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Journal Articles

Additional information to report on site tour of the Fukushima Daiichi Nuclear Power Station

Suto, Toshiyuki

Genshiryoku, hoshasen Bukaiho (Internet), (19), P. 15, 2016/12

The Tritiated Water Task Force under METI's Committee on Countermeasures for Contaminated Water Treatment for Fukushima Daiichi Nuclear Power Plant (1F) reported that the option of post-dilution offshore release could dispose the tritiated water at a smallest cost in the shortest amount of time. The amount of tritium in the contaminated water at 1F was compared with ones released from nuclear power plants and reprocessing plants as some help for grasping its level of magnitude.

Journal Articles

Report on site tour of the Fukushima Nuclear Power Station

Suto, Toshiyuki

Gijutsushi, 28(11), p.8 - 11, 2016/11

Five years have passed since the accident of the Fukushima Daiichi Nuclear Power Station. The Nuclear and Radiation section of the Institute of Professional Engineers hosted a site tour of the plant to make themselves sure what is going on in it and to disseminate information about it. The conditions of landscape during traveling between the gathering place and the plant, each reactor, contaminated water treatment, site, and work environment improvement will be reported.

Journal Articles

Penetration behavior of water solution containing radioactive species into dried concrete/mortar and epoxy resin materials

Sato, Isamu; Maeda, Koji; Suto, Mitsuo; Osaka, Masahiko; Usuki, Toshiyuki; Koyama, Shinichi

Journal of Nuclear Science and Technology, 52(4), p.580 - 587, 2015/04

 Times Cited Count:2 Percentile:66.76(Nuclear Science & Technology)

Penetration behavior of radionuclides such as $$^{137}$$Cs into dried concrete material, dried mortar material and epoxy paint for a few dozen days was observed using a solution containing fission products extracted from irradiated fuels to obtain fundamental information on the radionuclide penetration rate and depth. Hardly any radionuclides could penetrate into the epoxy paint. The radionuclide solution penetrated into concrete and mortar materials to a depth of a few millimeters for a few dozen days. The penetration behavior observed near the surface of concrete and mortar materials was similar to the diffusion of nuclides in media such as water-saturated concrete, bentonite and cement materials.

JAEA Reports

Penetration behavior of solution containing radioactive nuclides into floor and wall materials

Usuki, Toshiyuki; Sato, Isamu; Suto, Mitsuo; Maeda, Koji; Osaka, Masahiko; Koyama, Shinichi; Tokoro, Daishiro*; Sekioka, Ken*; Ishigamori, Toshio*

JAEA-Testing 2014-001, 29 Pages, 2014/05

JAEA-Testing-2014-001.pdf:5.33MB

The penetration tests with solution containing radioactive nuclides were experimented to understand basic data for floor and wall materials of Fukushima Daiichi reactor buildings. The solution prepared from irradiated fuels was used as solution containing radioactive nuclides. The solution was applied to surface of epoxy paint, dried concrete and mortar used as specimens. Dose-rate profiles of direction of depth were given by radiation measurement and grinding of the specimens. The penetrations of radioactive nuclides for epoxy paint specimens were not clearly observed and the penetration depths would be within 0.4 mm. The penetrations of radioactive nuclides for dried concrete specimens proceeded. The penetration rates were substantially decreased when 16 days have elapsed from start. The dose rates of penetrated dried concrete specimens were reduced to background by grinding-2.0 mm. $$gamma$$-ray spectrometry measurement showed that penetration behavior of near surface concrete are different among nuclides and the penetration behavior of radioactive nuclides into dried concrete and mortar materials through solution is similar to migration behavior of ions into those water-saturated materials.

JAEA Reports

Subcritical dimensions for design study of future reprocessing facility

Suto, Toshiyuki; Fukushima, Manabu*

JAEA-Data/Code 2011-021, 91 Pages, 2012/02

JAEA-Data-Code-2011-021.pdf:3.57MB

Future reprocessing facilities are considered to treat not only LWR uranium fuels but also LWR-MOX fuels and even FBR fuels. These spent fuels have various enrichments, isotopic compositions of plutonium. In order to make reasonable design and management of criticality safety for such multi-fuel reprocessing facilities, there needs to be critiricality safety data with a multiple of enrichments, isotopic compositions of plutonium. The authors calculated subcritical dimensions of uranium/plutonium mixed nitrate solution with a multiple set of enrichments and isotopic compositions of plutonium for cylinder, slab, and annular geometries and arranged the results in tables and graphs which are usable for the criticality safe design study of future reprocessing facilities. In addition, some findings useful for selecting criticality safety control methods were summarized.

JAEA Reports

Supplemental study on dose control for a criticality accident

Kanamori, Masashi; Suto, Toshiyuki; Tanaka, Kenichi*; Takada, Jun*

JAEA-Technology 2011-004, 12 Pages, 2011/03

JAEA-Technology-2011-004.pdf:0.97MB

In the previous report "A Study on Dose Control for JCO Criticality Accident Termination" (JAEA-Technology 2009-043), we discussed how to control the dose received during the termination work of the criticality accident. In this paper, we focused on the difference of the way in which dose rate attenuates between within 100 m from the source point and beyond 100 m and discussed the validity of using log-log plotting/semi-log plotting of dose rate - distance relation in order to extrapolate the dose rate at work place near the criticality accident point. In addition, we studied on the effect of the number of dose rate measurement data to be used for extrapolation. We recommend that about 10 mSv which is a half of 20 mSv annual dose limit should be used as worker's dose control target for the high neutron dose field work to ensure enough safety margin considering the following three points; (1) annual dose limit for workers, (2) dose received before, (3) measurement error.

JAEA Reports

A Study on dose control for Tokaimura criticality accident termination

Kanamori, Masashi; Suto, Toshiyuki; Tanaka, Kenichi*; Takada, Jun*

JAEA-Technology 2010-042, 11 Pages, 2011/01

JAEA-Technology-2010-042.pdf:0.94MB

JAEA-Technology 2009-043 "A Study on dose control for JCO criticality accident termination", the dose we discuss how to manage termination of the criticality accident at work or (we refer as previously report) As a result, based on the measurements from around 40 m to 100 m, we made a re-evaluation of the dose. Reevaluated doses match with the degree of accuracy 60% to 80% compared with the respective individual dose. In this paper, we validate by these doses by using log-log plots and semi-log plots for the distance from the source approximately 100 m and further attenuation. As a result, if the field is under high doses of neutrons, dose constraint assessment should consider some points, by using 10 mSv, half of the annual limit 20 mSv, as dose reference, the work performed could safely be managed. And summaries the valid range of log-log plots for intense neutron dose estimation.

Journal Articles

Commentaries for third secondary national examination on fiscal 2010 for the professional engineer of nuclear and radiation; Commentaries (Part 1) including key point for elective examinations

Sasaki, Satoru; Suto, Toshiyuki; Harada, Akio; Kurihara, Ryoichi; Yamamoto, Kazuyoshi; Tsuchida, Noboru; Shimizu, Isamu; Nomura, Toshibumi

Genshiryoku eye, 57(1), p.66 - 75, 2011/01

no abstracts in English

JAEA Reports

A Study on dose evaluation for Tokaimura criticality accident termination

Kanamori, Masashi; Suto, Toshiyuki; Tanaka, Kenichi*; Takada, Jun*

JAEA-Technology 2010-025, 11 Pages, 2010/08

JAEA-Technology-2010-025.pdf:1.55MB

Verification of dose control method for Tokaimura JCO criticality accident was performed. Personal dose estimation for Tokaimura criticality accident termination was performed based on measurements of neutron and $$gamma$$ ray doses taken before the work commenced, but the personal dose for the workers as a result of the termination work was found to be approximately 50 times higher than the previous estimation. For this report, we reevaluated doses based on the results of close range measurements from 40 meters to 100 meters, and the results were found to match personal doses with an accuracy of between 60-80%.

JAEA Reports

Fast Reactor Cycle Technology Development Project (FaCT Project); A Design study on an engineering-scale hot test facility (Interim report)

Nakamura, Hirofumi; Nagai, Toshihisa; Suto, Toshiyuki; Kosaka, Ichiro; Nakazaki, Katsutoshi; Suto, Shinya; Nakamura, Tomotaka; Nakabayashi, Hiroki; Hayashi, Naoto; Sumida, Daisaku

JAEA-Technology 2008-077, 276 Pages, 2008/12

JAEA-Technology-2008-077.pdf:25.66MB

Japan Atomic Energy Agency (JAEA) has been conducting "Fast Reactor Cycle Technology Development Project (FaCT Project)" for the purposes of researching and developing the technologies for the fast breeder reactor cycle commercialization since Japanese fiscal year (JFY) 2007. Based on the above R&D program for reprocessing system, the engineering-scale hot test would provide demonstration data on the specification, operation and maintenance of the adapted innovative technologies, system and plant. And more, these results would be fed to the design of the demonstration facility planning on the FaCT project road map. This report is the interim report of design studies about the engineering-scale hot test facility and includes not only design of the equipment and facility, but also consideration for design principle, requirements and facility basic planning.

Oral presentation

Current status of the equipment failure rate database of reprocessing plant

Suto, Toshiyuki

no journal, , 

Current status about the contract work with the JNES will present at the workshop of the risk information application for the nuclear fuel cycle facilities.

Oral presentation

Probabilistic safety assessment of the Tokai Reprocessing Plant

Ishida, Michihiko; Suto, Toshiyuki; Inano, Masatoshi; Aoshima, Atsushi

no journal, , 

In the Tokai Reprocessing Plant (TRP), process safety has been evaluated since the fire and explosion incident of the bituminization facility in 1997. In this report, both deterministic and probabilistic safety assessment approaches of the TRP are summarized.

Oral presentation

Current status of the component failure rates evaluation work based on the reprocessing plant maintenance data

Ishida, Michihiko; Suto, Toshiyuki; Inano, Masatoshi; Aoshima, Atsushi; Muramatsu, Ken; Ueda, Yoshinori*

no journal, , 

A study to evaluate component failure rates for a reprocessing plant has started at the Japan Atomic Energy Agency (JAEA). The study is sponsored by the Japan Nuclear Energy Safety Organization (JNES) as a contract research. The Tokai Reprocessing plant(TRP) has been operating over 30 years since 1977. Maintenance records of plant components have been accumulated and compiled in the Tokai Reprocessing Plant Maintenance Support System (TORMASS). Since the TORMASS was thought to be an effective source to pick up component failure information, the evaluation work of component failure rates started at 2005 based on the maintenance records in the TORMASS. By the end of 2006, failure rates for 17 types of 392 components had been evaluated. This result is expected to contribute to building up generic database for reprocessing plant component failure rates in order to promote the utilization of the risk information for reprocessing plants in Japan.

Oral presentation

Study on characteristics of criticality safety measures for fast reactor fuel reprocessing plant

Nakazaki, Katsutoshi; Suto, Toshiyuki; Kosaka, Ichiro; Fukushima, Manabu*

no journal, , 

no abstracts in English

Oral presentation

Preparation of equipment failure rate for a reprocessing facility, 2; Calculation of equipment failure rates based on the actual plant data

Ishida, Michihiko; Inano, Masatoshi; Aoshima, Atsushi; Suto, Toshiyuki; Muramatsu, Ken; Ueda, Yoshinori*

no journal, , 

Component failure rates are one of the most important inputs for PSA. Although there are many component failure rates databases for NPPs, there are few for a reprocessing plant. Therefore, in order to perform reliable PSA of a reprocessing plant, it is important to establish the failure rate database based on the actual plant maintenance data. The Tokai Reprocessing Plant (TRP) has been operating over 30 years since 1977. Maintenance records of plant components have been accumulated and compiled in the Tokai Reprocessing Plant Maintenance Support System (TORMASS). A basic study has started in the JAEA from 2005 by using the TORMASS data to contribute the development of reliable component failure rates. The outlines of the component failure estimation are summarized.

Oral presentation

A Study on dose control for Tokai mura criticality accident termination

Kanamori, Masashi; Suto, Toshiyuki

no journal, , 

no abstracts in English

Oral presentation

Decontamination experiment for floor of Fukushima Dai-ichi Reactor Buildings, 2; Penetration behavior of simulated-contaminated water into floor and wall materials

Usuki, Toshiyuki; Sato, Isamu; Kanayama, Fumihiko; Suto, Mitsuo; Maeda, Koji; Koyama, Shinichi; Kawatsuma, Shinji; Fukushima, Mineo; Tokoro, Daishiro*; Sekioka, Ken*; et al.

no journal, , 

no abstracts in English

Oral presentation

Penetration behavior of a solution containing radioactive nuclides into concrete and epoxy resin paint

Usuki, Toshiyuki; Sato, Isamu; Kanayama, Fumihiko; Suto, Mitsuo; Maeda, Koji; Koyama, Shinichi; Kawatsuma, Shinji; Fukushima, Mineo; Tokoro, Daishiro*; Sekioka, Ken*; et al.

no journal, , 

no abstracts in English

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