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Title: The Integral PWR SIR Transients: Comparisons Between CATHARE and RELAP Codes

Abstract

Within the framework of the research program on innovative light water reactors, the SERI (Service of Studies on Innovative Reactors) of the French Atomic Energy Commission (CEA), is presenting a predictive study on the modeling of a low-power integral Pressurized Water Reactor, using the CATHARE thermalhydraulic code. The concept selected for this study is that of the SIR reactor project, developed by AEA-T and ABB consortium. This very interesting concept is no doubt that which is the most complete to this date, and on which most information in the literature can be obtained. Many safety calculations made with the RELAP code are also available and represent a highly interesting base for comparison purposes, in order to improve the approach on the results obtained with CATHARE. A comparison of the behavior of the two codes is thus presented in this article. This study therefore shows that CATHARE finely models this type of new PWR concept. The transients studied cover a large area, ranging from natural circulation to loss of primary coolant accidents. The ATWS and a power transient have also been calculated. The comparison made between the CATHARE and RELAP results shows a very good agreement between the two codes, andmore » leads to a very positive conclusion on the pertinence of simulating an integral PWR. Moreover, even though this study is a thorough investigation on the subject, it confirms the potentially safe nature of the SIR reactor. (author)« less

Authors:
 [1]
  1. CEA Cadarache, RD 952, 13108 Saint Paul lez Durance Cedex (France)
Publication Date:
Research Org.:
The ASME Foundation, Inc., Three Park Avenue, New York, NY 10016-5990 (United States)
OSTI Identifier:
21064575
Resource Type:
Conference
Resource Relation:
Conference: ICONE-10: 10. international conference on nuclear engineering, Arlington - Virginia (United States), 14-18 Apr 2002; Other Information: Country of input: France
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; COMPARATIVE EVALUATIONS; INFORMATION; NATURAL CONVECTION; PRIMARY COOLANT CIRCUITS; PWR TYPE REACTORS; REACTOR SAFETY; RESEARCH PROGRAMS; SIMULATION

Citation Formats

Pignatel, Jean-Francois. The Integral PWR SIR Transients: Comparisons Between CATHARE and RELAP Codes. United States: N. p., 2002. Web.
Pignatel, Jean-Francois. The Integral PWR SIR Transients: Comparisons Between CATHARE and RELAP Codes. United States.
Pignatel, Jean-Francois. 2002. "The Integral PWR SIR Transients: Comparisons Between CATHARE and RELAP Codes". United States. doi:.
@article{osti_21064575,
title = {The Integral PWR SIR Transients: Comparisons Between CATHARE and RELAP Codes},
author = {Pignatel, Jean-Francois},
abstractNote = {Within the framework of the research program on innovative light water reactors, the SERI (Service of Studies on Innovative Reactors) of the French Atomic Energy Commission (CEA), is presenting a predictive study on the modeling of a low-power integral Pressurized Water Reactor, using the CATHARE thermalhydraulic code. The concept selected for this study is that of the SIR reactor project, developed by AEA-T and ABB consortium. This very interesting concept is no doubt that which is the most complete to this date, and on which most information in the literature can be obtained. Many safety calculations made with the RELAP code are also available and represent a highly interesting base for comparison purposes, in order to improve the approach on the results obtained with CATHARE. A comparison of the behavior of the two codes is thus presented in this article. This study therefore shows that CATHARE finely models this type of new PWR concept. The transients studied cover a large area, ranging from natural circulation to loss of primary coolant accidents. The ATWS and a power transient have also been calculated. The comparison made between the CATHARE and RELAP results shows a very good agreement between the two codes, and leads to a very positive conclusion on the pertinence of simulating an integral PWR. Moreover, even though this study is a thorough investigation on the subject, it confirms the potentially safe nature of the SIR reactor. (author)},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2002,
month = 7
}

Conference:
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  • In order to describe with the same code the whole sequence of severe LWR accidents, up to the vessel failure, the Institute of Protection and Nuclear Safety has performed a coupling of the severe accident code ICARE2 to the thermalhydraulics code CATHARE2. The resulting code, ICARE/CATHARE, is designed to be as pertinent as possible in all the phases of the accident. This paper is mainly devoted to the description of the ICARE2-CATHARE2 coupling.
  • RELAP/MOD1.5 (Cycle 31 and 34) calculations were made to assess the assumptions used by Westinghouse (W) to analyze mainsteam line break transients. Models of a W 3-loop and 4-loop nuclear steam supply system were used. Sensitivity studies were performed to determine the effect of the availability of offsite power, break size and initial core power. Comparison with W results indicated that if the assumptions used by W are replicated within the RELAP5 framework, then the W methodology for prediction of the Nuclear Steam Supply System (NSSS) response is conservative for steam line break transients.
  • Regardless low probability of occurrence the severe accident phenomena are investigated for all types of nuclear reactors in the world because the consequences of such accident could be catastrophic. Most of research is performed for the prevailing vessel-type light water reactors like PWRs and BWRs. Less research is performed for the channel-type reactors like CANDUs and RBMKs as they are operated just in a few countries. Up to now the phenomena that could occur in case of a severe accident in RBMK reactors were not analysed in detail and little literature is available on this topic. The paper presents onemore » of the first integrated analyses of severe accident in RBMK-1500 reactor. RELAP/SCDAPSIM code is used to simulate the phenomena in the reactor core and reactor cooling system and COCOSYS code is used to simulate the confinement phenomena during the same accident scenario. The performed analysis provided information regarding code acceptability for the severe accident analysis in RBMK reactor and assessment of the timing of the key events, i.e. core uncover, fuel cladding rupture, etc, and provided assessment regarding hydrogen distribution in confinement. (authors)« less
  • No abstract prepared.
  • A numerical study of the thermal and electrochemical performance of a single-tube Integrated Planar Solid Oxide Fuel Cell (IP-SOFC) has been performed. Results obtained from two finite-volume computational fluid dynamics (CFD) codes FLUENT and SOHAB and from a two-dimensional inhouse developed finite-volume GENOA model are presented and compared. Each tool uses physical and geometric models of differing complexity and comparisons are made to assess their relative merits. Several single-tube simulations were run using each code over a range of operating conditions. The results include polarization curves, distributions of local current density, composition and temperature. Comparisons of these results are discussed,more » along with their relationship to the respective imbedded phenomenological models for activation losses, fluid flow and mass transport in porous media. In general, agreement between the codes was within 15% for overall parameters such as operating voltage and maximum temperature. The CFD results clearly show the effects of internal structure on the distributions of gas flows and related quantities within the electrochemical cells.« less