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

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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
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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

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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
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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)

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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
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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

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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
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





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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 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)

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