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Journal Articles

Behavior of fuel with zirconium alloy cladding in reactivity-initiated accident and loss-of-coolant accident

Fuketa, Toyoshi*; Nagase, Fumihisa

Zirconium in the Nuclear Industry; 18th International Symposium (ASTM STP 1597), p.52 - 92, 2018/01

Extensive research programs have been performed for more than two decades in JAEA and a better understanding has been developed for fuel behavior under accident conditions. The program is comprised of: RIA studies including pulse-irradiation experiments in the NSRR, cladding mechanical tests, and development and verification of a computer code RANNS; LOCA tests including integral thermal shock tests, oxidation rate measurements, and cladding mechanical tests; development and verification of a computer code FEMAXI-6, etc. Data and findings from the research programs provided technical basis directly and indirectly for regulatory criteria in Japan and other countries. This paper reviews and summarizes the major outcome from the research programs and identifies further research needs, as the acceptance technical paper for the Kroll Medal award of ASTM.

Journal Articles

Properties of an Irradiated Heat-Treated Zr-2.5Nb Pressure Tube Removed From the NPD Reactor

Koike, Mitsutaka; Colema, C. E.*; Causey, A. R.*; Ells, C. E.*; Hosbon, R. R.*

Zirconium in the Nuclear Industry; 11th International Symposium (ASTM STP 1295), p.469 - 491, 1998/00

None

Journal Articles

Estimation of conservatism of present embrittlement criteria for zircaloy fuel cladding under LOCA

; ;

Zirconium in the Nuclear Industry, p.734 - 746, 1984/00

no abstracts in English

Journal Articles

None

Koike, Mitsutaka; ; Nagamatsu, Kenji; Shibahara, Itaru

10th International Symposium on Zirconium in the Nuclear Industry, , 

None

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