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Collaborative Laboratories for Advanced Decommissioning Science; Hokkaido University*
JAEA-Review 2025-028, 66 Pages, 2025/11
JAEA-Review-2025-028.pdf:3.59MB
The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2023. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station (1F), Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2023, this report summarizes the research results of the "Development of a prototype shielding-free radiation-resistant diamond neutron measurement system" conducted in FY2023. The present study aims to develop a prototype of a shielding-free neutron measurement system for 1F. The system consists of diamond neutron detectors and radiation-resistant silicon integrated circuits, and has radiation resistance of more than 10 MGy and 4 MGy, respectively, at the component level in terms of integrated dose, and has a track record of stable operation under 
-ray dose rate environment of 1.5 kGy/h. Future applications are expected to include neutron detectors for debris investigation, criticality proximity monitoring monitors, and neutron detectors for dry tube investigation in pressure vessels. In this development, a prototype consisting of 100 diamond detector elements of 5 mm square will be developed to obtain system construction technology and to evaluate system performance. In addition, a subcriticality evaluation method will be developed. This development will lead to the completion of system development, development of the actual system in collaboration with the manufacturer, and introduction of the system into 1F decommissioning project.
Collaborative Laboratories for Advanced Decommissioning Science; NAIS*
JAEA-Research 2025-004, 102 Pages, 2025/08
JAEA-Research-2025-004.pdf:3.33MB
For planning radioactive waste management at the Fukushima Daiichi Nuclear Power Station of the Tokyo Electric Power Company Holdings, Incorporated, estimation of radioactivity is essential with considering both contamination from the damaged fuel and activation during reactor operation; with regard to the latter, biological shielding is an important object due to its large amount. It is difficult to conduct field investigations or collect analysis samples at the site, hence the radioactivity should be estimated by calculative analysis with considering the actual conditions of the constituent materials, especially for activation of minor components and water, which affects the neutron flux. Besides it is important to assess the uncertainties involved in the calculation analysis. In this study, the trace composition and water content in the biological shielding concrete were investigated, and a three-dimensional computational model was constructed for the Unit 2 reactor building at the site to estimate the radioactivity concentration. In order to evaluate the uncertainty in the results, the factors contributing to the uncertainty were extracted and the uncertainty resulted from those factors on the calculation results, i.e. the influence of the diversity of the calculation model the parameters used in the calculation model. Based on the results, the dominant factors contributing to the uncertainty were extracted, and the handling as radioactive waste was discussed.
Tuya, D.; Nagaya, Yasunobu
Nuclear Science and Engineering, 198(5), p.1021 - 1035, 2024/05
Times Cited Count：1 Percentile：15.37(Nuclear Science & Technology)In Monte Carlo neutron transport calculations for local response or deep penetration problems, some estimation of an importance function is generally required in order to improve their efficiency. In this work, a new recursive Monte Carlo (RMC) method, which is partly based on the original RMC method, for estimating an importance function for local variance reduction (i.e., source-detector type) problems has been developed. The new RMC method has been applied to two sample problems of varying degrees of neutron penetrations, namely a one-dimensional iron slab problem and a three-dimensional concrete-air problem. The biased Monte Carlo calculations with variance reduction parameters based on the obtained importance functions by the new RMC method have been performed to estimate detector responses in these problems. The obtained results are in agreement with those by the reference unbiased Monte Carlo calculations. Furthermore, the biased calculations offered an increase in efficiency on the order of 1 to 10
in terms of the figure of merit (FOM). The results also indicated that the efficiency increased as the neutron penetration became deeper.
Furuta, Takuya
Igaku Butsuri, 41(4), P. 194, 2021/12
Number of medical uses of Particle and Heavy Ion Transport code System (PHITS) has been increased due to the recent high demands of medical use of radiations. The summary of such research works was described in the review article on medical application of Particle and Heavy Ion Transport code System PHITS published in Radiological Physics and Technology in 2021. There was a request from the editorial board of Japan Society of Medical Physics (JSMP) for writing an introductory article of this article in their internal journal. The research works on medical applications described in the review article, useful functions for medical application in PHITS, and newly opened user forum of PHITS have been introduced.
Furuta, Takuya; Sato, Tatsuhiko
Radiological Physics and Technology, 14(3), p.215 - 225, 2021/09
Number of the PHITS users has steadily increased since 2010 from when it is officially counted. Among them, increase of new users in medical physics is outstanding. Many research works in medical physics using PHITS have been published and the applications are widely spread in different fields such as applications to different types of radiotherapy, shielding calculations of medical facilities, application to radiation biology, and research and development of medical tools. In this article, we will introduce useful functions for medical application in PHITS by referring to examples of various medical applications.
Kochiyama, Mami; Okada, Shota; Sakai, Akihiro
JAEA-Technology 2021-010, 61 Pages, 2021/07
JAEA-Technology-2021-010.pdf:3.56MBJAEA-Technology-2021-010(errata).pdf:0.75MB
It is necessary to evaluate the radioactivity inventory in wastes in order to dispose of radioactive wastes generated from dismantling nuclear reactor in the shallow ground. In this report, we examined radioactivity evaluation method for near surface disposal about biological shield concrete near the core generated from the dismantling of JPDR. We calculated radioactive concentration of the target biological concrete using the DORT code and the ORIGEN-S code, and we estimated radioactivity concentration Di (Bq/t). For DORT calculation, the cross-section library created from the MATXSLIB-J40 file from JENDL-4.0 was used, and for ORIGEN-S, the attached library of SCALE6.0 was used. As a result of comparing the calculation results of the radioactivity concentration with the past measured values in the radial direction and the vertical direction, we found that the trends were generally the same. We calculated radioactive concentration of the target biological concrete Di (Bq/t), and we compared with the estimated Ci (Bq/t) equivalent to the dose criteria of trench disposal calculated for 140 nuclides. As a result we inferred that the except for about 2% of target waste could be disposed of in the trench disposal facility. We also preselected important nuclides for trench disposal based on the ratios (Di/Ci) for each nuclide, H-3, C-14, Cl-36, Ca-41, Co-60, Sr-90, Eu-152 and Cs-137 were selected as important nuclides.
Hashimoto, Shintaro; Sato, Tatsuhiko
EPJ Web of Conferences, 239, p.03015_1 - 03015_4, 2020/09
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Particle transport simulation codes based on the Monte Carlo technique have been successfully applied to shielding calculations in accelerator facilities. Estimation of not only statistical uncertainties, which depend on the number of trials, but also systemic uncertainties, which are caused by uncertainty of total cross section models, is required to confirm the reliability of the simulation results. We evaluated unclear quantities of internal parameters included in the total cross section model by the KALMAN code, which is based on the least squares technique, comparing with experimental data of the total cross section. The uncertainties in the total cross sections obtained by the new model are comparable to the experimental errors. In the present study, the systematic uncertainty included in the simulation results can be estimated by performing the transport calculations with variation of the internal parameters within their unclear quantities.
Collaborative Laboratories for Advanced Decommissioning Science; Kyushu University*
JAEA-Review 2019-039, 104 Pages, 2020/03
JAEA-Review-2019-039.pdf:5.57MB
The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Center of World Intelligence Project for Nuclear Science/Technology and Human Resource Development (hereafter referred to "the Project") in FY2018. The Project aims to contribute to solving problems in nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2018, this report summarizes the research results of the "Research and Development of Transparent Materials for Radiation Shield using Nanoparticles". The present study aims to reduce radiation exposure of workers in debris retrieval/analysis and reduce deterioration of optical and electronic systems in remote cameras. For these purposes, we develop transparent radiation shield by making the shield materials into nanoparticles, and dispersing/solidifying them in epoxy resin. By making B
C and W into nanoparticles, we will also develop a radiation shield that shields both neutrons and gamma-rays, and also suppresses secondary gamma-rays produced from neutrons.
Hashimoto, Shintaro; Sato, Tatsuhiko
Journal of Nuclear Science and Technology, 56(4), p.345 - 354, 2019/04
Times Cited Count：9 Percentile：59.49(Nuclear Science & Technology)Particle transport simulations based on the Monte Carlo method have been applied to shielding calculations. Estimation of not only statistical uncertainty related to the number of trials but also systematic one induced by unclear physical quantities is required to confirm the reliability of calculated results. In this study, we applied a method based on analysis of variance to shielding calculations. We proposed random- and three-condition methods. The first one determines randomly the value of the unclear quantity, while the second one uses only three values: the default value, upper and lower limits. The systematic uncertainty can be estimated adequately by the random-condition method, though it needs the large computational cost. The three-condition method can provide almost the same estimate as the random-condition method when the effect of the variation is monotonic. We found criterion to confirm convergence of the systematic uncertainty as the number of trials increases.
Maekawa, Fujio
Hamon, 28(4), p.208 - 211, 2018/11
Adequate shielding of neutrons and associated -rays is of importance from viewpoints of the radiation safety of researchers and good experimental data taking by reducing the background. This article introduces basics of neutron shielding, physics and suitable materials for neutron and -ray shielding, and an example of conceptual shielding design for the 1-MW spallation neutron source of J-PARC MLF.
Yoshida-Ouchi, Hiroko*; Matsuda, Norihiro; Saito, Kimiaki
Journal of Environmental Radioactivity, 187, p.32 - 39, 2018/07
Times Cited Count：22 Percentile：20.60(Environmental Sciences)Shinohara, Masanori; Ishitsuka, Etsuo; Shimazaki, Yosuke; Sawahata, Hiroaki
JAEA-Technology 2016-033, 65 Pages, 2017/01
JAEA-Technology-2016-033.pdf:11.14MB
To reduce the neutron exposure dose for workers during the replacement works of the startup neutron sources of the High Temperature Engineering Test Reactor, calculations of the exposure dose in case of temporary neutron shielding at the bottom of fuels handling machine were carried out by the PHITS code. As a result, it is clear that the dose equivalent rate due to neutron radiation can be reduced to about an order of magnitude by setting a temporary neutron shielding at the bottom of shielding cask for the fuel handling machine. In the actual replacement works, by setting temporary neutron shielding, it was achieved that the cumulative equivalent dose of the workers was reduced to 0.3 man mSv which is less than half of cumulative equivalent dose for the previous replacement works; 0.7 man mSv.
Shimazaki, Yosuke; Ono, Masato; Tochio, Daisuke; Takada, Shoji; Sawahata, Hiroaki; Kawamoto, Taiki; Hamamoto, Shimpei; Shinohara, Masanori
Proceedings of International Topical Meeting on Research Reactor Fuel Management and Meeting of the International Group on Reactor Research (RRFM/IGORR 2016) (Internet), p.1034 - 1042, 2016/03
In High Temperature Engineering Test Reactor (HTTR), three neutron holders containing 
Cf with 3.7 GBq for each are loaded in the graphite blocks and inserted into the reactor core as a neutron startup source which is changed at the interval of approximately ten years. These neutron holders containing the neutron sources are transported from the dealer's hot cell to HTTR using the transportation container. The holders loading to the graphite block are carried out in the fuel handling machine maintenance pit of HTTR. There were two technical issues for the safety handling work of the neutron holder. The one is the radiation exposure caused by significant movement of the container due to an earthquake, because the conventional transportation container was so large (
1240 mm, h1855 mm) that it can not be fixed on the top floor of maintenance pit by bolts. The other is the falling of the neutron holder caused by the difficult remote handling work, because the neutron holder capsule was also so long (155 mm, h1285 mm) that it can not be pulled into the adequate working space in the maintenance pit. Therefore, a new and low cost transportation container, which can solve the issues, was developed. To avoid the neutron and ray exposure, smaller transportation container (820mm, h1150 mm) which can be fixed on the top floor of maintenance pit by bolts was developed. In addition, to avoid the falling of the neutron holder, smaller neutron holder capsule (75 mm, h135 mm) with simple handling mechanism which can be treated easily by manipulator was also developed. As the result of development, the neutron holder handling work was safely accomplished. Moreover, a cost reduction for manufacturing was also achieved by simplifying the mechanism of neutron holder capsule and downsizing.
Nishitani, Takeo; Yamauchi, Michinori*; Nishio, Satoshi; Wada, Masayuki*
Fusion Engineering and Design, 81(8-14), p.1245 - 1249, 2006/02
Times Cited Count：14 Percentile：66.48(Nuclear Science & Technology)no abstracts in English
Hayashi, Takao; Tobita, Kenji; Nishio, Satoshi; Ikeda, Kazuki*; Nakamori, Yuko*; Orimo, Shinichi*; DEMO Plant Design Team
Fusion Engineering and Design, 81(8-14), p.1285 - 1290, 2006/02
Times Cited Count：29 Percentile：85.50(Nuclear Science & Technology)Neutron transport calculations were carried out to evaluate the capability of metal hydrides and borohydrides as an advanced shielding material. Some hydrides indicated considerably higher hydrogen content than polyethylene and solid hydrogen. The hydrogen-rich hydrides show superior neutron shielding capability to the conventional materials. From the temperature dependence of dissociation pressure, ZrH
and TiH can be used without releasing hydrogen at the temperature of less than 640 
C at 1 atm. ZrH and Mg(BH) can reduce the thickness of the shield by 30% and 20% compared to a combination of steel and water, respectively. Mixing some hydrides with F82H produces considerable effects in -ray shielding. The neutron and -ray shielding capabilities decrease in order of ZrH 
Mg(BH) and F82H TiH and F82H water and F82H.
Sasa, Toshinobu; Yang, J. A.*; Oigawa, Hiroyuki
Radiation Protection Dosimetry, 116(1-4), p.256 - 258, 2005/12
Times Cited Count：0 Percentile：0.00(Environmental Sciences)The proton beam duct of the accelerator-driven system (ADS) acts a streaming path for spallation neutrons and photons and causes the activation of the magnets and other devices above the subcritical core. We have performed a streaming analysis at the upper section of the lead-bismuth target/cooled ADS (800MWth). MCNPX was used to calculate the radiation dose from streamed neutrons and photons through the beam duct. For the secondary photon production calculation, cross sections for several actinides were substituted for plutonium because of the lack of gamma production cross section. From the results of this analysis, the neutron dose from the beam duct is about 20 orders higher than that of the bulk shield. The magnets and shield plug were heavily irradiated by streaming neutrons according to the DCHAIN-SP analysis.
Nakashima, Hiroshi; Nakane, Yoshihiro; Masukawa, Fumihiro; Matsuda, Norihiro; Oguri, Tomomi*; Nakano, Hideo*; Sasamoto, Nobuo*; Shibata, Tokushi*; Suzuki, Takenori*; Miura, Taichi*; et al.
Radiation Protection Dosimetry, 115(1-4), p.564 - 568, 2005/12
Times Cited Count：8 Percentile：47.35(Environmental Sciences)The High Intensity Proton Accelerator Project, named as J-PARC, is in progress, aiming at studies on the latest basic science and the advancing nuclear technology. In the project, the high-energy proton accelerator complex of the world highest intensity is under construction. In order to establish a reasonable shielding design, both simplified and detailed design methods were used in the shielding design of J-PARC. This paper reviews the present status of the radiation safety design study for J-PARC.
Liu, J. C.*; Fasso, A.*; Prinz, A.*; Rokni, S.*; Asano, Yoshihiro
Radiation Protection Dosimetry, 116(1-4), p.658 - 661, 2005/12
Times Cited Count：7 Percentile：43.50(Environmental Sciences)no abstracts in English
Yamauchi, Michinori*; Nishitani, Takeo; Nishio, Satoshi
Denki Gakkai Rombunshi, A, 125(11), p.943 - 946, 2005/11
Considering the geometrical characteristics of tokamak reactors with low aspect ratio, a basic neutronics strategy was derived to construct the inboard structure mainly for neutron shielding and produce enough tritium in the outboard blanket. The designs for optimal inboard shield were surveyed and necessary thickness was estimated to make the neutron flux low enough on the super-conducting magnet. In addition, the outer blanket designs were studied to attain the tritium breeding ratio (TBR) large enough for a self-sustaining fusion reactor on the basis of the advanced fusion reactor materials.
Morioka, Atsuhiko; Sakurai, Shinji; Okuno, Koichi*; Tamai, Hiroshi
Purazuma, Kaku Yugo Gakkai-Shi, 81(9), p.645 - 646, 2005/09
A 300 C heat-resistant neutron shielding material is newly developed, which consists of phenol-based resin with 5 weight-% boron. The neutron shielding performance of the developed resin, examined by the Cf neutron source, is almost the same as that of the polyethylene. The resin is applicable to the port section of vacuum vessel of the DD plasma device to suppress the streaming neutrons and to reduce the nuclear heating of the superconducting coils.
