Validation of core cooling capability analysis in Monju during guillotine pipe break at primary heat transport system

Yamada, Fumiaki; Arikawa, Mitsuhiro*; Fukano, Yoshitaka

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

In sodium-cooled fast reactor, since the coolant does not need to be pressurized, a pipe break due to the internal pressure does not occur physically. For safety margin in Japanese prototype fast breeder reactor (Monju), the guillotine pipe break accident, i.e., loss of integrity (LOPI) has been analyzed as an extreme assumption for beyond design basis accidents (B-DBAs) in the licensing application for the permit. The cooling capability of the core was re-evaluated in this paper during a large-scale, more specifically guillotine pipe break at the primary heat transport system (PHTS) in Monju, newly considering the following latest findings: (a) Experimental data on sodium boiling in fuel assemblies, (b) Actual PHTS pump coast-down characteristics, and (c) Transient burst test data on irradiated fuel claddings. The analysis models were the validated and simulations were re-performed also using the actual Monju data such as the response time to the trip signals, etc. As a result, it was clarified that the ratio of failed fuel claddings does not exceed around 3% of all of fuel assemblies, as in the past licensing analysis. The safety has been reconfirmed to be secured without significant core damage even under an extreme assumption of a double-ended guillotine pipe break at the PHTS in Monju.

Evaluation method of performance of siphon break value as core covering system for water-cooled test and research reactors

; Kumada, Hiroaki; Kaminaga, Fumito*

Nihon Genshiryoku Gakkai-Shi, 42(4), p.325 - 333, 2000/04

https://doi.org/10.3327/jaesj.42.325

no abstracts in English

Flow regime transition in high-pressure large-diameter horizontal two-phase flow

Anoda, Yoshinari; Kukita, Yutaka; Nakamura, Hideo; Tasaka, Kanji

Proc. on 1989 National Heat Transfer Conf., Vol. 4, 8 Pages, 1989/00

no abstracts in English

Thermal-Hydrauluc Analysis of Loss-of-Coolant Accident in the JMTR

;

JAERI-M 85-001, 33 Pages, 1985/02

JAERI-M-85-001.pdf:0.74MB

no abstracts in English

Verification study on alternative ECCS concepts for a PWR

; ; ; ;

Nucl.Eng.Des., 54(3), p.419 - 427, 1979/00

https://doi.org/10.1016/0029-5493(79)90184-5

no abstracts in English

ROSA-II Test Date Report,7;Runs 318,320,321,322,323

*

JAERI-M 7106, 181 Pages, 1977/06

JAERI-M-7106.pdf:3.91MB

no abstracts in English

ROSA-II Test Data Report,2; Run 307,308,309

*

JAERI-M 6241, 75 Pages, 1975/09

JAERI-M-6241.pdf:1.56MB

no abstracts in English