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
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
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
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 BC 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
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.
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
Shinohara, Masanori; Ishitsuka, Etsuo; Shimazaki, Yosuke; Sawahata, Hiroaki
JAEA-Technology 2016-033, 65 Pages, 2017/01
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
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
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.
Liu, J. C.*; Fasso, A.*; Prinz, A.*; Rokni, S.*; Asano, Yoshihiro
Radiation Protection Dosimetry, 116(1-4), p.658 - 661, 2005/12
no abstracts in English
Sasa, Toshinobu; Yang, J. A.*; Oigawa, Hiroyuki
Radiation Protection Dosimetry, 116(1-4), p.256 - 258, 2005/12
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
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.
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.
Hamon, 15(1), p.10 - 13, 2005/01
Most parts of the 1 MW pulsed spallation neutron source JSNS are regarded as radiation shield in complicated 3-D geometry. We have developed a shielding calculation method with a particle simulation code that is based on the Monte Carlo method. The method enabled us shielding designs for the 3-D shielding structure of JSNS with high accuracy. Basic structure of JSNS was optimized by the design calculations.
Advanced Radiation Technology Center
JAERI-Review 2004-025, 374 Pages, 2004/11
This annual report describes research and development activities which have been performed with the JAERI TIARA (Takasaki Ion Accelerators for Advanced Radiation Application) facilities from April 1, 2003 to March 31, 2004. Summary reports of 115 papers and brief descriptions on the status of TIARA in the period are contained. A list of publications, the type of research collaborations and organization of TIARA are also given as appendices.
Fujimoto, Nozomu; Tachibana, Yukio; Saikusa, Akio*; Shinozaki, Masayuki; Isozaki, Minoru; Iyoku, Tatsuo
Nuclear Engineering and Design, 233(1-3), p.273 - 281, 2004/10
From a viewpoint of heat leakage, there were two incidents during HTTR power-rise-tests. One was a temperature rise of the primary upper shielding, and the other was a temperature rise of the core support plate. Causes of the both incidents were small amount of helium flow in structures. For the temperature rise of the primary upper shielding, countermeasures to reduce the small amount of helium flow, enhancement of heat release and installation of thermal insulator were taken. For the temperature rise of the core support plate, temperature evaluations were carried out again considering the small amount of helium flow and design temperature of the core support plate was revised. By these countermeasures, the both temperatures were kept below their limits.