Development of advanced automatic control system for nuclear ship, 2; Perfect automatic operation after reactor scram events

Yabuuchi, Noriaki; Nakazawa, Toshio; Takahashi, Hiroki; Shimazaki, Junya; Hoshi, Tsutao

JAERI-Tech 97-057, 54 Pages, 1997/11

JAERI-Tech-97-057.pdf:2.03MB

no abstracts in English

Progress report of the design study on a large reactor

; Hayashi, Hideyuki; ; ;

PNC TN9410 94-222, 355 Pages, 1994/07

PNC-TN9410-94-222.pdf:14.85MB

A design study on a large scale fast reactor was performed with focusing on enhancement of passive safety and capital cost reduction. The passive safety feature in the plant design of next generation fast reactors is one of the important subjects to be sought. In FY 1993, studies on 1300MWe class lage fast reactor were performed aiming at passive shutdown in a typical ATWS such as the unprotected loss of flow accident (ULOF). This report describes the core design, systems design, equipments design and the technical assessment in terms of the passive safety feature. It also includes the evaluation of the seismic as well as thermal capacity of reactor building of the large fast reactor.

None

; Kamide, Hideki

PNC TN9410 91-227, 16 Pages, 1991/07

PNC-TN9410-91-227.pdf:0.44MB

None

Reliability test on control rod driving mechanism of HTTR with HENDEL

Hino, Ryutaro; *;

Nihon Genshiryoku Gakkai-Shi, 33(7), p.685 - 694, 1991/07

https://doi.org/10.3327/jaesj.33.685

no abstracts in English

Reliability test on control rod driving mechanism of HTTR with HENDEL

Hino, Ryutaro; *;

JAERI-M 90-151, 36 Pages, 1990/09

JAERI-M-90-151.pdf:1.67MB

no abstracts in English

Fuel centerline temperature response of LWR fuel rods at reactor scram

Kawamura, Hiroshi; Ando, Hiroei

Nihon Genshiryoku Gakkai-Shi, 31(7), p.852 - 860, 1989/07

https://doi.org/10.3327/jaesj.31.852

no abstracts in English

Test Results and Their Analyses on Performance of JEFR Safety Rod Drive Mechanism

Nihon Genshiryoku Gakkai-Shi, 13(4), p.182 - 189, 1971/00

https://doi.org/10.3327/jaesj.13.182

no abstracts in English

Benchmark analysis of FFTF unprotected loss of flow without scram test No.13 with fast reactor plant dynamics analysis code Super-COPD

Hamase, Erina; Ohgama, Kazuya; Kawamura, Takumi*; Doda, Norihiro; Yamano, Hidemasa; Tanaka, Masaaki

no journal, , 

Validation of an analysis model for a plant dynamic analysis code named Super-COPD including neutronics calculation of a one-point reactor kinetics model necessitates the further work on the beyond design basis accident. Therefore, JAEA participated in IAEA benchmark for Loss of Flow without Scram (LOFWOS) test No.13 performed at the Fast Flux Test Facility (FFTF), and the transient analysis at the first blind phase considering with major reactivity feedback mechanisms was carried out. It was observed that the whole plant dynamics analysis could follow the measured data. As a future work, the gap conductance model for transient, the upper plenum of reactor vessel with dividing several regions or multi-dimensional modeling, and the core model that can evaluate the radial heat transfer rate more accurately will be refined.