csni/wg-risk – level 2 psa and accident management workshop – march 20041 status of irsn level 2...

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

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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


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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


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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


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Content

General methodology initially based on NUREG 1150

1. Binning of level 1 PSA sequences in PDS2. Representation of important severe accident

events in an APET3. Binning of level 2 PSA into Release Categories4. Assessment of radioactive releases for each

release category5. Uncertainties assessment by Monte-Carlo method


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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 sizeSF – Component cooling or essential service water systems

PL – RCS break localization AP – Water makeup to RCS availability

RT – SGTR number BA – Safety injection water tankVL – V-LOCA SE – Secondary system breakAS – 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


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A high level of description of system states

Examples AS variables values

1 = CHRS available and in service2 = CHRS available and not in service3 = CHRS not available, failure occurred at

demand4 = CHRS not available, failure occurred in

function – not contaminated5 = CHRS not available, failure occurred in

function – contaminated


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150 Plant Damage States have been defined for power states. A representative thermal-hydraulics transient is defined for each PDS

Number of PDS Number ofthermal-hydraulics

transients

LOCA (large break) 17 9

LOCA (medium break) 24 14

LOCA (small break) 8 8

LOCA (very small break) 10 10

SGTR 20 15

Secondary break 13 13

Loss of heat sink 13 10

Loss of steam generator water injection

17 17

Total loss of electrical power 12 6


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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)


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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


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Physical models of APET

Level 1 PSAPlant Damage State

Before Core degradation

During Core degradation

Vessel Rupture

Corium-Concrete Interaction

Before core degradation

I- SGTR

During Core Degradationn

Advanced core

degradatio

CombustionH2

In-vessel steam

explosion

Direct ContaintHeating

Containment mechanical behavior

Corium concrete

interaction

Combustion

Ex-vessels.e.


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Physical models of APETCodes

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

In-vessel progressionIn-vessel

progressionAdvanced core

degradationAdvanced core

degradation

In-vesselexplosionIn-vesselexplosion

Direct containmentheating


Ex-vesselprogressionEx-vessel

progression

Containment mechanicalbehaviour


VULCAIN+CPA(ASTEC V0)

specific models

MC3D + EUROPLEXUS

RUPUICUV+CPA(ASTEC V0)

CORCON+CPA

CAST3M

Induced breaksInduced breaks

ThermalhydraulicsThermalhydraulics

SCAR

ICARE-CATHARE + specific mechanical calculations

In-vessel progressionIn-vessel

progressionAdvanced core

degradationAdvanced core

degradation

In-vesselexplosionIn-vesselexplosion



Ex-vesselprogressionEx-vessel

progression





specific modelsspecific models

MC3D + EUROPLEXUSMC3D + EUROPLEXUS



CORCON+CPACORCON+CPA

CAST3MCAST3M

Induced breaksInduced breaks

ThermalhydraulicsThermalhydraulics

SCARSCAR




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Physical models of APETTwo 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


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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

TH-system transient

Core degradation transient with actions recommended by severe accident management guides

PDS

Transient N°

Identification variables values

DCD downstream (results) variables values


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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


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Leakage through containment penetrations« 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 (, failure in operation, , failure on demand)

– Severe (100 % section) and non severe (1% section) are distinguished)


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Leakage through containment penetrationsAPET 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


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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.


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Fission product emission

Noble

Gases

Melt - corium 1100 °C

First corium

flow

Vessel Break

Volatil

molecular iodineProgressive Aerosol Emission


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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


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Graphical interface

A graphical interface allows interactive calculation in function of APF variables values


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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


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KANTExample of results vizualization


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KANTPerspectives

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


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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 20041 status of irsn level 2...

Documents

binning of level

csniwgrisk level

updated level

study level

psa slide

psa pwr

high level of description

status of irsn level