Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Mori, Tomoaki; Nishihara, Tetsuo; Kunitomi, Kazuhiko
Nihon Genshiryoku Gakkai Wabun Rombunshi, 6(3), p.253 - 261, 2007/09
Design studies of the hydrogen cogeneration high temperature gas cooled reactor (GTHTR300C), which can produce both electric power and hydrogen, have proceeded in Japan Atomic Energy Agency. In future, it will be obliged to operate using not enriched uranium but plutonium to coexist with fast reactors after the full deployment of fast reactor cycle. Therefore, a nuclear and thermal design has been performed to confirm the feasibility of the reactor core using Mixed-Oxide (MOX) fuel. The reactor core with operation period of 450 days and average burn-up of 123 GWd/ton for discharged fuel was designed. The reactor core met safety requirements of maximum fuel temperature of less than 1400 
C during normal operation, maximum power density of less than 13 W/cm
, shutdown margin of more than 1.0 %k/kk' and negative reactivity coefficient. The results proved that it is possible to operate the GTHTR300C using MOX fuels without consuming natural uranium resources.
Nishihara, Tetsuo; Mori, Tomoaki; Kunitomi, Kazuhiko
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 5 Pages, 2007/04
Japan Atomic Energy Agency (JAEA) has been carried out the design study of the HTGR hydrogen cogeneration system (GTHTR300C). GTHTR300C is 600MWth and the hydrogen production plant utilizes 370MW to supply 52,000m/h of hydrogen. Industrial hydrogen production capacity in Japan will decrease by 15 Bm/y in 2030. And the hydrogen demand for fuel cell vehicle in 2030 is estimated at 15Bm/y at a maximum. This hydrogen shortage is a potential market for the GTHTR300C. Hydrogen cost of the GTHTR300C is estimated at 20.5 JPY/m which has a economic conpetitiveness against other industrial hydorgen processes. 38 units of the GTHTR300C can supply a half of this shortage which accounts for the 33% of hydrogen demand for FCV in 2100.
Kunitomi, Kazuhiko; Yan, X.; Nishihara, Tetsuo; Sakaba, Nariaki; Mori, Tomoaki
Nuclear Engineering and Technology, 39(1), p.9 - 20, 2007/02
Design study on the Gas Turbine High Temperature Reactor 300-Cogeneration (GTHTR300C) aiming at producing both electricity by a gas turbine and hydrogen by a thermochemical water splitting method (IS process method) has been conducted. It is expected to be one of the most attractive systems to provide hydrogen for fuel cell vehicles after 2020s. The GTHTR300C employs a block type Very High Temperature Reactor (VHTR) with thermal power of 600MW and outlet coolant temperature of 950C. The maximum 370 MW to the secondary system is used for hydrogen production and the balance of the reactor thermal power is used for electricity generation. This paper describes the original design features focusing on the plant layout and plant cycle of the GTHTR300C together with present development status of the gas turbine, IHX, etc.
Takei, Masanobu*; Kosugiyama, Shinichi*; Mori, Tomoaki; Katanishi, Shoji; Kunitomi, Kazuhiko
Nihon Genshiryoku Gakkai Wabun Rombunshi, 5(2), p.109 - 117, 2006/06
no abstracts in English
Nojiri, Naoki; Fujimoto, Nozomu; Mori, Tomoaki; Obata, Hiroyuki*
JAERI-Data/Code 2004-012, 65 Pages, 2004/10
JAERI-Data-Code-2004-012.pdf:7.77MB
DELIGHT code is a fuel cell burnup analysis code which can produce the group constants necessary for High Temperature Gas-cooled Reactors (HTGR) core analyses. Collision probability method is used to the lattice calculation. The lattice calculation model is a cylinder type fuel or a ball type fuel of the HTGR. This code characterizes the burnup calculation considering the double heterogeneity caused by coated fuel particles of the HTGR fuel. DELIGHT code has updated its nuclear data library to the latest JENDL-3.3 data, and included new burnup chain models in order to calculate high burnup HTGR cores. The material regions of the periphery burnable poisons (BPs) were divided into details in order to improve calculation accuracy of the BP lattice calculation. This report presents the revised points of the DELIGHT-8 and can be used as user's manual.
Hazama, Taira; Mori, Tomoaki; ; Aihara, Nagafumi; ; Yoshida, Mamoru; *
JNC TN9400 2001-044, 136 Pages, 2001/05
JNC-TN9400-2001-044.pdf:3.97MB
A Subcriticality measurement technique was developed to improve safety and efficiency of criticality safety control in nuclear fuel processing facilities. In the development, two measurement techniques based on reactor noise analysis were selected as candidates of subcriticality measurement technique applicable to severe situations in FBR fuel reprocessing plants. The research activity was performed in Deuterium Critical Assembly (DCA) which was partly reconstructed from the original core of the advanced thermal reactor, so that light water and FBR type fuel could be used as in the FBR fuel reprocessing plants. Through the research, each technique was improved to satisfy criteria for subcriticality monitoring technique in FBR fuel reprocessing plant. Since the two techniques have basically different features while using common devices,thier combination would be a simple and reliable measurement system. This report summarizes processes and results of the research activity in DCA.
Mori, Tomoaki*; Takemura, Morio*
JNC TJ9450 2000-001, 96 Pages, 2000/03
JNC-TJ9450-2000-001.pdf:2.04MB
This report is intended to make it easier to apply the measured data obtained from the Special Materials Experiment, which was conducted at the Oak Ridge National Laboratory (ORNL) during about a month beginning at the end of June, 1992 as the last one of a series of eight experiments planned for the Japanese-American Shielding Program for Experimental Research (JASPER) which was started in 1986. For this reason. the information presented includes specifications and measurement data for all configurations, compositions of all materials, characteristics of the measurement system. and daily-basis records of measurements. The Special Materials Experiment was planned to obtain the data of neutron attenuation characteristics of selected shielding materials for use in advanced fast reactors. The material of particular interest for the experiment was zirconium hydride that is rich in hydrogen. The mockup slabs for the special materials were preceded by the spectrum modifier behind the TSR-II reactor of Tower Shielding Facility. The layer of zirconium hydride was simulated with a combination of zirconium and polyethylene slabs. The thick layer of polyethylene with no zirconium was installed in some configurations.Neutron flux was measured behind the configurations with various types of detectors. The integral neutron flux in wide energy region was measured in eight configurations and neutron spectrum in high energy region was measured also in almost all configurations. Information presented in this report is based mainly on a report issued by ORNL (ORNL/TM-12277. "Measurements for the JASPER Program Special Materials Experiment"). Additional information reported by the assignee is utilized also.
Mori, Tomoaki; Hazama, Taira
JNC TN9400 99-086, 77 Pages, 1999/12
The sub-criticality monitoring system has been developed for criticality safety management of nuclear fuel handling facilities. Low enriched MOX or uranium fuels have been used for sub-criticality measurements in the Deuterium Critical Assembly (DCA). This report describes the results of sub-criticality measurements for the core loaded with high enriched MOX fuels (JOYO MK-I fuels). The results of measurements showed that the prompt decay constant(
), which was an index of the sub-criticality, was detected between Keff=0.64 and Keff=0.82 and the difference of about 0.08
Keff could be detected. A simulation of sub-criticality measurements was performed with the Monte Calro code MCNP4A and the results were compared with measured values. The differences of values between calculations and measurements were less than 5%. It was confirmed that the sub-criticality measurements by Feynman- method could be simulated by MCNP4A code with enough accuracy. Feynman- method was applied to the DCA system as the neutron flux level was changed with time. It was confirmed that the common procedure of data processing had an enough ability to detect values for slow changes of neutron flux. On the other hand, the application of the differential filter was effective for fast changes of neutron flux.
Mori, Tomoaki;
Journal of Nuclear Science and Technology, 36(7), p.555 - 559, 1999/00
Times Cited Count：12 Percentile：66.07(Nuclear Science & Technology)None
Mori, Tomoaki;
PNC TN9410 98-056, 72 Pages, 1998/06
The sub-criticality monitoring system has been developed for criticality safety control in nuclear fuel handling plants. In the past experiments performed with the Deuterium Critical Assembly(DCA), it was confirmed that the detection of sub-criticality was possible to k
=0.3. To investigate the applicability of the method to more generalized system, experiments were performed in the light-water-moderated system of the modified DCA core. From these experiments, it was confirmed that the prompt decay constant(), which was a index of the sub-criticality, was detected between k=0.623 and k 0.870 and the difference of 0.05
0.1k could be distinguished. The values were numerically calculated with 2D transport code TWODANT and monte carlo code KENO V.a, and the results were compared with the measured values. The differences between calculated and measured values were proved to be less than 13%, which was sufficient accuracy in the sub-criticality monitoring system. It was confirmed that Feynman- method was applicable to sub-critical measurement of the light-water-moderated system.
Mori, Tomoaki;
PNC TN9450 98-003, 118 Pages, 1998/04
The nuclear characteristics of heavy-water-moderated cluster type lattice were systematically investigated by unit cen calculation with one-dimensional Sn transport code. The results of flux distribution, energy spectrum, and void reactivity were summarized for major parameters, such as fuel type, lattice pitch, coolant and moderator.
analysisMori, Tomoaki; Otani, Nobuo
PNC TN9410 97-095, 44 Pages, 1997/11
The Feynman- method has been developed for monitoring sub-criticality in nuclear fuel facilities. It is difficult to apply the Feynman- method which estimates statistical variation of the number of neutron counts per unit time, to the system in transient condition such that the averaged neutron flux varies with time. In the application of Feynman- method to such system, it is suggested to remove the averaged variation of neutron flux from neutron count data by the use of the difference filter. In this study, we applied the difference filter to reactor noise data at sub-criticality near to criticality, where the prompt decay constant was difficult to estimate due to the large effect of delayed neutron. With the difference filter, accurate prompt decay constants for effective multiplication factors from 0.999 to 0.994 were obtained by Feynman- method. It was cleared that the difference filter is effective to estimate accurate prompt decay constant, so that there is the prospect to be able to apply Feynman- method having the difference filter to the system in the transient condition.
Mori, Tomoaki
PNC TN9410 96-293, 101 Pages, 1996/11
In criticality Engineering Section, experiments for sub-criticality measurements by use with DCA (Deuterium Critical Assmbly) two-region core loaded with the test fuel assembly using MK-I fuels of JOYO are planed for the purpose of performing the study of sub-criticality measurements, Through this report, nuclear characteristics of DCA two-region core loaded with MK-I test fuels bave been understood with the satisfaction of DCA nuclear limits. And also, the composition of test fuel assembly in DCA core was decided from these results. SCALE4.2 code system including KENO-V with multi-group monte carlo method was used in order to calculate these nuclear characteristics. The estimated items of nuclear characteristics are critical heavy water height, heavy water level reactivity coefficient, heavy water dump reactivity and safety rod reactivity worth.
Mori, Tomoaki*; Takemura, Morio*
PNC TJ9055 96-002, 142 Pages, 1996/03
With use of a standard groupwise shielding design library JSSTDL produced from the latest evaluated nuclear data library JENDL-3.2, experimental analyses for the Radial Shield Attenuation Experiment (Configuration II; B
C, configuration V; Na+BC) and the Special Materials Experiment (Configuration III; Polyethylene) were performed. The results were compared with those obtained by the same analysis method and input data using JSDJ2 library that had been applied consistently to the JASPER experiment analyses. In general, the results with JSSTDL analyses are higher than those by JSDJ2. In comparison of the analysis for the configuration including fissionable layer, it was made clear that the cross section libraries with only one group structure in the thermal energy region may give inadequate results when the thermal neutron flux level is relatively high in the fissionable layer neighbouring to such material with big slowing-down effect as polyethylene. Also it was confirmed in the thick sodium configuration that the balanced mesh sizes between axial and radial directions are important for accurate analysis. After a brief review of the state of the JASPER experiment analyses up to this time, important configurations for experiment reanalyses were selected and items of such data as input for the reanalyses were also listed up. The data for reanalyses of some of the selected configurations were arranged in computer files.
Mori, Tomoaki*; Takemura, Morio*
PNC TJ9055 96-001, 96 Pages, 1996/03
This report is intended to make it easier to apply the measured data obtained from the Special Materials Experiment, which was conducted at the Oak Ridge National Laboratory (ORNL) during about a month beginning at the end of June, 1992 as the last one of a series of eight experiments planned for the Japanese-American Shielding Program for Experimental Research (JASPER) which was started in 1986. For this reason. the information presented includes specifications and measurement data for all configurations, compositions of all materials, characteristics of the measurement system. and daily-basis records of measurements. The Special Materials Experiment was planned to obtain the data of neutron attenuation characteristics of selected shielding materials for use in advanced fast reactors. The material of particular interest for the experiment was zirconium hydride that is rich in hydrogen. The mockup slabs for the special materials were preceded by the spectrum modifier behind the TSR-II reactor of Tower Shielding Facility. The layer of zirconium hydride was simulated with a combination of zirconium and polyethylene slabs. The thick layer of polyethylene with no zirconium was installed in some configurations. Neutron flux was measured behind the configurations with various types of detectors. The integral neutron flux in wide energy region was measured in eight configurations and neutron spectrum in high energy region was measured also in almost all configurations. Information presented in this report is based mainly on a report issued by ORNL (ORNL/TM-12277. "Measurements for the JASPER Program Special Materials Experiment"). Additional information reported by the assignee is utilized also.
Shindo, Ryuichi; Murata, Isao; Sawa, Kazuhiro; Shiozawa, Shusaku; *; Mori, Tomoaki*
Proc. of the 8th Int. Conf. on Radiation Shielding, p.351 - 358, 1994/00
no abstracts in English
Nakano, Masaaki*; Takada, Eiji*; Mori, Tomoaki; Kunitomi, Kazuhiko
no journal, ,
no abstracts in English
日野 竜太郎; 寺田 敦彦; 上地 優; 西畑 保雄
毛利 智聡*; 平田 慎吾*; 五十嵐 実*; 佐藤 学*
JP, 2014-205637 Patent licensing information Patent publication (In Japanese)
【課題】放射性廃棄物貯槽等のようにその内部空間領域が狭く且つ気密性が求められる設備への設置に適した水素と酸素との触媒式の再結合装置の提供。 【解決手段】容器50に取り付けて容器内部で発生した水素を触媒を用いて酸素と結合するための再結合装置1であって、容器50から流入した水素および酸素を含むガスがその内部を流通する第一の筒状体2と、第一の筒状体2の出口に連結され、第一の筒状体2から流入したガスを容器50に戻す戻し部3とを備え、第一の筒状体2が、入口側に水素と酸素とを結合する触媒部9を有している。
日野 竜太郎; 寺田 敦彦; 上地 優; 西畑 保雄
毛利 智聡*; 平田 慎吾*; 五十嵐 実*; 佐藤 学*
JP, 2014-205635 Patent licensing information Patent publication (In Japanese)
【課題】装置全体の小型化を図ることができる触媒式の再結合装置を提供する。 【解決手段】水素と酸素を含む処理対象ガスを導入して水素と酸素を再結合する触媒部3と、触媒部3により処理された後の処理対象ガスがその内部を流通する筒状部2とを備える。
