Imaizumi, Tomomi; Miyauchi, Masaru; Ito, Masayasu; Watahiki, Shunsuke; Nagata, Hiroshi; Hanakawa, Hiroki; Naka, Michihiro; Kawamata, Kazuo; Yamaura, Takayuki; Ide, Hiroshi; et al.
JAEA-Technology 2011-031, 123 Pages, 2012/01
The number of research reactors in the world is decreasing because of their aging. However, the planning to introduce the nuclear power plants is increasing in Asian countries. In these Asian countries, the key issue is the human resource development for operation and management of nuclear power plants after constructed them, and also the necessity of research reactor, which is used for lifetime extension of LWRs, progress of the science and technology, expansion of industry use, human resources training and so on, is increasing. From above backgrounds, the Neutron Irradiation and Testing Reactor Center began to discuss basic concept of a multipurpose low-power research reactor for education and training, etc. This design study is expected to contribute not only to design tool improvement and human resources development in the Neutron Irradiation and Testing Reactor Center but also to maintain and upgrade the technology on research reactors in nuclear power-related companies. This report treats the activities of the working group from July 2010 to June 2011 on the multipurpose low-power research reactor in the Neutron Irradiation and Testing Reactor Center and nuclear power-related companies.
Fukuda, Yota*; Tamada, Taro; Takami, Hideto*; Suzuki, Shinichiro*; Inoue, Tsuyoshi*; Nojiri, Masaki
Acta Crystallographica Section F, 67(6), p.692 - 695, 2011/06
Takafumi, Nakano,; Ishida, Michihiko; Kazuyuki, Morimoto,; Inano, Masatoshi; Nojiri, Ichiro
JNC TN8410 2003-017, 190 Pages, 2004/03
Probabilistic Safety Assessment (PSA) methodology has been applied to evaluate the relative importance of safety functions that prevent the progress of event causing to postulate accidents.
Ishida, Michihiko; ; Kazuyuki, Morimoto,; Nojiri, Ichiro
Probabilistic Safety Assessment and Management, p.543 - 548, 2004/00
Probabilistic Safety Assessment (PSA) methodology has been applied on the Tokai Reprocessing Plant (TRP) to evaluate relative importance of safety functions that prevent the progress of events causing to postulated accidents. The relative importance of safety functions was evaluated by using two major importance measures, Fussell-Vesely and Risk Achievement Worth both generally used in PSA of nuclear power plants. Through these evaluations, some useful insights into the safety of the TRP have been obtained. The results of the relative importance measures would be utilized to qualify component/equipment as important safety function.
Sakurai, Kiyoshi; Nojiri, Ichiro*
JAERI-Conf 2003-019, p.855 - 857, 2003/10
This paper provides overview of sub-criticality safety analysis seminar (July 2000-July 2003, JAERI, total 40 engineers from universities, research institutes and enterprises) for nuclear fuel cycle facility with the Monte Carlo method in Japan. MCNP-4C2 system (MS-DOS version) was installed in each note-type personal computer. Fundamental theory of reactor physics and Monte Carlo simulation including MCNP-4C manual was lectured. Effective neutron multiplication factor and neutron spectrum were calculated for JCO deposit tank, JNC uranium solution storage tank, JNC plutonium solution storage tank and JAERI TCA core. In the seminar, methodology of safety management for nuclear fuel cycle facility was discussed in order to prevent criticality accident.
Sakurai, Kiyoshi; Kume, Etsuo; Maekawa, Fujio; Nojiri, Ichiro*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 2(2), p.196 - 201, 2003/06
New attempt on the Monte Carlo seminars is as following; authors prepared the practical lecture note and the text book of continuous energy Monte Carlo method. The exact evaluation method of lower weight bound was proposed in the shielding safety analysis seminar and the high energy calculation seminar. For the general Monte Carlo calculation, the neutron cross-section library including 340 nuclides with 293K, that was compiled from JENDL-3.2, was used for all Monte Carlo Calculation seminars. Also for the check of the calculation method, benchmark experiment problem was used for each calculation seminar.
Miura, Akihiko; Nojiri, Ichiro; Matsuo, Akiko*; *; *; *
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 0 Pages, 2003/00
Fukuda, Kazuhito; Tanabe, Yoji; Nojiri, Ichiro
Proceedings of International Conference on Advanced Nuclear Energy and Fuel Cycle Systems (GLOBAL 2003) (CD-ROM), p.115 - 118, 2003/00
Operation Testing Laboratory (OTL) is an examination laboratory located in the Tokai Reprocessing Plant (TRP). It has provided many research results since 1972 and contributed to TRP in order to safety operation, troubleshooting, and new technology for future reprocessing. In this report, we present facility outline of the laboratory,the several experimental results and future plan.
Sato, Yoshihiko; *; ; Nojiri, Ichiro
JNC TN8410 2001-027, 46 Pages, 2002/03
Safety study "Study of evaluation of abnormal occurrence for chemical substances in the nuclear fuel facilities" will be carried out from 2001 to 2005. In this study, the prediction of thermal hazards of chemical substances will be investigated and prepared. The hazard prediction method of chemical substances will be constructed from these results. Therefore, the hazard prediction methods applied in the chemical engineering in which the chemical substances with the hazard of fire and explosion were often treated were investigated. CHETAH (The ASTM Computer Program for Chemical Thermodynamic and Energy Release Evaluation) developed by ASTM (American Society for Testing and Materials) and TSS (Thermal Safety Software) developed by CISP (ChemInform St. Petersburg) were introduced and the fire and explosion hazards of chemical substances and reactions in the reprocessing process were evaluated. From these evaluated results, CHETAH could almost estimate the heat of reaction at 10% accuracy. It was supposed that CHETAH was useful as a screening for the hazards of fire and explosion of the new chemical substances and so on. TSS could calculate the reaction rate and the reaction behavior from the data measured by the various calorimeters rapidly. It was supposed that TSS was useful as an evaluation method for the hazards of fire and explosion of the new chemical reactions and so on.
; Nojiri, Ichiro
ANS Nuclear Criticality Safety Division Topical, 0 Pages, 2002/00
Ishida, Michihiko; Nojiri, Ichiro; ;
Kaku Nenryo Saikuru Anzensei Kokusai Wakushoppu, 0 Pages, 2002/00
Shirai, Nobutoshi; Sudo, Toshiyuki; Nojiri, Ichiro
Monte Karuroho Niyoru Ryoshi Shimyureshon No Genjo To Kadai, p.235 - 249, 2002/00
Miura, Akihiko; Nojiri, Ichiro; Matsuo, Akiko*; *
Pr.7-231 2002, 0 Pages, 2002/00
Ishida, Michihiko; Sudo, Toshiyuki; Omori, Eiichi; Nojiri, Ichiro
Vol.1 No.040, 1(40), 0 Pages, 2001/00
Dai-5-Kai Kakuritsuronteki Anzen Hyoka To Kanri Ni Kansuru Kokusai Kaigi (PSAM5), 0 Pages, 2000/00
Kosaka, Ichiro; Nojiri, Ichiro; Yamanouchi, Takamichi
JNC TN8410 99-017, 302 Pages, 1999/03
For the purpuse of the analysis the loss of the containment of the gas, the pressure transients had been caluculated, using FIRAC Code, caused by the first fire incident in the Bitiminization filling cell, R152, and by the corst down of the cell exhaustion, limited operation of the ventilation system, begining at 1006, 11th, March, 1997. The FIRAC Code calculates the pressure and temperature of the cell and ventilation system at the fire incident. The all of the facility was modeled in the calculations. The heat generation rate and other unknown terms were taken into as the caluculational parameters, and the process of the incident were evaluated by attention to the different results by each parameter. In the results, the constraction of the ventilation system affected the process of the fire incident, and in the case taking account of exhausting to Z-facility, abnormal reverse flow occurred in the system.
Maki, Akira; Nojiri, Ichiro; ; ; Yamanouchi, Takamichi
JNC TN8440 99-002, 366 Pages, 1998/11
The fire and explosion incident of the bituminization facility happened in March 1997 although JNC had taken enough care of the safety of TRP. JNC reflected on it and decided to evaluate the safety of TRP voluntarily. This evaluation has included five activities, that is, (1)confirmation of the structure and organization of TRP, (2)research of the data for operation, radiation and maintenance of TRP, (3)research of reflection of the accidents and troubles which have happened at the past, (4)evaluation on the prevention system, (5)evaluation on the mitigation system. We publish this report to contribute to inheritance of accumulated knowledge and techniques from generation to generation, and remind us of lesson from the fire and explosion incident of the bituminization.
Sudo, Toshiyuki; ; ; Nojiri, Ichiro; Maki, Akira; Yamanouchi, Takamichi
JNC TN8410 99-003, 69 Pages, 1998/11
As a part of the safety confirmation work of Tokai Reprocessing Plant, the appropriateness was checked on the basic data used in criticality safety and shielding design of early-designed facilities in the plant on the basis of recent knowledge and safety evaluation methods. In the criticality safety design, it was confirmed that critical and subcritical values concerning mass and concentration of U and Pu and equipment dimension were appropriate. In the shielding design, it was found that the relation between shielding thickness and permissible radioactivity might give underestimated results of shielding thickness necessary to limit dose rate to the designated one on some condition. In this cases, however, it was confirmed that necessary shielding thickness has been secured because of the conservative calculation conditions for the real conditions except the operation test laboratory (OTL). However, the amount of radioactivity handled at OTL needs to be limited. From a viewpoint of criticality safety, operational control for U and Pu transfer was also investigated, As a result of it, at the transfer route where erroneous batch-wise transfer of process solution might lead to a criticality accident, the reliability of U and Pu concentration measurement needs to be improved by multiple measurements. At other transfer routes, it was confirmed that single failure of equipment or operation error would not lead to a criticality problem.
Nojiri, Ichiro; *; *
PNC TN8410 98-031, 226 Pages, 1998/02
; Nojiri, Ichiro; Kurosawa, Naohiro*; *; Sasaki, Toshihisa*
PNC TN8410 98-022, 145 Pages, 1998/01
In plant designs and safety evaluations of nuclear fuel cycle facilities, it is important to evaluate the direct radiation and the skyshine (air-scattered photon radiation) from facilities reasonably. The Neutron and Photon Shielding Calculation System for Workstation (NPSS-W) was developed. The NPSS-W can carry out the shielding calculations of the photon and the neutron easily and rapidly. The NPSS-W can easily calculate the radiation source intensity by ORIGEN-S and the dose equivalent rate by S transport calculational codes, which are ANISN and DOT3.5. The NPSS-W consists of five modules, which named CAL1, CAL2, CAL3, CAL4, CAL5). Some kinds of shielding calculational systems are calculated. The user's manual of NPSS-W the examples of calculations for each module and the output data are appended.