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JAEA Reports

Report of TRACY operation

Aizawa, Eiju; Ogawa, Kazuhiko; Sakuraba, Koichi; Tsukamoto, Michio; Sugawara, Susumu; Takeuchi, Masaki*; Miyauchi, Masakatsu; Yanagisawa, Hiroshi; Ono, Akio

JAERI-Tech 2002-031, 120 Pages, 2002/03

JAERI-Tech-2002-031.pdf:4.32MB

TRACY (Transient Experiment Critical Facility) in NUCEF (Nuclear Safety Research Facility) is the pulse-type critical facility using uranyl nitrate solution which can carry out various supercritical experiments changing reactivity addition up to 3$.TRACY achieved its first criticality on 20th December 1995,and transient operations have been conducted Since1996.This report summarizes the operation data of 176 experiments from the first criticality to FY2000.

Journal Articles

High power transient characteristics and capability of NSRR

Nakamura, Takehiko; Katanishi, Shoji; Kashima, Yoichi; Yachi, Shigeyasu; Yoshinaga, Makio; Terakado, Yoshibumi

Journal of Nuclear Science and Technology, 39(3), p.264 - 272, 2002/03

https://doi.org/10.3327/jnst.39.264

Times Cited Count：9 Percentile：50.42(Nuclear Science & Technology)

In order to study fuel behavior under abnormal transients and accidents, the control system of the Nuclear Safety Research Reactor (NSRR) of the Japan Atomic Energy Research Institute (JAERI) was modified to achieve high power transients. With this new operational mode, called Shaped Pulse (SP), transients at the maximum power of 10 MW can be conducted for a few seconds. This new operational mode supplements the previous Natural Pulse (NP) operation at the maximum power of 23 GW for milliseconds. For high power transient operation, a simulator using a point kinetic model was developed, and characteristics of the NSRR in the new operational mode were examined through tests and calculations. With the new operational mode, new types of fuel irradiation tests simulating power oscillations of boiling water reactors (BWRs) can to be conducted in the NSRR. Reactor characteristics and capability, such as control rod worth, feedback reactivity, and operational limits of the NSRR for SP operations are discussed.

JAEA Reports

Outline of operation and control system and analytical investigation of transient behavior of an out-of-pile hydrogen production system for HTTR heat utilization system

Inagaki, Yoshiyuki; Hada, Kazuhiko; Nishihara, Tetsuo; Takeda, Tetsuaki; Hino, Ryutaro; Haga, Katsuhiro

JAERI-Tech 97-050, 125 Pages, 1997/10

JAERI-Tech-97-050.pdf:2.96MB

no abstracts in English

JAEA Reports

Decrease in coolability events analysis for the safety assessment of JRR-3 silicide core by THYDE-W code

Kaminaga, Masanori; Yamamoto, Kazuyoshi

JAERI-Tech 97-016, 120 Pages, 1997/03

JAERI-Tech-97-016.pdf:3.76MB

no abstracts in English

JAEA Reports

Reactivity initiated events analysis for the safety assessment of JRR-3 silicide core by EUREKA-2 code

Kaminaga, Masanori

JAERI-Tech 97-014, 125 Pages, 1997/03

JAERI-Tech-97-014.pdf:4.04MB

no abstracts in English

JAEA Reports

Reactivity initiated events analysis for the safety assessment of JRR-4 silicide LEU core

Kaminaga, Masanori; Yamamoto, Kazuyoshi; Watanabe, Shukichi; Nakano, Yoshihiro

JAERI-Tech 95-040, 79 Pages, 1995/07

JAERI-Tech-95-040.pdf:2.25MB

no abstracts in English

JAEA Reports

Safety analysis of JMTR-LEU cores, 1; Reactivity initiated accident analysis

Nagaoka, Yoshiharu; Komukai, Bunsaku; ; Saito, Minoru;

JAERI-M 92-095, 68 Pages, 1992/07

JAERI-M-92-095.pdf:1.52MB

no abstracts in English

JAEA Reports

Feedback control of primary circulation pump of PIUS-type reactor

; Anoda, Yoshinari; Murata, Hideo; Yonomoto, Taisuke; ; Kukita, Yutaka

JAERI-M 91-076, 34 Pages, 1991/05

JAERI-M-91-076.pdf:1.02MB

no abstracts in English

Journal Articles

Safety characteristics of the High Temperature Engineering Test Reactor

Shindo, Masami; ; Kunitomi, Kazuhiko; ; Sawa, Kazuhiro

Nucl. Eng. Des., 132, p.39 - 45, 1991/00

https://doi.org/10.1016/0029-5493(91)90293-Q

Times Cited Count：5 Percentile：53.46(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Feedback control of primary circulation pump of PIUS-type reactor during startup and steady state operation

; Anoda, Yoshinari; Murata, Hideo; Yonomoto, Taisuke; Kukita, Yutaka;

Thermal Hydraulics of Advanced Nuclear Reactors, p.85 - 89, 1990/11

no abstracts in English

Journal Articles

ROSA-IV large scale test facility(LSTF); Test program and first look of test results

Tasaka, Kanji; Koizumi, Yasuo

Nihon Genshiryoku Gakkai-Shi, 29(1), p.18 - 30, 1987/01

https://doi.org/10.3327/jaesj.29.18

Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)

The Large Scal Test Facility (LSTF) of the Rig of Safety Assessment (ROSA)-IV program is an integral test fcility to investigate thermal-hydraulic response of a pressurized water reactor (PWR) system during small break loss-of-coolant accidents (LOCAs) and operational transients.

Journal Articles