Initiatives toward final disposal project related to low level radioactive waste generated from research institutions, universities, private institutions, medical institutions, etc.

Nakata, Hisakazu

Genkan Senta Topikkusu, (147), p.4 - 11, 2023/10

https://www.rwmc.or.jp/library/file/Topics_No147.pdf

no abstracts in English

Common evaluation procedure radioactivity concentration by theoretical calculation for radioactive waste generated from the decommissioning of research reactors

Okada, Shota; Murakami, Masashi; Kochiyama, Mami; Izumo, Sari; Sakai, Akihiro

JAEA-Testing 2022-002, 66 Pages, 2022/08

JAEA-Testing-2022-002.pdf:2.46MB

Japan Atomic Energy Agency is an implementing organization of burial disposal for low-level radioactive waste generated from research, industrial and medical facilities in Japan. Radioactivity concentrations of the waste are essential information for design of the disposal facility and for licensing process. A lot of the waste subjected to the burial disposal is arising from dismantling of nuclear facilities. Radioactive Wastes Disposal enter has therefore discussed a procedure to evaluate the radioactivity concentrations by theoretical calculation for waste arising from the dismantling of the research reactors facilities and summarized the common procedure. The procedure includes evaluation of radioactive inventory by activation calculation, validation of the calculation results, and determination of the disposal classification as well as organization of the data on total radioactivity and maximum radioactivity concentration for each classification. For the evaluation of radioactive inventory, neutron flux and energy spectra are calculated at each region in the reactor facility using two- or three-dimensional neutron transport code. The activation calculation is then conducted for 140 nuclides using the results of neutron transport calculation and an activation calculation code. The recommended codes in this report for neutron transport calculation are two-dimensional discrete ordinate code DORT, three-dimensional discrete ordinate code TORT, or Monte Carlo codes MCNP and PHITS, and for activation calculation is ORIGEN-S. Other recommendation of cross-section libraries and calculation conditions are also indicated in this report. In the course of the establishment of the procedure, Radioactive Wastes Disposal Center has discussed the commonly available procedure at meetings. It has periodically held to exchange information with external operators which have research reactor facilities. The procedure will properly be reviewed and be revised by reflecting future situ

Think of issues for treatment various waste from nuclear plant

Suzuki, Masafumi*; Yoshinaka, Kazuyuki

Gijutsushi, (648), p.12 - 15, 2020/12

AA2020-0333.pdf:0.97MB

The site tour which is a part of CPD events regarding the issue concerning the radioactive waste from nuclear facilities was organized in autumn 2019, and we visited the facilities related to the specified waste due to the Fukushima-Daiichi Nuclear Power Plant accident in spring 2020. The tours made us reacknowledge the impact of radionuclides on environment at accident, the importance of the management of nuclear facilities under the basic premise of ensuring safety, the necessity of the discussion based on the common understanding, and that would be contributed to find the solution of those issues.

Study on the basic system of the common non-destructive radioactivity measuring equipment for disposal of radioactive wastes generated from research, industrial and medical facilities

Izumo, Sari; Hayashi, Hirokazu; Nakata, Hisakazu; Amazawa, Hiroya; Motoyama, Mitsushi*; Sakai, Akihiro

JAEA-Technology 2018-018, 39 Pages, 2019/03

JAEA-Technology-2018-018.pdf:2.8MB

JAEA has planed the near surface disposal of LLW generated from research, industrial, and medical facilities. Maximum radioactivity concentration of each waste and total radioactivity of disposed wastes are needed to be less than the permitted values in the license of disposal facility. Thus, it is important not to evaluate the radioactivity of each waste in unduly conservative ways so as to dispose of the total amount of the waste that is originally planned. Accordingly, the detection limit is required to be as low as the clearance level for the very low level radioactive waste planned to be disposed of trench-type. In this report, the feasibility of the non-destructive assay method is studied by model calculations for gamma emitters. It is confirmed that the detection limit less than the clearance level can be achieved as regards the box type metal container that is difficult to measure. This report summarizes the requirements for the non-destructive measuring equipment.

Development of waste acceptance criteria and current challenges relating to the disposal project of LLW generated in research, medical and industrial facilities

Nakata, Hisakazu; Amazawa, Hiroya; Izumo, Sari; Okada, Shota; Sakai, Akihiro

Dekomisshoningu Giho, (58), p.10 - 23, 2018/09

Low level radioactive wastes are generated in the research and development of the nuclear energy, medical and industrial use of radioisotope except NPP in Japan. The disposal of wastes arising from NPP has already been implemented while not the one for wastes from research institutes etc. Japan Atomic Energy Agency therefore has been assigned an implementing organization for the disposal legally in 2008 in order to promote the disposal program as quickly and firmly as possible. Since then, JAEA has conducted their activity relating to the disposal facility design on generic site conditions and developing Waste Acceptance Criteria for LLW from research institutes. This report summarizes the WAC and current challenges.

Waste Technical Standards Working Group annual report 2016

Waste Technical Standards Working Group

JAEA-Review 2017-017, 112 Pages, 2017/11

JAEA-Review-2017-017.pdf:2.87MB

In Japan Atomic Energy Agency, JAEA, a Waste Technical Standards Working Group has established since FY2015. The Working Group is composed of the members from waste management sections in each site in JAEA and from Radioactive Waste Management and Disposal Project Department. In this Working Group, we discussed quality management on conditioning waste packages, methodologies to evaluate the radioactivity concentration and measures for dismantling waste. This annual report summarizes the results of discussion in FY2016.

Demonstrative experiments on the migration of radiocesium from buried soil contaminated by the accident at Fukushima Daiichi Nuclear Power Station

Yamaguchi, Tetsuji; Shimada, Taro; Ishibashi, Makoto*; Akagi, Yosuke*; Kurosawa, Mitsuru*; Matsubara, Akiyoshi*; Matsuda, Yuki*; Sato, Shigeyoshi*

Genshiryoku Bakkuendo Kenkyu (CD-ROM), 22(2), p.21 - 27, 2015/12

https://doi.org/10.3327/jnuce.22.2_21

It is predictable from previous studies that radiocesium hardly migrate into surrounding soils and groundwater from soils contaminated by the Fukushima Daiichi Nuclear Power Plant accident if they are buried and covered with indigenous soils. This study demonstrated the prediction by performing in-situ migration experiments over a year in a public park in Miho, Ibaraki prefecture and in two public parks in Misato, Saitama prefecture. Contaminated soils were buried at a depth range of 0.3 - 1.0 m or at 0.3 - 1.3 m and covered with indigenous soil layer of 0.3 m, and were sprinkled with water to accelerate the radiocesium migration. Migration of radiocesium was not observed from radiometric analyses of boring cores and soil water samples. Laboratory column and sorption experiments revealed that the radiocesium hardly leach out of the soil and even if they leach out from the contaminated soil, radiocesium is sorbed on surrounding soils and hardly migrate through the soli layer. Simulation of Cs-137 migration for 100 years by an advection-diffusion model showed that Cs-137 hardly migrate and decay out in the contaminated soil.

Data on mechanical properties of rocks in Japan for subsurface disposal of radioactive waste

Hagiwara, Shigeru*; Sakamoto, Yoshiaki*; Takebe, Shinichi; Nakayama, Shinichi

JAERI-Data/Code 2004-004, 167 Pages, 2004/03

JAERI-Data-Code-2004-004.pdf:10.14MB

no abstracts in English

Derivation of clearance levels for solid materials in Japan

Sakai, Akihiro; Okoshi, Minoru

Radiation Risk Assessment Workshop Proceedings, p.175 - 186, 2003/00

To establish the clearance levels, the Nuclear Safety Commission (NSC) has been discussing the clearance levels since May 1997. The NSC derived the unconditional clearance levels for the solid materials, namely concrete and metal, arising from the operation and dismantling of nuclear reactors and post irradiation examination (PIE) facilities. Two destinations of the cleared materials, namely disposal and recycle/reuse, were considered. Deterministic calculation models were established to assess individual doses resulting from 73 exposure pathways, and realistic parameter values were selected considering the Japanese natural and social conditions. The clearance levels for 21 radionuclides of nuclear reactors and for 49 of PIE facilities were derived as radioactivity concentration equivalent to the individual doses of 10 Sv/y. Most of calculated clearance levels were nearly the same as those shown in IAEA-TECDOC-855. Some, however, were different. It is considered that the major reasons depend on differences of fixed scenarios and of selected values of parameters.

Long-term effect of creep displacement of host-rock on stability of engineered barrier system for TRU waste; Two-dimensional analysis by the non-linear viscoelasticity model

Aoyagi, Takayoshi*; *; Mihara, Morihiro; Okutsu, Kazuo*; Maeda, Munehiro*

JNC TN8400 2001-024, 103 Pages, 2001/06

JNC-TN8400-2001-024.pdf:8.84MB

In the disposal concept of TRU waste, concentrated disposal of wastes forms in large cross-section underground cavities is envisaged, because most of TRU waste is no-heat producing in spite of large generated volume as compared with HLW. In the design of engineered barrier system based on large cross-section cavities, it is necessary to consider the long-term mechanical process such as creep displacement of the host rock from the viewpoint of the stability of engineered barrier system. In this study, the long-term creep displacement of the host rock was calculated using the non-linear viscoelasticity model and the effects on the stability of engineered barrier system was evaluated. As a result, in the disposal concept of crystalline rock, no creep displacement occurred at the time after 1 milion year. On the other hand, in the disposal concept of sedimentary rock, creep displacement of 8090mm occurred at the time after 1 milion year. Also, in this calculation, a maximum reduction of 45mm concerned with the thickness of buffer material was estimated. But these values resulted within allowance of design values. Therefore, these results show that the effects of the creep displacement on the stability of engieered barrier system would not be significant.