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Journal Articles

Electrochemical impedance analysis on solid electrolyte oxygen sensor with gas and liquid reference electrodes for liquid LBE

Adhi, P. M.*; Okubo, Nariaki; Komatsu, Atsushi; Kondo, Masatoshi*; Takahashi, Minoru*

Energy Procedia, 131, p.420 - 427, 2017/12

 Times Cited Count:0 Percentile:0.03

The ionic conductivity of solid electrolyte may insufficient, and the sensor output signal will deviate from the theoretical one in low temperature. The performance of oxygen sensor with Ag/air reference electrode (RE) and liquid Bi/Bi$$_{2}$$O$$_{3}$$ RE was tested in low-temperature LBE at 300$$sim$$450$$^{circ}$$C and the charge transfer reactions impedance at the electrode-electrolyte interface was analyzed by electrochemical impedance analysis (EIS). After steady state condition, both of the sensors performed well and can be used at 300$$sim$$450$$^{circ}$$C. Bi/Bi/Bi$$_{2}$$O$$_{3}$$ RE has lower impedance than Ag/air RE. Therefore, the response time of the oxygen sensor with Bi/Bi/Bi$$_{2}$$O$$_{3}$$ RE is faster than the oxygen sensor with Ag/air RE in the low-temperature region.

Journal Articles

Re-evaluation of assay data of spent nuclear fuel obtained at Japan Atomic Energy Research Institute for validation of burnup calculation code systems

Suyama, Kenya; Murazaki, Minoru*; Okubo, Kiyoshi; Nakahara, Yoshinori*; Uchiyama, Gunzo

Annals of Nuclear Energy, 38(5), p.930 - 941, 2011/05

 Times Cited Count:13 Percentile:74.14(Nuclear Science & Technology)

The isotopic composition of spent nuclear fuels is vital data for studies on the nuclear fuel cycle and reactor physics. The Japan Atomic Energy Agency (JAEA) has been active in obtaining such data for pressurized water reactor (PWR) and boiling water reactor (BWR) fuels, and some data has already been published. These data have been registered with the international Spent Fuel Isotopic Composition Database (SFCOMPO) and widely used as international benchmarks for burnup calculation codes and libraries. In this paper, postirradiation examination (PIE) data from two fuel assemblies irradiated in the Ohi-1 and Ohi-2 PWRs in Japan are shown. The PIE data from Ohi-2 has already been published. However, these data were not suitable for the benchmarking of calculation codes and libraries because several important specifications and data were not included. This paper summarizes the details of PIE data from Ohi-1 and Ohi-2 and shows in detail the data and specifications required for analyses of isotopic composition. For precise burnup analyses, the burnup values of PIE samples were re-evaluated in this study. These PIE data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. This indicates that the quality of PIE data from Ohi-1 and Ohi-2 PWRs is high, and that these PIE data are suitable for the benchmarking of burnup calculation code systems.

JAEA Reports

SWAT3.1; The Integrated burnup code system driving continuous energy Monte Carlo codes MVP and MCNP

Suyama, Kenya; Mochizuki, Hiroki*; Takada, Tomoyuki*; Ryufuku, Susumu*; Okuno, Hiroshi; Murazaki, Minoru; Okubo, Kiyoshi

JAEA-Data/Code 2009-002, 124 Pages, 2009/05

JAEA-Data-Code-2009-002.pdf:14.09MB

Integrated burnup calculation code system SWAT is a system that combines neutronics calculation code SRAC widely used in Japan and point burnup calculation code ORIGEN2. It has been used to evaluate the composition of the uranium, plutonium, minor actinide and the fission products in the spent nuclear fuel. Because of the ability to treat the arbitrary fuel geometry and no requirement of generating the effective cross section data, there is a great advantage to introduce continuous energy Monte Carlo Code into the burnup calculation code. Based on this idea, the integrated burnup calculation code system SWAT3.1 was developed by combining the continuous energy Monte Carlo code MVP and MCNP and ORIGEN2. This report describes the outline, input data instruction and several example of the calculation.

Journal Articles

Active reduction of the end effect by local installation of neutron absorbers

Suyama, Kenya; Murazaki, Minoru; Okubo, Kiyoshi; Okuno, Hiroshi

Annals of Nuclear Energy, 35(9), p.1628 - 1635, 2008/09

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

In the analysis of the burnup credit, it has been pointed out that the neutron multiplication factor becomes greater if we consider an axial burnup distribution of spent fuel assemblies than the case under an assumption of an average burnup through the fuel assemblies. This phenomenon is called "end effect" and it is one of the main technical issues in the burnup credit study. In this study, the reason why the end effect occurs in the criticality calculation of spent fuel assemblies is discussed by analyses of neutron flux distribution measurement both fixed source and eigenvalue calculations. These calculations show us that the end effect is induced by the solution of neutron balance equation as eigenvalue problem and an actual neutron flux increase occurs only when the neutron multiplication factor is close to unity. Based on the discussion, reducing the end effect actively by local installation of neutron absorbers (LINA) around the end regions of the fuel assemblies are proposed and its effect was confirmed based on the several criticality calculations.

JAEA Reports

Research and development on reduced-moderation light water reactor with passive safety features (Contract research)

Iwamura, Takamichi; Okubo, Tsutomu; Akie, Hiroshi; Kugo, Teruhiko; Yonomoto, Taisuke; Kureta, Masatoshi; Ishikawa, Nobuyuki; Nagaya, Yasunobu; Araya, Fumimasa; Okajima, Shigeaki; et al.

JAERI-Research 2004-008, 383 Pages, 2004/06

JAERI-Research-2004-008.pdf:21.49MB

The present report contains the achievement of "Research and Development on Reduced-Moderation Light Water Reactor with Passive Safety Features", which was performed by Japan Atomic Energy Research Institute (JAERI), Hitachi Ltd., Japan Atomic Power Company and Tokyo Institute of Technology in FY2000-2002 as the innovative and viable nuclear energy technology (IVNET) development project operated by the Institute of Applied Energy (IAE). In the present project, the reduced-moderation water reactor (RMWR) has been developed to ensure sustainable energy supply and to solve the recent problems of nuclear power and nuclear fuel cycle, such as economical competitiveness, effective use of plutonium and reduction of spent fuel storage. The RMWR can attain the favorable characteristics such as high burnup, long operation cycle, multiple recycling of plutonium (Pu) and effective utilization of uranium resources based on accumulated LWR technologies.

Journal Articles

Rise-to-power test of the HTTR (High Temperature Engineering Test Reactor)

Fujikawa, Seigo; Okubo, Minoru; Nakazawa, Toshio; Kawasaki, Kozo; Iyoku, Tatsuo

Nihon Genshiryoku Gakkai Wabun Rombunshi, 1(4), p.361 - 372, 2002/12

no abstracts in English

Journal Articles

Design of small Reduced-Moderation Water Reactor (RMWR) with natural circulation cooling

Okubo, Tsutomu; Suzuki, Motoe; Iwamura, Takamichi; Takeda, Renzo*; Moriya, Kumiaki*; Kanno, Minoru*

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

A small scale around 300 MWe reduced-moderation water reactor (RMWR) concept has been developed. For the core, a BWR type core concept with the tight-lattice fuel rod arrangement and the high void fraction is adopted to attain a high conversion ratio over 1.0. The negative void reactivity coefficients are also required, and the very flat short core concept is adopted to make the natural circulation cooling (NC) possible. The core burn-up of 60 GWd/t and the operation cycle of 24 months are also attained. For the system, simplification of the system with the passive safety features is a basic approach to overcome the scale demerit as well as the NC. For example, the HPCF system is replaced with the passive accumulator system resulting in the expensive emergency DGs reduction. The cost evaluation for concerned NSSS components gives about 20% reduction. Since MOX fuels in the RMWR contains Pu around 30 wt% and is irradiated to a high burn-up, the fuel safety evaluation has been performed and the acceptable results have been obtained from the thermal feasibility point of view.

Journal Articles

JAERI acquired a certificate of the pre-operation test for the High Temperature Engineering Test Reactor (HTTR)

Kawasaki, Kozo; Iyoku, Tatsuo; Nakazawa, Toshio; Okubo, Minoru; Baba, Osamu

Nihon Genshiryoku Gakkai-Shi, 44(4), P. 310, 2002/04

no abstracts in English

Journal Articles

High Temperature Engineering Test Reactor (HTTR) of JAERI attained the maximum reactor thermal power of 30MW

Kawasaki, Kozo; Iyoku, Tatsuo; Nakazawa, Toshio; Okubo, Minoru; Baba, Osamu

Nihon Genshiryoku Gakkai-Shi, 44(1), P. 2, 2002/01

no abstracts in English

Journal Articles

Core and system design of Reduced-Moderation Water Reactor with passive safety features

Iwamura, Takamichi; Okubo, Tsutomu; Yonomoto, Taisuke; Takeda, Renzo*; Moriya, Kumiaki*; Kanno, Minoru*

Proceedings of International Congress on Advanced Nuclear Power Plants (ICAPP) (CD-ROM), 8 Pages, 2002/00

Research and developments of reduced-moderation water reactor (RMWR) have been performed. The RMWR can attain the favorable characteristics such as high burn-up, long operation cycle, multiple recycling of plutonium and effective utilization of uranium resources, based on the matured LWR technologies. MOX fuel assemblies in the tight-lattice fuel rod arrangement are used to reduce the moderation of neutron, and hence, to increase the conversion ratio. The conceptual design has been accomplished for the small 330MWe RMWR core with the discharge burn-up of 60GWd/t and the operation cycle of 24 months, under the natural circulation cooling of the core. A breeding ratio of 1.01 and the negative void reactivity coefficient are simultaneously realized in the design. In the plant system design, the passive safety features are intended to be utilized mainly to improve the economy. At present, a hybrid one under the combination of the passive and the active components, and a fully passive one are proposed. The former has been evaluated to reduce the cost for the reactor components.

Journal Articles

Performance test of the HTTR

Tanaka, Toshiyuki; Okubo, Minoru; Iyoku, Tatsuo; Kunitomi, Kazuhiko; Takeda, Takeshi; Sakaba, Nariaki; Saito, Kenji

Nihon Genshiryoku Gakkai-Shi, 41(6), p.686 - 698, 1999/00

 Times Cited Count:4 Percentile:36.04(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Air vent in water chamber and surface coating on liner slides concerning auxiliary cooling system for the high temperature engineering test reactor

Takeda, Takeshi; Kunitomi, Kazuhiko; Okubo, Minoru; *

Nucl. Eng. Des., 185(2-3), p.229 - 240, 1998/00

 Times Cited Count:12 Percentile:70.06(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Present status of the High Temperature engineering Test Reactor (HTTR)

Tanaka, Toshiyuki; Okubo, Minoru; Fujikawa, Seigo; Mogi, Haruyoshi; Suzuki, Hiroshi

Proc. of PBNC'98, 2, p.1203 - 1210, 1998/00

no abstracts in English

Journal Articles

Present status of the High Temperature engineering Test Reactor(HTTR)

Tanaka, Toshiyuki; Shiozawa, Shusaku; Okubo, Minoru; Fujikawa, Seigo; Mogi, Haruyoshi; Suzuki, Hiroshi

Proceedings of European Nuclear Conference (ENC'98), 4, 5 Pages, 1998/00

no abstracts in English

JAEA Reports

Counter-measure to prevent temperature rise of stand pipe and primary upper shielding in HTTR

Kunitomi, Kazuhiko; Tachibana, Yukio; *; Nakano, Masaaki*; Saikusa, Akio; Takeda, Takeshi; Iyoku, Tatsuo; ; Sawahata, Hiroaki; Okubo, Minoru; et al.

JAERI-Tech 97-040, 91 Pages, 1997/09

JAERI-Tech-97-040.pdf:2.51MB

no abstracts in English

Journal Articles

Construction of the HTTR and its testing program

Tanaka, Toshiyuki; Baba, Osamu; ; Shiozawa, Shusaku; Okubo, Minoru

10th Pacific Basin Nuclear Conf. (10-PBNC), 1, p.811 - 818, 1996/10

no abstracts in English

Journal Articles

Construction of the HTTR and its testing program

Tanaka, Toshiyuki; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru;

JAERI-Conf 96-010, 0, p.97 - 104, 1996/07

no abstracts in English

Journal Articles

Air cooling design on stand pipe used for control rod of HTTR

Takeda, Takeshi; Kunitomi, Kazuhiko; Okubo, Minoru

Nihon Genshiryoku Gakkai-Shi, 38(4), p.307 - 314, 1996/00

no abstracts in English

Journal Articles

High accuracy heat transfer correlation on shell side of heat transfer tubes for pressurized water cooler in high temperature use

Kunitomi, Kazuhiko; Takeda, Takeshi; *; Okubo, Minoru; *; *; *

Nihon Genshiryoku Gakkai-Shi, 38(8), p.665 - 672, 1996/00

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

no abstracts in English

Journal Articles

Design and present status of high-temperature engineering test reactor

Baba, Osamu; Kunitomi, Kazuhiko; Kawaji, Satoshi; Tanaka, Toshiyuki; Shiozawa, Shusaku; Okubo, Minoru

Proc. of ASME$$cdot$$JSME 4th Int. Conf. on Nuclear Engineering 1996 (ICONE-4), 2, p.281 - 287, 1996/00

no abstracts in English

52 (Records 1-20 displayed on this page)