Nakagawa, Akinori; Oyokawa, Atsushi; Murakami, Masashi; Yoshida, Yukihiko; Sasaki, Toshiki; Okada, Shota; Nakata, Hisakazu; Sugaya, Toshikatsu; Sakai, Akihiro; Sakamoto, Yoshiaki
JAEA-Technology 2021-006, 186 Pages, 2021/06
Radioactive wastes generated from R&D activities have been stored in Japan Atomic Energy Agency. In order to reduce the risk of taking long time to process legacy wastes, countermeasures for acceleration of waste processing and disposal were studied. Work analysis of waste processing showed bottleneck processes, such as evaluation of radioactivity concentration, segregation of hazardous and combustibles materials. Concerning evaluation of radioactivity concentration, a radiological characterization method using a scaling factor and a nondestructive gamma-ray measurement should be developed. The number of radionuclides that are to be selected for the safety assessment of the trench type disposal facility can decrease using artificial barriers. Hazardous materials, will be identified using records and nondestructive inspection. The waste identified as hazardous will be unpacked and segregated. Preliminary calculations of waste acceptance criteria of hazardous material concentrations were conducted based on environmental standards in groundwater. The total volume of the combustibles will be evaluated using nondestructive inspection. The waste that does not comply with the waste acceptance criteria should be mixed with low combustible material waste such as dismantling concrete waste in order to satisfy the waste acceptance criteria on a disposal facility average. It was estimated that segregation throughput of compressed waste should be increased about 5 times more than conventional method by applying the countermeasures. Further study and technology development will be conducted to realize the plan.
Okamura, Tomohiro*; Oizumi, Akito; Nishihara, Kenji; Nakase, Masahiko*; Takeshita, Kenji*
JAEA-Data/Code 2020-023, 32 Pages, 2021/03
Nuclear Material Balance code (NMB code) have been developed in Japan Atomic Energy Agency. The NMB code will be updated with the function of mass balance analysis at the backend process such as reprocessing, vitrification and geological disposal. In order to perform its analysis with high accuracy, it is necessary to expand the number of FP nuclides calculated in the NMB code. In this study, depletion calculation by ORIGEN code was performed under 3 different burn-up conditions such as spent uranium fuel from light water reactor, and nuclides were selected from 5 evaluation indexes such as mass and heat generation. In addition, the FP nuclides required to configure a simple burnup chain with the same calculation accuracy as ORIGEN in the NMB code was selected. As the result, two lists with different number of nuclides, such as "Detailed list" and a "Simplified list", were created.
Takahashi, Tomoyuki*; Uchida, Shigeo*; Takeda, Seiji; Nakai, Kunihiro*
KURNS-EKR-11, p.97 - 102, 2021/03
This paper outlines the status of IAEA database compilation for migration parameters depending elements in a biosphere such as soil-to-plant transfer factor and bioconcentration factor of marine products, and the status of utilization of the database in dose evaluation of radioactive waste disposal in Japan. Additionally, in the case of applying a new database to the dose evaluation for future radioactive waste disposal in a specific area. We summarized the opinions of specialists and result of general discussion about future strategies to make a new database for their parameters, perspectives to be considered in it, issues, etc.
Yamamoto, Masahiko; Do, V. K.; Taguchi, Shigeo; Kuno, Takehiko; Takamura, Yuzuru*
Journal of Radioanalytical and Nuclear Chemistry, 327(1), p.433 - 444, 2021/01
A simple, practical, and reliable analytical method for determination of Na, K, Ca, Sr, and Ba by liquid electrode plasma optical emission spectrometry is developed. Appropriate emission lines for quantification, interferences from co-existing elements, and effect of measurement conditions with cell damage have been investigated. The spike and recovery tests using actual sample have been performed for method validation, and negligible sample matrix effect has been observed. Consequently, the method is successfully applied to several radioactive wastes. The obtained data have been agreed well with data from computer calculation and inductively coupled plasma optical emission spectrometry within 10% difference.
Ogawa, Yusuke*; Suzuki, Satoru*; Taniguchi, Naoki; Kawasaki, Manabu*; Suzuki, Hiroyuki*; Takahashi, Rieko*
Materials and Corrosion, 72(1-2), p.52 - 66, 2021/01
Cast steel is one of the promising alternative to forged steel that is the current reference material for carbon steel overpack. In this study, the full-scale cast steel overpack was produced experimentally and the distribution of casting defects were investigated. The corrosion test regarding corrosion rate and stress corrosion cracking (SCC) susceptibility were also conducted using samples taken from the full-scale cast steel overpack and the corrosion resistance of cast steel was compared with that of forged steel. From above two corrosion tests, it can be said that the corrosion resistance of cast steel is mostly the same as that of forged steel.
"Yugai Haikibutsu, Hoshasei Haikibutsu Eno Semento, Konkurito Gijutsu No Tekiyo Kenkyu Iinkai" Hokokusho (CD-ROM), p.165 - 173, 2020/12
no abstracts in English
Suzuki, Masafumi*; Yoshinaka, Kazuyuki
Gijutsushi, (648), p.12 - 15, 2020/12
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.
Seki, Misaki; Ishikawa, Koji*; Sano, Tadafumi*; Nagata, Hiroshi; Otsuka, Kaoru; Omori, Takazumi; Hanakawa, Hiroki; Ide, Hiroshi; Tsuchiya, Kunihiko; Fujihara, Yasuyuki*; et al.
KURNS Progress Report 2019, P. 279, 2020/08
no abstracts in English
Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Tachi, Yoshiaki; Kaito, Takeji; Hashidate, Ryuta; Kato, Shoichi; Furukawa, Tomohiro; Ito, Chikara; et al.
2018 GIF Symposium Proceedings (Internet), p.305 - 314, 2020/05
Oxide Dispersion Strengthened (ODS) steel has been developed worldwide as a high-strength and radiation-tolerant steel used for advanced nuclear system. Japan Atomic Energy Agency (JAEA) has been developing ODS steel as the primary candidate material of Sodium cooled Fast Reactor (SFR) high burn-up fuel cladding tube. Application of high burn-up fuel to SFR core can contribute to improvement of economical performance of SFR in conjunction with volume and hazardousness reduction of radioactive waste. This paper described the current status and future prospects of ODS tempered martensitic steel development in JAEA for SFR fuel application.
Tobita, Minoru*; Haraga, Tomoko; Sasaki, Takayuki*; Seki, Kotaro*; Omori, Hiroyuki*; Kochiyama, Mami; Shimomura, Yusuke; Ishimori, Kenichiro; Kameo, Yutaka
JAEA-Data/Code 2019-016, 72 Pages, 2020/02
In the future, radioactive wastes which generated from research and testing reactors in Japan Atomic Energy Agency are planning to be buried for the near surface disposal. Therefore, it is required to establish the method to evaluate the radioactivity concentrations of radioactive wastes by the time it starts disposal. In order to contribute to this work, we collected and analyzed the samples generated from JRR-2, JRR-3 and Hot laboratory facilities. In this report, we summarized the radioactivity concentrations of 25 radionuclides (H, C, Cl, Co, Ni, Sr, Nb, Mo, Tc, Ag, Sn, I, Cs, Eu, Eu, U, U, U, Pu, Pu, Pu, Pu, Am, Am, Cm) which were obtained from radiochemical analysis of those samples.
Geological Disposal Research and Development Department
JAEA-Evaluation 2019-010, 69 Pages, 2020/01
Japan Atomic Energy Agency (JAEA) consulted the advisory committee, "Evaluation Committee on Research and Development (R&D) Activities for Geological Disposal of High-Level Radioactive Waste", for an interim review of R&D activities on high-level radioactive waste disposal in accordance with "General Guideline for the Evaluation of Government Research and Development (R&D) Activities" by the Cabinet Office, Government of Japan, "Guideline for Evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology" and JAEA's "Regulation on Conduct for Evaluation of R&D Activities". In response to JAEA's request, the Committee reviewed mainly the progress of the R&D project on geological disposal, the relevance of the project outcome during the period of FY2015-2018. This report summarizes the results of the assessment by the Committee with the Committee report attached.
Seki, Misaki; Ishikawa, Koji*; Nagata, Hiroshi; Otsuka, Kaoru; Omori, Takazumi; Hanakawa, Hiroki; Ide, Hiroshi; Tsuchiya, Kunihiko; Sano, Tadafumi*; Fujihara, Yasuyuki*; et al.
KURNS Progress Report 2018, P. 257, 2019/08
no abstracts in English
Murakami, Masashi; Hoshino, Yuzuru; Nakatani, Takayoshi; Sugaya, Toshikatsu; Fukumura, Nobuo*; Sanda, Toshio*; Sakai, Akihiro
JAEA-Technology 2019-003, 50 Pages, 2019/06
Toward the establishment of a common approach to determine the radioactivity concentrations in dismantling wastes arising from research reactors, radionuclide concentrations in the reactor structure materials of aluminum, carbon steel, shield concrete, and graphite of TRIGA Mark II reactor at Rikkyo University, Japan, were evaluated with both radiochemical analysis and theoretical calculation. The measured nuclides by the radiochemical analysis were H, Co, and Ni in aluminum, H, Co, Ni, and Eu in carbon steel, H, Co, and Eu in shield concrete, and H, C, Co, Ni, and Eu in graphite. Neutron-flux distributions and neutron-induced activities were computed with DORT and ORIGEN-ARP codes, respectively. Using the results of material composition analysis, radioactivity concentrations were conservatively predicted with good accuracy except for graphite material.
Osawa, Hideaki; Nogami, Toshinobu; Hoshino, Masato; Tokunaga, Hiroaki*; Horikoshi, Hidehiko*
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 26(1), p.45 - 55, 2019/06
Japan Atomic Energy Agency has performed risk communication at Horonobe Underground Research Center, using the Public Information House and Underground Research Laboratory (URL), to promote understanding R&D of geological disposal technology and waste disposal against public. In this paper, we conducted the analysis of questionnaire investigation performing after visiting those facilities from FY2013 to FY2017. The results show that long-term safety would gain prominent attention as agita factor by growing understanding. The results also shows that visiting to those facility would become valuable experience to understand geological disposal because, for example, respondents with visiting to URL positively evaluated necessity, appropriates and safety of geological disposal, compared with those without visiting URL.
Osawa, Hideaki; Otomo, Shoji*; Hirose, Yukio*; Onuma, Susumu*
Ningen Kankyogaku Kenkyu, 17(1), p.59 - 64, 2019/06
This study examined the determinants of public acceptance of siting a repository for High-level radioactive waste (HLW), focusing on procedural fairness and trust. To examine the presumption, the study implemented a hypothetical scenario experiment that manipulated two factors: an opportunity of voice as an antecedent of procedural fairness and similarity value to the authority as a component of trust. Results indicated that affective reaction, procedural fairness, and trust determined public acceptance. A process analysis indicated that the effect of procedural fairness was strengthened when the trust on similarity was negative.
Ota, Masakazu; Tanaka, Taku*
Journal of Environmental Radioactivity, 201, p.5 - 18, 2019/05
CH released from deep underground radioactive waste disposal facilities can be a belowground source of CO owing to microbial oxidation of CH to CO in soils. Environmental C models assume that the transfer of CO from soil to plant occurs via foliar uptake of CO. Nevertheless, the importance of CO root uptake is not well understood. In the present study, belowground transport and oxidation of CH were modeled and incorporated into an existing land surface CO model (SOLVEG-II) to assess the importance of root uptake on CO transfer to plants. Performance of the model in calculating the belowground dynamics of CH was validated by simulating a field experiment of CH injection into subsoil. The model was then applied to C transfer in a hypothetical ecosystem impacted by continuous CH input from the water table (bottom of one-meter thick soil). In a shallowly rooted ecosystem with rooting depth of 11 cm, foliar uptake of CO was significant, accounting for 80% of the C accumulation in the leaves. In a deeply rooted ecosystem (rooting depth of 97 cm), where the root penetrated to depths close to the water-table, more than half (63%) the C accumulated in the leaves was transferred by the root uptake. We found that CO root uptake in this ecosystem depended on the distribution of methane oxidation in the soil; all C accumulated in the leaves was transferred by the root uptake when methane oxidation occurred at considerable depths (e-folding depths of 20 cm, or 80 cm). These results indicate that CO root uptake contributes significantly to CO transfer to plants if CH oxidation occurs at great depths and roots penetrate deeply into the soil.
Sato, Kazuhiko; Yagi, Naoto; Nakagiri, Toshio
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 6 Pages, 2019/05
no abstracts in English
Izumo, Sari; Hayashi, Hirokazu; Nakata, Hisakazu; Amazawa, Hiroya; Motoyama, Mitsushi*; Sakai, Akihiro
JAEA-Technology 2018-018, 39 Pages, 2019/03
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.
Kato, Tomoko; Fukaya, Yukiko*; Sugiyama, Takeshi*; Nakai, Kunihiro*; Oda, Chie; Oi, Takao
JAEA-Data/Code 2019-002, 162 Pages, 2019/03
The radioactive waste generated from Fukushima Daiichi nuclear power station (FDNPS) accident have features such as wide range of radioactivity level (from low to high) and huge amount etc. It would be necessary for the waste from the FDNPS accident to develop suitable disposal concept and to be disposed safely and reasonably. When considering such appropriate disposal concepts in site-generic phase, it is necessary to appropriately develop models and parameters depending on the disposal concepts, such as disposal depth and specification of engineered barrier. In addition, it is desirable to evaluate the safety of repository with common models and parameters independent on the disposal concepts. In the safety assessment of disposal, it is useful to show the difference in performance of repository with "dose" as an indicator of safety assessment. Biosphere model and parameter set and flux-to-dose conversion factors calculated using them are originally dependent on the disposal concepts. However, the biosphere models and the parameter set in safety assessment of near-surface disposal, sub-surface disposal and geological disposal are prepared in each case, and are different according to the age and purpose of the discussion. In this study, an example of biosphere model and parameter-set of groundwater sceinario commonly applicable to various disposal concepts were shown, to calculate flux-to-dose conversion factors, as common indicators independent to disposal concept. And, a set of flux-to-dose conversion factors was also calculated by using the commonly available biosphere model and parameter set. By applying the flux-to-dose conversion factors, it is possible to compare the performance of disposal concepts to the waste generated from FDNPS accident, focusing on the parts depending on the disposal concepts.
ImPACT Fujita Puroguramu Kokai Seika Hokokukai "Kaku Henkan Niyoru Koreberu Hoshasei Haikibutsu No Ohaba Na Teigen, Shigenka" Seika Hokokusho, Shiryoshu, p.28 - 31, 2019/03
In this project, long-lived fission products (LLFP) contained in conventional high-level radioactive wastes are separated and their life is reduced, and elements that can be used as resources are separated. By shortening the life of LLFP, it has been shown that it may be possible to dispose in intermediate depth of several tens of meters, meeting safety requirements, instead of geological disposal. In addition, for reassuring recycling of usable elements, possible exposure pathways were evaluated to estimate the safe concentration level of radioactivity.