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Watanabe, Tomoaki; Suyama, Kenya; Tada, Kenichi; Ferrer, R. M.*; Hykes, J.*; Wemple, C. A.*
Nuclear Science and Engineering, 10 Pages, 2024/00
Times Cited Count：0 Percentile：0.18(Nuclear Science & Technology)A new nuclear data library for the advanced lattice physics code CASMO5 has been prepared based on JENDL-5. In JENDL-5, many essential nuclides for conventional LWR analysis have also been modified based on state-of-the-art evaluations. The new JENDL-5-based CASMO5 library was prepared by replacing as much of the nuclear data of the current CASMO5 ENDF/B-VII.1-based library as possible with JENDL-5. This study verified and validated the new library. Verifications were performed based on the OECD/NEA burnup credit criticality safety benchmark phase III-C, and the calculated k
and fuel compositions of the BWR fuel assembly were compared with reported benchmark results. Comparison with the MCNP6.2 result was also performed using the same benchmark model. In addition, the TCA critical experiment and Takahama-3 post-irradiation experiment were used for validation. The results indicate that the new library performs well and is comparable to the ENDF/B-VII.1-based library in predictions of reactivity and fuel compositions for LWR systems.
Okumura, Keisuke; Sakamoto, Yukio*; Tsukiyama, Toshihisa*
Shahei Kaiseki No V&V Gaidorain Sakutei Ni Mukete, p.4 - 8, 2023/03
no abstracts in English
Yamamoto, Akio*; Tada, Kenichi; Chiba, Go*; Endo, Tomohiro*
Transactions of the American Nuclear Society, 124(1), p.544 - 547, 2021/06
Verification calculations for the capability of multi-group cross section generation in FRENDY (FRENDY/MG) are carried out through the comparison of one-group reaction rates using the multi-group cross sections obtained by FRENDY/MG and NJOY2016. Three different neutron spectra (LWR, FR, and 1/E) are used to calculate one-group reaction rates. The discrepancies of one-group reaction rates are small for most cases, showing the validity of FRENDY/MG. The FRENDY/MG will be released as the part of FRENDY nuclear data processing system in the near future.
Chiba, Go*; Yamamoto, Akio*; Tada, Kenichi; Endo, Tomohiro*
Transactions of the American Nuclear Society, 124(1), p.556 - 558, 2021/06
The FRENDY nuclear data processing code has been used to generate multi-group cross section libraries for the CBZ reactor physics code system. The newly generated libraries have been applied to neutronics calculations of a fast reactor core MET-1000, and several neutronics parameters are calculated. Calculations with other libraries generated by NJOY2016 have been also conducted, and differences in obtained neutronics parameters between the FRENDY-based library and the NJOY-based library have been quantified. Generally reasonable agreement between them has been obtained, so it has been demonstrated that the multi-group libraries for fast reactor neutronics calculations can be generated successfully by FRENDY. Detailed investigation on the impact of the difference in the processing codes on k-effective has been also carried out with a help of the perturbation theory, and the causes of the differences have been identified.
Tanaka, Masaaki; Nakada, Kotaro*; Kudo, Yoshiro*
Nihon Kikai Gakkai-Shi, 123(1222), p.26 - 29, 2020/09
In the nuclear engineering, simulations are used in radiation, thermal hydraulic, chemical reaction, and structural fields, and the integrated fields thereof, to be applied to the design, construction and operation of nuclear facilities. This article describes brief history of discussion in the AESJ to the publication and introductory explanation of the procedures in the five major elements described in the "Guideline for Credibility Assessment of Nuclear Simulations (AESJ-SC-A008: 2015)". And also, a practical experience of the V&V activity according to the fundamental concept indicated in the Guideline is introduced.
Tanaka, Masaaki; Kudo, Yoshiro*; Nakada, Kotaro*; Koshizuka, Seiichi*
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.1473 - 1484, 2019/08
Verification and validation (V&V) including uncertainty quantification on modeling and simulation activities has been very much focused on. Due to increase of requirement for standardization of the procedures on the V&V and prediction process to enhance the simulation credibility, "Guideline for Credibility Assessment of Nuclear Simulations (AESJ-SC-A008: 2015)" was published on July 2016 from the AESJ through ten-year discussion. The paper describes brief history of discussion in the AESJ to the publication and introductory explanation of the procedures in the five major elements and one scheme described in the Guideline. And also, a practical experience of the V&V activity according to the fundamental concept indicated in the Guideline is introduced.
Chai, P.; Yamashita, Susumu; Nagae, Yuji; Kurata, Masaki
Proceedings of 9th Conference on Severe Accident Research (ERMSAR 2019) (Internet), 14 Pages, 2019/03
In order to obtain a precise understanding of molten material behavior inside RPV and to improve the accuracy of the SA code, a new computational fluid dynamics (CFD) code with multi-phase, multi-physics models, which is called JUPITER, was developed. It optimized the algorithms of the multi-phase calculation. Besides, the chemical reactions are also modeled carefully in the code so that the melting process could be treated precisely. A series of verification and validation studies are conducted, which show good agreement with analytical solutions and previous experiments. The capabilities of the multi-physics models in JUPITER code provide us another useful tool to investigate the molten material behaviors in the relevant severe accident scenario.
Nakada, Kotaro*; Kudo, Yoshiro*; Koshizuka, Seiichi*; Tanaka, Masaaki
Nihon Genshiryoku Gakkai-Shi ATOMO
, 60(3), p.173 - 177, 2018/03
The Atomic Energy Society of Japan (AESJ) published "Guideline for Credibility Assessment of Nuclear Simulations 2015" in June, 2016 which specifies the concepts on methodology for the prediction with uncertainty quantification and the quality management based on the concept of verification and validation (V&V) of modeling and simulation. In this report, the outlines of activities in AESJ for publication of the guideline and the expectation for effective implementation of the guideline are described including that of the lectures with major respondents of the questionnaires.
Ohgama, Kazuya; Aliberti, G.*; Stauff, N. E.*; Oki, Shigeo; Kim, T. K.*
Mechanical Engineering Journal (Internet), 4(3), p.16-00592_1 - 16-00592_9, 2017/06
Ohgama, Kazuya; Ikeda, Kazumi*; Ishikawa, Makoto; Kan, Taro*; Maruyama, Shuhei; Yokoyama, Kenji; Sugino, Kazuteru; Nagaya, Yasunobu; Oki, Shigeo
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
Ohgama, Kazuya; Aliberti, G.*; Stauff, N. E.*; Oki, Shigeo; Kim, T. K.*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04
Ichikawa, Shoichi; Chiba, Yusuke; Ono, Fumiyasu; Hatori, Masakazu; Kobayashi, Takanori; Uekura, Ryoichi; Hashiri, Nobuo*; Inuzuka, Taisuke*; Kitano, Hiroshi*; Abe, Hisashi*
JAEA-Research 2017-001, 40 Pages, 2017/03
JAEA-Research-2017-001.pdf:5.19MB
In order to reduce the influence on a plant schedule of the MONJU by the maintenance of dew point hygrometers, The JAEA examined a capacitance type dew point hygrometer as an alternative dew point hygrometer for a lithium-chloride type dew point hygrometer which had been used at the CV-LRT in the MONJU. As a result of comparing a capacitance type dew point hygrometer with a lithium-chloride type dew point hygrometer at the CV-LRT (Atmosphere: nitrogen, Testing time: 24 hours), there weren't significant difference between a capacitance type dew point hygrometer and a lithium-chloride type dew point hygrometer. As a result of comparing a capacitance dew point hygrometer with a high-mirror-surface type dew point hygrometer for long term verification (Atmosphere: air, Testing time: 24 months), the JAEA confirmed that a capacitance type dew point hygrometer satisfied the instrument specification (
2.04
C) required by the JEAC4203-2008.
Tanaka, Masaaki
Nihon Genshiryoku Gakkai Keisan Kagaku Gijutsu Bukai Nyusu Reta (Internet), (27), p.9 - 15, 2017/03
In this report, the outline of the contents in the lecture course for "Guideline for Credibility Assessment of Nuclear Simulations 2015" published in June, 2016 from the Atomic Energy Society of Japan (AESJ) and the result of the lectures held in Tokyo and Osaka are introduced with the results of the questionnaires from the audience.
Ohgama, Kazuya; Aliberti, G.*; Stauff, N. E.*; Oki, Shigeo; Kim, T. K.*
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 6 Pages, 2016/06
Under the cooperative effort of the Civil Nuclear Energy R&D Working Group within the framework of the U.S.-Japan bilateral, Argonne National Laboratory (ANL) and Japan Atomic Energy Agency (JAEA) have been performing benchmark study using Japan Sodium-cooled Fast Reactor (JSFR) design with metal fuel. In this benchmark study, core characteristic parameters at the beginning of cycle were evaluated by the best estimate deterministic and stochastic methodologies of ANL and JAEA. The results obtained by both institutions are agreed well with less than 200 pcm of discrepancy on the neutron multiplication factor, and less than 3% of discrepancy on the sodium void reactivity, Doppler reactivity, and control rod worth. The results by the stochastic and deterministic were compared in each party to investigate impacts of the deterministic approximation and to understand potential variations in the results due to different calculation methodologies employed. Impacts of the nuclear data libraries were also investigated using a sensitivity analysis methodology.
Tanaka, Masaaki
Nihon Genshiryoku Gakkai Keisan Kagaku Gijutsu Bukai Nyusu Reta (Internet), (24), p.16 - 28, 2015/09
In order to enhance the simulation credibility, it is necessary to establish the procedure on verification and validation including the estimation of the modeling uncertainty. Lessons learned from the Fukushima Daiichi NPP Accident have indicated that it was important to recognize the credibility of the simulation. By putting forward to standardize the procedure on verification and validation including the estimation of the modeling uncertainty, it is expected to establish the basis of the simulation technology to realize the world highest level of nuclear safety and continuous improvement. The recent activity in the Atomic Energy Society of Japan (AESJ) for the guideline establishment is introduced.
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.
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.
Okoshi, Minoru
Dekomisshoningu Giho, (31), p.32 - 44, 2005/03
The International Atomic Energy Agency (IAEA) published a Safety Guide (RS-G-1.7) on the application of the concepts of exclusion, exemption and clearance in 2004. In this Safety Guide, the IAEA proposes the activity concentrations can be applied to clearance. The Nuclear Safety Commission reevaluated their clearance levels to reflect the new technological information given in the IAEA Safety Guide and other reports. The Regulatory Authorities have been investigating and discussing regulatory rules on clearance including the verification system of clearance levels. And the Atomic Society of Japan has deliberated the technological standards for the verification of clearance levels performed by nuclear operators. In this paper the activities and outcomes by those organizations are outlined, and the issues to be solved on clearance are listed and the basic ideas for solution are discussed.
Tsutsumi, Hideaki*; Yamada, Hiroyuki; Teragaki, Toshio*; Ebisawa, Katsumi; Shibata, Katsuyuki
Seismic Engineering 2000 (PVP-Vol.402-1), p.141 - 146, 2000/00
no abstracts in English
Yamamoto, Yoichi; Mukaiyama, Takehiko
JAERI-Tech 98-035, 31 Pages, 1998/08
no abstracts in English
