An Analytical study on Russia's civilian nuclear energy utilization

Kawasaki, Nobuchika

JAEA-Review 2025-043, 74 Pages, 2025/10

JAEA-Review-2025-043.pdf:2.45MB

Russia is one of the most advanced countries in the civilian use of nuclear energy. However, understanding the internal mechanisms of its nuclear program remains difficult due to various reasons. Therefore, this study presents a historical overview of Russia's nuclear energy utilization, fuel supply, fuel manufacturing capabilities, and concepts regarding reprocessing and the nuclear fuel cycle. From this overview, insights have been extracted and analyzed. These insights are then organized under two strategic perspectives: "Strategic diversity and continuity in developments and demonstrations" and "Diversity in utilizations and deployments," with considerations of implications for Japan, as below. Russia's nuclear energy policy strategically utilizes a variety of reactor types and fuel cycle technologies to expand nuclear power generation both domestically and internationally. Currently, nuclear power, centered on light-water reactors (VVER series), accounts for about 20% of Russia's electricity supply, and there are plans to increase this share to 25% by 2045. A wide range of reactors, from large-scale to medium and small modular reactors, are being constructed in Russia. Russia is also actively developing fast reactor technologies, and focusing on the reprocessing and recycling of spent fuel. Internationally, VVER-1200 reactors are under construction in several countries, and cooperation with China is deepening in the field of fast reactors. Notably, Russia offers an integrated, or selectively customizable, package of nuclear technology services on the international stage. These include not only reactor deployment, but also fuel supply, reprocessing, waste management, and even the provision of radioisotopes. Rather than simply exporting products or technology, Russia fosters long-term relationships and trust by flexibly responding to the conditions and needs of partner countries. For this reason, Russia promotes the technology developments in advance within the country in areas anticipated for future overseas deployment. It carefully selects target technologies and services and systematically rolls them out. This flexible strategy, combining "technological diversity" and "strategic consistency", enables cooperation with countries across various geopolitical contexts. For Japan, this strategic approach offers valuable lessons on how to engage in comprehensive international nuclear cooperation, not merely through technology exports, but through integrated approaches that encompass the entire fuel cycle, and by combining elements such as fast reactors and RI supply.

Oxygen potential and oxygen diffusion data for guiding the manufacture of MOX fuel for fast neutron reactors

Vauchy, R.; Horii, Yuta; Hirooka, Shun; Akashi, Masatoshi; Sunaoshi, Takeo*; Nakamichi, Shinya; Saito, Kosuke

Journal of Nuclear Materials, 616, p.156115_1 - 156115_16, 2025/10

https://doi.org/10.1016/j.jnucmat.2025.156115

Development of evaluation method for transition behavior of non-condensable gas in primary coolant system of pool-type sodium-cooled fast reactor; Preliminary evaluation of bubble detachment behavior from free surface in cold plenum region

Matsushita, Kentaro; Ezure, Toshiki; Fujisaki, Tatsuya*; Nakamine, Yoshiaki*; Imai, Yasutomo*; Tanaka, Masaaki

Nihon Kikai Gakkai 2025-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2025/09

In the design of sodium-cooled fast reactors (SFRs), it is important to evaluate the transition behavior of non-condensable gas entrained into the primary coolant system due to cover gas entrainment and dissolution. In this study, trajectories of non-condensable gas bubbles in the cold plenum of the pool-type SFR evaluated by three-dimensional CFD analyses applying Discrete Phase Model. As the result of sensitivity analyses regarding bubble radius flowing into the cold plenum, it was clarified that the release rate of bubbles showed an increase according to the increase of bubble radius and an asymptotic increasing behavior in the large bubble radius cases.

Reaction behavior between sodium and molten salt caused by the heat transfer tube failure for sodium-cooled fast reactor coupled to thermal energy storage system

Sato, Rika; Kondo, Toshiki; Umeda, Ryota; Kikuchi, Shin; Yamano, Hidemasa

Progress in Nuclear Science and Technology (Internet), 8, p.137 - 142, 2025/09

https://doi.org/10.15669/pnst.8.137

In a sodium-cooled fast reactor (SFR) coupled to thermal energy storage (TES) system, the reaction between nitrate molten salt as thermal energy storage medium and sodium (Na) as reactor coolant might occur under postulated accidental conditions. Thus, the reaction behavior of Na-nitrate molten salt is one of the important phenomena in terms of safety assessment of the SFR with TES system. In this study, reaction experiments on Na-solar salt were performed. It was found that Na-solar salt reaction occurred after the NaNO-KNO eutectic melting. Based on the measured reaction temperature, the kinetic parameters and rate constant were obtained and compared with the sodium-water reaction. From the results of kinetic analysis, it could be assumed that Na-solar salt reaction occurs in the time frame of the accident such as the failure of heat transfer tube of sodium-molten salt heat exchanger.

Achievement of safety demonstration tests using HTTR; Loss of forced cooling test at 100% reactor power (30 MW)

Nagasumi, Satoru; Hasegawa, Toshinari; Nakagawa, Shigeaki; Kubo, Shinji; Iigaki, Kazuhiko; Shinohara, Masanori; Saikusa, Akio; Nojiri, Naoki; Saito, Kenji; Furusawa, Takayuki; et al.

JAEA-Research 2025-005, 23 Pages, 2025/07

JAEA-Research-2025-005.pdf:2.68MB

A safety demonstration test under abnormal operating conditions using the HTTR (High Temperature Engineering Test Reactor) was conducted to demonstrate safety features of the HTGRs (High Temperature Gas-cooled Reactors). Under a simulation of a control rod shutdown failure, all primary helium gas circulators were intentionally stopped during a steady-state operation at 100% reactor thermal power (30 MW), temporal changes of the reactor power and temperatures around the reactor pressure vessel (RPV) were obtained after the complete loss of forced heat removal from the reactor core. After the event (primary coolant flow stopped), the reactor power quickly decreased due to the negative reactivity feedback associated with the core temperature rise, and then the reactor power spontaneously shifted to a stable state of low power (about 1.2%) even after a recriticality. Heat dissipation from RPV surface to a surrounding vessel cooling system (water-cooled panels) ensured the amount of heat removal required to maintain the reactor temperature constant in the low power state. In this way, the transition from the event occurrence to the stable and safety state, i.e., inherent safety features of HTGRs, were demonstrated in the case of core forced cooling loss without active shutdown operations.

Non-condensable gas accumulation and distribution due to condensation in the CIGMA Facility; Implications for Fukushima Daiichi Unit 3 (1F3)

Hamdani, A.; Soma, Shu; Abe, Satoshi; Shibamoto, Yasuteru

Progress in Nuclear Energy, 185, p.105771_1 - 105771_13, 2025/07

https://doi.org/10.1016/j.pnucene.2025.105771

Steam Explosion Simulation Code JASMINE v.3 User's Guide; Revised for code version 3.3c

Iwasawa, Yuzuru; Matsumoto, Toshinori; Moriyama, Kiyofumi*

JAEA-Data/Code 2025-001, 199 Pages, 2025/06

JAEA-Data-Code-2025-001.pdf:9.71MB

A steam explosion is defined as a phenomenon that occurs when a hot liquid comes into contact with a low-temperature cold liquid with volatile properties. The rapid transfer of heat from the hot liquid to the cold liquid results in a chain reaction of the explosive vaporization of the cold liquid and fine fragmentation of the hot liquid. The explosive vaporization of the cold liquid initiates the propagation of shock waves in the cold liquid. The expansion of the hot and cold liquid mixture exerts mechanical forces on the surrounding structures. In severe accidents of light water reactors, the molten core (melt) is displaced into the coolant water, resulting in fuel-coolant interactions (FCIs). The explosive FCI, referred to as a steam explosion, has been identified as a significant safety assessment issue as it can compromise the integrity of the primary containment vessel. The JASMINE code is an analytical tool developed to evaluate the mechanical forces imposed by steam explosions in nuclear reactors. It performs numerical simulations of steam explosions in a mechanistic manner. The present report describes modeling concepts, basic equations, numerical solutions, and example simulations, as well as instructions for input preparation, code execution, and the use of supporting tools for practical purpose. The present report is the updated version of the "Steam Explosion Simulation Code JASMINE v.3 User's Guide, JAEA-Data/ Code 2008-014". The present report was compiled and updated based on the latest version of the code, JASMINE 3.3c, with corrections for minor errors of the distributed code JASMINE 3.3b, and conformance to recently widely used compilers on UNIX-like environments such as the GNU compiler. The numerical simulations described in the present report were obtained using the latest version JASMINE 3.3c. The latest parameter adjustment method for a model parameter proposed by the previous study is employed to conduct the numerical simulations.

Application study of adaptive mesh refinement method on unsteady wake vortex analysis

Alzahrani, H.*; Matsushita, Kentaro; Sakai, Takaaki*; Ezure, Toshiki; Tanaka, Masaaki

Nuclear Technology, 13 Pages, 2025/06

https://doi.org/10.1080/00295450.2025.2472582

Development of evaluation method for cover gas entrainment by vortices generated at free surface in upper plenum of sodium-cooled fast reactor is required, and an evaluation method by predicting vortices from flow velocity distribution obtained by CFD analysis is developed. In this study, Adaptive Mesh Refinement (AMR) method is examined to improve efficiency of CFD analysis. Initial mesh was refined with two indexes: the first index (Index-1) is when the second invariant of velocity gradient tensor, Q, is negative and the second one (Index-2) is pressure gradient index added to Index-1. As a result of applying AMR method to unsteady vortices system with a flat plate and performing transient analyses with refined meshes, the result of pressure distribution and velocity around the flat plate in mesh using Index-2 was similar to the result of all refined mesh. It was also confirmed that vortices generation and growth was better simulated by refining meshes around separation area.

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.

Annual report of Nuclear Science Research Institute, JFY 2022

Nuclear Science Research Institute

JAEA-Review 2024-057, 178 Pages, 2025/03

JAEA-Review-2024-057.pdf:8.51MB

Nuclear Science Research Institute (NSRI) is composed of Planning and Management Department and six departments, namely Department of Operational Safety Administration, Department of Radiation Protection, Engineering Services Department, Department of Research Reactor and Tandem Accelerator, Department of Criticality and Hot Examination Technology and Department of Decommissioning and Waste Management, and each department manages facilities and develops related technologies to achieve the "Medium- to Long-term Plan" successfully and effectively. And, four research centers which are Advanced Science Research Center, Nuclear Science and Engineering Center, Nuclear Engineering Research Collaboration Center and Materials Sciences Research Center, belong to NSRI. In order to contribute the future research and development and to promote management business, this annual report summarizes information on the activities of NSRI of JFY 2022 as well as the activity on research and development carried out by Collaborative Laboratories for Advanced Decommissioning Science, Nuclear Safety Research Center and activities of Nuclear Human Resource Development Center, using facilities of NSRI.