Trading cards of nuclear technology; A Science communication tool developed for gaining literacy to read nuclear policy documents

Fukaya, Yuji; Yan, X.; Matsuo, Yuji*; Hayashi, Reona*

Japanese Journal of Science Communication (Internet), (36), p.45 - 56, 2025/03

https://doi.org/10.14943/112887

From the 1956 "Long-Term Program on the Development and Utilization of Atomic Energy" to the present-day "Strategic Energy Plan", documents related to relevant committees and nuclear power policies have been accumulated, and many of them may be red on the Internet. Moreover, there are diverse explanations and opinions based on these policy documents, as well as many papers on research and development that reflect them. On the other hand, 1956 was a time when Japan was transitioning from post-war reconstruction to high economic growth and was clearly aware of the possibility of war in the form of the Cold War. Under such circumstance, unless one is knowledgeable to appreciate, to some extent, the context and the source from which the benefits of a particular technology claimed in a particular policy document originate, it would be difficult to predict the actual effects of the technology when it is implemented in society. This might also mislead the general public, making it difficult to reflect the public opinion correctly in new policies. To address these dilemmas, a trading card game has been developed in which the players participate in the role-playing process as a proponent of a particular nuclear technology. The process consists of four phases, namely technology selection, understanding the current situation, policy making, and justification of the selected technology under the policy direction. The game could allow the players to predict the intent of the technology and policy communicator.

An Evaluation on Inelastic Thermal Neutron Scattering Cross-Section Data of Crystalline Graphite

Okita, Shoichiro; Abe, Yutaka*; Tasaki, Seiji*; Fukaya, Yuji

Radioisotopes, 73(3), p.233 - 240, 2024/11

https://doi.org/10.3769/radioisotopes.73.233

Fuel cycle scenarios and back-end technologies of HTGR in Japan

Fukaya, Yuji; Goto, Minoru; Shibata, Taiju

IAEA-TECDOC-2040, p.133 - 136, 2023/12

https://www-pub.iaea.org/MTCD/Publications/PDF/TE-2040web.pdf

Japan has developed back-end technologies to establish a multi-recycling fuel cycle with fast breeder reactors (FBRs) to ensure energy resources. Even though the development of FBR has been retreated to one of fundamental research, the reprocessing technologies for uranium fuel and disposal technologies had been completed for Light Water Reactor (LWR) fuel cycle on the process. These technologies were inherited to utilities and are about to be practical. Now, Japan had been completed High Temperature Engineering Test Reactor (HTTR) a prototype and research reactor, a commercial High Temperature Gas-cooled Reactor (HTGR) design Gas Turbine High Temperature Reactor 300 (GTHTR300) with related reprocessing technologies, and is planning domestic demonstration reactor project. In this context, a representative fuel cycle policy is reprocessing in Japan. However, Japan has investigated various fuel cycle scenarios to expand the usage of the commercial HTGR. Then, we would like to introduce the scenarios and development status of related technologies in the present study.

Reactor physics experiment on a graphite-moderated core to construct integral experiment database for HTGR

Okita, Shoichiro; Fukaya, Yuji; Sakon, Atsushi*; Sano, Tadafumi*; Takahashi, Yoshiyuki*; Unesaki, Hironobu*

Nuclear Science and Engineering, 197(8), p.2251 - 2257, 2023/08

https://doi.org/10.1080/00295639.2022.2087836

Feasibility study on reprocessing of HTGR spent fuel by existing PUREX plant and technology

Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi

Annals of Nuclear Energy, 181, p.109534_1 - 109534_10, 2023/02

https://doi.org/10.1016/j.anucene.2022.109534

Feasibility of reprocessing of High Temperature Gas-cooled Reactor (HTGR) spent fuel by existing Plutonium Uranium Redox EXtraction (PUREX) plant and technology has been investigated. The spent fuel dissolved solution includes approximately 3 times amount of uranium-235 and 1.5 times amount of protonium because of the 3 times higher burnup compared with that of Light Water Reactor (LWR). Then, the heavy metal of the spent fuel is planned to be diluted to 3.1 times by depleted uranium to satisfy the limitation of Rokkasho Reprocessing Plant (RRP) plant. In the present study, recoverability of uranium and plutonium with the dilution is confirmed by a simulation with a reprocessing process calculation code. Moreover, the case without the dilution from the economic perspective is investigated. As a result, the feasibility is confirmed without the dilution, and it is expected that the reprocessed amount is reduced to 1/3 compared with a diluted case even though the facility should be optimized from the perspective of mass flow and criticality.

Reactor physics experiment in a graphite moderation system for HTGR, 3

Fukaya, Yuji; Okita, Shoichiro; Kanda, Shun*; Goto, Masaki*; Nakajima, Kunihiro*; Sakon, Atsushi*; Sano, Tadafumi*; Hashimoto, Kengo*; Takahashi, Yoshiyuki*; Unesaki, Hironobu*

KURNS Progress Report 2021, P. 101, 2022/07

https://www.rri.kyoto-u.ac.jp/PUB/report/PR/ProgRep2021/ProgRep2021.html

The Japan Atomic Energy Agency (JAEA) started the Research and Development (R&D) to improve nuclear prediction techniques for High Temperature Gas-cooled Reactors (HTGRs) in 2018. The objectives are to intro-duce the generalized bias factor method to avoid full mock-up experiment for the first commercial HTGR and to improve neutron instrumentation system by virtue of the particular characteristics due to a graphite moderation system. For this end, we composed B7/4"G2/8"p8EU(3)+3/8"p38EU in the B-rack of Kyoto University Critical Assembly (KUCA) in 2021.

Re-evaluation of electricity generation cost of HTGR

Fukaya, Yuji; Ohashi, Hirofumi; Sato, Hiroyuki; Goto, Minoru; Kunitomi, Kazuhiko

Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 21(2), p.116 - 126, 2022/06

https://doi.org/10.3327/taesj.J21.005

An improvement electricity generation cost evaluation method for High Temperature Gas-cooled Reactors (HTGRs) has been performed. Japan Atomic Energy Agency (JAEA) had completed the commercial HTGR concept named Gas Turbine High Temperature Reactor (GTHTR300) and the electricity generation cost evaluation method approximately a decade ago. The cost evaluation was developed based on the method of Federation of Electric Power Companies (FEPC). The FEPC method was drastically revised after the Fukushima Daiichi nuclear disaster. Moreover, the escalation of material and labor cost for the decade should be consider to evaluate the latest cost. Therefore, we revised the cost evaluation method for GTHTR300 and the cost was compared with that of Light Water Reactor (LWR). As a result, it was found that the electricity generation cost of HTGR of 7.9 yen/kWh is cheaper than that of LWR of 11.7 yen/kWh by approximately 30% at the capacity factor of 70%.

Computed tomography neutron detector system to observe power distribution in a core with long neutron flight path

Fukaya, Yuji; Okita, Shoichiro; Nakagawa, Shigeaki; Goto, Minoru; Ohashi, Hirofumi

Annals of Nuclear Energy, 168, p.108911_1 - 108911_7, 2022/04

https://doi.org/10.1016/j.anucene.2021.108911

A power distribution monitoring system by using a moving detector for a core with a long neutron flight path has been proposed. High Temperature Gas-cooled Reactor (HTGR) and Fast Reactor (FR) has a long neutron flight path and the neutrons reach to detector far from fuel assembly in the center of the core unlike Light Water Reactor (LWR). By using the feature, power distribution can be observed with a few detectors by moving the detector and computed tomography technology similar to X-ray Computed Tomography (CT). For a small-sized core, the power distribution can be evaluated only by an ex-core neutron detector. For a large-sized core with inner detectors, the power distribution can be observed with a small number of in-core detectors even if the deployment is limited due to material integrity conditions such as temperature environment. The feasibility is numerically confirmed by simulations of the HTGR core and its detector response. It is expected to observe the power distribution in the core of HTGR and FR, which is difficult continuously to deploy in-core detectors because of high temperature and/or high irradiation damage.

Reduction of the source term of an assumed criticality accident in a fuel fabrication facility with solution system

Fukaya, Yuji; Goto, Minoru

Annals of Nuclear Energy, 164, p.108617_1 - 108617_6, 2021/12

https://doi.org/10.1016/j.anucene.2021.108617

A reasonable source term of a hypothetical criticality accident for fuel fabrication facility with solution system has been proposed. The public exposure must not exceed the limitation of 5 mSv during an accident. Then, we proposed the reasonable source term of the first burst peak due to the hydrogen gas generation by radiation decomposition of water. With the criticality control system composed of the Criticality Accident Alarm System (CAAS) and soluble neutron absorber, safety is ensured by the reduced fission number. We confirmed the effect by environmental impact assessment during a criticality accident by using site condition of a fuel fabrication facility in Tokai-mura, Japan. As a result, the public exposure is reduced at a site boundary from 68 mSv to 0.6 mSv under the current regulatory guideline.

Manufacturability estimation on burnable poison mixed fuel for improving criticality safety of HTGR fuel fabrication

Hasegawa, Toshinari; Fukaya, Yuji; Ueta, Shohei; Goto, Minoru

Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 5 Pages, 2021/08

https://doi.org/10.1115/ICONE28-61763

Burnable poison (BP) credit concept has been proposed as a criticality safety measure for commercial high temperature gas-cooled reactor (HTGR) fuel fabrication, so we estimated manufacturability of the BP-mixed UO kernel for the practical use of the concept. As a BP, boron, gadolinium, erbium, and hafnium are investigated. Boron mixed fuel kernels are fabricated by mixing boric acid powder with UO powder. In the case of the other BPs, BP nitrate powder is mixed with UO powder. In order to confirm that BP remain in the kernels after the heat treatment processes, thermodynamic equilibrium analysis was performed. Above 450C, boron would melt and vaporize during the heat treatment processes, so it was found that the boron mixed fuel kernel fabrication is difficult. On the other hand, it was found that gadolinium, erbium, and hafnium would change to solid oxides that do not melt and vaporize even at 2000C, and there was no problem with manufacturability of the BP-mixed fuel kernel.