SESSION 4
COMPUTER MODELLING AND RESEARCH ON HEAT AND MASS TRANSFER IN SEVERE
ACCIDENTS
OVERVIEW OF SEVERE ACCIDENT RESEARCH
U.S. NRC SEVERE ACCIDENT RESEARCH PROGRAM-PROGRESS, STATUS
AND FUTURE DIRECTION OF MODELLING AND SUPPORTING EXPERIMENTAL ACTIVITIES
C. Ader, T. King, A. Rubin, C. Tinkler,
R. Wright
U.S. NRC
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.
SEVERE ACCIDENT RESEARCH ACTIVITIES AT JAERI
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.
SEVERE ACCIDENT RESEARCH IN RUSSIA
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.
DEVELOPMENT AND APPLICATION OF COMPUTER CODES FOR SEVERE
ACCIDENT ANALYSIS
TEXAS-V A MODEL FOR SIMULATION OF THE VAPOR EXPLOSION
PROCESS
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.
AEA CALCULATIONAL SUPPORT FOR THE CORA EARLY PHASE MELT
PROGRESSION EXPERIMENTS
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
PARALLEL COMPUTING APPLIED TO THE MODELLING OF THE
COMPLEX INTERACTING PHENOMENA IN SEVERE FUEL DAMAGE EXPERIMENTS
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.
VAPEX CODE FOR ANALYSIS OF STEAM EXPLOSIONS UNDER SEVERE
ACCIDENTS
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.
ACCIDENT PROGRESSION AND SOURCE-TERM ANALYSES FOR LWR
SEVERE ACCIDENTS
JAPANESE ACTIVITIES AND PROGRESS
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.
DEVELOPMENT OF THE NEW FAST-RUNNING FRENCH/GERMAN
INTEGRAL CODE ASTEC
H. G. Allein1, C. Renault2
1 GRS, Koln, Germany
2 IPSN, Cadarache, France
DEVELOPMENT OF THE RBMK-1500 MODELS USING STATE-OF-ART
CODES
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.
EXPERIMENTAL ANALYSIS OF FP AEROSOL BEHAVIORS IN PRIMARY
PIPING WITH ART CODE DURING SEVERE ACCIDENT
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.
DEVELOPMENT OF MODELS FOR OECD RASPLAV PROJECT
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.
R&D NEEDS RELATED TO THE CORE CATCHER CONCEPT BASED ON
CORIUM SPREADING
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) 76.88.52.51
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.
STATUS OF THE ESTER SEVERE ACCIDENT CODE SYSTEM
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.
ESCADRE MOD0 AND RALOC MOD2 ASSESSMENT
MAJOR FINDINGS AND RELEVANCE TO THE SAFETY OF LWRS
C. Renault, A. Maillat
IPSN/DRS
CE Cadarache, France
B. Schwinges
GRS
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.
STATUS AND FUTURE DEVELOPMENT OF COMPUTER CODES FOR SEVERE
ACCIDENT ANALYSIS
STATUS AND MAIN UNCERTAINTIES IN LEADING SEVERE ACCIDENT
ANALYSIS CODES
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.
SEVERE ACCIDENT CODES STATUS AND FUTURE DEVELOPMENT
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.
THE CONCEPT OF ADVANCED SIMULATORS FOR NUCLEAR REACTOR
SEVERE ACCIDENTS
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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