OECD Workshop on Level 2 PSA and SAM

1. OECD Workshop on Level 2 PSA and SAM OECD/NEA/CSNI-WGRisk “International Workshop on Level 2 PSA and Severe Accident Management”Cologne, Germany, March 29-31, 2004 A PROPOSAL TO ASSESS CONDITIONAL PROBABILITY RANGES FOR PLANT INTERNAL PHENOMENA DURING CORE MELT SCENARIOS FOR GERMAN LWR J. Eyink*, T. Froehmel**, H. Loeffler*** *Framatome-ANP GmbH, Erlangen, Germany **Bundesamt für Strahlenschutz (BfS), Salzgitter, Germany ***Gesellschaft für Reaktorsicherheit mbH (GRS), Koeln, Germany

2. 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

3. Guidance for Level 2 PSA in Germany Task force "PSR" of the Federal Com- mittee for Nuclear Energy, chaired by BMU published: 1997 > revision process draft: 2003 Guide Basics of the Periodic Safety Review Guide Safety Status Analysis Guide Probabilistic Safety Analysis Guide Analysisof Physical Protection Methods for proba-bilistic safety analysis for nuclear power plants, 12/96 Protection goal oriented structure of nuclear regulations - Overview of fundamental requirements, 12/96 expert groupschaired by BfS revision process draft: 2004 Data for quantification of event sequence diagrams and event trees, 4/97 • new: • Level 2 PSA, technical details on methods and data

4. APET Branching Probabilities for German LWR 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

5. Approach 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

6. Depressurisation of the RCS • 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

7. Depressurisation of the RCS • intervalls of temperature and pressure should be used broad enough in order to compensate considerable unscertainties of models and accident conditions

8. Arrest of core degradation in-vessel • 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 Ref.: Löffler, H. et al.: Untersuchung auslegungsüberschreitender Anlagenzustände mittels Ereignisbaumtechnik am Beispiel einer Konvoi-Anlage. BMU-2002-594, November 2002, ISSN 0724-3316

9. Molten-Core-Water-Interaction Consequence Quantification of containment failure probability by comparison of potential explosion loads and the RPV failure limits and using experimental findings: • 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)

10. 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

11. Hydrogen combustion

12. 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.

13. Core-concrete-interaction

14. Pressurisation of Containment • 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

15. 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.