Locally mesh-refined lattice Boltzmann method for fuel debris air cooling analysis on GPU supercomputer

Onodera, Naoyuki; Idomura, Yasuhiro; Uesawa, Shinichiro; Yamashita, Susumu; Yoshida, Hiroyuki

Mechanical Engineering Journal (Internet), 7(3), p.19-00531_1 - 19-00531_10, 2020/06

https://doi.org/10.1299/mej.19-00531

A dry method is one of practical methods for decommissioning the TEPCO's Fukushima Daiichi Nuclear Power Station. Japan Atomic Energy Agency (JAEA) has been evaluating the air cooling performance of the fuel debris by using the JUPITER code based on an incompressible fluid model and the CityLBM code based on the lattice Boltzmann method (LBM). However, these codes were based on a uniform Cartesian grid system, and required large computational time and cost to capture complicated debris structures. We develop an adaptive mesh refinement (AMR) version of the CityLBM code on GPU based supercomputers and apply it to thermal-hydrodynamics problems. The proposed method is validated against free convective heat transfer experiments at JAEA. It is also shown that the AMR based CityLBM code on 4 NVIDIA TESLA V100GPUs gives 6.7x speedup of the time to solution compared with the JUPITER code on 36 Intel Xeon E5-2680v3 CPUs.

Ice I without stacking disorder by evacuating hydrogen from hydrogen hydrate

Komatsu, Kazuki*; Machida, Shinichi*; Noritake, Fumiya*; Hattori, Takanori; Sano, Asami; Yamane, Ryo*; Yamashita, Keishiro*; Kagi, Hiroyuki*

Nature Communications (Internet), 11(1), p.464_1 - 464_5, 2020/02

https://doi.org/10.1038/s41467-020-14346-5

Water freezes below 0C at ambient pressure ordinarily to ice I, with hexagonal stacking sequence. Under certain conditions, ice with a cubic stacking sequence can also be formed, but ideal ice I without stacking-disorder has never been formed until recently. Here we demonstrate a route to obtain ice I without stacking-disorder by degassing hydrogen from the high-pressure form of hydrogen hydrate, C, which has a host framework isostructural with ice I. The stacking-disorder free ice I is formed from C via an intermediate amorphous or nano-crystalline form under decompression, unlike the direct transformations occurring in ice XVI from neon hydrate, or ice XVII from hydrogen hydrate. The obtained ice I shows remarkable thermal stability, until the phase transition to ice I at 250 K, originating from the lack of dislocations. This discovery of ideal ice I will promote understanding of the role of stacking-disorder on the physical properties of ice as a counter end-member of ice I.

Oxidation of silicon carbide in steam studied by laser heating

Pham, V. H.; Nagae, Yuji; Kurata, Masaki; Furumoto, Kenichiro*; Sato, Hisaki*; Ishibashi, Ryo*; Yamashita, Shinichiro

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.670 - 674, 2019/09

Fuel debris' air cooling analysis using a lattice Boltzmann method

Onodera, Naoyuki; Idomura, Yasuhiro; Kawamura, Takuma; Uesawa, Shinichiro; Yamashita, Susumu; Yoshida, Hiroyuki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 6 Pages, 2019/05

A dry method is one of practical methods for decommissioning the TEPCO's Fukushima Daiichi Nuclear Power Station. Japan Atomic Energy Agency (JAEA) has been evaluating the air cooling performance by using the JUPITER code. However, the JUPITER code requires a large computational cost to capture debris' structures. To accelerate such CFD analyses, we use the CityLBM code, which is based on the lattice Boltzmann method (LBM) and is highly optimized for GPUs. The CityLBM code is validated against free convective heat transfer experiments at JAEA, and the similar accuracy as the JUPITER code is confirmed regarding the prediction capability of heat transfer and the resulting temperature distributions. It is also shown that the elapse time of a CityLBM simulation on GPUs is reduced to 1/6 compared with that of the corresponding JUPITER simulation on CPUs with the same number of GPUs and CPUs. The results show that the LBM is promising for accelerating thermal convective simulations.

Corrosion behaviour of FeCrAl-ODS steels in nitric acid solutions with several temperatures

Takahatake, Yoko; Ambai, Hiromu; Sano, Yuichi; Takeuchi, Masayuki; Koizumi, Kenji; Sakamoto, Kan*; Yamashita, Shinichiro

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 9 Pages, 2018/10

https://euronuclear.org/events/topfuel/topfuel2018/fullpapers/TopFuel2018-A0080-fullpaper.pdf

The corrosion behaviour of FeCrAl-ODS steels for the accident tolerant fuel cladding of LWRs were investigated in nitric acid solutions for the reprocessing process of spent fuels. The corrosion tests were carried out at 60C, 80C and the boiling point of the solutions, and the specimens were then analysed by XPS. The corrosion remarkably progressed at the boiling point, and the highest corrosion rate was 0.22 mm/y. In the oxide film, the atomic concentration of Fe was lower, than that in the base material, and those of Cr and Al were higher. The results show that the corrosion of FeCrAl-ODS steels in hot nitric acid solution is not severe because of the high corrosion resistance of the oxide film formed on the material; hence, the corrosion resistance of the new cladding materials in the dissolution process of spent fuel is acceptable for reprocessing operations.

Ion irradiation effects on FeCrAl-ODS ferritic steel

Kondo, Keietsu; Aoki, So; Yamashita, Shinichiro; Ukai, Shigeharu*; Sakamoto, Kan*; Hirai, Mutsumi*; Kimura, Akihiko*

Nuclear Materials and Energy (Internet), 15, p.13 - 16, 2018/05

https://doi.org/10.1016/j.nme.2018.05.022

Radiation hardening and microstructural evolution of ion irradiated 12Cr-6Al ODS ferritic steel was studied. Ion irradiation experiments were performed using Fe ions up to the nominal displacement damage of 20 dpa at the irradiation temperature was 300C. The monotonical increase of radiation hardening with dose was confirmed by experimentally obtained hardness data. The radiation hardening was also calculated theoretically by introducing the microstructural character examined by TEM into the dispersed barrier hardening model. The results showed a good agreement with the experimentally obtained data up to 5 dpa, while a slight discrepancy was found between theoretical and experimental hardness values at 20 dpa. Radiation hardening was mainly caused by irradiation-induced defect clusters below the irradiation dose of 5 dpa. As the irradiation dose increased toward 20 dpa, an additional influence of the radiation appeared, which was assumed to be induced by ' phase transformation.

Fuel behavior analysis for accident tolerant fuel with sic cladding using adapted FEMAXI-7 code

Shirasu, Noriko; Saito, Hiroaki; Yamashita, Shinichiro; Nagase, Fumihisa

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 8 Pages, 2017/09

http://wrfpm2017.org/upload/draft_file/F-226-PD5.pdf

Silicon carbide (SiC) is an attractive candidate of accident tolerant fuel (ATF) cladding material because of its high chemical stability, high radiation resistance and low neutron absorption. FEMAXI-ATF has been developed to analysis SiC cladding fuel behaviors. The thermal, mechanical and irradiation property models were implemented to FEMAXI-7, which is a fuel behavior analysis code being developed in JAEA. Fuel rod behavior analysis was performed under typical boiling water reactor (BWR) operating conditions with a model based on a 99 BWR fuel (Step III Type B), in which the cladding material was replaced from Zircaloy to SiC. The SiC cladding shows large swelling by irradiation. It increases the gap size and decreases cladding thermal conductivity. The mechanism of relaxation of stress is also different from the Zircaloy cladding. The experimental data for SiC materials are still insufficient to construct the models, especially for evaluating fracture behavior.

Analytical study of the applicability of FeCrAl-ODS cladding for BWR

Takano, Sho*; Kusagaya, Kazuyuki*; Goto, Daisuke*; Sakamoto, Kan*; Yamashita, Shinichiro

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

http://www.wrfpm2017.org/upload/draft_file/F-032-PD1.pdf

We focused on one of accident tolerant fuel (ATF) materials, Oxide Dispersion Strengthened Fe-Cr-Al Steel (FeCrAl-ODS). There is a reasonable prospect that FeCrAl-ODS is applied to BWRs, but relatively high neutron absorption should be compensated. To decrease adverse neutron economic impact, thin FeCrAl-ODS cladding was designed, and we evaluated characteristics of a core into which 99 Advanced BWR (ABWR) bundles with thin FeCrAl-ODS claddings were loaded. Thin FeCrAl-ODS water rods and channel boxes were also applied. We confirmed that FeCrAl-ODS core reactivity was sufficient by increasing enrichment of UO fuel under the limit of 5 wt%. Moreover, some representative FeCrAl-ODS core characteristics were comparable to zircaloy core. We also confirmed that fuel thermal-mechanical behaviors of thin FeCrAl-ODS cladding at normal operation and transient conditions were acceptable. These results led to a conclusion that FeCrAl-ODS was applicable to BWR in the analysis range of this study.

Welding technology R&D of Japanese accident tolerant fuel claddings of FeCrAl-ODS steel for BWRS

Kimura, Akihiko*; Yuzawa, Sho*; Sakamoto, Kan*; Hirai, Mutsumi*; Kusagaya, Kazuyuki*; Yamashita, Shinichiro

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

http://www.wrfpm2017.org/upload/draft_file/F-059-PD1.pdf

The effect of Al addition on the PRW weldability of ODS steel is shown with the discussion focusing on the microstructure changes by the welding. The ordinary welding methods including electron beam (EB) welding and tungsten inert gas (TIG) welding were also applied to the SUS430 endcap welding to cladding tube made of FeCrAl-ODS steel. The endcap welded ODS steel tube samples were tensile tested at RT. The EB welded FeCrAl-ODS/SUS430 samples broke in the ODS steel tube, indicating that the weld bond is stronger than the ODS base metal. However, the TIG welded FeCrAl-ODS/SUS430 samples broke at a weld bond. X-ray CT scan analysis was performed for the weld bond, and the bonding strength was correlated with the X-ray CT results in order to assess the feasibility of those welding methods for ATF-ODS steel cladding.

Overview of Japanese development of accident tolerant FeCrAl-ODS fuel claddings for BWRs

Sakamoto, Kan*; Hirai, Mutsumi*; Ukai, Shigeharu*; Kimura, Akihiko*; Yamaji, Akifumi*; Kusagaya, Kazuyuki*; Kondo, Takao*; Yamashita, Shinichiro

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 7 Pages, 2017/09

http://www.wrfpm2017.org/upload/draft_file/D_033-PD1.pdf

This paper will show the overview of current status of development of accident tolerant FeCrAl-ODS fuel claddings for BWRs (boiling water reactors) in the program sponsored and organized by the Ministry of Economy, Trade and Industry (METI) of Japan. This program is being carried out to create the technical basis for the practical use of the accident tolerant fuels and the other components in LWRs through multifaceted activities. In the development of FeCrAl-ODS fuel claddings both the experimental and the analytical studies have been performed. The acquisition and accumulation of key material properties of FeCrAl-ODS fuel claddings were conducted by using bar, sheet and tube shaped FeCrAl-ODS materials fabricated in this program to support the evaluations in the analytical studies. A neutron irradiation test was also started in the ORNL High Flux Isotope Reactor (HFIR) to examine the effect of neutron irradiation on the mechanical properties.