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Nakamura, Shoji; Shibahara, Yuji*; Endo, Shunsuke; Kimura, Atsushi
Journal of Nuclear Science and Technology, 58(10), p.1061 - 1070, 2021/10
Times Cited Count：5 Percentile：64.12(Nuclear Science & Technology)In a well-thermalized neutron field, it is principally possible to drive a thermal-neutron capture cross-section without considering an epithermal neutron component. This was demonstrated by a neutron activation method using the graphite thermal column (TC-Pn) of the Kyoto University Research Reactor. First, in order to confirm that the graphite thermal column was a well-thermalized neutron field, neutron irradiation was performed with neutron flux monitors: 
Au, 
Co, 
Sc, 
Cu, and 
Mo. The TC-Pn was confirmed to be extremely thermalized on the basis of Westcott's convention, because the thermal-neutron flux component took a constant value regardless of the sensitivity of each flux monitor to epithermal neutrons. Next, as a demonstration, the thermal-neutron capture cross section of 
Ta(n,
)
Ta reaction was measured using the graphite thermal column, and then derived to be 20.5
0.4 barn, which supported the evaluated value of 20.40.3 barn. The Ta nuclide could be useful as a flux monitor that complements the sensitivity between Au and Mo monitors.
Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.666 - 677, 2019/08
An evaluation methodology of critical heat fluxes (CHFs) based on a mechanism for fuel assemblies in light water reactors (LWRs) is needed in order to design and evaluate the safety for the fuel assemblies in LWRs. In our study, the numerical simulation with surface-tracking will be applied for the two-phase flow in fuel assemblies in order to obtain the detail data relating to the size and velocity of bubbles in the subchannel, which is needed to predict the CHF based on the mechanism. In this study, the numerical simulation of two-phase flow in 4
4 bundle was implemented by using JUPITER in order to establish the evaluation method of the size and velocity of bubbles by the numerical simulation, which is the multi-physics simulation code and enable to track the gas-liquid surface. The simulation results are validated by the curve of flow regime for air-water under the adiabatic condition. The bubble and velocity of bubbles obtained by simulation results are analyzed.
Maeda, Makoto; Komeda, Masao; Tobita, Hiroshi; Ozu, Akira; Kureta, Masatoshi; Bogucarska, T.*; Crochemore, J. M.*; Varasano, G.*; Pedersen, B.*
Dai-37-Kai Kaku Busshitsu Kanri Gakkai Nihon Shibu Nenji Taikai Rombunshu (CD-ROM), 7 Pages, 2017/02
JAEA and EC/JRC are carrying out collaborative research to develop NDA techniques that can be utilized for quantification of high radioactive special nuclear materials such as spent fuel and next generation minor actinide fuels. In the research, reliability of neutron transport codes is important because it is utilized for design and development of a demonstration system of next-generation Differential Die-away (DDA) technique in JAEA. In order to evaluate the reliability, actual neutron flux distribution in a sample cavity was examined in PUNITA device using JRC type DDA technique and JAWAS-T device using JAEA type DDA technique, and then the measurement results were compared with the simulation results obtained by the neutron transport codes. The neutron flux distribution in the target matrix was also examined in the PUNITA and compared with the simulation results. We report on the measurement and simulation results of the neutron flux distribution and evaluation results of the reliability of the neutron transport codes.
Maeda, Makoto; Komeda, Masao; Tobita, Hiroshi; Ozu, Akira; Kureta, Masatoshi; Bogucarska, T.*; Crochemore, J. M.*; Varasano, G.*; Pedersen, B.*
Proceedings of INMM 57th Annual Meeting (Internet), 9 Pages, 2016/07
The JAEA and EC/JRC have started collaborative research to develop a technique that can be utilized for quantification of high radioactive special nuclear materials such as next generation minor actinide fuels. In the study of a Differential Die-Away (DDA) technique, which is one of the techniques to be improved in the collaborative research, JRC type and JAEA type DDA techniques are compared. In the JRC type DDA technique, large amount of thermal neutron is generated using D-T neutron generator and graphite moderator to accomplish high detection sensitivity for small amount of fissile material. On the other hand, in JAEA type, relatively hard neutron spectrum and moderation of neutron in the target matrix are utilized to minimize position dependence of detection efficiency. Estimation of the neutron field is important to evaluate the performance of the system in DDA technique. The purpose of this study is to validate simulation results by experimental results and evaluate neutron flux distribution in the system by the simulation and the experiment. In this paper, we present the evaluation results of the neutron flux distributions in PUNITA which utilizes JRC type DDA technique and JAWAS-T which utilizes JAEA type DDA technique obtained by Monte Carlo simulation and activation method.
Kobayashi, Noboru; Okubo, Tsutomu; Uchikawa, Sadao
JAERI-Review 2005-029, 119 Pages, 2005/09
JAERI-Review-2005-029.pdf:11.01MB
The research on Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed in JAERI for the development of future innovative reactors. The workshop on the FLWRs has been held every year since 1998 aiming at information exchange between JAERI and other organizations. The 8th workshop was held on Feb. 10, 2005 under the joint auspices of JAERI and North Kanto and Kanto-Koetsu branches of Atomic Energy Society of Japan with 75 participants. The workshop began with 3 presentations on FLWRs entitled "Framework and Status of Research and Development on FLWRs", "Long-Term Fuel Cycle Scenarios for Advanced Utilization of Plutonium from LWRs", and "Experiments on Characteristics on Hydrodynamics in Tight-Lattice Core". Then 3 lectures followed: "Development of Evaluation Method for Accuracy in Predicting Neutronics Characteristics of Tight-Lattice Core" by Osaka University, "Development of Cost-Reduced Low-Moderation Spectrum Boiling Water Reactor" by Toshiba Corporation and "Design and Analysis on Super-Critical Water Cooled Power Reactors" by Tokyo University.
SiMatsuda, Tatsuma; Haga, Yoshinori; Ikeda, Shugo; Galatanu, A.; Yamamoto, Etsuji; Shishido, Hiroaki*; Yamada, Mineko*; Yamaura, Junichi*; Hedo, Masato*; Uwatoko, Yoshiya*; et al.
Journal of the Physical Society of Japan, 74(5), p.1552 - 1556, 2005/05
Times Cited Count：24 Percentile：72.12(Physics, Multidisciplinary)We have succeeded in growing a high-quality single crystal of UCuSi with the tetragonal structure by the Sn-flux method and measured the electrical resistivity, magnetic susceptibility, magnetization and specific heat. UCuSi is found to order antiferromagnetically below 
= 106 K, and follows a successive ferromagnetic ordering at 
= 100 K. The magnetic properties are highly anisotropic, reflecting the crystal structure. An easy-axis of magnetization is found to be the [001] direction (
-axis) both in the antiferromagnetic and ferromagnetic phases, while the [100] direction (
-axis) corresponds to the hard-axis in magnetization. The magnetization curve in the antiferromagnetic phase indicates a clear metamagnetic transition at a low field of about 1 kOe and changes smoothly into a ferromagnetic magnetization curve below = 100 K. The saturation moment is determined as 1.75
/U at 2 K. The electronic specific heat coefficient is also determined as 20 mJ/K
mol.
*; Kunugi, Tomoaki
Nihon Kikai Gakkai Rombunshu, B, 64(617), p.65 - 70, 1998/01
no abstracts in English
Ando, Masaki; *; Nishina, Kojiro*; *
Journal of Nuclear Science and Technology, 34(5), p.445 - 453, 1997/05
Times Cited Count：6 Percentile：47.9(Nuclear Science & Technology)no abstracts in English
Fujimura, Toichiro
JAERI-Data/Code 95-019, 54 Pages, 1996/01
JAERI-Data-Code-95-019.pdf:1.47MB
no abstracts in English
Yamashita, Kiyonobu; Murata, Isao; Shindo, Ryuichi; *; H.Werner*
Journal of Nuclear Science and Technology, 31(5), p.470 - 478, 1994/05
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
Obara, Toru*; Nakajima, Ken; *; Sekimoto, Hiroshi*; Suzaki, Takenori
JAERI-M 94-004, 40 Pages, 1994/02
no abstracts in English
Suzuki, Tomoo; ; Tanaka, Shunichi;
JAERI 1327, 110 Pages, 1992/05
no abstracts in English
Miura, Yukitoshi; Okano, Fuminori
Review of Scientific Instruments, 61(11), p.3581 - 3585, 1990/11
Times Cited Count：12 Percentile：75.58(Instruments & Instrumentation)no abstracts in English
*; ;
JAERI-M 84-101, 46 Pages, 1984/06
no abstracts in English
*;
Nuclear Science and Engineering, 83, p.333 - 349, 1983/00
Times Cited Count：4 Percentile：50.97(Nuclear Science & Technology)no abstracts in English
H(d,n)
He neutron source by total absorption method using water bath; ; ;
Journal of Nuclear Science and Technology, 12(8), p.491 - 501, 1975/08
Times Cited Count：0no abstracts in English
Ogawa, Hiroaki; Kitazawa, Sin-iti; Sugie, Tatsuo; Katsunuma, Atsushi*; Kitazawa, Daisuke*; Omori, Keisuke*; Itami, Kiyoshi
no journal, ,
no abstracts in English
4 simulated fuel bundleOno, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
no journal, ,
An evaluation methodology of critical heat fluxes (CHFs) based on a mechanism for fuel assemblies in light water reactors (LWRs) is needed in order to design and evaluate the safety for the fuel assemblies in LWRs. JAEA is embarking on the establishment of methodology to simulate the two-phase flow in the fuel bundles which applies to prediction CHF. In this study, the numerical simulation of two-phase flow in 44 bundle was implemented by using JUPITER in order to establish the evaluation method of the size and velocity of bubbles by the numerical simulation, which is the multi-physics simulation code and enable to track the gas-liquid surface. It was confirmed that the JUPITER enables to simulate the two-phase flow in the 44 simulated fuel bundle which is large calculation domain over 3 m.
