considering the dose benefits which these features introduce, the updated Optimized Predicted Dose Estimate (EDPOa) is estimated at 17.9 man.mSv per year per unit, which is a 35% improvement over the Initial Predicted Dose Estimate (EDPI).

2.3.2.2. Worksite logistics

The "worksite logistics" activity consists of a collection of wide-ranging elementary operations:

- worksite logistics operation which includes installation of change area tenting, security arrangements, and equipment preparation and monitoring (breathing air, materials, consumables), ALARA support and working conditions (remote monitoring, radio communications, general lighting…),

- "scaffolding erection and removal" operation,

- "shielding installation and removal" operation, which includes installing and removing supports and shielding devices,

- "nuclear logistics" operation for power operating conditions.

This activity represents between 13 and 16% of the total annual dose depending on the types of outages (ROO, NRO, and ISIO), and 11% of the dose for an operating unit.

It is a high dose operation in terms of collective and individual doses. Actually, logistic personnel represent one of the most exposed worker group, after the laggers and welders.

Owing to the type of logistic tasks, optimizing the individual dose of this worker group is based on reducing exposure time and source term.

Apart from the source term, the following measures have already been implemented:

- permanent platforms installed around the SGs at 5.15 m and 8.70 m

- decontaminable lead blankets.

Optimizing this activity relies on adapting for the EPR good practices currently used in operating units. The modifications currently considered have been defined from the analysis of these good practices.

The logistics activities with the greatest dose uptakes are:

- operations on the refuelling machine,

- Inspections of the reactor vessel head,

- maintenance activities around the SG water boxes,

- maintenance on the RCP [RCS], RCV [CVCS], and RIS-RRA [SIS-RHR] valves and components,

- Inspections of the RIS-RRA [SIS-RHR] exchangers

- maintenance activities on the GMPP [RCP] pumps.

The main modifications being studied to optimize logistic operations are:




- provision of shielding anchoring points for the activities with the highest dose uptake,

- use of PDMS (3D) and AUTOCAD (2D) computer tools to establish location maps for these anchoring points

- installation of fast, movable assembly-disassembly scaffolding around the highest dose uptake activities; particularly for inspections around the RIS [SIS] exchangers and pumps

- installation of permanent platforms around the SGs eyeholes, handholes, and primary manways openings as well as around the SGs

- to provide for fast-assembly/disassembly change area tentings,

- to include sufficient lighting (supplementary or permanent), electrical sockets, and air inlets,

- to consider during outage planning, activities associated with the installation and removal of shields (circuits full of water );

- to use containers / flasks designed for transporting contaminated material

- consideration of human factor and of conventional safety measures.

The EDPOa is estimated at 43.5 man.mSv per year per unit, the improvement percentage relative to the EDPI being 15% compared to the EDPI.

2.3.2.3. Valve and Components Operations

The "RCP [RCS], RCV [CVCS], and RIS-RRA [SIS-RHR] valves and components" activity represents an average of 8.5% of the total annual dose for units with an ROO type outage, 13% for a NRO type outage, and 12 % for a ISIO outage. It is a high dose operation in terms of collective and individual doses.

The RCP–RCV–RIS/RRA [RCS-CVCS-SIS/RHR] valves and components activity includes several operations performed during unit outages, mainly in the RCD (Core completely unloaded) state.

Each operation covers the set of maintenance tasks carried out on all gate valves, check valves, and globe valves located on the RCP [RCS], RCV [CVCS], and RIS-RRA [SIS/RHR] systems, respectively.

As a result, this RCP–RCV–RIS/RRA [RCS-CVCS-SIS/RHR] valves and components activity is complex, firstly due to the technological diversity of the installed valve equipment (gate valves, pressure relief valves, swing check valves…), as well as due to the range of maintenance activities carried out:

- tightening the press-packing or packing replacement

- internal valve inspection (visual examinations, dimensional checks…)

- valve-seat grinding

- complete replacement of the valve.




The diversity of these operations is conveyed by their duration, the number of workers needed, and the types of worker groups involved (fitters, welders, NDT monitoring…).

The workers absorbed doses around valve worksites are high because the workers are in direct contact with the materials on which they are working.

Furthermore, the environment and accessibility of those sites which are located in the reactor building often close to areas of the unit with more or less elevated dose rates, have a significant impact on the working conditions and consequently on the workers dose uptake.

The optimisation measures relating to valves and components activities are the following:

- use of Stellite will be limited as much as possible, particularly through the replacement:

- of gate valves with globe valves (which do not need cobalt-based hard facing) for ND between 50 and 150

- of stellite in gate valves with another hard facing cobalt-free material.

- elimination of pipe joints by means of socket welding, on all pipework carrying radioactive fluid: joints will be made on all piping with DN > 25 by butt welding, thus eliminating those clamping welds which are possible sources of hot spots;

- improvement of the watertightness of valves: valves carrying radioactive fluid will be provided with a double watertight barrier: press packing with a leak collection device. In the case of globe valves, this assembly will be supplemented with a watertight bellows seal. These measures contribute to radiological cleanliness (limiting leaks, which "dirty" the plants) and to a reduction of the amount of exposed work (fewer entries for unplanned repair of leaking valves);

- the installation of modular-maintenance globe valves: in radioactive systems, the use of modular-maintenance globe valves will be preferred. These valves have a detachable sub-assembly linking the valve internals, including the seat, which thus avoids maintenance activities such as replacing packing, in-situ grinding of the seat (with the risk of contaminating the plant), or even complete valve replacement. The replacement of the valve stem is performed in 30 min, compared to 2 hr minimum for a simple packing replacement;

- the development of elective maintenance with diagnosis tools: a certain number of electrically actuated on-off valves will be fitted with diagnosis tools monitoring the equipment operability and thus reducing the amount of maintenance (fewer regularly scheduled monitoring operations)

- consideration of the characteristics of the four EPR trains:

- the EPR is characterized by four trains for safety and support systems; these trains are independent and can be checked separately underwater, hence limiting of dose uptake,

- RRA [RHR] operation is provided by the RIS [SIS], which limits the large valves in these operations.




To date, only the benefits derived from the installation of modular-maintenance globe valves and components have been quantified (30%). The other modifications however are going qualitatively in the right direction.

Thus, the EDPOa of the RCP [RCS], RCV [CVCS], and RIS/RRA [SIS/RHR] valve activity has a value of 26.1 man.mSv per year.

2.3.2.4. Steam generator worksite

In the French fleet, the "SG preparation and inspection" activity represents an average of 18% of the total annual dose for a ROO type outage, 12% for a NRO type outage, and 7% for a ISIO outage. When considering the frequency of EPR-type outages, this is a job which represents an average of 13% of the total dose uptake for a unit's outage, two-thirds of which are for activities on the primary side and one third on the secondary side. It is therefore a high dose activity in terms of collective dose which must be optimized.

The total volume of maintenance activities applicable to the EPR SG remains comparable to current practice, without taking into account the less onerous maintenance frequency of units equipped with SGs with Inconel 690 tubes.

The SG preparation and inspection activity is divided into two jobs performed under APR (shut-down for refuelling) and RCD (core completely unloaded) outage conditions:

- primary side SG preparation and inspections, which consists principally of opening and closing the primary manways and associated maintenance,

- secondary side SG preparation and inspections, consisting primarily of opening and closing of the secondary openings (eyeholes, handholes and secondary manways) and the associated maintenance including sludge lancing.

Workers absorbed doses at a SG worksite are significant because the SG bunkers are tight spaces close to the primary circuit components. The primary circuit water level being very low during primary side operations, the primary fluid no longer fulfils its shielding function, and hot spots may appear in the system voids (SG, pipes and valves located in the SG bunker). In particular, among the highest dose activities, are those requiring man entry of the waterbox using "jumpers" (installation of taps, humidity detectors, etc.). For these activities, the exposure time is measured in seconds.

The collective reference dose for SG preparation and inspections is 40.1 man.mSv per year per unit.

Special measures have been taken for the EPR steam generators in order to reduce the source term, as well as the duration or frequency of high dose activities. They are considered to represent well proven modifications for the EPR. These measures are presented below.

- Optimisation of the source term: From the design stage, the ALARA approach is considered to optimize the source term (TS), particularly for 60Co, which, along with 58Co, contributes to approximately 80% of collective doses. Several engineering options have been proposed aiming to reduce the source term as low as reasonably possible and thus reducing the dose rate. The principal options considered aim at reducing:

- the residual cobalt content in the stainless steels making up the primary circuit,




- stellite-based hard facing materials.

In the case of the EPR SG, Inconel 690 alloy was finally preferred over Inconel 800 on stress-corrosion resistance, on overall steam-generator design, and on industrial feasibility grounds. Recommendations aiming at optimizing the behaviour of the 690 alloy in terms of dose uptake have been associated with this choice.

In addition, the temperature-bypass lines, which could make up as much as 13% of the area dose rate in some locations of the 1300-MW SG bunkers, have been removed.

- Optimizing the quantity of exposed work: The geographic location of pipes and equipment as well as the size of the worksites have been designed such that the amount of time exposed is as low as possible. Hence the objective aiming at reducing the duration and frequency of maintenance activities has been considered in the design of the EPR steam generators:

- To facilitate access to the interior of the SG by increasing the size of access openings:

- increasing the outside diameter of the primary manways (THPs) to 533 mm instead of 450 mm on the N4

- increasing the outside diameter of the secondary manways (THSs) to 600 mm instead of 530 mm on the N4.

These improvements are also proceeding in the same direction as human-factors studies.

- To reduce the frequency of tube bundle cleaning operations on the secondary side by reducing the production of crud:

- selection of materials limiting corrosion,

- optimisation of secondary water chemistry (and especially the All Volatile Treatment),

- re-use of the N4 design for blocking water run-off and the flow partition plate (N4 OPEX currently demonstrates that with sediment amounts lower than 12 kg, sludge lancing can be performed at every other shut-down, when it was initially anticipated for every shut-down).

- To facilitate access to the outermost tubes by modifying the geometry of the waterbox: adding a cylindrical section below the tube plate.

- To reduce the risk of mechanical damage to the ARE [MFWS] and ASG [EFWS] plants:

- positioning ARE [MFWS] tapping points in the tapered shell (inclined tube sheet) in order to limit the thermal layering phenomenon and thus to improve the fatigue behaviour of the points,




- installing a thermal sleeve welded onto the ARE [MFWS] and ASG [EFWS]: the absence of leakage between the sleeve and the tube sheet is beneficial vis-à-vis fatigue behaviour and catastrophic failure,




- limiting the usage of the ASG [EFWS] system, which is used exclusively for back-up and during periodic tests.

- To reduce the risk of damage to the tube bundle:

- installing traps for loose parts (PCM) in the main feedwater ring,

- use of high-permeability support plates (increasing the size of the cross-section for the passage of secondary water, reducing deposits).

These engineering measures are considered to represent proven modifications for the EPR. The in-service inspection program for the EPR SGs will be set up incorporating these improvements, with all the measures being included into the design to allow in-service SG monitoring and inspection.

By considering the latest available OPEX values and incorporating the proven modifications described above, the EDPI associated with this activity is estimated at 30.2 man.mSv per year and per unit, which is a 25% lower value than the reference dose.

Other modifications being considered should allow this activity dose to be optimized:

- requirement to use automated remote-control tools for the NDT operations,

- design of shielding measures adapted to eddy current SG tube monitoring ,

- installation of removable floors for the activities requiring eyehole/handhole opening (handling the gap left after the removal of the thermal insulation) and appropriate shielding protection,

- consideration of nozzle dams with inflatable joints or of lighter nozzle dams,

- improving reliability (electronic failures), simplifying (facilitating manoeuvrability), and improving performance (reducing execution times) for the opening and closing manways machine (MOFTH),

- providing BTR type screws on sealed joins in order to allow the use of pneumatic or electric equipment for their removal and installation,

- optimizing the design of thermal insulation (with a separate section at the eyehole/handhole level for partial disassembly),

- reducing the number of tube inspections by relying on operational feedback from SGs equipped with similar tubes. This should in particular remove the requirement to open waterboxes during ASR type outages. The dose benefits have not been estimated.

Taking these modifications into account, the EDPOa has been estimated at as 24.7 man.mSv per year and per unit, which is a reduction of 38.5% compared to the reference dose.

2.3.2.5. Worksite for Opening and Closing the Reactor Vessel

The opening and closing the reactor vessel activity, including inter-vessel work, represents an average of 8.8% of the total dose for a ROO type outage, 5.4% for a NRO type outage, and 3.8% for a ISIO outage. It is a high dose operation in terms of collective and individual doses.




The collective dose associated with the opening and closing the reactor vessel is mainly due to operations at the bottom of the pool at the o-ring level and to operations performed near the reactor vessel head, particularly when it is on its stand.

Vessel opening and closing is a worksite-type of activity, performed during plant outage and included in the technical work scope for integrated vessel maintenance. This job consists of preparing for maintenance, opening the reactor vessel for defuelling, and closing it for re-start following refuelling. It consists of three important phases, which are subdivided into nearly 90 elementary operations.

Aside from the source term, the following sources of development may be mentioned:

- Optimizing the transfer of upper (EISs) and lower internals (EIIs) underwater:

- the upper internals are systematically taken out of the vessel during unit refuelling outages (ROO, NRO, and ISIO). The EPR design allows the upper internals to be transferred from the vessel to their stand while still maintaining a water level of about 3 meters above the control rods. This design reduces the dose rate (DDD) at the pool-floor level compared with that in units at current French NPPs;

- The lower internals are removed from the vessel during a ten-year inspection (ISIO) type outage. As for the upper internals, the EPR design allows for their transfer underwater (approximately 300 mm of water above the rod-cluster assemblies). This advantage allows the dose rate to be reduced at the pool-floor level compared with that in units at current French NPPs.

- The addition of an area dedicated to storing the reactor vessel head: The pressure vessel head has a dedicated location for its storage on a stand after opening of the vessel. This storage area is located 4.45 m above the service floor and includes an appropriate concrete shielding. Thus it allows the increase in area dose rate from the reactor vessel head to be limited at the service-floor access routes. Furthermore, the sitting of the EPR stand will allow the dose rate due to stud cleaning and lubrication operations to be reduced during ROO type outages.

- Optimizing the removal and replacement of the reactor vessel head thermal insulation: the annular section of the thermal insulation removed in order to detach the vessel studs is removable as a single unit. This development constitutes an advantage over previous designs, in which the removal of thermal insulation was done piece by piece (four to six pieces, depending on the thermal insulation stage). In addition, the top part of the thermal insulation will be reinforced to support the weight of maintenance personnel, and the provision of access and safety measures (a ladder, guardrails). This design eliminates the on-site fitting problems encountered during removal and replacement of the thermal insulation, due to the deformation of sections, and it thus reduces the amount of exposed work at the pool base. Moreover, personnel safety is improved compared to the N4 units, where a guardrail is not provided for the thermal insulation (human factor consideration).

- Absence of CRDM ventilation air ducts (opening and closing the CDRM ventilation hatches): The mechanisms adopted for the EPR include coils capable of operating at higher temperatures than the coils for the N4 mechanisms, while still producing a lower Joule effect and thus not requiring forced-air ventilation on the head.




- Reactor vessel level measurement of the KONVOI type: The EPR design uses the in-core instrumentation of the KONVOI design, which is different from that at French power stations, particularly in the case of reactor-vessel level measurement (measured by ΔP on the N4; measured by level probe on the EPR). One of the impact on the EPR reactor-vessel opening and closing activities is the elimination of removal and reinstallation operations for the reactor vessel level pipe, necessary prior to opening the N4 vessel (reducing the amount of exposed work).

- Routing of the ex-core instrumentation through the pool's concrete wall: in contrast to the P’4 and N4 stages, the EPR ex-core instrumentation is routed into the reactor-vessel pit through the pool's concrete wall. As a result, closing the nuclear instrumentation covers at the pool bottom in the preparatory phase, prior to opening the vessel, is removed in the EPR (reducing the amount of exposed work).

- N4-type MST (Multistud Tensioning machine): The EPR MST will benefit from the N4 improvements, especially by reducing studs seizing risk during vessel opening and closing operations. The amount of exposed work at the bottom of the pool should therefore reduce during vessel opening and closing operations.

- Routing of the in-core instrumentation through the vessel head: one of the essential differences between the EPR design and that of the P’4 and N4 series is the routing of internal core instrumentation through the vessel head. This is to eliminate penetrations through the bottom of the EPR reactor vessel, for safety reasons. This has also led to the use of the aeroball system, as mobile instrumentation for measuring neutron flux. The resulting consequences associated with reactor vessel opening and closing operations are as follows:

- a greater number of disassembly and re-assembly operations for leaktightness at the instrumentation adapter level (17 instead of the four on the N4 units), as well as their cleaning and assessment during outage,

- the addition of disconnection operations for the aeroball tubes, first at their connection panel located in the upper part of the pool, then at the instrumentation adapter level, during the preparatory for the pressure vessel opening (reconnections during the closing phase),

- the addition of raising and lowering operations for the platform above the head which is dedicated to the aeroball ducts,

- The need to remove and install the aeroball nozzles respectively before the upper internals removal and after their reintroduction into the vessel, (the aeroball nozzles cannot remain connected to the upper internals during their transfer, an operation performed with the pool filled).

- Increased number of control rod drive shafts and of control rods drive mechanisms (CRDMs):

- the increased size of the EPR core (241 fuel assemblies instead of 205 for the N4 units) and its operating mode (increased cycle duration,…), leads to an increase in the number of CRDMs (89 for the EPR and 73 for the N4 units), hence the need to double the number of cable trays on the EPR (2 for the EPR, 1 for the N4 units),




- like the CRDMs, the number of control rods drive shafts increases from 73 to 89 on the EPR. This has the result of increasing the amount of exposed work for the control rods disconnecting and connecting operations before and after removing the upper internals.

- Penetrations in the BR pool walls: In contrast to the P’4 and N4 series, the walls of the EPR pool includes penetrations. These nine penetrations are for:

- the two connection panels (for CRDM cables and instrumentation) located opposite the head cable trays,

- the connection panel for the aeroball ducts located opposite the aeroball dedicated platform,

- the six ventilation ducts blowing fresh air in the direction of the head, four of which exit at the pool wall lower level, the other two being at the same level as the electrical connection panels.

These penetrations are blocked by manually closing watertight doors prior to opening the pressure vessel, then are reopened in the vessel closure phase. The impact of this design on outage activities is therefore the addition of in-pool opening and closing operations.

The EDPI obtained for the pressure vessel opening and closing activity is 21 man.mSv per year.

This increase of 8% compared to the reference dose (19.4 man.mSv) is due to the routing of neutronics instrumentation from the core through the pressure-vessel head on the EPR. Actually, all the operations linked to the instrumentation and required prior to opening the pressure vessel, impact the pressure vessel opening activities. Moreover, most of these additional operations are performed at the vessel flange level (a high dose rate area).

This system however offers the advantage of eliminating high dose operations associated with the instrumentation and performed beneath the pressure vessel.

The modifications being studied consider optimizing the worksite ergonomics beneath the pressure vessel head while the head is on its stand. Aside from the potential dose benefit, this is in particular a gain in the human factors area, which has been heavily involved in this practice; as a result, the EDPOa remains equal to the EDPI, which is 21 man.mSv per year.

2.3.2.6. Fuel Posting out Worksite

The "fuel posting out" activity at the Fuel Building transfer station represents approximately 25% of the total annual dose during power operation. It is a high dose operation in terms of collective and individual doses.

The dose rates for this activity are generated by:

- radiation from the fuel assemblies. This radiation has a significant neutron component.

- the radiation from fluid systems located above the trolley. Circulation of contaminated water leads to the formation of several hot spots above the DMK trolley. Radiation maps have shown that these hot spots contribute significantly to the total dose received.




In addition, the fuel posting out worksite is sensitive to radiological cleanliness. Indeed, the area and DMK trolley are vulnerable to contamination through water splashing, particularly during fluid systems disconnection operations on the cart, flask lid installation, flask skirt draining, and setting the penetration into water.

For the EPR, as for the P´4 and N4 series, the transport flask loading principles with depleted fuel assemblies are based on below-pit loading: The flask is transferred under the Fuel Building pool and brought into contact with a penetration located at the bottom of the pool (see sketch below).

Pool

Jack station Hopper station

Penetration

Transport flask

Transport vehicle

The different stages of this fuel posting out activity may be summarized as follows:

- Flask arrival: disassembly of the guards on the transport vehicle, tipping of the flask, positioning the flask above the DMK trolley, and transfer to the Fuel Building,

- Flask preparation: fitting of the valve tools, fluid connection, setting the flask in the water, adjusting the flask position, lid removal,

- Fuel loading: flask position adjusting, connection to the penetration, loading the flask, draining the penetration,

- Flask treatment: lid installation, water-tightness checks, draining and drying, flooding with nitrogen, radiological monitoring,

- Flask transfer out: transfer to the lifting station, tipping of the flask, loading onto the transport vehicle, radiological monitoring.

The collective absorbed dose during fuel posting out is 7 man.mSv on average in the P’4 units. This dose is variable, mainly depending on the residual activity of the twelve assemblies being removed (the neutron dose rate increases with residual activity) and on the trolley frequency of use (the trolley's contamination increases with the number of fuel transfers). For the P’4 units, the fuel posting out activity corresponds to an annual dose of 24.5 man.mSv at a rate of 3.5 fuel removals a year. This value represents the reference dose for the fuel posting out activity.




The known dose increases and reductions associated with the EPR design are listed below:

- Dose and contamination optimisation: - 40 %

- fluid flow improvement,

- design simplification,

- valve automation,

- sslow leakage fast connections,

- modification of relief tank.

- Operations optimisation for preparing and monitoring the trolley: -20 %

- installation of shielding,

- activity duration optimisation,

- simplification of radiological controls.

- Use of MOX: The use of MOX fuel leads to an increased neutron dose rate (+ 25%).

The Initial Predicted Dose Estimate (EDPI) for the fuel posting out activity is estimated at 15.8 man.mSv, which is a 35% improvement compared to the reference dose.

Modifications currently being considered allow a revised value for the Optimized Predicted Dose Estimate (EDPO) to be determined. A 10% improvement is achieved by optimising the cart position i.e.:

- blocking and centring ring at the penetration,

- lid lifting equipment horizontally adjustable.

The EDPO achieved for the fuel posting out activity is estimated at 14.2 man.mSv, which is a 42% improvement over the reference dose.

2.3.2.7. Waste Treatment Operations

Waste treatment operations represent an important activity with regard to the need to achieve radiological cleanliness.

Current development studies for the EPR are based on the following principles:

- waste items only leave the controlled area after being properly treated for transport and for its end user,

- the adoption of an intended or equivalent flux is preferred by design,

- waste treatment is performed in well-defined geographic areas,

- optimised ergonomics of transit and handling must be ensured.




The first results of these studies and the OPEX analysis for existing facilities have led to the following design choices for the EPR:

- in order to eliminate all surface transport of unlocked hulls, the Effluent Treatment Building has been located immediately adjacent to the Nuclear Auxiliaries Building and linked to it by underground transfer. This solution allows unlocked hulls to be transferred without leaving the controlled area,

- measures for treating low-activity waste are incorporated into Effluent Treatment Building; however a sorting area has been maintained in the Nuclear Auxiliaries Building for this type of waste,

- optimisation of shielding around the hull storage area.

3. SUMMARY OF RESULTS

The results in the following table are given in man.mSv.

Predicted type Reference dose Initial Predicted Dose Estimate

Optimized Predicted Dose

Estimate

Predicted Dose Target Estimate

ROO 323.3 ROO 290.4 ROO 249.0 ROO 235.8NRO: 517.7 NRO: 463.3 NRO: 400.4 NRO: 381.4

Dose by outage type

(average value per

year) 1327.7

(352.7)ISIO 1190.4

(316.1)ISIO 1038.3

(273.8) ISIO 1005.

(262.2)

Dose for unit in operation

per year 87.4 72.7 92.7 90.7

Total average per year 440 388 366 353

The Optimized Predicted Dose Estimate is a value representative of the proven modifications and radiological protection modifications being considered for the EPR; the value obtained is 0.366 man.Sv per year and per unit.

The calculated dose target of 0.353 man.Sv per year and per unit shows that the target of 0.35 man.Sv per year per unit is realistic, based on current studies in progress.

SUB-CHAPTER: L.4 SECTION : - FIGURE : 1 PAGE : 21 / 21

UK-EPR


FIG. 1: PRINCIPLE OF THE EPR OPTIMISATION INITIATIVE

Dose statistics

for P’4 and N4 series units

Reference DoseHigh dose activities Main Priorities

(best unit of French fleet)

Site visits with workers groups Initial EPR dose with known Research of adapted solutions with benefits from design development designers, mainly MA and IGC (EDPI) Analysis of solutions (cost- benefit)

EPR dose updated by new Results of optimization studies design developments (GC, PIS…)

Optimized Predicted Dose Estimation (EDPO)

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SUB-SECTION L.4 DOSE UPTAKE PREDICTION

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