IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making “Overview of Level 2 PSA” Workshop Information IAEA Workshop City, Country XX - XX Month, Year Lecturer Lesson IV 3_3 Lecturer Lesson IV 3_3

IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making 2 Levels of Risk Analysis LEVEL 1 PSA LEVEL 2 PSA LEVEL 3 PSA The assessment of plant failures leading to core damage and the determination of core damage frequency (CDF). The assessment of containment response leading, together with the results of Level 1 analysis, to the determination of release magnitudes and frequencies. The assessment of off-site consequences leading, together with the results of Level 2 analysis, to estimates of risk to the public.

IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making 3 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

IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making 4 PSA Framework

IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making 5 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).

IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making 6 PSA Level 1/Level 2 Interface Diagram ETs IE1. IE2... IE X CD/PDS PDS DIAGRAM PDS 1 PDS 2. PDS 7. PDS X X STCCETs X STC DIAGRAM STC frequency Source Terms LEVEL 1 LEVEL 2

IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making 7 Example of the PDS Grouping Criteria

IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making 8 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

IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making 9 Severe Accident Phenomena in Containment

IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making 10 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

IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making 11 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

IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making 12 CET Construction

IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making 13 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.

IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making 14 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

IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making 15 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.

IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making 16 Source Term Calculation – – 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. Two basic methods to define the source term magnitude, composition, and timing:

IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making 17 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.

IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making 18 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

IAEA Training Course on Safety Assessment of NPPs to Assist Decision Making 19 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)