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Miyahara, Naoya; Miwa, Shuhei; Horiguchi, Naoki; Sato, Isamu*; Osaka, Masahiko
Journal of Nuclear Science and Technology, 56(2), p.228 - 240, 2019/02
Times Cited Count：7 Percentile：61.18(Nuclear Science & Technology)In order to improve LWR source term under severe accident conditions, the first version of a fission product (FP) chemistry database named "ECUME" was developed. The ECUME is intended to include major chemical reactions and their effective kinetic constants for representative SA sequences. It is expected that the ECUME can serve as a fundamental basis from which FP chemical models in the SA analysis codes can be elaborated. The implemented chemical reactions in the first version were those for representative gas species in Cs-I-B-Mo-O-H system. The chemical reaction kinetic constants were evaluated from either literature data or calculated values using ab-initio calculations. The sample chemical reaction calculation using the presently constructed dataset showed meaningful kinetics effects at 1000 K. Comparison of the chemical equilibrium compositions by using the dataset with those by chemical equilibrium calculations has shown rather good consistency for the representative Cs-I-B-Mo-O-H species. From these results, it was concluded that the present dataset should be useful to evaluate FP chemistry in Cs-I-B-Mo-O-H system under LWA SA conditions.
Ohgama, Kazuya; Ikeda, Kazumi*; Ishikawa, Makoto; Kan, Taro*; Maruyama, Shuhei; Yokoyama, Kenji; Sugino, Kazuteru; Nagaya, Yasunobu; Oki, Shigeo
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
Ohgama, Kazuya; Nakano, Yoshihiro; Oki, Shigeo
Journal of Nuclear Science and Technology, 53(8), p.1155 - 1163, 2016/08
Times Cited Count：1 Percentile：10.6(Nuclear Science & Technology)The power distribution and core characteristics in various configurations of fuel subassemblies with an innerduct structure in the Japan Sodium-cooled Fast Reactor were evaluated using a Monte Carlo code for neutron transport and burnup calculation. The correlation between the fraction of fuel subassemblies facing outward and the degree of power increase at the core center was observed regardless of the compositions. This indicated that the spatial fissile distribution caused by innerduct configurations was the major factor of the difference in the power distribution. A power increase was also found in an off-center region, and it tended to be greater than that at the core center because of the steep gradient of neutron flux intensity. The differences in the worth of control rods caused by innerduct configurations were confirmed.
Hayashi, Takao; Tobita, Kenji; Nishio, Satoshi; Ikeda, Kazuki*; Nakamori, Yuko*; Orimo, Shinichi*; DEMO Plant Design Team
Fusion Engineering and Design, 81(8-14), p.1285 - 1290, 2006/02
Times Cited Count：22 Percentile：79.98(Nuclear Science & Technology)Neutron transport calculations were carried out to evaluate the capability of metal hydrides and borohydrides as an advanced shielding material. Some hydrides indicated considerably higher hydrogen content than polyethylene and solid hydrogen. The hydrogen-rich hydrides show superior neutron shielding capability to the conventional materials. From the temperature dependence of dissociation pressure, ZrH
and TiH can be used without releasing hydrogen at the temperature of less than 640 
C at 1 atm. ZrH and Mg(BH
) can reduce the thickness of the shield by 30% and 20% compared to a combination of steel and water, respectively. Mixing some hydrides with F82H produces considerable effects in 
-ray shielding. The neutron and -ray shielding capabilities decrease in order of ZrH 
Mg(BH) and F82H TiH and F82H water and F82H.
Kishimoto, Katsumi; Arigane, Kenji*
JAERI-Tech 2005-016, 83 Pages, 2005/03
JAERI-Tech-2005-016.pdf:10.52MB
Revaluation to activation activity of reactor evaluated at the notification of dismantling submitted in 1997 was carried out in JRR-2 where decommissioning was advanced now. In the revaluation, estimation accuracy on neutron streaming at various horizontal experimental tubes was improved by applying 3 dimensional model to neutron transport calculation that had been carried out by 2 dimensional model, and calculating with TORT. As the result, excessive overestimations on horizontal experimental tubes and biological shield that had greatly contributed to total activation activity in evaluation at the notification of dismantling was revised, sum of their activation activities in the revaluation decreased to 1/18(case after 1 year from the permanent shutdown of reactor) of evaluation at the notification of dismantling, and the structural materials that had large activation activity were changed. By the above, it was shown that introducing 3 dimensional model was effective in evaluation on activation activity of the research reactor that had a lot of various experimental tubes.
Iijima, Susumu*; Kato, Yuichi*; Takasaki, Kenichi*; Okajima, Shigeaki
JAERI-Data/Code 2004-016, 91 Pages, 2004/12
JAERI-Data-Code-2004-016.pdf:7.45MB
The calculation code system "EXPARAM" was designed to analyze the experimental results systematically measured at the fast critical assembly (FCA). Some calculation codes developed independently in JAERI and in US research institutes were collected and arranged as the fast reactor physics calculation code system. The multi-group core calculation code and the perturbation calculation code based on the diffusion theory and the transport theory calculate the reactor physics parameters such as eigenvalue, reaction rate, Doppler reactivity worth and sodium void worth. The dynamic physics parameters such as prompt neutron lifetime and effective delayed neutron fraction are also calculated. Input and Output data of calculation codes are transferred to each other using a direct access file on UNIX computer system.
Matsubayashi, Masahito; Hibiki, Takashi*; Mishima, Kaichiro*; Yoshii, Koji*; Okamoto, Koji*
Nuclear Instruments and Methods in Physics Research A, 533(3), p.481 - 490, 2004/11
Times Cited Count：5 Percentile：30.85(Instruments & Instrumentation)The validity of a fast neutron radiography quantification method, the 
-scaling method, which was originally proposed for thermal neutron radiography was examined with Monte Carlo calculations and experiments conducted at the YAYOI fast neutron source reactor. Water and copper were selected as comparative samples for a thermal neutron radiography case and a dense object, respectively. Although different characteristics on effective macroscopic cross-sections were implied by the simulation, the -scaled experimental results with the fission neutron spectrum cross-sections were well fitted to the measurements for both the water and copper samples. This indicates that the -scaling method could be successfully adopted for quantitative measurements in fast neutron radiography.
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.
Kitada, Takanori*; Okumura, Keisuke; Unesaki, Hironobu*; Saji, Etsuro*
Proceedings of International Conference on Physics of Fuel Cycles and Advanced Nuclear Systems; Global Developments (PHYSOR 2004) (CD-ROM), 8 Pages, 2004/04
Burnup calculation benchmark has been carried out for the LWR next generation fuels aiming at high burnup up to 70 GWd/t with UO and MOX. Based on the submitted results by many benchmark participants, the present status of calculation accuracy has been confirmed for reactor physics parameters of the LWR next generation fuels, and the factors causing the calculation differences were analyzed in detail. Moreover, the future experiments and research subjects necessary to reduce the calculation differences were discussed and proposed.
Research Committee on Reactor Physics
JAERI-Research 2004-004, 409 Pages, 2004/03
JAERI-Research-2004-004.pdf:28.53MB
This report summarizes the second phase (FY2001-2002) activity of "the Working Party (WP) on Reactor Physics for LWR Next Generation Fuels". The next generation fuels mean the ones aiming at further extended burn-up such as 70GWd/t over the current design. In the WP, the benchmark activity has been conducted to investigate and improve the calculation accuracy of the nuclear characteristics of the next generation fuels. In the second phase activity, all benchmark results were compiled and compared. Based on the comparison, the present status of calculation accuracy for the next generation fuels has been confirmed, and the factors causing the calculation differences were analyzed in detail. Moreover, analyses of the post irradiation and critical experiments with the codes used in the benchmark were reviewed, and future experiments and research subjects necessary to reduce the calculation differences were discussed and proposed.
Sato, Satoshi
JAERI-Research 2003-014, 223 Pages, 2003/09
JAERI-Research-2003-014.pdf:10.94MB
no abstracts in English
Unesaki, Hironobu*; Okumura, Keisuke; Kitada, Takanori*; Saji, Etsuro*
Transactions of the American Nuclear Society, 88, p.436 - 438, 2003/06
In order to investigate the calculation accuracy of the nuclear characteristics of LWR next generation fuels, the Research Committee on Reactor Physics organized by JAERI 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 UO or MOX fuels whose fissile enrichments may exceed the Japanese regulatory limitations for the current LWR fuels such as 5wt.% U-235. Until now, twelve organizations have pariticipated in the benchmark activity. From the comparison with the cell burn-up calculation results using different codes and library data, status of the calculation accuracy and future subjects are clarified.
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.
Research Committee on Reactor Physics
JAERI-Review 2001-047, 180 Pages, 2002/02
JAERI-Review-2001-047.pdf:10.03MB
Under the Research Committee on Reactor Physics, the Working Party on Reactor Physics of Accelerator-Driven System (ADS-WP) was set in July 1999 to review and investigate special subjects related to reactor physics research for the Accelerator-Driven Subcritical System (ADS).The ADS-WP, at the first meeting, discussed a task guideline of its activity for two years and decided to concentrate upon three subjects: (1) neutron transport calculations in high energy range, (2) static and kinetic (safety-related) characteristics of subcritical system, and (3) system design including ADS concepts and elemental technology developments required.The activity of ADS-WP continued from July 1999 to March 2001. In this duration, the members of ADS-WP met together four times and discussed the above subjects. In addition, the ADS-WP conducted a questionnaire on requests and proposals for the plan of Transmutation Physics Experimental Facility in the High-Intensity Proton Accelerator Project, which is a joint project between JAERI and KEK (High Energy Accelerator Research Organization).This report summarizes the results obtained by the above ADS-WP activity. The report will be useful to overview those results and moreover to set up a new guideline of future research activity in this field.
Sukegawa, Takenori; Hatakeyama, Mutsuo; Yanagihara, Satoshi
JAERI-Tech 2001-058, 81 Pages, 2001/09
JAERI-Tech-2001-058.pdf:5.98MB
In general, neutron transport and activation calculation codes are used for residual radioactive inventory estimation; however, it is essential to verify calculations by measurement results because of geometrical complexity of the reactor and so on. The comparison between measured and calculated radioactivity in the JPDR core components showed a relatively good agreement (factor of 2), and it was cleared that water content and weight ratio of steel bars to concrete materials significantly influenced the neutron flux distribution in the biological shield (factor of 2-10 error). The measured radioactivity inside of the reactor pressure vessel wall and at the inner part of the biological shield was compared in detail with the calculations to verify the methodology applied to calculations of radioisotope production. Then it was found that the radioactive inventory could be estimated accurately with combination of calculations and measurement of radioactivity in samples and dose rate distribution for planning of dismantling activities.
T.V.Hung*; Sakamoto, Yukio; Yasuda, Hideshi
JAERI-Research 98-057, 25 Pages, 1998/10
JAERI-Research-98-057.pdf:1.04MB
no abstracts in English
T.V.Hung*
JAERI-Research 98-037, 14 Pages, 1998/07
JAERI-Research-98-037.pdf:0.67MB
no abstracts in English
*; Kaji, Yoshiyuki; *; ; *; Kaburaki, Hideo
Keisan Kogaku Koenkai Rombunshu, 3(1), p.59 - 62, 1998/05
no abstracts in English
Nakano, Yoshihiro; Soyama, Kazuhiko; *
JAERI-Conf 96-008, 0, p.86 - 91, 1996/03
no abstracts in English
*; *; *; *; Kugo, Teruhiko;
PHYSOR 96: Int. Conf. on the Physics of Reactors, 1, p.B92 - B101, 1996/00
no abstracts in English
