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Shiota, Yuki; Ariyoshi, Hideo; Shiohama, Yasutaka; Isobe, Yuta; Takeuchi, Ryotaro; Kudo, Junki; Hanaki, Shotaro; Hamano, Tomoharu; Takagi, Tsuyohiko
JAEA-Technology 2022-019, 95 Pages, 2022/09
JAEA-Technology-2022-019.pdf:7.59MB
In the first stage of "Monju" decommissioning project, "Fuel Unloading Operations" have been carrying out. The operations consists of two processes. The first process is "Fuel Treatment and Storage" is that the fuel assemblies unloaded from the Ex-Vessel fuel Storage Tank (EVST) are canned after sodium cleaning, and then transferred to the storage pool. The second process is "Fuel Unloading" that the fuel assemblies in the reactor core are replaced with dummy fuel assemblies and stored in the EVST. "Fuel Treatment and Storage" and "Fuel Unloading" are performed alternately until 370 fuel assemblies in the core and 160 fuel assemblies in the EVST are all transferred to the storage pool. This is a summary of "Fuel Unloading" in the third quarter of "Fuel Unloading Operation". In fiscal 2020, as "Fuel Unloading", 72 fuel assemblies and 74 blanket fuel assemblies were unloaded from the core, and stored in the EVST. From the EVST, 145 dummy fuel assemblies and 1 fixed absorber were loaded in the core instead. During these operations, a total of 36 cases alarming or equipment malfunctions classified into 4 types occurred. However, these events were estimated in advance, there were no significant events that menaces to safety of fuel assemblies and equipment. Therefore, there were no serious problem like fall of fuel assemblies and events that may affect schedule of the project like stick of gripper of ex-vessel fuel transfer machine. When equipment's work or performance fail, the operation continued with safety by elimination of causes of problem. Fuel handling system of Monju has function that is endemic to sodium cooling fast breeding reactor. Because continuous operations of fuel handling system with actual fuel assemblies start recently, we don't have as much experience as PWR and BWR. With estimation of various troubles, reduction of frequency of trouble occurrence and minimization of impacts on schedule performed.
Koga, Kazuhiro*; Suzuki, Kazunori*; Takagi, Tsuyohiko; Hamano, Tomoharu
FAPIG, (196), p.8 - 15, 2020/01
The prototype fast breeder reactor Monju has already started (from June 2017) the unloading operation period (about 5.5 years: until the end of 2022) of the fuel assembly, which is the first stage of decommission. Among them, the first "Processing of fuel assembly" operation (86 in total) was conducted from August 2018 to January 2019 as the first handling of the fuel assembly. Fuji Electric provided technical support, such as dispatching technicians throughout the period, in cooperation with Japan Atomic Energy Agency for the "Processing of fuel assembly" operation, and contributed to the completion of the operation while experiencing various troubles. This manuscript introduces the contents of the first "Processing of fuel assembly" operation and the overview of the trouble status. This manuscript is a sequel to FAPIG No.194 "Prototype Fast Breeder Reactor Monju Decommissioning and Unloading Operation of the Fuel Assembly from the Core", please refer to it.
Ohshima, Hiroyuki; Sakai, Takaaki; ; Yamaguchi, Akira; Nishi, Yoshihisa*; Ueda, Nobuyuki*; *
JNC TN9400 2000-077, 223 Pages, 1999/05
JNC-TN9400-2000-077.pdf:6.24MB
The feasibility study (Phase l) is being carried out at JNC to build up new design concepts of practical fast reactors (FRs) from the viewpoint of economy, safety, effective use of resources, reduction of environmental burden and non-proliferation. This report describes the results of the investigation, related to decay heat removal, core/fuel-assembly thermal-hydraulics and thermal-hydraulic correlations, that was performed in fiscal l999 as a part of the feasibility study. ln the study of the decay heat removal, the effects of several design parameters on the performance of the reactor vessel auxiliary cooling system (RVACS) in a middle-scale sodium-cooled FR were clarified by using a plant dynamic analysis code. The upper limit of RVACS performance was preliminarily estimated at approximately 0.5
0.6 MWe. Numerical methods for the plant dynamic analysis of gas-and heavy-metal-cooled FRs were also developed. They were applied to the preliminary calculations of the transition from scram to natural circulation and the transient characteristics in tentative plant design concepts were clarified. ln addition, a dimensionless number indicating natural circulation performance was deduced for the comparison of several plant design concepts. With respect to the core/fuel-assembly thermal-hydraulics, numerical analysis methods were improved for the pin-type fuel assembly of gas-and heavy-metal-cooled FRs, the coated-particle- type fuel assembly of helium-gas-cooled FR, and the ductless core of sodium-and heavy-metal-cooled FRs. As preliminary evaluations, thermal-hydraulics in the heavy-metal-cooled FR fuel assembly was compared with sodium-cooled one and thermal-hydraulic analyses of carbon-dioxide- and helium-gas-cooled FR fuel assemblies were performed. The analysis for the fuel assembly with inside duct of sodium-cooled FR was also carried out. The correlations of pressure loss and heat transfer coefficient were investigated for the thermal-hydraulic ...
Hunter
PNC TN9410 97-057, 106 Pages, 1997/05
This study was based on a 'pancake' type fast reactor core design of 600 MW(e), which had been optimized for Pu burning with a feed Pu vector appropriate to once-through irradiation of MOX fuel in a PWR. The purpose of the study was to investigate the effects of varying the Pu vector, examining various methods of offsetting the effects of such a change, and finally to produce fuel cycles optimized for the different qualities of Pu vector within the same basic design. In addition to the reference (once-through) Pu vector, two extreme Pu vectors were examined: high quality Pu from military stockpiles; low quality Pu corresponding to the equilibrium point of multiple recycling in a Pu burning fast reactor. Variations in Pu quality were overcome by changing the fuel inventory - replacing some of the fuel by diluent material, and altering the fuel pin size. Using absorber material (
B
C) as diluent improves the rod worth shutdown margin but degrades the Na void and Doppler safety parameters, a non-absorber diluent has the opposite effects, so a mix of the 2 material types was used to optimize the core characteristics. Of the non-absorber diluent materials examined, ZrH gave significantly better performance than all others; 
BC was the second choice for non-absorber diluent, because of its compatibility with BC absorber. It was not possible to accommodate the lower quality (multi-recycled) Pu vector without a significant increase in the fuel pin volume. It was not generally possible, especially with the increased fuel pin size, to achieve positive rod worth shutdown margins - this was overcome by increasing the number of control rods. For the higher quality Pu vectors to maintain ratings within limits, it was necessary to adopt hollow fuel pellets, or else to use the diluent material as an inert matrix in the fuel pellets. It proved possible to accommodate both extremes of Pu vector within a single basic design, maintaining ...
Murata, Isao*; Mori, Takamasa; Nakakawa, Masayuki; *
JAERI-Data/Code 96-016, 79 Pages, 1996/03
JAERI-Data-Code-96-016.pdf:2.81MB
no abstracts in English
; Hino, Ryutaro; Inagaki, Yoshiyuki; Takase, Kazuyuki; Ioka, Ikuo; Takada, Shoji; Suzuki, Kunihiro; Kunitomi, Kazuhiko; Maruyama, So;
JAERI 1333, 196 Pages, 1995/03
no abstracts in English
Takase, Kazuyuki; Akino, Norio
Proc. of the 30th Intersociety Energy Conversion Engineering Conf., 0, P. 95_169, 1995/00
no abstracts in English
Nomura, Yasushi; Okuno, Hiroshi
Nihon Genshiryoku Gakkai-Shi, 35(2), p.155 - 163, 1993/02
no abstracts in English
Ono, Akio; Osugi, Toshitaka;
JAERI-M 91-186, 63 Pages, 1991/11
no abstracts in English
Hino, Ryutaro; Takase, Kazuyuki;
Nihon Genshiryoku Gakkai-Shi, 32(10), p.996 - 998, 1990/10
Times Cited Count：0 Percentile：0.02(Nuclear Science & Technology)no abstracts in English
test section, V; Channel blockage test result on HTTR fuel columnHino, Ryutaro; Takase, Kazuyuki;
JAERI-M 90-163, 45 Pages, 1990/09
no abstracts in English
test section, Part IV; Crossflow test in parallel gapTakase, Kazuyuki; Hino, Ryutaro;
JAERI-M 90-118, 34 Pages, 1990/08
no abstracts in English
Sawa, Kazuhiro; *; *; Shiozawa, Shusaku
JAERI-M 90-063, 42 Pages, 1990/03
no abstracts in English
test section, II; Test results in helium gas conditions heated to 1000
CHino, Ryutaro; Takase, Kazuyuki; Maruyama, So;
JAERI-M 90-032, 46 Pages, 1990/03
no abstracts in English
CHino, Ryutaro; Maruyama, So; Takase, Kazuyuki; ; Shimomura, Hiroaki
Nihon Genshiryoku Gakkai-Shi, 31(4), p.470 - 476, 1989/04
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
; Akino, Norio; Ogawa, Masuro; ; ; Sanokawa, Konomo; Okamoto, Yoshizo
Kikai No Kenkyu, 39(1), p.154 - 160, 1987/01
no abstracts in English
), I; Test results of a single-channel mock-up fuel rod with uniform power distribution in the axial direction; ; ; ; ;
Nihon Genshiryoku Gakkai-Shi, 28(5), p.428 - 435, 1986/00
Times Cited Count：4 Percentile：48.02(Nuclear Science & Technology)no abstracts in English
;
JAERI-M 85-145, 47 Pages, 1985/09
no abstracts in English
):(1st Report,Test Results of a Fuel Rod with Uniform Power Distribution in the Axial Direction.); ; ; ; ; ;
JAERI-M 85-084, 41 Pages, 1985/06
no abstracts in English
Test Section,1; Uniform Power Distribution Test Conducted with Moderate-Temperature Heater Rod; ; ; ; ; ;
JAERI-M 85-067, 29 Pages, 1985/06
no abstracts in English
