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Watanabe, Sota*; Kubo, Koji*; Okajima, Satoshi; Wakai, Takashi
Nihon Kikai Gakkai M&M 2017 Zairyo Rikigaku Kanfarensu Koen Rombunshu (Internet), p.581 - 585, 2017/10
no abstracts in English
Hada, Kazuhiko
Nihon Kikai Gakkai Doryoku Enerugi Shisutemu Bumon Nyusu Reta, (25), p.2 - 3, 2002/10
no abstracts in English
Ohira, Shigeru; Tada, Eisuke; Hada, Kazuhiko; Neyatani, Yuzuru; Maruo, Takeshi; Hashimoto, Masayoshi*; Araki, Takao*; Nomoto, Kazuhiro*; Tsuru, Daigo; Ishida, Toshikatsu*; et al.
Fusion Engineering and Design, 54(3-4), p.515 - 522, 2001/04
Times Cited Count：3 Percentile：27.1(Nuclear Science & Technology)no abstracts in English
Kakehi, Isao; Nakabayashi, Hiroki
JNC TN9400 2000-051, 237 Pages, 2000/04
JNC-TN9400-2000-051.pdf:8.14MB
In this study, we have proposed the concept of safety systems (solutions of safety problems) in pyrochemical reprocessing systems (lt consists of pyrochemical reprocessing methods and the injection casting process for the metal fuel fabrication, or vibro-packing process for the oxide fuel fabrication.) which has different concept from the existing PUREX reprocessing method and pellet fuel fabrication process. And we performed its safety evaluations. FoIlowing the present Japanese safety regulations for reprocessing facilities, we pointed out functions, design requirements and equipments relating to its safety systems and picked up subjects. For the survey of safety evaluations, we first selected anticipated events and accident events, and second by evaluated 6the correspondence of the limitation of the public exposure to the accidents above, by using two parameters, the safety design parameter (the filter performance to confine radioactive matelials) and the leak inventory of radioactivities, and last by picked up its problems. ln addition to the above evaluations we performed basic criticality analyses for its systems to utilize these results for the design and evaluation of the criticality safety management system. Thus this study specified the concept of safety systems for pyrochemical reprocessing processes and then issues in order to establish safety design policies (matters which must consider for the safety design) and guides and to advance more definite safety design.
; ; *;
JNC TN9400 2000-043, 23 Pages, 2000/03
ln the feasibility study on commercialized fast breeder reactor (FBR) cycle systems conducted in JNC, it is required for candidate FBR plants that the level of safety should be enhanced so as to assure: (1)Comparative or superior safety level to that of light water reactors (LWRs), and (2)releaf of the public from anxiety about potential nuclear hazard. Adopting Passive safety characteristics is one of the measures. To attain the above safety objective, we considered implication of the basic safety principles for nuclear power plants that were created by the international nuclear safety advisory group of IAEA. The way to relieve from the anxiety was also taken into account. Then a definite safety objective was set from the standpoint of prevention of core disruptive accident (CDA). Furthermore, as a definite safety goal relating to reactor coresafety, elimination of re-criticality issues under CDA was set by considering characteristics of FBR in comparison with those of LWR. To examine measures for elimination of re-criticality issues, we developed a quick method to estimate possibility of re-criticality under CDA, by drawing a map about criticality characteristics under CDA in various degraded cores. Then hopeful measures were proposed for elimination of re-criticality issues in sodium-cooled FBR with mixed-oxide fuel. Molten fuel discharge behavior of their measures was preliminarily analyzed. We concluded that discharge capability of "a subassembly with an internal duct" was effective, and that "partial removal of axial blanket" was also effective as one of the measures though it has small effect on core performance.
Abe, Hitoshi; Tashiro, Shinsuke; Nagai, Hitoshi; Koike, Tadao; Okagawa, Seigo; Murata, Mikio
JAERI-Tech 99-067, p.23 - 0, 1999/09
no abstracts in English
Isozaki, Kazunori;
PNC TN9410 97-069, 134 Pages, 1997/07
Hyougo-ken southern earthquake broke out in 1997/01/17. The Atomic Energy Safety Co㎜ission considered reasonable of the design guide for seismic design. And the Science and Technology Agency(STA) required reevaluation of atomic power facilities built by old design guide according to the new seismic design guide. JOYO obtained the construction license in 1970/02. Heat transport system and buildings of JOYO was reevaluated by the new seismic design guide for the MK-III project. So, JOYO was not required reevaluation by STA. But, this evaluation of MK-III was limited to reconstruction area, and the seismic design was reevaluated extensively to confirm earthquake proof characteristics. The structural integrity of buildings and equipments was confirmed by the result of reevaluation by the new seismic design guide. The analysis model conditions were established according to the 1987 and 1991 version of JEAG. This was done by ground investigation result and buildings vaibration test. It was made clear that the analysis model conditions were reasonable and conservative from a technical view point.
Miyakawa, Shunichi; ; Soga, Tomonori
PNC TN9410 97-068, 113 Pages, 1997/07
Since the first control rod design for the Joyo Mk-II core (about twenty years ago), there have been several challenging improvements; for example, a helium venting mechanism and a flow induced vibration prevention mechanism. Forty-four control rods with these various modifications have been fabricated. To date, thirty-four have been irradiated and the sixteen have been examined, This experience and effort has produced fruitful results: (1)Efficiency and reliability of the diving-bell type Helium venting mechanism (2)Efficiency of the flow induced vibration prevention mechanism (3)Efficiency of the improvement for scram damping mechanism (4)Clarification of absorvber-pellet-cladding-mechanical-interaction (ACMI)phenomena and preventive methods The fourth result listed above has been a subject of investigation for fifteen years in several countries, that is a main phenomena to dominate control rod life time. The results of this investigation of ACMI in absorber elements are summarized below: (a)In five of Joyo Mk-II control rods, cladding cracks were found in fifteen of the elements. These cracks were caused by a acceleration ACMI, due to B
C fragments relocation. They occurred over a wide burnup range from 5E+26 Cap./m
to 45E+26Cap./m in a nearly typical provability distribution. The cladding cracked because of its low ductility (approximately 1/4 lower than the uniform elongation of usual tensile testing for irradiated 316SS cladding) due to neutron irradiation and the ultra slow ACMI induced strain rate. (b)In this case the crack growth rate is extremely slow and the ACMI induced cracking in absorber elements do not influence either the reactor or plant operations. It is on this basis that a strict limitation to avoid the cladding crack is not necessary. According1y, it is suggested that a realistic design standard should consider the ACMI phenomena and the burnup limit be based on the nominal base calculation for average plastic strain use ...
Iwata, Koji; *; *; *; *; *; *
PNC TN9410 97-042, 8 Pages, 1997/03
A design guide for flow-induced vibration of thermometer wells is proposed to prevent the recurrence of the failure of thermometer wells, which was the direct cause of the 1995 sodium leak incident of the secondary main piping of the prototype fast breeder reactor MONJU. As a supplement to the technical standards in force for MONJU, the design guide specifies the methods of evaluation and the design criteria on structural integrity against flow-induced vibration for thermometer wells, which are inserted into pipes of fast breeder reactors. The design guide is a PNC's (Power Reactor and Nuclear Fuel Development Corporation) internal guide for MONJU, which is to be used, with the permission of outside authorities, to confirm the integrity of the existing equipments as well as to make an improved design of thermometer wells. The proposed design guide was prepared by the Special Working Group on Thermometer Design Guide, organized in PNC during the period from May to November, 1996.
Miya, Kenzo*; ; Takatsu, Hideyuki; Hada, Kazuhiko; Koizumi, Koichi; Jitsukawa, Shiro; ; Okawa, Yoshinao; Shimakawa, T.*; Aoto, Kazumi*; et al.
Fusion Engineering and Design, 31, p.145 - 165, 1996/00
Times Cited Count：3 Percentile：32.72(Nuclear Science & Technology)no abstracts in English
Power Reactor and Nuclear Fuel Development Corporation
PNC TN9360 95-002, 98 Pages, 1995/11
no abstracts in English
Power Reactor and Nuclear Fuel Development Corporation
PNC TN9360 95-001, 104 Pages, 1995/11
no abstracts in English
Miya, Kenzo*; ; Takatsu, Hideyuki
Kikai No Kenkyu, 47(1), p.179 - 184, 1995/00
no abstracts in English
; Hayashi, Hideyuki; ; ;
PNC TN9410 94-222, 355 Pages, 1994/07
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.
*; *; *; *; *; *; *
PNC TJ9164 94-006, 133 Pages, 1994/03
This report gives an applicability of SWAT-3 facility and contents of the reconstruction in order to confirm a DBL (Design Basis Leak) for the demonstration reactor SG. (1).Test Cndition and test case. Evaluation of the wall temperature for adjacent heat transfer tubes under the sodium-water reaction event was performed. (a)As the effect of tube rupture due to overheating, failure of upper part of the helical coil was severer than one of the lower part. (b)The wall temperature depends on the water side condition. (c)Reference test condition, whici is water leak rale about 1 kg/s, failure of upper part of the helical coil and 30% partial load, was selected. A total of ten test cases were decided. (2).System and Components Design. (a)Large leak sodium-water reaction analyses including water injection rate analysis and quasi-steady pressure analysis were performed. The maximum water leak rate of 1 DEG was 7.2 kg/s and the water leak rate at the quasi-steady state was 3.1 kg/s. The maximum pressure was 18.1kg/cm
a at the piping between the reaction vessel and IHX, the pressure was within the design condition of SWAT-3 facility. (b)Based on the results of the large leak sodium-water reaction analyses, a reaction vessel, water heaters and a dump tank were designed and their design specification were clarified. The reaction vessel was a scale of one third of the demonstration reactor SG and it was designed to be able to conduct the water injection test twice with one test unit. (c)A system and piping diagram, and many kinds of list (Piping list, Valve list, instrumentlist) were made up. (3).Reconstruction scope and arrangement plan. The reconstruction scope and a layout for the components and piping were clarfied and the arrange ment plans were made up. (4)Reconstruction period. The recoastruction period and man power for the design, fabrication, inspection and installation were studied and the reconstruction schedule was made up.
