Overview of Level 2 PSA by xQ826l5g

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									IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making




       “Overview of Level 2 PSA”




                            Lecturer
                          Lesson IV 3_3

           Workshop Information
             IAEA Workshop XX City , Month, Year
                                - XX
                                     Country
Levels of Risk Analysis

                                        The assessment of plant failures leading
LEVEL 1 PSA                             to core damage and the determination of
                                        core damage frequency (CDF).
                                        The assessment of containment response
                                        leading, together with the results of Level
LEVEL 2 PSA
                                        1 analysis, to the determination of release
                                        magnitudes and frequencies.
                                        The assessment of off-site consequences
                                        leading, together with the results of Level
LEVEL 3 PSA
                                        2 analysis, to estimates of risk to the
                                        public.


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     Level 2 Analysis Task
     –        Level 1/Level 2 Interface
              (Plant Damage State Grouping)
     –        Containment Response Analysis
              (Containment Strength)
     –        Containment Accident Progression
              (Containment Event Trees)
     –        Source Term Analysis



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                                  PSA Framework




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     Level 1/Level 2 Interface
     Core Damage sequences identified by Level 1 are
     grouped with respect to probable containment responses
     to
                        Plant Damage State Groups (PDS)
     Each Plant Damage State (PDS) is the entry point to a
                           Containment Event Tree (CET)
     The PDS grouping criteria can be best displayed in
     sorting tree diagram (PDS Logic Diagram).



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       PSA Level 1/Level 2 Interface Diagram
        LEVEL 1                                     LEVEL 2
          ETs           CD/PDS                           CETs             STC

IE1                                         PDS 1                             1
                               1
 .                                          PDS 2                             2
                               3
                                                                              3
IE2                            7               .
                                                                              .
                                                                                      STC frequency
                            11
 .                                          PDS 7                             .

 .
                               7
                                               .                              .       Source Terms
                               5
                                                                              .
 .                             4               .
                                                                              X
                            23
IE X                                        PDS X



      PDS DIAGRAM                                    STC DIAGRAM
                           1                                                      1
                           2                                                      2
                           3                                                      3
                           .                                                      .
                           7                                                      7
                           .                                                      .
                           23                                                     X


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Example of the PDS Grouping Criteria




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    Containment Overpressure Capacity
      –     A probabilistic evaluation of the containment ultimate
            pressure capacity using finite element modeling
      –     The potential failure modes examined for e.g. VVER
            containment are:
                             • Membrane failures of the containment shell
                             • Failure at the containment wall - basemat junction
                             • Failure of the containment wall - upper ring
                               junction
                             • Failure of the dome - upper ring junction
                             • Failure of the basemat

      –     They were evaluated for three temperatures at the
            inside liner: 150 o, 215 o and 260 o C
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    Severe Accident Phenomena in
    Containment




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     Possible Containment Failure Modes
      The Containment Building failure modes are usually derived
      from NUREG-1335 list of potential containment failure modes
      and mechanisms. These are:
               •    Direct Bypass (ADVs, ISLOCA)
               •    Failure to Isolate
               •    Steam Explosions
               •    Combustion Processes
               •    Hydrogen deflagration and detonation conditions
               •    Direct Containment Heating
               •    Steam Overpressurization
               •    Core-Concrete Interaction (Basemat Melt-through)
               •    Blowdown Forces (Vessel Thrust Force)
               •    Liner Melt-Through
               •    Piping Penetration Melt
               •    Failure of Containment Building Penetrations


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  Containment Event Trees (CETs)
    –    CET delineates the possible accident sequences
    –    CET entry conditions are defined by a PDS
    –    Headings consist of the important "events" which can lead to
         significantly different outcomes (timing and mode of
         containment failure and the atmospheric release of
         radionuclides - the source terms)
    –    Event timing is also a key factor in organizing the events on
         the CET. The time periods considered:
                  •    Prior to Reactor Pressure Vessel (RPV) failure
                  •    At or within a few hours of the time of RPV failure
                  •    Late - many hours after RPV failure


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  CET Construction




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  CET Quantification
   –    The relative probability of each containment end state is
        quantified separately for each Plant Damage State associated
        with the CET.
   –    Each branch is assigned a probability (branch fraction).
   –    The probability assigned to each branch is the analyst's degree-
        of-belief, for a given set of accident conditions, that the
        specific event outcome will occur.
   –    The probabilities are combined for each pathway leading to a
        distinct containment end state.
   –    Each event of the CET has a subordinate tree called a
        decomposition event tree, or DET.
   –    The quantification of a CET event is carried out in the
        Decomposition Event Tree associated to each CET event.

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  Decomposition ET (DET)
   –    Two types of branching in DET event:
          • Sorting event which assigns one branch the value of one and all other
            branches a value of zero depending on the values for PDS attributes
            and prior event decisions in the CET.
          • Split fractions for which a probability is assigned to each of the event
            branches by the analyst.
   –    The sources of "data" for the split fractions include:
                 A. Results of Past Studies
                 B. Temelín Specific Calculations by ÚJV
                 C. Separate Effects Calculations
                 D. Engineering Assessment/Judgment



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  Source Term Analysis
   –    The next data flow step after CET frequency quantification is
        source term binning. In a process analogous to the earlier
        binning of plant damage sequences, the large number of
        containment sequence end points is grouped into a smaller
        number of source term categories (STCs).
   –    The source term categories are defined according to important
        radionuclide release characteristics: timing, energy content,
        magnitude, etc.
   –    As final step in STC binning, the CET frequencies of all
        sequences assigned to the same STC are summed to yield
        source term frequency.


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  Source Term Calculation
   Two basic methods to define the source term magnitude,
   composition, and timing:
   – Deterministic    analysis of representative sequences
        representing STCs with a code such as the STCP,
        MELCOR, CATHARE, etc.
   – By reference to past analysis results such as NUREG-
        1150, IDCOR, IPPSS, other past PRAs, etc.




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   How to evaluate consequence
   severity of individual accidents?
   Accident sequences analyzed in the Level 1 and 2 PSA can be
   reviewed for their significance based on the following criteria:
           sequences with the highest frequency, i.e. most frequent
            sequences
           sequences with the highest impact, i.e. unfrequent
            sequences leading to the most severe source term
           most significant sequences, i.e. with the highest source
            term related to the frequency of occurence

   NOTE: Each Level 1 sequence divides into various STC with certain probability.
         So some accident sequence could make e.g. 1.11% of STC 1, 20.46% of
         STC3, 12.80% of STC5, 25.31% of STC8, 2.21% of STC 9, etc.


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   Examples of Integrated Codes for
   Severe Accidents Analyses
     – STCP (Battelle, Columbus Division, USNRC)
     – MELCOR (Sandia National Laboratory, USNRC)
     – MAAP (Molecular Accident Analysis Program,
       EPRI)
     – THALES/ART (Japan, JAERI)
     OTHER CODES:
     – SCDAP-RELAP5
     – ATHLET-SA
     – CATHARE-ICARE

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    References

   – IAEA Safety Series No. 50-P-8 "Procedures for
     Conducting Probabilistic Safety Assessments in
     Nuclear Power Plants (Level 2)"
   – NUREG-1150-Vol 2 "Severe Accident Risk: An
     Assessment for Five U.S. Nuclear Power Plants"
     (1990)




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