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Shimada, Kazumasa; Sasaki, Toshihisa*; Iijima, Masashi*; Munakata, Masahiro
JAEA-Research 2018-012, 68 Pages, 2019/02
JAEA-Research-2018-012.pdf:4.15MB
The external exposure dose of off-site emergency responders at Fukushima Dai-ichi Nuclear Power Station accident were evaluated in order to consider a radiation protection of emergency responders. The maximum value of individual daily dose of emergency responders whose activities details were recorded from 12th to 31th March 2011 was 650 
Sv engaged in evacuation support in Futaba Town on 12th. Next, atmospheric concentrations and deposition of radionuclides were calculated from the source terms estimated by previous studies using atmospheric diffusion and deposit calculation codes, and air dose rates at off-site were estimated. Then, the external exposure dose was calculated for 6 emergency responders whose daily activities and personal doses were continuously recorded. Furthermore, the maximum value and the average value of the calculation external dose of emergency responders in the activity area were compared with the measurement value of the personal dosimeter of them. These results showed that the calculated value of the external exposure dose of emergency responders calculated from the maximum value of the dose rate in the active area roughly includes the measured value.
Sugaya, Toshikatsu; Nakatani, Takayoshi; Sasaki, Toshihisa*; Nakamura, Yasuo*; Sakai, Akihiro; Sakamoto, Yoshiaki
JAEA-Technology 2016-036, 126 Pages, 2017/02
JAEA-Technology-2016-036.pdf:7.28MB
At the Radioactive Waste Management and Disposal Project Department Sector of Decommissioning and Radioactive Waste Management, we performed the technological study about the disposal measures of the low-level radioactive waste targeted for uranium-bearing waste and intermediate depth disposal-based waste occurring from the process of the nuclear fuel cycle.
Takebe, Shinichi; Sasaki, Toshihisa; Saito, Tatsuo; Yamaguchi, Naoko
JAEA-Technology 2013-033, 87 Pages, 2013/11
JAEA-Technology-2013-033.pdf:2.49MB
Materials, etc. in a non-controlled area is, if the criteria (10 microsieverts per year) listed below that in "A guideline regarding treatment of materials in nuclear facilities considering the influence of fallout released from the accident of TEPCO's Fukushima Dai-ichi Nuclear Power Station" (March 30, 2012), that in accordance with relevant laws and regulations, such as "Waste Disposal and Cleaning Act" (Act No. 137 of 1970), effective use as a resource or be properly disposed of is required. In this paper, in order to effectively use as resources or properly disposed of the materials, etc. in a non-controlled area, radioactivity concentration of materials, etc. in you see " For clearance level specified in Radiation Hazards Prevention Law" and "Radionuclide Concentrations for Materials not Requiring Treatment as Radioactive Wastes Generated from Dismantling etc. of Reactor Facilities and Nuclear Fuel Use Facilities (in Japanese), NSC Japan, 2005." corresponds to the dose of the above criteria the presented results have been estimated as an example.
Takeda, Seiji; Sawaguchi, Takuma; Sasaki, Toshihisa*; Kimura, Hideo
JAEA-Data/Code 2011-008, 166 Pages, 2011/08
JAEA-Data-Code-2011-008.pdf:3.33MB
The upper bound of radioactivity concentration during disposal means the maximum concentration of radionuclides of waste repository allowable in a license application for that repository. For transuranium (TRU) and uranium wastes, it is necessary to amend the law to set the upper bound of radioactivity concentration for three concepts to dispose of low-level waste categories: near surface disposal without an artificial barrier (trench disposal), near surface disposal with an artificial barrier (concrete vault disposal), and intermediate depth disposal. We developed an assessment code (GSA-GCL ver.2) to derive the upper bound of radioactivity concentration for TRU and uranium wastes, according to geological and artificial barrier features in three disposal concepts. This document summaries descriptions of the capabilities and structure of GSA-GCL ver.2 code, mathematical models, user information for execution of the code system (users' manual), input and output file system and the result of verification for the models in this code. We provide the evaluated result of the upper bounds of radioactivity concentration for radionuclides in TRU wastes, which reflects to the regulation report on the result for three concepts to dispose of low-level waste categories, reviewed by Nuclear Safety Commission of Japan in May 2007.
Takeda, Seiji; Sasaki, Toshihisa*; Nagasawa, Hirokazu; Kimura, Hideo
JAEA-Data/Code 2010-019, 61 Pages, 2010/11
JAEA-Data-Code-2010-019.pdf:1.47MB
In deep geological disposal or intermediate depth disposal, human intrusion, i.e. accidental excavation or drilling into the disposal site, may make a direct or indirect effect on the disposal system. Assessment of the human intrusion would need the examination of institutional control to reduce the probability of the intrusion occurring, the estimation of its probability, and the development of method to estimate its associated exposure of persons, as well as some foreign countries. Assuming that drilling action into the disposal site especially leads to the human proximity to the radioactive waste or the damage to the barrier system (drilling scenario), we have developed the evaluation code of radiological effect from the human intrusion into radioactive waste disposal system (HUINT). The evaluation code of human intrusion scenario, HUINT, can calculate the radiation exposure dose for the workers of a series of actions accompanied with drilling and for the workers and public people concerned with the reuse of drilling cores, identified with the actual information on drilling action in Japan. This report provides the descriptions of mathematical models on the drilling scenario, input specification and user information for execution of HUINT (user manual), and the result of verification for calculation with the models in HUINT.
Nagasawa, Hirokazu; Takeda, Seiji; Kimura, Hideo; Sasaki, Toshihisa*
JAEA-Data/Code 2010-018, 85 Pages, 2010/11
JAEA-Data-Code-2010-018.pdf:1.94MB
In deep geological disposal or intermediate depth disposal, human intrusion, i.e. accidental excavation or drilling into the disposal site, may make a direct or indirect effect on the disposal system. Assessment of the human intrusion would need the examination of institutional control to reduce the probability of the intrusion occurring, the estimation of its probability, and the development of method to estimate its associated exposure of persons, as well as some foreign countries. Assuming that drilling action into the disposal site especially leads to the human proximity to the radioactive waste or the damage to the barrier system (drilling scenario), we have collected both the data on borehole drilling implemented in Japan and information on actual situation of drilling activities. We have developed an assembly of the information mentioned above as database, including the model parameters used in the code to assess radiation dose for drilling scenario.
Takeda, Seiji; Yamaguchi, Tetsuji; Nagasawa, Hirokazu; Watanabe, Masatoshi; Sekioka, Yasushi; Kanzaki, Yutaka; Sasaki, Toshihisa; Ochiai, Toru; Munakata, Masahiro; Tanaka, Tadao; et al.
JAEA-Research 2009-034, 239 Pages, 2009/11
JAEA-Research-2009-034.pdf:33.52MB
In safety assessment for geological disposal of high level radioactive waste, it is of consequence to estimate the uncertainties due to the long-term frame associated with long-lived radionuclides and the expanded geological environment. The uncertainties result from heterogeneity intrinsic to engineered and natural barrier materials, insufficient understanding of phenomena occurring in the disposal system, erroneous method of measurement, and incomplete construction. It is possible to quantify or to reduce the uncertainties according to scientific and technological progress. We applied a deterministic and a Monte Carlo-based probabilistic method simulation techniques to the uncertainty analysis for performance of hypothetical geological disposal system for high level radioactive waste. This study provides the method to evaluate the effects of the uncertainties with respect to scenarios, models and parameters in engineering barrier system on radiological consequence. The results also help us to specify prioritized models and parameters to be further studied for long-term safety assessment.
Sawaguchi, Takuma; Takeda, Seiji; Sasaki, Toshihisa; Ochiai, Toru; Watanabe, Masatoshi; Kimura, Hideo
JAEA-Research 2008-046, 62 Pages, 2008/03
JAEA-Research-2008-046.pdf:3.34MB
The Atomic Energy Commission of Japan states that the transuranium waste and uranium waste are to be disposed of by either near surface disposal without artificial barrier (trench disposal), near surface disposal with artificial barrier (concrete vault disposal), or intermediate depth disposal, depending on the radionuclides and their radioactivity concentration in the wastes. The ranges of radioactivity concentration for these different disposal concepts are, therefore, required to be determined for the regulation. The radioactivity concentration limits define the upper bound of radioactivity concentrations for licensing application of the disposal of radioactive waste. This document summaries the concept and method for estimation of the radioactivity concentration limits for concrete vault disposal of transuranium and uranium wastes, and provides the derived values of the radioactivity concentration limits for each radionuclide in the wastes.
Takeda, Seiji; Sasaki, Toshihisa; Sawaguchi, Takuma; Ochiai, Toru; Kimura, Hideo
JAEA-Research 2008-045, 60 Pages, 2008/03
JAEA-Research-2008-045.pdf:3.51MB
This document summaries the concept and method (scenario selection, model/code description and parameter selection) for estimation of the radioactivity concentration limits for intermediate depth disposal of transuranium and uranium wastes, and provide the derived values of the radioactivity concentration limit for each radionuclide in the wastes. The values for the transuranium waste are published in a Nuclear Safety Commission of Japan report.
Takeda, Seiji; Watanabe, Masatoshi; Sawaguchi, Takuma; Sasaki, Toshihisa; Ochiai, Toru; Kimura, Hideo
JAEA-Research 2008-044, 64 Pages, 2008/03
JAEA-Research-2008-044.pdf:5.21MB
This document summaries the concept and method (scenario selection, model/code description and parameter selection) for estimation of the radioactivity concentration limits for trench disposal of transuranium and uranium wastes, and provide the derived values of the radioactivity concentration limit for each radionuclide in the wastes. The values for the transuranium waste are published in a Nuclear Safety Commission of Japan report.
Sasaki, Toshihisa; Watanabe, Masatoshi; Takeda, Seiji; Sawaguchi, Takuma; Ochiai, Toru; Kimura, Hideo
JAEA-Data/Code 2008-003, 29 Pages, 2008/02
JAEA-Data-Code-2008-003.pdf:3.7MBJAEA-Data-Code-2008-003(errata).pdf:0.04MB
In this report, external effective dose conversion factors necessary for examining the activity concentration limits are derived for three disposal concepts. After this, the activity concentration limits that constitute a permissible range of radioactive concentration to typical land disposal concept (for radioactive wastes containing transuranic nuclides from reprocessing and MOX fuel manufacturing and uranium waste from enrichment and fuel manufacturing) are calculated. External effective dose conversion factors are derived in consideration with analysis that conforms to laws that use the conversion coefficients of ICRP Publication 74 for effective dose conversion, and adoption of the latest data i.e. 
-ray's and X-ray's energies and intensities of "JAERI-Data/Code 2001-004" as photon energy data. This document summarizes calculation method, conditions, and results of external effective dose conversion factors for transuranium and uranium wastes disposal.
Takeda, Seiji; Kanno, Mitsuhiro*; Sasaki, Toshihisa*; Minase, Naofumi*; Kimura, Hideo
JAEA-Data/Code 2006-003, 137 Pages, 2006/02
JAEA-Data-Code-2006-003.pdf:7.4MB
no abstracts in English
Amano, Hikaru; Ikeda, Hiroshi*; Sasaki, Toshihisa*; Matsuoka, Shungo*; Kurosawa, Naohiro*; Takahashi, Tomoyuki*; Uchida, Shigeo*
KEK Proceedings 2003-11, p.239 - 244, 2003/11
A Code MOGRA (Migration Of GRound Additions) is a migration prediction code for toxic ground additions including radioactive materials in a terrestrial environment, which consists of computational codes that are applicable to various evaluation target systems, and can be used on personal computers. The computational code has the dynamic compartment analysis block at its core, the graphical user interface (GUI) for model formation, computation parameter settings, and results displays. The code MOGRA has varieties of databases, which is called MOGRA-DB. Another additional code MOGRA-MAP can take in graphic map and calculate the square measure about the target land.
Yokoyama, Sumi; Noguchi, Hiroshi; Ryufuku, Susumu*; Sasaki, Toshihisa*; Kurosawa, Naohiro*
JAERI-Data/Code 2002-022, 87 Pages, 2002/11
JAERI-Data-Code-2002-022.pdf:4.26MB
Tritium, which is used as a fuel of a D-T burning fusion reactor, is the most important radionuclide for the safety assessment of a nuclear fusion experimental reactor such as ITER. Thus, a computer code, ACUTRI, which calculates the radiological impact of tritium released accidentally to the atmosphere, has been developed, aiming to be of use in a discussion on licensing of a fusion experimental reactor and an environmental safety evaluation method in Japan. ACUTRI calculates an individual tritium dose based on transfer models specific to tritium in the environment. A Gaussian plume model is used for calculating the atmospheric dispersion of tritium gas (HT) and/or tritiated water (HTO). The environmental pathway model in ACUTRI considers the following internal exposures: inhalation from a primary plume (HT and/or HTO) released from the facilities and inhalation from a secondary plume (HTO) reemitted from the ground following deposition of HT and HTO. This report describes an outline of the ACUTRI code, a user guide and the results of test calculation.
Kuroda, Toshimasa*; Sato, Kazuyoshi; Akiba, Masato; Ezato, Koichiro; Enoeda, Mikio; Osaki, Toshio*; Kosaku, Yasuo; Sato, Satoshi; Sato, Shinichi*; Suzuki, Satoshi*; et al.
JAERI-Tech 2002-044, 25 Pages, 2002/03
JAERI-Tech-2002-044.pdf:2.68MB
no abstracts in English
; 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.
; 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.
Nagasawa, Hirokazu; Takeda, Seiji; Kimura, Hideo; Sasaki, Toshihisa
no journal, ,
no abstracts in English
Kimura, Hideo; Takeda, Seiji; Sasaki, Toshihisa; Ochiai, Toru
no journal, ,
no abstracts in English
Nagasawa, Hirokazu; Takeda, Seiji; Kimura, Hideo; Sasaki, Toshihisa
no journal, ,
no abstracts in English