C. Ader, T. King, A. Rubin, C. Tinkler, R. Wright

The U.S. NRC has invested considerable resources over the past 15 years to develop a qualitative and quantitative understanding of severe accident phenomena and issues for LWRs. This understanding has resulted in the development of analytical models and integrated computer codes (e.g., SCDAP/RELAP, CONTAIN, MELCOR) and data sufficient to support plant risk assessments, resolve certain severe accident issues for operating and future plants and identify areas requiring further research and modelling improvement. This paper focuses on what has been accomplished and what remains to be done to further our understanding and ability to predict certain severe accident phenomena. It addresses severe accident phenomena and issues from the perspective of how well they can be modelled and predicted and what are the remaining uncertainties. Risk perspectives and future experimental and modelling work are discussed along with the factors considered in deciding where our remaining resources should be applied.


J. Sugitomo
Department of Reactor Safety Research
Japan Atomic Energy Research Institute
Tokai-mura, Ibaraki-ken,Japan 319-11

Severe accident research at JAERI was initiated soon after the TMI-2 accident in 1979 and has been accelerated after the Chernobyl accident in 1986. JAERI's research activities aim at the confirmation of the safety margin, the quantification of the risk, and the evaluation of the effectiveness of the accident management measures of the nuclear power reactors.

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.


V. A. Sidorenko, V. G. Asmolov*
Ministry of Russian Federation on Atomic Energy, Moscow, Russia
* Nuclear Safety Research Center "Kurchatov Institute" Moscow, Russia

Variety of probable scenario of thermal, structural, hydraulic, chemical processes accompanying severe accidents causes large uncertainties at their description by means of calculation models. This is a reason why so serious attention is paid to the research clarification and reduction of this uncertainties. An important issue of the researches conductive in the work in the field of nuclear safety is understanding the due to complexity of problems, their decision not under capacities of a one separate country. Even such countries as USA, France, Germany, Great Britain, Japan, almost all experimental programs related to the safety of nuclear power plants carry out in close cooperation. One of the most important considerations is a large money investment in work. In Russia, before Chernobyl accident, the works on research of safety of nuclear power plants were conducted almost in total isolation from other world and limited by the analysis only design basis limits. The lesson of Chernobyl, weeks and months followed accident, initiation of express-researches simultaneously with the progression of accident have demonstrated to all community lack of necessary base of knowledge on severe accidents subject.

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.



M. L. Corradini, S. Nisowankosit, B. Shanoum, J. Tang
University of Wisconsin-Madison, Madison, Wisconsin, USA

Over the course of the last ten years the TEXAS model has been developed to model the mixing, triggering, propation and expansion processes of a vapor explosion. It was originally developed for analysis of FCI propagation behavior for LMFBR UO2-sodium interactions by Sandia in 1983, and modified by us for LWR applications. The vapor explosion process is modelled in a mechanistic fashion for each phase of the process with the fuel, coolant liquid and coolant vapor treated as separate fluids in a unique and novel Lagrangian-Eulerian framework, under the limitation of one-dimensional fluid mechanics. The current models of TEXAS will be discussed in the paper along with our analysis of the KROTOS experiments compared to thermodynamic analyses and the recent MFCI experiments performed here at the University of Wisconsin with simulant materials.


T. J. Haste*, R. P. Hiles
AEA Technology, Winfrit Technology Centre,United Kingdom
S. Hagen, P. Hofmann
Kemforschungszentrum Karlsruhe GmbH, Germany

The CORA series of electrically heated melt progression bundle experiments performed at KfK Karlsruhe was aimed at identifying and quantifying the mechanisms and sequence of events causing severe fuel damage to light water reactor(LWR) fuel rods during heat-up and flooding. It was supported by a program of separate effect tests to measure the kinetics of chemical reactions identified as important in the integral experiments.

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".

* Lead Author


J. W. DeVaal, J. R. Gauld, M. E. Klein, B. H. McDonald
AECL Chalk River Laboratories
Chalk River, Ontario K0J 1J0
H.-W. Chiang, D. W. Dormuth, M. E. Lavack, D. R. Whitehouse, G. B. Wilkin
AECL Whiteshell Laboratories
Pinawa, Manitoba R0E 1L0

Severe fuel damage experiments are being conducted in the Blowdown Test Facility (BTF) in the NRU reactor at Chalk River Laboratories (CRL) to improve our understanding of postulated accidents in CANDU* reactors. Full analysis of the data obtained from these experiments requires the ability to model reactor neutronics, system thermalhydraulics, fuel performance, fission product release and fission product transport and deposition in the BTF circuit. Best-estimate calculations of these processes require intimate coupling and simultaneous solution of all the equations describing the entire range of physical and chemical phenomena involved. The ELOCA code has been developed at CRL for calculation of CANDU fuel performance under accident conditions and, at Whiteshell Laboratories (WL), the CATHENA code has been developed for analysis of primary circuit thermalhydraulics during these accidents. Also within AECL, the VICTORIA code is used for primary circuit fission product transport analysis, while the GOTHIC code is in use for analyzing containment thermalhydraulics. This paper describes a technology under development at AECL for coupling these codes together, along with a reactor physics kinetics code, for performing integrated analyses of our severe fuel damage experiments. The current technology includes coupling codes to each other in modular routines within a single executable code (i.e., as "subroutine constructs") and by coupling separate executable codes together in a parallel, distributed computing environment (i.e., as "network constructs"). An example of a coupled analysis of a BTF experiment is presented to illustrate the application of the latest coupled code system. This technology, which is being tested in analyzing BTF experiments, will be further developed and ultimately will be available for performing integrated safety analyses of CANDU reactors.
* CANada Deuterium Uranium, Registered Trademark of Atomic Energy of Canada Limited.


B. I. Nigmatulin, V. I. Melikhov, O. I. Melikhov
Research & Engineering Centre of Nuclear Plants Safety
142530 Electrogorsk, Bezymyannaya 6, Moscow Region, Russia
phone: (09643) 31679 fax.: (09643) 30515

This paper is devoted to the presentation of steam explosion investigations implemented in EREC. It is given a description of the VAPEX code developed in EREC to simulate steam explosions under NPPs severe accident conditions. Mathematical models of steam explosions (both premixing and propagation) developed for the VAPEX code are presented. Numerical aspects of the VAPEX code are given. Verification of VAPEX code on the experimental data for premixing and propagation is described. Sample calculations of ex-vessel steam explosions under LWR conditions are given.


M. Kajimoto, N. Tanaka, O. Furukawa, K. Saito, Y. Takechi, M. Hirano

Since the TMI-2 accident, experimental and analytical researches for LWR severe accidents (SAs) have been carried out in Japan by the Japan Atomic Energy Research Institute (JAERI), industries and Nuclear Power Engineering Corperation (NUPEC). During 1980s, significant researches on SAs were performed by the organizations as a part of developing PSA methodology. In 1994, the industries set fourth PSA-based individual plant examination (IPE) programs, and have selected viable measures of accident managements (AMs). this issue together with supportive R&Ds are being discussed by the organizations. The present paper describes Japanese activities and progress of the SA analyses.

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.


H. G. Allein1, C. Renault2
1 GRS, Koln, Germany
2 IPSN, Cadarache, France

Abstract not available.


E. Uspuras
Ignalina Safety Analysis Group
Lithuanian Energy Institute
3 Breslaujos, 3035 Kaunas, Lithuania

The Ignalina nuclear power plant is a twin-unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. Several important design futures of RBMK-1500 are unique and extremely complex with respect to western reactors. The design full thermal power level for each reactor was 4800 MW, however, actual full thermal power level is 4200 MW. A safety assessment of RBMK-1500 reactors was performed by Russian design institute using their own home-made codes only for 4800 MW power level. A safety reassessment of the Ignalina NPP using state-of-art techniques to determine priorities for safety enhancement is necessary. A program of data verification and analysis is presently being conducted by Ignalina Safety Analysis Group. The state-of-art codes,such as RELAP5/MOD3 (USA) and ATHLET (Germany), were originally designed for Pressurized Water Reactors. Because of unique RBMK designs, the application of these codes to RBMK-1500 encountered several problems. The paper deals with the development of suitable RELAP5 and ATHLET nodalization schemes for Ignalina nuclear power plant and investigation of loss-of-coolant accidents as well as safety-related operational transients. A successful best estimate RELAP5 and ATHLET models of the Ignalina NPP has been developed. These models include the reactor main circulation circuit, reactor control systems and plant safety systems required for transient and accident analysis. Calculations performed with those models compare favorable with plant data.


A. Hidaka1, M. Igarashi1, K. Hashimoto1, H. Sato1, T. Yoshino2, J. Sugimoto1
1 Severe Accident Research Laboratory
Department of Reactor Safety Research
Japan Atomic Energy Research Institute
Tokai-mura, Naka-gun, Ibaraki-ken, 319-11, Japan
Telephone 81-292-82-6778, Facsimile 81-292-82-5570
2 Toshiba Advanced System Corporation
1-2-4 Isago Kawasaki-ku, Kawasaki-shi,
Kanagawa-ken, 210, Japan
Telephone 81-44-246-7070, Facsmile 81-44-246-7078

The WAVE (Wide range Aerosol model VErification) experiments have been performed at JAERI to investigate the CsI deposition onto the inner surface of the pipe wall under typical severe accident conditions. It was shown that relatively large amount of CsI was deposited at the upstream of the pipe and that larger amount of CsI was deposited on the ceiling than the floor at the downstream. Analyses of the experiments have also been conducted with the three-dimensional thermohydraulic code, SPRAC, and the radionuclide transport analysis code, ART. The experimental results were well reproduced with ART by using peripherally subdivided pipe cross section and associated representative thermohydraulic information from SPRAC prediction. It was clarified through the present experiment and analyses that major deposition mechanisms for the chemical form of CsI are thermophoresis and condensation. Accordingly, the coupling of the FP behavior and the detailed thermohydraulic analyses was found to be essential in order to accurately predict the CsI deposition in the pipe, to which little attention has been paid in the previous studies.


V. F. Strizhov
Nuclear Safety Institute, Russian Academy of Sciences
Moscow, Russia

This paper deals with the models being developed for the OECD RASPLAV project on the problem lower head behavior and failure. The main purposes of the modeling efforts are as follows: 1) To produce models for integral experimental facilities, 2) To apply models to optimize experimental design, 3) To analyze experimental results and refine models and 4) To validate integral models able to extrapolate experimental results to plant scale.

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.


J. C. Micaelli & J. M. Seiler (DRN/DTP-Grenoble); H. Bung, C. Maunier, J. M. Humbert (DRN/DMT-Saclay); F. Valin (DTA-Saclay); G. Cognet, A. Forestier, I. Szabo, J. P. Van Dorsselaere (DRN/DER-Cadarache); Y. Philipponneau (DEC-Cadarache)
GAREC: CEA-DRN Group for Analysis of R&D needs related to Core Catcher Systems
Connection: J. M. Seiler
Commissariat à l'Energie Atomique
Centres d'Etudes Nucléaires de Grenoble
Service de Thermohydraulique des Réacteurs
17 rue des Martyrs-38054 GRENOBLE CEDEX 9(France)
Fax: (33)

The analysis is devoted to the core-catcher concepts based on spreading with special emphasis for the EPR (European Pressurized Reactor) concept. The paper will present the general strategy of the work which has been derived from the ISTIR methodology initiated by NRC. Scenario analyses reveal that the corium may drain from the vessel in several steps under different forms (liquid jets and solid debris) with variable compositions and temperatures depending on accident sequences and time. Corium accumulations due to remelting of debris or overlapping of several flows are highly credible. Therefore the demonstration strategy is more directed to prove that accumulations will be limited and have consequences which are compatible with the core-catcher requirements. Real material spreading experiments are necessary to investigate the potential for formation of accumulations and tests with sustain heating are required (VULCANO). Coolability investigations must also be purchased (MACE). The papers will also focus on three specific items which are judged to be of importance: 1)the methodology used for the scaling of the spreading tests, 2) the necessity of coupling Thermalhydraulics, Metallurgy and Physico-Chemistry, and 3) the effect of the mechanical properties of the crust on the equilibrium height of an accumulation of molten corium.


A. V. Jones, I. Shepherd, M. Delaval, J. Sangregorio, S. Treta (JRC)

ESTER is a system of codes developed in various laboratories of the EU and elsewhere assembled in host framework for the purpose of calculating LWR severe accident phenomena both in reactors and in small-scale experiments such as Phebus-FP. Codes being incorporated as ESTER modules need to be changed only minimally into their internal working, but are constrained to receive their input data from and supply their output data to an in-core database managed by the framework. This arrangement practically eliminates unwanted "crosstalk" between modules, and allows all modules access to a set of common facilities including a user interface, an input data reader and checker, printed and graphical output facilities (including an on-line display) and restart and post-processing tools.

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.


C. Renault, A. Maillat
CE Cadarache, France
B. Schwinges
Köln, Germany

ESCADRE is a set of specialized codes each one dealing with a particular aspect of the severe accident phenomenological domain, from core uncovery up to radioactive release out of the containment of LWRs. Recently, these codes have been linked to each other into a self-standing integral code.

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.



Dr. U. Brockmeier
Department of Nuclear and New Energy Systems
Ruhr University Bochum, Germany

The SFD (Severe Fuel Damage) codes ATHLET/CD, ICARE, MELCOR and SCDAP/RELAP5 underlay intensive development and verification activities. At the same time, the experimental database is broadened particularly as regards late phase core melt progression and damage (e.g. the Sandia MP- and the French PHEBUS-FP-experiments). Having access to an extended database, model development focusses on late phase phenomena and on the transition between early and late phase; e.g. ceramic material dissolution, relocation and blockage formation. Further activities focus on quench induced hydrogen production, non fuel eutectics and specifics on eastern type PWR bundle (CORA-W1,-W2). Thermalhydraulics and numerical methods are enhanced with view on new reactor designs and natural convection driven heat removal systems. Concerning early phase damage progression, a comparative assessment of the leading SFD codes shows that the modelling basis in roughly adequate. However, recent OECD-ISP (International Standard Problem) activities still highlight the need for further model refinement to increase code quality towards a reliable tool for design and licensing issues.


L. A. Bolshov, V. F. Strizhov
Nuclear Safety Institute, Russian Academy of Sciences
Moscow, Russia

New demands of acceptance of nuclear power require deterministic evidence of nuclear power plants (NPP) safety. From this point of view the role of deterministic analysis of NPP safety plays a very important role both for existing and future generation of NPPs. Considering current status of existing severe accident codes one may make a conclusion that their capabilities are quite limited and not sufficient to proof NPP safety. This conclusion is based on the experience of usage these codes, analysis of models and experimental database supporting codes and used for their validation. At the same time modern level of development of computer technique and numeric methods allows to use equations based on the first principles rather than correlation. Transition to the physical modeling appears to be more effective to approve in the cases of designing and validation of codes using both separate effect and integral tests and allows to increase predictive power of codes and reduce the range of uncertainties. Moreover physical modeling allows to understand critical points of models and codes, and to plan integral tests to resolve severe accident and management issues.

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


L. A. Bolshov, V. F. Strizhov
Nuclear Safety Institute, Russian Academy of Sciences
Moscow, Russia

Simulation of severe accidents of nuclear reactors is fundamentally difficult because of lack of knowledge on physical processes and coupling of a wide variety of disciplines. What we do at present is to model the processes macroscopically and introduce conservative hypotheses. Mechanistic modeling of the processes and methods to analyze coupled phenomena are needed to overcome these difficulties.

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.

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