JAERI has conducted wide range of severe accident research activities both in experiment and analysis, such as core degradation behavior, fission product behavior both in coolant system and containment, containment integrity and assessment of accident management measures.

For the core degradation behavior JAERI has conducted in-pile experiments using NSRR (Nuclear Safety Research Reactor). For the high temperature behavior of fuel and core materials, core material interaction tests has been performed. TMI debris sample examinations and analysis have been performed at the Hot laboratory. Three-dimensional thermal and inelastic stress analysis of the lower head of the TMI-2 vessel has been completed.

For the fission product behavior, the pool scrubbing experiments at elevated temperature and pressure of TMI-2 accident have been conducted. An overall aerosol behavior in containment has been investigated as part of ALPHA (Assessment of Loads and Performance of Containment in Hypothetical Accidents) Program. Experiment on deposition of aerosols in the piping of 40 mm diameter has been conducted. It was found that three-dimensional flow in the piping is important to predict the deposition behavior. A new project on FP aerosol behaviors in the primary piping named WIND (Wide Range Piping Integrity Demonstration) Program has been initiated. This project aims at the revaporization of aerosol due to decay heating and also the integrity of the piping from this heat source.

Ex-vessel phenomena such as molten core and coolant interaction, molten core and concrete interaction (MCCI), and the leakage through containment penetrations have been investigated in ALPHA Program. Efforts are being made for the evaluation of void fraction measurement in the pre mixing zone before the occurrence of steam explosion for the development of analytical model.

In order to investigate the effectiveness of the mitigative accident management measures such as water addition onto the molten core has been investigated in ALPHA Program.

Code developmental activities are promoted for an integrated source term analysis code, THALES-2, fission product analysis code, ART, and containment aerosol analysis code, REMOVAL. Application of detailed and mechanistic codes, such as SCDAP/RELAP5, CONTAIN, and CORCON, to experimental analyses has made progress by participating international standard problem and code comparison exercises

In the present paper, the recent highlighted results obtained at JAERI will be presented with the emphasize of large uncertainties still remaining in the severe accident phenomenologies.

Unique opportunity to carry out in short terms a enormous complex of work was immediate inclusion in international cooperation, adaptation and application of methodologies and calculational programs developed on West countries for reactors designed in Russia, verification of codes on experimental data, received as during execution of international researches, as within the frame work of realization of the domestic program of experimental work, aimed on research to the specific features of the domestic reactors (differences in geometry of a circuit, constructional materials, equipment, fuel and etc.).

The main objective of the program of necessary scientific researches was related to creation of base of knowledge for the justification of projects of domestic nuclear power plants with light water reactors of a new generation. The level of safety incorporated in these projects should be sufficient to meet the criteria and requirements fixed in international community standards on nuclear power plant safety, in particular from the point of view of severe accident. For realization of this purpose with minimum costs, within the framework of the Russian program the following subjects were scheduled:

• Study and understanding of the methodologies incorporated in calculational programs transferred to Russia by the West countries, their adaptation with reference to the specific Russian nuclear power plant's features and verification within the framework of a created severe accident database;
• Development of domestic models for the description of the separate processes accompanying severe accidents, their introduction in used calculational programs;
• Creation of domestic versions of calculational codes for severe accident simulation;
• Creation of indicated codes in Russian supervision bodies to create the normative status of codes utilization;
• Introduction of certificated codes to the practice of the organizations that involved in design process.

For understanding of key processes, described by foreign codes, possibilities of their adaptation on the account of design of domestic nuclear power plants and the verifications on Russian base of experiments, the following priorities were defined in the experimental program:

• Analysis of processes preceding active core damage;
• Study of physical and chemical processes accompanying release of fission products from the fuel matrix and element, behavior of fission products during heat-up and degradation of an active core;
• Study of main processes, accompanying destruction of fuel elements and fuel assemblies;
• Study of behavior of fragments of active core and corium during interaction with a bottom of a reactor;
• Study of processes at interaction of corium with a concrete of a reactor shaft;
• Study of behavior of a hydrogen at concentrations, close to occurrence of detonation effects;
• Study of behavior containment depressurisation and filtration of the fission products systems potentially suspended in containment atmosphere during severe accident progression.

Report presents the results of R&D researches on the problems listed above.

During the conduct of the test series, which ran from August 1987 to April 1993, AEA provided calculational support at the request of KfK for five of the PWR-related experiments, namely CORA-15, CORA-7, CORA-29, CORA-30 and CORA-10. For CORA-15, recommendations were provided regarding the rod fill pressure to be used in this experiment to study the effect of ballooning on melt progression. For the CORA-7 large bundle test, the SCDAP code was used to advise on the scaling of the power and coolant conditions required in going from the standard 25-rod configuration to the 57-rod bundle employed. For the CORA-29 test which examined the effect of a limited degree of cladding pre-oxidation, SCDAP/RELAP5 calculations were made of the likely effect of the pre-oxidation used. For the CORA-30 low heat-up rate test, and for the CORA-10 test which simulated melt progression in a partially flooded bundle, SCDAP/RELAP5 was used to provide recommendations regarding the power and coolant conditions required to achieve the test objectives.

This report provides an overview of the calculational support provided for each of these five experiments, and comments on the lessons learnt from the experiments and from the performance of the codes involved.

The technical work was funded by the UK Health and Safety Executive, except that for CORA-15 which was funded jointly by the Central Electricity Generating Board and the UK Department of Energy under the Thermal Reactor Agreement. The synthesis of the paper was funded by the Commission of the Eurepean Communities under the Re-inforced Concerted Action "Core Degradation".

Institute of Nuclear Safety (INS/NUPEC) developed and applied the WETBERAN code to the source-term analysis of TMI-2 accident. This analysis pointed out that the diffusion rate of iodine in water was a dominant factor to predict source-term of the TMI-2 accident. This study suggested also the applicability and usefulness of the computer code for predicting source-term of real LWRs. STCP and MELCOR were introduced to INS and widely used to perform PSA for typical domestic LWRs since 1987. In this PSA, about 20 containment failure scenarios for a typical PWR and and about 30 scenarios for a typical BWR were analyzed by MELCOR to find timings of major events such as the reactor vessel and containment failures. Various AM measures and their effectiveness/viability were also discussed by based on MELCOR analyses. JAERI initiated analytical researches on SAs in 1981, and the PSA Research Program has been carried out since 1983. the THALES/ART code that was developed by JAERI, was intensively used to identify important parameters that control effects on accident progression and source-term. About 300 hundred containment failure scenarios for a typical BWR were calculated by THALES/ART. This study concluded that accident progressions and source-term can be categorized into 5 groups in terms of initiating events and available ECCSs. Recently, an updated version of THALES-2 has been completed and used to examine the effects of re-vaporization of radio-nuclides on source-term. The industries introduced the MAAP code in 1983. In the IPE program performed by the industries through 1994, MAAP was used as a primary tool to quantify containment event trees and AM procedures. These results have been reviewed at INS against information obtained with MELCOR.

Validation studies of computer codes have been carried out through experimental analyses such as PHEBUS-FP, Severe Fuel Damage Tests (SNL) and NSPP aerosol experiments (ORNL). besides, a comperative study among THALES/ART, STCP, MAAP was endeavored by a co-operation of JAERI, industries and INS in 1990. This study identified many important factors that may affect SA progression and source-term behavior such as debris/water heat and transfer, debris/concrete interactions, and re-vaporization of radionuclides.

Analytical researches on SAs in Japan have been done with both introduced and originally developed codes. There still remain some differences between the predicted results among the codes due to phenomenological uncertainties such as debris/water heat transfer and debris/concrete interactions, but common and qualitative understandings of SA phenomena have been established and utilization in performing level 2 PSA and discussing AMs by taking account of experimental results.

Proposed experimental facilities are two principle kinds. The first is slice geometry A and the second is hemispherical B geometry. Two possible ways to produce internal heat generation are proposed for A geometry - sidewall heating method and direct electric heating (DEH). For B geometry only DEH method is considered.

In accordance with the objectives of the Project and proposed designs the following models are being developed:

• 2D and 3D natural circulation code for A geometry;
• 2D axisymmetrical model for B geometry with account of electrodynamic terms for consistent account of Lorentz forces and current distribution.

In the paper 3D natural circulation model is presented for slice A geometry. This model allows to analyze effects of sidewalls on the flow pattern both in the case of sidewall heating and volumetric heating. Obtained results were compared with existing correlations. Comparison of different facilities is presented on the base of 3D calculations for slice geometry.

Comparison is the similarity of the physical processes vs. different heating methods and different geometries.

Recently an official version of ESTER has been released incorporating a long list of modules, together with test cases, installation procedures for various machines and extensive documentation. The paper describes the facilities of ESTER, the modules which it contains, and certain powerful combinations of modules which are offered in the code, and then goes on to give a brief survey of ESTER applications, illustrated by sample results. The applications range from reactor sequence calculations (comparison with MELCOR) to Phebus pre and post-test analyses and calculations of separate effects experiments such as FALCON.

The paper concludes with some general remarks on the experience gained with the construction of a large code system using components from several sources.

RALOC is a thermalhydraulic code for the simulation of containment behavior, including multicompartment thermal-hydraulics and hydrogen burning/deflagration. The implementation of version mod2 into ESCADRE is under progress.

ESCADRE and RALOC have been extensively validated against a large set of experiments. A substantial amount of results have been in the field of:

• fission product release and retention in the reactor cooling system,
• containment threat due to hydrogen burning,
• fission product depletion in the containment,
• containment melt-through due to molten core concrete interaction,
• contribution of volatile iodine species to source term.

This paper is a review of the most significant outcomes from this validation work with regard to the relevance to the evaluation of source term and containment integrity for LWRs.

Some examples to illustrate mentioned above problems are presented and discussed in this paper.

The supersimulator is a concept of innovative simulators capable of analyzing scenarios from normal operation to severe accidents incorporating highly advanced methods of physical and multidisciplinary modeling. We have made a feasibility study of the supersimulator. In this paper we discuss the plan, methods, status and issues of the supersimulator project.