Development of nuclear security technologies for a large public event (Collaboration research)

Takahashi, Tone; Mochimaru, Takanori*; Koizumi, Mitsuo; Yoshimi, Yuki*; Yamanishi, Hirokuni*; Wakabayashi, Genichiro*; Ito, Fumiaki*

JAEA-Review 2025-039, 34 Pages, 2025/11

JAEA-Review-2025-039.pdf:2.18MB

To prevent acts of terrorism involving nuclear or radioactive materials at major public events, it is required to have surveillance technologies that either prevent these materials from being brought in or detect quickly if somebody brings them in secretly. Setting radiation gate monitors to survey pedestrians and vehicles is one of the effective methods. However, considering the possibility of individuals bypassing these monitors, complementary technologies are needed to continuously survey areas inside the gates. To survey extensive areas, radiation mapping is effective. By using multiple detectors and aggregating the data, the survey becomes much more efficient. We have developed mobile detectors capable of simultaneously measuring location data and radiation levels outdoors, with the ability to aggregate measurement results via a network and immediately visualize them on a map. For indoor environments, we have developed a technology that integrates radiation measurement results with environmental mapping created using SLAM (Simultaneous Localization and Mapping) to produce 3D maps of the surveyed areas. Additionally, we have been working on the development of a source search technology using a fast neutron detector to quickly detect neutron sources, including nuclear materials. In this report, we describe a concept of the wide area survey system and report technology development results so far.

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.

Difference in accumulation of plutonium and curium isotopes formed in americium targets irradiated in Joyo and JMTR

Onishi, Takashi; Koyama, Shinichi*; Yokoyama, Keisuke; Morishita, Kazuki; Watanabe, Masashi; Maeda, Shigetaka; Yano, Yasuhide; Oki, Shigeo

Nuclear Engineering and Design, 432, p.113755_1 - 113755_17, 2025/02

https://doi.org/10.1016/j.nucengdes.2024.113755

Applicability of mechanistic analysis code seraphim to sodium-water reaction in tube bundle system

Kosaka, Wataru; Uchibori, Akihiro; Watanabe, Akira*; Shiina, Yoshimi*

Proceedings of 10th Workshop on Computational Fluid Dynamics for Nuclear Reactor Safety (CFD4NRS-10) (Internet), 12 Pages, 2025/00

First freezing experiments with a molten mixture of boron carbide and stainless steel in core disruptive accidents of sodium-cooled fast reactors

Emura, Yuki; Matsuba, Kenichi; Kikuchi, Shin; Yamano, Hidemasa

Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 8 Pages, 2024/11

High-temperature strength of modified type 316 steel for fast reactor fuel before and after neutron irradiation

Miyazawa, Takeshi; Uwaba, Tomoyuki; Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Onizawa, Takashi; Ando, Masanori; Kaito, Takeji

JAEA-Technology 2024-009, 140 Pages, 2024/10

JAEA-Technology-2024-009.pdf:8.03MB

For the purpose of enhancing the reliability of fast reactor fuel designing using modified type 316 steel, the out-of-pile and in-pile mechanical data of modified type 316 steel cladding tubes and wrapper tubes were statistically analyzed with the knowledge on material science and engineering; the high-temperature strength equations of modified type 316 steel, which can be applied to high-dose neutron irradiation environment, were derived. The out-of-pile high-temperature tensile and creep data of modified type 316 steel cladding tubes and wrapper tubes were derived up to 900C, which is higher than the upper limit temperature of anticipated transient event of fast reactor. Using the extended database, the best-fit equation and the lower limit equation were derived for out-of-pile 0.2% proof strength, ultimate tensile strength and creep rupture strength while the best-fit equation and the upper and lower limit equations for creep strain. Furthermore, the degradation factors for tensile and creep strength, which will be produced by in-reactor environment (i.e., neutron irradiation in liquid sodium), were evaluated using the existing neutron irradiation data of modified type 316 steel, which were derived using the experimental fast reactor Joyo, the French proto-type fast reactor Phenix, the American experimental fast reactor FFTF. The derived equations were validated by the comparison with the experimental data.

Study on eutectic melting behavior of control rod materials in severe accidents of sodium-cooled fast reactors, 3; Material analysis of boron carbide immersed in molten stainless steel

Takai, Toshihide; Emura, Yuki; Yamano, Hidemasa

Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NUTHOS-14) (Internet), 11 Pages, 2024/08

Application of the GIF safety design criteria and safety design guidelines on decay heat removal system to next generation sodium-cooled fast reactor in Japan

Yamano, Hidemasa; Futagami, Satoshi; Higurashi, Koichi*

Proceedings of Advanced Reactor Safety (ARS 2024), p.121 - 130, 2024/08

This paper describes the application of safety design criteria (SDC) and safety design guideline (SDG) developed in the Generation-IV international forum on decay heat removal system (DHRS) enhancing reliability to sodium-cooled fast reactors (SFRs) recently designed in Japan.

Numerical simulation of accidents involving core damage with integrative severe accident analysis code, SPECTRA

Ishida, Shinya; Uchibori, Akihiro; Okano, Yasushi

Dai-28-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2024/06

no abstracts in English