Effective dose coefficients for internal exposure dose assessment in accordance with ICRP 2007 recommendations (Contract research)

Takahashi, Fumiaki; Manabe, Kentaro; Sato, Kaoru

JAEA-Review 2020-068, 114 Pages, 2021/03

JAEA-Review-2020-068.pdf:2.61MB

Radiation safety regulations have been currently established based on the 1990Recommendation by the International Commission on Radiological Protection (ICRP) in Japan. Meanwhile, ICRP released the 2007 Recommendation that replaces the 1990 Recommendation. Thus, the Radiation Council, which is established under the Nuclear Regulation Authority (NRA), has made discussions to incorporate the purpose of the 2007 Recommendation into Japanese regulations for radiation safety. As ICRP also has published effective dose coefficients for internal exposure assessment in accordance with the 2007recommendation, the technical standards are to be revised for the internal exposure assessment method in Japan. Currently, not all of the effective doses have been published to revise concentration limits for internal exposure protections of workers and public. The published effective dose coefficients are applied to radionuclides which are important in radiation protection for internal exposure of a worker. Thus, we review new effective dose coefficients as well as basic dosimetry models and data based upon Occupational Intakes of Radionuclides (OIR) parts 2, 3 and 4 that have been published from 2016 to 2019 by ICRP. In addition, issues are sorted out to provide information for revision of the technical standards for internal exposure assessment based on the 2007 Recommendations in future.

Online measurement of the atmosphere around geopolymers under gamma irradiation

Cantarel, V.; Lambertin, D.*; Labed, V.*; Yamagishi, Isao

Journal of Nuclear Science and Technology, 58(1), p.62 - 71, 2021/01

https://doi.org/10.1080/00223131.2020.1801531

The gas production of wasteforms is a major safety concern for encapsulating active nuclear wastes. For geopolymers and cements, the H produced by radiolytic processes is a key factor because of the large amount of water present in their porous structure. Herein, the gas composition evolution around geopolymers was monitored on line under Co gamma irradiation. Transient evolution of the hydrogen production yield was measured for samples with different formulations. The rate of its evolution and the final values are consistent with the presence of a chemical reaction of the pseudo-first order consuming hydrogen in the samples. The results show this phenomenon can significantly reduce the hydrogen source term of geopolymer wasteform provided their diffusion constant remains low. Lower hydrogen production rates and faster kinetics were observed with geopolymers formulations in which pore water pH was higher. Besides hydrogen production, a steady oxygen consumption was observed for all geopolymers samples. The oxygen consumption rates are proportional to the diffusion constants estimated in the modelization of hydrogen recombination by a pseudo first order reaction.

Activity report of the task group of radiation protection about wastes containing natural radioactive nuclides

Saito, Tatsuo; Kobayashi, Shinichi*; Zaitsu, Tomohisa*; Shimo, Michikuni*; Fumoto, Hiromichi*

Hoken Butsuri (Internet), 55(2), p.86 - 91, 2020/06

https://doi.org/10.5453/jhps.55.86

Safety cases for disposal of uranium bearing waste and NORM with uranium has not yet been fully developed in Japan, because of safety assessment of extraordinary long timespan and uncertainty in unexpected incidents with uncompleted radon impact evaluation measures arising from uranium waste disposal facility in far future. Our task group of radiation protection for wastes with natural radioactive nuclides studied some safety cases with disposal of uranium bearing waste and NORM in terms of nuclides, U-235, U-238 and their progenies, and comprehensively discussed the current state of their disposal in comparison to the ideas of international organizations such as ICRP and IAEA. We developed our ideas for long term uncertainty and radon with the knowledge of experts in each related area of direction, repeating discussions, focusing out the orientation of each directions, and outlined the recommendations with our suggestions of solving important issues in the future to be addressed.

Effects of -radiation on a direct disposal system for spent nuclear fuel, 2; Review of research into safety assessments of direct disposal of spent nuclear fuel in Europe and North America

Kitamura, Akira; Takase, Hiroyasu*; Metcalfe, R.*; Penfold, J.*

Journal of Nuclear Science and Technology, 53(1), p.19 - 33, 2016/01

https://doi.org/10.1080/00223131.2015.1014874

Not only geological disposal of vitrified waste generated by spent fuel (SF) reprocessing, but also the possibility of disposing of SF itself in deep geological strata (hereinafter "direct disposal of SF") may be considered in the Japanese geological disposal program. In the case of direct disposal of SF, the radioactivity of the waste is higher and the potential effects of the radiation are greater. Specific examples of the possible effects of radiation include: increased amounts of canister corrosion; generation of oxidizing chemical species in conjunction with radiation degradation of groundwater and accompanying oxidation of reducing groundwater; and increase in the dissolution rate and the solubility of SF. Therefore, the influences of radiation, which are not expected to be significant in the case of geological disposal of vitrified waste, must be considered in safety assessments for direct disposal of SF. Focusing especially on the effects of -radiation in safety assessment, this study has reviewed safety assessments in countries other than Japan that are planning direct disposal of SF. The review has identified issues relevant to safety assessment for the direct disposal of SF in Japan.

Radiation safety design for the J-PARC project

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

https://doi.org/10.1093/rpd/nci237

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.

Proceedings of the 6th IEA International Workshop on Beryllium Technology for Fusion; December 2-5, 2003, Miyazaki City, Japan

Ishitsuka, Etsuo; Kawamura, Hiroshi; Tanaka, Satoru*

JAERI-Conf 2004-006, 347 Pages, 2004/03

JAERI-Conf-2004-006.pdf:35.63MB

This report is the Proceedings of the Sixth International Energy Agency International Workshop on Beryllium Technology for Fusion. The workshop was held on December 2-5, 2003, at SEAGAIA in Miyazaki City, Japan with 69 participants who attended from Europe, the Russian Federation, Kazakhstan, Ukraine, China, the United States and Japan. The topics for papers were arranged into nine sessions; Status of beryllium study, Plasma and tritium interactions, ITER oriented issues, Neutron irradiation effects, Beryllide application, Disposal and recycling, Molten salt, Health and safety issues and Panel discussion. The issues in these topics were discussed intensively on the bases of 49 presentations. In the Panel discussion, the international collaboration for three topics, i.e., Neutron irradiation effects, Beryllide application, Recycling and Disposal, were discussed, and necessary items for the international collaboration were proposed.

Overview of nuclear safety research 2001-2002

Nuclear Safety Research Center Planning and Analysis Division

JAERI-Review 2003-040, 298 Pages, 2004/01

JAERI-Review-2003-040.pdf:24.14MB

JAERl is conducting nuclear safety research primarily at the Nuclear Safety Research Center in close cooperation with the related departments in accordance with the Annual Plan for Nuclear Safety Research issued by the Nuclear Safety Commission. The fields of conducting safety research at JAERl are the engineering safety of nuclear power plants and nuclear fuel cycle facilities, and radioactive waste management as well as advanced technology for safety improvement or assessment. This report summarizes the nuclear safety research activities of JAERI from April 2001 through March 2003.

Behavior of high-burnup PWR fuels with low-tin zircaloy-4 cladding under reactivity-initiated-accident conditions

Fuketa, Toyoshi; Sasajima, Hideo; Sugiyama, Tomoyuki

Nuclear Technology, 133(1), p.50 - 62, 2001/01

https://doi.org/10.13182/NT01-A3158

no abstracts in English

Characteristics of shielding design calculation for the SPring-8 synchrotron radiation beamlines

Asano, Yoshihiro; Sasamoto, Nobuo

Radiation Protection Dosimetry, 82(3), p.167 - 174, 1999/00

https://doi.org/10.1093/oxfordjournals.rpd.a032622

no abstracts in English

NSRR pulse irradiation experiments and tube burst tests

Fuketa, Toyoshi; Nagase, Fumihisa; Nakamura, Takehiko; Uetsuka, Hiroshi; Ishijima, Kiyomi

NUREG/CP-0166, 3, p.223 - 241, 1998/00

no abstracts in English