STATUS OF IRSN LEVEL 2 PSA (PWR 900) - PowerPoint PPT Presentation

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STATUS OF IRSN LEVEL 2 PSA (PWR 900)

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Title: WG risk workshop level 2 PSA -IRSN Nb 1 Author: IRSN/DSR/SAGR/BEPAG Last modified by: RAIMOND Created Date: 10/21/2003 3:03:21 PM Document presentation format – PowerPoint PPT presentation

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Title: STATUS OF IRSN LEVEL 2 PSA (PWR 900)


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

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

3
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

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

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
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

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 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
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)

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

10
Physical models of APET
11
Physical models of APETCodes
  • Construction of physical model based on results
    obtained by validated codes calculations.
    Experts judgments are used for result
    interpretation or when direct code calculations
    are note possible

12
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

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
TH-system transient
PDS
Core degradation transient with actions
recommended by severe accident management guides
Transient N Identification variables values Identification variables values Identification variables values Identification variables values Identification variables values DCD downstream (results) variables values DCD downstream (results) variables values DCD downstream (results) variables values DCD downstream (results) variables values DCD downstream (results) variables values DCD downstream (results) variables values DCD downstream (results) variables values





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

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)

16
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

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.

18
The radioactive releases calculation
modelFission product emission
Noble Gases
Volatil molecular iodine
Progressive Aerosol Emission
Melt - corium
First corium flow
Vessel Break
1100 C
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

20
The radioactive releases calculation
modelGraphical interface
  • A graphical interface allows interactive
    calculation in function of APF variables values

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

22
KANTExample of results vizualization
23
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

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 (?)
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