Shimada, Taro; Sukegawa, Takenori
Journal of Nuclear Science and Technology, 52(3), p.396 - 415, 2015/03
Radiation source models in DecDose code for assessing public and worker exposure doses during the decommissioning of nuclear facilities were improved in this study. A segmentation model evaluating the length, volume, and surface area of kerfs in the object to be dismantled was improved to deal with seven shapes of objects simulating most of the components and the structures in nuclear facilities. Models for the evaluation of the external dose by direct and skyshine radiation were also improved to deal with the distribution of waste containers temporarily placed in the building and the quantity of radionuclides stored in the individual container. Good agreement was observed between actual and calculated kerf volumes in cutting some components such as the reactor pressure vessel of the Japan Power Demonstration Reactor. It is an indication of the validity of the model improved in this study. On the other hand, some discrepancies were observed between actual and calculated quantities of radionuclides discharged into the ocean, indicating the necessity of further validation of the model.
Tanaka, Tadao; Shimada, Taro; Sukegawa, Takenori
Progress in Nuclear Science and Technology (Internet), 4, p.832 - 835, 2014/04
According to a basic policy of Japan, nuclear power plant sites are allowed to be released from nuclear safety regulations after the plants are decommissioned. It is necessary to confirm that there is no significant radioactivity remaining on the sites, for the site release beforehand. Cobalt 60 is one of the typical radionuclide for nuclear power plants. In the evaluation concept, all of cobalt 60, which is in reality distributed across the area of interest, are assumed to be the single point source located at the furthest position on the surface of the area from a Ge detector. In such a configuration, minimum detectable time was supplied by Monte Carlo calculations, and the minimum detectable time was approximately equal to the actual measurement time of the point source by the Ge detector. These results mean that the proposed evaluation method was reasonable for the conservative evaluation of cobalt 60 remaining in the nuclear power plant sites.
Ishigami, Tsutomu; Sukegawa, Takenori*; Mukai, Masayuki
JAEA-Technology 2013-027, 124 Pages, 2013/10
In order to safely and efficiently implement decommissioning of nuclear installations, it is important to beforehand predict decommissioning project management data (PMD) and to develop a decommissioning plan based on the predicted results. The PMD prediction is made with PMD evaluation equations including model parameters such as unit work activity coefficients. Although model parameter values developed so far include uncertainties, little evaluation of the uncertainties and resulted uncertainties in predicted PMD has been made. However information on the uncertainties is valuable in flexibly studying and developing a decommissioning plan. We therefore studied and evaluated uncertainties in model parameters by analyzing the JPDR decommissioning experience data. This report describes an evaluation method of the model parameter uncertainties and their evaluated results.
Ishigami, Tsutomu; Mukai, Masayuki; Sukegawa, Takenori; Matsubara, Takeshi*
JAEA-Data/Code 2012-023, 83 Pages, 2012/11
Verification for site release is one of procedures to confirm termination of decommissioning of nuclear installations. The verification procedure would need to confirm that the radioactive concentration at the site is lower than the criterion value by measurement. Then to efficiently perform the measurement and verification it is one of important issues how to efficiently estimate and evaluate overall spatial radioactivity distribution using a sampling method. For the efficient estimation and evaluation we have applied a Kriging technique which in the geostatistics, and have developed a computer program ESRAD (Estimation of Spatial RadioActivity Distribution). The ESRAD program is designed to support sample selection, calculate a variogram, and estimate a radioactivity distribution for the area concerned. This report describes the Kriging technique, structure and functions of ESRAD, input file format, output examples, execution procedure of ERSAR, and sample run with ESRAD.
Sukegawa, Takenori; Shimada, Taro; Ito, Takeshi*; Tanaka, Tadao
JAEA-Technology 2011-025, 41 Pages, 2011/09
Nuclear facility sites after decommissioning are allowed to be released from nuclear safety regulations after confirming that sites have been decontaminated to acceptable levels. In-situ measurement with the use of a portable Ge detector is a suitable technology for confirmatory survey. A conservative method to evaluate residual radioactivity was proposed in this study. In the evaluation method concept, all of the radionuclide, which are in reality distributed across the area of interest, is assumed to be the single point source located at the furthest position of the area from a Ge detector. Based on this assumption, the detectable minimum time of the interest radionuclide were predicted by the calculation. If radiation from the point source is not detected for longer than the predicted detectable time, it can be proven that the radioactivity remaining in the interest area is lower than the radioactivity corresponding to the assumed point source. Results of the field test in JAEA site indicate that the proposed method was reasonable for the conservative evaluation of residual radioactivity.
Tanaka, Tadao; Shimada, Taro; Ito, Takeshi*; Hirano, Takahiro*; Sukegawa, Takenori
Progress in Nuclear Science and Technology (Internet), 1, p.408 - 411, 2011/02
Nuclear power plant sites are allowed to be released from nuclear safety regulations after the plants are decommissioned in Japan. The regulatory compliance will require confirming that there is no significant radioactivity remaining on the sites for the site release. In the present study, we propose an evaluation method of Cs-137 remaining on sites of decommissioned nuclear power plants. The method is time-efficient and gives a conservative result. In the evaluation method concept, all of the Cs-137, which is in reality distributed across the area of interest, is assumed to be the single point source located at the furthest position on the surface of the area from a detector. For such a configuration, the counting time that the Cs-137 point source is detectable is predicted using Monte Carlo calculations. If radiation from the Cs-137 point source is not detected for longer than the predicted counting time, it can be proven that the radioactivity remaining on the surface of the area is lower than the radioactivity corresponding to the assumed Cs-137 point source. A Cs-137 radiation source was placed at a fixed distance from the Ge detector, and the peak counting rate of Cs-137 were measured. The detectable time predicted by Monte Carlo calculations was approximately equal to the actual measurement time by the Ge detector, which means that the proposed evaluation method was reasonable for the conservative evaluation of remaining radioactivity.
Tanaka, Tadao; Shimada, Taro; Ito, Takeshi*; Sukegawa, Takenori
Proceedings of 13th International Conference on Environmental Remediation and Radioactive Waste Management (ICEM 2010) (CD-ROM), p.551 - 557, 2010/10
-ray from the progeny radionuclides of U-238, such as Th-234, Pa-234m and Ra-226, has been utilized for the evaluation of uranium concentration. In the present study, we proposed an evaluation method for U-238 concentration of background level in environment and for probate of vast land areas, in which the -ray radiations from Th-234, Pa-234m, Ra-226 is measured with the portable Ge detector. The U-238 concentration estimated by the in-situ metrology with portable Ge detector was in the order of 0.01 Bq/g in radioactive concentration, and was in comparable level with the concentrations decided by the ICP-MS established as high sensitive uranium analytical method. The method may be available for the U-238 concentration determination in vast land areas for the site release after decommissioning nuclear fuel handling facilities.
Sukegawa, Takenori; Shimada, Taro; Katsurai, Kiyomichi; Tanaka, Tadao; Nakayama, Shinichi
JAEA-Review 2009-075, 86 Pages, 2010/03
In the field of safety regulation systems for nuclear facilities after completion of operation, criteria of residual radioactivities and confirmation and verification procedures for termination of decommissioning are important problems that should concretely be made a study. Safety standards and criteria in IAEA, USA, etc., and practical examples of site release of power reactors in USA were studied, therefore, problems for introducing the regulation system in Japan were discussed. In this report, final status survey of Trojan power plant were investigated as a particular case of site release, and, concept of specifying survey areas to be measured radioactivities and demonstrated to compliance with release criteria was discussed. In addition, the idea of confirmation and verification procedure for termination of decommissioning in Japan was proposed referring to the US guidance (MARSSIM).
Shimada, Taro; Oshima, Soichiro*; Sukegawa, Takenori
Journal of Power and Energy Systems (Internet), 4(1), p.40 - 53, 2010/02
A safety assessment code, DecDose, for decommissioning of nuclear facilities has been developed, based on the experiences of the decommissioning project of Japan Power Demonstration Reactor (JPDR) at Japan Atomic Energy Research Institute DecDose evaluates the annual exposure dose of the public and workers according to the progress of decommissioning, and also evaluates the public dose at accidental situations including fire and explosion. As for the public, both the internal and the external doses are calculated by considering inhalation, ingestion, direct radiation from radioactive aerosols and radioactive depositions, and skyshine radiation from waste containers. For external dose for workers, the dose rate from contaminated components and structures to be dismantled is calculated. Internal dose for workers is calculated by considering dismantling conditions, e.g. cutting speed, cutting length of the components and exhaust velocity. DecDose was partially verified by comparison with the actual external dose of workers which were acquired during JPDR Decommissioning Project.
Shimada, Taro; Oshima, Soichiro; Sukegawa, Takenori
Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 8 Pages, 2009/07
A safety assessment code, DecDose, for decommissioning of nuclear facilities has been developed, based on the experiences of the decommissioning project of Japan Power Demonstration Reactor (JPDR). DecDose evaluates the annual exposure dose of the public and workers according to the progress of decommissioning of the plant, and also evaluates the public dose at accidental situations including fire and explosion. The DecDose is expected to contribute to utilities in formulating rational dismantling plans and to the safety authority in estimating conservativeness in safety assessment of licensing application or risk-based regulatory criteria.
Sukegawa, Takenori; Shimada, Taro; Shiraishi, Kunio; Tachibana, Mitsuo; Ishigami, Tsutomu
JAEA-Data/Code 2008-009, 57 Pages, 2008/03
We developed the RADO code system for evaluating residual radioactive inventory in decommissioning of nuclear reactor. The code system consists of computer programs which calculate macroscopic effective cross section, neutron flux, and radioactive inventory. This report describes an evaluation method of radioactive inventory, structure and functions of RADO, input and output of RADO, and sample run with RADO.
Mizukoshi, Seiji; Sukegawa, Takenori
Dekomisshoningu Giho, (34), p.26 - 39, 2006/09
Technical information including available dismantling technologies and residual radioactive contamination is required to establish safety standards for the regulatory review for the future decommissioning of nuclear fuel cycle facilities. We have acquired the technical information on the past and on-going decommissioning projects of domestic and foreign uranium enrichment facilities and spent fuel reprocessing facilities. This report introduces the information and the study on the safety issues concerning the decommissioning.
Shiraishi, Kunio; Sukegawa, Takenori; Ishigami, Tsutomu
JAEA-Data/Code 2005-002, 162 Pages, 2006/01
In order to efficiently develop a decommissioning plan of a nuclear facility, it is useful to develop a database containing information on decommissioning technology, cost and risk analysis results, and decommissioning projects such as the JPDR decommissioning project by collecting the information systematically. A decommissioning database has been developed by collecting decommissioning related information and analyzing it. The database provides information on not only data of decommissioning technology and decommissioning projects but also laws and safety standards for decommissioning in each country and international organizations. The database is released in a Homepage on Web and is available for use via intranet with functions of retrieval, display and printing.
Oshima, Soichiro; Shiraishi, Kunio; Shimada, Taro; Sukegawa, Takenori; Yanagihara, Satoshi
JAERI-Tech 2005-046, 46 Pages, 2005/09
A model for estimating decommissioning costs consisting of labor cost, device cost and expense, was developed for items which OECD/NEA had standardized, and was installed into the computer system for planning and management of reactor decommissioning (COSMARD). Input data files and databases for the decommissioning of JPDR were prepared, and the decommissioning cost was calculated with COSMARD. In addition, the decommissioning cost for a large scale BWR power plant was also calculated on the assumption of the advantage of scale. The calculations have shown that it is useful and efficient for studying the decommissioning costs for nuclear reactors to apply the COSMARD with database for cost estimation to the decommissioning cost calculation.
Akutsu, Atsushi; Kishimoto, Katsumi; Sukegawa, Takenori; Shimada, Taro
JAERI-Tech 2003-090, 75 Pages, 2004/01
The Japan Research Reactor No.1 (JRR-1) that was constructed first in Japan was permanently shut down after operation from 1957 to 1968. At present, the reactor part is in safe store conditions. The JRR-1 facility is being used as an exhibition room for the time being, and will be dismantled in the future. In consideration of future dismantling of the facility, the radioactive inventory in reactor part was calculated using computer codes that are Two-Dimensional Discrete Ordinates Transport Code (DORT) and Oak Ridge Isotope Generation and Depletion Code (ORIGEN-MD). The average concentration of radioactivity is estimated to be 6.40 Bq/g in the core tank as of April, 2002. It is also expected that the low level waste (LLW) weights approximately 400kg and very low level waste (VLLW) weights approximately 14,000kg, and the waste which doesn't need to deal as a radioactive material weights approximately 250,000kg.
Shimada, Taro; Sukegawa, Takenori; Yanagihara, Satoshi; Sato, Tadamichi*; Sakai, Shinichi*
Proceedings of 9th Biennial International Conference on Nuclear and Hazardous Waste Management (Spectrum '02) (CD-ROM), 6 Pages, 2002/08
no abstracts in English
Oshima, Soichiro; Sukegawa, Takenori; Shiraishi, Kunio; Yanagihara, Satoshi
JAERI-Tech 2001-086, 83 Pages, 2001/12
Project management data on dismantling the Japan Power Demonstration Reactor (JPDR) was calculated using the Code System for Planning and Management of Reactor Decommissioning (COSMARD), and then its validity was studied by comparing the calculation results with actual data. In addition, work breakdown structure models and database were modified to meet an evaluation with changing work difficulty of preparation and cleanup activities, and calculations were further conducted to analyze feasibility by changing various conditions on cutting and conditioning activities. As the results, COSMARD was verified to be useful by confirming calculation capability on reflection of actual work conditions and relatively good agreement between actual data and calculations. Moreover, it was cleared that main parameters such as work difficulty of preparation and cleanup activities and the cutting speed in demolition work could affect to manpower within 30% in each calculations.
Shiraishi, Kunio; Sukegawa, Takenori; Yanagihara, Satoshi
JAERI-Data/Code 2001-028, 86 Pages, 2001/11
The data on worker dose in dismantling of the Japan Power Demonstration Reactor (JPDR) was analyzed to characterize its features. It was appeared from the study that the collective dose to the workers was 306 man-mSv, in which maximum individual dose was 8.5 mSv, almost all doses were received in the activities for dismantling of reactor internals, the reactor pressure vessel and the biological shield, and that the worker dose distribution was similar to that in the maintenanee of the facilities which was characterized by the hybrid log normal distribution model. Farthermore, it was found that the dismantling activities were categorized into three groups depending on dose rates in workplaces, then contribution factors for radiation exposure in terms of dose rates in different groups were derived based on the analysis. The study would be useful for estimation of worker dose in future decommissioning of commercial nuclear power plants in Japan.
Sukegawa, Takenori; Hatakeyama, Mutsuo; Yanagihara, Satoshi
JAERI-Tech 2001-058, 81 Pages, 2001/09
In general, neutron transport and activation calculation codes are used for residual radioactive inventory estimation; however, it is essential to verify calculations by measurement results because of geometrical complexity of the reactor and so on. The comparison between measured and calculated radioactivity in the JPDR core components showed a relatively good agreement (factor of 2), and it was cleared that water content and weight ratio of steel bars to concrete materials significantly influenced the neutron flux distribution in the biological shield (factor of 2-10 error). The measured radioactivity inside of the reactor pressure vessel wall and at the inner part of the biological shield was compared in detail with the calculations to verify the methodology applied to calculations of radioisotope production. Then it was found that the radioactive inventory could be estimated accurately with combination of calculations and measurement of radioactivity in samples and dose rate distribution for planning of dismantling activities.
Yanagihara, Satoshi; Oshima, Soichiro; Sukegawa, Takenori; Tanabe, Norio*; Takaya, Junichi*; Kiuchi, Yoshio*; Yokota, Shuichi*
Nihon Genshiryoku Gakkai-Shi, 43(5), p.493 - 502, 2001/05
no abstracts in English