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JNC TN9410 2000-010, 72 Pages, 2000/03
JNC-TN9410-2000-010.pdf:2.14MB
The experimental fast reactor JOYO served as the MK-II irradiation bed core for testing fuel and material for FBR development for 16 years from 1982 to 1997. During the MK-II core operation, extensive data were accumulated from the plant characteristic tests. Tests conducted at JOYO included operating characteristic tests for confirming operational safety, performance tests for confirming design performance of the MK-II core, and special tests for research and development ofthe plant. In this report, the outline and the results of each test item are shown. These test data can be provided by the magnet-optical disk.
; ; Saikawa, Takuya*; Sukegawa, Kazuya*
JNC TN9410 2000-008, 66 Pages, 2000/03
JNC-TN9410-2000-008.pdf:1.39MB
The experimental fast reactor "JOYO" served as the MK-II irradiation bed core for testing fuel and material for FBR development for 15 years from 1982 to 1997. During the MK-II operation, impurities concentrations in the sodium and the argon gas were determined by 67 samples of primary sodium, 81 samples of secondary sodium, 75 samples of primary argon gas, 89 samples of secondary argon gas (the overflow tank) and 89 samples of secondary argon gas (the dump tank). The sodium and the argon gas purity control data were accumulated from in thirty-one duty operations, thirteen special test operations and eight annual inspections. These purity control results and related plant data were compiled into database, which were recorded on CD-ROM for user convenience. Purity control data include concentration of oxygen, carbon, hydrogen, nitrogen, chlorine, iron, nickel and chromium in sodium, concentration of oxygen, hydrogen, nitrogen, carbon monoxide, carbon dioxide, methane and helium in argon gas with the reactor condition.
Tsukimori, Kazuyuki; Furuhashi, Ichiro*
JNC TN9400 2000-049, 93 Pages, 2000/03
JNC-TN9400-2000-049.pdf:2.82MB
lt is one of the important key points to reduce thermal stress of the primary piping system in the design of sodium coolant loop-type FBR plants. The objectives of this study are to understand the characteristics of the thermal stresses in the simple S-shaped hot leg piping systems which run from the outlet nozzle of the reactor vessel (R/V) to the inlet nozzle of the intermediate heat exchanger (IHX), and to propose some recommendable routings of piping systems. Results are summarized as follows. (1)Generally, the thermal stresses in elbows are severer than those at nozzles. The tendency was observed that the stress in elbow decreases with the increase of the distance between the outlet nozzle of R/V and the inlet nozzle of IHX and also the distance between the outlet nozzle of R/V and the liquid surface level. (2)lt is expected to reduce thermal stresses in elbow to big extent by adopting super 90 degree elbows. Therefore, in these cases the dimension region which satisfies the allowable stress is broad compared with that in the case of the conventional 90 degree elbow. (3)The stress estimations in elbow based on 'MITl notice No.501' become excessively large compared with the results by FEA using shell elements, when the maximum stress occurs at the end of elbow. ln these cases, the estimation can be rationalized by replacing the maximum stress by the mean of stresses at the end and at the middle of the elbow. (4)Two routings with 105 degree elbows are recommended. 0ne has the advantage from the view point of reduction of length of pipe and the other does from the view point of reduction of thermal stresses, compared with the routing with 90 degree elbows.
; Hada, Kazuhiko; Koikegami, Hajime*; Kisamori, Hiroyuki*
JAERI-Tech 96-042, 41 Pages, 1996/10
no abstracts in English
JAERI-Tech 96-006, 86 Pages, 1996/02
no abstracts in English
Kaji, Yoshiyuki; Ioka, Ikuo;
The 3rd JSME/ASME Joint Int. Conf. on Nuclear Engineering (ICONE),Vol. 1, 0, p.363 - 368, 1995/00
no abstracts in English
Ioka, Ikuo; Kaji, Yoshiyuki; Terunuma, Isao*; Nekoya, Shinichi;
JAERI-Data/Code 94-010, 60 Pages, 1994/09
JAERI-Data-Code-94-010.pdf:3.76MB
no abstracts in English
Mogi, Haruyoshi; Tanaka, Toshiyuki; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru
Doryoku, Enerugi Gijutsu No Saizensen : Shimpojiumu Koen Rombunshu 1994, 0, p.305 - 310, 1994/00
no abstracts in English
Nihon Kikai Gakkai Dai-72-Ki Zenkoku Taikai Koen Rombunshu, Vol.III, 0, p.649 - 651, 1994/00
no abstracts in English
Kaji, Yoshiyuki;
Proc. of the 12th Int. Conf. on Structural Mechanics in Reactor Technology,Vol. L; SMiRT 12, p.129 - 134, 1993/00
no abstracts in English
Kimura, Hidetaka; *; *; Kawasaki, Hirotsugu; Aoto, Kazumi;
PNC TN9450 91-003, 28 Pages, 1991/03
None
Kondo, Tatsuo
Nihon Tekko Kyokai Nishiyama Kinen Gijutsu Koza, p.247 - 276, 1990/00
no abstracts in English
*
PNC TN9410 88-132, 132 Pages, 1988/09
This report summarizes the results of "Study on the Main Design Parameter of Large Scale Fast Reactor" which is a two year program from 1986 to 1987. This study was performed to contribute to the selection of main specifications and R&D items for demonstration FBR plant based on the experiences developed on "JOYO" and "MONJU".
*; *; Nakanishi, Seiji; *
PNC TN9410 88-103, 115 Pages, 1988/08
Current emphasis in the development of liquid-metal cooled fast breeder reactors (LMFBRs) is placed on the reduction of the plant construction cost without spoiling the safety in view of its practical application. So much effort is paid for the cost reduction. The reduction of piping length of heat transport system piping is considered as an effective measure for the cost reduction. An application of component floating support piping which brings good results in LWR leads to reduction of piping length and it is an effective measure as well as bellows expansion joint piping and high-chrome-steel piping in LMFBR. A design study for the application of the IHX floating support piping system to primary main heat transport system piping of LMFBR was conducted to demonstrate the adequacy of the piping system by using the improved structural design method considering the characteristics of LMFBR. The stress analysis of piping due to dead weight, thermal expansion at steady and transient conditions, and earthquake was performed, while the nozzles stress due to internal pressure, dead weight, earthquake and thermal expansion reaction force, and thermal transients was analyzed. It was confirmed that the analytical results were satisfied the allowable values and the piping support equipments were highly put into practical use. Therefore it was concluded that the adequacy of the IHX floating support piping system was demonstrated.
Nucl.Eng.Des., 97, p.279 - 296, 1986/00
Times Cited Count：2 Percentile：32.47(Nuclear Science & Technology)no abstracts in English
*; *; *; *
PNC TN941 83-08, 51 Pages, 1983/03
Experimental Fast Reactor "JOYO" completed its 75MWt power operation with the Mark-I core (breeding core) in Dec., 1981. The Auxiliary Core Cooling System (ACCS) has been operated satisfactorily since the first sodium charge in Jan., 1977. This paper describes the operation history until the end of Mark-I Operation. (1)Accumulated operation time of Primary Auxiliary Cooling System with circulating pump on during outages for annual inspection or others was only 530 hours, while the rest being occupied by the counterflow from main circulating punps. Auxiliary circulating pump experienced the automatical start only when the reactor scrammed and reactor sodium level lowered because of the failure of overflow make-up pump in July, 1981. (2)Secondary auxiliary cooling system had been operated approximately 39,140 hours in full flow rate, meanwhile circulating pump failed 4 times because of power loss.
;
Nihon Kikai Gakkai Rombunshu, A, 49(439), p.293 - 303, 1983/00
no abstracts in English
; ; Suzuki, Katsuo;
JAERI-M 82-043, 44 Pages, 1982/05
no abstracts in English
