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Journal Articles

Development of gas entrainment evaluation model based on distribution of pressure along vortex center line; Application to a gas entrainment experiment with traveling vortices in an open water channel flow?

Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Sakai, Takaaki*

Nuclear Engineering and Design, 432, p.113785_1 - 113785_16, 2025/02

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

Establishing an evaluation method for the gas entrainment (GE) of argon cover gas due to surface vortices is required in terms of safety design of sodium-cooled fast reactors. To modify the evaluation model in an in-house evaluation tool for GE, StreamViewer, a modified evaluation model on the pressure distribution along the vortex center line (PVL model) was proposed to identify the vortex center lines by connecting continuous vortex center points from the suction port to the surface and evaluate gas core length based on the balance between the hydrostatic pressure and the pressure decrease distribution along the vortex center line. PVL model was applied the three-dimensional numerical analysis results for the experiments where a plate induced unsteady traveling vortices in the open channel flow. Consequently, the GE evaluation using StreamViewer with PVL model could reproduce the relation between the inlet flow velocity and the gas core length in the unsteady vortex flow experiments.

Journal Articles

First freezing experiments with a molten mixture of boron carbide and stainless steel in core disruptive accidents of sodium-cooled fast reactors

Emura, Yuki; Matsuba, Kenichi; Kikuchi, Shin; Yamano, Hidemasa

Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 8 Pages, 2024/11

Journal Articles

Thinning behavior of solid boron carbide immersed in molten stainless steel for core disruptive accident of sodium-cooled fast reactor

Emura, Yuki; Takai, Toshihide; Kikuchi, Shin; Kamiyama, Kenji; Yamano, Hidemasa; Yokoyama, Hiroki*; Sakamoto, Kan*

Journal of Nuclear Science and Technology, 61(7), p.911 - 920, 2024/07

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

Journal Articles

Numerical simulation of accidents involving core damage with integrative severe accident analysis code, SPECTRA

Ishida, Shinya; Uchibori, Akihiro; Okano, Yasushi

Dai-28-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2024/06

no abstracts in English

Journal Articles

Numerical study of initiating phase of core disruptive accident in small sodium-cooled fast reactors with negative void reactivity

Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi

Journal of Nuclear Science and Technology, 61(5), p.582 - 594, 2024/05

 Times Cited Count:1 Percentile:25.62(Nuclear Science & Technology)

JAEA Reports

Investigation of the core neutronics analysis conditions for evaluation of burn-up nuclear characteristics of the next-generation fast reactors

Takino, Kazuo; Oki, Shigeo

JAEA-Data/Code 2023-003, 26 Pages, 2023/05

JAEA-Data-Code-2023-003.pdf:1.66MB

Since next-generation fast reactors aim to achieve a higher core discharge burn-up than conventional reactors do, core neutronics design methods must be refined. Therefore, a suitable analysis condition is required for the analysis of burn-up nuclear characteristics to accomplish sufficient estimation accuracy while maintaining a low computational cost. We investigated the effect of the analysis conditions on the accuracy of estimation of the burn-up nuclear characteristics of next-generation fast reactors in terms of neutron energy groups, neutron transport theory, and spatial mesh. This study treated the following burn-up nuclear characteristics: criticality, burn-up reactivity, control rod worth, breeding ratio, assembly-wise power distribution, maximum linear heat rate, sodium void reactivity, and Doppler coefficient for the equilibrium operation cycle. As a result, it was found that the following conditions were the most suitable: 18-energy-group structure, 6 spatial meshes per assembly with diffusion approximation. Additionally, these conditions should apply to correction factors for energy group structure, spatial mesh and transport effects.

Journal Articles

A 3D particle-based simulation of heat and mass transfer behavior in the EAGLE ID1 in-pile test

Zhang, T.*; Morita, Koji*; Liu, X.*; Liu, W.*; Kamiyama, Kenji

Annals of Nuclear Energy, 179, p.109389_1 - 109389_10, 2022/12

 Times Cited Count:3 Percentile:42.88(Nuclear Science & Technology)

Journal Articles

Analysis on cooling behavior for simulated molten core material impinging to a horizontal plate in a sodium pool

Matsushita, Hatsuki*; Kobayashi, Ren*; Sakai, Takaaki*; Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 9 Pages, 2022/09

During core disruptive accidents in sodium-cooled fast reactors, the molten core material flows through flow channels, such as the control rod guide tubes, into the core inlet plenum under the core region. The molten core material can be cooled and solidified while impinging on a horizontal plate of the inlet plenum in a sodium coolant. However, the solidification and cooling behaviors of molten core materials impinged on a horizontal structure have not been sufficiently studied thus far. Notably, this is an important phenomenon that needs to be elucidated from the perspective of improving the safety of sodium-cooled fast reactors. Accordingly, a series of experiments on discharging a simulated molten core material (alumina: Al$$_{2}$$O$$_{3}$$) into a sodium coolant on a horizontal structure was conducted at the experimental facility of the National Nuclear Center of the Republic of Kazakhstan. In this study, analyses on the sodium experiments using SIMMER-III as the fast reactor safety evaluation code were performed. The analysis methods were validated by comparing the results and experiment data. In addition, the cooling and solidification behaviors during jet impingement were evaluated. The results indicated that the molten core material exhibited fragmentation owing to the impingement on the horizontal plate and was, therefore, scattered toward the periphery. Furthermore, the simulated molten core material was evaluated to be cooled by sodium and subsequently solidified.

Journal Articles

Study on initiating phase of core disruptive accident (Validation study of SAS4A code for the unprotected transient overpower accident)

Ishida, Shinya; Fukano, Yoshitaka

Nihon Kikai Gakkai Rombunshu (Internet), 88(911), p.21-00304_1 - 21-00304_11, 2022/07

In previous studies, the reliability and validity of the SAS4A code was enhanced by applying Phenomena Identification and Ranking Table (PIRT) approach to the Unprotected Loss of Flow (ULOF). SAS4A code has been developed to analyze the early stage of Core Disruptive Accident (CDA), which is named Initiating Phase (IP). In this study, PIRT approach was applied to Unprotected Transient over Power (UTOP), which was one of the most important and typical events in CDA as well as ULOF. The phenomena were identified by the investigation of UTOP event progression and physical phenomena relating to UTOP were ranked. 8 key phenomena were identified and the differences in ranking between UTOP and ULOF were clarified. The code validation matrix was completed and an SAS4A model, which was not validated in ULOF, was identified and validated. SAS4A code became applicable to various scenarios by using PIRT approach to UTOP and the reliability and validity of SAS4A code were significantly enhanced.

Journal Articles

Boundary condition free homogenization and evaluation of its performance in fast reactor core analysis

Maruyama, Shuhei

Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) (Internet), 10 Pages, 2022/05

This paper proposes a new homogenization method, "Boundary Condition Free Homogenization (BCFH)". The traditional homogenization method separates the core calculation and the cell (assembly) calculation by assuming a specific boundary condition or a peripheral region in the cell calculation. Nevertheless, there are ambiguities and approximation in these assumptions, and they can also cause a decline in accuracy. BCFH aims to avoid these problems and improve the accuracy in the cell calculation such as homogenization. We imposed the conditions that the physical quantities in the cell related to the reaction rate preservation is preserved for any incoming partial current, during the homogenization. That is, the response matrices of cell average (or total) flux and outgoing partial current, to be the same form between heterogeneous and homogeneous system. As a result, homogenized parameters, such as cross-sections, superhomgenization factors, and discontinuity factors, are no longer dependent on a specific boundary condition. The new homogenized parameters obtained in this way are extended from the conventional vector form to the matrix form in BCFH. To investigate the performance of BCFH, numerical tests are done for the simplified models which originates in 750MW-class sodium-cooled fast reactor with MOX fuel core in Japan. It is found that BCFH is particularly effective in evaluating control rod reactivity worth and reaction rate distribution, compared to the traditional method. We conclude that the BCFH can be a promising homogenization concept for core neutronic analysis.

JAEA Reports

Development of the unified cross-section set ADJ2017R

Yokoyama, Kenji; Maruyama, Shuhei; Taninaka, Hiroshi; Oki, Shigeo

JAEA-Data/Code 2021-019, 115 Pages, 2022/03

JAEA-Data-Code-2021-019.pdf:6.21MB
JAEA-Data-Code-2021-019-appendix(CD-ROM).zip:435.94MB

In JAEA, several versions of unified cross-section set for fast reactors have been developed so far; we have developed a new unified cross-section set ADJ2017R, which is an improved version of the unified cross-section setADJ2017 for fast reactors. The unified cross-section set is used for reflecting information of C/E values (analysis / experiment values) obtained by integral experiment analyses in reactor core design via the cross-section adjustment methodology; the values are stored in the standard database for FBR core design. In the methodology, the cross-section set is adjusted by integrating the information such as uncertainty (covariance) of nuclear data, uncertainty of integral experiment / analysis, sensitivity of integral experiment with respect to nuclear data. ADJ2017R basically has the same performance as ADJ2017, but we conducted an additional investigation on ADJ2017 and revised the following two points. The first is to unify the evaluation method of the correlation coefficient of uncertainty caused by experiments (hereinafter referred to as the experimental correlation coefficient). Because it was found that the common uncertainty used in the evaluation of the experimental correlation coefficient was evaluated by two different methods, the experimental correlation coefficients were revised for all experimental data, and the evaluation method was unified. The second is the review of the integral experiment data used for the cross-section adjustment calculation. It was found that one of the experimental values of composition ratio after irradiation of the Am-243 sample has a problem in uncertainty evaluation because its experimental uncertainty is extremely small compared to the others. The cross-section adjustment calculation was, therefore, redone by excluding the experimental value. In the creation of ADJ2017, a total of 719 data sets were analyzed and evaluated, and eventually adopted 620 integral experimental data sets. In contrast, a total of 61

Journal Articles

A 3D particle-based analysis of molten pool-to-structural wall heat transfer in a simulated fuel subassembly

Zhang, T.*; Morita, Koji*; Liu, X.*; Liu, W.*; Kamiyama, Kenji

Extended abstracts of the 2nd Asian Conference on Thermal Sciences (Internet), 2 Pages, 2021/10

For the Japanese sodium cooled fast reactor, a fuel subassembly with an inner duct structure (FAIDUS) was designed to avoid the re-criticality by preventing the large-scale pool formation. In the present study, using the finite volume particle method, the EAGLE ID1 test which was an in-pile test performed to demonstrate the effectiveness of FAIDUS was numerically simulated and the thermal-hydraulic mechanisms underlying the heat transfer process were analyzed.

Journal Articles

Numerical simulation of heat transfer behavior in EAGLE ID1 in-pile test using finite volume particle method

Zhang, T.*; Funakoshi, Kanji*; Liu, X.*; Liu, W.*; Morita, Koji*; Kamiyama, Kenji

Annals of Nuclear Energy, 150, p.107856_1 - 107856_10, 2021/01

 Times Cited Count:6 Percentile:56.80(Nuclear Science & Technology)

Journal Articles

Validation of analysis models on relocation behavior of molten core materials in sodium-cooled fast reactors based on the melt discharge experiment

Igarashi, Kai*; Onuki, Ryoji*; Sakai, Takaaki*; Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08

Journal Articles

Routing study of above core structure with mock-up experiment for ASTRID

Takano, Kazuya; Sakamoto, Yoshihiko; Morohoshi, Kyoichi*; Okazaki, Hitoshi*; Gima, Hiromichi*; Teramae, Takuma*; Ikarimoto, Iwao*; Botte, F.*; Dirat, J.-F.*; Dechelette, F.*

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 8 Pages, 2019/05

ASTRID has the objective to integrate innovative options in order to prepare the 4th generation reactors. In ASTRID, large number of tubes are installed above each fuel subassembly to monitor the core. These instrumentations such as thermocouples (TC) and Failed Fuel Detection and Location (FFDL) systems are integrated into Above Core Structure (ACS) with various sizes tubes. In the present study, the routing study for TC tubes and FFDL tubes was performed with 3D modeling and mock-up experiment of the ACS designed for ASTRID with 1500 MW thermal power in order to clarify the integration process and secure the design hypotheses. Although some problems on fabricability were found in the mock-up experiment, the possible solutions were proposed. The present study gives manufacturing feedback to design team and will contribute to increase the knowledge for ACS design and fabricability.

Journal Articles

Particle-based simulation of heat transfer behavior in EAGLE ID1 in-pile test

Morita, Koji*; Ogawa, Ryusei*; Tokioka, Hiromi*; Liu, X.*; Liu, W.*; Kamiyama, Kenji

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 11 Pages, 2018/10

The EAGLE in-pile ID1 test has been performed by Japan Atomic Energy Agency to demonstrate early fuel discharge from a fuel subassembly with an inner duct structure, which is named FAIDUS. It was deduced that early duct wall failure observed in the test was initiated by high heat flux from the molten pool of fuel and steel mixture. The posttest analyses suggest that molten pool-to-duct wall heat transfer might be enhanced effectively by the molten steel with large thermal conductivity in the pool without the presence of fuel crust on the duct wall. In this study, mechanisms of heat transfer from the molten pool to the duct wall was analyzed using a fully Lagrangian approach based on the finite volume particle method for multi-component, multi-phase flows. A series of pin disruption, molten pool formation and duct wall failure behaviors was simulated to investigate mixing and separation behavior of molten steel and fuel in the pool, and their effect on molten pool-to-duct wall heat transfer. The present 2D particle-based simulations demonstrated that large thermal load beyond 10 MW/m$$^{2}$$ on the duct wall was caused by effective heat transfer due to direct contact of liquid fuel with nuclear heat to the duct wall.

Journal Articles

Fast reactor core seismic experiment and analysis under strong excitation

Yamamoto, Tomohiko; Iwasaki, Akihisa*; Kawamura, Kazuki*; Matsubara, Shinichiro*; Harada, Hidenori*

Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 8 Pages, 2018/07

To design fast reactor (FR) core components, seismic response must be evaluated in order to ensure structural integrity. Thus, a core seismic analysis method has been developed to evaluate 3D core vibration behavior considering fluid structure interaction and vertical displacements (rising). 1/1.5 scale 37 core element mock-ups hexagonal-matrix experiment was performed to validate the core elements vibration analysis code in three dimensions (REVIAN-3D). Based on the test data, the analysis model newly incorporated to respond to strong excitation was verified.

Journal Articles

Core seismic experiment and analysis of full scale single model for fast reactor

Yamamoto, Tomohiko; Kitamura, Seiji; Iwasaki, Akihisa*; Matsubara, Shinichiro*; Okamura, Shigeki*

Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 10 Pages, 2017/07

To design fast reactor (FR) components, seismic response must be evaluated in order to ensure structural integrity. Therefore, a sophisticated analysis method has to be developed to study the seismic response of FR core. The fast reactors are made of several hundred core assemblies in hexagonal arrangement. When a big earthquake occurs, large horizontal displacement and impact force of each core assembly may cause a trouble for control rod insertability and core assembly intensity. Therefore, a seismic analysis method of fast reactor core considering horizontal nonlinear behavior, such as impact, fluid-structure interaction, etc. is needed. Validation of the core assembly vibration analysis code in three dimension (REVIAN-3D) was conducted by a full scale experiment. In this validation, the vertical behavior (raising displacement) and horizontal behavior (Impact force, horizontal response) of the analysis result agreed very well with the experiments.

Journal Articles

Study on reactor vessel coolability of sodium-cooled fast reactor under severe accident condition; Water experiments using a scale model

Ono, Ayako; Kurihara, Akikazu; Tanaka, Masaaki; Ohshima, Hiroyuki; Kamide, Hideki; Miyake, Yasuhiro*; Ito, Masami*; Nakane, Shigeru*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

The water experiment apparatus simulating the thermal hydraulics in a reactor vessel under operating the decay heat removal systems (DHRSs) was fabricated. The theoretical evaluation for similarity and results of basic experiments show applicability for a scale model experiment of a sodium-cooled fast reactor. This paper, moreover, describes the results of flow visualization experiment under operating a dipped-type passive DHX, which is planned to be installed in both a loop type reactor and pool type reactor, and the calculation results using FLUENT comparing with the result of water experiment.

Journal Articles

Experimental investigation on characteristics of mixed particle debris in sedimentation and bed formation behavior

Sheikh, M. A. R.*; Son, E.*; Kamiyama, Motoki*; Morioka, Toru*; Matsumoto, Tatsuya*; Morita, Koji*; Matsuba, Kenichi; Kamiyama, Kenji; Suzuki, Toru

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 12 Pages, 2016/10

This paper reports an experimental evaluation on debris bed formation characteristics in core-disruptive accidents cogitating the heterogeneous mixture of particles. In the present study, to appraise the characteristics, a series of experiments was accomplished by gravity driven discharge of solid binary mixtures of particles as simulant debris from a nozzle into a quiescent water pool in isothermal condition at room temperature. Currently, two types of spherical particles, namely Alumina and stainless steel with different diameter are employed to study the effect of key experimental parameters on bed mound shape. In experimental investigation both convex and concave mound shapes were perceived based on the effect of particle size and nozzle diameter. The present outcomes could be useful to validate numerical models and simulation codes of particulate debris sedimentation.

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