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Takahashi, Naoki; Yoshinaka, Kazuyuki; Harada, Akio; Yamanaka, Atsushi; Ueno, Takashi; Kurihara, Ryoichi; Suzuki, Soju; Takamatsu, Misao; Maeda, Shigetaka; Iseki, Atsushi; et al.
Nihon Genshiryoku Gakkai Homu Peji (Internet), 64 Pages, 2016/00
no abstracts in English
Sanada, Yukihisa; Tsujimura, Norio; Shimizu, Yoshio; Izaki, Kenji; Furuta, Sadaaki
Journal of Nuclear Science and Technology, 45(Suppl.5), p.74 - 77, 2008/06
Times Cited Count：1 Percentile：10.05(Nuclear Science & Technology)The purpose of this study is the establishment of the determination procedures for the placements of CAAS detectors in PCDF. The dose of detection point was evaluated the simple equation which was formulated in calculated factors by MCNP and ANISN. When the alarm trip point was 2.0 mGy/h, the detection area was covered 30 m distances from the equipment to the CAD and 100 cm concrete shielding. This result will be reflected in the determination of the CAD placements and three CADs were placed in PCDF.
Shimizu, Yoshio; Suitsu, Yuichi; Murakami, Tatsutoshi; Yuri, Akiya
Proceedings of 8th International Conference on Nuclear Criticality Safety (ICNC 2007), p.335 - 340, 2007/05
Nuclear criticality safety evaluations for the fast breeder reactor's MOX fuel fabrication facility were performed. In the parametric studies made with SCALE, following three cases are here introduced. (1) The effect of the plutonium isotopic ratio on Pu
(
U, 
Pu, 
Pu) mass control were evaluated. The design conditions of the plutonium isotopic ratio under the operating condition were adjusted from the evaluation. (2) The moderation effects of organic materials in MOX fuel fabrication facility were evaluated. The water contents equivalent moderation effects to organic contents were obtained from the evaluation. (3) The effects of nonuniform mixture of MOX powder and water were evaluated by two-phase model and SMORES. The 
k, which differ from the k
of uniform model, were evaluated.
Shimizu, Yoshio; Oka, Tsutomu
Journal of Nuclear Science and Technology, 41(Suppl.4), p.105 - 108, 2004/03
None
Shimizu, Yoshio; Oka, Tsutomu
Proceedings of 7th International Conference on Nuclear Criticality Safety (ICNC 2003), p.858 - 862, 2003/00
None
; Nojiri, Ichiro
ANS Nuclear Criticality Safety Division Topical, 0 Pages, 2002/00
None
Omori, Eiichi; Sudo, Toshiyuki; ; Kosaka, Ichiro; ; ;
JNC TN8410 99-005, 274 Pages, 1999/02
Through the investigation of the cause of the fire and explosion incident at Bituminization Demonstration Facility of JNC Tokai Works, the lesson learned is that the safety assessment is necessary even for the licensed facilities by recent knowledge. The safety assessment has been conducted for the facilities in Tokai Reprocessing Plant by recent knowledge and operational experience. This report describes the evaluation results of the incident mitigation systems and the hypothetical accidents. In the evaluation of the incident mitigation system, supposed incidents were solvent fire, rapid reaction of hydrazine decomposition, leakage of radioactive material and loss of power supply. The evaluation was focused on the integrity of the filters in case of the fire and the rapid leaction, the availability of the recovery system in case of the radioactive leakage, and so on. As a result of evaluation, several improvements were pointed out to be necessary for the prevention of incident magnification. In the evaluation of the hypothetical accidents, criticality at a dissolver and fire at solvent extraction mixer-settlers were hypothesized. It was confirmed that the Tokai Reprocessing Plant is still distant enough from the population.
; Nagai, Takayuki*; ; Sasaki, Toshihisa*
JNC TN8400 99-007, 216 Pages, 1999/02
If the fire accident was occurred in the Glove Box (GB) in the nuclear fuel cycle facilities, it is important to clear the fluctuation of the negative pressure in GB and the influence of the ventilation system. In Japan Nuclear Cycle Development Institute, the fire and extinguishment experiments about the GB ventilation system were executed. The simulations with a calculation code of these experiments were also performed. In this report, FIRAC were improved and these experiments were evaluated with FIRAC. FIRAC, which was developed in Los Alamos National Laboratory in U.S., is a computer code to simulate fire accidents in nuclear facilities. The original FIRAC can not simulate the GB ventilation system adequately. The original FIRAC can not simulate the inflow of the suffocative gas for the extinguishment experiments. The control damper model, the correction of storage of heat, the heat conduction of the construction materials, the model of the hot layer and cold layer, the model of inflow of the suffocative gas, etc., were improved, and the FIRAC are performed to simulate these experiments fitly.
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; 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.
Shimizu, Yoshio; Suitsu, Yuichi; Murakami, Tatsutoshi; Yuri, Akiya
no journal, ,
no abstracts in English
Suitsu, Yuichi; Shimizu, Yoshio; Murakami, Tatsutoshi; Yuri, Akiya
no journal, ,
no abstracts in English
Yuri, Akiya; Shimizu, Yoshio; Suitsu, Yuichi; Murakami, Tatsutoshi; Ninagawa, Junichi
no journal, ,
no abstracts in English
Sanada, Yukihisa; Tsujimura, Norio; Shimizu, Yoshio; Izaki, Kenji; Jin, Kazumi; Mikami, Satoshi; Kobayashi, Hirohide; Kawai, Keiichi*
no journal, ,
no abstracts in English
Shimizu, Yoshio
no journal, ,
no abstracts in English
