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Saito, Yuki*; Ishiwata, Tobimaru*; Horiuchi, Misato*; Nishiki, Yuto*; Kikuchi, Ryosuke*; Otake, Tsubasa*; Kawakita, Ryohei; Takayama, Yusuke; Mitsui, Seiichiro; Sato, Tsutomu*
Shigen, Sozai Koenshu (Internet), 11(1), 7 Pages, 2024/03
no abstracts in English
Aoyagi, Kazuhei; Ishii, Eiichi
Environmental Earth Sciences, 83(3), p.98_1 - 98_15, 2024/02
Times Cited Count：0 Percentile：0.04(Environmental Sciences)The long-term geological disposal of high-level radioactive waste relies on predictions of future changes in a disposal facility's hydro-mechanical characteristics to assess potential leakage through fractures in the excavation damaged zone (EDZ) after backfilling the facility. This study evaluated the transmissivity of EDZ fractures using in situ hydraulic tests around the area of a full-scale, experimental, engineered barrier system in the Horonobe Underground Research Laboratory, Hokkaido, Japan. After their installation, the buffer blocks swelled, altering the stresses within the EDZ fractures. The effects of these changing stresses on the fractures' transmissivity were assessed over a period of 4 years. The transmissivity continuously decreased in this period to about 41% of its value measured prior to the swelling. Using the Barton-Bandis normal-stress-dependent fracture-closure model, the decrease in transmissivity is quantitatively attributed to closure of the EDZ fractures, which was caused by the swelling pressure increasing up to 0.88 MPa. Evidence of fracture closure came from seismic tomography surveying, which revealed a slight increase in seismic velocity in the study area with increasing swelling pressure. The results show that EDZ fractures were closed by swelling of the full-scale buffer material. They also demonstrate the applicability of the Barton-Bandis model to preliminary estimation of the long-term transmissivity of EDZ fractures in facilities for the geological disposal of radioactive waste.
Nagai, Takayuki; Hasegawa, Takehiko*
JAEA-Research 2023-008, 41 Pages, 2023/12
JAEA-Research-2023-008.pdf:7.52MB
To reduce the risks posed by stored the high-level radioactive liquid waste (HAW), Tokai Vitrification Facility (TVF) is working to produce the HAW into vitrified bodies. With the aim of steady vitrification of HAW in TVF, the vitrification technology section has manufactured a new 3rd melter with an improved bottom structure and is working to verify the performance of this melter. In this study, solidified glass samples were taken from simulated vitrified bodies produced by flowing molten glass during the bottom drain-out test in the operation confirmation of the TVF 3rd melter. And the properties of the surface layer and fracture surface of the vitrified bodies were evaluated by using Raman spectroscopy, synchrotron radiation XAFS measurement, and laser ablation inductively coupled plasma atomic emission spectroscopy (LA ICP-AES) analysis. As a result of measuring the surface layer and fracture surface of the solidified samples produced on an actual scale, a slight difference was confirmed between the properties of the surface layer and those of the fracture surface. Since the chemical composition of these simulated vitrified bodies does not contain platinum group elements, it is expected that the glass structure of solidified glass samples is different from that of the actual vitrified body. However, this sample measuring was a valuable opportunity to evaluate samples produced by using the direct energized joule heating method. The properties of cullet used the operation confirmation of the TVF 3rd melter and the cullet of another production lot were measured and analyzed in the same manner under the measuring conditions of solidified glass samples. As a result, it was confirmed that cullet with different producing histories have different glass structures even with the same chemical composition, and that differences in glass structures remain in the glass samples after melting these cullet.
Matsueda, Makoto
Chino To Joho, 35(4), P. 88, 2023/11
Radioactive waste is what contains radioactive materials generated through nuclear activities, radiopharmacy, research and development. The treatment and disposal of the waste are one of the key challenges facing people. This glossary describes the classification of radioactive waste, the challenges and the current efforts of its disposal and current efforts, and so on.
Nakata, Hisakazu
Genkan Senta Topikkusu, (147), p.4 - 11, 2023/10
no abstracts in English
Tachi, Yukio
Kagaku To Kyoiku, 71(10), p.420 - 423, 2023/10
no abstracts in English
Nakazawa, Osamu; Takiya, Hiroaki; Murakami, Masashi; Donomae, Yasushi; Meguro, Yoshihiro
JAEA-Review 2023-012, 6 Pages, 2023/08
JAEA-Review-2023-012.pdf:0.93MB
The selection of back-end technology development issues to be prioritized and their schedule of the Japan Atomic Energy Agency (JAEA) have been put together as the "Strategic Roadmap for Back-end Technology Development." The results of questionnaires on development technologies (seeds) and technical issues (needs) within JAEA conducted in FY2022 were reflected in the selection. The issues were extracted from among those that match the seeds and needs, from the perspective of early implementation in the work front and the perspective of common issues, and nine themes were selected. We will build a cross-organizational implementation framework within JAEA and aim to implement the development results in the work front as well as social implementation.
Fukaya, Yuji; Maruyama, Takahiro; Goto, Minoru; Ohashi, Hirofumi; Higuchi, Hideaki
JAEA-Research 2023-002, 19 Pages, 2023/06
JAEA-Research-2023-002.pdf:1.48MB
A study on disposal of waste derived from commercial High Temperature Gas-cooled Reactor ("HTGR") has been performed. Because of significant difference between the reprocessing of Light Water Reactor ("LWR") and that of HTGR due to difference in structures of the fuel, adoptability of the laws relating to reprocessing waste disposal, which is enacted for LWR, to HTGR waste should be confirmed. Then, we compared the technologies and waste of reprocessing and evaluated radioactivity concentration in graphite waste by activation and contamination based on whole core burn-up calculation. As a result, it was found that SiC residue waste should be disposed of into a geological repository as 2nd class designated radioactive waste in the Designated Radioactive Waste Final Disposal Act (Act No.117 of 2000), by way of amendment of the applicable order, same as hull and end-piece of LWR, and graphite waste should be shallowly disposed of than geological disposal as 2nd class waste for pit disposal in the Act on the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors (Act No.166 of 1957) same as a channel box of LWR.
Sakamoto, Yoshiaki
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 30(1), p.15 - 18, 2023/06
The entire process of nuclear power generation is called the nuclear fuel cycle, and each process generates various types of radioactive waste. These radioactive wastes are generated from the operation and decommissioning of these facilities, and are treated and disposed of appropriately according to their radioactivity concentrations and properties. This paper describes the basic outline of the nuclear fuel cycle and the fundamentals of the treatment and disposal of radioactive waste (including radioactive waste from the use of radioactive materials in facilities other than the nuclear fuel cycle), called the back end of the nuclear fuel cycle.
Murakami, Hiroaki; Nishiyama, Nariaki; Takeuchi, Ryuji; Iwatsuki, Teruki
Oyo Chishitsu, 64(2), p.60 - 69, 2023/06
In order to confirm the quality control items for borehole closure in radioactive waste disposal projects, in-situ borehole sealing tests using bentonite material were conducted. As a result, the closure performance was successfully demonstrated by comparing the data of water injection tests conducted before and after the installation of the closure material. However, the breakthrough was observed after closing, probably due to high differential pressure applied to the seal section. Thus, it is important to ascertain throughout the entire operation that the borehole is adequately closed. The placement and specifications of the closure material should be determined according to the hydrogeological structure in the borehole. The confirmation items to use bentonite for sealing material are identified to be: to consider swelling and density loss in the borehole; to place the planned depth using appropriate emplacement technique; to be placed without damage to seals when use some backfilling materials, considering effect of permeability on adjacent seals.
Ono, Ayato; Takayanagi, Tomohiro; Sugita, Moe; Ueno, Tomoaki*; Horino, Koki*; Yamamoto, Kazami; Kinsho, Michikazu
JAEA-Technology 2022-036, 31 Pages, 2023/03
JAEA-Technology-2022-036.pdf:8.77MB
In the Japan Atomic Energy Agency (JAEA), many electrical facilities such as power receiving equipment and power supply units are installed in experimental facilities such as the Nuclear Science Research Institute (NSRI) and the Japan Proton Accelerator Research Complex (J-PARC). However, some facilities have been in operation for more than half a century since they were manufactured, some have already been closed or deactivated, and others are still in operation while replacing parts and taking other aging measures. In these facilities, materials that were used because of their excellent properties at the time of manufacture are now designated as hazardous substances and require special management when disposed of. One of them is polychlorinated biphenyl (PCB). PCB were used in a very wide range of fields because of their stability against heat, high electrical insulation, and chemical resistance. However, it was found that PCB have persistent properties and may cause damage to human health and the living environment, and the government has enacted the "Act on Special Measures for Promotion of Proper Treatment of PCB Wastes (PCB Special Measures Law)" to promote reliable and proper disposal. JAEA has almost completed the excavation survey of high-concentration PCB waste and is in the process of excavating low-concentration PCB waste. However, there are still new relevant items to be discovered. This report summarizes and reports the knowledge necessary for identifying PCB waste and points to be noted when handling capacitors, etc., based on examples of actual disassembly and investigation work conducted on power supply units and other electrical equipment, such as capacitors attached to power supply units, etc.
Nagai, Takayuki
JAEA-Research 2022-014, 84 Pages, 2023/02
JAEA-Research-2022-014.pdf:22.26MB
Most of the simulated waste glasses used for physical property evaluation are processed into a shape suitable for the measurement method from glass gob obtained by slowly cooling molten glass to room temperature. However, the actual vitrified waste glass material is obtained by cooling and being coagulated the glass drained from the bottom of glass melter into the canister. In this study, Raman spectroscopy was performed on the coagulated surface of molten simulated waste glass in the depth direction to evaluate the state of the Si-O bridging structure near the coagulated glass surface. The Raman spectra measured from the surface to the depth direction near the surface of the glasses produced by several melting and coagulation conditions of molten simulated waste glass cullet in the air atmosphere, and it was confirmed that there were changes in these spectra. On the other hand, the raw material glass cullet and the surface of the glass solidified in argon gas atmosphere showed little change in the spectrum in the depth direction, and the Si-O bridging structure near the glass surface was similar. It was also confirmed that the spectrum change in the depth direction measurement was small for the cut surface of the glass, and that the change in the spectrum for the broken glass fracture surface was also small. For glasses with a large change in Raman spectra in the depth direction near the coagulated surface, the molten glass was cooled from the molten state to room temperature in a muffle furnace with air atmosphere. That is, the magnitude of the spectral change with respect to the depth direction depends on the time from the molten state to coagulation. In order to confirm the reason why the number of bridging oxygen in the Si-O bridging structure is small on the coagulated surface of glass, the XANES spectra of Si-K edge and Ce-L3 edge were measured by XAFS on the coagulated surface and the cutting face. As a result, the Si-K edge peak on the coagulated surface is
Sakai, Akihiro; Kamei, Gento; Sakamoto, Yoshiaki
Nihon Genshiryoku Gakkai-Shi ATOMO
, 65(1), p.25 - 29, 2023/01
Currently, radioactive waste generated from research institutes, etc. is keeping in storage facilities without being disposed of. In order to solve this problem, the Japan Atomic Energy Agency (JAEA) is proceeding with the project for concrete-pit disposal and trench disposal of these waste. This paper introduces the characteristics of the waste and disposal facilities planned by the JAEA, as well as the status of development of the siting criteria for the disposal facility.
Kochiyama, Mami
Kaku Deta Nyusu (Internet), (133), p.76 - 81, 2022/10
The outline of the presentation at the joint session of Research Committee for Nuclear Data and Subcommittee on Nuclear Data in the Atomic Energy Society of Japan 2022 Autumn Meeting was contributed to Nuclear Data News. As part of the study on the near surface disposal of waste from research facilities, we are studying a method for evaluating the radioactivity inventory of waste generated by the dismantling of research reactors. In the radioactivity evaluation of the research reactor, we have investigated the method of calculating the neutron transport in the reactor and using the obtained neutron spectrum to calculate the activation of the internal structure by the ORIGEN-S code. In recent years, we have introduced and evaluated libraries created based on JENDL-4.0 and JENDL/AD-2017, and we will introduce the status of their examination. And we will introduce how to apply the results obtained by the radioactivity evaluation calculation to burial disposal.
Yokoyama, Kaoru; Ohashi, Yusuke
Annals of Nuclear Energy, 175, p.109240_1 - 109240_7, 2022/09
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Decommissioning is planned at nuclear facilities that have been discontinued. We examined the evaluation method of uranium radioactivity for concrete waste generated by the decommissioning of nuclear facilities. Since the peaks of Ac-228, Tl-208, and K- 40 are derived from concrete waste, it is difficult to distinguish the 1001 keV peak emitted from the uranium source. We have derived a formula to correct gamma rays from concrete and the environment, and the amount of uranium was quantified. When the weight of concrete waste is about 300 kg, if the weight of uranium is 3 g or more, it can be quantified within a relative error of about 30%. Measurement tests were performed using homogeneous simulated concrete waste. Since uranium contamination is on the concrete surface at the uranium processing facility and small chunks generated by scraping the concrete surface will be stored in a drum and measured, it seems that the test of homogeneous concrete reflects the actual waste.
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
Ogawa, Rina; Nakata, Hisakazu; Sugaya, Toshikatsu; Sakai, Akihiro
JAEA-Technology 2022-010, 54 Pages, 2022/07
JAEA-Technology-2022-010.pdf:11.07MB
Japan Atomic Energy Agency has considered trench disposal as one of the disposal methods for radioactive wastes generated from research facilities and other facilities. The trench disposal facility is regulated by "Act on the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors". In particular, the design of the trench facility is regulated by a rule under the law. When the rule was amended in 2019, the design of the trench disposal facility required equipment to reduce ingress of rain water and groundwater. In the report, studies on the design of a trench disposal facility to adapt to the amended rule were performed. The trench disposal facility has considered being established in a place lower than groundwater level. Therefore, it was decided to study covering soil at the upper part of the trench facility, because the ingress water in the facility is mainly derived from rain water. In this study, it was decided to evaluate the design of covering soil of the radioactive waste categorized into chemically stable materials. Therefore, as the examination method, a parameter study on varying the permeability coefficient and thickness of the layers composing cover soil. In the parameter study, the velocity of the water infiltrating into the trench facility was evaluated. Based on the results, more efficient design of the layers composing the covering soil was considered. The result showed that the impermeable efficiency of the covering soil was different depending on the thickness and the permeability conductivity of each layer. As a result, it was possible to understand the impermeable performance of covering soil by the permeability coefficient and thickness of each layer. We will plan to decide the specification of the cover soil while examination of future tasks and cost in the basic design.
Sakasegawa, Hideo
ENEKAN, 20, p.20 - 23, 2022/07
no abstracts in English
Murakami, Masashi; Sasaki, Toshiki
JAEA-Review 2022-004, 106 Pages, 2022/06
JAEA-Review-2022-004.pdf:3.95MB
A numerous analytical data will be required for the characterization of the radioactive waste stored in Japan Atomic Energy Agency toward their processing and disposal. A "Data Quality Objectives (DQO) Process" is widely applied as a tool for the development of a cost-effective characterization plan in the overseas nuclear sites. The DQO Process is a multi-step planning process developed by the United States Environmental Protection Agency (EPA), and can be used for the planning of a scientifically rigorous and cost-effective data collection program for the various projects involving the collection of the environmental data. We have considered to reduce the cost required for the future characterization of the stored waste by applying the statistical methods and have performed a literature survey on the DQO Process. The survey effort was focused on the guidance documents of the DQO Process published by the EPA and was also spent for the related matters such as a quality system of the EPA and the activities beyond the DQO Process as well as the examples of the application at the nuclear sites. In this report, the details on the planning procedure using the DQO Process are reviewed together with the background information such as why DQO Process was developed, what kind of transition was occurred, and how it is currently used in the EPA. The examples on the application for various objects at Hanford Site in the United States, where has the various legacy waste generated in the past activities and has the big environmental problems, are also reviewed. This report summarizes the important matters and methodology on the planning with the statistical sampling methods. It also provides the examples of the approaches for the complex objects, and will therefore be helpful in the future planning for the various kind of the waste characterization.
Sakai, Akihiro
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 29(1), p.48 - 54, 2022/06
no abstracts in English
