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Mikake, Shinichiro; Yamamoto, Masaru; Ikeda, Koki; Sugihara, Kozo; Takeuchi, Shinji; Hayano, Akira; Sato, Toshinori; Takeda, Shinichi; Ishii, Yoji; Ishida, Hideaki; et al.
JAEA-Technology 2010-026, 146 Pages, 2010/08
JAEA-Technology-2010-026.pdf:41.08MB
JAEA-Technology-2010-026-appendix(CD-ROM).zip:83.37MB
The Mizunami Underground Research Laboratory (MIU), one of the main facilities in Japan for research and development of the technology for high-level radioactive waste disposal, is under construction in Mizunami City. In planning the construction, it was necessary to get reliable information on the bedrock conditions, specifically the rock mass stability and hydrogeology. Therefore, borehole investigations were conducted before excavations started. The results indicated that large water inflow could be expected during the excavation around the Ventilation Shaft at GL-200m and GL-300m Access/Research Gallery. In order to reduce water inflow, pre-excavation grouting was conducted before excavation of shafts and research tunnels. Grouting is the injection of material such as cement into a rock mass to stabilize and seal the rock. This report describes the knowledge and lessons learned during the planning and conducting of pre-excavation grouting.
Takamura, Shuichi*; Kado, Shinichiro*; Fujii, Takashi*; Fujiyama, Hiroshi*; Takabe, Hideaki*; Adachi, Kazuo*; Morimiya, Osamu*; Fujimori, Naoji*; Watanabe, Takayuki*; Hayashi, Yasuaki*; et al.
Kara Zukai, Purazuma Enerugi No Subete, P. 164, 2007/03
no abstracts in English
Kuznetsov, I.*; Shvetsov, Y. E.*; Ashurko, Yu. M.*; Volkov, A. V.*; Kashcheev, M. V.*; Tsykunov, A. G.*; Kamanin, Y. L.*; Bakhmetyev, A. M.*; Zamyatin, V. A.*; Niwa, Hajime; et al.
Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10
None
Konomura, Mamoru; Ogawa, Takashi; Okano, Yasushi; Yamaguchi, Hiroyuki; Murakami, Tsutomu; Takaki, Naoyuki; Nishiguchi, Youhei; Sugino, Kazuteru; Naganuma, Masayuki; Hishida, Masahiko; et al.
JNC TN9400 2004-035, 2071 Pages, 2004/06
JNC-TN9400-2004-035.pdf:76.42MB
The attractive concepts for Sodium-, lead-bismuth-, helium- and water-cooled FBRs have been created through using typical plant features and employing advanced technologies. Efforts on evaluating technological prospects of feasibility have been paid for these concepts. Also, it was comfirmed if these concepts satisfy design requierments of capability and performance presumed in the feasibilty study on commertialization of Fast Breeder Reactor Systems. As results, it was concluded that the selection of sodium-cooled reactor was most rational for practical use of FBR technologies in 2015.
Ishida, Masayoshi; Kawada, Kenichi; Niwa, Hajime
JNC TN9400 2003-059, 74 Pages, 2003/07
JNC-TN9400-2003-059.pdf:1.58MB
Safety characteristics in core disruptive accidents (CDAs) of mid-sized MOX fueled liquid metal reactor core of high converter type have been examined by using the CDA initiating phase analysis code SAS4A. The design concept of high converter type reactor core has been studied as one of options in the category of sodium-cooled reactor in Phase II of Feasibility Study on Commercialized Fast Reactor Cycle System.An unprotected loss-of-flow accident (ULOF) has been selected as a representative CDA initiator for this study. A core concept of high converter type, which employed a large diameter fuel pin of 11.1mm with 1.2m core height to get a large fuel volume fraction in the core to achieve high internal conversion ratio was proposed in JFY2001. Each fuel subassembly of the core (abbreviated here as UPL120) was provided with an upper sodium plenum directly above the core to reduce the sodium void reactivity worth. Because of the large fuel pin diameter, average specific fuel power density (31 kW/kg-MOX) of UPL120 is about one half of those of conventional large MOX cores. The reactivity worth of sodium voiding is 6$ in the whole core, and -1$ in the all upper plenums. Initiating phase of ULOF accident in UPL120 under the conditions of nominal design and best estimate analysis resulted in a slightly super-prompt critical power burst. The causes of the super-prompt criticality have been identified twofold: (a) the low specific fuel power density of core reduced the effectiveness of prompt negative reactivity feedback of Doppler and axial fuel expansion effects upon increase in reactor power, and (b) the longer core height compared with conventional 1m cores brought, together with the lower specific power density, a remarkable delay in insertion of negative fuel dispersion reactivity after the onset of fuel disruption in sodium voided subassembly due to the lower linear heat rating in the top portion of the core. During the delay, burst-type fuel failures in sodium un-v
Ohira, Shigeru; Tada, Eisuke; Hada, Kazuhiko; Neyatani, Yuzuru; Maruo, Takeshi; Hashimoto, Masayoshi*; Araki, Takao*; Nomoto, Kazuhiro*; Tsuru, Daigo; Ishida, Toshikatsu*; et al.
Fusion Science and Technology, 41(3), p.642 - 646, 2002/05
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
; Tobita, Yoshiharu; ; ; Ishida, Masayoshi; ;
JNC TN9400 2001-056, 64 Pages, 2001/03
JNC-TN9400-2001-056.pdf:2.66MB
The Phase I of the Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System is being performed for two years from Fisca1 Year (FY) 1999. This report describes the results obtained in FY 1999 as an interim report of the Phase I from the viewpoint of reactor safety for various FBR plant condidates. The objectives of the study are to understand the safety charaeteristics of advanced fuel and to examine the fulfillment of the target level of reactor safety in each plant concept. The items studied are the recriticality characteristics of degraded core for various core concepts, investigation of the measures for avoiding recriticality event, safety analysis of sodium cooled MOX fueled cores, target of void worth in core design for sodium cooled reactors, and investigation of core disruptive accident sequences in various reactor concepts. The results of this study have been reflected properly to the core and plant design. In FY 2000, the study will be continued along with the progress of the plant design in order to prepare for the judgment of the candidates from the viewpoint of reactor safety.
Fukahori, Tokio; Kunieda, Satoshi; Chiba, Satoshi; Harada, Hideo; Nakashima, Hiroshi; Mori, Takamasa; Shimakawa, Satoshi; Maekawa, Fujio; Watanabe, Yukinobu*; Shigyo, Nobuhiro*; et al.
no journal, ,
The latest version of JENDL High Energy File (JENDL/HE) and JENDL Photonuclear Data File (JENDL/PD) is being planed to be released as JENDL/HE-2007 and JENDL/PD-2007. In JENDL/HE-2007, nuclear data for about 100 nuclides, which are newly ecaluated and revised from JENDL/HE-2004 will be stored. The JENDL/PD-2007 will have nuclear data for about 170 nuclides.
