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Journal Articles

Verification of detailed core-bowing analysis code ARKAS_cellule with IAEA benchmark problems

Ota, Hirokazu*; Ohgama, Kazuya; Yamano, Hidemasa

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.30 - 39, 2019/09

Journal Articles

A Validation study of a neutronics design methodology for fast reactors using reaction rate distribution measurements in the prototype fast reactor Monju

Ohgama, Kazuya; Takegoshi, Atsushi; Katagiri, Hiroki*; Hazama, Taira

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 8 Pages, 2019/05

Journal Articles

A Design study on a mixed oxide fuel sodium-cooled fast reactor core partially loading highly concentrated MA-containing metal fuel

Ohgama, Kazuya; Ota, Hirokazu*; Oki, Shigeo; Iizuka, Masatoshi*

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 9 Pages, 2019/05

Journal Articles

Study on heterogeneous minor actinide loading fast reactor core concepts with improved safety

Ohgama, Kazuya; Oki, Shigeo; Kitada, Takanori*; Takeda, Toshikazu*

Proceedings of 21st Pacific Basin Nuclear Conference (PBNC 2018) (USB Flash Drive), p.942 - 947, 2018/09

Journal Articles

Comparative study on neutronics characteristics of a 1500 MWe metal fuel sodium-cooled fast reactor

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

Journal Articles

Model verification and validation procedure for a neutronics design methodology of next generation fast reactors

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

Journal Articles

Comparative study on burnup characteristics of a 1500 MWe metal fuel sodium-cooled fast reactor

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

Journal Articles

Tradeoff analysis of metal-fueled fast reactor design concepts

Stauff, N. E.*; Ohgama, Kazuya; Aliberti, G.*; Oki, Shigeo; Kim, T. K.*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

Journal Articles

Design study of a 750 MWe Japan sodium-cooled fast reactor with metal fuel

Ohgama, Kazuya; Ota, Hirokazu*; Ikusawa, Yoshihisa; Oki, Shigeo; Ogata, Takanari*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04

Journal Articles

Core design of the next-generation sodium-cooled fast reactor in Japan

Kan, Taro*; Ogura, Masashi*; Hibi, Koki*; Oki, Shigeo; Maeda, Seiichiro; Maruyama, Shuhei; Ohgama, Kazuya

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

Journal Articles

Analysis of fuel subassembly innerduct configurational effects on the core characteristics and power distribution of a sodium-cooled fast breeder reactor

Ohgama, Kazuya; Nakano, Yoshihiro; Oki, Shigeo

Journal of Nuclear Science and Technology, 53(8), p.1155 - 1163, 2016/08

 Times Cited Count:1 Percentile:82.04(Nuclear Science & Technology)

The power distribution and core characteristics in various configurations of fuel subassemblies with an innerduct structure in the Japan Sodium-cooled Fast Reactor were evaluated using a Monte Carlo code for neutron transport and burnup calculation. The correlation between the fraction of fuel subassemblies facing outward and the degree of power increase at the core center was observed regardless of the compositions. This indicated that the spatial fissile distribution caused by innerduct configurations was the major factor of the difference in the power distribution. A power increase was also found in an off-center region, and it tended to be greater than that at the core center because of the steep gradient of neutron flux intensity. The differences in the worth of control rods caused by innerduct configurations were confirmed.

Journal Articles

Comparative study on neutronics characteristics of a 1500 MWe metal fuel sodium-cooled fast reactor

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.

Journal Articles

Revision of the JENDL FP fission yield data

Katakura, Junichi*; Minato, Futoshi; Ohgama, Kazuya

EPJ Web of Conferences (Internet), 111, p.08004_1 - 08004_5, 2016/03

 Times Cited Count:9 Percentile:0.74

The JENDL FP Fission Yield Data Library (JENDL/FPY-2011) was released in 2011. Although the data of the yield file are successfully applied to decay heat summation calculations, there have been reported some inadequacies when applied to delayed neutron related subjects. And also from sensitivity analyses of summation calculation, some fission yield data in the JENDL file have been claimed to have some problems. In order to remedy those problems some yield data of the JENDL file have been re-examined and revised. The following is one example. The yield data of $$^{86}$$Ge for thermal neutron fission of $$^{235}$$U is given to be 6.277$$times10^{-3}$$ in JENDL/FPY-2011 which has been claimed to be too large. The re-examination and re-calculation of the fission yield data now give the new value of 3.437$$times10^{-6}$$ which seems to be more reasonable. There are some other nuclides indicated by the sensitivity analyses. The process of the re-examination of those nuclides and the revised yield data will be presented in the workshop WONDER 2015.

Journal Articles

Improvement of transient analysis method of a sodium-cooled fast reactor with FAIDUS fuel sub-assemblies

Ohgama, Kazuya; Kawashima, Katsuyuki*; Oki, Shigeo

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

In order to evaluate transient behavior of Japan sodium-cooled fast reactor (JSFR) with fuel sub-assemblies with the innerduct structure (FAIDUS) precisely, a new model for a plant dynamics code HIPRAC was developed. In this new model, inner core and outer core channels can be divided into three channels, respectively, such as interior, edge and near innerduct channel, and calculate coolant redistribution and coolant temperature in each channel. Coolant temperature distribution of interior and edge channels calculated by this model was compared with previous study by the general-purpose thermal-hydraulics code $$alpha$$-FLOW. Coolant temperature behavior inside the innerduct was analyzed by a commercial thermal hydraulics code STAR-CD ver. 3.26. Based on this result, horizontally-uniformed coolant temperature in the innerduct was assumed as a heat transfer model of the innderduct. Reactivity coefficients for 750 MWe JSFR with low -decontaminated transuranic (TRU) fuel were evaluated. Transient behaviors of an unprotected loss-of-flow (ULOF) accident for JSFR with 750 MWe output calculated by previous and new models were compared. The results showed that the detailed evaluation of coolant temperature improved overestimation of the coolant temperature and coolant temperature feedback reactivity of the peripheral channels including coolant inside the innerduct and in the inter-wrapper gap.

Journal Articles

Info session on human networking held in Japan-IAEA Joint Nuclear Energy Management School; Aiming to develop human network among nuclear young generation in the world

Nishiyama, Jun*; Ohgama, Kazuya; Sakamoto, Tatsujiro*; Watanabe, Rin*

Nippon Genshiryoku Gakkai-Shi, 57(2), p.123 - 125, 2015/02

no abstracts in English

Journal Articles

Comparison and sensitivity analysis of the core characteristics of a sodium-cooled fast breeder reactor with 750 MWe output evaluated by JENDL-4.0 and ADJ2000R

Ohgama, Kazuya; Oki, Shigeo; Sugino, Kazuteru; Okubo, Tsutomu

Journal of Nuclear Science and Technology, 51(4), p.558 - 567, 2014/04

 Times Cited Count:1 Percentile:87.59(Nuclear Science & Technology)

Core characteristics of a sodium-cooled fast breeder reactor (FBR) with 750 MWe output using highly decontaminated uranium and plutonium and highly minor-actinide-containing compositions were evaluated using the fast reactor cross-section set generated by the new Japanese nuclear data library JENDL-4.0. The core characteristics were compared with those obtained using the unified cross-section set ADJ2000R in order to investigate the differences between both the results. The effects on the core characteristics caused by the differences in the nuclear data of important reactions and nuclides in the cross-section sets were analyzed by a burnup sensitivity analysis. It was confirmed that adopting JENDL-4.0 to the FBR core design improves the breeding ratio, the burnup reactivity, and the reactivity control balance, because of the differences in the capture cross-sections of U-238 and Pu-239 of both the libraries. The difference in the sodium void reactivity evaluated with both the libraries was less than 1% because the increase caused by the differences in the elastic scattering cross-sections of sodium, the inelastic scattering cross-section, and the $$mu$$-average value of U-238 was practically cancelled out by the decrease caused by the differences in the capture cross-sections of Pu-239, the inelastic scattering cross-section of iron, and the capture cross- sections of Am-241.

JAEA Reports

The International Atomic Energy Agency Nuclear Energy Management School 2012

Ohgama, Kazuya; Ando, Yoko; Yamaguchi, Mika; Ikuta, Yuko; Shinohara, Nobuo; Murakami, Hiroyuki; Yamashita, Kiyonobu; Uesaka, Mitsuru*; Demachi, Kazuyuki*; Komiyama, Ryoichi*; et al.

JAEA-Review 2013-004, 76 Pages, 2013/05

JAEA-Review-2013-004.pdf:13.53MB

JAEA together with the Japan Nuclear Human Resource Development Network (JN-HRD Net), the University of Tokyo (UT) and the Japan Atomic Industrial Forum (JAIF) cohosted the IAEA-Nuclear Energy Management School in Tokai Village, aiming that Japan will be the center of nuclear HRD in the Asian region. In the school, not only lectures by IAEA experts, but also lectures by Japanese experts and technical visits were included for foreign participants. The school contributed to the internationalization of Japanese young professionals, development of nuclear human resource of other countries, and enhancement of cooperation between IAEA and Japan. Additionally, collaborative relationship within JN-HRD Net was strengthened by the school. In this report, findings obtained during the preparatory work and the school period are reported for future international nuclear HRD activities in Japan.

Journal Articles

IAEA Nuclear Energy Management School holding in Japan; Aiming for Nuclear Human Resource Development Center in Asia

Uesaka, Mitsuru*; Yamashita, Kiyonobu; Ohgama, Kazuya; Hattori, Takuya*

Nippon Genshiryoku Gakkai-Shi, 55(1), p.53 - 56, 2013/01

no abstracts in English

Journal Articles

Fast Breeder Reactor core design study using JENDL-4.0

Ohgama, Kazuya; Oki, Shigeo; Okubo, Tsutomu

JAEA-Conf 2012-001, p.21 - 26, 2012/07

Journal Articles

The Experience of World Nuclear University Summer Institute 2010

Ohgama, Kazuya; Ogino, Haruyuki*; Sato, Takahiko*; Suzuki, Ayako*

Nippon Genshiryoku Gakkai-Shi, 53(7), p.509 - 513, 2011/07

no abstracts in English

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