Denis Kolchinsky
Project Chief Engineer
Fulfillment of User
requirement UR 1.4 – release
into containment in AES-
2006 design
19-22 November, 2013 INPRO Forum, IAEA, Vienna
State Atomic Energy Corporation ROSATOM
Branch of Joint Stock Company «East-European leading scientific research and design
institute for energy technology»
Saint-Petersburg R&D Institute “Atomenergoproject” (SPbAEP)
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All information and data in this presentation take from design
and analytical documents for NPPs of AES-2006 design:
Leningradskaja NPP-2 (LAES-2) - two units are under
construction near Saint-Petersburg (RF)
Baltic NPP (BtAES) – two units in Kaliningrad Region
(RF)
Belorusskaja NPP – construction has recently started in
Republic of Belarus
And also from open information for Balakovo NPP (with
reactor type V-320).
References
The frequency of a major release of radioactivity into the
containment / confinement of an INS due to internal events
should be reduced. Should a release occur, the consequences
should be mitigated.
User requirement UR1.4
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Meeting to Criterion CR1.4.1: major release into containment
Activity release into containment atmosphere under LOCA accidents is ever
determined by presence of damaged fuel cladding in the core. The following
acceptance criteria are justified in the design:
For design conditions of category 3 - number of damaged fuel rods shall not
exceed 1 % of the total number of fuel rods in the core
For design conditions of category 4 - number of damaged fuel rods shall not
exceed 10 % of the total number of fuel rods in the core.
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In case of design accidents the expected annual irradiation dose of a limited
part of population at the protection zone boundary (NPP site) is:
For category 3 – effective dose below 1 mSv/event
For category 4 - effective dose below 5 mSv/event
The results of accidents analyzes are presented in PSAR,
Chapter 15 for units LAES-2 and BtAES
LOCA accidents are managed by active safety systems ensured reactor
coolant inventory and heat removal.
Meeting to Criterion CR1.4.1: major release into containment (continuation)
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If safety systems are working in design basis mode, the acceptance criteria
are implemented. Binding of volatile radionuclides in containment
atmosphere is ensured by adding of chemicals (alcali) into boric water of containment spray system.
Safety systems configuration has forth-
trains configuration and ensures
functional margin as N+2 (in V-320
design it is N+1 only)
Two main reasons which had helped to significantly decrease Core
Damage Frequency (CDF) for AES-2006 in comparison with V-320
design are:
N+2 safety systems configuration;
Using of Passive Heat Removal System via Steam Generators (SG
PHRS) for mitigation some of safety systems multiply failures
consequences (such as: black-out, emergency feed water failure, high
pressure emergency injection failure at small LOCA, etc).
Meeting to Criterion CR1.4.1: major release into containment (continuation)
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Multiply failures (included common cause failures) in safety systems may
lead to severe core damage. The frequency of core damage is a
quantitative indicator of criteria CR1.4.1
Main functions of Passive Heat Removal System via Steam Generators
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Residual heat removal and reactor shut-down
cooling at full-loss-of-power conditions;
Residual heat removal and reactor shut-down
cooling at full-loss-of-feedwater-supply conditions;
Prevention of radioactive coolant atmospheric
injection through BRU-A or SG safety valves during
accident with coolant leak from primary to
secondary circuit;
Minimization of radioactive coolant release
during accident with coolant leak from primary to
secondary circuit and simultaneously steam pipeline
break in not cutting part out of containment;
Redundancy of active safety systems in the case
of their failure for emergency reactor cool-down
conditions during accidents.
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Probability Safety Characteristics of NPP Designs
Operation Mode
CDF, 1/(r *y)
V-320 (BNPP) AES-91
(TNPP)
AES-2006
(LAES-2)
Power operation 2,55·10-5 1,35·10-6 1,36·10-7
Shutdown 2,33·10-5 3,50·10-7 4,58·10-7
BNPP – Balokovskaja NPP (Ref.: PSA-1)
TNPP – Tianwan NPP in China (Ref.: PSA-1)
LAES-2 – Leningradskaja NPP-2 (Ref.: PSA-1)
Conclusion:
Due to new safety measures in AES-2006 design the CDF had been
decreased to almost times in comparison with AES-91 and to
hundred times in comparison with V-320!
Criteria CR1.4.2: processes
Indicator IN1.4.2: Natural or engineered processes
sufficient for controlling relevant system parameters and
activity levels in containment/confinement.
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Assurance of containment integrity and containment heat removal
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The main aims
Containment
integrity keeping
(at the initial
period of accident
without operator
actions)
Heat removal from
containment at the late
stage of accident
Corium cooling (inside
reactor vessel or in
core catcher)
For mentioned aims reaching the following design basis
are stated:
Steam explosions during interacting between water and
corium is to be excluded;
Direct heating of containment is to be excluded;
Detonation of combustible mixtures is to be excluded;
Formation of non-condensing gases is to be limited
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Assurance of containment integrity and containment heat removal (continuation)
Meeting to these statements is approved in PSAR, Chapter
15.
Activity Localization Means Inside the Containment
- Containment spray system
- Chemical injection systems for iodine binding
- Hydrogen removal system
- Passive heat removal system from the containment (in
case if spray system does not efficient)
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This function is ensured by:
Emergency Spray System
Intends for:
Pressure reduction in the
containment in case of LOCA
Removal of fission products from
the containment atmosphere .
Control of chemical water
composition in the sump-tank by
adding chemicals.
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Passive Heat Removal System from the Containment
Main functions:
Reducing and maintaining pressure inside the
containment within the design limits during BDBA
including severe some with core damage;
Removal to the ultimate heat sink the heat released
into the containment during BDBA including severe
some with core damage;
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Redundancy of
containment spray
system.
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0 10000 20000 30000
Время (с)
100000
200000
300000
400000
500000
600000
Давление (Па)
Давление в контейнментеDu850 cold (2JND+2JNG+2ГЕ)
С работой системы JMN
Без JMN и JMP
C работой JMP
1) LLOCA (DBA) 2) LLOCA without Spray
system, without PHRS/C 3) LLOCA without
Spray system, with PHRS/C
Pressure in the containment
Design basis:
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In DBAs detonation and deflagration is excluded in the Containment.
In BDBAs detonation of hydrogen is excluded and deflagration is
permitted if the localizing safety systems perform their design
functions.
Capacity of Hydrogen removal system is designated as if 1000 kg of
H2 generates in the Containment during 5-7 hours.
Hydrogen safety
Passive catalytic
hydrogen recombiners
Employment of natural
inertization of gas
medium by steam
The Passive catalytic hydrogen recombiners
recombiners
frame
catalytic element
Recombiners arangement in
the containment
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0 2000 4000 6000 8000 10000Время, с
0
0.04
0.08
0.12
0.16
0.2
Объем
ная концентрация
водорода,
%10
0 об
.
H2 concentration during accident
0 2000 4000 6000 8000 10000Время, с
0
100
200
300
400
Масса
водорода,
кг
Mass of H2: 1-generated; 2-recobained;
3-not recombined
Criterion CR1.4.3 (accident management)
Acceptance limit AL1.4.3: Procedures, equipment and training
sufficient to prevent large release outside containment /
confinement and regain control of the facility.
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Means for BDBA management
1 – PHRS/C, 2 – PHRS/SG, 3 – EHRT, 8 – Core catcher, 5- PARs
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BDBA Management Systems Core Catcher [8]
Hydrogen Removal System (passive recombiners) [5]
System of primary circuit overpressure protection and emergency gas removal system [7]
Passive Heat Removal System via Steam Generators [2]
Passive Heat Removal System from Containment [3]
Core Catcher Water Supply System [9]
System of emergency chemical agents supply [4]
Special Measures for Fuel Pool and Emergency Heat Removal Tanks Make-Up
Special electrical train with moving diesel-generator and accumulators;
Special I&C equipment and control panel in the MCR (SAMS).
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The Main Functions of the Core Catcher
Reactor vault protected against corium thermomechanical interaction
Reception and localization of solid and liquid corium components
Heat transfer from corium
Сore subcriticality ensuring
Decreased hydrogen and radionuclides release into the containment
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Core Catcher Main Technical Solutions
dry, water-cooled, crucible-type, located in the sub-reactor space mixture of iron and aluminum oxides, steel and oxide as part of sacrificial materials two-layer CC vessel resistance to thermal stresses
heat shield for protected the upper part of the vessel against thermal radiation
gadolinium oxide as part of sacrificial material to ensure subcriticality of molten core
1
2
3
4
5
1- reactor vessel, 2 –
dry reactor protection,
3- console framework,
4 – service area, 5 –
core catcher vessel
CC does not require periodic service works. CC equipment and the
measurement system testing should be carried out after accidents, which are
accompanied by leakage in the containment.
1
2
3
4
5
1
2
3
5
4
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1 - reactor
2 –core catcher
3 – fuel pool
4 – reactor internals inspection vault
5 – pit-tanks
6 – core catcher flooding pipes (water
supply to the corium surface)
7 –core catcher heat exchanger feeding
pipelines
8 – steam removal (pipes)
Filling and Cooling of Core Catcher
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8
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Basic Codes for Justification of Passive Systems
KORSAR-B1- best-estimate code a fully non-equilibrium two-fluid model of coolant flow;
SOKRAT/V1 - best-estimate code for modeling severe accident conditions;
KUPOL-M - containment code (lumped paramers);
ANSYS SFX, STAR-CD, PGS-TK, LOGOS, FIRECON special CFD codes for containment calculation;
Program unit KORSAR – KUPOL, SOKRAT-KUPOL integrate codes for joint calculations RV & Conteinment;
DANKO, ANSYS - strength calculations of building constructions, equipment
FIRECON, FIRECON- Calculation of loads on hydrogen
combustion,determination of possible combustion modes
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Large-scale test rig “SPOT”
for HRS/SG, CKTI, St.
Petersburg
Test rig “PHRS/C” in OKBM,
Nizhniy Novgorod;
Large-scale test rig “KMS” in
NITI, Sosnovyi Bor;
Experimental validation of Passive Systems
Additional Power Supply Channel from Mobile Diesel Generator
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Constantly switched-on switcher
Switched-on by operator switcher
Information and Measurement System
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The special control panels for severe accident
management (SAM) are located a the Main and
Emergency Control Rooms
In severe accident at ex-vessel stage the
following parameters are monitoring at the
SAMS panel in MCR:
Temperature in core catcher;
Cooling water level around the core catcher
vessel.
Information and Measurement System (continuation)
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Rapid temperature increasing (>
25 oC/s) registered by sensors
in core catcher (or/and their
failure) is the criterion of
reactor vessel brake down and
corium dropping into core
catcher.
BDBA Management Instructions
As a necessary part of design the following documents are developed and
supplied for the NPP owner:
Emergency Operating Procedures (EOPs) – guidance for managing of
design basis accidents and beyond design basis accidents (before core
damage occurs) for core damage prevention.
Severe Accident Management Guidelines (SAMG) - propose a range of
possible mitigating actions and allow make an additional evaluation and
choose alternative actions.
19-22 November, 2013 INPRO Forum, IAEA, Vienna
AES-2006 Design has significantly better probabilistic safety characteristic then V-320 design – passive heat removal safety systems allow to decrease the frequency of core damage;
AES-2006 contains the special measures, equipment and procedures for BDBA (including severe accidents) management missing in V-320 design.
AES-2006 fully meets to INPRO criteria 1.4.1-1.4.3
CONCLUSIONS
19-22 November, 2013 INPRO Forum, IAEA, Vienna
THANK YOU FOR THE ATTENTION !
19-22 November, 2013 INPRO Forum, IAEA, Vienna
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