Challenge of novel hybrid-waste-solidification of mobile nuclei generated in Fukushima Nuclear Power Station and establishment of rational disposal concept and its safety assessment (Contract research); FY2023 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Institute of Science Tokyo*

JAEA-Review 2025-016, 143 Pages, 2025/10

JAEA-Review-2025-016.pdf:10.71MB

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 (hereafter referred to "1F"), 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 "Challenge of novel hybrid-waste-solidification of mobile nuclei generated in Fukushima Nuclear Power Station and establishment of rational disposal concept and its safety assessment" conducted from FY2021 to FY2023. This study aims to establish the rational waste disposal concept of various wastes generated in 1F based on the hybrid-waste-solidification by the Hot Isostatic Press (HIP) method. The ceramics form with target elements, mainly iodine, which is challenging to immobilize, and Minor Actinides such as Am, an alpha emitter and heat source, are HIPed with well-studied materials such as SUS and zircaloy, which make the long-term stability evaluation and safety assessment possible. In 2024, the project's final year, we demonstrated the effectiveness of the hybrid solidification concept by linking all the sub-themes, from waste synthesis to disposal considerations. The compatibility of various wastes, such as ALPS, AREVA sediment wastes, AgI, waste silver adsorbent, ceria adsorbent, and iodine apatite, with metals and oxide matrices was investigated. which involves investigating the HIPed hybrid wastes after exploring the compatibility of various metals and oxide matrices using the rapid sintering method, spark plasma sintering (SPS), proposed in this project. It revealed that hybrid waste solidification with SUS matrix was superior for many wastes. Furthermore, we studied waste disposal concepts based on nuclide migration calculations. Finally, we could connect the waste fabrication to safety assessment for the first time, leading to finding an appropriate waste disposal scenario for 1F decommissioning.

Heat transfer coefficient modeling for downward saturated boiling flows in vertical pipes

Wada, Yuki; Shibamoto, Yasuteru; Hibiki, Takashi*

International Journal of Heat and Mass Transfer, 249, p.127219_1 - 127219_16, 2025/10

https://doi.org/10.1016/j.ijheatmasstransfer.2025.127219

Simple technique for the preparation of uranium-impregnated porous silica particles and their application as working standard particles for analysis of the safeguards environmental samples

Tomita, Jumpei; Tomita, Ryohei; Suzuki, Daisuke; Yasuda, Kenichiro; Miyamoto, Yutaka

Journal of Nuclear Science and Technology, 12 Pages, 2025/09

https://doi.org/10.1080/00223131.2025.2557376

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.

Fuel debris criticality analysis technology using non-contact measurement method (Contract research); FY2023 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Institute of Science Tokyo*

JAEA-Review 2025-010, 62 Pages, 2025/08

JAEA-Review-2025-010.pdf:3.63MB

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 FY2021, this report summarizes the research results of the "Fuel debris criticality analysis technology using non-contact measurement method" conducted from FY2021 to FY2023. The purpose of research was to improve the fuel debris criticality analysis technology using non-contact measurement method by the development of the fuel debris criticality characteristics measurement system and the multi-region integral kinetic analysis code. It was performed by Institute of Science Tokyo, Tokyo City University, National Institute of Advanced Industrial Science and Technology, and Nagaoka University of Technology. We developed the fuel debris criticality characteristics measurement system which has a two layer structure surrounding a canister containing fuel debris fragments with He proportional counters. The operational validation and performance evaluation were performed on the developed detector system. We have demonstrated the feasibility and accuracy of measuring the amount of fissile material and water content. MIK2.0-MVP code, which can calculate fission reaction rate attributed to both prompt and delayed neutrons and also can take the movement of fuel debris into calculation, was developed. After parallelizing the tally process of C() function, MIK2.0-MVP code will be applicable to weakly coupled reactors which include moving fuel debris particles if a supercomputer will be used for the tally process of C() function and if the coupling of MIK2.0-MVP code with MPS will be weak.

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.

Detailed computational models for nuclear criticality analyses on the first startup cores of NSRR: A TRIGA annular core pulse reactor

Yanagisawa, Hiroshi; Motome, Yuiko

JAEA-Research 2025-001, 99 Pages, 2025/06

JAEA-Research-2025-001.pdf:1.98MB

The detailed computational models for nuclear criticality analyses on the first startup cores of NSRR (Nuclear Safety Research Reactor), which is categorized as a TRIGA-ACPR (Annular Core Pulse Reactor), were created for the purposes of deeper understandings of safety inspection data on the neutron absorber rod worths of reactivity and improvement of determination technique of the reactivity worths. The uncertainties in effective neutron multiplication factor (k) propagated from errors in the geometry, material, and operation data for the present models were evaluated in detail by using the MVP version 3 code with the latest Japanese nuclear data library, JENDL-5, and the previous versions of JENDL libraries. As a result, the overall uncertainties in k for the present models were evaluated to be in the range of 0.0027 to 0.0029 k. It is expected that the present models will be utilized as the benchmark on k for TRIGA-ACPR. Moreover, it is confirmed that the overall uncertainties were sufficiently smaller than the values of absorber rod worths determined in NSRR. Thus, it is also considered that the present models are applicable to further analyses on the absorber rod worths in NSRR.

Scalar spin chirality Nernst effect

Go, G.*; Goli, D. P.*; Esaki, Nanse; Tserkovnyak, Y.*; Kim, S. K.*

Physical Review Research (Internet), 7(2), p.L022066_1 - L022066_7, 2025/06

https://doi.org/10.1103/y3m6-25rw

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

https://doi.org/10.1016/j.clay.2025.107862

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.

Crystal structures and phase stability of antiferroelectric phases in undoped and Ca-modified sodium niobates

Aso, Seiyu*; Matsuo, Hiroki*; Yoneda, Yasuhiro; Morikawa, Daisuke*; Tsuda, Kenji*; Oyama, Kenji*; Ishigaki, Toru*; Noguchi, Yuji*

Physical Review B, 111(17), p.174114_1 - 174114_12, 2025/05

https://doi.org/10.1103/PhysRevB.111.174114

We investigate the crystal structures, phase transitions, and phase stability of undoped and Ca-modified NaNbO through a combined analysis of high-resolution synchrotron radiation X-ray and neutron diffraction, convergent-beam electron diffraction, and density functional theory (DFT) calculations. It is demonstrated that the antiferroelectric (AFE)- phase is stabilized over a wide temperature range of 200 to 800 K by Ca modification, and that the NaNbO is stabilized by temperature-driven isostatic pressure accompanied by lattice expansion, whereas the Ca-modified NaNbO is induced by composition-induced chemical pressure along with lattice shrinkage.