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Journal Articles

Long-term changes in the chemical, microstructural, and transport properties of a low-pH cement shotcrete during operation of the Horonobe Underground Research Laboratory, Japan

Mochizuki, Akihito; Matsui, Hiroya; Nakayama, Masashi; Sakamoto, Ryo*; Shibata, Masahito*; Motoshima, Takayuki*; Jo, Mayumi*

Case Studies in Construction Materials, 22, p.e04648_1 - e04648_20, 2025/07

 Times Cited Count:0 Percentile:0.00(Construction & Building Technology)

The properties of low-pH cement used in the geological disposal of radioactive waste may change through atmospheric carbonation and degradation caused by groundwater during the long-term operation of a repository. In this study, we investigated the effects of atmospheric carbonation and groundwater contact on the chemical, microstructural, and transport properties of shotcrete made from low-pH, high-fly-ash silica-fume cement (HFSC) over a period of 16 years in an underground research laboratory. In both carbonated and degraded zones of the HFSC shotcrete, capillary porosity increased for pores of $$<$$300 nm in diameter, and the total porosity was higher than in undegraded zones. These changes in porosity may be associated with the decalcification of calcium-silicate-hydrate and decomposition of ettringite. Such changes were minor in altered zones of OPC shotcrete, indicating that HFSC shotcrete is less resistant to atmospheric carbonation and groundwater leaching under the studied conditions. However, the hydraulic conductivity in HFSC was low enough to fulfill the specific functional requirements of low-pH cements for geological disposal.

JAEA Reports

Verification of analytical model of MELCOR code for accident of evaporation to dryness by boiling of reprocessed high level liquid waste

Yoshida, Kazuo; Hiyama, Mina*; Tamaki, Hitoshi

JAEA-Research 2025-003, 24 Pages, 2025/06

JAEA-Research-2025-003.pdf:2.06MB

An accident of evaporation to dryness by boiling of high-level radioactive liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (RuO$$_{4}$$) are released from the tanks with water and nitric-acid mixed vapor into the atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. RuO$$_{4}$$ is expected to be absorbed chemically into water dissolving nitrous acid. Condensation of mixed vapor plays an important role for Ru transporting behavior in the facility building. The thermal-hydraulic behavior in the facility building is simulated with MELCOR code. The latent heat, which is a governing factor for vapor condensing behavior, has almost same value for nitric acid and water at the temperature range under 120 centigrade. Considering this thermal characteristic, it is assumed that the amount of nitric acid is substituted with mole-equivalent water in MELCOR simulation. Compensating modeling induced deviation by this assumption have been assembled with control function features of MELCOR. The comparison results have been described conducted between original simulation and modified simulation with compensating model in this report. It has been revealed that the total amount of pool water in the facility was as same as both simulations.

Journal Articles

Numerical simulation of coupled THM behaviour of full-scale EBS in backfilled experimental gallery in the Horonobe URL

Sugita, Yutaka; Ono, Hirokazu; Beese, S.*; Pan, P.*; Kim, M.*; Lee, C.*; Jove-Colon, C.*; Lopez, C. M.*; Liang, S.-Y.*

Geomechanics for Energy and the Environment, 42, p.100668_1 - 100668_21, 2025/06

 Times Cited Count:1 Percentile:0.00(Energy & Fuels)

The international cooperative project DECOVALEX 2023 focused on the Horonobe EBS experiment in the Task D, which was undertaken to study, using numerical analyses, the thermo-hydro-mechanical (or thermo-hydro) interactions in bentonite based engineered barriers. One full-scale in-situ experiment and four laboratory experiments, largely complementary, were selected for modelling. The Horonobe EBS experiment is a temperature-controlled non-isothermal experiment combined with artificial groundwater injection. The Horonobe EBS experiment consists of the heating and cooling phases. Six research teams performed the THM or TH (depended on research team approach) numerical analyses using a variety of computer codes, formulations and constitutive laws.

Journal Articles

DECOVALEX-2023: An International collaboration for advancing the understanding and modeling of coupled thermo-hydro-mechanical-chemical (THMC) processes in geological systems

Birkholzer, J. T.*; Graupner, B. J.*; Harrington, J.*; Jayne, R.*; Kolditz, O.*; Kuhlman, K. L.*; LaForce, T.*; Leone, R. C.*; Mariner, P. E.*; McDermott, C.*; et al.

Geomechanics for Energy and the Environment, 42, p.100685_1 - 100685_17, 2025/06

 Times Cited Count:0

JAEA Reports

Seismic reinforcement works of Waste Treatment Facility No.2

Kinoshita, Junichi; Sakamoto, Yu; Suzuki, Ichiro; Nakajima, Ryota; Morita, Yusuke; Irie, Hirobumi

JAEA-Technology 2024-027, 55 Pages, 2025/05

JAEA-Technology-2024-027.pdf:8.49MB

The Waste Treatment Facility No.2 has equipment that can process solid waste with relatively high radioactive levels generated within the Japan Atomic Energy Agency. This facility had been constructed under the old Building Standards Act. Seismic evaluation based on a new regulatory requirements enforced in December 2013 was executed, thereby, it was found that the seismic resistance requirements was insufficient according to the current Building Standards Act. Therefore, seismic reinforcement works was carried out from November 2018 to February 2020. In this report, seismic reinforcement design, works, test and inspection was complied.

Journal Articles

A Review on the effect of iron on bentonite stability

Wilson, J.*; Sasamoto, Hiroshi; Tachi, Yukio; Kawama, Daisuke*

Applied Clay Science, 275, p.107862_1 - 107862_15, 2025/05

 Times Cited Count:0

High-Level Radioactive Waste (HLW) repositories include iron or steel-based containers/overpack and bentonite buffers. Over the last 25 years or so, research efforts have attempted to elucidate the nature of iron-bentonite interactions, especially the potential for the deleterious alteration of the swelling clay component (smectite), to iron-rich layer silicates, some of which lack the capacity for intracrystalline swelling. This could result in a reduction or loss in swelling pressure in the bentonite buffer which is designed to protect waste containers from shear forces and also acts to restrict water and solute transport, as part of an engineered barrier system. Most data on iron-bentonite interactions come from experimental and geochemical modelling studies, as natural analogue data are lacking. The data suggests that there is the potential for the development of an iron-rich bentonite alteration zone with smectite (generally present as the aluminous montmorillonite type) undergoing alteration to iron-rich solids, including layer silicates and steel corrosion products such as green rust or magnetite. The evidence available is complex, arguably incomplete, with many potential complex couplings. Many uncertainties remain despite efforts taken over the last 25 years, but plausible scenarios for iron-bentonite interactions have been identified and possible implications for buffer properties have been suggested.

Journal Articles

New filter concept for removal of fine particle generated in high level radioactive solution

Takahatake, Yoko; Watanabe, So; Watanabe, Masayuki; Sano, Yuichi; Takeuchi, Masayuki

Progress in Nuclear Science and Technology (Internet), 7, p.195 - 198, 2025/05

Extraction chromatgraphy technology for trivalent minor actinide (MA(III) ; Am(III) and Cm(III)) recovery from the solution generated by an extraction process in reprocessing of spent nuclear fuel has been developed. A fine particle is generated in the solution. The fine particle must be removed before MA recovery operation, because that leads clogging of the extraction chlomatography column. In order to prevent clogging the column, filtration system utilizing porous silica beads packed column has been designed. In this study, a fine particle trapping system was developed and particle removal performance of the system was experimentally evaluated using alumina particles as simulated fine particle. Column experiments revealed that the fine particle with the particle size from 0.12 to 15 $$mu$$m is cause of clogging of the filtration column. Since simulated fine particles were trapped on filtration experiments, a filtration system using the porous silica beads column is practical,

Journal Articles

Achievements and status of the STRAD project for radioactive liquid waste management

Arai, Yoichi; Watanabe, So; Nakahara, Masaumi; Funakoshi, Tomomasa; Hoshino, Takanori; Takahatake, Yoko; Sakamoto, Atsushi; Aihara, Haruka; Hasegawa, Kenta; Yoshida, Toshiki; et al.

Progress in Nuclear Science and Technology (Internet), 7, p.168 - 174, 2025/05

The Japan Atomic Energy Agency (JAEA) has been conducting a project named "Systematic Treatment of RAdioactive liquid waste for Decommissioning (STRAD)" project since 2018 for fundamental and practical studies for treating radioactive liquid wastes with complicated compositions. Fundamental studies have been conducted using genuine liquid wastes accumulated in a hot laboratory of the JAEA called the Chemical Processing Facility (CPF), and treatment procedures for all liquid wastes in CPF were successfully designed on the results obtained. As the next phase of the project, new fundamental and practical studies on primarily organic liquid wastes accumulated in different facilities of JAEA are in progress. This paper reviews the representative achievements of the STRAD project and introduces an overview of ongoing studies.

JAEA Reports

Study on the evaluation method of radioactivity for dismantling wastes generated from test and research reactors using ORIGEN attached to SCALE6.2.4

Tomioka, Dai; Kochiyama, Mami; Ozone, Kenji; Nakata, Hisakazu; Sakai, Akihiro

JAEA-Technology 2024-023, 38 Pages, 2025/03

JAEA-Technology-2024-023.pdf:1.54MB

Japan Atomic Energy Agency is an implementing organization of near-surface disposal for low-level radioactive wastes generated from research, industrial and medical facilities in Japan. Information on the radioactivity concentration of these radioactive wastes is dispensable for the design and conformity assessment of the waste disposal facilities for the licensing application of the disposal project and its safety review. Radioactive Wastes Disposal Center has been improving the radioactivity evaluation procedure for the dismantling waste generated from the research reactors based on the activation calculation. In order to investigate the applicability of the ORIGEN code (included in SCALE6.2.4), which enables more accurate activation calculations using multigroup neutron spectra, we performed activation calculations with the ORIGEN-code and the ORIGEN-S code (included in SCALE6.0), which has been widely used in the past, for the dismantled wastes from the Rikkyo University Research Reactor, where radioactivity analysis data for the structural materials around the reactor core were compiled. As a result, the calculation time difference between ORIGEN and ORIGEN-S was small and the evaluated radioactivity concentrations of the former were in the range of 0.8-1.0 times those of the latter, which was in good agreement with those of radiochemical analysis in the range of 0.5-3.0 times. The applicability of ORIGEN was confirmed. In addition, activation calculations assuming trace elements in structural materials of nuclear reactor were performed with ORIGEN and ORIGEN-S and the results were compared. The causes of the large differences among 170 nuclides that are important for dose assessment in near-surface disposal were assessed each nuclide.

JAEA Reports

Update of fire receiver panel (The Waste Safety Testing Facility)

Hatakeyama, Yuichi; Hirai, Koki; Ikegami, Yuta*; Sano, Naruto; Tomita, Takeshi; Usami, Koji; Tagami, Susumu

JAEA-Technology 2024-020, 33 Pages, 2025/03

JAEA-Technology-2024-020.pdf:2.21MB

The Waste Safety Testing Facility (WASTEF) is a facility that began operation in December 1982 with the aim of conducting safety testing research on the long-term storage and subsequent geological disposal of high-level radioactive waste generated by the reprocessing of spent fuel. This facility is composed of five concrete cells, one lead cell, six glove boxes, and seven hoods, and is a large-scale facility capable of using nuclear fuel materials including uranium and plutonium, as well as radioisotopes such as neptunium and americium. The facility is equipped with an automatic fire alarm system for the entire building in accordance with the Fire Service Act and regulations on technical standards for facilities used. This is an important aspect of safety management, and it is required that the equipment be sufficiently sound and reliable. However, after more than 30 years of use since its installation, the fire receiving panel, one of the components of the automatic fire alarm system, has deteriorated significantly. Furthermore, many of the parts used have been discontinued and are no longer available, making it difficult to procure them, making it difficult to maintain the equipment's performance. Therefore, in order to ensure the safe and stable operation of WASTEF, the fire receiving panel was updated. This report summarizes the update of the fire receiving panel among the automatic fire alarm equipment that was implemented in FY2022.

JAEA Reports

Assessment report of research and development activities in FY2023; Activity "Implementation and technological development of decommissioning of nuclear facilities" and "Implementation and technological development of radioactive waste processing" (In-advance evaluation)

Nuclear Backend Technology Development

JAEA-Evaluation 2024-002, 86 Pages, 2025/03

JAEA-Evaluation-2024-002.pdf:8.29MB

Japan Atomic Energy Agency (hereinafter referred to as "JAEA") consulted the "Evaluation Committee on Research and Development Activities for Decommissioning" (hereinafter referred to as "Committee") about the in-advance evaluation of research themes which had started after FY2021 of R&D plan for the 4th medium- and long-term objectives period concerning "Decommissioning of nuclear facilities and development of technologies for the treatment of related radioactive waste" project in accordance with the "Guideline for evaluation of government R&D activities". In response, the Committee evaluated to the evaluation points of view made by JAEA.

Journal Articles

Reduction and phase transformation of Ce-doped zirconolites

Hayashizaki, Kohei; Hirooka, Shun; Yamada, Tadahisa*; Sunaoshi, Takeo*; Murakami, Tatsutoshi; Saito, Kosuke

Ceramics (Internet), 8(1), p.24_1 - 24_12, 2025/03

Journal Articles

Feasibility study of reactor radiation photon spectroscopy in Fugen for nuclear decommissioning

Kaburagi, Masaaki; Miyamoto, Yuta; Mori, Norimasa; Iwai, Hiroki; Tezuka, Masashi; Kurosawa, Shunsuke*; Tagawa, Akihiro; Takasaki, Koji

Journal of Nuclear Science and Technology, 62(3), p.308 - 316, 2025/03

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Journal Articles

Evolution of radionuclide transport and retardation processes in uplifting granitic rocks, Part 2; Modelling coupled processes in uplift scenarios

Metcalfe, R.*; Benbow, S. J.*; Kawama, Daisuke*; Tachi, Yukio

Science of the Total Environment, 958, p.177690_1 - 177690_17, 2025/01

Uplifting fractured granitic rocks occur in substantial areas of countries such as Japan. A repository site would be selected in such an area only if it is possible to make a safety case, accounting for the changing conditions during uplift. The safety case must include robust arguments that chemical processes in the rocks around the repository will contribute sufficiently to minimise radiological doses to biosphere receptors. To provide confidence in the safety arguments, numerical models need to be sufficiently realistic, but also parameterised conservatively (pessimistically). However, model development is challenging because uplift involves many complex couplings between groundwater flow, chemical reactions between water and rock, and changing rock properties. The couplings would affect radionuclide mobilisation and retardation, by influencing diffusive radionuclide fluxes between groundwater flowing in fractures and effectively immobile porewater in the rock matrix and radionuclide partitioning between water and solid phases, via: (i) mineral precipitation/dissolution; (ii) mineral alteration; and (iii) sorption/desorption. It is difficult to represent all this complexity in numerical models while showing that they are parameterised conservatively. Here we present a modelling approach, illustrated by simulation cases for some exemplar radioelements, to identify realistically conservative process conceptualisations and model parameterisations.

Journal Articles

Development of a theoretical scaling factor method for the inventory estimation of difficult-to-measure nuclide Cs-135 in fuel debris and radioactive wastes

Sakamoto, Masahiro; Okumura, Keisuke; Kanno, Ikuo; Matsumura, Taichi; Terashima, Kenichi; Riyana, E. S.; Kaneko, Junichi*; Mizokami, Masato*; Mizokami, Shinya*

Journal of Nuclear Science and Technology, 10 Pages, 2025/00

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Journal Articles

Evolution of radionuclide transport and retardation processes in uplifting granitic rocks, Part 1; Key processes, conceptual models and scenario

Metcalfe, R.*; Tachi, Yukio; Sasao, Eiji; Kawama, Daisuke*

Science of the Total Environment, 957, p.177375_1 - 177375_17, 2024/12

A safety case for an underground radioactive waste repository must show that groundwater will not in future transport radionuclides from the repository to the near-surface environment (the biosphere) in harmful quantities. Safety cases are developed step-wise throughout a programme to site and develop a repository. At early stages, before a site is selected, safety cases are generic and based on simplified safety assessment models of the disposal system that have conservative parameter values. Later, when site-specific conditions are known, more realistic models are needed for the long-term geo-environmental evolution and their impacts on radionuclide migration/retention. Uplift is one such environmental change, which may be particularly important in countries near active tectonic plate boundaries, such as Japan. Here we review the state of knowledge about how the properties of fractured granitic rocks evolve during uplift, based on studies in Japan. Hence, we present conceptual models and a generic scenario for mass transport and retardation processes in uplifting granitic rocks as a basis for realistic numerical models to underpin safety assessment.

JAEA Reports

Horonobe Underground Research Laboratory Project Investigation Report for the 2023 Fiscal Year

Nakayama, Masashi

JAEA-Review 2024-042, 111 Pages, 2024/11

JAEA-Review-2024-042.pdf:7.83MB

The Horonobe Underground Research Laboratory (URL) Project is being pursued by the Japan Atomic Energy Agency (JAEA) to enhance the reliability of relevant technologies for geological disposal of high-level radioactive waste through investigating the deep geological environment within the host sedimentary rocks at Horonobe Town in Hokkaido, north Japan. In the fiscal year 2023, we continued R&D on "Study on near-field system performance in geological environment", "Demonstration of repository design options", and "Understanding of buffering behaviour of sedimentary rock to natural perturbations". These are identified as key R&D on challenges to be tackled in the Horonobe underground research plan for the fiscal year 2020 onwards. Specifically, "full-scale engineered barrier system (EBS) performance experiment" and "solute transport experiment with model testing" were carried out as part of "Study on nearfield system performance in geological environment". "Demonstration of engineering feasibility of repository technology" and "evaluation of EBS behaviour over 100$$^{circ}$$C" were addressed for "Demonstration of repository design options". The validation of a method for assessing permeability using the Ductility Index and a method for estimating the state of in-situ ground pressure from hydraulic perturbation tests were investigated as part of the study "Understanding of buffering behaviour of sedimentary rock to natural perturbations". In FY2023, we resumed construction of the subsurface facilities, 3 new tunnels in the 350 m gallery and resumed excavation of the East Access Shaft and the Ventilation Shaft. By the end of FY2023, the 350 m gallery extension (tunnel extension 66 m) had been completed, and the depths of the East Access Shaft and Ventilation Shaft were GL-424 m and GL-393 m respectively.

JAEA Reports

Study on water stopping, repair and stabilization of lower PCV by geopolymer, etc. (Contract research); FY2022 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; The University of Tokyo*

JAEA-Review 2024-021, 126 Pages, 2024/11

JAEA-Review-2024-021.pdf:6.51MB

The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2022. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2021, this report summarizes the research results of the "Study on water stopping, repair and stabilization of lower PCV by geopolymer, etc" conducted in FY2022. The present study aims to propose a construction method to stop jet deflectors by improved geopolymer and ultra-heavy muddy water, and to repair the lower part of the dry well. In addition, in order to increase the options for on-site construction in unknown situations such as deposition conditions, we will examine a wide range of construction outside the pedestal, and evaluate the feasibility of the construction method by the latest thermal flow simulation method.

Journal Articles

Initial verification of Cyclus and NMB fuel cycle simulators

Bachmann, A. M.*; Richards, S.*; Feng, B.*; Nishihara, Kenji; Abe, Takumi

Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10

This work demonstrates the value of code verification as an initial step in utilizing fuel cycle simulation. Cyclus and NMB are open-source fuel cycle simulators that provide computational modeling of nuclear fuel cycle alternatives and were chosen by Argonne National Laboratory and Japan Atomic Energy Agency (JAEA), respectively, for a multi-year collaboration on fuel cycle benchmarks. Both are relatively new and can be improved after conducting a rigorous code-to-code comparison. Initial verification of these simulators was performed using a set of hypothetical scenarios for once-through and multi-recycle fuel cycles. The results of this work identify how differences in scenario definitions and the modeling methodologies of the two simulators lead to differences in results in material inventories, mass flows, and other important metrics for fuel cycle assessments.

JAEA Reports

Horonobe Underground Research Laboratory Project; Investigation program for the fiscal year 2024

Nakayama, Masashi

JAEA-Review 2024-033, 64 Pages, 2024/09

JAEA-Review-2024-033.pdf:5.15MB

The Horonobe Underground Research Laboratory Project is being pursued by the Japan Atomic Energy Agency to enhance the reliability of relevant technologies for geological disposal of high-level radioactive waste through investigating the deep geological environment within the host sedimentary rocks at Horonobe Town in Hokkaido, north Japan. In the fiscal year 2024, we continue R&D on "Study on near-field system performance in geological environment", "Demonstration of repository design options", and "Understanding of buffering behaviour of sedimentary rocks to natural perturbations". These are identified as key R&D challenges to be tackled in the Horonobe underground research plan for the fiscal year 2020 onwards. In the "Study on near-field system performance in geological environment", we continue to obtain data from the full-scale engineered barrier system performance experiment, and work on the specifics of the full-scale engineered barrier system dismantling experiment. We summarise the solute transport experiments for the excavation damaged zone and the effects of organic matter, micro-organisms and colloids, and develop the assessment methodology. We summarise the evaluation methodology using the deep Wakkanai Formation as a case study for block-scale solute transport experiments. As for "Demonstration of repository design options", we summarise the results of investigations and experiments on changes in the geological environment after tunnel excavation and closure, and summarise the applicability and technical challenges of the closure technology for boreholes excavated from tunnels. The systematic integration of technologies towards EBS emplacement, including the organisation of investigation and evaluation methods and analysis, will be promoted. Experiments to confirm the performance of the engineered barrier system under critical conditions, such as high temperatures ($$>$$100$$^{circ}$$C), continue the in-situ tests started in 2023.

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