Severe Accident Research Research Articles

Modeling of atmosphere stratification in containments of nuclear power plants using lumped-parameter code

One of the current high priority safety issues of nuclear power plants is atmosphere stratification in the containments. It requires extensive experimental and numerical investigations. Numerical investigation of stratification using lumped-parameter codes is complicated due to inherent limitations of the codes. These limitations have to be taken into account in developing nodalisation of the containments. The paper presents M5 experiment simulations using lumped parameter code COCOSYS. Experiment M5 was performed in the MISTRA test facility in the frames of FP-6 project SARNET (Severe Accident Research Network). Obtained results are presented in the context of evaluation of the lumped-parameter code COCOSYS capabilities to model the formation of stratified atmosphere and to give recommendations for the development of nodalisations in similar cases using any lumped-parameter code. The paper describes the MISTRA test facility and the performed experiment M5, the gas flow model of the COCOSYS code, the developed nodalisations, and the obtained results. The performed analysis showed that detailed vertical nodalisation and special treatment of injected gas flow is required for simulation of gas mixing phenomena.

Sustainable integration of EU research in severe accident phenomenology and management

In order to optimise the use of the available means and to constitute sustainable research groups in the European Union, the Severe Accident Research NETwork of Excellence (SARNET) has gathered, between 2004 and 2008, 51 organizations representing most of the actors involved in severe accident (SA) research in Europe plus Canada. This project was co-funded by the European Commission (EC) under the 6th Euratom Framework Programme. Its objective was to resolve the most important pending issues for enhancing, in regard of SA, the safety of existing and future nuclear power plants (NPPs). SARNET tackled the fragmentation that existed between the national R&D programmes, in defining common research programmes and developing common computer codes and methodologies for safety assessment. The Joint Programme of Activities consisted in: – Implementing an advanced communication tool for accessing all project information, fostering exchange of information, and managing documents; – Harmonizing and re-orienting the research programmes, and defining new ones; – Analyzing the experimental results provided by research programmes in order to elaborate a common understanding of relevant phenomena; – Developing the ASTEC code (integral computer code used to predict the NPP behaviour during a postulated SA) by capitalizing in terms of physical models the knowledge produced within SARNET; – Developing scientific databases, in which the results of research experimental programmes are stored in a common format; – Developing a common methodology for probabilistic safety assessment of NPPs; – Developing short courses and writing a text book on severe accidents for students and researchers; – Promoting personnel mobility amongst various European organizations. This paper presents the major achievements after four and a half years of operation of the network, in terms of knowledge gained, of improvements of the ASTEC reference code, of dissemination of results and of integration of the research programmes conducted by the various partners. Most initial objectives were reached but the continuation of the SARNET network, co-funded by EC in the 7th Framework Programme (SARNET2 project that started in April 2009 for 4 years), will consolidate the first assets and focus mainly on the highest priority pending issues as determined during the first period. The objective will be also to make the network evolve towards a complete self-sustainability.

Application and Validation of ASTEC Code in the Analysis of Fission Product Release and Transport Processes Taking Place in the Phébus and the STORM Facilities

This paper presents an overview of the activities carried out in the framework of the SARNET project by the CIEMAT, INR, JRC/IE, GRS, UJV, and VUJE partners involved in the validation of ASTEC on fission product (FP) release and transport experiments simulating severe accident conditions in the reactor circuit and containment.These activities were mainly devoted to the analysis of the Phébus experiments, FPT0, FPT1, and FPT2, which provided fundamental reference data for the severe accident research. The ELSA, SOPHAEROS, CPA, and IODE modules were used for FP release from the bundle, transport in the circuit, containment thermal hydraulics and aerosol behavior, and iodine behavior in containment, respectively. Studies on aerosol behavior in the STORM experiments and iodine behavior in the ThAI experiments are also summarized.The paper describes not only the results of validation of some stand-alone or several coupled code modules but also the results of first integral calculations, when all the relevant modules of the ASTEC code were used to model the FP release and transport. In the integral calculations, no boundary conditions are to be defined by the code users for most of the code modules, but only at such interfaces were the boundary conditions applied in the experiment. The integral calculation allows more objective judgment about the combined uncertainties of the calculated results.Together with overview of the progress in the validation of the main ASTEC modules, this paper also points out what needs to be improved in the modeling of future ASTEC V2 code versions.

ASTEC, COCOSYS, and LIRIC Interpretation of the Iodine Behavior in the Large-Scale THAI Test Iod-9

The large-scale iodine test Iod-9 of the German Thermal hydraulics, Hydrogen, Aerosols, Iodine (THAI) program was jointly interpreted by means of post-test analyses within the THAI Circle of the Severe Accident Research NETwork (SARNET)/Work Package 16. In this test, molecular iodine (I2) was injected into the vessel dome of the 60 m3 THAI vessel to observe the evolution of its distribution between water, gas, and surfaces. The main processes addressed in Iod-9 are (a) the mass transfer of I2 between the gas and the two sumps, (b) the iodine transport in the main sump when it is stratified and then mixed, and (c) the I2 adsorption onto, and desorption from, the vessel walls in the presence and absence of wall condensation. The codes applied by the THAI Circle partners were the Accident Source Term Evaluation Code (ASTEC)-IODE (IRSN, Saint Paul Lez Durance, France), Containment Code System (COCOSYS)-Advanced Iodine Model (AIM) (GRS, Garching, Germany), and Library of Iodine Reactions in Containment (LIRIC; AECL, Chalk River, ON, Canada). ASTEC-IODE and the Advanced Iodine Model (AIM) are semi-empirical iodine models integrated in the lumped-parameter codes ASTEC and COCOSYS, respectively. With both codes multicompartment iodine calculations can be performed. LIRIC is a mechanistic iodine model for single stand-alone calculations. The simulation results are compared with each other and with the experimental measurements. Special issues that were encountered during this work were studied in more details: I2 diffusion in the sump water, I2 reaction with the steel of the vessel wall in gaseous and aqueous phases, and I2 mass transfer from the gas to the sump. Iodine transport and behavior in THAI test Iod-9 are fairly well simulated by ASTEC-IODE, COCOSYS-AIM, and LIRIC in post-test calculations. The measured iodine behavior is well understood and all measured data are found to be consistent. The very slow iodine transport within the stratified main sump was simulated with COCOSYS only, in a qualitative way. Consequently, this work highlighted the need to improve modeling of (a) the wet iodine adsorption and the washdown from the walls, (b) the I2 mass transfer between gas and sump, and (c) the I2/steel reaction in the gaseous and aqueous phases. In any case, the analysis of the large-scale iodine test Iod-9 has been an important validation step for the codes applied.

Uncertainty analysis for fission products transport in CANDU primary heat transport during a severe accident

Abstract The work realized under the Severe Accident Research Network of excellence (SARNET) project has shown that the SOPHAEROS module, part of Accident Source Term Evaluation Code (ASTEC) can be fully used to simulate the fission products transport and deposition phenomena in the CANDU Primary Heat Transport (PHT) system. This paper presents an uncertainty analysis for the fission products transport in the CANDU PHT system during a severe accident to obtain the domains of the output parameters, for this study masses of Caesium, Strontium and Iodine deposited in the PHT system and its nodes, taking into account the associated input parameters uncertainties. Five uncertain parameters, the starting time for the releasing process, the duration of the releasing process, the releasing fractions for Cs, Sr and I have been chosen. To generate aleatory values for the uncertain parameters, a method and software have been developed and Monte Carlo simulations to determine uncertainties propagation through the SOPHAEROS module has been carried out.

Modelling of sprays in containment applications with A CMFD code

During the course of a hypothetical severe accident in a Pressurized Water Reactor (PWR), spray systems are used in the containment in order to prevent overpressure in case of a steam line break, and to enhance the gas mixing in case of the presence of hydrogen. In the frame of the Severe Accident Research Network (SARNET) of the 6th EC Framework Programme, two tests was produced in the TOSQAN facility in order to study the spray behaviour under severe accident conditions: TOSQAN 101 and TOSQAN 113. The TOSQAN facility is a closed cylindrical vessel. The inner spray system is located on the top of the enclosure on the vertical axis. For the TOSQAN 101 case, an initial pressurization in the vessel is performed with superheated steam up to 2.5 bar. Then, steam injection is stopped and spraying starts simultaneously at a given water temperature (around 25 °C) and water mass flow-rate (around 30 g/s). The depressurization transient starts and continues until the equilibrium phase, which corresponds to the stabilization of the average temperature and pressure of the gaseous mixture inside the vessel. The purpose of the TOSQAN 113 cold spray test is to study helium mixing due to spray activation without heat and mass transfers between gas and droplets. We present in this paper the spray modelling implemented in NEPTUNE_CFD, a three-dimensional multi-fluid code developed especially for nuclear reactor applications. A new model dedicated to the droplet evaporation at the wall is also detailed. Keeping in mind the Best Practice Guidelines, 1 1 Best Practice Guidelines for the use of CFD in Nuclear Reactor Safety Applications, NEA/CSNI/R5 (2007). closure laws have been selected to ensure a grid-dependence as weak as possible. For the TOSQAN 113 case, the time evolution of the helium volume fraction calculated shows that the physical approach described in the paper is able to reproduce the mixing of helium by the spray. The prediction of the transient behaviour should be improved by including in the model corrections based on better understanding of the influence of the dispersed phase on the turbulence of the continuous phase. For the TOSQAN 101 case, droplet velocity, steam volume fraction and gas temperature profiles compare favourably with the experimental results. In the frame of the SARNET network, the results obtained with the physical modelling implemented in the NEPTUNE_CFD code reproduce correctly the entrainment phenomena and the condensation zone ( Malet and Métier, 2007).

Ranking of severe accident research priorities

The objectives of the SARNET network are to define common research programmes in the field of severe accidents and to develop common computer tools and methodologies for safety assessment in this field. To reach these objectives, one of the work packages, named “Severe Accident Research Priorities” (SARP), aimed at reviewing and reassessing the priorities of research issues as a basis to harmonize and to re-orient research programmes, to define new ones, and to close – if possible – resolved issues on a common basis. The work was performed in close collaboration with 8 participating institutions, led by GRS, representing technical safety organisations, industry and utilities (IRSN, CEA, EDF, FZK, GRS, KTH, TUS, VTT). This action made use notably of (1) the outcomes of the EURSAFE project in the 5th Framework Programme, i.e. the Phenomena Identification and Ranking Tables (PIRT) on severe accidents, (2) the results of the validation and benchmarking activities on ASTEC, (3) the results of reactor calculations carried out in the other SARNET tasks, and (4) the outcome of the research performed in the three thematic sub-domains of SARNET (corium, containment and source term). The main outcome of EURSAFE was a list of 21 topics which included recommendations for experimental programmes and code developments. This list formed the basis of the work in SARP. Also the methodology applied in EURSAFE to consider both the risk potential and the severe accident issues where large uncertainties still subsist was adopted. The analyses of the progress of research and development activities considered whether (1) any research issue was resolved due to reduction of uncertainties or gain of scientific insights, (2) any new issue had to be added to the list of needed research, (3) any new process or phenomenon had to be included in the general PIRT list taking into account the safety relevance and the lack of knowledge, and (4) any new accident management program has to be developed to cope with unresolved problems. Furthermore a strategy plan was elaborated to ensure a wide consensus with the end-user requirements and the objectives of SARNET research activities.

Coolability of particulate beds in severe accidents: Status and remaining uncertainties

Particulate debris beds may form during different stages of a severe core melt accident; e.g. in the degrading hot core, due to thermal stresses during reflooding, in the lower plenum, by melt flow from the core into water in the lower head, and in the cavity by melt flow out of a failing RPV into a wet cavity. Deep water pools in the cavity are used in Nordic BWRs as an accident management measure aiming at particulate debris formation and coolability. It has been elaborated in the joint work of the European Severe Accident Research Network (SARNET) in Work Package (WP) 11.1 that coolability of particulate debris, reflooding of hot debris as well as boil-off under decay heat (long-term coolability), is strongly favoured by 2D/3D effects in beds with non-homogeneous structure and shape. Especially, water inflow from the sides and via bottom regions strongly improves coolability as compared to 1D situations with top flooding, the latter being in the past the basis of analyses on coolability. Data from experiments included in the SARNET network (DEBRIS at IKE and STYX at VTT) and earlier ones (e.g. POMECO at KTH) have been used to validate key constitutive laws in 2D codes as WABE (IKE) and ICARE/CATHARE (IRSN), especially concerning flow friction and heat transfer. Major questions concern the need of the explicit use of interfacial friction to adequately treat the various flow situations in a unified approach, as well as the adequate characterization of realistic debris composed of irregularly shaped particles of different sizes. Joint work has been supported by transfer of the WABE code to KTH and VTT. Concerning realistic debris, the formation from breakup of melt jets in water is investigated in the DEFOR experiments at KTH. Present results indicate that porosities in the debris might be much higher than previously assumed, which would strongly support attainment of coolability. Calculations have been performed with IKEJET/IKEMIX describing jet breakup, mixing and settling of resulting particles. Models about debris bed formation and porosity are developed at KTH. The codes have been applied to reactor conditions for analysing the potential for coolability in the different phases of a severe accident. Calculations have been performed with WABE (MEWA) implemented in ATHLET-CD and with ICARE/ICATHARE for degraded cores and debris beds in the lower plenum, under reflooding and boil-off. Ex-vessel situations have also been analysed. Strong effects of lateral water inflow and cooling by steam in hot areas have been demonstrated. In support, some typical basic configurations have been analysed, e.g. configurations with downcomers considered as possible AM measures. Melt pool formation or coolability of particulate debris is a major issue concerning melt retention in the core and the lower head. Present conclusions from those analyses for adequate modelling in ASTEC are outlined as well as remaining uncertainties. Experimental and analysis efforts and respective continued joint actions are discussed, which are needed to reach resolution of the coolability issue.

SARNET: Severe accident research network of excellence

Fifty-one organisations network in SARNET (Severe Accident Research NETwork of Excellence) their research capacities in order to resolve the most important pending issues for enhancing, with regard to Severe Accidents (SA), the safety of existing and future Nuclear Power Plants (NPPs). This project, co-funded by the European Commission (EC) under the 6th Framework Programme, has been defined in order to optimise the use of the available means and to constitute sustainable research groups in the European Union. SARNET tackles the fragmentation that may exist between the different national R&D programmes, in defining common research programmes and developing common computer tools and methodologies for safety assessment. SARNET comprises most of the organisations involved in SA research in Europe, plus Canada. To reach these objectives, all the organisations networked in SARNET contributed to a Joint Programme of Activities, which consisted of: • Implementation of an advanced communication tool for accessing all project information, fostering exchange of information, and managing documents; • Harmonization and re-orientation of the research programmes, and definition of new ones; • Analysis of the experimental results provided by research programmes in order to elaborate a common understanding of relevant phenomena; • Development of the ASTEC code (integral computer code used to predict the NPP behaviour during a postulated SA), which capitalizes in terms of physical models the knowledge produced within SARNET; • Development of Scientific Databases in which all the results of research programmes are stored in a common format (DATANET); • Development of a common methodology for Probabilistic Safety Assessment of NPPs; • Development of short courses and writing a textbook on Severe Accidents for students and researchers; • Promotion of personnel mobility amongst various European organisations. This paper presents the major achievements after four and a half years of operation of the network, in terms of knowledge gained, of improvement of the ASTEC reference code, of dissemination of results and of integration of the research programmes conducted by the various partners. After this first period (2004–2008), co-funded by the EC, a further contract SARNET2 with the EC for the next four years started in April 2009 as part of the 7th Framework Programme. During this period, the networking activities will focus mainly on the remaining pending issues as determined during the first period, experimental activities will be directly included in the common work and the network will evolve toward complete self-sustainability. The bases for such an evolution are presented in the last part of the paper.

SARNET integrated European Severe Accident Research—Conclusions in the source term area

The overall aim of the SARNET (Severe Accident Research NETwork), in the EU 6th Framework programme was to integrate in a sustainable manner the research capabilities of fifty-one European organisations from eighteen member states of the European Union (EU) plus the Joint Research Centres, with one Canadian company, to resolve important remaining uncertainties and safety issues concerning existing and future nuclear plant, especially water-cooled reactors, under hypothetical severe accident conditions. It emphasised integrating activities, spreading of excellence (including knowledge transfer) and jointly executed research, with the knowledge gained being encapsulated in the European severe accident modelling code ASTEC. This paper summarises the achievements over the whole project in the Source Term Topic, which dealt with potential radioactive release to the environment, covering release of fission products and structural materials from the core, their transport in the primary circuit, and their behaviour in the containment. The main technical areas covered, as emphasised by the earlier EURSAFE project, were the effect of oxidative conditions on fission product release and transport (especially the behaviour of the highly radiotoxic ruthenium under air ingress conditions), iodine volatility in the primary circuit, control rod aerosol release (Ag–In–Cd) that affects iodine transport, containment by-pass in the case of steam generator tube rupture, aerosol retention in containment cracks, aerosol remobilisation in the circuit, and iodine/ruthenium behaviour in the containment especially concerning the volatile fraction in the atmosphere. The studies also covered performance of new experiments, analysis of existing data, and formulation and improvement of theoretical models. Significant progress was made in each area. Looking to the future, the 7th Framework successor project SARNET2 covers the remaining issues concerning iodine and ruthenium, including practical application of the results. The results outlined here will make a good basis for this continued endeavour.

INVESTIGATIONS ON THE RESOLUTION OF SEVERE ACCIDENT ISSUES FOR KOREAN NUCLEAR POWER PLANTS

Under the government supported long-term nuclear R&D program, the severe accident research program at KAERI is directed to investigate unresolved severe accident issues such as core debris coolability, steam explosions, and hydrogen combustion both experimentally and numerically. Extensive studies have been performed to evaluate the in-vessel retention of core debris through external reactor vessel cooling concept for APR1400 as a severe accident management strategy. Additionally, an improvement of the insulator design outside the vessel was investigated. To address steam explosions, a series of experiments using a prototypic material was performed in the TROI facility. Major parameters such as material composition and void fraction as well as the relevant physics affecting the energetics of steam explosions were investigated. For hydrogen control in Korean nuclear power plants, evaluation of the hydrogen concentration and the possibility of deflagration-to-detonation transition occurrence in the containment using three-dimensional analysis code, GASFLOW, were performed. Finally, the integrated severe accident analysis code, MIDAS, has been developed for domestication based on MELCOR. The data transfer scheme using pointers was restructured with the modules and the derived-type direct variables using FORTRAN90. New models were implemented to extend the capability of MIDAS.

The ASTEC Integral Code for Severe Accident Simulation

For several years the French Institut de Radioprotection et de Sûreté Nucléire (IRSN) and the German Gesellschaft für Anlagen und Reaktorsicherheit (GRS) mbH have been jointly developing a system of calculation codes—the integral Accident Source Term Evaluation Code (ASTEC)—to simulate the complete scenario of a hypothetical severe accident in a nuclear light water reactor, from the initial event until the possible radiological release of fission products out of the containment, i.e., the source term. ASTEC has progressively reached a larger European dimension through projects of the European Commission Framework Programme. In particular, in the frame of the European Severe Accident Research NETwork of Excellence (SARNET), jointly executed research activities were performed with the ultimate objectives of providing physical models for integration into ASTEC and making the code the European reference. This effort will go on in the frame of the SARNET2 next network. The ASTEC models are today at the state of the art, except for reflooding of a degraded core. Many applications have been performed by IRSN for significant safety studies, including the probabilistic safety analysis level 2 on a French pressurized water reactor. The first version V2.0 of the new ASTEC series, released in spring 2009, will allow simulation of the European Pressurized Reactor (EPR) and will include advanced core degradation models. Then, ASTEC will remain the repository of knowledge gained from international research and development. Other long-term objectives are on one hand extension of the scope of application to boiling water reactors and CANada Deuterium Uranium (CANDU) reactors, to accidents in the ITER Fusion facility, and to Very High Temperature Reactor (VHTR) Generation IV reactors, and on the other hand to the use for emergency response tools and for severe accident simulators.

Analyses of severe accident scenarios in RBMK-1500

Abstract Even though research of severe accidents in light water reactors is performed around the world for several decades many questions remain. Research is mostly performed for vessel-type reactors. RBMK is a channel type light water reactor, which differs from the vessel-type reactors in several aspects. These differences impose some specifics in the accident phenomena and processes that occur during severe accidents. Severe accident research for RBMK reactors is taking first steps and very little information is available in the open literature. The existing severe accident analysis codes are developed for vessel-type reactors and their application to the analysis of accidents in RBMK is not straightforward. This paper presents the results of an analysis of large loss-of-coolant accident scenarios with loss of coolant injection to the core of RBMK-1500. The analysis performed considers processes in the reactor core, in the reactor cooling system and in the confinement until the fuel melting started. This paper does not aim to answer all the questions regarding severe accidents in RBMK but rather to start a discussion, identify the expected timing of the key phenomena.

Review of debris bed cooling in the TMI-2 accident

Still today after decades of severe accident research it is not well understood why the molten corium did not attack or even penetrate the lower head vessel wall in the TMI 2 accident. The findings can only be explained by additional assumptions which have been proposed by various authors. This paper is also looking for an explanation by examining the role of the debris cooling in the TMI lower head. The present knowledge of debris cooling is based on small-scale experiments with simulant debris. It is argued that the experiments have been stopped to early and therefore do not reveal the potential of debris bed coolability in case of a corium debris bed. It is also argued that in TMI the debris bed was in a state of turbulence and fluidization such that the coolabilty was much higher than in the small-scale experiments. Basically this paper hypothesizes that the unknown phenomenon in the TMI 2 accident is the strong interaction of the debris particles of a wide range of grain sizes and the strong turbulent motion of the cooling fluid and vapor mixture in conjunction with a “virtual gap” at the vessel wall.

European Validation of the Integral Code ASTEC (EVITA): First experience in validation and plant sequence calculations

The main objective of the European Validation of the In tegral Code ASTEC (EVITA) project is to distribute the severe accident integral code ASTEC to European partners in order to apply the validation strategy issued from the VASA project (4th EC FWP). Partners evaluate the code capability through validation on reference experiments and plant applications accounting for severe accident management measures, and compare results with reference codes. The basis version V0 of ASTEC ( Accident Source Term Evaluation Code)—commonly developed and basically validated by GRS and IRSN—was made available in late 2000 for the EVITA partners on their individual platforms. Users’ training was performed by IRSN and GRS. The code portability on different computers was checked to be correct. A “hot line” assistance was installed continuously available for EVITA code users. The actual version V1 has been released to the EVITA partners end of June 2002. It allows to simulate the front-end phase by two new modules: • for reactor coolant system 2-phase simplified thermal hydraulics (5-equation approach) during both front-end and core degradation phases, • for core degradation, based on structure and main models of ICARE2 (IRSN) reference mechanistic code for core degradation and on other simplified models. Next priorities are clearly identified: code consolidation in order to increase the robustness, extension of all plant applications beyond the vessel lower head failure and coupling with fission product modules, and continuous improvements of users’ tools. As EVITA has very successfully made the first step into the intention to provide end-users (like utilities, vendors and licensing authorities) with a well validated European integral code for the simulation of severe accidents in NPPs, the EVITA partners strongly recommend to continue validation, benchmarking and application of ASTEC. This work will continue in Severe Accident Research Network (SARNET) in the 6th Framework Programme where ASTEC plays a key role as the reference European integral code.

Accomplishments and challenges of the severe accident research

This paper briefly describes the progress of the severe accident research since 1980, in terms of the accomplishments made so far and the challenges that remain. Much has been accomplished: many important safety issues have been resolved and consensus is near on some others. However, some of the previously identified safety issues remain as challenges, while some new ones have arisen due to the shift in focus from containment to vessel integrity. New reactor designs have also created some new challenges. In general, the regulatory demands for new reactor designs are stricter, thereby requiring much greater attention to the safety issues concerned with the containment design of the new large reactors, and to the accident management procedures for mitigating the consequences of a severe accident. We apologize for not providing references to many fine investigations that contributed to the great progress made so far in the severe accident research.

PHEBUS FP: a severe accident research programme for current and advanced light water reactors

The ongoing PHEBUS FP programme is the centrepiece of an international co-operation investigating, through a series of integral in-pile experiments, key-phenomena involved in the progression of a postulated severe accident in a light water reactor (LWR). The dedicated PHEBUS facility offers the capability to study the degradation of real core material, from the early phase of cladding oxidation and hydrogen production up to the late phase of melt progression and molten pool formation. The subsequent release of fission products (FPs) and structural materials is also experimentally studied, including their physicochemical interactions, their transport in the cooling system, and their deposition in the containment. The revolatilisation of iodine due to radiochemical effects in the water of the sump and the amount of low-volatility FPs and transuranium elements reaching the containment are receiving a special interest, as large uncertainties related to their modelling subsist. FPT-0 and FPT-1, the first experiments of the programme, performed in December 1993, and July 1996, respectively, have demonstrated that the PHEBUS FP facility is capable of successfully attaining these objectives. They reached very advanced states of degradation, comparable to what was observed in TMI 2, and generated a wealth of results on FP behaviour. The resulting database has been—and will be—applied to develop and validate the computer codes used to assess the safety of the currently operating plants and to check the efficiency of accident management procedures. They will also support the design of future plants having the capacity to confine core melt-down accidents within their containments.

An integrated structure and scaling methodology for severe accident technical issue resolution: Development of methodology

Scaling has been identified as a particularly important element of the Severe Accident Research Program because of its relevance not only to experimentation, but also to analyses based on code calculations or special models. Recognizing the central importance of severe accident scaling issues, the United States Regulatory Commission implemented a Severe Accident Scaling Methodology (SASM) development program involving a lead laboratory contractor and a Technical Program Group to guide the development and to demonstrate its practicality via a challenging application. The Technical Program Group recognized that the Severe Accident Scaling Methodology was an integral part of a larger structure for technical issue resolution and, therefore, found the need to define and document this larger structure, the Integrated Structure for Technical Issue Resolution (ISTIR). The larger part of the efforts have been devoted to the development and demonstration of the Severe Accident Scaling Methodology, which is Component II of the ISTIR. The ISTIR and the SASM have been tested and demonstrated, by their application to a postulated direct containment heating scenario. The ISTIR objectives and process are summarized in this paper, as is its demonstration associated directly with the SASM. The objectives, processes and demonstration for the SASM are also summarized in the paper. The full body of work is referenced.

Development of the methodology and approaches to validate safety and accident management

The article compares the development of the methodology and approaches to validate the nuclear power plant safety and accident management in Russia and advanced industrial countries. It demonstrates that the development of methods of safety validation is dialectically related to the accumulation of the knowledge base on processes and events during NPP normal operation, transients and emergencies, including severe accidents. The article describes the Russian severe accident research program (1987–1996), the implementation of which allowed Russia to reach the world level of the safety validation efforts, presents future high-priority study areas. Problems related to possible approaches to the methodological accident management development are discussed.

シビアアクシデントに関する熱流動研究の最近の動向

Research on severe accident for light water reactors has been initiated after the accident of Three Mile Island in 1979 and accelerated all over the world after the Chernobyl accident in 1986. For the severe accident management Nuclear Safety Commission of Japan recommended electric utilities to investigate the possible measures in 1992. In response to this they proposed accident management strategies based on their investigations in 1994 to be voluntarily implemented in several years by reflecting findings from severe accident research. Since the severe accident phenomena largely involve the thermal-hydraulic aspects, it is of great importance to evaluate the thermal-hydraulic behaviors of severe accident in order to reduce the associated uncertainties. In the present article thermal-hydraulic behaviors during severe accident are described in terms of phenomenology, severe accident management, key issues, research methodology and remaining research items. The main phenomena include core melt progression, in-vessel core melt behaviors, molten core coolant interaction, molten core concrete interaction, direct containment heating, hydrogen behaviors, and fission product behaviors.