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Nakamura, Hideo; Bentaib, A.*; Herranz, L. E.*; Ruyer, P.*; Mascari, F.*; Jacquemain, D.*; Adorni, M.*
Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 10 Pages, 2022/10
and MVP on the MCNP criticality validation suiteBrun, E.*; Zoia, A.*; Trama, J. C.*; Lahaye, S.*; Nagaya, Yasunobu
Proceedings of International Conference on Nuclear Criticality Safety (ICNC 2015) (DVD-ROM), p.351 - 360, 2015/09
This paper presents a joint work conducted at CEA Saclay and JAEA Tokai aimed at comparing the Monte Carlo codes TRIPOLI and MVP on a selection of ICSBEP benchmarks. Our goal is to establish a common set of Monte Carlo input decks, as a basis for rigorous inter-code comparison in criticality-safety. As a reference, we will use the MCNP Criticality Validation Suite: other Monte Carlo developers might easily join that effort in the future. For the purpose of inter-code comparison, the TRIPOLI and MVP input decks have been translated from those of MCNP, without any further assumptions. Both TRIPOLI and MVP have been run with the same ENDF/B-VII.0 evaluated nuclear data, and as far as possible the same simulation options as in the original LANL work. In this abstract, we present preliminary results on the BIGTEN benchmark. In the full paper these will be extended to the 31 benchmarks of the MCNP Criticality Validation Suite. In the future, this database will also help in the analysis of sensitivity to nuclear data, CPU times and figures of merit.
Kamide, Hideki; Ohshima, Hiroyuki; Sakai, Takaaki; Tanaka, Masaaki
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.8141 - 8155, 2015/08
In this paper, the authors focus on four kinds of thermal-hydraulic issues associated with the SDC, i.e. fuel assembly thermal-hydraulics, natural circulation decay heat removal, thermal striping phenomena, and core disruptive accidents, and provide a description of their evaluation method developments including verification and validation and necessary experimental studies for the Japan Sodium-cooled Fast Reactor (JSFR). These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all phenomena envisioned in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing down of knowledge/technologies.
Kobayashi, Jun; Tanaka, Masaaki; Ohno, Shuji; Ohshima, Hiroyuki; Kamide, Hideki
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.6664 - 6677, 2015/08
Numerical simulation is recognized an essential tool for the physical phenomena analysis and plant design study of a sodium-cooled fast reactor (SFR). In order to enhance credibility of the numerical results in the activities for plant design by using numerical simulations, it is recognized that verification and validation (V&V) process is very important. In this study, experiments for planar triple parallel jets mixing phenomena conducted in JAEA were proposed as benchmark problems for the code validation in the area of thermal striping study in the SFR development.
Shimazaki, Yosuke; Isaka, Kazuyoshi; Nomoto, Yasunobu; Seki, Tomokazu; Ohashi, Hirofumi
JAEA-Technology 2014-038, 51 Pages, 2014/12
JAEA-Technology-2014-038.pdf:3.84MB
The analytical models for the evaluation of graphite oxidation were implemented into the THYTAN code, which employs the mass balance and a node-link computational scheme to evaluate tritium behavior in the High Temperature Gas-cooled Reactor (HTGR) systems for hydrogen production, to analyze the graphite oxidation during the air or water ingress accidents in the HTGR systems. This report describes the analytical models of the THYTAN code in terms of the graphite oxidation analysis and its verification and validation (V&V) results. Mass transfer from the gas mixture in the coolant channel to the graphite surface, diffusion in the graphite, graphite oxidation by air or water, chemical reaction and release from the primary circuit to the containment vessel by a safety valve were modeled to calculate the mass balance in the graphite and the gas mixture in the coolant channel. The computed solutions using the THYTAN code for simple questions were compared to the analytical results by a hand calculation to verify the algorithms for each implemented analytical model. A representation of the graphite oxidation experimental was analyzed using the THYTAN code, and the results were compared to the experimental data and the computed solutions using the GRACE code, which was used for the safety analysis of the High Temperature Engineering Test Reactor (HTTR), in regard to corrosion depth of graphite and oxygen concentration at the outlet of the test section to validate the analytical models of the THYTAN code. The comparison of THYTAN code results with the analytical solutions, experimental data and the GRACE code results showed the good agreement.
Suyama, Kenya; Mochizuki, Hiroki*; Okuno, Hiroshi; Miyoshi, Yoshinori
Proceedings of International Conference on Physics of Fuel Cycles and Advanced Nuclear Systems; Global Developments (PHYSOR 2004) (CD-ROM), 10 Pages, 2004/04
This paper provides validation results of SWAT2, the revised version of SWAT, which is a code system combining point burnup code ORIGEN2 and continuous energy Monte Carlo code MVP, by the analysis of post irradiation examinations (PIEs). Some isotopes show differences of calculation results between SWAT and SWAT2. However, generally, the differences are smaller than the error of PIE analysis that was reported in previous SWAT validation activity, and improved results are obtained for several important fission product nuclides. This study also includes comparison between an assembly and a single pin cell geometry models.
Okuno, Hiroshi; Nomura, Yasushi
Proceedings of the 2001 Topical Meeting on Practical Implementation of Nuclear Criticality Safety (CD-ROM), 8 Pages, 2001/11
The nuclear criticality safety handbook of Japan was first published in 1988, which was translated into English in 1995. This paper intends to introduce the American community of nuclear criticality safety to activities for revising the Japanese handbook, putting an emphasis on practical use of code validation. They include (1) publication of "Nuclear Criticality Safety Handbook, Version 2" and its English translation, (2) publication of "A Guide Introducing Burnup Credit, Preliminary Version," and (3) preparation of "Nuclear Criticality Safety Handbook, -Data Collection-, Version 2."
Takada, Hiroshi; Meigo, Shinichiro; Niita, Koji*
Advanced Monte Carlo for Radiation Physics, Particle Transport Simulation and Applications, p.949 - 954, 2001/00
no abstracts in English
Takeda, Takeshi; Nakagawa, Shigeaki; Tachibana, Yukio; Takada, Eiji*; Kunitomi, Kazuhiko
JAERI-Research 2000-016, p.80 - 0, 2000/03
JAERI-Research-2000-016.pdf:2.72MB
no abstracts in English
Noguchi, Hiroshi; Yokoyama, Sumi
KURRI-KR-30, p.204 - 209, 1998/00
no abstracts in English
Katakura, Junichi; Okuno, Hiroshi; Naito, Yoshitaka
Proc.Int.Seminar on Nuclear Criticality Safety, p.138 - 144, 1987/00
no abstracts in English
Nomura, Yasushi; Katakura, Junichi; Naito, Yoshitaka; Komuro, Yuichi; Okuno, Hiroshi
JAERI 1303, 152 Pages, 1986/11
no abstracts in English