J. Eyink*, T. Froehmel**, H. Loeffler*** *Framatome-ANP GmbH, Erlangen, Germany

**Bundesamt für Strahlenschutz (BfS), Salzgitter, Germany***Gesellschaft für Reaktorsicherheit mbH (GRS), Koeln, Germany

OECD/NEA/CSNI-WGRisk “International Workshop on Level 2 PSA and Severe Accident Management”

Cologne, Germany, March 29-31, 2004

OECD Workshop on Level 2 PSA and SAM

A PROPOSAL TO ASSESS CONDITIONAL PROBABILITY RANGES FOR PLANT INTERNAL PHENOMENA DURING

CORE MELT SCENARIOS FOR GERMAN LWR

Content

Overview on regulatory Guidance for Level 2 PSA in Germany

Methods for Quantification of Branching Probabilities in APET for German LWR

Recommendations and Examples

GuideBasics of the

Periodic Safety Review

GuideSafety Status

Analysis

GuideProbabilistic

Safety Analysis

GuideAnalysisof

Physical Protection

Protection goal oriented structure of nuclear regulations - Overview of fundamental requirements, 12/96

Methods for proba-bilistic safety analysis for nuclear power plants, 12/96

Data for quantification of event sequence diagrams and event trees, 4/97

Task force"PSR" of the Federal Com-mittee for NuclearEnergy, chaired byBMUpublished: 1997> revision processdraft: 2003

expert groupschaired by BfS

revision processdraft: 2004

new:• Level 2 PSA, technical

details on methods and data

Guidance for Level 2 PSA in Germany

Main objectives of the guidance are:

to support a systematical assessment of branching probabilities for severe accident progression event tree analysis for German LWR and

broadening the information and database for this analysis step of Level 2 PSA

to specify for which branching probabilities generic, plant-typ specific or plant specific numbers need to be used

to reduce the potential of controversial expert views on complex and not well known severe accident phenomena which might be difficult to resolve in the frame of Periodic Safety Review-process

APET Branching Probabilities for German LWR

selected phenomena: Depressurization of the RCS Arrest of core degradation in-vessel Molten Core-Water-Interaction (in-vessel steam explosion) Hydrogen combustion Loss of RPV integrity under high pressure Coolability of core debris (ex-vessel) Arrest of core-concrete-interaction Pressurisation of the containment

description: every phenomenon is described in a qualitative way addressing the key physical

and chemical features. available methods to treat the problem are presented methods are illustrated by examples based on available plant specific PSA general recommendations and - if possible - quantitative values are issued on how

to deal with the problem in PSAs

Approach

Phenomenon orProcess

Recommendation Rationale

Depressurisation of theRCS

Evaluation of the available timefor the operator actionEvaluation of the success takinginto account previous operatorerrors

This analysis is necessaryfor each PDS.Also for similar plants adifferent mix of scenarioswith different timeprogression may contributeto each PDS

Calculations of RCS-pressure progression shall be performed with / without depressurisation for a number of scenarios and results shall be plant-specific

Determination of the period of time is important to get the points of time for available information and latest initiation of active primary bleed in order to avoid RPV failure at high pressure

Determination of probability of successful operator action depending on time and various stress-situations by using e.g. SWAIN-data

Depressurisation of the RCS

Depressurisation of the RCS

Phenomenonor Process

Recommendation Rationale

Failure ofcomponents ofthe RCS (e.g.hot leg, SGT) incase of highpressure in theRCS

Basis should be plant specific calculations onsurface temperature history. Creep failure canbe either calculated with the thermohydrauliccode, if possible, or general correlation can beapplied considering the proper material.For SGT at these conditions:Tube-temperature: 700 – 800 °CRCS-pressure: 80 barfollowing generic values can be applied5% fractile: 0.0150% fractile: 0.0595% fractile: 0.5The distribution of failure probability considerspre existing mechanical damages as well asintact tubes.

The effort devotedto modelling RCSfailure should becommensurable toits importance

intervalls of temperature and pressure should be used broad enough in order to compensate considerable unscertainties of models and accident conditions

Arrest of core degradation in-vessel

Phenomenon or Process Recommendation RationaleArrest of core degradationin-vessel

In case of PWR a genericcorrelation is given:if core degradation is lessthan 20% and coolabilitycan stabilised by waterinjection RPV integrity canbe assumed.

TMI as example;Similar correlations forBWR are to be established.The large degree ofuncertainty does not justifyplant specific investigation

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Ref.: Löffler, H. et al.: Untersuchung auslegungsüberschreitender Anlagenzustände mittels Ereignisbaumtechnik am Beispiel einer Konvoi-Anlage. BMU-2002-594, November 2002, ISSN 0724-3316

reliability data resulting from level 1 degree of core degradation is the ratio of

fuel which lost pellet geometry and the entire fuel mass

determination of likelihood of coolability depending on degree of core degradation by deterministical analyses:- of degradation gradient (dep. on scenario)- of starting point of water injection- of keeping the partially destroyed core in configuration

Molten-Core-Water-Interaction Consequence

determination of debris mass (incl. thermal energy) which can interact with water in the RPV-bottom head

determination of mechanical energy resulting from explosions by using conversion factors for calculated thermal energy,

comparison of mechanical energy and RPV failure limits (energy transmission by accelerated compact water plug)

Quantification of containment failure probability by comparison of potential explosion loads and the RPV failure limits and using experimental findings:

Phenomenonor Process

Recommendation Rationale

In vesselsteamexplosion

For the probability alpha mode failureof the containment the followinggeneric values can be applied for thePWR:5% fractile: 10-5

50% fractile: 10-4

95% fractile: 10-3

A method is presented that can beapplied to BWR.

The numbers are in line withinternational assessments(e.g. SERG2) and newexperimental findings(BERDA, ECO)The low probabilities and thelarge uncertainties justifygeneric numbers.

RPV failure and Containment failure modes

Relevant factors of RPV failure mode (PWR/BWR): mode of bottom head failure at increased pressure with large or small leak wetted or unwetted bottom head

Phenomenonor Process

Recommendation Rationale

Failure of theRPV underhigh pressure(> 80 bar)

a) leak size PWR5% percentile: 0.04 m2

50% percentile: 0.3 m2

95% percentile: 3.6 m2

b) leak size BWR5% percentile: 50 cm2

50% percentile: 0.05 m2 (10 penetrations)95% percentile: 3.6 m2

c) containment rupture by missiles (1 m2)large leak (>3.6 m2): 90%small leak (< 3.6 m2): 10%

d) containment rupture due to DCH (1 m2)plant typ specific calculation required taking intoaccount Zr oxidation and hydrogen combustion.

a) Sandia LHFexperiments asbasis

d) Simplifiedmethodsavailable in theliterature

Hydrogen combustion

Phenomenonor Process

Recommendation Rationale

Hydrogencombustion

Plant specific calculations are required inorder to get: The range of hydrogen production (in

total and for each PDS) The likelihood of ignition as function of

time taking into account plant specificsof the electrical installations and therecombiners

Plant specific calculations of thepressure distributions for each PDS incase of combustion

Plant type specific analysis of thestructural behaviour of thecontainment (failure probability andmode as function of the pressure)

The large contribution ofcontainment failure due tohydrogen combustion to therisk justifies a partly plantspecific approach;validated tools for suchcalculations are available

Core-concrete-interaction

Core-concrete-interaction in dry cavity

It has to be assumed, that core material – without water covering – completely melts through the basemat and that released gaseous components increase the containment pressure (PWR, BWR Typ 72). In case of one BWR-type (Typ 69) fire hazards shall be considered in adjacent buildings (e.g. reactor building). Important parameters (erosion rate, gas release rate) relevant to containment integrity are to be estimated.

A ground-level leakage of about 1 sqm should be postulated. The resulting fission product release rate strongly depends on successful prior containment venting.

Core-concrete-interaction in cavity filled with water

In case that the cavity is water flooded, a continous coolability can be realised and the erosion process can be stopped if the so called EPRI-criterion for long time debris coolability is fulfilled. An uncertainty distribution is suggested around this criterion.

In case of water flooded cavity without long time coolability, erosion rate and gas release should be handled in the same way as in the case with dry cavity, taking into account a steam source relevant to containment pressure.

Core-concrete-interaction

Phenomenonor Process

Recommendation Rationale

Coolability ofcore debris andarrest of coreconcreteinteraction

Dependent on the particle size distribution:5% percentile: 0.05 MW/m2

50% percentile: 0.2 MW/m2

95% percentile: 1 MW/m2

A large set ofexperiments areavailable in theliterature

Pressurisation of Containment

Phenomenonor Process

Recommendation Rationale

Pressurisation ofthe containment

Plant specific calculations necessary to assessthe time period available for counter measures,e.g. venting

pressure build-up is caused by - heat rate (residual heat generation is unknown),- ratio of steam production and condensation (heat sinks) as well as- produced non-condensable gases (MCCI)

containment failure pressure is type specific

Conclusions

The extension of probabilistic analyses in the frame of PSR regarding severe accident sequences with core melt as well as all requirements regarding extent, methods and procedural steps are oriented on basic international recommendations on procedures for conducting level 2 PSA, published e.g. in /IAEA 95/ or /NEA 97/.

Detailed methodological requirements based on plant-specific German experiences and current knowledge about core melt phenomena according to the state of the art complete the level 2-part of the guidance.

A systematical approach is recommended to derive reasonable conditional probability ranges for these phenomena and processes.

Focus is on those phenomena that are of importance with respect to their consequences and to which a large uncertainty is associated. As far as possible quantitative values are introduced.