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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

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

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

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

Distillation of cadmium from uranium-plutonium-cadmium alloy

Kato, Tetsuya*; Iizuka, Masatoshi*; Inoue, Tadashi*; Iwai, Takashi; Arai, Yasuo

Journal of Nuclear Materials, 340(2-3), p.259 - 265, 2005/04

 Times Cited Count:22 Percentile:82.13(Materials Science, Multidisciplinary)

no abstracts in English

JAEA Reports

Neutronic study on seed-blanket type reduced-moderation water reactor fuel assembly

Shelley, A.; Kugo, Teruhiko; Shimada, Shoichiro*; Okubo, Tsutomu; Iwamura, Takamichi

JAERI-Research 2004-002, 47 Pages, 2004/03

JAERI-Research-2004-002.pdf:3.08MB

Neutronic study has been done for a PWR-type reduced-moderation water reactor with seed-blanket fuel assemblies to achieve a high conversion ratio, a negative void coefficient and a high burnup by using a MOX fuel. The results of the precise assembly burnup calculations show that the recommended numbers of seed and blanket layers are 15(S15) and 5(B5), respectively. By the optimization of axial configuration, the S15B5 assembly with the seed of 1000$$times$$2 mm high, internal blanket of 150 mm high and axial blanket of 400$$times$$2 mm high is recommended. In this configuration, the conversion ratio is 1.0 and the core average burnup is 38 GWd/t. The S15B5 assembly can attain the core average burnup of 45 GWd/t by decreasing the height of seed to 500$$times$$2 mm, however, the conversion ratio becomes 0.97. The void and fuel temperature coefficients are negative for both of the configurations. Effect of metal or T-MOX (PuO$$_{2}$$+ThO$$_{2}$$) fuel has been also investigated. Metal improves the conversion ratio but makes the void coefficient worse. T-MOX improves the void coefficient, but decreases the conversion ratio.

JAEA Reports

Study on the stress corrosion cracking of zirconium in nuclear fuel reprocessing environment

Kato, Chiaki

JAERI-Research 2003-013, 143 Pages, 2003/08

JAERI-Research-2003-013.pdf:22.12MB

This study is investigation about stress corrosion cracking (SCC) of zirconium in nuclear fuel reprocessing. Chapter 1 is described background. Chapter 2 is explained experimental apparates. Chapter 3 is described the increased oxidization potential on the heat-transfer surface and suggested the initiation of SCC on a boiling heat-transfer surface. Chapter 4 is described that the SCC susceptibility increased with increasing nitric acid concentration and solution temperature on notched specimen by SSRT. In addition, the SCC susceptibility effected by the crystal anisotropy by the hot rolling direction and increased on a parallel face to the rolling direction. Chapter 5 is described that the SCC susceptibility increased in HAZ/base metal boundary in order to the preferential orientation of cleavage plane (0002). Chapter 6 is described that the increased oxidization potential on the heat-transfer surface is attributed to the reduction of nitrous acid concentration by the thermal decomposition on the surface and the removal of the decomposition product from solution by boiling bubbles.

JAEA Reports

Modeling of fuel performance and fission product release behaviour during HTTR normal operation; A Comparative study on the FZJ and JAERI Modeling approach

Verfondern, K.*; Sumita, Junya; Ueta, Shohei; Sawa, Kazuhiro

JAERI-Research 2000-067, 127 Pages, 2001/03

JAERI-Research-2000-067.pdf:6.64MB

no abstracts in English

Journal Articles

Irradiation behavior of microspheres of U-Zr alloys

Ogawa, Toru; *; Ito, Akinori; ; Sekino, Hajime; Nishi, Masahiro; Ishikawa, Akiyoshi; Akabori, Mitsuo

Journal of Alloys and Compounds, 271-273, p.670 - 675, 1998/00

 Times Cited Count:7 Percentile:51.14(Chemistry, Physical)

no abstracts in English

JAEA Reports

Measurement test on creep strain rate of uranium-zirconium solid solutions

*; Akabori, Mitsuo; Ogawa, Toru

JAERI-Tech 96-052, 18 Pages, 1996/11

JAERI-Tech-96-052.pdf:0.92MB

no abstracts in English

JAEA Reports

Experimental study on sodium void reactivity worth in mockup cores of metallic-fueled and MOX-fueled fast reactors using FCA

Oigawa, Hiroyuki; Iijima, Susumu

JAERI-Research 95-007, 93 Pages, 1995/02

JAERI-Research-95-007.pdf:2.7MB

no abstracts in English

Journal Articles

Verification of fission product release model from High Temperature Engineering Test Reactor fuel

Sawa, Kazuhiro; Shiozawa, Shusaku; Fukuda, Kosaku;

Journal of Nuclear Science and Technology, 29(9), p.842 - 850, 1992/09

no abstracts in English

Journal Articles

Demixing of U-Zr alloys under a thermal gradient

Ogawa, Toru; Iwai, Takashi; *

J. Less-Common Met., 175, p.59 - 69, 1991/00

no abstracts in English

JAEA Reports

Core design study for hybrid type trans-uranium nuclides incineration plant, Part 1; Concept

Takada, Hiroshi; Kanno, Ikuo; Takizuka, Takakazu; Akabori, Mitsuo; ; Kaneko, Yoshihiko

JAERI-M 90-131, 18 Pages, 1990/08

JAERI-M-90-131.pdf:0.59MB

no abstracts in English

Journal Articles

Development of the SESAME metallic fuel performance code

*; *; *; Ogawa, Toru; *; *; *; Saito, Hioraki*

Nuclear Technology, 89, p.183 - 193, 1990/02

 Times Cited Count:24 Percentile:89.15(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Fuel elements and fuel cycle concepts of actinide burner reactors; Conceptual study of actinide burner reacotors, IV

Ogawa, Toru; Mukaiyama, Takehiko; Takano, Hideki; Takizuka, Takakazu; ; *

JAERI-M 89-123, 38 Pages, 1989/09

JAERI-M-89-123.pdf:1.25MB

no abstracts in English

Journal Articles

Theoretical estimatuion of thermodynamic properties of metallic fuel

Ogawa, Toru; Iwai, Takashi

FBR Kinzoku Nenryo Saikuru No Seiritsusei Hyoka, Kinzoku Nenryo Kyodo Kaiseki Kodo No Kaihatsu, Sono 2; Kani Kaiseki Kodo No Kairyo, Kensho To Shosai Kaiseki Kodo No Kihon Sekkei, Denryoku Chuo Kenkyusho Hokoku, p.232 - 244, 1989/00

no abstracts in English

Journal Articles

Cladding embrittlement and fuel rod failure threshold under reactivity initiated accident condition

;

Journal of Nuclear Science and Technology, 24(11), p.897 - 905, 1987/11

 Times Cited Count:2 Percentile:30.22(Nuclear Science & Technology)

no abstracts in English

28 (Records 1-20 displayed on this page)