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Journal Articles

Analysis of fast reactor fuel irradiation behavior in the MA recycle system

Ozawa, Takayuki

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 8 Pages, 2017/07

In a recycle system for minor actinides (MAs) currently studied to reduce the degree of hazard and the amount of high-level radioactive wastes, MAs will be recycled by reprocessing and irradiating as mixed oxide (MOX) with plutonium (Pu) and uranium (U) in a fast reactor. MA content is expected to be $$sim$$5 wt.% in the future recycle system, and MAs might affect irradiation behavior of MA-MOX fuels. The main influences of MA-containing would be increase of fuel temperature and cladding stress, and the important behaviors would be fuel restructuring, redistribution, helium (He) generation and cladding corrosion. The MA-containing influences were evaluated with CEPTAR.V2, including fuel properties and analysis models to evaluate the MA-MOX fuel irradiation behavior, by using the results of highly americium (Am) containing MOX irradiation experiment, B8-HAM, performed in Joyo. The irradiation behavior of Am-MOX fuels could be precisely analyzed and revealed the influences of Am-containing.

Journal Articles

LOCA and RIA studies at JAERI

Sugiyama, Tomoyuki; Nagase, Fumihisa; Nakamura, Jinichi; Fuketa, Toyoshi

HPR-362, Vol.2, 12 Pages, 2004/05

To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as loss of coolant accident (LOCA) and reactivity-initiated accident (RIA) is being studied at the Japan Atomic Energy Research Institute (JAERI). The LOCA program consists of integral thermal shock tests and other separate tests for oxidation rate and mechanical property of fuel claddings. Prior to the tests on irradiated claddings, the tests have been conducted on non-irradiated claddings to examine separate effects of corrosion and hydrogen absorption during reactor operation. The tests on irradiated claddings have recently been started and results have been obtained. As for an RIA study, a series of experiments with high burnup fuel rods is being performed by using pulse irradiation capability of the NSRR. This paper presents recent results obtained from the LOCA and RIA studies at JAERI.

JAEA Reports

None

Yamanaka, Shinsuke*; Abe, Kazuyuki

JNC-TY9400 2000-004, 78 Pages, 2000/03

JNC-TY9400-2000-004.pdf:2.39MB

no abstracts in English

JAEA Reports

Modification of the evaluation model for Pu redistribution phenomena

; *;

JNC-TN9400 2000-045, 64 Pages, 2000/03

JNC-TN9400-2000-045.pdf:2.47MB

During the irradiation, the Pu redistribution phenomena would occur in the FBR MOX fuel pellets. The phenomena would considerably affect on the thermal properties of the fuels, therefore, it is need to establish the evaluation method for Pu redistribution phenomena. ln JNC, the efforts for development of the evaluation model for the phenomena had been continued and the simple evaluation model was constructed in 1992. In this work, the modification of the simple model developed in JNC has been done and the following results were obtained. (1)Based on the recent data of the MOX fuel irradiation tests, the evaluation model for Pu redistribution phenomena constructed in l992 is modified. And the model is included into the fuel performance analysis code "CEDAR". (2)To calibrate the modified CEDAR code, it is confirmed that the uncertainty in the Pu concentration evaluation for the center of the fuel pellet at EOL is about $$pm$$3wt.%. (3)Based on the results of the evaluations using the modified CEDAR code, it is found that, in the early stage of the irradiation, the Pu redistribution is controlled by the vapor transportation mechanism via pores, and after that, the Pu redistribution is kept in progress due to the thermal diffusion mechanism with the change of the Pu concentration due to the degradation of U and Pu by fissions. And it is also found that the O/M ratio dependence of the U-Pu inter diffusion coefficients would affect on the Pu redistribution mechanisms, in especial, in the early stage of the irradiation.

JAEA Reports

Development of sodium facilities for NSRR fast reactor fuel tests, 2; Sodium capsule

Yoshinaga, Makio; Nakamura, Takehiko; Yamazaki, Toshi*

JAERI-Tech 2000-017, p.59 - 0, 2000/03

JAERI-Tech-2000-017.pdf:2.31MB

no abstracts in English

JAEA Reports

None

*; Iriya, Yoshikazu*

JNC-TJ8440 99-003, 156 Pages, 1999/03

JNC-TJ8440-99-003.pdf:2.72MB

no abstracts in English

Journal Articles

MOX fuel behavior under reactivity accident conditions

Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; *; Abe, Tomoyuki*

Proceedings of 7th International Conference on Nuclear Engineering (ICONE-7) (CD-ROM), 10 Pages, 1999/00

no abstracts in English

JAEA Reports

Post irradiation examination data of high burnup PWR fuel rod; Rod No.:B15 (assembly No.:NO1G13)

*; Ishijima, Kiyomi; Yamahara, Takeshi

JAERI-Data/Code 98-002, 24 Pages, 1998/02

JAERI-Data-Code-98-002.pdf:1.02MB

no abstracts in English

Journal Articles

Report on international meeting 「International Seminar on Thermal Performance of (high burnup) LWR Fuel」

Nakamura, Jinichi

Kaku Nenryo, (29), P. 15, 1998/00

no abstracts in English

JAEA Reports

Behavior of stainless steel cladding fuel under a fast power transient condition; NSRR SC-1 test results

Katanishi, Shoji; Ishijima, Kiyomi; ; Kikuchi, Teruo;

JAERI-Research 94-039, 54 Pages, 1994/11

JAERI-Research-94-039.pdf:5.2MB

no abstracts in English

JAEA Reports

Investigation on Halden LWR ramp test by means of FEMAXI-III code(PWR version)

Nakamura, Jinichi; *; Furuta, Teruo; *

JAERI-M 91-027, 36 Pages, 1991/03

JAERI-M-91-027.pdf:1.03MB

no abstracts in English

JAEA Reports

Static strain measurement tests for the capsule used in the shock structural tests

Tanzawa, Sadamitsu; *; ; *; *

JAERI-M 90-232, 30 Pages, 1991/01

JAERI-M-90-232.pdf:0.85MB

no abstracts in English

Journal Articles

Reactor accident and simulation; Severe accident of a light water reactor

Soda, Kunihisa

Shimyureshon, 9(2), p.79 - 86, 1990/00

no abstracts in English

Journal Articles

LWR fuel safety research with paticular emphasis on RIA/LOCA and other conditions

Ichikawa, Michio; ; Kawasaki, Satoru

Journal of Nuclear Science and Technology, 26(1), p.118 - 125, 1989/01

no abstracts in English

Journal Articles

Thermal-hydraulics in uncovered core of light water reactor in severe core damage accident, IV; Analysis of core damage behavior in TMI-2 accident with SEFDAN code

; *

Journal of Nuclear Science and Technology, 24(1), p.12 - 22, 1987/01

 Times Cited Count:1 Percentile:80(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Analysis of power excursion event in Chernobyl accident with RETRAN code

; ; Hirano, Masashi;

Journal of Nuclear Science and Technology, 23(12), p.1107 - 1109, 1986/12

 Times Cited Count:2 Percentile:56.69(Nuclear Science & Technology)

no abstracts in English

Oral presentation

Fuel safety research at JAEA

Amaya, Masaki

no journal, , 

The objectives of the fuel safety research at JAEA are to evaluate the appropriateness of current Japanese regulatory criteria and their safety margins regarding light-water-reactor fuels, to provide the data for regulation for improved fuels which consist of cladding and fuel pellet with new materials, and to provide fuel behavior analysis codes which are applicable to regulatory activities. In this presentation, the outline of fuel safety research activities at JAEA is presented in addition to the progress in reactivity-initiated accident (RIA) and loss-of-coolant accident (LOCA) experiments.

Oral presentation

Development of technologies for reduction in volume and toxicity of high-level radioactive wastes, 2; Challenges in fast reactor cycle

Maeda, Seiichiro

no journal, , 

The uranium-plutonium mixed oxide (MOX) fuel bearing several percent minor actinides is adopted as the fast reactor fuel to reduce the high level radioactive waste volume and its toxicities. Their physical properties such as melting points and thermal conductivities are investigated symmetrically. It is getting clear that the effect of MA bearing is not so significant. The MOX fuel pins including up to 5% of Am were irradiated in Joyo to confirm their irradiation performance. Some types of irradiation experiments using MA-MOX fuels are planned after Joyo restart. The obtained data will contribute to develop fuel pin performance codes for MA-MOX fuels. SmART (Small Amount of Reuse Fuel Test) cycle program is in progress to demonstrate MA recycles starting from FR spent fuels. Furthermore, long life cladding with ODS steel is under development to enhance the transmutation efficiency of MA in one cycle. Thus, R&Ds in FR systems advances steadily to solve the waste issue.

Oral presentation

Fuel safety research at JAEA

Amaya, Masaki

no journal, , 

no abstracts in English

Oral presentation

Fuel safety research at JAEA

Amaya, Masaki

no journal, , 

no abstracts in English

23 (Records 1-20 displayed on this page)