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Fukuda, Kodai
Annals of Nuclear Energy, 208(1), p.110748_1 - 110748_10, 2024/12
Gunji, Satoshi; Araki, Shohei; Izawa, Kazuhiko; Suyama, Kenya
Annals of Nuclear Energy, 209, p.110783_1 - 110783_7, 2024/12
Since the compositions and properties of the fuel debris are uncertain, critical experiments are required to validate calculation codes and nuclear data used for the safety evaluation. For this purpose, the Japan Atomic Energy Agency (JAEA) has been modifying a critical assembly called "STACY." The first criticality of the modified STACY is scheduled for spring 2024. This paper reports the consideration results of the specifications of the basic core configurations of the modified STACY at the first criticality. We prepared two types of gird plates with different neutron moderation conditions (their intervals are 1.50 cm and 1.27 cm). However, there is a limitation on the number of available UO
fuel rods. The core configurations for the first criticality satisfying these experimental constraints were designed by computational analysis. A cylindrical core configuration with a 1.50 cm grid plate close to the optimum moderation condition needs 253 fuel rods to reach criticality. As to the 1.27 cm grid plate, we considered core configurations with 2.54 cm intervals by using doubled pitches of the grid plate. It will need 213 fuel rods for the criticality. In addition, we considered the experimental core configuration with steel/concrete simulant rods to simulate fuel debris conditions. This paper shows these core configurations and their evaluated specifications.
Yoshida, Naoki; Ono, Takuya; Amano, Yuki; Yoshida, Ryoichiro; Abe, Hitoshi; Yamane, Yuichi
Nuclear Technology, 210(10), p.1999 - 2007, 2024/10
A malfunction of the cooling system of high-level liquid waste (HLLW) and failure of countermeasures may lead to the "evaporation to dryness due to the loss of cooling functions" (EDLCF) of HLLW. In the EDLCF, ruthenium (Ru) can be released at a greater fraction to initial amount than other elements in HLLW by forming gaseous Ru. It is important to identify the chemical form of the released gaseous Ru to achieve a comprehensive understanding of the events impacting the source term assessment of Ru in this accident, such as particle formation, gas absorption and deposition on migration pathways. In this study, we observed the ultraviolet/visible spectroscopy of the off-gas generated during the heating of an HLLW simulant. Employing a program that allows the separation and quantification of known components within the spectrum (ruthenium tetroxide (RuO
), nitrogen dioxide, and nitric acid), we attempted to analyze the composition of gaseous Ru within the generated off-gas. Our findings reveal RuO as the main component of the gaseous Ru in off-gas after comparing the total amount of released Ru and the RuO released amount obtained via spectroscopic analysis.
Takeda, Takeshi; Shibata, Taiju
JAEA-Review 2024-040, 29 Pages, 2024/09
JAEA-Review-2024-040.pdf:1.33MB
An important theme of Japan's 6th strategic energy plan is to indicate the energy policy path towards carbon neutrality by 2050. Policy responses for Japan's nuclear energy research and development (R&D) towards 2030 contain the demonstrations of technologies for small modular reactors (SMRs) through international cooperation by 2030. In light of this energy plan, basic policy initiatives over the next 10 years have been compiled to realize Green Transformation (GX), which simultaneously achieves decarbonization and economic growth. Looking overseas, activities of SMR R&D are active internationally, mainly in the US, Canada, Europe, China, and Russia. These activities are not only by heavy industry manufactures and R&D institutes, but also by venture companies. Under these circumstances, the NEA CSNI has gathered an Expert Group on SMRs (EGSMR) to help estimate the safety effects of SMRs. The EGSMR efforts required the submission of responses to several questionnaires whose main purpose was to collect the latest information on the efforts of SMR deployment and research. The first author of this report responded to this based on information from Hitachi-GE Nuclear Energy, Ltd. and Mitsubishi Heavy Industries, Ltd. as well as JAEA. Most of the responses from Japan to the questionnaires are the information that serves as the basis of CSNI Technical Opinion Paper No. 21 (TOP-21). In this report, the Japan's publicly available responses to the questionnaires arranged and additional information are explained, which complements some of the content of the TOP-21. In this manner, the investigation results of R&D related to SMR in Japan, focusing on the EGSMR activities (2022-2023), are summarized. The target of this report is to provide useful information for future discussions on international cooperation concerning SMR as well as nuclear power field human resources development internationally and domestically.
Motegi, Kosuke; Shibamoto, Yasuteru; Hibiki, Takashi*; Tsukamoto, Naofumi*; Kaneko, Junichi*
JAEA-Review 2024-039, 45 Pages, 2024/09
JAEA-Review-2024-039.pdf:2.23MB
Several heat transfer correlations have been reported related to single-phase opposing flow; however, these correlations are based on experiments conducted in various channel geometries, working fluids, and thermal flow parameter ranges. Therefore, establishing a guideline for deciding which correlation should be selected based on its range of applicability and extrapolation performance is important. This study reviewed the existing heat transfer correlations for turbulent opposing-flow mixed convection. Furthermore, the authors evaluated the predictive performance of each correlation by comparing them with the experimental data obtained under various experimental conditions. The Jackson and Fewster, Churchill, and Swanson and Catton correlations can accurately predict all the experimental data. The authors confirmed that heat transfer correlations using the hydraulic-equivalent diameter as a characteristic length can be used for predictions regardless of channel-geometry differences. Furthermore, correlations described based on nondimensional dominant parameters can be used for predictions regardless of the differences in working fluids.
Yoshida, Kazuo; Hiyama, Mina*; Tamaki, Hitoshi
JAEA-Research 2024-007, 24 Pages, 2024/08
JAEA-Research-2024-007.pdf:2.1MB
An accident of evaporation to dryness by boiling of high-level radioactive liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into the atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. It has been observed experimentally that volatility of RuO is suppressed by HNO generated by HNO
radiolysis. The analysis of chemical reactions of NO
including HNO and HNO in the waste tank is essential to simulate of these phenomena. To resolve this issue, an analytical approach has been attempted to couple dynamically two computer codes SHAWED and SCHERN. The simulation of boiling behavior in the tank is conducted with SHAWED. SCHERN simulates chemical behaviors of HNO, HNO and NO in the tank. A programmatic coupling algorithm and a trial simulation of the accident are presented in this report.
Kang, Z.; Okuda, Yukihiko; Nishida, Akemi; Tsubota, Haruji; Ito, Masaharu; Li, Y.
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 9 Pages, 2024/08
Most studies conducted till now on local damage of reinforced concrete (RC) slab structures subjected to missile impact are about normal impact, while few research related to oblique impact can be found. The objective of this study is to carry out impact tests under different impact conditions including oblique impacts, to confirm the different impact behaviors of the RC slab structure, to develop an analysis method by investigating the test results and analytical conditions, and to validate the analysis method through comparison with the test results. This study focuses on the effect of the stiffness of the supports for oblique impact on the reaction forces of RC slab. Until now, static loading tests were conducted to confirm the stiffnesses of the supporting parts that supported the RC specimen during oblique impact tests. Based on the obtained load-displacement relationships, and so on, the stiffnesses of the supporting parts are estimated.
Sugiura, Ayumu*; Takito, Kiyotaka; Furuya, Osamu*; Nakamura, Izumi*; Okuda, Yukihiko
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 8 Pages, 2024/08
Ueki, Taro
Progress in Nuclear Energy, 173, p.105236_1 - 105236_10, 2024/08
The theme of this paper is how to efficiently analyse extreme realizations of neutron effective multiplication factor (keff) over random media replicas modelled by incomplete randomized Weierstrass function (IRWF). To this end, a new bounded amplification (BA) technique is applied to IRWF. Numerical results indicate that the BA-applied IRWF reduces a required number of random media replicas at least by an order of magnitude. To validate this efficiency gain, generalized extreme value (GEV) analysis is applied to a data set of keff values obtained without applying BA. It turns out that the extreme values of these keff values follow the Weibull distribution. Therefore, the theory of GEV guarantees the existence of the upper limit of these keff values, and the actually computed upper limit is indeed smaller than the top two keff values obtained from an order-of magnitude reduced number of BA-applied IRWF random media replicas. This means that the efficiency gain via BA has been confirmed by GEV analysis.
Tanimura, Yoshihiko; Yoshitomi, Hiroshi; Nishino, Sho; Tsuji, Tomoya; Fukami, Tomoyo; Shinozuka, Tomoki; Oishi, Kohei; Ishii, Masato; Takamiya, Kei; Onuki, Takaya; et al.
Radiation Measurements, 176, p.107196_1 - 107196_6, 2024/08
The ICRU has proposed to change the definitions of the operational quantities used for the area and individual monitoring for external exposure in the ICRU Report 95. As introducing the new operational quantities into the radiation monitoring may affect the dose assessment results using the present personal dosimeters, it is necessary to characterize the energy spectrum in the workplace and the energy dependency of the dosimeters to be used. In this work the photon spectra were measured using a NaI(Tl) scintillation detector or a LaBr
(Ce) scintillation detector at the workplaces in the Japanese Research Reactor No.3 (JRR-3) and the Japan Proton Accelerator Research Complex (J-PARC) at Japan Atomic Energy Agency (JAEA). Then the present and new operational quantities were evaluated using the above mention spectra at the workplaces and compared each other.
Tsujimura, Norio
JAEA-Review 2024-028, 19 Pages, 2024/07
JAEA-Review-2024-028.pdf:1.36MB
The surface density limits currently set by Japan's regulations for radioactive surface contamination, 4 Bq/cm
for 
-emitting radionuclides and 40 Bq/cm for 
/
-emitting radionuclides, are based on the derived working limits studied by Dunster of the UKAEA in the 1950s and 1960s. This derived working limit was introduced into Japan's and the UK's legislation at about the same time. Since then, however, the regulatory content in Japan has remained almost unchanged, while in the UK it has been amended several times over the past few decades. The most recent UK legislation does not include specific numerical requirements for surface contamination density; therefore, nuclear operators are now taking voluntary measures such as setting their radiological control standards for surface contamination. This report provides primary reference material for discussing Japan's future regulatory framework for surface contamination density, including the historical evolution of the differences between Japan's and the UK's regulations.
Risk Analysis Research Group, Reactor Safety Research Division, Nuclear Safety Research Center
JAEA-Data/Code 2024-006, 40 Pages, 2024/07
JAEA-Data-Code-2024-006.pdf:1.92MB
The Risk Analysis Research Group, Reactor Safety Research Division, Nuclear Safety Research Center, Sector of Nuclear Safety Research and Emergency Preparedness, Japan Atomic Energy Agency has been developing OSCAAR, a probabilistic risk assessment program for nuclear facility accidents. OSCAAR has the feature to calculate atmospheric concentrations of radioactive materials using an atmospheric dispersion model. This feature requires the input of meteorological data about wind speed, precipitation rate, atmospheric stability and so on. However, to use numerical weather prediction data created from the Japan Meteorological Agency (JMA) on OSCAAR, it is necessary to convert the data format to match OSCAAR input format in advance. Therefore, we developed GPV2OSC, a pre-processing program for OSCAAR, to create meteorological data converted from JMA weather prediction data format to OSCAAR input format when the target region and period are specified. This report describes the outline and usage of GPV2OSC.
Suyama, Kenya; Gunji, Satoshi; Watanabe, Tomoaki; Araki, Shohei; Fukuda, Kodai; Shimada, Kazuya; Fujita, Tatsuya; Ueki, Taro; Nguyen, H.
JAEA-Conf 2024-001, 40 Pages, 2024/07
JAEA-Conf-2024-001.pdf:1.28MB
JAEA-Conf-2024-001-appendix(CD-ROM).zip:163.97MB
The 12th International Conference on Nuclear Criticality Safety (ICNC2023) was held from October 1 to October 6, 2023, at the Sendai International Center (Aobayama, Aoba-ku, Sendai, Miyagi-prefecture 980-0856, Japan), organized by Japan Atomic Energy Agency (JAEA) and co-organized by the Reactor Physics Division of the Atomic Energy Society of Japan (AESJ) and the Nuclear Energy Agency of the Organization for Economic Co-operation and Development (OECD/NEA). 224 presentations passed peer review and 273 technical session registrations, bringing the total number of registered participants to 289, including accompanying persons. Technical tours were also conducted to i) Fukushima Daiichi Nuclear Power Station of TEPCO holdings and Interim Storage Facility Information Center, ii) Nuclear Science Research Institute of JAEA (STACY Renewable Reactor and FCA), iii) NanoTerasu of Tohoku University (synchrotron radiation facility) and Onagawa Nuclear Power Station of Tohoku Electric Power Co., Inc. This report summarizes the conference and compiles the papers that were presented and agreed to be published in the Proceedings.
Abe, Satoshi; Shibamoto, Yasuteru
Annals of Nuclear Energy, 202, p.110461_1 - 110461_16, 2024/07
Times Cited Count：0 Percentile：0.05Watanabe, Tomoaki; Yamane, Yuichi
Journal of Nuclear Science and Technology, 61(7), p.958 - 966, 2024/07
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)The total fission energy released in a criticality accident involving fissile solution boiling tends to be high because the relatively high fission power continues during boiling. Simulating fission power change correctly during boiling seems essential to estimate the total fission energy. Fission power during boiling changes depending on fissile concentration and volume as the solution evaporates. In this study, we investigated the effect of concentration and volume change on estimated total fission energy for a long time of boiling. We introduced a model calculating the evaporation of fissile solution into the modified quasi-steady-state method to simulate power change during boiling. Three CRAC experiments and the Idaho Chemical Processing Plant (ICPP) criticality accident in 1959 were analyzed. As a result, the calculated energy considering concentration and volume change during boiling reproduced the measured energy well.
Lacroix, V.*; Dulieu, P.*; Hasegawa, Kunio
Proceedings of ASME 2024 Pressure Vessels & Piping Conference (PVP 2024) (Internet), 5 Pages, 2024/07
When a subsurface flaw is detected close to the surface of a pressure retaining component, the subsurface flaw shall be transformed into a surface flaw in accordance with flaw to surface proximity rules. This re-characterization process from subsurface to surface flaw is adopted in ASME Code Section XI. The limit value for this proximity rule depends on ratio between the subsurface flaw to the free surface distance and the flaw depth. This limit value does not depend on the flaw aspect ratio. This paper proposes an interaction limit criterion and presents a new limit value for the flaw-to-surface proximity factor, depending on the flaw depth, the remaining ligament and the flaw aspect ratio.
Nakamura, Izumi*; Takito, Kiyotaka; Shimazu, Ryuya*; Okuda, Yukihiko; Sakai, Michiya*; Otani, Akihito*; Watakabe, Tomoyoshi; Okuda, Takahiro; Shibutani, Tadahiro*; Shiratori, Masaki*
Proceedings of ASME 2024 Pressure Vessels & Piping Conference (PVP 2024) (Internet), 9 Pages, 2024/07
Takito, Kiyotaka; Okuda, Yukihiko; Nishida, Akemi; Li, Y.
Proceedings of ASME 2024 Pressure Vessels & Piping Conference (PVP 2024) (Internet), 9 Pages, 2024/07
In probabilistic risk assessment against earthquakes (seismic PRA), one of the important issues is the development of a realistic response analysis method for evaluating the fragility of equipment and piping systems subject to input ground motions exceeding design assumption. Particularly, piping systems have complex plant-specific three-dimensional geometries. The arrangement and rigidity of piping support structures have significant impact on the response characteristics of the entire piping system. Therefore, it is necessary to develop a seismic response analysis method for piping systems, including piping support structures. To this end, the authors have been working on the development of elasto-plastic analysis method for piping support structures in order to establish a realistic response analysis method for piping systems including piping support structures. In this study, parametric studies on important analytical parameters were conducted to understand the influence of the analytical parameters on the deformation in the elasto-plastic analysis of piping support structures. At first, a three-dimensional finite element analysis model was created for cyclic loading tests of a gate type piping support structure assuming large deformations. Using the model, parametric studies were conducted for analysis parameters such as yield stress, secondary stiffness and element size. As the result, the influences of the analysis parameters on the analysis results were quantitatively clarified, which provided insight into the methodology for setting appropriate analysis parameters on elasto-plastic analysis for piping support structures.
Morley, A.*; Negyesi, M.*; Hasegawa, Kunio
Proceedings of ASME 2024 Pressure Vessels & Piping Conference (PVP 2024) (Internet), 7 Pages, 2024/07
Fatigue crack growth rates for stainless steels in air environment are provided by the ASME Code in Section XI. The crack growth rates increase with increasing temperature, and the temperature effect is expressed by a parameter ST. The current parameter ST is cumbersome, and original basis for the temperature function is questionable, consisting mostly of testing outside of the bounds of the temperatures of interest to its application in ASME Code Section XI. An alternative equation for ST for stainless steels in an air environment is proposed for ASME Code Section XI, based on the literature experimental data surveyed in this study.
Shimada, Taro; Shimada, Asako; Miwa, Kazuji*; Nabekura, Nobuhide*; Sasaki, Toshihisa*; Takai, Shizuka; Takeda, Seiji
JAEA-Research 2024-004, 115 Pages, 2024/06
JAEA-Research-2024-004.pdf:6.02MB
We have studied the confirmation method for the termination of decommissioning of nuclear facilities based on the site release flow presented at the Nuclear Regulation Authority (NRA) study team meeting in 2017, and organized it as a procedure for the site soil. First, the effects of radionuclides released by the Fukushima Daiichi Nuclear Power Station accident are excluded as background radioactivity, and the distribution of radioactivity concentration of facility origin on the site is evaluated using geostatistical method kriging. Then, considering the downstream transport of sediment by surface runoff generated by rainfall that exceeds the infiltration capacity of the ground surface, a series of evaluation procedures are presented to evaluate the exposure dose reflecting future changes from the evaluated radioactivity concentration distribution, and a comparison method with the assumed 0.01 mSv/y as a dose criterion is proposed. Furthermore, an example of the procedure for evaluating the distribution of contamination in the subsurface was also presented for the case where groundwater is affected.
