Estimating the corrosion rate of stainless steel R-SUS304ULC in nitric acid media under concentrating operation

Irisawa, Eriko; Kato, Chiaki

Journal of Nuclear Materials, 591, p.154914_1 - 154914_10, 2024/04

https://doi.org/10.1016/j.jnucmat.2024.154914

The amount of corrosion of austenitic stainless-steel R-SUS304ULC was evaluated considering the changes in solution composition and boiling during actual concentration operations. Austenitic stainless-steel R-SUS304ULC is the structural material of the highly radioactive liquid waste concentrator in Japanese spent fuel reprocessing plant, which treats highly corrosive nitric acid solutions during enrichment operations. The study results show that it is necessary to focus on nitric acid concentrations, oxidizing metal ion concentrations, and decompression boiling as factors that accelerate the corrosion rate of stainless steel because of cathodic reaction activation.

Estimation method for residual sodium amount on unloaded dummy fuel assembly

Kawaguchi, Munemichi; Hirakawa, Yasushi; Sugita, Yusuke; Yamaguchi, Yutaka

Nuclear Technology, 210(1), p.55 - 71, 2024/01

https://doi.org/10.1080/00295450.2023.2214261

This study has developed an estimation method for residual sodium film and sodium lumps on dummy fuel pins in Monju and demonstrated sodium draining behavior through gaps among the pins, experimentally. The amounts of the residual sodium on the surface of the pins were measured using the three-type test specimens: (a) single pin, (b) 7-pin assembly, and (c) 169-pin assembly. The experiments revealed that the withdrawal speed of the pins and improvement of the sodium wetting increased drastically the amounts of the residual sodium. Furthermore, the experiments using the 169-pin assembly measured the practical amounts of the residual sodium in the dummy fuel assembly of short length and demonstrated sodium draining behavior through the dummy fuel assembly. The estimation method includes four models: a viscosity flow model, Landau-Levich-Derjaguin (LLD) model, an empirical equation related to the Bretherton model, and a capillary force model in a tube. The calculation predicted comparable amounts of the residual sodium with the experiments. An uncertain of the sodium wetting effects were close to 1.8 times the estimation values of the LLD model. With this estimation method, the amounts of the residual sodium on the unloaded Monju dummy fuel assembly can be evaluated.

Criticality safety evaluation of high active liquid waste during the evaporation to dryness process at Tokai Reprocessing Plant

Miura, Takatomo; Kudo, Atsunari; Koyama, Daisuke; Obu, Tomoyuki; Samoto, Hirotaka

Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 10 Pages, 2023/10

Tokai Reprocessing Plant (TRP) had reprocessed 1,140 tons of spent fuel discharged from commercial reactors (BWR, PWR) and Advanced Thermal Reactor "Fugen" from 1977 to 2007. TRP had entered decommissioning stage in 2018. In order to reduce the risk of High Active Liquid Waste (HALW) held at the facility, the vitrification of HALW is given top priority. HALW generated from reprocessing of spent fuel contains not only fission products (FPs) but also trace amounts of uranium (U) and plutonium (Pu) within the liquid and insoluble residues (sludge). Under normal conditions, concentrations of U and Pu in HALW are very low so that it can not reach criticality. Since FPs with high neutron absorption effect coexists in HALW, even if the cooling function is lost due to serious accident and HALW evaporates to dryness, it is considered that criticality would not been reached. In order to confirm this estimation quantitatively, criticality safety evaluations were carried out for the increase of U and Pu concentrations by evaporation of HALW to the point of dryness. In this evaluation, infinite multiplication factors were calculated for each of solution system and sludge system of HALW with respect to the concentration change through evaporation to dryness. It is confirmed it could not reach criticality. The abundance ratios of U, Pu and FPs were set conservatively based on analytical data and ORIGEN calculation results. Multiplation factors for two-layer infinite slab model of solution and sludge systems of HALW were also calculated, and it was confirmed it could not reached criticality. In conclusion, the result was gaind that there could be no criticality even in the process through evaporation to dryness of HALW in TRP.

Development of stable solidification technique of ALPS sediment wastes by apatite ceramics (Contract research); FY2021 Nuclear Energy Science & Technology and Human Resource Development Project

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

JAEA-Review 2022-076, 227 Pages, 2023/03

JAEA-Review-2022-076.pdf:9.42MB

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 FY2021. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station (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 FY2019, this report summarizes the research results of the "Development of stable solidification technique of ALPS sediment wastes by apatite ceramics" conducted from FY2019 to FY2021. Since the final year of this proposal was FY2021, the results for three fiscal years were summarized. The present study aims to establish an apatite solidification process of radioactive sediment wastes, which were generated from the ALPS process manipulating the large amount of contaminated water from 1F. We selected the precipitation method and post stabilization for engineering-scale process. Investigation on composition, structure and elution properties of apatite and related phosphate waste forms fabricated from the simulated ALPS sediment wastes were implemented.

Nuclear data processing code FRENDY version 2

Tada, Kenichi; Yamamoto, Akio*; Kunieda, Satoshi; Nagaya, Yasunobu

JAEA-Data/Code 2022-009, 208 Pages, 2023/02

JAEA-Data-Code-2022-009.pdf:3.87MB

The nuclear data processing code has an important role to connect evaluated nuclear data libraries and neutronics calculation codes. Japan Atomic Energy Agency (JAEA) has developed the nuclear data processing code FRENDY since 2013 to generate cross section files from evaluated nuclear data libraries, such as JENDL, ENDF/B, JEFF, and TENDL. The first version of FRENDY was released in 2019. FRENDY version 1 generates ACE files which are used for continuous energy Monte Carlo codes such as PHITS, Serpent, and MCNP. FRENDY version 2 generates multi-group neutron cross-section files from ACE files. The other major improvements are as follows: (1) uncertainty quantification for the probability tables of the unresolved resonance cross-section; (2) perturbation of the ACE file for the uncertainty quantification using a continuous Monte Carlo code; (3) modification of the ENDF-6 formatted nuclear data file. This report describes an overview of the nuclear data processing methods and input instructions for FRENDY.

Liquid phase sintering of alumina-silica co-doped cerium dioxide CeO ceramics

Vauchy, R.; Hirooka, Shun; Watanabe, Masashi; Yokoyama, Keisuke; Sunaoshi, Takeo*; Yamada, Tadahisa*; Nakamichi, Shinya; Murakami, Tatsutoshi

Ceramics International, 49(2), p.3058 - 3065, 2023/01

https://doi.org/10.1016/j.ceramint.2022.09.295

Present status of JAEA's R&D toward HTGR deployment

Shibata, Taiju; Nishihara, Tetsuo; Kubo, Shinji; Sato, Hiroyuki; Sakaba, Nariaki; Kunitomi, Kazuhiko

Nuclear Engineering and Design, 398, p.111964_1 - 111964_4, 2022/11

https://doi.org/10.1016/j.nucengdes.2022.111964

Japan Atomic Energy Agency (JAEA) has been promoting the research and development (R&D) of High Temperature Gas-cooled Reactor (HTGR). R&D on reactor technologies is carried out by using High Temperature engineering Test Reactor (HTTR). The HTTR was resumed without significant reinforcements in 2021. On January 2022, a safety demonstration test under the OECD/NEA LOFC project was carried out. JAEA is promoting R&D on a carbon-free hydrogen production by thermochemical water splitting Iodine-Sulfur process (IS process). JAEA conducts design study for various HTGR systems toward commercialization. A new test program about demonstration of hydrogen production by the HTTR was launched. Steam methane reforming hydrogen production system was selected for the first demonstration by 2030.

Analysis of microparticles generated by laser processing and development of a methodology for their nuclear identification (Contract research); FY2020 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; The University of Tokyo*

JAEA-Review 2022-015, 119 Pages, 2022/09

JAEA-Review-2022-015.pdf:6.62MB

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 FY2020. 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 FY2018, this report summarizes the research results of the "Analysis of microparticles generated by laser processing and development of a methodology for their nuclear identification" conducted from FY2018 to FY2021 (this contract was extended to FY2021). Since the final year of this proposal was FY2021, the results for four fiscal years were summarized. Although laser processing has various advantages, one well-known disadvantage is that it generates a large amount of microparticles during the processing. Therefore, the application of laser processing to decommissioning waste contaminated with radioactive materials has been hesitant because the mechanism generating the microparticles has not been fully understood.

Phase-field mobility for crystal growth rates in undercooled silicates, SiO and GeO liquids

Kawaguchi, Munemichi; Uno, Masayoshi*

Journal of Crystal Growth, 585, p.126590_1 - 126590_7, 2022/05

https://doi.org/10.1016/j.jcrysgro.2022.126590

Phase-field mobility, , and crystal growth rates in crystallization of 11 oxides or mixed oxides in undercooled silicates, SiO and GeO liquids were calculated with a simple phase-field model (PFM), and material dependence of the was discussed. Ratios between experimental crystal growth rates and the PFM simulation with were confirmed to be proportional to a power of on the solid/liquid interface process during the crystal growth in a log-log plot. We determined that parameters, and , of the were to mJs and to , which were unique for the materials. It was confirmed that our PFM simulation with the determined reproduced quantitively the experimental crystal growth rates. The has a proportional relationship with the diffusion coefficient of a cation molar mass average per unit an oxygen molar mass at in a log-log graph. The depends on the sum of the cation molar mass per the oxygen molar mass, , in a compound. In , the decreases with the cation molar mass increasing. The assumed cause is that the B represents the degree of the temperature dependence of the . Since the cation molar mass is proportional to an inertial resistance of the cation transfer, the decreases with inverse of the cation molar mass. In crystallization of the silicates of heavy cation in , the saturates at approximately 0.67, which leads to .

Gas entrainment phenomenon from free liquid surface in a sodium-cooled fast reactor; Measurements and evaluation on a gas core growth form the liquid surface

Uchida, Mao*; Alzahrani, H.*; Shiono, Mikihito*; Sakai, Takaaki*; Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03

Gas entrainment from cover gas is one of key issues for sodium-cooled fast reactors design to prevent unexpected effects to core reactivity. A vortex model based evaluation method has been developed to evaluate the surface vortex gas core growth at the free surface in the reactor vessel. In this study, water experiments were performed to clarify the prediction accuracy for the vortex gas core growth during the vortex drift motion using a circulating water tunnel with an open flow channel test section. Gas core growth were predicted by applying the evaluation method to the numerical analyses performed in the same geometry of the experiments, and compared with the experimental results. It was observed the gas core growth became large at downstream region where downward velocity became large in experiment. However, the gas core length which were predicted from numerical result showed a discrepancy with the experimental result on the peak position and an overestimation of peak value.