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Journal Articles

Numerical analysis of natural convective heat transfer with porous medium using JUPITER

Uesawa, Shinichiro; Yamashita, Susumu; Sano, Yoshihiko*; Yoshida, Hiroyuki

Journal of Nuclear Science and Technology, 62(6), p.523 - 541, 2025/06

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Japan Atomic Energy Agency (JAEA) has developed a numerical method with the JUPITER code with a porous medium model to calculate the thermal behavior in PCVs of 1F. In this study, we performed an experiment and numerical simulation of the natural convective heat transfer with the porous medium to validate JUPITER with the porous medium model. In comparison of the temperature and velocity distributions between the experiment and simulation, the temperature distribution in the simulation was in good agreement with the distribution in the experiment except the temperature near the top surface of the porous medium. The velocity distribution also agreed qualitatively with the experimental result. In addition, we also performed the numerical simulations with various effective thermal conductivity models to discuss the effect of the conductivity based on the internal structure of porous media on the natural convective heat transfer. The result indicated that the temperature distribution in the porous medium and the velocity distribution of the natural convection were significantly different for each model, and thus the conductivity of the fuel debris was one of the key parameters of in the thermal behavior analysis in 1F.

Journal Articles

Numerical investigation of the accuracy of a conductance-type wire-mesh sensor for a single spherical bubble and bubbly flow

Uesawa, Shinichiro; Ono, Ayako; Nagatake, Taku; Yamashita, Susumu; Yoshida, Hiroyuki

Journal of Nuclear Science and Technology, 62(5), p.432 - 456, 2025/05

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

We performed electrostatic simulations of a wire-mesh sensor (WMS) for a single spherical bubble and bubbly flow to clarify the accuracy of the WMS. The electrostatic simulation for the single bubble showed the electric current density distribution and the electric current path from the excited transmitter to receivers for various bubble locations. It indicated systematic errors based on the nonuniform current density distribution around the WMS. The electrostatic simulation for the bubbly flow calculated by the computational fluid dynamics code, JAEA Utility Program for Interdisciplinary Thermal-hydraulics Engineering and Research (JUPITER), indicated that the WMS had difficulty in quantitatively measuring the intermediate values of the instantaneous void fraction between 0 and 1 because they cannot be estimated by previous transformation methods from the WMS signal to the void fraction, such as linear approximation or Maxwell's equation, and have a significant deviation of the void fraction of $$pm$$0.2 for the WMS signal. However, the electrostatic simulation indicated that the time-averaged void fractions around the center of the flow channel can be estimated using linear approximation, and the time-averaged void fraction near the wall of the flow channel can be estimated using Maxwell's equation.

Journal Articles

Development of advanced AI-based segmentation and prediction method for air entrainment in plunging water jets

Zhou, W.*; Miwa, Shuichiro*; Yamashita, Susumu; Okamoto, Koji*

Progress in Nuclear Energy, 177, p.105441_1 - 105441_17, 2024/12

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Understanding air entrainment phenomena induced by plunging water jets is critical in the fields of nuclear and hydraulic engineering. Air entrainment is one of the key safety design parameters for nuclear systems. However, most existing studies rely on empirical correlations or curve-fitting models to estimate bubble penetration depth, and no agreed-upon calculation principle exists for varying jet conditions. To address these limitations, this research developed two advanced AI approaches: an improved YOLOv5 for segmenting air entrainment and the NSGA-III-BPNN method for predicting penetration depth. The improved YOLOv5 enables real-time segmentation and extraction of air entrainment motion and dynamics under diverse conditions, demonstrating high scalability and robustness. The penetration depth estimated using the improved YOLOv5 shows greater accuracy compared to conventional empirical correlations and is more efficient than traditional image post-processing techniques for classifying shape regimes based on dynamic air entrainment patterns. To overcome the limitations of object segmentation, which typically relies on video or image data, the NSGA-III-BPNN method predicts maximum penetration depths with greater accuracy than YOLOv5, offering a more effective prediction model for air entrainment penetration depth. By leveraging advanced AI techniques, the research not only provides valuable segmentation data for refining computational fluid dynamics (CFD) modeling but also paves the way for significant advancements in both nuclear and hydraulic engineering.

Journal Articles

Numerical investigation of accuracy of conductance-typed wire-mesh sensor using CFD and electrostatic simulations

Uesawa, Shinichiro; Ono, Ayako; Yamashita, Susumu; Yoshida, Hiroyuki

Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 7 Pages, 2024/11

A conductance-typed wire-mesh sensor (WMS), utilizing the difference in conductivity between gas and liquid phases between the electrodes, is one of the practical measurement techniques of a cross-sectional void fraction distribution in a flow path. In this study, we performed two-phase computational fluid dynamics (CFD) and electrostatic simulations around a WMS for a single spherical bubble and bubbly flow to clarify the systematic error in the WMS. The results for the single bubble indicated that there were systematic errors based on the non-uniform current density distribution around the WMS. The correlation between instantaneous void fractions and WMS signals is not uniquely determined for positions of the single bubble moving across the WMS, even for the same bubble. Moreover, the correlation between the instantaneous void fractions and the WMS signals did not fit in a linear approximation and Maxwell's equation, which traditionally used transformation methods from the WMS signal to the void fraction. The results for the bubbly flow indicated that the WMS had difficulty in quantitative measurements of the instantaneous void fraction because the values had a significant deviation of the void fraction of approximately $$pm$$0.2. On the other hand, time-averaged void fraction values had relatively small deviation. Thus, we concluded that the WMS, using existing transformation methods, can measure time-averaged void fractions, but it is difficult to measure quantitatively instantaneous void fractions.

Journal Articles

Bubble flow analysis using multi-phase field method

Sugihara, Kenta; Onodera, Naoyuki; Sitompul, Y.; Idomura, Yasuhiro; Yamashita, Susumu

EPJ Web of Conferences, 302, p.03002_1 - 03002_10, 2024/10

 Times Cited Count:0 Percentile:0.00(Computer Science, Interdisciplinary Applications)

In simulations of gas-liquid two-phase flows using conventional interface capture methods, we observed that when bubbles come close to each other, they tend to merge numerically, despite experimental evidence indicating that they would repel each other. Given the significant impact of sequential numerical coalescence on flow patterns, it is necessary to regulate the merging behavior of close bubbles. To address this issue, we introduced the Multi- Phase Field (MPF) method, which mitigates bubble coalescence by applying an independent fluid fraction function to each bubble. In this study, we employed the MPF based on the N-phase model to minimize numerical errors associated with surface interactions at triple junction points. Additionally, we implemented the Ordered Active Parameter Tracking (OAPT) method to efficiently store several hundreds of fluid fraction functions. To validate the MPF method, we conducted analysis of turbulent bubbly pipe flows and compared the results against experimental data from Colin et al. The validation results showed reasonable agreements with respect to the bubble distribution and the flow velocity profiles.

Journal Articles

R&D status of digital technology on inverse estimation of radioactive source distributions and related source countermeasures; Fast Digital Twin Tech. in Decommissioning Field: 3D-ADRES-Indoor FrontEnd

Machida, Masahiko; Yamada, Susumu; Kim, M.; Tanaka, Satoshi*; Tobita, Yasuhiro*; Iwata, Ayako*; Aoki, Yuto; Aoki, Kazuhisa; Yanagisawa, Kenichi*; Yamaguchi, Takashi; et al.

RIST News, (70), p.3 - 22, 2024/09

Inside the Fukushima Daiichi Nuclear Power Plant (1F), there are many locations with high radiation levels due to contamination by radioactive materials that leaked from the reactor. These pose a significant obstacle to the smooth progress of decommissioning work. To help solve this issue, the Japan Atomic Energy Agency (JAEA), under a subsidy from the Ministry of Economy, Trade, and Industry's decommissioning and contaminated water management project, is conducting research and development on digital technologies to improve the radiation environment inside the decommissioning site. This project, titled "Development of Technology to Improve the Environment Inside Reactor Buildings (Enhancing Digital Technology for Environment and Source Distribution to Reduce Radiation Exposure)," began in April of FY 2023. In this project, the aim is to develop three interconnected systems: FrontEnd, Pro, and BackEnd. The FrontEnd system, based on the previously developed 3D-ADRES-Indoor (prototype) from FY 2021-2022, will be upgraded to a high-speed digital twin technology usable on-site. The Pro system will carry out detailed analysis in rooms such as the new office building at 1F, while the BackEnd system will serve as a database to centrally manage the collected and analyzed data. This report focuses on the FrontEnd system, which will be used on-site. After point cloud measurement, the system will quickly create a 3D mesh model, estimate the radiation source from dose rate measurements, and refine the position and intensity of the estimated source using recalculation techniques (re-observation instructions and re-estimation). The results of verification tests conducted on Unit 5 are also presented. Furthermore, the report briefly discusses the future research and development plans for this project.

Journal Articles

Evaluation of interface capturing schemes of VOF method through application to bubble flows with single orifice

Fukuda, Takanari; Yamashita, Susumu; Yoshida, Hiroyuki

Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 12 Pages, 2024/08

The VOF method is a type of CFDs that is most widely applied to multiphase flow analysis involving advective interfaces, and several interface-capturing schemes have been developed for an accurate advection of VOF values. However, the performance of these schemes has typically been evaluated only for limited numerical problems where velocity fields are spatially orderly and fixed in time. Few studies have been conducted to evaluate the performance of these schemes for more realistic and complex conditions, such as gas-liquid two-phase flows in nuclear reactors. Therefore, in this study, three-dimensional analysis of bubble flows has been conducted using the interface-capturing schemes of THINC and THINC/WLIC, which have been developed relatively recently. Evaluation is performed using more engineering indicators such as the number, volume, and trajectory of bubbles, which can influence the void fraction distribution in reactor cores. The results of these comparisons showed that the VOF value could be significantly diffused, leading to numerical brake-up and dissipation of the bubbles, with the influence of interface-capturing scheme.

Journal Articles

Development of a simplified boiling model applied for large-scale detailed two-phase flow simulations based on the VOF method

Ono, Ayako; Sakashita, Hiroto*; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki

Mechanical Engineering Journal (Internet), 11(4), p.24-00188_1 - 24-00188_12, 2024/07

Japan Atomic Energy Agency (JAEA) is developing the evaluation method for a two-phase flow in the reactor core using simulation codes based on the Volume Of Fluid (VOF) method. JAEA started developing a Simplified Boiling Model (SBM) for the large-scale two-phase flow in the fuel assemblies. In the SBM, the motion and growth equations of the bubble are solved to obtain their diameter and time length at the detachment, of which size scale is within/around the calculation grid size of the numerical simulation. JUPITER calculates the bubble behavior with a scale of more than several $$mu$$m. In this study, the convection boiling on a vertical heating surface is simulated using the developed SBM. The comparison between the simulation and experimental results showed good reproducibility of the heat flux and velocity dependency on the passage period of the bubble.

Journal Articles

Benchmark simulation code for the thermal-hydraulics design tool of the accelerator-driven system; Validation and benchmark simulation of flow behavior around the beam window

Yamashita, Susumu; Kondo, Nao; Sugawara, Takanori; Monji, Hideaki*; Yoshida, Hiroyuki

Journal of Nuclear Science and Technology, 61(6), p.740 - 761, 2024/06

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

To confirm the validity of the thermal-hydraulics design tool based on the Ansys Fluent, we used a detailed computational fluid dynamics code named JAEA Utility Program for Interdisciplinary Thermal-hydraulics Engineering and Research (JUPITER) for the thermal-hydraulics around the beam window (BW) of the Accelerator-Driven System (ADS). The Fluent uses the Reynolds-Averaged Navier-Stokes (RANS) model and can quickly calculate the turbulent flow around the BW as a BW design tool. At first, we compared the results of JUPITER with the experimental results using a mock-up BW system in water to confirm the validity of JUPITER. As a result, we confirmed that numerical results are in good agreement with the experimental results. Thus, we showed that JUPITER could be used as a benchmark code. We also performed a benchmark simulation for the Fluent calculation using validated JUPITER to show the applicability of JUPITER as an alternative of experiments. As a result, the mean values around the BW agreed with each other, e.g., the mean velocity profile for stream and horizontal directions. Therefore, we confirmed that JUPITER showed a good performance in validating the thermal-hydraulics design tool as a fluid dynamics solver. Moreover, Fluent has enough accuracy as a thermal-hydraulics design tool for the ADS.

JAEA Reports

Utilization of gamma ray irradiation at the WASTEF Facility

Sano, Naruto; Yamashita, Naoki; Watanabe, Masaya; Tsukada, Manabu*; Hoshino, Kazutoyo*; Hirai, Koki; Ikegami, Yuta*; Tashiro, Shinsuke; Yoshida, Ryoichiro; Hatakeyama, Yuichi; et al.

JAEA-Technology 2023-029, 36 Pages, 2024/03

JAEA-Technology-2023-029.pdf:2.47MB

At the Waste Safety Testing Facility (WASTEF), the gamma ray irradiation device "Gamma Cell 220" was relocated from the 4th Research Building of the Nuclear Science Research Institute in FY2019, and the use of gamma ray irradiation has begun. Initially, Fuel Cycle Safety Research Group, Fuel Cycle Safety Research Division, Nuclear Safety Research Center, Sector of Nuclear Safety Research and Emergency Preparedness, the owner of this device, conducted the tests as the main user, but since 2022, other users, including those outside the organization, have started using it. The gamma ray irradiation device "Gamma Cell 220" is manufactured by Nordion International Inc. in Canada. Since it was purchased in 1989, the built-in $$^{60}$$Co radiation source has been updated once, and safety research related to nuclear fuel cycles, etc. It is still used for this purpose to this day. This report summarizes the equipment overview of the gamma ray irradiation device "Gamma Cell 220", its permits and licenses at WASTEF, usage status, maintenance and inspection, and future prospects.

JAEA Reports

Development of an electrochemical measurement method for carbon steels in radiation source dissolved solution and a corroded specimen analysis method using an imaging plate

Yamashita, Naoki; Aoyama, Takahito; Kato, Chiaki; Sano, Naruto; Tagami, Susumu

JAEA-Technology 2023-028, 22 Pages, 2024/03

JAEA-Technology-2023-028.pdf:1.9MB

At the Fukushima Daiichi Nuclear Power Station (1F), which is currently undergoing decommissioning, there is growing interest in the effects of radiation-emitting radionuclides such as $$^{90}$$Sr and $$^{137}$$Cs on the structural integrity. In particular, the corrosion behavior of carbon steel, which is used in many parts of 1F, is known to change depending on metal cations in solution, but the effects of $$^{90}$$Sr and $$^{137}$$Cs on corrosion are not yet understood. In addition, it is important to investigate the distribution of $$^{90}$$Sr and $$^{137}$$Cs in the rust layer in order to understand the corrosion behavior, but the method has not yet been established. In this study, a glove box was prepared to conduct corrosion tests of carbon steel in NaCl containing $$^{90}$$Sr and $$^{137}$$Cs in the glove box. In addition, in order to clarify the influence of $$^{90}$$Sr and $$^{137}$$Cs, which exist as metal cations in the solution, on the corrosion behavior of carbon steel, we attempted to establish a detection method for radioactive materials in the rust layer using an imaging plate.

Journal Articles

Summary report in FY2022 of subsidy program for "the Project of Decommissioning and Contaminated Water Management (Development of Analysis and Estimation Technology for Characterization of Fuel Debris (Development of Technologies for Enhanced Analysis Accuracy, Thermal behavior Estimation, and Simplified Analysis of Fuel Debris)" started in FY2011

Koyama, Shinichi; Ikeuchi, Hirotomo; Mitsugi, Takeshi; Maeda, Koji; Sasaki, Shinji; Onishi, Takashi; Tsai, T.-H.; Takano, Masahide; Fukaya, Hiroyuki; Nakamura, Satoshi; et al.

Hairo, Osensui, Shorisui Taisaku Jigyo Jimukyoku Homu Peji (Internet), 216 Pages, 2023/11

In FY 2021 and 2022, JAEA perfomed the subsidy program for "the Project of Decommissioning and Contaminated Water Management (Development of Analysis and Estimation Technology for Characterization of Fuel Debris (Development of Technologies for Enhanced Analysis Accuracy, Thermal Bahavior Estimation, and Simplified Analysis of Fuel Debris)" started in FY 2021. This presentation material summarized the results of the project, which will be available shortly on the website of Management Office for the Project of Decommissioning, Contaminated Water and Treated Water Management.

JAEA Reports

Optimized phase-field modeling using a modified conservative Allen-Cahn equation for two-phase flows

Sugihara, Kenta; Onodera, Naoyuki; Idomura, Yasuhiro; Yamashita, Susumu

JAEA-Research 2023-006, 47 Pages, 2023/10

JAEA-Research-2023-006.pdf:3.28MB

This report presents a new surface capturing method based on the phase field model for gas-liquid two-phase flows simulation. In the conventional phase field model, the interface correction strength parameter was determined from the maximum flow velocity in the computational domain, but because the interface correction was applied uniformly to the entire space, it was also applied to locations that did not require correction. In the new method, the phase field parameter or the intensity of the phase field model is extended to have a spatial distribution, allowing us to set the optimal parameters depending on the local flow velocity fields. We also propose a method to derive the optimal phase field parameter based on systematic parameter scans using error analysis of the interface advection test and bubble rising calculations. Through benchmark tests of gas-liquid two-phase flows, the proposed model is verified, and it is shown that the proposed model has higher accuracy than the conventional phase field model.

Journal Articles

Numerical simulation method using a Cartesian grid for oxidation of core materials under steam-starved conditions

Yamashita, Susumu; Sato, Takumi; Nagae, Yuji; Kurata, Masaki; Yoshida, Hiroyuki

Journal of Nuclear Science and Technology, 60(9), p.1029 - 1045, 2023/09

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Journal Articles

Development of a numerical simulation method for air cooling of fuel debris by JUPITER

Yamashita, Susumu; Uesawa, Shinichiro; Ono, Ayako; Yoshida, Hiroyuki

Mechanical Engineering Journal (Internet), 10(4), p.22-00485_1 - 22-00485_25, 2023/08

A detailed evaluation for air cooling of fuel debris in actual reactors will be essential in fuel debris retrieval under dry conditions. To understand the heat transfer in and around fuel debris, which is assumed as a porous medium in the primary containment vessel (PCV) mechanistically, we newly applied the porous medium model to the multiphase and multicomponent computational fluid dynamics code named JUPITER (JAEA Utility Program for Interdisciplinary Thermal-hydraulics Engineering and Research). We applied the Darcy-Brinkman model as for the porous medium model. This model has high compatibility with JUPITER because it can treat both a pure fluid and a porous medium phase simultaneously in the same manner as the one-fluid model in multiphase flow simulation. We addressed the case of natural convection with a high-velocity flow standing out nonlinear effects by implementing the Forchheimer model, including the term of the square of the velocity as a nonlinear effect to the momentum transport equation of JUPITER. We performed some simple verification and validation simulations, such as the natural convection simulation in a square cavity and the natural convective heat transfer experiment with the porous medium, to confirm the validity of the implemented model. We confirmed that the result of JUPITER agreed well with these simulations and experiments. In addition, as an application of the updated JUPITER, we performed the preliminary simulation of air cooling of fuel debris in the condition of the Fukushima Daiichi Nuclear Power Station unit 2 including the actual core materials. As a result, JUPITER calculated the temperature and velocity field stably in and around the fuel debris inside the PCV. Therefore, JUPITER has the potential to estimate the detailed and accurate thermal-hydraulics behaviors of fuel debris.

Journal Articles

Development of numerical analysis method of oxygen concentration near wall of lead-bismuth eutectic channel

Watanabe, Nao; Yamashita, Susumu; Uesawa, Shinichiro; Nishihara, Kenji; Yoshida, Hiroyuki

Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.3522 - 3534, 2023/08

Accelerator-driven system (ADS), the coolant of which is lead-bismuth eutectic (LBE), has been designed by Japan Atomic Energy Agency. Estimating corrosion rate at the wall surface of LBE channel is an important issue in considering safety and the life of the entire structure. The corrosion rate depends on state of oxygen layers forming at the material surface. Therefore, this study aims to develop a method to evaluate the corrosion rate in ADS for the design study by estimation of the oxide layer growth and dissolution (OLGD) rates by means of numerical analysis. The OLGD rates, mass transfer rates of oxygen and iron between the material and LBE and advection-diffusion rates of them in LBE depend on each other. Therefore, in order to estimate OLGD rates, the three numerical analysis models should be coupled. For the advection-diffusion calculation, to use CFD code should be reasonable approach to analyze complex flow in ADS, while for the OLGD and the mass transfer calculation, to use some correlation equations should be reasonable because their scales are much smaller than the advection-diffusion. The present work has developed the analysis method of OLGD rates by using JUPITER code, which is CFD code developed in JAEA. In terms of the correlation equations of OLGD and mass transfer rates, existing models used in a previous study were used with modified.

Journal Articles

Development of numerical simulation method of natural convection around heated porous medium by using JUPITER

Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05

For contaminated water management in decommissioning Fukushima Daiichi Nuclear Power Stations, reduction in water injection, intermittent injection water and air cooling are considered. However, since there are uncertainties of fuel debris in the PCV, it is necessary to examine and evaluate optimal cooling methods according to the distribution state of the fuel debris and the progress of the fuel debris retrieval work in advance. We have developed a method for estimating the thermal behavior in the air cooling, including the influence of the position, heat generation and the porosity of fuel debris. Since a large-scale thermal-hydraulics analysis of natural convection is necessary for the method, JUPITER developed independently by JAEA is used. It is however difficult to perform the large-scale thermal-hydraulics analysis with JUPITER by modeling the internal structure of the debris which may consist of a porous medium. Therefore, it is possible to analyze the heat transfer of the porous medium by adding porous models to JUPITER. In this study, we report the validation of JUPITER applied the porous model and discuss which heat transfer models are most effective in porous models such as series, parallel and geometric mean models. To obtain validation data of JUPITER for the natural convective heat transfer analysis around the porous medium, we performed the heat transfer and the flow visualization experiments of the natural convection in the experimental system including the porous medium. In the comparison between the experiment and the numerical analysis with each model, the numerical result with the geometric mean model was the closest of the models to the experimental results. However, the numerical results of the temperature and the velocity were overestimated for those experimental results. In particular, the temperature near the interface between the porous medium and air was more overestimated.

Journal Articles

Double diffusive dissolution model of UO$$_{2}$$ pellet in molten Zr cladding

Ito, Ayumi*; Yamashita, Susumu; Tasaki, Yudai; Kakiuchi, Kazuo; Kobayashi, Yoshinao*

Journal of Nuclear Science and Technology, 60(4), p.450 - 459, 2023/04

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

JAEA Reports

Development of a test method for electrochemical measurements of stainless steel in nitric acid solution containing neptunium-237 under gamma-ray irradiation

Yamashita, Naoki; Irisawa, Eriko; Kato, Chiaki; Sano, Naruto; Tagami, Susumu

JAEA-Technology 2022-035, 29 Pages, 2023/03

JAEA-Technology-2022-035.pdf:2.54MB

In the treatment process of the current commercial reprocessing plant (Rokkasho Reprocessing Plant), the high-level liquid waste concentrator is the equipment that treats the most corrosive solution. In the high-level liquid waste concentrator, the extracted liquid waste after separation of uranium and plutonium is heated, concentrated, and reduced in volume. Therefore, the amount of gamma- rays emitted from fission products and the concentration of corrosive metal ion species such as neptunium-237 ($$^{237}$$Np) are the highest in the reprocessing process, and the amount of corrosion in the high-level liquid waste concentrate canner is expected to be large. In this study, in order to clarify the effect of gamma-rays on the corrosion reaction of stainless steel in nitric acid solutions containing $$^{237}$$Np from the electrochemical viewpoint, the corrosion test apparatus for heat transfer surfaces in an airtight concrete cell at the Waste Safety TEsting Facility (WASTEF) of Nuclear Science Research Institute was modified to enable electrochemical measurements under gamma-ray irradiation. The effect of gamma-rays on the corrosion reaction taking place on the stainless steel surface was discussed from the electrochemical test results obtained. As a result, changes in the immersion potentials of stainless steel and the polarization curves due to chemical species caused by radiolysis of gamma-ray irradiation were confirmed.

JAEA Reports

R&D and maintenance management of the WASTEF Facility (FY2021)

Sano, Naruto; Yamashita, Naoki; Hoshino, Kazutoyo*; Tsukada, Manabu*; Sawauchi, Fumiya*; Otake, Yoshinori; Ichise, Kenichi; Tagami, Susumu

JAEA-Technology 2022-034, 47 Pages, 2023/03

JAEA-Technology-2022-034.pdf:2.81MB

The Waste Safety Testing Facility (WASTEF) was established in 1982 as an experimental facility for long-term storage of solidified high-level radioactive waste generated in the reprocessing of spent light water reactor fuel and the subsequent safety assessment of geological disposal. It is a historic facility that started operation in 1982. This facility consists of 5 concrete cells, 1 lead cell, 6 glove boxes, and 7 hoods, and is a large-scale facility that can use nuclear fuel materials including uranium and plutonium and radioactive isotopes including TRU. In this facility, research and development requested by the research department is carried out in the Hot Material Examination Section. In addition, patrol inspections, self-inspections, etc. are also carried out as maintenance management based on safety regulations. This report summarizes the overview of WASTEF facilities, the results of operation, maintenance and management work in FY2021, and the future outlook.

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