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Watanabe, Tomoaki; Aizawa, Naoto*; Chiba, Go*; Tada, Kenichi; Yamamoto, Akio*
Proceedings of International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2025) (Internet), 10 Pages, 2025/04
Currently, a major burnup calculation method for the nuclide composition of nuclear fuel conducts neutron transport calculations at each burnup step to account for changes in the neutron spectrum. While this method is highly accurate, the large computational cost of neutron transport calculations can be problematic. Therefore, a fast burnup calculation method based on neutron spectrum reconstruction with the proper orthogonal decomposition (POD) and regression model is investigated. In this method, dimensionality reduction by POD is applied to many neutron fluxes obtained from detailed burnup calculations for various input parameter sets, and regression models are constructed to connect the dimensionality-reduced neutron fluxes and parameters. By substituting arbitrary input parameters to the regression models, the neutron flux is reconstructed and the burnup calculation is performed. This method performs burnup calculations that consider changes in the neutron spectrum based on input conditions without neutron transport calculations. The present method was applied to a PWR UO
fuel pin cell model. The results show the nuclide inventory can be calculated with a prediction accuracy within a few percent. In addition, it is found that the calculation error is dominated by the regression models, which implies the further improvement of the regression models leads to improving the accuracy.
Koda, Yuya; Nakamura, Yasuyuki; Iguchi, Yukihiro*; Yanagihara, Satoshi*
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 10 Pages, 2024/11
"Fugen" has been proceeding with decommissioning work since receiving approval for the decommissioning plan in 2008, and is currently in the second step of dismantling the reactor peripheral equipment. In decommissioning, project management is important to select strategies and optimize processes, waste, safety, costs, etc. In Japan, it is expected that decommissioning work for nuclear facilities will become more serious in the future, and the knowledge and project management data obtained from the decommissioning work carried out at "Fugen" to date will be useful for planning. Since the dismantling and removal of the main turbine equipment and the peripheral equipment in preparation for the reactors have been completed so far, the actual data has been organized and analyzed and compiled into unit work coefficients. The project management data, collected during the decommissioning work carried out at "Fugen" is organized and calculated as a unit work coefficient.
Yada, Hiroki; Takaya, Shigeru; Machida, Hideo*
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 7 Pages, 2024/11
Matsuba, Kenichi; Kato, Shinya; Kamiyama, Kenji; Akaev, A. S.*; Vurim, A. D.*; Baklanov, V. V.*
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 7 Pages, 2024/11
During a severe accident in sodium-cooled fast reactors, molten core materials could be discharged from the core region toward the lower sodium region of the reactor vessel through coolant channels, such as control rod guide tubes. Typical SFRs have a sodium plenum with limited depth and volume, such as the core inlet plenum located under the core region. Therefore, it is important to evaluate the coolability of molten core materials discharged into a depth- and volume-limited sodium plenum. In the present study, to deepen the understanding on the coolability of molten core materials discharged into such a sodium plenum, conditions under which molten core materials form solidified fragments were discussed based on an experiment discharging a molten fuel simulant (molten Al2O3) into a test vessel filled with liquid sodium.
Morita, Keisuke; Aoki, Takeshi; Shimizu, Atsushi; Sato, Hiroyuki
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 6 Pages, 2024/11
Okita, Shoichiro; Aoki, Takeshi; Fukaya, Yuji; Tachibana, Yukio
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 5 Pages, 2024/11
Isobe, Yuta; Tanigaki, Takanori; Tone, Kohei; Joboji, Yuya; Matsui, Kazuaki; Obata, Ikuhito
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 8 Pages, 2024/11
The prototype fast breeder reactor Monju is a loop-type, sodium-cooled nuclear reactor in Japan. Monju transitioned; to decommissioning in 2018, and the first phase of decommissioning was completed in FY2022 by unloading all the fuel assemblies from the reactor core and Ex-vessel storage tank (EVST) to spent fuel pool. In FY2023, Monju has moved to the second phase of decommissioning, and currently we are working to unload neutron shields and other core elements from the reactor core and EVST to spent fuel pool, and to dismantle water and steam system such as turbine generator, condensers, feed water heaters, etc. Monju has a primary sodium system that cools the reactor core and fuel, and a secondary sodium system that transfers heat. Sodium in the primary system is radioactive, but sodium in the secondary system is non-radioactive. We plan to extract non-radioactive sodium first, so we are first considering this in advance. At present, we have decided on the sodium extraction area where the ISO tank for extraction of non-radioactive sodium will be installed. In addition, we also considered the route for transferring sodium from the existing tanks to the ISO tanks. New piping for these operations will be added, and the sodium will be extracted using existing gas systems and electromagnetic pumps. We will take appropriate measures to sodium leakage and continue to consider safe and rational extraction and transportation methods.
Nakamura, Hiroki; Machida, Masahiko
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 5 Pages, 2024/11
Efficiency and safety improvements in nuclear fuel development demand a comprehensive understanding of the thermal properties of PuO
. Because of difficulty of experiments at high temperatures, they have been supplemented by numerical simulations, such as density functional theory (DFT). While DFT struggles to replicate the nonmagnetic insulating ground state of PuO, DFT+U can reproduce nonmagnetic insulating state. However, the obtained state remains less stable than magnetic states and is not the ground state. Adiabatic connection fluctuation-dissipation theory (ACFDT) is expected to be a promising solution, addressing higher-order correlations and exact exchange energies. In this study, we evaluate the ground state energy using ACFDT and find that the nonmagnetic state can be the ground state. This result successfully reproduces the observed nonmagnetic ground state of PuO and has the potential to improve predictions of the thermal properties of nuclear fuel materials.
Kang, Z.; Okuda, Yukihiko; Nishida, Akemi; Tsubota, Haruji; Ito, Masaharu; Li, Y.
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 9 Pages, 2024/11
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, 2024/11
Kamiya, Tomohiro; Nagatake, Taku; Ono, Ayako; Tada, Kenichi; Kondo, Ryoichi; Nagaya, Yasunobu; Yoshida, Hiroyuki
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 7 Pages, 2024/11
We have developed the JAEA Advances Multi-Physics Analysis platform for Nuclear systems (JAMPAN) to realize high-fidelity neutronics/thermal-hydraulics coupling simulations. We will perform MVP/JUPITER coupling simulation for a single BWR fuel assembly in order to confirm that the neutronics/thermal-hydraulics coupling through JAMPAN is feasible. This presentation explains how to send and receive data between MVP and JUPITER through JAMPAN and simulation results.
Kobayashi, Hideharu; Naruse, Keiji; Hirako, Kazuhito; Sawazaki, Hiromasa; Goto, Takehiro; Obata, Ikuhito; Matsui, Kazuaki
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 9 Pages, 2024/11
The prototype fast breeder reactor MONJU is a loop-type sodium-cooled nuclear reactor to be decommissioned in Japan. Japan Atomic Energy Agency (JAEA) started decommissioning of MONJU in 2018, and will complete four phases of the decommissioning in about 30 years. The first Phase is the Fuel Assembly Removal Period, during which secondary sodium was drained, the fuel assemblies in the reactor core were removed and put in the storage in the spent fuel pool, and the distribution of contamination in buildings, components, equipment, etc. was evaluated. In the second Phase, the Preparation for dismantling period, the neutron shields in the reactor core will be moved and put in the storage in the spent fuel pool in preparation for the dismantling of the sodium equipment, the transport of sodium, and the dismantling and removal of the water and steam power generation system. We also continue to assess the distribution of contamination in buildings, components, equipments, etc. The third Phase is the Decommissioning Period I, which includes the dismantling of sodium equipment, the transport of spent fuel, and the removal of the water and steam power generation equipment. The final Phase is the Decommissioning Period II. The radiation controlled area will be freed and all the buildings will be dismantled and removed. This paper provides an overview of MONJU decommissioning, the results of its First Phase, and details of the second Phase, which is currently underway.
Wen, J.*; Kamada, Yuto*; Yokoyama, Kosei*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Imaizumi, Yuya; Tagami, Hirotaka; Matsuba, Kenichi; Kamiyama, Kenji
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 8 Pages, 2024/11
Yamada, Takeshi*; Li, X.; Yamashita, Takuya; Yamaji, Akifumi*
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 10 Pages, 2024/11
In this study, a new crust model is being developed to analyze MCCI, which involves continuous concrete ablation with presence of the crust layer between the corium and the concrete walls, which may gradually move with the slow concrete wall ablation process over long time. The new crust model must enable accumulation of physical displacement of the crust particle over long time (i.e., enable physical creeping) while preventing accumulation of numerical displacement of the crust particles over long time (i.e., preventing numerical creeping), Hence, in the new crust model, the PS has been effectively disabled for the crust particles. Qualitative validity of such numerical modeling was confirmed through some trial analyses of VULCANO-VBS test using a set of tentative calculation conditions and parameters, which should be carefully revised for future quantitative discussions including validation of the analysis results with experimental results.
Yuki, Kohei*; Horiguchi, Naoki; Yoshida, Hiroyuki; Yuki, Kazuhisa*
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 4 Pages, 2024/11
Fuel debris in the Fukushima Nuclear Power Station is cooled under immersion condition. However, in the event of an unexpected decrease in water level, coolant contacts high-temperature fuel debris having porous structure. In this event, although fuel debris needs to be cooled rapidly, thermal behavior at liquid-solid contact, such as capillary phenomenon, remains unclear. In this paper, as basic research, we evaluate droplet evaporation characteristics after contact with metal porous media with small pores less than 1 mm. In experiment, to obtain life time curve of a droplet, bronze or stainless steel porous media having 1, 40, or 100 
m pore diameter are utilized. Experimental results show that Leidenfrost phenomenon is suppressed on the porous surfaces because generated vapor can be discharged from the pores. Further, for bronze porous media, capillary phenomenon is observed as the temperature of the porous media increase because of generation of oxide film having fine structure. On the other hand, due to low wettability of stainless steel porous media, capillary phenomenon does not occur, and the droplet was not sucked and spread into pore. This indicates that rapid cooling by the capillary phenomenon can not be expected if fuel debris has the same characteristics as the stainless steel porous media.
C-SS and its relocation behaviorAhmed, Z.*; Sharma, A. K.*; Pellegrini, M.*; Yamano, Hidemasa; Okamoto, Koji*
Proceedings of Saudi International Conference On Nuclear Power Engineering (SCOPE2023) (Internet), 8 Pages, 2023/11
In this study, the eutectic behavior and subsequent melt structure of boron migration are observed by a quantitative and high-resolution visualization method using radiative heating. Experiments were conducted using B4C pellet and powder within SS tubes, replicating the actual control rod design in the temperature range of 1150
C to 1372C to study long-duration melting and relocation behavior. The visualization technique accurately identified the time of eutectic melting onset and the related temperature, pointing out different values for the pellet and the powder cases.
Kondo, Ryoichi; Nagaya, Yasunobu
Proceedings of International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2023) (Internet), 10 Pages, 2023/08
A functional expansion tally (FET) method with numerical basis functions generated by singular value decomposition (SVD) is newly proposed. Traditionally, analytical functions were used for the FET calculations, e.g., Legendre polynomials for a one-dimensional distribution. However, the expansion terms could increase to reconstruct steep or complex distributions with these functions. A basis set that can well represent the target distribution with lower order expansion is desired to achieve high accuracy with the small computational resource. In the present study, a numerical basis set is generated from snapshot data by using SVD. This approach is based on the reduced order modeling (ROM). We applied ROM to the FET method in Monte Carlo calculations. The numerical result showed the applicability of the proposed method, on the other hand, some issues were revealed, e.g., discretization of the snapshot data.
Sugawara, Takanori; Kunieda, Satoshi
Proceedings of International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2023) (Internet), 7 Pages, 2023/08
This study investigates the impact of the change from JENDL-4 to JENDL-5 on neutronics analysis of transmutation systems. As the transmutation systems, the following two systems are targeted: JAEA-ADS, a lead-bismuth cooled accelerator-driven system, and MARDS, a molten salt chloride accelerator-driven system. For the JAEA-ADS, the k-eff value increased 189 pcm from JENDL-4 to JENDL-5. It was found that the revisions of various nuclides affected to this difference. For example, the revision of 
N indicated an increase of 200 pcm from the JENDL-4 result. For the MARDS, it was found that the major revision of 
Cl and 
Cl cross sections was the main cause of the k-eff differences. This study confirmed that the difference in the nuclear data libraries still indicated differences in calculation results for the transmutation systems.
CLuu, V. N.; Nakajima, Kunihisa
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05
Ota, Hirokazu*; Ogata, Takanari*; Yamano, Hidemasa; Futagami, Satoshi; Shimada, Sadae*; Yamada, Yumi*
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 8 Pages, 2023/05
