Valerii KorobeinikovState Scientific Center
Institute of Physics and Power Engineering
Russia
1st Consultancy Meeting forReview of Innovative Reactor Concepts for Prevention of Severe
Accidents and Mitigation of their Consequences (RISC)
31 March - 2 April 2014
IAEA Headquarters Vienna Austria
Innovative Concepts Based on Fast Reactor Technology
Contents
bull Sodium Fast Reactorsbull Stages in the development of fast reactors in Russia
bull BN-1200 fast reactor
bull The stages of heavy metal coolant technology development
bull First rdquocommercialrdquo LFR SVBR-100
bull Proposals for estimation
bull RUSSIAN APPROACH TO FAST REACTOR SAFETY ANALYSIS
bull Major safety issues (after the Fukushima accident)
2
Stages in the development of fast
reactors in Russia
1Creation of fundamental basis of FRs (1950-
1970) Critical zero-power BR-1 experimental reactors
BR-510 BOR-60
2Engineering and technical familiarization of
Sodium Fast Reactors (1970-1990)First prototype of fast reactor BN-350 power fast reactor
BN-600 of Beloyarsk NPP (up to 2020)
3Discussions and conceptual investigations
(1990-2010)
4Current Russian Program (2010-2020)
commercialization SFR and development
new type FRReactor BN-800 with sodium coolant commercial reactor
BN-1200 with sodium coolant prototype of LVFR type
reactor with Pb-Bi coolant3
Fast reactor BR-510
BR-2 (1956-1957)
fuel ndash metallic Pu
power - 100 kW
coolant ndash Mercury (Hg)
BR-5 (1959-1964)
fuel ndash plutonium oxide
power - 5 MW
coolant ndash Sodium (Na)
BR-10 (1964-1998)
fuel ndash monocarbide and
mononitride U
power - 10 MW
coolant ndash Sodium (Na)
BR-10 experience oxide and
nitride fuel
neutron flux
up to 151015 cm-2s-14
Fast reactor BN-350
1000 MW(th) 350 MW(e)) was
the Worldrsquos first fast reactor-
prototype the loop-type reactor
was cooled through 6 separate
loops with sodium coolant
BN-350 confirmed technical and
engineering reliability of fast
reactors and gave first real
experience of operation
Problems with reliability of steam
and gas generators were
successfully solved5
BN-600 reactorRussian BN reactor
bull integral type of layout
bull three circulation loops (radioactive
Na-Na heat exchangers and
nonradioactive Na-H2O steam
generators)
bull oxide fuel (UO2 )
bull three zones of enrichment of core
bull availability of radial and axial
blankets parameters of a core
bull power density ~ 450 kWm3 and ~
48 kWm
bull burnup ~ 10-11 hm damage of
cladding -80-90 dpa
bull output temperature of Na ~550С
steel ~700С
bull operation time between fuel
reloading ndash frac12 year6
Key problems
1evaluation of feasibility of inherently safe fast reactors
2choice of coolant sodium heavy liquid metal gas or steam
3choice of fuel type МОХ carbide nitride or metal
4expediency of use of fertile blankets
5expediency and method (hetero- homo-geneous) of MA
transmutation
6fuel breeding level from BR~1 to BR~15
7fuel breeding level in the core core with equilibrium fuel and
BRcore ~1
8power density in the core from ~500 MWm3
to ~250 MWm3 and lower
9optimal fuel burn-up value from ~10 to ~15 and to ~20
10fuel cycle duration from 1-3 years to 5 years and more
11depth of fuel purification in reprocessing from 10-4 to 10-8 7
Breeding trend
BR-1 BR ~ 25
BR-2 BR ~16
BN-350 BR ~12
BN-800 BR ~ 10 divide 11
Factors
spectrum softening oxide fuel
(light element - O) sodium
coolant ( - Na)
priority of safety Na plenum
and exclusion of upper blanket
thin radial blanket
Is the high breeding actual goal
8
Conception of ldquofast reactor start
from U-235rdquo
9
Main conceptual ways of reactor
safety improvement
bull Minimization of excess reactivity for the fuel burn-up
bull Decrease of sodium void reactivity effect
bull Use of passive devices for reactivity control
bull Use of passive devices for decay heat removal
10
BN-800 fast reactor
BN-800 is first post-Soviet fast
reactor in Russia (start up at 2014)
This project is aimed at the
development of the fuel cycle
infrastructure and mastering of the
new types of fuel (MOX fuel)
Sodium plenum making it possible
to assure zero void reactivity effect
and passive safety systems are
special features of BN-800 reactor
design
Tests of these elements would lead to
the progress in the area of fast
reactor safety
1 - vessel 2 -guard vessel 3 - core 4 -
core diagrid 5 - core catcher 6 - silo 7 -
main sodium pump 8 - upper
stationary shielding 9 - large rotating
plug 10 ndashcentral rotating plug 11 -
protection cap 12 - refueling
mechanism 13 - small rotating plug 14
ndash intermediate heat exchanger11
BN-1200 fast reactor
12
Fuel cyclebullfuel ndash mixed oxide or nitridebulllow power density in the corebullexternal fuel cycle duration - 3 yearsbullBR ndash 12 (oxide) -13 (nitride BRcore ~1) bullMA utilization in the basic fuelSafetybull2 types of passive control rodsbullflattened core sodium plenumbullintegration of all primary sodium systems in the reactor vessel to eliminate radioactive sodium leaksEconomical characteristicsbulloptimization of layout approachesbullincrease of load factor by transition to one-year refuelling intervalbullincrease of the fuel burn-up
1 - intermediate heat exchanger 2 - reactor vessel3 - guard vessel 4 - silo 5 - core diagrid 6 - core catcher 7 - reactor core 8 ndash pump nozzle 9 ndash main sodium pump 10 ndash cold trap 11 ndashcontrol rod drives 12 ndash rotating plug
APL-705 experimental
(1971)
Test facility (1951)
APL-705 serial
(1976-1996)
SVBR-100 (plan - 2017)
The stages of heavy metal
coolant technology
development
13
First rdquocommercialrdquo LFR SVBR-100
Based on silicon oxide protected
ferritic-martensitic steel cladding the
Soviet submarine reactor design has been
converted to a commercial concept
SVBR-100 lead-bismuth cooled
reactor with 100 MWe power using
MOX or nitride fuel MOX gives
breeding ratio ~ 084 nitride ~ 10
Development financed by consortium
between Rosatom and private investors
Construction of prototype to start 2017
in Dimitrovgrad
httpwwwakmeengineeringcomsvbr10
0html14
SVBR Project is a pilot project for SC Rosatom in terms of implementation of
large-scale high-tech projects in the nuclear industry jointly with a commercial
partner
The project implementation form is also new and is connected not only with
outsourcing of commercial investments but also with establishing a project
management and business development joint venture
The project is a part of the Federal Target Program Nuclear Power
Technologies of the New Generation for 2010 - 2015 and until 2020 It is also
one of the components of New technological platform closed fuel cycle and
fast reactors project which is being realized under Commission for
Modernization and Technological Development of Russias Economy
The final product of the project is the basic technology of a lead-bismuth fast-
neutron reactor adapted to a civil project including construction of a
100MW(e) modular power generating plant and associated range of 100
MW(e) aliquot products
15
The intrinsic security and inherent safety properties
(at physical level) of RU SVBR-100 allow making the structural
design of power generating units much less elaborate and using
the modular design principle
The RU SVBR-100 structural design includes the requirements
of versatility in terms of the fuel used to enable fast conversion
to MOX and later to nitride fuel as well to serve the basis for
fuel self-sufficiency in the closed nuclear fuel cycle
The RU design and parameters allow setting up factory
manufacture of RU modules and their delivery to the site of
installation by railroad or vehicle which significantly reduces
the labor costs and lead time of nuclear power plant
construction Serial manufacture will presumably allow
achieving much lower production cost and stable quality of the
product16
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (1)
The main effect in providing a high safety level of the SVBR-100 RF
is achieved due to use of fast neutron reactor heavy liquid-metal
coolant and integral design of the reactor with total elimination of
pipelines with radioactive coolant beyond the monoblock vessel
The reactor possesses a negative void reactivity effect and
negative feedbacks the efficiency of the strongest absorbing rod does
not exceed 1 $ And that coupled with technical realization of the
control and protection system (CPS) eliminates prompt neutron
runaway of the reactor
The high boiling point of coolant heightens reliability of heat
removal from the core and safety due to lack of the heat transfer
crisis Also being coupled with a provided safeguard casing of the
monoblock that eliminates loss of coolant accidents (LOCA) and
high pressure radioactive exhausts
17
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (2)
The low pressure in the primary circuit reduces the risk
of its tightness failure and makes possible lessening the thickness
of reactor vesselrsquos walls and diminishing the limitations imposed
on the rate of temperature change according to thermal-cycling
strength conditions
The RF components do not contain materials releasing hydrogen
as a result of thermal and radiation effects and chemical reactions
with coolant water and air Therefore in an event of tightness
failure in the primary circuit the likelihood of chemical explosions
and fires is virtually eliminated
The circulation scheme of lead-bismuth coolant (LBC) provides
elimination of watersteam ingress into the core in an event of steam
generator (SG) leak due to effective gravitational separation of
steam on a free LBC level in the monoblock18
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
Contents
bull Sodium Fast Reactorsbull Stages in the development of fast reactors in Russia
bull BN-1200 fast reactor
bull The stages of heavy metal coolant technology development
bull First rdquocommercialrdquo LFR SVBR-100
bull Proposals for estimation
bull RUSSIAN APPROACH TO FAST REACTOR SAFETY ANALYSIS
bull Major safety issues (after the Fukushima accident)
2
Stages in the development of fast
reactors in Russia
1Creation of fundamental basis of FRs (1950-
1970) Critical zero-power BR-1 experimental reactors
BR-510 BOR-60
2Engineering and technical familiarization of
Sodium Fast Reactors (1970-1990)First prototype of fast reactor BN-350 power fast reactor
BN-600 of Beloyarsk NPP (up to 2020)
3Discussions and conceptual investigations
(1990-2010)
4Current Russian Program (2010-2020)
commercialization SFR and development
new type FRReactor BN-800 with sodium coolant commercial reactor
BN-1200 with sodium coolant prototype of LVFR type
reactor with Pb-Bi coolant3
Fast reactor BR-510
BR-2 (1956-1957)
fuel ndash metallic Pu
power - 100 kW
coolant ndash Mercury (Hg)
BR-5 (1959-1964)
fuel ndash plutonium oxide
power - 5 MW
coolant ndash Sodium (Na)
BR-10 (1964-1998)
fuel ndash monocarbide and
mononitride U
power - 10 MW
coolant ndash Sodium (Na)
BR-10 experience oxide and
nitride fuel
neutron flux
up to 151015 cm-2s-14
Fast reactor BN-350
1000 MW(th) 350 MW(e)) was
the Worldrsquos first fast reactor-
prototype the loop-type reactor
was cooled through 6 separate
loops with sodium coolant
BN-350 confirmed technical and
engineering reliability of fast
reactors and gave first real
experience of operation
Problems with reliability of steam
and gas generators were
successfully solved5
BN-600 reactorRussian BN reactor
bull integral type of layout
bull three circulation loops (radioactive
Na-Na heat exchangers and
nonradioactive Na-H2O steam
generators)
bull oxide fuel (UO2 )
bull three zones of enrichment of core
bull availability of radial and axial
blankets parameters of a core
bull power density ~ 450 kWm3 and ~
48 kWm
bull burnup ~ 10-11 hm damage of
cladding -80-90 dpa
bull output temperature of Na ~550С
steel ~700С
bull operation time between fuel
reloading ndash frac12 year6
Key problems
1evaluation of feasibility of inherently safe fast reactors
2choice of coolant sodium heavy liquid metal gas or steam
3choice of fuel type МОХ carbide nitride or metal
4expediency of use of fertile blankets
5expediency and method (hetero- homo-geneous) of MA
transmutation
6fuel breeding level from BR~1 to BR~15
7fuel breeding level in the core core with equilibrium fuel and
BRcore ~1
8power density in the core from ~500 MWm3
to ~250 MWm3 and lower
9optimal fuel burn-up value from ~10 to ~15 and to ~20
10fuel cycle duration from 1-3 years to 5 years and more
11depth of fuel purification in reprocessing from 10-4 to 10-8 7
Breeding trend
BR-1 BR ~ 25
BR-2 BR ~16
BN-350 BR ~12
BN-800 BR ~ 10 divide 11
Factors
spectrum softening oxide fuel
(light element - O) sodium
coolant ( - Na)
priority of safety Na plenum
and exclusion of upper blanket
thin radial blanket
Is the high breeding actual goal
8
Conception of ldquofast reactor start
from U-235rdquo
9
Main conceptual ways of reactor
safety improvement
bull Minimization of excess reactivity for the fuel burn-up
bull Decrease of sodium void reactivity effect
bull Use of passive devices for reactivity control
bull Use of passive devices for decay heat removal
10
BN-800 fast reactor
BN-800 is first post-Soviet fast
reactor in Russia (start up at 2014)
This project is aimed at the
development of the fuel cycle
infrastructure and mastering of the
new types of fuel (MOX fuel)
Sodium plenum making it possible
to assure zero void reactivity effect
and passive safety systems are
special features of BN-800 reactor
design
Tests of these elements would lead to
the progress in the area of fast
reactor safety
1 - vessel 2 -guard vessel 3 - core 4 -
core diagrid 5 - core catcher 6 - silo 7 -
main sodium pump 8 - upper
stationary shielding 9 - large rotating
plug 10 ndashcentral rotating plug 11 -
protection cap 12 - refueling
mechanism 13 - small rotating plug 14
ndash intermediate heat exchanger11
BN-1200 fast reactor
12
Fuel cyclebullfuel ndash mixed oxide or nitridebulllow power density in the corebullexternal fuel cycle duration - 3 yearsbullBR ndash 12 (oxide) -13 (nitride BRcore ~1) bullMA utilization in the basic fuelSafetybull2 types of passive control rodsbullflattened core sodium plenumbullintegration of all primary sodium systems in the reactor vessel to eliminate radioactive sodium leaksEconomical characteristicsbulloptimization of layout approachesbullincrease of load factor by transition to one-year refuelling intervalbullincrease of the fuel burn-up
1 - intermediate heat exchanger 2 - reactor vessel3 - guard vessel 4 - silo 5 - core diagrid 6 - core catcher 7 - reactor core 8 ndash pump nozzle 9 ndash main sodium pump 10 ndash cold trap 11 ndashcontrol rod drives 12 ndash rotating plug
APL-705 experimental
(1971)
Test facility (1951)
APL-705 serial
(1976-1996)
SVBR-100 (plan - 2017)
The stages of heavy metal
coolant technology
development
13
First rdquocommercialrdquo LFR SVBR-100
Based on silicon oxide protected
ferritic-martensitic steel cladding the
Soviet submarine reactor design has been
converted to a commercial concept
SVBR-100 lead-bismuth cooled
reactor with 100 MWe power using
MOX or nitride fuel MOX gives
breeding ratio ~ 084 nitride ~ 10
Development financed by consortium
between Rosatom and private investors
Construction of prototype to start 2017
in Dimitrovgrad
httpwwwakmeengineeringcomsvbr10
0html14
SVBR Project is a pilot project for SC Rosatom in terms of implementation of
large-scale high-tech projects in the nuclear industry jointly with a commercial
partner
The project implementation form is also new and is connected not only with
outsourcing of commercial investments but also with establishing a project
management and business development joint venture
The project is a part of the Federal Target Program Nuclear Power
Technologies of the New Generation for 2010 - 2015 and until 2020 It is also
one of the components of New technological platform closed fuel cycle and
fast reactors project which is being realized under Commission for
Modernization and Technological Development of Russias Economy
The final product of the project is the basic technology of a lead-bismuth fast-
neutron reactor adapted to a civil project including construction of a
100MW(e) modular power generating plant and associated range of 100
MW(e) aliquot products
15
The intrinsic security and inherent safety properties
(at physical level) of RU SVBR-100 allow making the structural
design of power generating units much less elaborate and using
the modular design principle
The RU SVBR-100 structural design includes the requirements
of versatility in terms of the fuel used to enable fast conversion
to MOX and later to nitride fuel as well to serve the basis for
fuel self-sufficiency in the closed nuclear fuel cycle
The RU design and parameters allow setting up factory
manufacture of RU modules and their delivery to the site of
installation by railroad or vehicle which significantly reduces
the labor costs and lead time of nuclear power plant
construction Serial manufacture will presumably allow
achieving much lower production cost and stable quality of the
product16
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (1)
The main effect in providing a high safety level of the SVBR-100 RF
is achieved due to use of fast neutron reactor heavy liquid-metal
coolant and integral design of the reactor with total elimination of
pipelines with radioactive coolant beyond the monoblock vessel
The reactor possesses a negative void reactivity effect and
negative feedbacks the efficiency of the strongest absorbing rod does
not exceed 1 $ And that coupled with technical realization of the
control and protection system (CPS) eliminates prompt neutron
runaway of the reactor
The high boiling point of coolant heightens reliability of heat
removal from the core and safety due to lack of the heat transfer
crisis Also being coupled with a provided safeguard casing of the
monoblock that eliminates loss of coolant accidents (LOCA) and
high pressure radioactive exhausts
17
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (2)
The low pressure in the primary circuit reduces the risk
of its tightness failure and makes possible lessening the thickness
of reactor vesselrsquos walls and diminishing the limitations imposed
on the rate of temperature change according to thermal-cycling
strength conditions
The RF components do not contain materials releasing hydrogen
as a result of thermal and radiation effects and chemical reactions
with coolant water and air Therefore in an event of tightness
failure in the primary circuit the likelihood of chemical explosions
and fires is virtually eliminated
The circulation scheme of lead-bismuth coolant (LBC) provides
elimination of watersteam ingress into the core in an event of steam
generator (SG) leak due to effective gravitational separation of
steam on a free LBC level in the monoblock18
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
Stages in the development of fast
reactors in Russia
1Creation of fundamental basis of FRs (1950-
1970) Critical zero-power BR-1 experimental reactors
BR-510 BOR-60
2Engineering and technical familiarization of
Sodium Fast Reactors (1970-1990)First prototype of fast reactor BN-350 power fast reactor
BN-600 of Beloyarsk NPP (up to 2020)
3Discussions and conceptual investigations
(1990-2010)
4Current Russian Program (2010-2020)
commercialization SFR and development
new type FRReactor BN-800 with sodium coolant commercial reactor
BN-1200 with sodium coolant prototype of LVFR type
reactor with Pb-Bi coolant3
Fast reactor BR-510
BR-2 (1956-1957)
fuel ndash metallic Pu
power - 100 kW
coolant ndash Mercury (Hg)
BR-5 (1959-1964)
fuel ndash plutonium oxide
power - 5 MW
coolant ndash Sodium (Na)
BR-10 (1964-1998)
fuel ndash monocarbide and
mononitride U
power - 10 MW
coolant ndash Sodium (Na)
BR-10 experience oxide and
nitride fuel
neutron flux
up to 151015 cm-2s-14
Fast reactor BN-350
1000 MW(th) 350 MW(e)) was
the Worldrsquos first fast reactor-
prototype the loop-type reactor
was cooled through 6 separate
loops with sodium coolant
BN-350 confirmed technical and
engineering reliability of fast
reactors and gave first real
experience of operation
Problems with reliability of steam
and gas generators were
successfully solved5
BN-600 reactorRussian BN reactor
bull integral type of layout
bull three circulation loops (radioactive
Na-Na heat exchangers and
nonradioactive Na-H2O steam
generators)
bull oxide fuel (UO2 )
bull three zones of enrichment of core
bull availability of radial and axial
blankets parameters of a core
bull power density ~ 450 kWm3 and ~
48 kWm
bull burnup ~ 10-11 hm damage of
cladding -80-90 dpa
bull output temperature of Na ~550С
steel ~700С
bull operation time between fuel
reloading ndash frac12 year6
Key problems
1evaluation of feasibility of inherently safe fast reactors
2choice of coolant sodium heavy liquid metal gas or steam
3choice of fuel type МОХ carbide nitride or metal
4expediency of use of fertile blankets
5expediency and method (hetero- homo-geneous) of MA
transmutation
6fuel breeding level from BR~1 to BR~15
7fuel breeding level in the core core with equilibrium fuel and
BRcore ~1
8power density in the core from ~500 MWm3
to ~250 MWm3 and lower
9optimal fuel burn-up value from ~10 to ~15 and to ~20
10fuel cycle duration from 1-3 years to 5 years and more
11depth of fuel purification in reprocessing from 10-4 to 10-8 7
Breeding trend
BR-1 BR ~ 25
BR-2 BR ~16
BN-350 BR ~12
BN-800 BR ~ 10 divide 11
Factors
spectrum softening oxide fuel
(light element - O) sodium
coolant ( - Na)
priority of safety Na plenum
and exclusion of upper blanket
thin radial blanket
Is the high breeding actual goal
8
Conception of ldquofast reactor start
from U-235rdquo
9
Main conceptual ways of reactor
safety improvement
bull Minimization of excess reactivity for the fuel burn-up
bull Decrease of sodium void reactivity effect
bull Use of passive devices for reactivity control
bull Use of passive devices for decay heat removal
10
BN-800 fast reactor
BN-800 is first post-Soviet fast
reactor in Russia (start up at 2014)
This project is aimed at the
development of the fuel cycle
infrastructure and mastering of the
new types of fuel (MOX fuel)
Sodium plenum making it possible
to assure zero void reactivity effect
and passive safety systems are
special features of BN-800 reactor
design
Tests of these elements would lead to
the progress in the area of fast
reactor safety
1 - vessel 2 -guard vessel 3 - core 4 -
core diagrid 5 - core catcher 6 - silo 7 -
main sodium pump 8 - upper
stationary shielding 9 - large rotating
plug 10 ndashcentral rotating plug 11 -
protection cap 12 - refueling
mechanism 13 - small rotating plug 14
ndash intermediate heat exchanger11
BN-1200 fast reactor
12
Fuel cyclebullfuel ndash mixed oxide or nitridebulllow power density in the corebullexternal fuel cycle duration - 3 yearsbullBR ndash 12 (oxide) -13 (nitride BRcore ~1) bullMA utilization in the basic fuelSafetybull2 types of passive control rodsbullflattened core sodium plenumbullintegration of all primary sodium systems in the reactor vessel to eliminate radioactive sodium leaksEconomical characteristicsbulloptimization of layout approachesbullincrease of load factor by transition to one-year refuelling intervalbullincrease of the fuel burn-up
1 - intermediate heat exchanger 2 - reactor vessel3 - guard vessel 4 - silo 5 - core diagrid 6 - core catcher 7 - reactor core 8 ndash pump nozzle 9 ndash main sodium pump 10 ndash cold trap 11 ndashcontrol rod drives 12 ndash rotating plug
APL-705 experimental
(1971)
Test facility (1951)
APL-705 serial
(1976-1996)
SVBR-100 (plan - 2017)
The stages of heavy metal
coolant technology
development
13
First rdquocommercialrdquo LFR SVBR-100
Based on silicon oxide protected
ferritic-martensitic steel cladding the
Soviet submarine reactor design has been
converted to a commercial concept
SVBR-100 lead-bismuth cooled
reactor with 100 MWe power using
MOX or nitride fuel MOX gives
breeding ratio ~ 084 nitride ~ 10
Development financed by consortium
between Rosatom and private investors
Construction of prototype to start 2017
in Dimitrovgrad
httpwwwakmeengineeringcomsvbr10
0html14
SVBR Project is a pilot project for SC Rosatom in terms of implementation of
large-scale high-tech projects in the nuclear industry jointly with a commercial
partner
The project implementation form is also new and is connected not only with
outsourcing of commercial investments but also with establishing a project
management and business development joint venture
The project is a part of the Federal Target Program Nuclear Power
Technologies of the New Generation for 2010 - 2015 and until 2020 It is also
one of the components of New technological platform closed fuel cycle and
fast reactors project which is being realized under Commission for
Modernization and Technological Development of Russias Economy
The final product of the project is the basic technology of a lead-bismuth fast-
neutron reactor adapted to a civil project including construction of a
100MW(e) modular power generating plant and associated range of 100
MW(e) aliquot products
15
The intrinsic security and inherent safety properties
(at physical level) of RU SVBR-100 allow making the structural
design of power generating units much less elaborate and using
the modular design principle
The RU SVBR-100 structural design includes the requirements
of versatility in terms of the fuel used to enable fast conversion
to MOX and later to nitride fuel as well to serve the basis for
fuel self-sufficiency in the closed nuclear fuel cycle
The RU design and parameters allow setting up factory
manufacture of RU modules and their delivery to the site of
installation by railroad or vehicle which significantly reduces
the labor costs and lead time of nuclear power plant
construction Serial manufacture will presumably allow
achieving much lower production cost and stable quality of the
product16
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (1)
The main effect in providing a high safety level of the SVBR-100 RF
is achieved due to use of fast neutron reactor heavy liquid-metal
coolant and integral design of the reactor with total elimination of
pipelines with radioactive coolant beyond the monoblock vessel
The reactor possesses a negative void reactivity effect and
negative feedbacks the efficiency of the strongest absorbing rod does
not exceed 1 $ And that coupled with technical realization of the
control and protection system (CPS) eliminates prompt neutron
runaway of the reactor
The high boiling point of coolant heightens reliability of heat
removal from the core and safety due to lack of the heat transfer
crisis Also being coupled with a provided safeguard casing of the
monoblock that eliminates loss of coolant accidents (LOCA) and
high pressure radioactive exhausts
17
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (2)
The low pressure in the primary circuit reduces the risk
of its tightness failure and makes possible lessening the thickness
of reactor vesselrsquos walls and diminishing the limitations imposed
on the rate of temperature change according to thermal-cycling
strength conditions
The RF components do not contain materials releasing hydrogen
as a result of thermal and radiation effects and chemical reactions
with coolant water and air Therefore in an event of tightness
failure in the primary circuit the likelihood of chemical explosions
and fires is virtually eliminated
The circulation scheme of lead-bismuth coolant (LBC) provides
elimination of watersteam ingress into the core in an event of steam
generator (SG) leak due to effective gravitational separation of
steam on a free LBC level in the monoblock18
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
Fast reactor BR-510
BR-2 (1956-1957)
fuel ndash metallic Pu
power - 100 kW
coolant ndash Mercury (Hg)
BR-5 (1959-1964)
fuel ndash plutonium oxide
power - 5 MW
coolant ndash Sodium (Na)
BR-10 (1964-1998)
fuel ndash monocarbide and
mononitride U
power - 10 MW
coolant ndash Sodium (Na)
BR-10 experience oxide and
nitride fuel
neutron flux
up to 151015 cm-2s-14
Fast reactor BN-350
1000 MW(th) 350 MW(e)) was
the Worldrsquos first fast reactor-
prototype the loop-type reactor
was cooled through 6 separate
loops with sodium coolant
BN-350 confirmed technical and
engineering reliability of fast
reactors and gave first real
experience of operation
Problems with reliability of steam
and gas generators were
successfully solved5
BN-600 reactorRussian BN reactor
bull integral type of layout
bull three circulation loops (radioactive
Na-Na heat exchangers and
nonradioactive Na-H2O steam
generators)
bull oxide fuel (UO2 )
bull three zones of enrichment of core
bull availability of radial and axial
blankets parameters of a core
bull power density ~ 450 kWm3 and ~
48 kWm
bull burnup ~ 10-11 hm damage of
cladding -80-90 dpa
bull output temperature of Na ~550С
steel ~700С
bull operation time between fuel
reloading ndash frac12 year6
Key problems
1evaluation of feasibility of inherently safe fast reactors
2choice of coolant sodium heavy liquid metal gas or steam
3choice of fuel type МОХ carbide nitride or metal
4expediency of use of fertile blankets
5expediency and method (hetero- homo-geneous) of MA
transmutation
6fuel breeding level from BR~1 to BR~15
7fuel breeding level in the core core with equilibrium fuel and
BRcore ~1
8power density in the core from ~500 MWm3
to ~250 MWm3 and lower
9optimal fuel burn-up value from ~10 to ~15 and to ~20
10fuel cycle duration from 1-3 years to 5 years and more
11depth of fuel purification in reprocessing from 10-4 to 10-8 7
Breeding trend
BR-1 BR ~ 25
BR-2 BR ~16
BN-350 BR ~12
BN-800 BR ~ 10 divide 11
Factors
spectrum softening oxide fuel
(light element - O) sodium
coolant ( - Na)
priority of safety Na plenum
and exclusion of upper blanket
thin radial blanket
Is the high breeding actual goal
8
Conception of ldquofast reactor start
from U-235rdquo
9
Main conceptual ways of reactor
safety improvement
bull Minimization of excess reactivity for the fuel burn-up
bull Decrease of sodium void reactivity effect
bull Use of passive devices for reactivity control
bull Use of passive devices for decay heat removal
10
BN-800 fast reactor
BN-800 is first post-Soviet fast
reactor in Russia (start up at 2014)
This project is aimed at the
development of the fuel cycle
infrastructure and mastering of the
new types of fuel (MOX fuel)
Sodium plenum making it possible
to assure zero void reactivity effect
and passive safety systems are
special features of BN-800 reactor
design
Tests of these elements would lead to
the progress in the area of fast
reactor safety
1 - vessel 2 -guard vessel 3 - core 4 -
core diagrid 5 - core catcher 6 - silo 7 -
main sodium pump 8 - upper
stationary shielding 9 - large rotating
plug 10 ndashcentral rotating plug 11 -
protection cap 12 - refueling
mechanism 13 - small rotating plug 14
ndash intermediate heat exchanger11
BN-1200 fast reactor
12
Fuel cyclebullfuel ndash mixed oxide or nitridebulllow power density in the corebullexternal fuel cycle duration - 3 yearsbullBR ndash 12 (oxide) -13 (nitride BRcore ~1) bullMA utilization in the basic fuelSafetybull2 types of passive control rodsbullflattened core sodium plenumbullintegration of all primary sodium systems in the reactor vessel to eliminate radioactive sodium leaksEconomical characteristicsbulloptimization of layout approachesbullincrease of load factor by transition to one-year refuelling intervalbullincrease of the fuel burn-up
1 - intermediate heat exchanger 2 - reactor vessel3 - guard vessel 4 - silo 5 - core diagrid 6 - core catcher 7 - reactor core 8 ndash pump nozzle 9 ndash main sodium pump 10 ndash cold trap 11 ndashcontrol rod drives 12 ndash rotating plug
APL-705 experimental
(1971)
Test facility (1951)
APL-705 serial
(1976-1996)
SVBR-100 (plan - 2017)
The stages of heavy metal
coolant technology
development
13
First rdquocommercialrdquo LFR SVBR-100
Based on silicon oxide protected
ferritic-martensitic steel cladding the
Soviet submarine reactor design has been
converted to a commercial concept
SVBR-100 lead-bismuth cooled
reactor with 100 MWe power using
MOX or nitride fuel MOX gives
breeding ratio ~ 084 nitride ~ 10
Development financed by consortium
between Rosatom and private investors
Construction of prototype to start 2017
in Dimitrovgrad
httpwwwakmeengineeringcomsvbr10
0html14
SVBR Project is a pilot project for SC Rosatom in terms of implementation of
large-scale high-tech projects in the nuclear industry jointly with a commercial
partner
The project implementation form is also new and is connected not only with
outsourcing of commercial investments but also with establishing a project
management and business development joint venture
The project is a part of the Federal Target Program Nuclear Power
Technologies of the New Generation for 2010 - 2015 and until 2020 It is also
one of the components of New technological platform closed fuel cycle and
fast reactors project which is being realized under Commission for
Modernization and Technological Development of Russias Economy
The final product of the project is the basic technology of a lead-bismuth fast-
neutron reactor adapted to a civil project including construction of a
100MW(e) modular power generating plant and associated range of 100
MW(e) aliquot products
15
The intrinsic security and inherent safety properties
(at physical level) of RU SVBR-100 allow making the structural
design of power generating units much less elaborate and using
the modular design principle
The RU SVBR-100 structural design includes the requirements
of versatility in terms of the fuel used to enable fast conversion
to MOX and later to nitride fuel as well to serve the basis for
fuel self-sufficiency in the closed nuclear fuel cycle
The RU design and parameters allow setting up factory
manufacture of RU modules and their delivery to the site of
installation by railroad or vehicle which significantly reduces
the labor costs and lead time of nuclear power plant
construction Serial manufacture will presumably allow
achieving much lower production cost and stable quality of the
product16
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (1)
The main effect in providing a high safety level of the SVBR-100 RF
is achieved due to use of fast neutron reactor heavy liquid-metal
coolant and integral design of the reactor with total elimination of
pipelines with radioactive coolant beyond the monoblock vessel
The reactor possesses a negative void reactivity effect and
negative feedbacks the efficiency of the strongest absorbing rod does
not exceed 1 $ And that coupled with technical realization of the
control and protection system (CPS) eliminates prompt neutron
runaway of the reactor
The high boiling point of coolant heightens reliability of heat
removal from the core and safety due to lack of the heat transfer
crisis Also being coupled with a provided safeguard casing of the
monoblock that eliminates loss of coolant accidents (LOCA) and
high pressure radioactive exhausts
17
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (2)
The low pressure in the primary circuit reduces the risk
of its tightness failure and makes possible lessening the thickness
of reactor vesselrsquos walls and diminishing the limitations imposed
on the rate of temperature change according to thermal-cycling
strength conditions
The RF components do not contain materials releasing hydrogen
as a result of thermal and radiation effects and chemical reactions
with coolant water and air Therefore in an event of tightness
failure in the primary circuit the likelihood of chemical explosions
and fires is virtually eliminated
The circulation scheme of lead-bismuth coolant (LBC) provides
elimination of watersteam ingress into the core in an event of steam
generator (SG) leak due to effective gravitational separation of
steam on a free LBC level in the monoblock18
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
Fast reactor BN-350
1000 MW(th) 350 MW(e)) was
the Worldrsquos first fast reactor-
prototype the loop-type reactor
was cooled through 6 separate
loops with sodium coolant
BN-350 confirmed technical and
engineering reliability of fast
reactors and gave first real
experience of operation
Problems with reliability of steam
and gas generators were
successfully solved5
BN-600 reactorRussian BN reactor
bull integral type of layout
bull three circulation loops (radioactive
Na-Na heat exchangers and
nonradioactive Na-H2O steam
generators)
bull oxide fuel (UO2 )
bull three zones of enrichment of core
bull availability of radial and axial
blankets parameters of a core
bull power density ~ 450 kWm3 and ~
48 kWm
bull burnup ~ 10-11 hm damage of
cladding -80-90 dpa
bull output temperature of Na ~550С
steel ~700С
bull operation time between fuel
reloading ndash frac12 year6
Key problems
1evaluation of feasibility of inherently safe fast reactors
2choice of coolant sodium heavy liquid metal gas or steam
3choice of fuel type МОХ carbide nitride or metal
4expediency of use of fertile blankets
5expediency and method (hetero- homo-geneous) of MA
transmutation
6fuel breeding level from BR~1 to BR~15
7fuel breeding level in the core core with equilibrium fuel and
BRcore ~1
8power density in the core from ~500 MWm3
to ~250 MWm3 and lower
9optimal fuel burn-up value from ~10 to ~15 and to ~20
10fuel cycle duration from 1-3 years to 5 years and more
11depth of fuel purification in reprocessing from 10-4 to 10-8 7
Breeding trend
BR-1 BR ~ 25
BR-2 BR ~16
BN-350 BR ~12
BN-800 BR ~ 10 divide 11
Factors
spectrum softening oxide fuel
(light element - O) sodium
coolant ( - Na)
priority of safety Na plenum
and exclusion of upper blanket
thin radial blanket
Is the high breeding actual goal
8
Conception of ldquofast reactor start
from U-235rdquo
9
Main conceptual ways of reactor
safety improvement
bull Minimization of excess reactivity for the fuel burn-up
bull Decrease of sodium void reactivity effect
bull Use of passive devices for reactivity control
bull Use of passive devices for decay heat removal
10
BN-800 fast reactor
BN-800 is first post-Soviet fast
reactor in Russia (start up at 2014)
This project is aimed at the
development of the fuel cycle
infrastructure and mastering of the
new types of fuel (MOX fuel)
Sodium plenum making it possible
to assure zero void reactivity effect
and passive safety systems are
special features of BN-800 reactor
design
Tests of these elements would lead to
the progress in the area of fast
reactor safety
1 - vessel 2 -guard vessel 3 - core 4 -
core diagrid 5 - core catcher 6 - silo 7 -
main sodium pump 8 - upper
stationary shielding 9 - large rotating
plug 10 ndashcentral rotating plug 11 -
protection cap 12 - refueling
mechanism 13 - small rotating plug 14
ndash intermediate heat exchanger11
BN-1200 fast reactor
12
Fuel cyclebullfuel ndash mixed oxide or nitridebulllow power density in the corebullexternal fuel cycle duration - 3 yearsbullBR ndash 12 (oxide) -13 (nitride BRcore ~1) bullMA utilization in the basic fuelSafetybull2 types of passive control rodsbullflattened core sodium plenumbullintegration of all primary sodium systems in the reactor vessel to eliminate radioactive sodium leaksEconomical characteristicsbulloptimization of layout approachesbullincrease of load factor by transition to one-year refuelling intervalbullincrease of the fuel burn-up
1 - intermediate heat exchanger 2 - reactor vessel3 - guard vessel 4 - silo 5 - core diagrid 6 - core catcher 7 - reactor core 8 ndash pump nozzle 9 ndash main sodium pump 10 ndash cold trap 11 ndashcontrol rod drives 12 ndash rotating plug
APL-705 experimental
(1971)
Test facility (1951)
APL-705 serial
(1976-1996)
SVBR-100 (plan - 2017)
The stages of heavy metal
coolant technology
development
13
First rdquocommercialrdquo LFR SVBR-100
Based on silicon oxide protected
ferritic-martensitic steel cladding the
Soviet submarine reactor design has been
converted to a commercial concept
SVBR-100 lead-bismuth cooled
reactor with 100 MWe power using
MOX or nitride fuel MOX gives
breeding ratio ~ 084 nitride ~ 10
Development financed by consortium
between Rosatom and private investors
Construction of prototype to start 2017
in Dimitrovgrad
httpwwwakmeengineeringcomsvbr10
0html14
SVBR Project is a pilot project for SC Rosatom in terms of implementation of
large-scale high-tech projects in the nuclear industry jointly with a commercial
partner
The project implementation form is also new and is connected not only with
outsourcing of commercial investments but also with establishing a project
management and business development joint venture
The project is a part of the Federal Target Program Nuclear Power
Technologies of the New Generation for 2010 - 2015 and until 2020 It is also
one of the components of New technological platform closed fuel cycle and
fast reactors project which is being realized under Commission for
Modernization and Technological Development of Russias Economy
The final product of the project is the basic technology of a lead-bismuth fast-
neutron reactor adapted to a civil project including construction of a
100MW(e) modular power generating plant and associated range of 100
MW(e) aliquot products
15
The intrinsic security and inherent safety properties
(at physical level) of RU SVBR-100 allow making the structural
design of power generating units much less elaborate and using
the modular design principle
The RU SVBR-100 structural design includes the requirements
of versatility in terms of the fuel used to enable fast conversion
to MOX and later to nitride fuel as well to serve the basis for
fuel self-sufficiency in the closed nuclear fuel cycle
The RU design and parameters allow setting up factory
manufacture of RU modules and their delivery to the site of
installation by railroad or vehicle which significantly reduces
the labor costs and lead time of nuclear power plant
construction Serial manufacture will presumably allow
achieving much lower production cost and stable quality of the
product16
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (1)
The main effect in providing a high safety level of the SVBR-100 RF
is achieved due to use of fast neutron reactor heavy liquid-metal
coolant and integral design of the reactor with total elimination of
pipelines with radioactive coolant beyond the monoblock vessel
The reactor possesses a negative void reactivity effect and
negative feedbacks the efficiency of the strongest absorbing rod does
not exceed 1 $ And that coupled with technical realization of the
control and protection system (CPS) eliminates prompt neutron
runaway of the reactor
The high boiling point of coolant heightens reliability of heat
removal from the core and safety due to lack of the heat transfer
crisis Also being coupled with a provided safeguard casing of the
monoblock that eliminates loss of coolant accidents (LOCA) and
high pressure radioactive exhausts
17
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (2)
The low pressure in the primary circuit reduces the risk
of its tightness failure and makes possible lessening the thickness
of reactor vesselrsquos walls and diminishing the limitations imposed
on the rate of temperature change according to thermal-cycling
strength conditions
The RF components do not contain materials releasing hydrogen
as a result of thermal and radiation effects and chemical reactions
with coolant water and air Therefore in an event of tightness
failure in the primary circuit the likelihood of chemical explosions
and fires is virtually eliminated
The circulation scheme of lead-bismuth coolant (LBC) provides
elimination of watersteam ingress into the core in an event of steam
generator (SG) leak due to effective gravitational separation of
steam on a free LBC level in the monoblock18
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
BN-600 reactorRussian BN reactor
bull integral type of layout
bull three circulation loops (radioactive
Na-Na heat exchangers and
nonradioactive Na-H2O steam
generators)
bull oxide fuel (UO2 )
bull three zones of enrichment of core
bull availability of radial and axial
blankets parameters of a core
bull power density ~ 450 kWm3 and ~
48 kWm
bull burnup ~ 10-11 hm damage of
cladding -80-90 dpa
bull output temperature of Na ~550С
steel ~700С
bull operation time between fuel
reloading ndash frac12 year6
Key problems
1evaluation of feasibility of inherently safe fast reactors
2choice of coolant sodium heavy liquid metal gas or steam
3choice of fuel type МОХ carbide nitride or metal
4expediency of use of fertile blankets
5expediency and method (hetero- homo-geneous) of MA
transmutation
6fuel breeding level from BR~1 to BR~15
7fuel breeding level in the core core with equilibrium fuel and
BRcore ~1
8power density in the core from ~500 MWm3
to ~250 MWm3 and lower
9optimal fuel burn-up value from ~10 to ~15 and to ~20
10fuel cycle duration from 1-3 years to 5 years and more
11depth of fuel purification in reprocessing from 10-4 to 10-8 7
Breeding trend
BR-1 BR ~ 25
BR-2 BR ~16
BN-350 BR ~12
BN-800 BR ~ 10 divide 11
Factors
spectrum softening oxide fuel
(light element - O) sodium
coolant ( - Na)
priority of safety Na plenum
and exclusion of upper blanket
thin radial blanket
Is the high breeding actual goal
8
Conception of ldquofast reactor start
from U-235rdquo
9
Main conceptual ways of reactor
safety improvement
bull Minimization of excess reactivity for the fuel burn-up
bull Decrease of sodium void reactivity effect
bull Use of passive devices for reactivity control
bull Use of passive devices for decay heat removal
10
BN-800 fast reactor
BN-800 is first post-Soviet fast
reactor in Russia (start up at 2014)
This project is aimed at the
development of the fuel cycle
infrastructure and mastering of the
new types of fuel (MOX fuel)
Sodium plenum making it possible
to assure zero void reactivity effect
and passive safety systems are
special features of BN-800 reactor
design
Tests of these elements would lead to
the progress in the area of fast
reactor safety
1 - vessel 2 -guard vessel 3 - core 4 -
core diagrid 5 - core catcher 6 - silo 7 -
main sodium pump 8 - upper
stationary shielding 9 - large rotating
plug 10 ndashcentral rotating plug 11 -
protection cap 12 - refueling
mechanism 13 - small rotating plug 14
ndash intermediate heat exchanger11
BN-1200 fast reactor
12
Fuel cyclebullfuel ndash mixed oxide or nitridebulllow power density in the corebullexternal fuel cycle duration - 3 yearsbullBR ndash 12 (oxide) -13 (nitride BRcore ~1) bullMA utilization in the basic fuelSafetybull2 types of passive control rodsbullflattened core sodium plenumbullintegration of all primary sodium systems in the reactor vessel to eliminate radioactive sodium leaksEconomical characteristicsbulloptimization of layout approachesbullincrease of load factor by transition to one-year refuelling intervalbullincrease of the fuel burn-up
1 - intermediate heat exchanger 2 - reactor vessel3 - guard vessel 4 - silo 5 - core diagrid 6 - core catcher 7 - reactor core 8 ndash pump nozzle 9 ndash main sodium pump 10 ndash cold trap 11 ndashcontrol rod drives 12 ndash rotating plug
APL-705 experimental
(1971)
Test facility (1951)
APL-705 serial
(1976-1996)
SVBR-100 (plan - 2017)
The stages of heavy metal
coolant technology
development
13
First rdquocommercialrdquo LFR SVBR-100
Based on silicon oxide protected
ferritic-martensitic steel cladding the
Soviet submarine reactor design has been
converted to a commercial concept
SVBR-100 lead-bismuth cooled
reactor with 100 MWe power using
MOX or nitride fuel MOX gives
breeding ratio ~ 084 nitride ~ 10
Development financed by consortium
between Rosatom and private investors
Construction of prototype to start 2017
in Dimitrovgrad
httpwwwakmeengineeringcomsvbr10
0html14
SVBR Project is a pilot project for SC Rosatom in terms of implementation of
large-scale high-tech projects in the nuclear industry jointly with a commercial
partner
The project implementation form is also new and is connected not only with
outsourcing of commercial investments but also with establishing a project
management and business development joint venture
The project is a part of the Federal Target Program Nuclear Power
Technologies of the New Generation for 2010 - 2015 and until 2020 It is also
one of the components of New technological platform closed fuel cycle and
fast reactors project which is being realized under Commission for
Modernization and Technological Development of Russias Economy
The final product of the project is the basic technology of a lead-bismuth fast-
neutron reactor adapted to a civil project including construction of a
100MW(e) modular power generating plant and associated range of 100
MW(e) aliquot products
15
The intrinsic security and inherent safety properties
(at physical level) of RU SVBR-100 allow making the structural
design of power generating units much less elaborate and using
the modular design principle
The RU SVBR-100 structural design includes the requirements
of versatility in terms of the fuel used to enable fast conversion
to MOX and later to nitride fuel as well to serve the basis for
fuel self-sufficiency in the closed nuclear fuel cycle
The RU design and parameters allow setting up factory
manufacture of RU modules and their delivery to the site of
installation by railroad or vehicle which significantly reduces
the labor costs and lead time of nuclear power plant
construction Serial manufacture will presumably allow
achieving much lower production cost and stable quality of the
product16
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (1)
The main effect in providing a high safety level of the SVBR-100 RF
is achieved due to use of fast neutron reactor heavy liquid-metal
coolant and integral design of the reactor with total elimination of
pipelines with radioactive coolant beyond the monoblock vessel
The reactor possesses a negative void reactivity effect and
negative feedbacks the efficiency of the strongest absorbing rod does
not exceed 1 $ And that coupled with technical realization of the
control and protection system (CPS) eliminates prompt neutron
runaway of the reactor
The high boiling point of coolant heightens reliability of heat
removal from the core and safety due to lack of the heat transfer
crisis Also being coupled with a provided safeguard casing of the
monoblock that eliminates loss of coolant accidents (LOCA) and
high pressure radioactive exhausts
17
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (2)
The low pressure in the primary circuit reduces the risk
of its tightness failure and makes possible lessening the thickness
of reactor vesselrsquos walls and diminishing the limitations imposed
on the rate of temperature change according to thermal-cycling
strength conditions
The RF components do not contain materials releasing hydrogen
as a result of thermal and radiation effects and chemical reactions
with coolant water and air Therefore in an event of tightness
failure in the primary circuit the likelihood of chemical explosions
and fires is virtually eliminated
The circulation scheme of lead-bismuth coolant (LBC) provides
elimination of watersteam ingress into the core in an event of steam
generator (SG) leak due to effective gravitational separation of
steam on a free LBC level in the monoblock18
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
Key problems
1evaluation of feasibility of inherently safe fast reactors
2choice of coolant sodium heavy liquid metal gas or steam
3choice of fuel type МОХ carbide nitride or metal
4expediency of use of fertile blankets
5expediency and method (hetero- homo-geneous) of MA
transmutation
6fuel breeding level from BR~1 to BR~15
7fuel breeding level in the core core with equilibrium fuel and
BRcore ~1
8power density in the core from ~500 MWm3
to ~250 MWm3 and lower
9optimal fuel burn-up value from ~10 to ~15 and to ~20
10fuel cycle duration from 1-3 years to 5 years and more
11depth of fuel purification in reprocessing from 10-4 to 10-8 7
Breeding trend
BR-1 BR ~ 25
BR-2 BR ~16
BN-350 BR ~12
BN-800 BR ~ 10 divide 11
Factors
spectrum softening oxide fuel
(light element - O) sodium
coolant ( - Na)
priority of safety Na plenum
and exclusion of upper blanket
thin radial blanket
Is the high breeding actual goal
8
Conception of ldquofast reactor start
from U-235rdquo
9
Main conceptual ways of reactor
safety improvement
bull Minimization of excess reactivity for the fuel burn-up
bull Decrease of sodium void reactivity effect
bull Use of passive devices for reactivity control
bull Use of passive devices for decay heat removal
10
BN-800 fast reactor
BN-800 is first post-Soviet fast
reactor in Russia (start up at 2014)
This project is aimed at the
development of the fuel cycle
infrastructure and mastering of the
new types of fuel (MOX fuel)
Sodium plenum making it possible
to assure zero void reactivity effect
and passive safety systems are
special features of BN-800 reactor
design
Tests of these elements would lead to
the progress in the area of fast
reactor safety
1 - vessel 2 -guard vessel 3 - core 4 -
core diagrid 5 - core catcher 6 - silo 7 -
main sodium pump 8 - upper
stationary shielding 9 - large rotating
plug 10 ndashcentral rotating plug 11 -
protection cap 12 - refueling
mechanism 13 - small rotating plug 14
ndash intermediate heat exchanger11
BN-1200 fast reactor
12
Fuel cyclebullfuel ndash mixed oxide or nitridebulllow power density in the corebullexternal fuel cycle duration - 3 yearsbullBR ndash 12 (oxide) -13 (nitride BRcore ~1) bullMA utilization in the basic fuelSafetybull2 types of passive control rodsbullflattened core sodium plenumbullintegration of all primary sodium systems in the reactor vessel to eliminate radioactive sodium leaksEconomical characteristicsbulloptimization of layout approachesbullincrease of load factor by transition to one-year refuelling intervalbullincrease of the fuel burn-up
1 - intermediate heat exchanger 2 - reactor vessel3 - guard vessel 4 - silo 5 - core diagrid 6 - core catcher 7 - reactor core 8 ndash pump nozzle 9 ndash main sodium pump 10 ndash cold trap 11 ndashcontrol rod drives 12 ndash rotating plug
APL-705 experimental
(1971)
Test facility (1951)
APL-705 serial
(1976-1996)
SVBR-100 (plan - 2017)
The stages of heavy metal
coolant technology
development
13
First rdquocommercialrdquo LFR SVBR-100
Based on silicon oxide protected
ferritic-martensitic steel cladding the
Soviet submarine reactor design has been
converted to a commercial concept
SVBR-100 lead-bismuth cooled
reactor with 100 MWe power using
MOX or nitride fuel MOX gives
breeding ratio ~ 084 nitride ~ 10
Development financed by consortium
between Rosatom and private investors
Construction of prototype to start 2017
in Dimitrovgrad
httpwwwakmeengineeringcomsvbr10
0html14
SVBR Project is a pilot project for SC Rosatom in terms of implementation of
large-scale high-tech projects in the nuclear industry jointly with a commercial
partner
The project implementation form is also new and is connected not only with
outsourcing of commercial investments but also with establishing a project
management and business development joint venture
The project is a part of the Federal Target Program Nuclear Power
Technologies of the New Generation for 2010 - 2015 and until 2020 It is also
one of the components of New technological platform closed fuel cycle and
fast reactors project which is being realized under Commission for
Modernization and Technological Development of Russias Economy
The final product of the project is the basic technology of a lead-bismuth fast-
neutron reactor adapted to a civil project including construction of a
100MW(e) modular power generating plant and associated range of 100
MW(e) aliquot products
15
The intrinsic security and inherent safety properties
(at physical level) of RU SVBR-100 allow making the structural
design of power generating units much less elaborate and using
the modular design principle
The RU SVBR-100 structural design includes the requirements
of versatility in terms of the fuel used to enable fast conversion
to MOX and later to nitride fuel as well to serve the basis for
fuel self-sufficiency in the closed nuclear fuel cycle
The RU design and parameters allow setting up factory
manufacture of RU modules and their delivery to the site of
installation by railroad or vehicle which significantly reduces
the labor costs and lead time of nuclear power plant
construction Serial manufacture will presumably allow
achieving much lower production cost and stable quality of the
product16
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (1)
The main effect in providing a high safety level of the SVBR-100 RF
is achieved due to use of fast neutron reactor heavy liquid-metal
coolant and integral design of the reactor with total elimination of
pipelines with radioactive coolant beyond the monoblock vessel
The reactor possesses a negative void reactivity effect and
negative feedbacks the efficiency of the strongest absorbing rod does
not exceed 1 $ And that coupled with technical realization of the
control and protection system (CPS) eliminates prompt neutron
runaway of the reactor
The high boiling point of coolant heightens reliability of heat
removal from the core and safety due to lack of the heat transfer
crisis Also being coupled with a provided safeguard casing of the
monoblock that eliminates loss of coolant accidents (LOCA) and
high pressure radioactive exhausts
17
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (2)
The low pressure in the primary circuit reduces the risk
of its tightness failure and makes possible lessening the thickness
of reactor vesselrsquos walls and diminishing the limitations imposed
on the rate of temperature change according to thermal-cycling
strength conditions
The RF components do not contain materials releasing hydrogen
as a result of thermal and radiation effects and chemical reactions
with coolant water and air Therefore in an event of tightness
failure in the primary circuit the likelihood of chemical explosions
and fires is virtually eliminated
The circulation scheme of lead-bismuth coolant (LBC) provides
elimination of watersteam ingress into the core in an event of steam
generator (SG) leak due to effective gravitational separation of
steam on a free LBC level in the monoblock18
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
Breeding trend
BR-1 BR ~ 25
BR-2 BR ~16
BN-350 BR ~12
BN-800 BR ~ 10 divide 11
Factors
spectrum softening oxide fuel
(light element - O) sodium
coolant ( - Na)
priority of safety Na plenum
and exclusion of upper blanket
thin radial blanket
Is the high breeding actual goal
8
Conception of ldquofast reactor start
from U-235rdquo
9
Main conceptual ways of reactor
safety improvement
bull Minimization of excess reactivity for the fuel burn-up
bull Decrease of sodium void reactivity effect
bull Use of passive devices for reactivity control
bull Use of passive devices for decay heat removal
10
BN-800 fast reactor
BN-800 is first post-Soviet fast
reactor in Russia (start up at 2014)
This project is aimed at the
development of the fuel cycle
infrastructure and mastering of the
new types of fuel (MOX fuel)
Sodium plenum making it possible
to assure zero void reactivity effect
and passive safety systems are
special features of BN-800 reactor
design
Tests of these elements would lead to
the progress in the area of fast
reactor safety
1 - vessel 2 -guard vessel 3 - core 4 -
core diagrid 5 - core catcher 6 - silo 7 -
main sodium pump 8 - upper
stationary shielding 9 - large rotating
plug 10 ndashcentral rotating plug 11 -
protection cap 12 - refueling
mechanism 13 - small rotating plug 14
ndash intermediate heat exchanger11
BN-1200 fast reactor
12
Fuel cyclebullfuel ndash mixed oxide or nitridebulllow power density in the corebullexternal fuel cycle duration - 3 yearsbullBR ndash 12 (oxide) -13 (nitride BRcore ~1) bullMA utilization in the basic fuelSafetybull2 types of passive control rodsbullflattened core sodium plenumbullintegration of all primary sodium systems in the reactor vessel to eliminate radioactive sodium leaksEconomical characteristicsbulloptimization of layout approachesbullincrease of load factor by transition to one-year refuelling intervalbullincrease of the fuel burn-up
1 - intermediate heat exchanger 2 - reactor vessel3 - guard vessel 4 - silo 5 - core diagrid 6 - core catcher 7 - reactor core 8 ndash pump nozzle 9 ndash main sodium pump 10 ndash cold trap 11 ndashcontrol rod drives 12 ndash rotating plug
APL-705 experimental
(1971)
Test facility (1951)
APL-705 serial
(1976-1996)
SVBR-100 (plan - 2017)
The stages of heavy metal
coolant technology
development
13
First rdquocommercialrdquo LFR SVBR-100
Based on silicon oxide protected
ferritic-martensitic steel cladding the
Soviet submarine reactor design has been
converted to a commercial concept
SVBR-100 lead-bismuth cooled
reactor with 100 MWe power using
MOX or nitride fuel MOX gives
breeding ratio ~ 084 nitride ~ 10
Development financed by consortium
between Rosatom and private investors
Construction of prototype to start 2017
in Dimitrovgrad
httpwwwakmeengineeringcomsvbr10
0html14
SVBR Project is a pilot project for SC Rosatom in terms of implementation of
large-scale high-tech projects in the nuclear industry jointly with a commercial
partner
The project implementation form is also new and is connected not only with
outsourcing of commercial investments but also with establishing a project
management and business development joint venture
The project is a part of the Federal Target Program Nuclear Power
Technologies of the New Generation for 2010 - 2015 and until 2020 It is also
one of the components of New technological platform closed fuel cycle and
fast reactors project which is being realized under Commission for
Modernization and Technological Development of Russias Economy
The final product of the project is the basic technology of a lead-bismuth fast-
neutron reactor adapted to a civil project including construction of a
100MW(e) modular power generating plant and associated range of 100
MW(e) aliquot products
15
The intrinsic security and inherent safety properties
(at physical level) of RU SVBR-100 allow making the structural
design of power generating units much less elaborate and using
the modular design principle
The RU SVBR-100 structural design includes the requirements
of versatility in terms of the fuel used to enable fast conversion
to MOX and later to nitride fuel as well to serve the basis for
fuel self-sufficiency in the closed nuclear fuel cycle
The RU design and parameters allow setting up factory
manufacture of RU modules and their delivery to the site of
installation by railroad or vehicle which significantly reduces
the labor costs and lead time of nuclear power plant
construction Serial manufacture will presumably allow
achieving much lower production cost and stable quality of the
product16
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (1)
The main effect in providing a high safety level of the SVBR-100 RF
is achieved due to use of fast neutron reactor heavy liquid-metal
coolant and integral design of the reactor with total elimination of
pipelines with radioactive coolant beyond the monoblock vessel
The reactor possesses a negative void reactivity effect and
negative feedbacks the efficiency of the strongest absorbing rod does
not exceed 1 $ And that coupled with technical realization of the
control and protection system (CPS) eliminates prompt neutron
runaway of the reactor
The high boiling point of coolant heightens reliability of heat
removal from the core and safety due to lack of the heat transfer
crisis Also being coupled with a provided safeguard casing of the
monoblock that eliminates loss of coolant accidents (LOCA) and
high pressure radioactive exhausts
17
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (2)
The low pressure in the primary circuit reduces the risk
of its tightness failure and makes possible lessening the thickness
of reactor vesselrsquos walls and diminishing the limitations imposed
on the rate of temperature change according to thermal-cycling
strength conditions
The RF components do not contain materials releasing hydrogen
as a result of thermal and radiation effects and chemical reactions
with coolant water and air Therefore in an event of tightness
failure in the primary circuit the likelihood of chemical explosions
and fires is virtually eliminated
The circulation scheme of lead-bismuth coolant (LBC) provides
elimination of watersteam ingress into the core in an event of steam
generator (SG) leak due to effective gravitational separation of
steam on a free LBC level in the monoblock18
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
Conception of ldquofast reactor start
from U-235rdquo
9
Main conceptual ways of reactor
safety improvement
bull Minimization of excess reactivity for the fuel burn-up
bull Decrease of sodium void reactivity effect
bull Use of passive devices for reactivity control
bull Use of passive devices for decay heat removal
10
BN-800 fast reactor
BN-800 is first post-Soviet fast
reactor in Russia (start up at 2014)
This project is aimed at the
development of the fuel cycle
infrastructure and mastering of the
new types of fuel (MOX fuel)
Sodium plenum making it possible
to assure zero void reactivity effect
and passive safety systems are
special features of BN-800 reactor
design
Tests of these elements would lead to
the progress in the area of fast
reactor safety
1 - vessel 2 -guard vessel 3 - core 4 -
core diagrid 5 - core catcher 6 - silo 7 -
main sodium pump 8 - upper
stationary shielding 9 - large rotating
plug 10 ndashcentral rotating plug 11 -
protection cap 12 - refueling
mechanism 13 - small rotating plug 14
ndash intermediate heat exchanger11
BN-1200 fast reactor
12
Fuel cyclebullfuel ndash mixed oxide or nitridebulllow power density in the corebullexternal fuel cycle duration - 3 yearsbullBR ndash 12 (oxide) -13 (nitride BRcore ~1) bullMA utilization in the basic fuelSafetybull2 types of passive control rodsbullflattened core sodium plenumbullintegration of all primary sodium systems in the reactor vessel to eliminate radioactive sodium leaksEconomical characteristicsbulloptimization of layout approachesbullincrease of load factor by transition to one-year refuelling intervalbullincrease of the fuel burn-up
1 - intermediate heat exchanger 2 - reactor vessel3 - guard vessel 4 - silo 5 - core diagrid 6 - core catcher 7 - reactor core 8 ndash pump nozzle 9 ndash main sodium pump 10 ndash cold trap 11 ndashcontrol rod drives 12 ndash rotating plug
APL-705 experimental
(1971)
Test facility (1951)
APL-705 serial
(1976-1996)
SVBR-100 (plan - 2017)
The stages of heavy metal
coolant technology
development
13
First rdquocommercialrdquo LFR SVBR-100
Based on silicon oxide protected
ferritic-martensitic steel cladding the
Soviet submarine reactor design has been
converted to a commercial concept
SVBR-100 lead-bismuth cooled
reactor with 100 MWe power using
MOX or nitride fuel MOX gives
breeding ratio ~ 084 nitride ~ 10
Development financed by consortium
between Rosatom and private investors
Construction of prototype to start 2017
in Dimitrovgrad
httpwwwakmeengineeringcomsvbr10
0html14
SVBR Project is a pilot project for SC Rosatom in terms of implementation of
large-scale high-tech projects in the nuclear industry jointly with a commercial
partner
The project implementation form is also new and is connected not only with
outsourcing of commercial investments but also with establishing a project
management and business development joint venture
The project is a part of the Federal Target Program Nuclear Power
Technologies of the New Generation for 2010 - 2015 and until 2020 It is also
one of the components of New technological platform closed fuel cycle and
fast reactors project which is being realized under Commission for
Modernization and Technological Development of Russias Economy
The final product of the project is the basic technology of a lead-bismuth fast-
neutron reactor adapted to a civil project including construction of a
100MW(e) modular power generating plant and associated range of 100
MW(e) aliquot products
15
The intrinsic security and inherent safety properties
(at physical level) of RU SVBR-100 allow making the structural
design of power generating units much less elaborate and using
the modular design principle
The RU SVBR-100 structural design includes the requirements
of versatility in terms of the fuel used to enable fast conversion
to MOX and later to nitride fuel as well to serve the basis for
fuel self-sufficiency in the closed nuclear fuel cycle
The RU design and parameters allow setting up factory
manufacture of RU modules and their delivery to the site of
installation by railroad or vehicle which significantly reduces
the labor costs and lead time of nuclear power plant
construction Serial manufacture will presumably allow
achieving much lower production cost and stable quality of the
product16
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (1)
The main effect in providing a high safety level of the SVBR-100 RF
is achieved due to use of fast neutron reactor heavy liquid-metal
coolant and integral design of the reactor with total elimination of
pipelines with radioactive coolant beyond the monoblock vessel
The reactor possesses a negative void reactivity effect and
negative feedbacks the efficiency of the strongest absorbing rod does
not exceed 1 $ And that coupled with technical realization of the
control and protection system (CPS) eliminates prompt neutron
runaway of the reactor
The high boiling point of coolant heightens reliability of heat
removal from the core and safety due to lack of the heat transfer
crisis Also being coupled with a provided safeguard casing of the
monoblock that eliminates loss of coolant accidents (LOCA) and
high pressure radioactive exhausts
17
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (2)
The low pressure in the primary circuit reduces the risk
of its tightness failure and makes possible lessening the thickness
of reactor vesselrsquos walls and diminishing the limitations imposed
on the rate of temperature change according to thermal-cycling
strength conditions
The RF components do not contain materials releasing hydrogen
as a result of thermal and radiation effects and chemical reactions
with coolant water and air Therefore in an event of tightness
failure in the primary circuit the likelihood of chemical explosions
and fires is virtually eliminated
The circulation scheme of lead-bismuth coolant (LBC) provides
elimination of watersteam ingress into the core in an event of steam
generator (SG) leak due to effective gravitational separation of
steam on a free LBC level in the monoblock18
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
Main conceptual ways of reactor
safety improvement
bull Minimization of excess reactivity for the fuel burn-up
bull Decrease of sodium void reactivity effect
bull Use of passive devices for reactivity control
bull Use of passive devices for decay heat removal
10
BN-800 fast reactor
BN-800 is first post-Soviet fast
reactor in Russia (start up at 2014)
This project is aimed at the
development of the fuel cycle
infrastructure and mastering of the
new types of fuel (MOX fuel)
Sodium plenum making it possible
to assure zero void reactivity effect
and passive safety systems are
special features of BN-800 reactor
design
Tests of these elements would lead to
the progress in the area of fast
reactor safety
1 - vessel 2 -guard vessel 3 - core 4 -
core diagrid 5 - core catcher 6 - silo 7 -
main sodium pump 8 - upper
stationary shielding 9 - large rotating
plug 10 ndashcentral rotating plug 11 -
protection cap 12 - refueling
mechanism 13 - small rotating plug 14
ndash intermediate heat exchanger11
BN-1200 fast reactor
12
Fuel cyclebullfuel ndash mixed oxide or nitridebulllow power density in the corebullexternal fuel cycle duration - 3 yearsbullBR ndash 12 (oxide) -13 (nitride BRcore ~1) bullMA utilization in the basic fuelSafetybull2 types of passive control rodsbullflattened core sodium plenumbullintegration of all primary sodium systems in the reactor vessel to eliminate radioactive sodium leaksEconomical characteristicsbulloptimization of layout approachesbullincrease of load factor by transition to one-year refuelling intervalbullincrease of the fuel burn-up
1 - intermediate heat exchanger 2 - reactor vessel3 - guard vessel 4 - silo 5 - core diagrid 6 - core catcher 7 - reactor core 8 ndash pump nozzle 9 ndash main sodium pump 10 ndash cold trap 11 ndashcontrol rod drives 12 ndash rotating plug
APL-705 experimental
(1971)
Test facility (1951)
APL-705 serial
(1976-1996)
SVBR-100 (plan - 2017)
The stages of heavy metal
coolant technology
development
13
First rdquocommercialrdquo LFR SVBR-100
Based on silicon oxide protected
ferritic-martensitic steel cladding the
Soviet submarine reactor design has been
converted to a commercial concept
SVBR-100 lead-bismuth cooled
reactor with 100 MWe power using
MOX or nitride fuel MOX gives
breeding ratio ~ 084 nitride ~ 10
Development financed by consortium
between Rosatom and private investors
Construction of prototype to start 2017
in Dimitrovgrad
httpwwwakmeengineeringcomsvbr10
0html14
SVBR Project is a pilot project for SC Rosatom in terms of implementation of
large-scale high-tech projects in the nuclear industry jointly with a commercial
partner
The project implementation form is also new and is connected not only with
outsourcing of commercial investments but also with establishing a project
management and business development joint venture
The project is a part of the Federal Target Program Nuclear Power
Technologies of the New Generation for 2010 - 2015 and until 2020 It is also
one of the components of New technological platform closed fuel cycle and
fast reactors project which is being realized under Commission for
Modernization and Technological Development of Russias Economy
The final product of the project is the basic technology of a lead-bismuth fast-
neutron reactor adapted to a civil project including construction of a
100MW(e) modular power generating plant and associated range of 100
MW(e) aliquot products
15
The intrinsic security and inherent safety properties
(at physical level) of RU SVBR-100 allow making the structural
design of power generating units much less elaborate and using
the modular design principle
The RU SVBR-100 structural design includes the requirements
of versatility in terms of the fuel used to enable fast conversion
to MOX and later to nitride fuel as well to serve the basis for
fuel self-sufficiency in the closed nuclear fuel cycle
The RU design and parameters allow setting up factory
manufacture of RU modules and their delivery to the site of
installation by railroad or vehicle which significantly reduces
the labor costs and lead time of nuclear power plant
construction Serial manufacture will presumably allow
achieving much lower production cost and stable quality of the
product16
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (1)
The main effect in providing a high safety level of the SVBR-100 RF
is achieved due to use of fast neutron reactor heavy liquid-metal
coolant and integral design of the reactor with total elimination of
pipelines with radioactive coolant beyond the monoblock vessel
The reactor possesses a negative void reactivity effect and
negative feedbacks the efficiency of the strongest absorbing rod does
not exceed 1 $ And that coupled with technical realization of the
control and protection system (CPS) eliminates prompt neutron
runaway of the reactor
The high boiling point of coolant heightens reliability of heat
removal from the core and safety due to lack of the heat transfer
crisis Also being coupled with a provided safeguard casing of the
monoblock that eliminates loss of coolant accidents (LOCA) and
high pressure radioactive exhausts
17
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (2)
The low pressure in the primary circuit reduces the risk
of its tightness failure and makes possible lessening the thickness
of reactor vesselrsquos walls and diminishing the limitations imposed
on the rate of temperature change according to thermal-cycling
strength conditions
The RF components do not contain materials releasing hydrogen
as a result of thermal and radiation effects and chemical reactions
with coolant water and air Therefore in an event of tightness
failure in the primary circuit the likelihood of chemical explosions
and fires is virtually eliminated
The circulation scheme of lead-bismuth coolant (LBC) provides
elimination of watersteam ingress into the core in an event of steam
generator (SG) leak due to effective gravitational separation of
steam on a free LBC level in the monoblock18
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
BN-800 fast reactor
BN-800 is first post-Soviet fast
reactor in Russia (start up at 2014)
This project is aimed at the
development of the fuel cycle
infrastructure and mastering of the
new types of fuel (MOX fuel)
Sodium plenum making it possible
to assure zero void reactivity effect
and passive safety systems are
special features of BN-800 reactor
design
Tests of these elements would lead to
the progress in the area of fast
reactor safety
1 - vessel 2 -guard vessel 3 - core 4 -
core diagrid 5 - core catcher 6 - silo 7 -
main sodium pump 8 - upper
stationary shielding 9 - large rotating
plug 10 ndashcentral rotating plug 11 -
protection cap 12 - refueling
mechanism 13 - small rotating plug 14
ndash intermediate heat exchanger11
BN-1200 fast reactor
12
Fuel cyclebullfuel ndash mixed oxide or nitridebulllow power density in the corebullexternal fuel cycle duration - 3 yearsbullBR ndash 12 (oxide) -13 (nitride BRcore ~1) bullMA utilization in the basic fuelSafetybull2 types of passive control rodsbullflattened core sodium plenumbullintegration of all primary sodium systems in the reactor vessel to eliminate radioactive sodium leaksEconomical characteristicsbulloptimization of layout approachesbullincrease of load factor by transition to one-year refuelling intervalbullincrease of the fuel burn-up
1 - intermediate heat exchanger 2 - reactor vessel3 - guard vessel 4 - silo 5 - core diagrid 6 - core catcher 7 - reactor core 8 ndash pump nozzle 9 ndash main sodium pump 10 ndash cold trap 11 ndashcontrol rod drives 12 ndash rotating plug
APL-705 experimental
(1971)
Test facility (1951)
APL-705 serial
(1976-1996)
SVBR-100 (plan - 2017)
The stages of heavy metal
coolant technology
development
13
First rdquocommercialrdquo LFR SVBR-100
Based on silicon oxide protected
ferritic-martensitic steel cladding the
Soviet submarine reactor design has been
converted to a commercial concept
SVBR-100 lead-bismuth cooled
reactor with 100 MWe power using
MOX or nitride fuel MOX gives
breeding ratio ~ 084 nitride ~ 10
Development financed by consortium
between Rosatom and private investors
Construction of prototype to start 2017
in Dimitrovgrad
httpwwwakmeengineeringcomsvbr10
0html14
SVBR Project is a pilot project for SC Rosatom in terms of implementation of
large-scale high-tech projects in the nuclear industry jointly with a commercial
partner
The project implementation form is also new and is connected not only with
outsourcing of commercial investments but also with establishing a project
management and business development joint venture
The project is a part of the Federal Target Program Nuclear Power
Technologies of the New Generation for 2010 - 2015 and until 2020 It is also
one of the components of New technological platform closed fuel cycle and
fast reactors project which is being realized under Commission for
Modernization and Technological Development of Russias Economy
The final product of the project is the basic technology of a lead-bismuth fast-
neutron reactor adapted to a civil project including construction of a
100MW(e) modular power generating plant and associated range of 100
MW(e) aliquot products
15
The intrinsic security and inherent safety properties
(at physical level) of RU SVBR-100 allow making the structural
design of power generating units much less elaborate and using
the modular design principle
The RU SVBR-100 structural design includes the requirements
of versatility in terms of the fuel used to enable fast conversion
to MOX and later to nitride fuel as well to serve the basis for
fuel self-sufficiency in the closed nuclear fuel cycle
The RU design and parameters allow setting up factory
manufacture of RU modules and their delivery to the site of
installation by railroad or vehicle which significantly reduces
the labor costs and lead time of nuclear power plant
construction Serial manufacture will presumably allow
achieving much lower production cost and stable quality of the
product16
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (1)
The main effect in providing a high safety level of the SVBR-100 RF
is achieved due to use of fast neutron reactor heavy liquid-metal
coolant and integral design of the reactor with total elimination of
pipelines with radioactive coolant beyond the monoblock vessel
The reactor possesses a negative void reactivity effect and
negative feedbacks the efficiency of the strongest absorbing rod does
not exceed 1 $ And that coupled with technical realization of the
control and protection system (CPS) eliminates prompt neutron
runaway of the reactor
The high boiling point of coolant heightens reliability of heat
removal from the core and safety due to lack of the heat transfer
crisis Also being coupled with a provided safeguard casing of the
monoblock that eliminates loss of coolant accidents (LOCA) and
high pressure radioactive exhausts
17
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (2)
The low pressure in the primary circuit reduces the risk
of its tightness failure and makes possible lessening the thickness
of reactor vesselrsquos walls and diminishing the limitations imposed
on the rate of temperature change according to thermal-cycling
strength conditions
The RF components do not contain materials releasing hydrogen
as a result of thermal and radiation effects and chemical reactions
with coolant water and air Therefore in an event of tightness
failure in the primary circuit the likelihood of chemical explosions
and fires is virtually eliminated
The circulation scheme of lead-bismuth coolant (LBC) provides
elimination of watersteam ingress into the core in an event of steam
generator (SG) leak due to effective gravitational separation of
steam on a free LBC level in the monoblock18
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
BN-1200 fast reactor
12
Fuel cyclebullfuel ndash mixed oxide or nitridebulllow power density in the corebullexternal fuel cycle duration - 3 yearsbullBR ndash 12 (oxide) -13 (nitride BRcore ~1) bullMA utilization in the basic fuelSafetybull2 types of passive control rodsbullflattened core sodium plenumbullintegration of all primary sodium systems in the reactor vessel to eliminate radioactive sodium leaksEconomical characteristicsbulloptimization of layout approachesbullincrease of load factor by transition to one-year refuelling intervalbullincrease of the fuel burn-up
1 - intermediate heat exchanger 2 - reactor vessel3 - guard vessel 4 - silo 5 - core diagrid 6 - core catcher 7 - reactor core 8 ndash pump nozzle 9 ndash main sodium pump 10 ndash cold trap 11 ndashcontrol rod drives 12 ndash rotating plug
APL-705 experimental
(1971)
Test facility (1951)
APL-705 serial
(1976-1996)
SVBR-100 (plan - 2017)
The stages of heavy metal
coolant technology
development
13
First rdquocommercialrdquo LFR SVBR-100
Based on silicon oxide protected
ferritic-martensitic steel cladding the
Soviet submarine reactor design has been
converted to a commercial concept
SVBR-100 lead-bismuth cooled
reactor with 100 MWe power using
MOX or nitride fuel MOX gives
breeding ratio ~ 084 nitride ~ 10
Development financed by consortium
between Rosatom and private investors
Construction of prototype to start 2017
in Dimitrovgrad
httpwwwakmeengineeringcomsvbr10
0html14
SVBR Project is a pilot project for SC Rosatom in terms of implementation of
large-scale high-tech projects in the nuclear industry jointly with a commercial
partner
The project implementation form is also new and is connected not only with
outsourcing of commercial investments but also with establishing a project
management and business development joint venture
The project is a part of the Federal Target Program Nuclear Power
Technologies of the New Generation for 2010 - 2015 and until 2020 It is also
one of the components of New technological platform closed fuel cycle and
fast reactors project which is being realized under Commission for
Modernization and Technological Development of Russias Economy
The final product of the project is the basic technology of a lead-bismuth fast-
neutron reactor adapted to a civil project including construction of a
100MW(e) modular power generating plant and associated range of 100
MW(e) aliquot products
15
The intrinsic security and inherent safety properties
(at physical level) of RU SVBR-100 allow making the structural
design of power generating units much less elaborate and using
the modular design principle
The RU SVBR-100 structural design includes the requirements
of versatility in terms of the fuel used to enable fast conversion
to MOX and later to nitride fuel as well to serve the basis for
fuel self-sufficiency in the closed nuclear fuel cycle
The RU design and parameters allow setting up factory
manufacture of RU modules and their delivery to the site of
installation by railroad or vehicle which significantly reduces
the labor costs and lead time of nuclear power plant
construction Serial manufacture will presumably allow
achieving much lower production cost and stable quality of the
product16
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (1)
The main effect in providing a high safety level of the SVBR-100 RF
is achieved due to use of fast neutron reactor heavy liquid-metal
coolant and integral design of the reactor with total elimination of
pipelines with radioactive coolant beyond the monoblock vessel
The reactor possesses a negative void reactivity effect and
negative feedbacks the efficiency of the strongest absorbing rod does
not exceed 1 $ And that coupled with technical realization of the
control and protection system (CPS) eliminates prompt neutron
runaway of the reactor
The high boiling point of coolant heightens reliability of heat
removal from the core and safety due to lack of the heat transfer
crisis Also being coupled with a provided safeguard casing of the
monoblock that eliminates loss of coolant accidents (LOCA) and
high pressure radioactive exhausts
17
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (2)
The low pressure in the primary circuit reduces the risk
of its tightness failure and makes possible lessening the thickness
of reactor vesselrsquos walls and diminishing the limitations imposed
on the rate of temperature change according to thermal-cycling
strength conditions
The RF components do not contain materials releasing hydrogen
as a result of thermal and radiation effects and chemical reactions
with coolant water and air Therefore in an event of tightness
failure in the primary circuit the likelihood of chemical explosions
and fires is virtually eliminated
The circulation scheme of lead-bismuth coolant (LBC) provides
elimination of watersteam ingress into the core in an event of steam
generator (SG) leak due to effective gravitational separation of
steam on a free LBC level in the monoblock18
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
APL-705 experimental
(1971)
Test facility (1951)
APL-705 serial
(1976-1996)
SVBR-100 (plan - 2017)
The stages of heavy metal
coolant technology
development
13
First rdquocommercialrdquo LFR SVBR-100
Based on silicon oxide protected
ferritic-martensitic steel cladding the
Soviet submarine reactor design has been
converted to a commercial concept
SVBR-100 lead-bismuth cooled
reactor with 100 MWe power using
MOX or nitride fuel MOX gives
breeding ratio ~ 084 nitride ~ 10
Development financed by consortium
between Rosatom and private investors
Construction of prototype to start 2017
in Dimitrovgrad
httpwwwakmeengineeringcomsvbr10
0html14
SVBR Project is a pilot project for SC Rosatom in terms of implementation of
large-scale high-tech projects in the nuclear industry jointly with a commercial
partner
The project implementation form is also new and is connected not only with
outsourcing of commercial investments but also with establishing a project
management and business development joint venture
The project is a part of the Federal Target Program Nuclear Power
Technologies of the New Generation for 2010 - 2015 and until 2020 It is also
one of the components of New technological platform closed fuel cycle and
fast reactors project which is being realized under Commission for
Modernization and Technological Development of Russias Economy
The final product of the project is the basic technology of a lead-bismuth fast-
neutron reactor adapted to a civil project including construction of a
100MW(e) modular power generating plant and associated range of 100
MW(e) aliquot products
15
The intrinsic security and inherent safety properties
(at physical level) of RU SVBR-100 allow making the structural
design of power generating units much less elaborate and using
the modular design principle
The RU SVBR-100 structural design includes the requirements
of versatility in terms of the fuel used to enable fast conversion
to MOX and later to nitride fuel as well to serve the basis for
fuel self-sufficiency in the closed nuclear fuel cycle
The RU design and parameters allow setting up factory
manufacture of RU modules and their delivery to the site of
installation by railroad or vehicle which significantly reduces
the labor costs and lead time of nuclear power plant
construction Serial manufacture will presumably allow
achieving much lower production cost and stable quality of the
product16
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (1)
The main effect in providing a high safety level of the SVBR-100 RF
is achieved due to use of fast neutron reactor heavy liquid-metal
coolant and integral design of the reactor with total elimination of
pipelines with radioactive coolant beyond the monoblock vessel
The reactor possesses a negative void reactivity effect and
negative feedbacks the efficiency of the strongest absorbing rod does
not exceed 1 $ And that coupled with technical realization of the
control and protection system (CPS) eliminates prompt neutron
runaway of the reactor
The high boiling point of coolant heightens reliability of heat
removal from the core and safety due to lack of the heat transfer
crisis Also being coupled with a provided safeguard casing of the
monoblock that eliminates loss of coolant accidents (LOCA) and
high pressure radioactive exhausts
17
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (2)
The low pressure in the primary circuit reduces the risk
of its tightness failure and makes possible lessening the thickness
of reactor vesselrsquos walls and diminishing the limitations imposed
on the rate of temperature change according to thermal-cycling
strength conditions
The RF components do not contain materials releasing hydrogen
as a result of thermal and radiation effects and chemical reactions
with coolant water and air Therefore in an event of tightness
failure in the primary circuit the likelihood of chemical explosions
and fires is virtually eliminated
The circulation scheme of lead-bismuth coolant (LBC) provides
elimination of watersteam ingress into the core in an event of steam
generator (SG) leak due to effective gravitational separation of
steam on a free LBC level in the monoblock18
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
First rdquocommercialrdquo LFR SVBR-100
Based on silicon oxide protected
ferritic-martensitic steel cladding the
Soviet submarine reactor design has been
converted to a commercial concept
SVBR-100 lead-bismuth cooled
reactor with 100 MWe power using
MOX or nitride fuel MOX gives
breeding ratio ~ 084 nitride ~ 10
Development financed by consortium
between Rosatom and private investors
Construction of prototype to start 2017
in Dimitrovgrad
httpwwwakmeengineeringcomsvbr10
0html14
SVBR Project is a pilot project for SC Rosatom in terms of implementation of
large-scale high-tech projects in the nuclear industry jointly with a commercial
partner
The project implementation form is also new and is connected not only with
outsourcing of commercial investments but also with establishing a project
management and business development joint venture
The project is a part of the Federal Target Program Nuclear Power
Technologies of the New Generation for 2010 - 2015 and until 2020 It is also
one of the components of New technological platform closed fuel cycle and
fast reactors project which is being realized under Commission for
Modernization and Technological Development of Russias Economy
The final product of the project is the basic technology of a lead-bismuth fast-
neutron reactor adapted to a civil project including construction of a
100MW(e) modular power generating plant and associated range of 100
MW(e) aliquot products
15
The intrinsic security and inherent safety properties
(at physical level) of RU SVBR-100 allow making the structural
design of power generating units much less elaborate and using
the modular design principle
The RU SVBR-100 structural design includes the requirements
of versatility in terms of the fuel used to enable fast conversion
to MOX and later to nitride fuel as well to serve the basis for
fuel self-sufficiency in the closed nuclear fuel cycle
The RU design and parameters allow setting up factory
manufacture of RU modules and their delivery to the site of
installation by railroad or vehicle which significantly reduces
the labor costs and lead time of nuclear power plant
construction Serial manufacture will presumably allow
achieving much lower production cost and stable quality of the
product16
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (1)
The main effect in providing a high safety level of the SVBR-100 RF
is achieved due to use of fast neutron reactor heavy liquid-metal
coolant and integral design of the reactor with total elimination of
pipelines with radioactive coolant beyond the monoblock vessel
The reactor possesses a negative void reactivity effect and
negative feedbacks the efficiency of the strongest absorbing rod does
not exceed 1 $ And that coupled with technical realization of the
control and protection system (CPS) eliminates prompt neutron
runaway of the reactor
The high boiling point of coolant heightens reliability of heat
removal from the core and safety due to lack of the heat transfer
crisis Also being coupled with a provided safeguard casing of the
monoblock that eliminates loss of coolant accidents (LOCA) and
high pressure radioactive exhausts
17
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (2)
The low pressure in the primary circuit reduces the risk
of its tightness failure and makes possible lessening the thickness
of reactor vesselrsquos walls and diminishing the limitations imposed
on the rate of temperature change according to thermal-cycling
strength conditions
The RF components do not contain materials releasing hydrogen
as a result of thermal and radiation effects and chemical reactions
with coolant water and air Therefore in an event of tightness
failure in the primary circuit the likelihood of chemical explosions
and fires is virtually eliminated
The circulation scheme of lead-bismuth coolant (LBC) provides
elimination of watersteam ingress into the core in an event of steam
generator (SG) leak due to effective gravitational separation of
steam on a free LBC level in the monoblock18
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
SVBR Project is a pilot project for SC Rosatom in terms of implementation of
large-scale high-tech projects in the nuclear industry jointly with a commercial
partner
The project implementation form is also new and is connected not only with
outsourcing of commercial investments but also with establishing a project
management and business development joint venture
The project is a part of the Federal Target Program Nuclear Power
Technologies of the New Generation for 2010 - 2015 and until 2020 It is also
one of the components of New technological platform closed fuel cycle and
fast reactors project which is being realized under Commission for
Modernization and Technological Development of Russias Economy
The final product of the project is the basic technology of a lead-bismuth fast-
neutron reactor adapted to a civil project including construction of a
100MW(e) modular power generating plant and associated range of 100
MW(e) aliquot products
15
The intrinsic security and inherent safety properties
(at physical level) of RU SVBR-100 allow making the structural
design of power generating units much less elaborate and using
the modular design principle
The RU SVBR-100 structural design includes the requirements
of versatility in terms of the fuel used to enable fast conversion
to MOX and later to nitride fuel as well to serve the basis for
fuel self-sufficiency in the closed nuclear fuel cycle
The RU design and parameters allow setting up factory
manufacture of RU modules and their delivery to the site of
installation by railroad or vehicle which significantly reduces
the labor costs and lead time of nuclear power plant
construction Serial manufacture will presumably allow
achieving much lower production cost and stable quality of the
product16
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (1)
The main effect in providing a high safety level of the SVBR-100 RF
is achieved due to use of fast neutron reactor heavy liquid-metal
coolant and integral design of the reactor with total elimination of
pipelines with radioactive coolant beyond the monoblock vessel
The reactor possesses a negative void reactivity effect and
negative feedbacks the efficiency of the strongest absorbing rod does
not exceed 1 $ And that coupled with technical realization of the
control and protection system (CPS) eliminates prompt neutron
runaway of the reactor
The high boiling point of coolant heightens reliability of heat
removal from the core and safety due to lack of the heat transfer
crisis Also being coupled with a provided safeguard casing of the
monoblock that eliminates loss of coolant accidents (LOCA) and
high pressure radioactive exhausts
17
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (2)
The low pressure in the primary circuit reduces the risk
of its tightness failure and makes possible lessening the thickness
of reactor vesselrsquos walls and diminishing the limitations imposed
on the rate of temperature change according to thermal-cycling
strength conditions
The RF components do not contain materials releasing hydrogen
as a result of thermal and radiation effects and chemical reactions
with coolant water and air Therefore in an event of tightness
failure in the primary circuit the likelihood of chemical explosions
and fires is virtually eliminated
The circulation scheme of lead-bismuth coolant (LBC) provides
elimination of watersteam ingress into the core in an event of steam
generator (SG) leak due to effective gravitational separation of
steam on a free LBC level in the monoblock18
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
The intrinsic security and inherent safety properties
(at physical level) of RU SVBR-100 allow making the structural
design of power generating units much less elaborate and using
the modular design principle
The RU SVBR-100 structural design includes the requirements
of versatility in terms of the fuel used to enable fast conversion
to MOX and later to nitride fuel as well to serve the basis for
fuel self-sufficiency in the closed nuclear fuel cycle
The RU design and parameters allow setting up factory
manufacture of RU modules and their delivery to the site of
installation by railroad or vehicle which significantly reduces
the labor costs and lead time of nuclear power plant
construction Serial manufacture will presumably allow
achieving much lower production cost and stable quality of the
product16
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (1)
The main effect in providing a high safety level of the SVBR-100 RF
is achieved due to use of fast neutron reactor heavy liquid-metal
coolant and integral design of the reactor with total elimination of
pipelines with radioactive coolant beyond the monoblock vessel
The reactor possesses a negative void reactivity effect and
negative feedbacks the efficiency of the strongest absorbing rod does
not exceed 1 $ And that coupled with technical realization of the
control and protection system (CPS) eliminates prompt neutron
runaway of the reactor
The high boiling point of coolant heightens reliability of heat
removal from the core and safety due to lack of the heat transfer
crisis Also being coupled with a provided safeguard casing of the
monoblock that eliminates loss of coolant accidents (LOCA) and
high pressure radioactive exhausts
17
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (2)
The low pressure in the primary circuit reduces the risk
of its tightness failure and makes possible lessening the thickness
of reactor vesselrsquos walls and diminishing the limitations imposed
on the rate of temperature change according to thermal-cycling
strength conditions
The RF components do not contain materials releasing hydrogen
as a result of thermal and radiation effects and chemical reactions
with coolant water and air Therefore in an event of tightness
failure in the primary circuit the likelihood of chemical explosions
and fires is virtually eliminated
The circulation scheme of lead-bismuth coolant (LBC) provides
elimination of watersteam ingress into the core in an event of steam
generator (SG) leak due to effective gravitational separation of
steam on a free LBC level in the monoblock18
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (1)
The main effect in providing a high safety level of the SVBR-100 RF
is achieved due to use of fast neutron reactor heavy liquid-metal
coolant and integral design of the reactor with total elimination of
pipelines with radioactive coolant beyond the monoblock vessel
The reactor possesses a negative void reactivity effect and
negative feedbacks the efficiency of the strongest absorbing rod does
not exceed 1 $ And that coupled with technical realization of the
control and protection system (CPS) eliminates prompt neutron
runaway of the reactor
The high boiling point of coolant heightens reliability of heat
removal from the core and safety due to lack of the heat transfer
crisis Also being coupled with a provided safeguard casing of the
monoblock that eliminates loss of coolant accidents (LOCA) and
high pressure radioactive exhausts
17
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (2)
The low pressure in the primary circuit reduces the risk
of its tightness failure and makes possible lessening the thickness
of reactor vesselrsquos walls and diminishing the limitations imposed
on the rate of temperature change according to thermal-cycling
strength conditions
The RF components do not contain materials releasing hydrogen
as a result of thermal and radiation effects and chemical reactions
with coolant water and air Therefore in an event of tightness
failure in the primary circuit the likelihood of chemical explosions
and fires is virtually eliminated
The circulation scheme of lead-bismuth coolant (LBC) provides
elimination of watersteam ingress into the core in an event of steam
generator (SG) leak due to effective gravitational separation of
steam on a free LBC level in the monoblock18
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (2)
The low pressure in the primary circuit reduces the risk
of its tightness failure and makes possible lessening the thickness
of reactor vesselrsquos walls and diminishing the limitations imposed
on the rate of temperature change according to thermal-cycling
strength conditions
The RF components do not contain materials releasing hydrogen
as a result of thermal and radiation effects and chemical reactions
with coolant water and air Therefore in an event of tightness
failure in the primary circuit the likelihood of chemical explosions
and fires is virtually eliminated
The circulation scheme of lead-bismuth coolant (LBC) provides
elimination of watersteam ingress into the core in an event of steam
generator (SG) leak due to effective gravitational separation of
steam on a free LBC level in the monoblock18
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (3)
Safety systems do not contain elements which actuation can be
blocked in an event of their failure or under impact of human
factors
- Removal of heat decay is provided passively by natural circulation
of LBC in the primary circuit This is realized by transferring heat
over four independent channels in the SG to the secondary circuit
water and then to the water tank of the passive heat removal system
(PHRS) with removal of generated steam into the atmosphere
- In an event of large leak in several SG tubes localization of SG
leak is provided passively while increasing the steam pressure in the
gas system over 05 MPa This is provided by using a safeguarding
membrane and discharging steam into the bubbling device (It
should be highlighted that operating experience has revealed that in
an event of small leak in the SG the RF does not need to be shut
down at once)19
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
INHERENT SELF-PROTECTION AND PASSIVE SAFETY
OF RF SVBR-100 (4)
When LBC temperature is increased over a specified value the rods
of the additional emergency protection system which are mounted
in ldquodryrdquo channels and are without drivers on the reactor lid
actuate passively by gravity due to fusible locks made of the alloy
with a corresponding melting temperature and holding the rods
in the upper position at normal temperature modes
In an event of postulated failure of all four channels of the PHRS
it is provided to flood the reactor vault by water from the tank
mounted above and transfer heat via the monoblock vessel
air gap and safeguard casing to the water with further removal
of generated steam into the atmosphere
20
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
Proposals for estimation
SVBR-100
21
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
RUSSIAN APPROACH TO FAST
REACTOR SAFETY ANALYSIS
SAFETY is its capability of keeping radiation doses
of personnel inhabitants and environment within
permissible limits under normal operating
conditions abnormal operating conditions and in
case of accident (OPB-8897)
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
1 Russian approach to fast reactor safety analysis was
formed on the basis of the large experience gained in designing and
operating of Nuclear reactors and in particular fast neutron reactors
(This experience included in itself large sodium leaks leading to radioactive sodium
releases from the primary circuit of the reactor as well as failures of steam generator
tubes causing water and steam penetration to the secondary sodium)
International experience collected in the IAEA recommendations
2 Regulatory documents and reactor operation regulations are
periodically updating on the base Russian and worldwide experience gained in preventing and mitigating of
abnormalities and accidents
Russian and international experience in designing and safety analysis of advanced
projects (GEN-4 INPRO)
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
1 Regulatory Documents for NPP Safety Provision determine the set of design and
safety criteria
2 One of the Main document that determined common regulatory approach and
common requirements to safety analysis of fast reactors in Russia is OPB-8897
It includes brief list of specific requirements regulating characteristics of various
type reactors with regard to safety It also requires that the SAFETY REPORT
should be issued for each reactor SAFETY REPORT should be developed in
accordance with another regulatory document ndash ldquoSpecial standard contents of
safety analysis reportrdquo
3 Large number of computer codes are involved in the process of safety
justification In accordance with Russian regulations those codes should be
certified The procedure of certification includes in itself verification of the code
and its expertise by the team of independent experts (Team leader is normally the
representative of ldquoROSTEHNADZORrdquo) and expertise by Special Certification
Board
4 Finally it should be proved that project characteristics and reactor behavior
under normal and accidental conditions satisfies the set of design and safety
criteria
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
PBYa RU AS-89 Rules on Nuclear Safety of Nuclear Power
Plants Reactor Units
OPB-8897 General Statements on Nuclear Power Plant Safety
Provision
NRB-99 Radiation Safety Codes
OSPORB-99 Basic Sanitary Rules for Radiation Safety
Provision
SP AS-03 Sanitary Rules for Nuclear Power Plant Design and
Operation
NP-032-01 NPP Siting Basic Criteria and Requirements for
Safety Provision
Russian Regulatory Documents for
NPP Safety Provision
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
Major safety issues (after the Fukushima accident)
26
After the Fukushima accident checks were made on Russian nuclear plants Following these in mid June 2011 Rosenergoatom announced a RUR 15 billion ($530 million) safety upgrade program for additional power and water supply back-up Rosenergoatom spent RUR 26 billion on 66 mobile diesel generator sets 35 mobile pumping units and 80 other pumps
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
Measures to improve accident
management preparedness
27
Implementation of additional emergency equipment for
emergency water and power supply at nuclear power plants
Confining system reliability improvement
Implementation of emergency and post-accident sampling
Analysis of feasibility and expediency for implementation of the
reactor pressure vessel outer cooling
Enhancement of main control room and emergency control room
protection
Qualification of safety system components for lsquoharshrsquo
environmental conditions
Improvement of the emergency response interaction system
Development and implementation of Guidelines for severe
accident management
Improvement of personnel competences and preparedness
28
THANK YOU FOR YOURATTENTION
28
THANK YOU FOR YOURATTENTION
![17-04-10 Schedule Articleprjmgrcap.com/images/17-04-10_Schedule_Article.pdf2010/04/17 · í $ *HQHUDOL]HG (3& 6FKHGXOH 7RRO 7 ( :ROI 6XJDU /DQG 7H[DV 3OHDVH QRWH WKLV LV FRS\ULJKWHG](https://static.documents.pub/doc/80x56/5eb54a29889e20707617b830/17-04-10-schedule-20100417-hqhudolhg-3-6fkhgxoh-7rro-7-roi.jpg)
