Training using JMTR and related facilities in FY2021 and FY2022

Nakano, Hiroko; Fujinami, Kyoko; Yamaura, Takayuki; Kawakami, Jun; Hanakawa, Hiroki

JAEA-Review 2023-036, 33 Pages, 2024/03

JAEA-Review-2023-036.pdf:2.47MB

A practical training course using the JMTR (Japan Materials Testing Reactor) and other research infrastructures was held from November 29 to December 2 in 2021 for Asian young researchers and engineers. This course was adopted as International Youth Exchange Program in Science (SAKURA SCIENCE Exchange Program) which is the project of the Japan Science and Technology Agency, and this course aims to enlarge the number of high-level nuclear researchers/engineers in Asian countries which are planning to introduce a nuclear power plant, and to promote the use of facilities in future. In this year, from the viewpoint of preventing the spread of COVID-19 infection, it was decided to hold the event online. 53 young researchers and engineers joined the course from 6 countries. In FY2022, training programs with invitations were held due to the easing of restrictions on entry into Japan from overseas. 7 young researchers and engineers from4 Asian countries participated in the training from February 1 to 10, 2023.The common curriculum in the training course of FY2021 and FY2022 included lectures on nuclear energy, irradiation testing, safety management, JMTR decommissioning plan, etc. In the online session, conducted in FY2021, information exchange on the energy situation in each country was conducted. On-site training conducted in FY2022, included practical training on operation using simulations, environmental monitoring, etc. and facility tours of the JMTR, etc. Many participants could join the online training course, they created a diversity of expertise and made lively discussions during the information exchange. On-site training, while limited in number of participants, provided a good opportunity for personnel exchange through practical training and face-face communication. It is desirable to hold on-site training as long as circumstances permit. This report summarizes the training conducted in FY2021 and FY2022.

Development of neutronics and thermal-hydraulics coupled analysis method on platform for design optimization in fast reactor

Doda, Norihiro; Hamase, Erina; Yokoyama, Kenji; Tanaka, Masaaki

Keisan Kogaku Koenkai Rombunshu (CD-ROM), 25, 4 Pages, 2020/06

With the aim of advancing the design optimization in fast reactors, neutronics and thermal-hydraulics coupled analysis method which can consider the temporal change of neutron flux distribution in the core has been developed. A three-dimensional neutronics analysis code and a plant dynamics analysis code are coupled on a platform using Python programing. In this report, outlines of the coupling method of analysis codes, the results of its application to the actual plant under a virtual accidental condition, and the future development is described.

Study on nuclear analysis method for high temperature gas-cooled reactor and its nuclear design (Thesis)

Goto, Minoru

JAEA-Review 2014-058, 103 Pages, 2015/03

JAEA-Review-2014-058.pdf:22.36MB

The following issues were investigated using experimental data of HTTR, which is a Japan's HTGR with 30 MW thermal power. (1)Applicability of nuclear data libraries to nuclear analysis for HTGR, (2) Applicability of the improved nuclear analysis method for HTGR, (3) Effectiveness of a rod-type burnable poison on HTGR reactivity control. Using these investigation results, a nuclear design of a small-sized HTGR with 50 MW thermal power (HTR50S) was performed. In the nuclear design of HTR50S, we challenged to decrease the number of the fuel enrichments and to increase the power density compared with HTTR. As a result, the nuclear design was completed successfully by reducing the number of the fuel enrichment to only three from twelve of HTTR and increasing the power density by 1.4 times of HTTR.

Achievements of element technology development for breeding blanket

Department of Fusion Engineering Research; Department of Materials Science

JAERI-Review 2005-012, 143 Pages, 2005/03

JAERI-Review-2005-012.pdf:11.74MB

no abstracts in English

Evaluation of neutronic characteristics of STACY 80-cm-diameter cylindrical core fueled with 6% enriched uranyl nitrate solution

Yanagisawa, Hiroshi; Sono, Hiroki

JAERI-Tech 2003-057, 39 Pages, 2003/06

JAERI-Tech-2003-057.pdf:2.76MB

no abstracts in English

A Neutronics and burnup analysis of the accelerator-driven transmutation system with different cross section libraries

Sasa, Toshinobu; Tsujimoto, Kazufumi; Kaneko, Kunio*; Takano, Hideki

Journal of Nuclear Science and Technology, 39(Suppl.2), p.1183 - 1186, 2002/08

http://www.tandfonline.com/doi/abs/10.1080/00223131.2002.10875313

no abstracts in English

Introduction of research group; Fusion Neutronics Laboratory, Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute

Nishitani, Takeo; Kutsukake, Chuzo; Hori, Junichi

Purazuma, Kaku Yugo Gakkai-Shi, 77(6), p.609 - 610, 2001/06

http://ci.nii.ac.jp/naid/110003825647

no abstracts in English

Neutronic study of DRX (Deep Sea Reactor X) core for deep sea research vessel

Odano, Naoteru; Ishida, Toshihisa

JAERI-Research 2001-004, 36 Pages, 2001/03

JAERI-Research-2001-004.pdf:1.51MB

no abstracts in English

ComparaUve analyses on nuclear charaderistics of water-cooled breeder cores

; Sato, Wakaei*;

JNC TN9400 2000-037, 87 Pages, 2000/03

JNC-TN9400-2000-037.pdf:3.48MB

ln order to compare the nuclear characteristics of water-cooled bleeder cores with that of LMFBR, MOX fuel cell models are established for boiling and non-boiling LWR, non-boiling HWR and sodium-cooled reactor. Frst, the comarison is made between the heterogeneous cell calculation results by SRAC and those by SLAROM. The results show some differences as for neutron energy spectrum, one-grouped cross section and conversion ratio due to the different grouped cross section library (both are based on JENDL-3.2, though) used for each code, however, the difference is acceptably small for grasping the basic characteristics of the above-mentioned cores. Second, using the SLAROM code, main core parameters such as mean neutron energy, ratio of fast neutron and -value, are analyzed. The comparison between the cores show that softened neutron spectrum by the scattering effect of hydrogen or heavy hydrogen increase the contribution of nuclear reaction (especially for neutron capture reaction rather than fission reaction) in lower energy region comparing with LMFBR. ln order to overcome the effect, tighter lattice than LMFBR is necessary for water-cooled cores to realize the breeding of fissile nuclides. Third, effects of Pu isotopic composition on the breeding ratio are evaluated using SRAC burnup calculation. From the results, it is confirmed that degraded Pu (larger ratio of Pu-240) show the larger breeding ratio. At last, sensitivity analyses are made for k-effective and main reaction ratios. As for k-effective, using a temporary covariance data of JENDL-3.2, uncertainty resulting from the cross sections' error is analyzed for a boiling LWR and a sodium-cooled reactor. The boiling LWR core shows larger sensitivity in lower energy region than the sodium-cooled reactor (especially for the energy region lower than 1kev), And, 18-group analysis that is considered acceptably good for LMFBR analysis, should not be enough for accurate sensitivity estimation of ...

Preparation of next generation set of group cross sections; A Task report to the Japan Nuclear Cycle Development Institute)

*

JNC TJ9400 2000-005, 182 Pages, 2000/03

JNC-TJ9400-2000-005.pdf:4.74MB

The SLAROM code, performing fast reactor cell calculation based on a deterministic methodology, has been revised by adding the universal module PEACO of generating Ultra-fine group neutron spectra. The revised SLAROM, then, was utilized for evaluating reaction rate distributions in ZPPR-13A simulated by a 2-dim RZ homogeneous model, although actually ZPPR-13A composed of radial heterogereous cells. The reaction rate distributions of ZPPR-13A were also calculated by the code MVP, that is a continuous energy Monte Carlo calculation code based on a probabilistic methodology. By coparing both results, it was concluded that the module PEACO has excellent capability for evaluating highly accurate effective cross sections. Also it was proved that the use of a new fine group cross section library set (next generation set), reflecting behavior of cross sections of structural materials, such as Fe and O, in the fast neutron energy region, is indispensable for attaining a better agreement within 1% between both calculation methods. Also, for production of a next generation set of group cross sections, the code NJOY97.V107 was added to the group cross section production system and both front and end processing parts were prepared. This system was utilized to produce the new 70 group JFS-3 library using the evaluated nuclear data library JENDL-3.2. Furthermore, to confirm the capability of this new group cross section production system, the above new JFS-3 library was applied to core performance analysis of ZPPR-9 core with a 2-dim RZ homogeneous model and analysis of heterogeneous cells of ZPPR-9 core by using the deterministic method. Also the analysis using the code MVP was performed. Bycoaparison of both results the following conclusion has been derived; the deterministic method, with the PEACO module for resonance cross sections, contributes to improve accuracy of predicting reaction rate distributions and Na void reactivity in fast reactor cores. And it ...