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 1

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 2

Steps

•2000 - version (1.0) based on IRSN level 1 PSA published in

1990 – power states of reactor

•2003 – version (1.1) - revision of 1.0 - power states of reactor •2004 – version 2.0 - 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 3

Content

General methodology initially based on NUREG 1150 1. Binning of level 1 PSA sequences in PDS 2. Representation of important severe accident events in an APET 3. Binning of level 2 PSA into Release Categories 4. Assessment of radioactive releases for each release category 5. Uncertainties assessment by Monte-Carlo method CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004 4

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 PL – RCS break localization RT – SGTR number VL – V-LOCA AS – CHRS availability BP – Low pressure safety injection availability HP – High pressure safety injection availability GV – SG availability LC – Electrical board availability (low voltage) LH – Electrical board availability (high voltage) SF – Component cooling or essential service water systems AP – Water makeup to RCS availability BA – Safety injection water tank SE – Secondary system break SO – Pressurizer safety valve availability IE – Containment isolation CR – Core criticity PR – Residual power PU – Emergency plan RS – Electrical network availability CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004 5

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 6

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 LOCA (large break) LOCA (medium break) LOCA (small break) LOCA (very small break) SGTR Secondary break Loss of heat sink Loss of steam generator water injection Total loss of electrical power 17 24 8 10 20 13 13 17 12 Number of thermal-hydraulics transients 9 14 8 10 15 13 10 17 6 CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004 7

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 8

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 9

Physical models of APET

Level 1 PSA Plant Damage State Before Core degradation Before core degradation During Core degradation I- SGTR During Core Degradationn Combustion H2 Advanced core degradatio Vessel Rupture In-vessel steam explosion Ex-vessel s.e.

Direct Containt Heating Containment mechanical behavior CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004 10 Corium-Concrete Interaction Combustion Corium concrete interaction

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

VULCAIN+CPA (ASTEC V0) ICARE-CATHARE + specific mechanical calculations

CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004 11

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 12

Physical models of APET Example of Core degradation

STEP 1 : DEFINITION OF CALCULATIONS 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 STEP 2 : CONSTITUTION OF A RESULT GRID Transient N° Identification variables values DCD downstream (results) variables values CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004 13

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 14

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 15

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 16

The radioactive releases calculation model

A simplified model has been developed for level 2 PSA.

Each level 2 sequence is characterized by failure.

« APF » variables that give information on accident progression and containment 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 17

The radioactive releases calculation model Fission product emission

Noble Gases Volatil molecular iodine Progressive Aerosol Emission 1100 °C Melt - corium CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004 First corium flow Vessel Break 18

Fission products behavior in containment

Containment atmosphere composition

– Aerosol mass in suspension depends on :

explosion) or in containment (Combustion), natural deposition, spray system (CSHRS) efficiency and containment leakage emission, energetic phenomena in RCS (steam

– Molecular iodine depends on

: emission, painting adsorption, spray system (CSHRS) efficiency and containment leakage

– Organic iodine depends on

containment leakage

– Noble gases depends on

: adsorbed molecular iodine to organic iodine and : 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

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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 20

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 21

KANT Example of results vizualization

CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004 22

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 23

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 – – 2004 – – 2005 – – 2006 ?

Analysis of French Utility approach for level 2 PSA Version 2.0 for power states of reactor (recombiner, …) Version 2.1 for shutdown states of reactor Improvement of methods (dynamic fiability ?, interface ?), Other plant application (?) CSNI/WG-RISK – LEVEL 2 PSA AND ACCIDENT MANAGEMENT WORKSHOP – MARCH 2004 24