1. STATUS OF IRSN LEVEL 2 PSA (PWR 900)
General objectives
Content of the study
Level 1 to Level 2 Interface
Quantification of physical phenomena with uncertainties in APET
A model for containment leakage through containment penetrations
Radioactive releases model
KANT : a quantification software for level 2 PSA
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

2. General objectives A level 2 PSA for French 900 MW PWR
to contribute to reactor safety level assessment,
to estimate the benefits of accident management procedures,
to provide quantitative elements about advantages of any reactor design or operation modifications,
to acquire quantitative knowledge for emergency management teams,
to help in definition of RD programs in the severe accident field
learning from detailed studies are also extended to other French Plants
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

3. Steps
version (1.0) based on IRSN level 1 PSA published in 1990 – power states of reactor
2003 – version (1.1) - revision of power states of reactor
2004 – version updated level 1 PSA – response surfaces method for uncertainties assessment - hydrogen recombiners
2005 – version 2.1 – shutdown states of reactor
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

4. Content General methodology initially based on NUREG 1150
Binning of level 1 PSA sequences in PDS
Representation of important severe accident events in an APET
Binning of level 2 PSA into Release Categories
Assessment of radioactive releases for each release category
Uncertainties assessment by Monte-Carlo method
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

5. A detailed interface between level 1 to level 2 PSA
20 interfaces variables serve to define the Plant Damage States and concern initiator event, system and containment state, residual power, activation of emergency plan.
PT – RCS break size
SF – Component cooling or essential service water systems
PL – RCS break localization
AP – Water makeup to RCS availability
RT – SGTR number
BA – Safety injection water tank
VL – V-LOCA
SE – Secondary system break
AS – CHRS availability
SO – Pressurizer safety valve availability
BP – Low pressure safety injection availability
IE – Containment isolation
HP – High pressure safety injection availability
CR – Core criticity
GV – SG availability
PR – Residual power
LC – Electrical board availability (low voltage)
PU – Emergency plan
LH – Electrical board availability (high voltage)
RS – Electrical network availability
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

6. A detailed interface between level 1 to level 2 PSA
A high level of description of system states
Examples AS variables values
1 = CHRS available and in service
2 = CHRS available and not in service
3 = CHRS not available, failure occurred at demand
4 = CHRS not available, failure occurred in function – not contaminated
5 = CHRS not available, failure occurred in function – contaminated
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

7. A detailed interface between level 1 to level 2 PSA
150 Plant Damage States have been defined for power states. A representative thermal-hydraulics transient is defined for each PDS
Number of PDS
Number of
thermal-hydraulics transients
LOCA (large break) 17 9
LOCA (medium break) 24 14
LOCA (small break) 8
LOCA (very small break) 10
SGTR 20 15
Secondary break 13
Loss of heat sink
Loss of steam generator water injection
Total loss of electrical power 12 6
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

8. A detailed interface between level 1 to level 2 PSA
Thermal-hydraulics transient are calculated with the SCAR version of the simulator SIPA 2 (that includes CATHARE 2).
Advantages of this approach :
to obtain a better evaluation of accident kinetics and delays before releases,
to consolidate level 1 PSA assumptions,
to define more precise conditions for severe acc. Phenomena,
to provide a large panel of « best-estimated » transients for use in other context (accident management team, safety analysis)
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

9. APET – Quantification of physical phenomena with uncertainties
The different physical phenomena are organized in « physical models » :
each physical model represents a set of physical phenomena that are tightly coupled ;
2 separated models are linked by a limited numbers of variables transmitted by the APET
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

10. Physical models of APET
Level 1 PSA
Plant Damage State
Before Core degradation
During Core degradation
Vessel Rupture
Corium-Concrete Interaction
Before core degradation
I- SGTR
During Core Degradationn
Advanced core degradatio
Combustion H2
In-vessel steam explosion
Direct Containt
Heating
Containment mechanical behavior
Corium concrete interaction
Ex-vessel
s.e.
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

11. Physical models of APET Codes
Construction of physical model based on results obtained by validated codes calculations. Expert’s judgments are used for result interpretation or when direct code calculations are note possible
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

12. Physical models of APET Two methods are employed
METHODE 1 : RESPONSE SURFACES
Downstream variables values = F(upstream variables values)
(Details provided in second workshop presentation)
METHODE 2 : GRID OF RESULTS
For core degradation progression strong scenario effects and discontinuities have to be taken into account (valve opening, RCS cooling by SG, RCS water injection …)
Construction of response surfaces would be a very difficult task
Grid of result approach is used
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

13. Physical models of APET Example of Core degradation
STEP 1 : DEFINITION OF CALCULATIONS
STEP 2 : CONSTITUTION OF A RESULT GRID
Core degradation transient without actions recommended by severe accident management guides
PDS
TH-system transient
Core degradation transient with actions recommended by severe accident management guides
Transient N°
Identification variables values
DCD downstream (results) variables values
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

14. Physical models of APET Example of Core Degradation
STEP 3 : RESULT GRID IN THE APET
ONE SCENARIO DEPENDS ON SYSTEM AVAILIBILITY, HUMAN ACTIONS, RESIDUAL POWER …
A SELECTION TREE SELECTS THE MOST REPRESENTATIVE TRANSIENT IN THE RESULTS GRID
THE DOWNSTREAM VALUES ARE EXTRACTED FROM THE RESULTS GRID FOR THE REPRESENTATIVE TRANSIENT
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

15. Leakage through containment penetrations « b mode »
A specific method has been developed to take into account pre-existing leakage or isolation failure during the accident
A specific software, BETAPROB has been developped
A model is constructed :
System description (hydraulics components, valves, pumps, sumps, rooms of auxiliary building and ventilation/filtration level)
Failure probabilities (l, failure in operation, g, failure on demand)
Severe (100 % section) and non severe (1% section) are distinguished)
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

16. Leakage through containment penetrations APET Model
For each system configuration, BETAPROB calculates all the possible leakage paths and proposes a classification of leakage paths as a function of
Nature of release source (liquid from RCS or gaseous from containment atmosphere)
Transfer mode to environment in function of ventilation systems and filtration
Leakage section
In the APET, for each systems configurations are calculated
Probabilities of leak categories in term of leakage section
Probabilities of leak categories in term of filtration efficiency
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

17. The radioactive releases calculation model
A simplified model has been developed for level 2 PSA.
Each level 2 sequence is characterized by « APF » variables that give information on accident progression and containment failure.
The model can calculate radiaoactive releases as a time function of time for each combination of APF variables.
Uncertainties have been taken into account for most influent parameters.
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

18. The radioactive releases calculation model Fission product emission
Noble
Gases
Volatil
molecular iodine
Progressive Aerosol Emission
Melt - corium
1100 °C
First corium
flow
Vessel
Break
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

19. Fission products behavior in containment
Containment atmosphere composition
Aerosol mass in suspension depends on : emission, energetic phenomena in RCS (steam explosion) or in containment (Combustion), natural deposition, spray system (CSHRS) efficiency and containment leakage
Molecular iodine depends on : emission, painting adsorption, spray system (CSHRS) efficiency and containment leakage
Organic iodine depends on : adsorbed molecular iodine to organic iodine and containment leakage
Noble gases depends on : emission and containment leakage
Radioactive releases depend on
Containment leakage size (mass flow),
Containment atmosphere composition,
Aerosol filtration and iodine retention,
Activity as a function of delay after SCRAM
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

20. The radioactive releases calculation model Graphical interface
A graphical interface allows interactive calculation in function of APF variables values
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

21. KANT A software for level 2 PSA quantification
A specific software, able to take into account the specifities of the IRSN methodologies has been developed.
The software is linked with the releases model
Operational for Windows operating system (C++, MFC, Access)
3 main modules :
APET development (subtrees, specific language for model)
APET quantification (Monte-Carlo method)
Results vizualization
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

22. KANT Example of results vizualization
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

23. KANT Perspectives Future Improvements
Extension of functionalities in terms of results presentation
Identification and quantification of early radioactive releases
Graphical presentation of the APET
A convivial interface to give access to main results
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004

24. Conclusions
A detailed level 2PSA for French 900 MW is performed by IRSN with some specifities
Systematic use of validated codes
Original models (containment leakage, human factor)
Detailed interface and large transient calculation
A specific software, KANT, operational since 1998, with a development program
Future
2004 – Analysis of French Utility approach for level 2 PSA
2004 – Version 2.0 for power states of reactor (recombiner, …)
2005 – Version 2.1 for shutdown states of reactor
2006 ? Improvement of methods (dynamic fiability ?, interface ?),
Other plant application (?)
CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004