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Journal Articles

Molecular dynamics analysis of reactor graphite for preparing thermal neutron scattering law

Okita, Shoichiro; Goto, Minoru

Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 10 Pages, 2023/10

Journal Articles

Thermal-neutron capture cross-section measurements of neptunium-237 with graphite thermal column in KUR

Nakamura, Shoji; Shibahara, Yuji*; Endo, Shunsuke; Kimura, Atsushi

Journal of Nuclear Science and Technology, 59(11), p.1388 - 1398, 2022/11

https://doi.org/10.1080/00223131.2022.2058639

Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)

The present study selected $^{237}$ Np among radioactive nuclides and aimed to measure the thermal-neutron capture cross-section for $^{237}$ Np in a well-thermalized neutron field by an activation method. A $^{237}$ Np standard solution was used for irradiation samples. A thermal-neutron flux at an irradiation position was measured with neutron flux monitors: $^{45}$ Sc, $^{59}$ Co, $^{98}$ Mo, $^{181}$ Ta and $^{197}$ Au. The $^{237}$ Np sample and flux monitors were irradiated together for 30 minutes in the graphite thermal column equipped with the Kyoto University Research Reactor. The similar irradiation was carried out twice. After the irradiations, the $^{237}$ Np samples were quantified using 312-keV gamma ray emitted from $^{233}$ Pa in a radiation equilibrium with $^{237}$ Np. The reaction rates of $^{237}$ Np were obtained from gamma-ray peak net counts given by $^{238}$ Np, and then the thermal-neutron capture cross-section of $^{237}$ Np was found to be 173.84.4 barn by averaging the results obtained by the two irradiations. The present result was in agreement with the reported data given by a time-of-flight method within the limit of uncertainty.

Journal Articles

Measurements of thermal-neutron capture cross-section of the $^{237}$ Np(n, ) reaction with TC-Pn in KUR

Nakamura, Shoji; Endo, Shunsuke; Kimura, Atsushi; Shibahara, Yuji*

KURNS Progress Report 2021, P. 93, 2022/07

https://www.rri.kyoto-u.ac.jp/PUB/report/PR/ProgRep2021/ProgRep2021.html

In terms of nuclear transmutation studies of minor actinides in nuclear wastes, the present work selected $^{237}$ Np among them and aimed to measure the thermal-neutron capture cross-section of $^{237}$ Np using a well-thermalized neutron field by a neutron activation method because there have been discrepancies among reported cross-section data. A $^{237}$ Np standard solution was used for irradiation samples. The thermal-neutron flux at an irradiation position was measured with flux monitors: $^{45}$ Sc, $^{59}$ Co, $^{98}$ Mo, $^{181}$ Ta and $^{197}$ Au. The $^{237}$ Np sample was irradiated together with the flux monitors for 30 minutes in the graphite thermal column equipped in the Kyoto University Research Reactor. The similar irradiation was repeated once more to confirm the reproducibility of the results. After irradiation, the $^{237}$ Np samples were quantified using 312-keV gamma-ray emitted from $^{233}$ Pa in radiation equilibrium with $^{237}$ Np. The reaction rates of $^{237}$ Np were obtained from the peak net counts of gamma-rays emitted from generated $^{238}$ Np, and then the thermal-neutron capture cross-section of $^{237}$ Np was found to be 173.84.7 barn by averaging the results obtained by the two irradiations. The present result was in agreement with the reported data given by a time-of-flight method within a limit of uncertainty.

JAEA Reports

Separation of $^{14}$ C from irradiated graphite materials, 1; Oxidation behaviors and the changes in pore structure of Q1 and IG-110 graphite due to the air reaction (Joint research)

Fujii, Kimio

JAERI-Tech 2005-048, 108 Pages, 2005/09

JAERI-Tech-2005-048.pdf:25.05MB

The graphite-moderated power reactor was shut down in 1998 and its decommissioning program is being planned. Various graphites are used in the core of magnox-type reactors and HTTR as core-support structural materials and moderating materials of fast neutrons. For the nuclear graphite disposal, it is necessary to determine especially the treatment of long-lived nuclides, such as $^{14}$ C which are generated in the graphite components during reactor operation. As a research, which solves the problem of the $^{14}$ C concentration, the cooperative research is concluded between JAERI and Japan Nuclear Power Corp. in 1999, and the research for the basic data acquisition has been advanced up to the present. To find the optimum conditions for $^{14}$ C reduction, basic data on oxidation reaction and the structure of graphite materials are indispensable. In the present experiment, we measure the air oxidation characteristics in the temperature range 450800 $^{circ}$ C in Quality1 graphite and IG-110 graphite. Changes in pore diameter and pore size distribution due to air oxidation are discussed.

JAEA Reports

Research and development plan for advanced high temperature gas cooled reactor fuels and graphite components (Contract research)

Sawa, Kazuhiro; Ueta, Shohei; Shibata, Taiju; Sumita, Junya; Ohashi, Jumpei; Tochio, Daisuke

JAERI-Tech 2005-024, 34 Pages, 2005/03

JAERI-Tech-2005-024.pdf:2.15MB

The Very-High-Temperature Reactor (VHTR) is one of the strong candidates for the Generation IV Nuclear Energy System. JAERI has developed Zirconium carbide (ZrC)-coated fuel particle and ZrC coating layer is expected to maintain its intactness under higher temperature and burn-up comparing conventional SiC-coating layer. JAERI carries out (1) ZrC-coating process development by large-scale coater, (2) inspection method development and (3) irradiation test and post irradiation experiment of ZrC coated particles. Also, JAERI carries out reactivity insertion tests to clarify the coating failure mechanism and tries to increase allowable temperature limit in case of reactivity insertion accident. Furthermore, JAERI develops non-destructive evaluation methods for mechanical properties of graphite components by ultrasonic testing and micro-indentation technique. This report describes these research and development plan and results of FY 2004 as a MEXT contact research.

Journal Articles

Reactor internals design

Sumita, Junya; Ishihara, Masahiro; Nakagawa, Shigeaki; Kikuchi, Takayuki; Iyoku, Tatsuo

Nuclear Engineering and Design, 233(1-3), p.81 - 88, 2004/10

https://doi.org/10.1016/j.nucengdes.2004.07.013

Times Cited Count：4 Percentile：29.18(Nuclear Science & Technology)

A High Temperature Gas-cooled Reactor is particularly attractive due to its capability of producing high temperature helium gas and its possibility to exploit inherent safety characteristic. To achieve high temperature helium-gas, reactor internals are made of graphite and heat resistant materials, its surroundings are composed of metals. The reactor internals of the HTTR consist of graphite and metallic core support structures and shielding blocks. This paper describes the reactor internal design of the HTTR, especially the core support graphite structures, and the program of an in-service inspection.

Journal Articles

Heat removal performance of auxiliary cooling system for the High Temperature Engineering Test Reactor during scrams

Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki*

Annals of Nuclear Energy, 30(7), p.811 - 830, 2003/05

https://doi.org/10.1016/S0306-4549(02)00119-6

Times Cited Count：1 Percentile：10.87(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Carbon-14 in neutron-irradiated graphite for graphite-moderated reactors (Joint research)

Fujii, Kimio; Matsuo, Hideto*

JAERI-Review 2002-034, 44 Pages, 2002/12

JAERI-Review-2002-034.pdf:3.34MB

Graphite material is used as a moderator and reflector of graphite-moderated gas-cooled reactors such as the Tokai Nuclear Power Station of Japan Atomic Power Company. For the decommissioning of those reactors, it is very important to obtain the density of carbon-14 in the graphite of which half-time is very long, and the development of the reduction technique for density of carbon-14 is one of the main technical issues as well. The reactor ceased the commercial operation on March 1998, and is in the process of the decommissioning. Technical developement on nuclear graphite disposal is one of the main issues, and it is pointed out that carbon-14 density is an impotant factor for the treatment on the disposal of the nuclear graphite. In order to resolve the problem, literatures survey was carried out for describing the status of technical development on resolving the problem. This report describes the outline of the literatures surveyed and the list of the related literatures is attached as well.

Journal Articles