Proposition of confirmation items on the borehole sealing for the disposal of radioactive waste

Murakami, Hiroaki; Nishiyama, Nariaki; Takeuchi, Ryuji; Iwatsuki, Teruki

Oyo Chishitsu, 64(2), p.60 - 69, 2023/06

https://doi.org/10.5110/jjseg.64.60

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.

Review of performance assessment for engineered barrier systems to support future RD&D of radioactive waste management in Japan

Abe, Takeyasu; Iida, Yoshihisa

Journal of Advanced Concrete Technology, 20(3), p.236 - 253, 2022/03

https://doi.org/10.3151/jact.20.236

This paper is a state-of-the-art report on the performance assessment of cementitious and related materials as components of engineered barrier systems for radioactive waste management. In this paper, (1) the concept of safety functions is reviewed as the engineering background of discussion, (2) an overview of the postclosure performance assessment for Belgian low- and intermediate-level short-lived radioactive waste disposal is provided, and (3) a modeling methodology for engineered barrier systems is analyzed using the concept of "mandala for durability mechanics". According to these works, authors present technical suggestions for technical stakeholders of Japanese low-level radioactive waste disposal.

Importance of root uptake of CO on C transfer to plants impacted by below-ground CH release

Ota, Masakazu; Tanaka, Taku*

Journal of Environmental Radioactivity, 201, p.5 - 18, 2019/05

https://doi.org/10.1016/j.jenvrad.2019.01.012

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.

Investigation of system for volume-reduction and recycling of HLW

Nishihara, Kenji

ImPACT Fujita Puroguramu Kokai Seika Hokokukai "Kaku Henkan Niyoru Koreberu Hoshasei Haikibutsu No Ohaba Na Teigen, Shigenka" Seika Hokokusho, Shiryoshu, p.130 - 133, 2019/03

High level radioactive waste contains elements with various characteristics. It is possible to reduce the load on the disposal site by separating them according to those characteristics and appropriately dealing with them. In this project, we are working to shorten the life span of long-lived fission products (LLFP). When this technology is realized, high-level radioactive wastes will become new radioactive wastes with low radioactivity. As a result of investigation of disposal concept of new radioactive waste, it turned out that intermediate-depth disposal currently considered for low level radioactive waste may be suitable. Intermediate-depth disposal is a method of small-scale disposal in shallow locations as compared to geological disposal for conventional high-level radioactive waste. We conducted a safety assessment when this disposal is applied to new radioactive wastes, and found that it is possible to safely dispose of for the four LLFPs addressed by this project.

Study on the evaluation method to determine the radioactivity concentration in radioactive waste generated from post-irradiation examination facilities, 2

Tsuji, Tomoyuki; Hoshino, Yuzuru; Sakai, Akihiro; Sakamoto, Yoshiaki; Suzuki, Yasuo*; Machida, Hiroshi*

JAEA-Technology 2017-010, 75 Pages, 2017/06

JAEA-Technology-2017-010.pdf:2.31MB

It is necessary for reasonable disposal to be studied on evaluation methods to determine radioactivity concentrations in the radioactive wastes, which is generated from post-irradiation examination (PIE) facilities, for establishment of reasonable confirmation methods concerning radioactive wastes generated from research, industrial, and medical facilities. It has been chosen the PIE facilities of NUCLEAR DEVELOPMENT CORPORATION as a model for this study. As a result, it has been confirmed that the theoretical methods are applied for the important nuclides (H-3, C-14, Co-60, Ni-63, Sr-90, Tc-99, Cs-137, Eu-154, U-234, U-235, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Am-241 and Cm-244).

Recent activities in the field of nuclear waste management

Kitamura, Akira; Kirishima, Akira*

Journal of Nuclear Science and Technology, 52(3), p.448 - 450, 2015/03

https://doi.org/10.1080/00223131.2014.952700

The Journal of Nuclear Science and Technology covers a variety of subjects in the field of nuclear waste management, which includes radioactive waste treatment, radioactive waste disposal and environment, decommissioning and dismantling. This summary introduces activities presented in recent years.

Geochemistry research equipment for TRU waste elements

Akai, Masanobu; Ito, Nobuyuki*; Yamaguchi, Tetsuji; Tanaka, Tadao; Iida, Yoshihisa; Nakayama, Shinichi; Inagaki, Shingo*

JAERI-Tech 2004-058, 47 Pages, 2004/09

JAERI-Tech-2004-058.pdf:7.27MB

Geochemistry Research Equipment for TRU Waste Elements has been installed in Back-end Cycle Key Elements Research Facility (BECKY) of Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF). This equipment is designed to study geochemical behavior of TRU elements and other radionuclides contained in TRU waste (TRU waste elements) and to acquire data for safety assessments of radioactive wastes disposal. The equipment consists of anaerobic glove box systems, aerobic glove box systems equipped with built-in barrier performance testing apparatus, and analytical instruments. This report describes principles, structure, performance and safety designs of each component of the equipment, and results of research performed in the equipment.

Proceedings of the 6th NUCEF Seminar; February 20, 2003, JAERI, Tokai, Japan

6th NUCEF Seminar Working Group

JAERI-Conf 2003-018, 132 Pages, 2003/10

JAERI-Conf-2003-018.pdf:10.41MB

no abstracts in English

Passivation condition of carbon steel in bentonite/sand mixture

Taniguchi, Naoki; Kawakami, Susumu; *

JNC TN8400 2001-025, 27 Pages, 2002/03

JNC-TN8400-2001-025.pdf:1.16MB

It is essential to understand the corrosion type of carbon steel under the repository conditions for the lifetime assessment of carbon steel overpack used for geological isolation of high-level radioactive waste. According to the previous study, carbon steel is hard to passivate in buffer material assuming a chemical condition range of groundwater in Japan. However, concrete support will be constructed around the overpack in the case of repository in the soft rock system and groundwater having a higher pH may infiltrate to buffer material. There is a possibility that the corrosion type of carbon steel will be influenced by the rise of the pH in groundwater. In this study, anodic polarization experiments were performed to understand the passivation condition of carbon steel in buffer material saturated with water contacted with concrete. An ordinary concrete and a low-alkalinity concrete were used in the experiment. The results of the experiments showed that the carbon steel can passivate under the condition that water having pH 13 infiltrate to the buffer material assuming present property of buffer material. If the low-alkalinity concrete is selected as the support material, passivation can not occur on carbon steel overpack. The effect of the factors of buffer material such as dry density and mixing ratio of sand on the passivation of carbon steel was also studied. The results of the study showed that the present property of buffer material is enough to prevent passivation of carbon steel.

Dataset of the relationship between unconfined compressive strength and tensile strength of rock mass

Sugita, Yutaka; Yui, Mikazu

JNC TN8450 2001-007, 16 Pages, 2002/02

JNC-TN8450-2001-007.pdf:0.78MB

This report summary the dataset of the relationship between unconfined compressive strength and tensile strength of the rock mass described in supporting report 2; repository design and engineering technology of second progress report (H12 report) on research and development for the geological disposal of HLW in Japan.