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Sato, Ikken
Nuclear Engineering and Design, 383, p.111426_1 - 111426_19, 2021/11
Times Cited Count：7 Percentile：64.17(Nuclear Science & Technology)Takeda, Takeshi
JAEA-Data/Code 2020-019, 58 Pages, 2021/01
JAEA-Data-Code-2020-019.pdf:3.85MB
An experiment denoted as SB-SL-01 was conducted on March 27, 1990 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-IV (ROSA-IV) Program. The ROSA/LSTF experiment SB-SL-01 simulated a main steam line break (MSLB) accident in a pressurized water reactor (PWR). The test assumptions were made such as auxiliary feedwater (AFW) injection into secondary-side of both steam generators (SGs) and coolant injection from high pressure injection (HPI) system of emergency core cooling system into cold legs in both loops. The MSLB led to a fast depressurization of broken SG, which caused a decrease in the broken SG secondary-side wide-range liquid level. The broken SG secondary-side wide-range liquid level recovered because of the AFW injection into the broken SG secondary-side. The primary pressure temporarily decreased a little just after the MSLB, and increased up to 16.1 MPa following the closure of the SG main steam isolation valves. Coolant was manually injected from the HPI system into cold legs in both loops a few minutes after the primary pressure reduced to below 10 MPa. The primary pressure raised due to the HPI coolant injection, but was kept at less than 16.2 MPa by fully opening a power-operated relief valve of pressurizer. The core was filled with subcooled liquid through the experiment. Thermal stratification was seen in intact loop cold leg during the HPI coolant injection owing to the flow stagnation. On the other hand, significant natural circulation prevailed in broken loop. When the continuous core cooling was ensured by the successive coolant injection from the HPI system, the experiment was terminated. The experimental data obtained would be useful to consider recovery actions and procedures in the multiple fault accident with the MSLB of PWR. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment SB-SL-01.
Takeda, Takeshi
JAEA-Data/Code 2016-004, 59 Pages, 2016/07
JAEA-Data-Code-2016-004.pdf:3.34MB
The TR-LF-07 test simulated a loss-of-feedwater transient in a PWR. A SI signal was generated when steam generator (SG) secondary-side collapsed liquid level decreased to 3 m. Primary depressurization was initiated by fully opening a power-operated relief valve (PORV) of pressurizer (PZR) 30 min after the SI signal. High pressure injection (HPI) system was started in loop with PZR 12 s after the SI signal, while it was initiated in loop without PZR when the primary pressure decreased to 10.7 MPa. The primary and SG secondary pressures were kept almost constant because of cycle opening of the PZR PORV and SG relief valves. The PZR liquid level began to drop steeply following the PORV full opening, which caused liquid level formation at the hot leg. The primary pressure became lower than the SG secondary pressure, which resulted in the actuation of accumulator (ACC) system in both loops. The primary feed-and-bleed operation was effective to core cooling because of no core uncovery.
Takada, Shoji; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Seki, Tomokazu; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Sawa, Kazuhiro
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to demonstrate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. The local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1
C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.
Nakamura, Hirofumi; Nishi, Masataka
Journal of Nuclear Materials, 329-333(Part1), p.183 - 187, 2004/08
Times Cited Count：27 Percentile：82.07(Materials Science, Multidisciplinary)no abstracts in English
Kai, Tetsuya; Harada, Masahide; Maekawa, Fujio; Teshigawara, Makoto; Konno, Chikara; Ikeda, Yujiro
Journal of Nuclear Science and Technology, 41(Suppl.4), p.172 - 175, 2004/03
In J-PARC neutron source, intense protons (3 GeV,1 MW) pass through a proton-beam window and bombard a Hg target in a target-moderator-reflector-assembly (TMRA). The SS316 target chamber is the most highly activated. Decouplers (Ag-In-Cd (AIC) alloy) are also highly activated. Some neutron extraction holes of Be and AL-coated iron reflector are lined with AIC alloy. A SS316 shield is located outer the TMRA. All these components are cooled by H
O or DO. We estimated the induced-radioactivity of the TMRA components and the cooling water using NMTC/JAM, MCNP4 and DCHAIN-SP. As results, the remote maintenance and massive shields were indispensable. For example, a 30 cm thick Fe cask for the reflector assembly was necessary to attenuate the radiation less than 1 mSv/h. The cask required a 130-ton crane. The AL-coated Fe of the reflector was adopted instead of SS316 resulting in eliminating the high activity of Ni in SS316 and reduction of the cask weight. Based on these results, shielding wall designs and maintenance scenarios of the highly activated components are developed.
Hanawa, Satoshi; Tachibana, Yukio; Iyoku, Tatsuo; Ishihara, Masahiro; Ito, Haruhiko
JAERI-Tech 2003-064, 25 Pages, 2003/07
JAERI-Tech-2003-064.pdf:2.84MB
On the 147cycle operation, the water leakage was found at the pressure instrumentation pipe which is attached to the exit pipe of No.1 charge pump of the purification system of primary cooling system at JMTR in the Oarai establishment, JAERI. Then JMTR was shutted down manually on December 10th. It was predicted that the crack on the pressure instrumentation pipe was initiated and propagated by the cyclic load which was caused by the charge pump. Therefore, vibration and stress analyses of pressure instrumentation pipe were performed. From the vibration analysis, the natural frequency of the pressure instrumentation pipe of No.1 charge pump is between 53
58Hz, which is close to the resonance frequency of 50Hz. From the stress analysis results, total stress generated on the pressure instrumentation pipe is 112.2MPa at the natural frequency of 53Hz and 74.2Mpa at 58Hz. It was found that the stress of 112.2MPa is close to the fatigue limit of used materials.
Working Group for Investigation of Cause of Crack Initiation
JAERI-Tech 2003-060, 183 Pages, 2003/07
JAERI-Tech-2003-060.pdf:55.37MB
On December 10, 2002, the water leakage was found at the pressure instrumentation pipe attached to the exit pipe of No.1 charging pump of the purification system of a primary cooling system at JMTR, and the cracks were detected on the pressure instrumentation pipe by the visual observation. The Investigation Committee for Water Leakage from Instrumentation Pipe in JMTR was established and organized by specialists from inside and outside JAERI on December 16. In order to investigate the cause of crack initiation at the pressure instrumentation pipe, the Working Group was organized in the Department of JMTR. Visual inspection, fractgraphy test, metallographic observation and hardness test for the pressure instrumentation pipe and its weldment were carried out in the JMTR Hot Laboratory. This report summarized above data obtained by investigation on the cause of the crack initiation.
Hori, Junichi; Sato, Satoshi; Yamauchi, Michinori*; Ochiai, Kentaro; Nishitani, Takeo
JAERI-Research 2003-002, 50 Pages, 2003/03
JAERI-Research-2003-002.pdf:2.21MB
D-T neutron irradiation experiments have been performed with F82H and ODS ferritic steels and the effective cross sections for 
Co productions in those materials via the sequential reactions were measured. The effective cross sections for F82H and ODS ferritic steels were about 1.5 times larger than that for iron. The distributions of effective cross sections were measured for 6 materials (iron, copper, vanadium, titanium, tungsten and lead) and F82H. The sequential reaction rates in the region close to hydrogen compound became over 20 times larger than that in material itself. In the case of F82H, the increase ratio was about 50. It was indicated that the activity for the sequential reaction product Co will reach to 3-10 
of that for primary neutron reaction product 
Mn aound the surface of a cooling pipe in a fusion reactor. The effective cross sections were estimated by using (n,xp), (p,n) reaction cross sections, proton emission spectra, proton stopping power in the material. The estimated values were compared with experimental results.
Enoeda, Mikio; Kuroda, Toshimasa*; Moriyama, Koichi*; Ohara, Yoshihiro
Journal of Nuclear Science and Technology, 38(11), p.921 - 929, 2001/11
Times Cited Count：2 Percentile：19.28(Nuclear Science & Technology)Test module testing in ITER is one of the most important mile-stone for development of the DEMO blanket. In the design of test modules in ITER, it is very important to show that test modules do not cause additional safety concern to ITER. This work has been performed for the evaluation of the substantial safety of Test Module of Water Cooled Solid Blanket, which is the current candidate blanket for the DEMO blanket in Japan. Major issues of the evaluation were establishment of post accident cooling in TM, hydrogen gas generation by Be-steam reaction, and pressure increase and spilled water amount by Loss of Coolant Accident (LOCA) event. The evaluation was performed to derive the upper bound of consequences in significant events, of which scenario can be assumed by the similarity of the safety analysis of Shielding Blanket.
Okumura, Susumu; Arakawa, Kazuo; Fukuda, Mitsuhiro; Nakamura, Yoshiteru; Yokota, Wataru; Ishimoto, Takayuki*; Kurashima, Satoshi; Ishibori, Ikuo; Nara, Takayuki; Agematsu, Takashi; et al.
AIP Conference Proceedings 600, p.330 - 332, 2001/00
Frequent corrections of the magnetic field of the JAERI AVF cyclotron were required for keeping a beam current constant during long time operation. We observed correlation between the magnetic field and the temperature of the cyclotron magnet yoke by measuring the magnetic field with an NMR probe and the temperature with platinum resistance thermometers. The unstable phenomenon of a cyclotron beam was induced by temperature change in the magnet yoke caused mainly by thermal conduction from the main coil. To restrain the thermal conduction to the yoke, we have inserted temperature controlled copper plates between the yoke and the main coil. In addition, a temperature control system of the cooling water of the trim coils has been installed independent of the total cooling system for controlling the pole tip temperature. An optimum condition of the temperature control systems for stabilizing the magnetic field has been investigated.
; ;
JAERI-Tech 98-052, 69 Pages, 1998/11
no abstracts in English
Miura, H.*; Sato, Satoshi; Enoeda, Mikio; Kuroda, Toshimasa*; Takatsu, Hideyuki; Kawamura, Yoshinori; Tanaka, Satoru*
JAERI-Tech 97-051, 51 Pages, 1997/10
no abstracts in English
Takada, Shoji; Suzuki, Kunihiro; Inagaki, Yoshiyuki; Sudo, Yukio
Heat Transfer-Jpn. Res., 26(3), p.159 - 175, 1997/00
no abstracts in English
Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Watanabe, Hironori
Transactions of the American Nuclear Society, 69, p.539 - 540, 1993/00
no abstracts in English
Yamamoto, Katsumune; Yokouchi, Iichiro; ; ;
Nihon Genshiryoku Gakkai-Shi, 29(8), p.717 - 723, 1987/08
Times Cited Count：1 Percentile：19.15(Nuclear Science & Technology)no abstracts in English
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JAERI-M 86-158, 58 Pages, 1986/11
no abstracts in English
; Tasaka, Kanji; ; ; ; ; ; Koizumi, Yasuo
JAERI-M 86-038, 275 Pages, 1986/03
no abstracts in English
Yamamoto, Katsumune; ; ; Yokouchi, Iichiro; ;
Nihon Genshiryoku Gakkai-Shi, 28(5), p.425 - 427, 1986/00
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)no abstracts in English
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Proc.2nd Int.Topical Meeting on Nuclear Power Plant Thermal Hydraulics and Operations, p.2 - 129, 1986/00
no abstracts in English
