Numerical investigations on the coolability and the re-criticality of a debris bed with the density-stratified configuration

Li, C.-Y.; Uchibori, Akihiro; Takata, Takashi; Pellegrini, M.*; Erkan, N.*; Okamoto, Koji*

Dai-25-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2021/07

The capability of stable cooling and avoiding re-criticality on the debris bed are the main issues for achieving IVR (In-Vessel Retention). In the actual situation, the debris bed is composed of mixed-density debris particles. Hence, when these mixed-density debris particles were launched to re-distribute, the debris bed would possibly form a density-stratified distribution. For the proper evaluation of this scenario, the multi-physics model of CFD-DEM-Monte-Carlo based neutronics is established to investigate the coolability and re-criticality on the heterogeneous density-stratified debris bed with considering the particle relocation. The CFD-DEM model has been verified by utilizing water injection experiments on the mixed-density particle bed in the first portion of this research. In the second portion, the coupled system of the CFD-DEM-Monte-Carlo based neutronics model is applied to reactor cases. Afterward, the debris particles' movement, debris particles' and coolant's temperature, and the k-eff eigenvalue are successfully tracked. Ultimately, the relocation and stratification effects on debris bed's coolability and re-criticality had been quantitatively confirmed.

Development of unstructured mesh-based numerical method for sodium-water reaction phenomenon in steam generators of sodium-cooled fast reactors

Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohshima, Hiroyuki

Journal of Nuclear Science and Technology, 54(10), p.1036 - 1045, 2017/10

https://doi.org/10.1080/00223131.2017.1344155

To evaluate a sodium-water reaction phenomenon in a steam generator of sodium-cooled fast reactors, a computational fluid dynamics code SERAPHIM, in which a compressible multicomponent multiphase flow with sodium-water chemical reaction is computed, has been developed. The original SERAPHIM code is based on the difference method. In this study, unstructured mesh-based numerical method was developed to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Numerical analysis of an underexpanded jet experiment was performed as part of validation of the unstructured mesh-based numerical method. The calculated pressure profile showed good agreement with the experimental data. Applicability of the numerical method for the actual situation was confirmed through the analysis of water vapor discharging into liquid sodium. The effect of use of the unstructured mesh was also investigated by the two analyses using structured and unstructured mesh.

Proceedings of the 2014 Symposium on Nuclear Data; November 27-28, 2014, Conference hall, Hokkaido University, Sapporo Japan

Aikawa, Masayuki*; Iwamoto, Osamu; Ebata, Shuichiro*; Kunieda, Satoshi; Nakamura, Shoji; Koura, Hiroyuki

JAEA-Conf 2015-003, 332 Pages, 2016/03

JAEA-Conf-2015-003.pdf:30.22MB

The 2014 Symposium on Nuclear Data was held at Conference Hall, Hokkaido University, on November 27 and 28, 2014. The symposium was organized by the Nuclear Data Division of the Atomic Energy Society of Japan, Hokkaido Branch of the Atomic Energy Society of Japan, and Nuclear Reaction Data Centre, Faculty of Science, Hokkaido University in cooperation with Nuclear Science and Engineering Directorate of Japan Atomic Energy Agency. In the symposium, there were two tutorials, "Cross section measurement strategy for long lived fission product" and "Physics and Nuclear Data in Radiation Therapy" and four sessions, "A Neutron TOF Measurement Instrument desired by Nuclear Data Community", "Recent Topics", "Application of Nuclear Data", and "Nuclear Theory and Nuclear Data". In addition, recent research progress on experiments, evaluation, benchmark and application was presented in a poster session. Among 88 participants, all presentations and following discussions were very active and fruitful. This report consists of total 62 papers including 2 tutorials, 16 oral and 44 poster presentations.

Nuclear technology and potential ripple effect of superconducting magnets for fusion power plant

Nishimura, Arata*; Muroga, Takeo*; Takeuchi, Takao*; Nishitani, Takeo; Morioka, Atsuhiko

Fusion Engineering and Design, 81(8-14), p.1675 - 1681, 2006/02

https://doi.org/10.1016/j.fusengdes.2005.09.007

In a fusion reactor plant, a neutral beam injector (NBI) will be operated for a long time, and it will allow neutron streaming from NBI ports to outside of the plasma vacuum vessel. It requires the superconducting magnet to develop nuclear technology to produce stable magnetic field and to reduce activation of the magnet components. In this report, the back ground of the necessity and the contents of the nuclear technology of the superconducting magnets for fusion application are discussed and some typical investigation results are presented, which are the neutron irradiation effect on NbSn wire, the development of low activation superconducting wire, and the design concept to reduce nuclear heating and nuclear transformation by streaming. In addition, recent activities in high energy particle physics are introduced and potential ripple effect of the technology of the superconducting magnets is described briefly.

Introduction to modern nodal method and discontinuity factor

Okumura, Keisuke

Nihon Genshiryoku Gakkai Dai-36-Kai Robutsuri Kaki Semina Tekisuto, p.81 - 102, 2004/08

The modern node method which uses a discontinuous factor has come to be widely used recently in the reactor core analyses of commercial light water reactors. The basic theory, numerical computation technique and examples of calculation results are explained for biginners of the modern nodal method.

Experimental study on accelerator driven subcritical reactor, JAERI's nuclear research promotion program, H12-031 (Contract research)

Shiroya, Seiji*; Misawa, Tsuyoshi*; Unesaki, Hironobu*; Ichihara, Chihiro*; Kobayashi, Keiji*; Nakamura, Hiroshi*; Shin, Kazuo*; Imanishi, Nobutsugu*; Kanazawa, Satoshi*; Mori, Takamasa

JAERI-Tech 2004-025, 93 Pages, 2004/03

JAERI-Tech-2004-025.pdf:6.69MB

In view of the future plan of Research Reactor Institute, Kyoto University, the present study consisted of (1) the transmission experiments of high energy neutrons through materials, (2) experimental simulation of ADSR using the Kyoto University Critical Assembly(KUCA), and (3) conceptual neutronics design study on KUR type ADSR using the MCNP-X code. Through the present study, valuable knowledge on the basic nuclear characteristics of ADSR, which is indispensable to promote the study on ADSR, was obtained both theoretically and experimentally. For the realization of ADSR, it is considered to be necessary to accumulate results of research steadily. For this purpose, it is inevitable (1) to compile the more precise nuclear data for the wide energy range, (2) to establish experimental techniques for reactor physics study on ADSR including subcriticality measurement and absolute neutron flux measurement, and (3) to develop neutronics calculation tools which take into account the neutron generation process by the spallation reaction and the delayed neutron behavior.

Basic concept of Common reactor physics Code Systems; Final report of working party on Common reactor physics Code Systems (CCS)

Research Committee on Reactor Physics

JAERI-Review 2004-003, 152 Pages, 2004/03

JAERI-Review-2004-003.pdf:5.8MB

A working party was organized for two years (2001-2002) on common reactor physics code systems under the Research Committee on Reactor Physics of JAERI. This final report is compilation of activity of the working party on common reactor physics code systems during two years. Objectives of the working party is to clarify basic concept of common reactor physics code systems to improve convenience of reactor physics code systems for reactor physics researchers in Japan on their various field of research and development activities. We have held four meetings during 2 years, investigated status of reactor physics code systems and innovative software technologies, and discussed basic concept of common reactor physics code systems.

Nuclear Energy System Department annual report; April 1, 2002 - March 31, 2003

Department of Nuclear Energy System

JAERI-Review 2003-023, 232 Pages, 2003/09

JAERI-Review-2003-023.pdf:16.58MB

The Department has carried out researches and developments (R&Ds) of innovative nuclear energy system and their related fundamental technologies to ensure the long-term energy supply in Japan. The report deals with the R&Ds of an innovative water reactor, called Reduced-Moderation Water Reactor (RMWR), which has the capability of multiple recycling and breeding of plutonium using light water reactor technologies. In addition, as basic studies and fundamental researches of nuclear energy system in general, described are intensive researches in the fields of reactor physics, thermal-hydraulics, nuclear data, nuclear fuels, and materials. These activities are essential not only for the R&Ds of innovative nuclear energy systems but also for the improvement of safety and reliability of current nuclear energy systems. The maintenance and operation of reactor engineering facilities belonging to the Department support experimental activities.

Nuclear Energy System Department annual report

Department of Nuclear Energy System

JAERI-Review 2003-004, 236 Pages, 2003/03

JAERI-Review-2003-004.pdf:16.34MB

This report summarizes the research and development activities in the Department of Nuclear Energy System during the fiscal year of 2001 (April 1, 2001 - March 31, 2002). The Department has been organized from April 1998. The main research activity is aimed to build the basis of the development of future nuclear energy systems. The research activities of the Department cover basic nuclear data evaluation, conceptual design of a reduced-moderation water reactor, reactor physics experiments and development of the reactor analysis codes, experiment and analysis of thermal-hydrodynamics, energy system analysis and assessment, development of advanced materials for a reactor, lifetime reliability assessment on structural material, development of advanced nuclear fuel, design of a marine reactor and the research for a nuclear ship system. The maintenance and operation of reactor engineering facilities belonging to the Department are undertaken. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report.

Benchmark results of burn-up calculation for LWR next generation fuels

Okumura, Keisuke; Unesaki, Hironobu*; Kitada, Takanori*; Saji, Etsuro*

Proceedings of International Conference on the New Frontiers of Nuclear Technology; Reactor Physics, Safety and High-Performance Computing (PHYSOR 2002) (CD-ROM), 10 Pages, 2002/10

In order to investigate the calculation accuracy of the nuclear characteristics of LWR next generation fuels, the Research Committee on Reactor Physics organized by Japan Atomic Energy Research Institute has proposed "Reactor Physics Benchmark for LWR Next Generation Fuels". The next generation fuels aim at very high burn-up of about 70GWd/t in PWR or BWR with UO2 or MOX fuels whose fissile enrichments may exceed the Japanese regulatory limitations for the current LWR fuels such as 5wt.% U-235. Twelve organizations have carried out the analyses of the benchmark problems with different codes and data, and their submitted results have been compared. As a result, status of accuracy with the current data and method and some problems to be solved in the future were clarified.