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Journal Articles

Fission gas release from irradiated mixed-oxide fuel pellet during simulated reactivity-initiated accident conditions; Results of BZ-3 and BZ-4 tests

Kakiuchi, Kazuo; Udagawa, Yutaka; Amaya, Masaki

Annals of Nuclear Energy, 155, p.108171_1 - 108171_11, 2021/06

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Fracture-mechanics-based evaluation of failure limit on pre-cracked and hydrided Zircaloy-4 cladding tube under biaxial stress states

Li, F.; Mihara, Takeshi; Udagawa, Yutaka; Amaya, Masaki

Journal of Nuclear Science and Technology, 57(6), p.633 - 645, 2020/06

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Analytical study of SPERT-CDC test 859 using fuel performance codes FEMAXI-8 and RANNS

Taniguchi, Yoshinori; Udagawa, Yutaka; Amaya, Masaki

Annals of Nuclear Energy, 139, p.107188_1 - 107188_7, 2020/05

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

The Effect of base irradiation on failure behaviors of UO$$_{2}$$ and chromia-alumina additive fuels under simulated reactivity-initiated accidents; A Comparative analysis with FEMAXI-8

Udagawa, Yutaka; Mihara, Takeshi; Taniguchi, Yoshinori; Kakiuchi, Kazuo; Amaya, Masaki

Annals of Nuclear Energy, 139, p.107268_1 - 107268_9, 2020/05

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Behavior of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

Taniguchi, Yoshinori; Udagawa, Yutaka; Mihara, Takeshi; Amaya, Masaki; Kakiuchi, Kazuo

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.551 - 558, 2019/09

Journal Articles

Behavior of high-burnup advanced LWR fuels under design-basis accident conditions

Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Taniguchi, Yoshinori

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

JAEA Reports

The Evaluation of the influence of hydride rim and biaxial stress condition on the cladding failure under a reactivity-initiated-accident by using EDC test method

Shinozaki, Takashi; Mihara, Takeshi; Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki

JAEA-Research 2014-025, 34 Pages, 2014/12

JAEA-Research-2014-025.pdf:6.05MB

EDC test is a test method on the mechanical property of fuel cladding tube, and it focuses on the stress condition generated by PCMI under a RIA. We conducted EDC tests which simulate the mechanical conditions during a RIA by using the unirradiated cladding tubes which simulate hydride rim. Circumferential residual strains observed in post-test specimens tended to decrease with increasing the hydrogen concentration in the test cladding tubes and the thickness of the hydride rim. We also prepared RAG tube and performed EDC tests on it. It was observed that circumferential total strains at failure tended to decrease with increasing pre-crack depth on the outer surface of RAG tube specimen. We conducted biaxial stress tests by applying longitudinal tensile load onto RAG tube specimens. It was observed that circumferential total strains at failure under biaxial stress conditions tended to decrease compared to the results under uniaxial tensile condition.

JAEA Reports

Behavior of irradiated BWR fuel under reactivity-initiated-accident conditions; Results of tests FK-1, -2 and -3

Sugiyama, Tomoyuki; Nakamura, Takehiko; Kusagaya, Kazuyuki*; Sasajima, Hideo; Nagase, Fumihisa; Fuketa, Toyoshi

JAERI-Research 2003-033, 76 Pages, 2004/01

JAERI-Research-2003-033.pdf:17.46MB

Boiling water reactor (BWR) fuels with burnups of 41 to 45 GWd/tU were pulse-irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity-initiated-accident (RIA) conditions. BWR fuel segment rods of 8$$times$$8BJ (STEP I) type from Fukushima-Daiichi Unit 3 nuclear power plant were refabricated into short test rods, and they were subjected to prompt enthalpy insertion from 293 to 607 J/g (70 to 145 cal/g) within about 20 ms. The fuel cladding had enough ductility against the prompt deformation due to pellet cladding mechanical interaction. The plastic hoop strain reached 1.5% at the peak location. The cladding surface temperature locally reached about 600 deg C. Recovery of irradiation defects in the cladding due to high temperature during the pulse irradiation was indicated via X-ray diffractometry. Fission gas release during the pulse irradiation was from 3.1% to 8.2%, depending on the peak fuel enthalpy and the normal operation conditions.

Journal Articles

Failure thresholds of high burnup BWR fuel rods under RIA conditions

Nakamura, Takehiko*; Fuketa, Toyoshi; Sugiyama, Tomoyuki; Sasajima, Hideo

Journal of Nuclear Science and Technology, 41(1), p.37 - 43, 2004/01

 Times Cited Count:16 Percentile:72.41(Nuclear Science & Technology)

Transient deformation of high burnup BWR fuel rods was measured and failure limit was examined under simulated RIA conditions. Brittle cladding failure occurred at a small strain of about 0.4% during an early phase of the pulse irradiation tests at the NSRR. Strain rates were in an order of tens %/s at the time of the failure. Comparison of the results with thermal expansion of pellets suggested that the deformation was caused by thermal expansion of the pellets. In other words, the influence of fission gases in the pellets was small in the early phase of the deformation. Separate effect tests were conducted to examine influence of the cladding temperature on the failure behavior of cladding. Influence of the pulse width on the failure threshold was discussed in terms of the strain rate, magnitude of the deformation and temperature of the cladding for high burnup BWR fuel rods under RIA conditions.

Journal Articles

Phenomenon identification and ranking tables (PIRTs) for rod ejection accidents in pressurized water reactors containing high burnup fuel

Boyack, B. E.*; Motta, A. T.*; Peddicord, K. L.*; Alexander, C. A.*; Deveney, R. C.*; Dunn, B. M.*; Fuketa, Toyoshi; Higar, K. E.*; Hochreiter, L. E.*; Langenbuch, S.*; et al.

NUREG/CR-6742, 263 Pages, 2001/09

no abstracts in English

Journal Articles

Study of accident scenario ofChernobyl power reactor

; *; ; *;

Genshiryoku Kogyo, 32(12), p.17 - 31, 1986/00

no abstracts in English

JAEA Reports

Semiannual Progress Report on the NSRR Experiments,8

JAERI-M 8779, 75 Pages, 1980/03

JAERI-M-8779.pdf:2.86MB

no abstracts in English

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