Thermal conductivity evaluation of Am-doped oxide fuels

Yokoyama, Keisuke; Watanabe, Masashi; Onishi, Takashi; Yano, Yasuhide; Tokoro, Daishiro*; Sugata, Hiromasa*; Kato, Masato*

JAEA-Research 2025-002, 18 Pages, 2025/05

JAEA-Research-2025-002.pdf:1.73MB

It is advocated as a development target of fast reactors (FRs) to allow for the of use of mixed oxide (MOX) fuels containing minor actinide (MA) separated and recovered from spent fuels with the aim of reducing the volume and toxicity of high-level radioactive waste generated from nuclear reactors. In the development of MAMOX fuels, it is important behavior to understand the thermal properties such as thermal conductivity for fuel design and analysis of the irradiation. However, there are only a few reports on the thermal properties of MA-MOX fuels, and neither the effects of MA contents nor of oxygen non-stoichiometry in MOX fuels on their thermal conductivities have been fully understood. In this study, the thermal conductivities of MOX fuels with up to 15% Am content were measured at near-stoichiometric composition and the relationship between thermal conductivity and Am content was evaluated. Moreover, the thermal conductivities of Am-doped UO fuels were also measured and evaluated by comparison with Am-MOX to evaluate the effect of Am content. The fuel samples used in this study were three types of MOX with a Pu content of 30% and different Am contents (5%, 10%, and 15%), and UO containing 15% Am. The thermal conductivities of specimens were calculated from the thermal diffusivities measured by the laser flash method, the density of the specimens and, the heat capacity at constant pressure. The oxygen partial pressure during the measurement was controlled at that of the targeted near-stoichiometric composition. The thermal conductivities of all specimens exhibited a decline with increasing temperature and Am content, with a particularly pronounced reduction observed below 1,173 K. The results of the classical phonon scattering model analysis of the measured thermal conductivities showed that the effect of lattice strain due to the Am addition was significant on the thermal resistivity change, and the effect was comparable for both MOX and UO.

Microstructural observation of simulated fuel kernels for Pu-burner high temperature gas-cooled reactor in Japan

Aihara, Jun; Ueta, Shohei; Honda, Masaki*; Kasahara, Seiji; Okamoto, Koji*

JAEA-Research 2024-012, 98 Pages, 2025/02

JAEA-Research-2024-012.pdf:32.24MB

Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for the purpose of more safely reducing amount of recovered Pu. In Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO (PuO-YSZ) small particle and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. ZrC layer is oxygen getter. In research project of Pu-burner HTGR carried out from fiscal year of 2014 to fiscal year of 2017, simulated CFPs were fabricated using Ce to simulate Pu. Moreover, simulated fuel compacts were fabricated using fabricated simulated CFPs. In this report, results of microstructural observation of CeO-YSZ and ZrC layer at each fabrication step are reported.

Development of failure mitigation technologies for improving resilience of nuclear structures

Kasahara, Naoto*; Yamano, Hidemasa; Nakamura, Izumi*; Demachi, Kazuyuki*; Sato, Takuya*; Ichimiya, Masakazu*

International Journal of Pressure Vessels and Piping, 211, p.105298_1 - 105298_6, 2024/10

https://doi.org/10.1016/j.ijpvp.2024.105298

Development of safety design technologies for sodium-cooled fast reactor coupled to thermal energy storage system with sodium-molten salt heat exchanger, 2; Study of sodium-molten salt heat exchanger heat transfer performance

Hayashi, Masaaki*; Nakahara, Hirotaka*; Shirakura, Shota*; Yamano, Hidemasa

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

As part of the development of risk assessment technologies for sodium-cooled fast reactor coupled to thermal energy storage (TES) system with sodium-molten salt heat exchanger (HX), simple evaluation of heat transfer performance using heat transfer coefficient formula is performed. And Computational Fluid Dynamics (CFD) thermal analyses by STAR-CCM+ with partial HX model are performed to develop evaluation technology. The performance evaluation technology of a HX between sodium and molten salt and the confirmation of heat transfer improvement measures effects is developed.

Development of safety design technologies for sodium-cooled fast reactor coupled to thermal energy storage system with sodium-molten salt heat exchanger, 1; Overview and Consideration on tube failure

Yamano, Hidemasa; Takano, Kazuya; Kurisaka, Kenichi; Kikuchi, Shin; Kondo, Toshiki; Umeda, Ryota; Sato, Rika; Shirakura, Shota*

Dai-28-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2024/06

This project studies investigation on safety design guideline and risk assessment technology for sodium-cooled fast reactor with the molten-salt heat storage system, development of evaluation method for heat transferring performance between sodium and molten-salt and improvement of the performance, and evaluation of chemical reaction characteristic between sodium and molten-salt and improvement of its safety. This paper describes the effect of sodium-molten salt heat transfer tube failure in addition to the project overview and progress.

Formulation of high-temperature strength equation of 9Cr-ODS tempered martensitic steels using the Larson-Miller parameter and life-fraction rule for rupture life assessment in steady-state, transient, and accident conditions of fast reactor fuel

Miyazawa, Takeshi; Tanno, Takashi; Imagawa, Yuya; Hashidate, Ryuta; Yano, Yasuhide; Kaito, Takeji; Otsuka, Satoshi; Mitsuhara, Masatoshi*; Toyama, Takeshi*; Onuma, Masato*; et al.

Journal of Nuclear Materials, 593, p.155008_1 - 155008_16, 2024/05

https://doi.org/10.1016/j.jnucmat.2024.155008

Evaluation of thermal neutron scattering law of nuclear-grade isotropic graphite

Nakayama, Shinsuke; Iwamoto, Osamu; Kimura, Atsushi

EPJ Web of Conferences, 294, p.07001_1 - 07001_6, 2024/04

https://doi.org/10.1051/epjconf/202429407001

Graphite is a candidate of moderator in innovative nuclear reactors such as molten salt reactors. Scattering of thermal neutrons by the moderator material has a significant impact on the reactor core design. To contribute to the development of innovative nuclear reactors, an evaluation method of thermal neutron scattering law for reactor grade graphite was studied. The inelastic scattering component due to lattice vibration was evaluated based on the phonon density of states computed with first-principles lattice dynamics simulations. The simulations were performed for ideal crystalline graphite. The coherent elastic scattering component due to crystal structure was evaluated based on neutron transmission and scattering experiments recently performed in the J-PARC/MLF facility. In comparison with the neutron transmission experiments, it was found that the quantification of small-angle neutron scattering due to structures larger than crystal, such as pores in graphite, is important. Based on the above methods, thermal neutron scattering law data for reactor-grade graphite at room temperature were evaluated.

Development of failure mitigation technologies for improving resilience of nuclear structures, 6; Resilience improvements of fast reactors by failure mitigation for excessive earthquake

Yamano, Hidemasa; Futagami, Satoshi; Ando, Masanori; Kurisaka, Kenichi

Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 11 Pages, 2024/03

In this study, the dynamic structural analysis of the reactor vessel for excessive earthquake using the FINAS/STAR code has shown the elephant foot buckling deformation and calculated the cumulative fatigue failure fraction. Using the calculation results, this paper describes the fragility curve using the safety factor method, indicating the significantly improved curve compared the previous one.

Thermal conductivity measurement of uranium-plutonium mixed oxide doped with Nd/Sm as simulated fission products

Horii, Yuta; Hirooka, Shun; Uno, Hiroki*; Ogasawara, Masahiro*; Tamura, Tetsuya*; Yamada, Tadahisa*; Furusawa, Naoya*; Murakami, Tatsutoshi; Kato, Masato

Journal of Nuclear Materials, 588, p.154799_1 - 154799_20, 2024/01

https://doi.org/10.1016/j.jnucmat.2023.154799

The thermal conductivities of near-stoichiometric (U,Pu,Am)O doped with NdO/SmO, which is major fission product (FP) generated by a uranium-plutonium mixed oxides (MOX) fuel irradiation, as simulated fission products are evaluated at 1073-1673 K. The thermal conductivities are calculated from the thermal diffusivities that are measured using the laser flash method. To evaluate the thermal conductivity from a homogeneity viewpoint of Nd/Sm cations in MOX, the specimens with different homogeneity of Nd/Sm are prepared using two kinds of powder made by ball-mill and fusion methods. A homogeneous Nd/Sm distribution decreases the thermal conductivity of MOX with increasing Nd/Sm content, whereas heterogeneous Nd/Sm has no influence. The effect of Nd/Sm on the thermal conductivity is studied using the classical phonon transport model (A+BT). The dependences of the coefficients A and B on the Nd/Sm content (C and C, respectively) are evaluated as: A(mK/W)=1.70 10 + 0.93C + 1.20C, B(m/W)=2.39 10.

Development of probabilistic risk assessment methodology using artificial intelligence technology, 2; Automatic fault detection method for building reliability database

Ujita, Hiroshi*; Morimoto, Tatsuya*; Futagami, Satoshi; Yamano, Hidemasa; Kurisaka, Kenichi

Proceedings of PSAM 2023 Topical Conference AI & Risk Analysis for Probabilistic Safety/Security Assessment & Management, 8 Pages, 2023/10

This study is intended to develop PRA methodology using the AI technology. For this purpose, as a first step, the authors have been conducting a three-year program including the development of AI tools for automatic fault tree (FT) creation and automatic fault detection methodology for building reliability database. These AI tools are intended to enable any users to easily perform PRA with the same quality without user effect. For the automatic fault detection method, The AI tool is developed for extracting failure occurrence locations (system/equipment), failure modes, and causes from Japanese reliability databases of NUCIA(for light water reactors) and CORDS(for sodium-cooled fast reactors), and transforming them into a database using AI technologies.