A Study on the methodology for rational treatment/disposal of contaminated concrete waste considering volume reduction of waste (Contract research); FY2023 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Hokkaido University*

JAEA-Review 2025-037, 103 Pages, 2025/12

JAEA-Review-2025-037.pdf:7.28MB

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 FY2023. 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 FY2023, this report summarizes the research results of the "A study on the methodology for rational treatment/disposal of contaminated concrete waste considering volume reduction of waste" conducted in FY2023. The present study aims to evaluate rational waste management strategies incorporating reuse and recycling focusing on radioactive concrete waste, which will be massively generated from dismantling. Feasibility and challenges of aggregate recycling are considered assuming a typical recycled aggregate production technique, based on the characteristics of the concrete. In 2023, the migration behaviors of radionuclides and ions in cementitious materials having interfacial transition zones (ITZ) were investigated through diffusion and leaching experiments using radioactive and non-radioactive tracers and modeled by random walk particle tracking method with a sampling technique using a probabilistic distribution model for two media with an interface. Properties of surrogate contaminated concrete samples prepared by immersing in Cs solution were examined. Migration of ions was studied for surrogate contaminated aggregates and recycled concrete using the surrogate. In addition, surrogate waste package was prepared using by-product powder to study mechanical and chemical properties as well as leaching behavior of radionuclides. Information on properties of the contaminated concrete and tools to estimate the amount of concrete were organized in order to evaluate different waste management scenarios incorporating reuse/recycling.

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.

Data report of ROSA/LSTF experiment SB-PV-03; 0.2% pressure vessel bottom break LOCA with SG depressurization and gas inflow

Takeda, Takeshi

JAEA-Data/Code 2024-014, 76 Pages, 2024/12

JAEA-Data-Code-2024-014.pdf:4.0MB

An experiment denoted as SB-PV-03 was conducted on November 19, 2002 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-PV-03 simulated a 0.2% pressure vessel bottom small-break loss-of-coolant accident in a pressurized water reactor (PWR). The test assumptions included total failure of high pressure injection system of emergency core cooling system (ECCS) and noncondensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of ECCS. Secondary-side depressurization of both steam generators (SGs) as an accident management (AM) action to achieve the depressurization rate of 55 K/h in the primary system was initiated 10 min after the generation of a safety injection signal, and continued afterwards. Auxiliary feedwater injection into the secondary-side of both SGs was started for 30 min with some delay after the onset of the AM action. The AM action was effective on the primary depressurization until the ACC tanks began to discharge nitrogen gas into the primary system. The core liquid level recovered in oscillative manner because of intermittent coolant injection from the ACC system into both cold legs. Therefore, the core liquid level remained at a small drop. The pressure difference between the primary and SG secondary sides became larger after nitrogen gas ingress. Core uncovery occurred by core boil-off during reflux condensation in the SG U-tubes under nitrogen gas influx. When the maximum cladding surface temperature of simulated fuel rods exceeded the pre-determined value of 908 K, the core power was automatically reduced to protect the LSTF core. After the automatic core power reduction, coolant injection from low pressure injection (LPI) system of ECCS into both cold legs led to the whole core quench. After the continuous core cooling was confirmed through the actuation of the LPI system, the experiment was terminated.

Design policy of pilot plant for accelerator-driven system

Nishihara, Kenji; Sugawara, Takanori; Fukushima, Masahiro; Iwamoto, Hiroki; Katano, Ryota; Abe, Takumi

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

A pilot plant for the accelerator-driven system is proposed as a scaled-down version of a lead-bismuth cooled ADS with 800 MW thermal output for transmutation of minor actinides. In this presentation, the design policy of the pilot plant is presented.

Development of zeolite column adsorption dynamics simulation code (ZAC)

Yamagishi, Isao; Hato, Shinji*; Nishihara, Kenji; Tsubata, Yasuhiro; Sagawa, Yusuke*

JAEA-Data/Code 2024-002, 63 Pages, 2024/07

JAEA-Data-Code-2024-002.pdf:2.91MB
JAEA-Data-Code-2024-002-appendix(CD-ROM).zip:9.42MB

Adsorption columns filled with zeolite are used to treat contaminated water containing radioactive cesium generated by the Fukushima Daiichi Nuclear Power Station accident. As the contaminated water treatment progresses, the radioactive cesium in the adsorption column becomes highly concentrated, and the adsorption column becomes a high radiation source. To evaluate the radiation effects such as decay heat and radiolytic hydrogen production in the adsorption column, the concentration of radioactive cesium in the adsorption column is necessary, but since it is difficult to evaluate the concentration by measurement, it is estimated by simulation. In this research, a zeolite column adsorption dynamics simulation (Zeolite Adsorption Column: ZAC) code was developed to calculate the concentration of radioactive materials such as radioactive cesium in a zeolite filled adsorption column when they are injected into the column. The code was validated through comparison of calculation results with existing codes and experimental results of small column tests. This report presents the details of the model, the handling of the code, and the validity of the results for the developed code.

Report on the lecture of standard committee in the 2023 fall meeting entitled "Standardization Activities for Safe Long Term Operation"

Murakami, Kenta*; Onizawa, Kunio; Yamamoto, Akio*

Nihon Genshiryoku Gakkai-Shi ATOMO, 66(4), p.199 - 202, 2024/04

https://doi.org/10.3327/jaesjb.66.4_199

The Standards Committee of Atomic Energy Society of Japan has been leading activities related to long-term operation through the revision of the Code of Practice for Aging Management, and we believe that we must continue to make important contributions in light of recent changes in laws and regulations. This paper recapitulates the discussions in the special session conducted at the 2023 fall meeting, and describes the efforts toward safe long-term operation and the points to keep in mind in the standardization of such activities. The important points are (1) to make effective use of knowledge found over time, (2) not to overlook new knowledge that has a significant impact on safety, including obsolescence, (3) to assign a level of importance to the response based on the impact on safety and the likelihood of its occurrence, and (4) to contribute to the establishment of an international knowledge base.

Survey and proposal for Japanese-English bilingual translation of technical terms focusing on nuclear disaster prevention

Togawa, Orihiko; Okuno, Hiroshi

JAEA-Review 2023-043, 94 Pages, 2024/03

JAEA-Review-2023-043.pdf:1.53MB

In order to translate nuclear disaster prevention documents written in Japanese into English, the Basic Act on Disaster Management, the Act on Special Measures Concerning Nuclear Emergency Preparedness, and the Convention on Nuclear Safety were surveyed for corresponding terms in Japanese and English. The survey results were integrated and unified English translations were selected. As a result, a Japanese-English correspondence table of technical terms in the field of nuclear disaster prevention was prepared and proposed.

Reviewing codes and standards for long term operation in Japan

Murakami, Kenta*; Arai, Taku*; Yamada, Koji*; Momma, Kensuke*; Tsuji, Takashi*; Nakagawa, Nobuyuki*; Onizawa, Kunio

Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 3 Pages, 2024/03

This paper studied the future vision of codes and standards in Japan by systematically comparing Japanese regulatory rules, standards, and industry guides related to long term operation with international safety standards, and confirmed that the Japanese standard system generally meets their recommendations. The recommendation for the future improvements of Japanese codes and standards were summarized into five items.

Effect of preparation conditions and storage time on characteristic and rheological properties of carbonate slurries

Kato, Tomoaki; Yamagishi, Isao

JAEA-Technology 2023-018, 53 Pages, 2023/11

JAEA-Technology-2023-018.pdf:2.6MB

In the decommissioning of Fukushima Daiichi Nuclear Power Station, radioactive carbonate slurry waste was generated using the Advanced Liquid Processing System (ALPS) pretreatment and temporarily stored in a high integrity container (HIC). In 2015, overflow of supernatant from HIC estimate as bubble retention in the carbonate slurry was discovered, increasing the need for a safety assessment of the carbonate slurry stored the HIC (HIC slurry). In this study, a carbonate slurry (simulated slurry) was prepared according to the Mg/Ca mass ratio in the ALPS inlet water of the HIC slurry which overflew the HIC. The effects of reaction time during the pretreatment process, suspended solids concentration (SS concentration), and settling time on the particle composition, morphology and rheological properties of the slurry were investigated. Evaluating the effect of reaction time and concentration process on chemical properties in slurry production, the effect of the reaction time was not confirmed in the simulated slurry that had undergone the concentration process, and slurry prepared at SS concentration of 150 g/L was composed of formless particles have a particle diameter of 0.4 m or less. We also investigate the effect of SS concentration on sedimentability, decrease in SS concentration by dilution with processing solution contributed to an increase in the initial slurry settling velocity. Furthermore, two different flow characteristics were observed depending on the settling time, suggesting that the slurry at the initial settling time has non-Bingham flow properties, whereas it changes to Bingham flow properties as the settling time becomes longer. In addition, yield stress was increased with settling time, and this yield stress was found to be exponentially proportional to the density of the slurry. These results provide knowledge to estimate the current state of HIC slurry and are expected to contribute to the safety assessment.

Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR) (FY2021)

Department of HTTR

JAEA-Review 2023-016, 82 Pages, 2023/09

JAEA-Review-2023-016.pdf:2.31MB

The High Temperature Engineering Test Reactor (HTTR) is the first Japanese High Temperature Gas-cooled Reactor (HTGR) with 30MW in thermal power and 950C of maximum outlet coolant temperature that is constructed by the Japan Atomic Energy Agency located at Oarai-machi, Higashiibaraki-gun, Ibaraki-ken, Japan. The purpose of the HTTR is establishment of basic HTGR technologies, demonstration of HTGR safety characteristics and so on. The HTTR has had a lot of experience of HTGRs' operation and maintenance throughout rated power operations, safety demonstration tests, long-term high temperature operations and demonstration tests relevant to HTGRs' R&Ds. In the fiscal year 2021, as the HTTR completed activities to conform to the New Regulatory Requirements of Nuclear Regulation Authority, The HTTR restarted since the 2011 off the Pacific coast of Tohoku Earthquake and carried out the Loss-of-forced cooling test without Vessel Cooling System (VCS) operational at 9MW (Three gas circulators trip and VCS is stopped.) as the safety demonstration test. This report summarizes the activities carried out in the fiscal year 2021, which were the situation of the New Regulatory Requirements screening of the HTTR, the operation and maintenance of the HTTR, R&Ds relevant to commercial-scale HTGRs, the international cooperation on HTGRs and so on.