Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 67

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

JAEA Reports

Development of the versatile reactor analysis code system, MARBLE3

Yokoyama, Kenji; Hazama, Taira; Taninaka, Hiroshi; Oki, Shigeo

JAEA-Data/Code 2024-007, 41 Pages, 2024/10

JAEA-Data-Code-2024-007.pdf:1.1MB

The third version of the versatile reactor analysis code system, MARBLE3, has been developed. In the development of the former version of MARBLE, object-oriented scripting language Python (Python2) had been used and then the latest version of Python (Python3) was released. However, due to its backward incompatibility, MARBLE no longer worked with Python3. For this reason, MARBLE3 has been fully modified and maintained to work with Python3. In MARBLE3, newly developed analysis codes and newly proposed calculation methods were incorporated, and the user interface was extended and solvers were reimplemented for maintainability, extensibility, and flexibility. In MARBLE3, the three-dimensional hexagonal/triangular transport code MINISTRI Ver.7 (MINISTRI7) and the three-dimensional hexagonal/triangular diffusion code D-MINISTRI are available as the new analysis codes. These codes can be used in the neutronics analysis system SCHEME and the fast reactor burnup analysis system OPRHEUS, which are the subsystems of MARBLE. In addition, the user interface of CBG, a core analysis system embedded in MARBLE, was extended so that the diffusion and transport calculation solvers for the 2-dimensional RZ system of CBG can be used on SCHEME. On the other hand, MARBLE3 has extended the functionality of the burnup calculation solver so that it can use the numerical methods proposed in the papers on the improvement of the Chebyshev rational function approximation method and the minimax polynomial approximation method. From the viewpoint of maintainability, the point reactor kinetics solver POINTKINETICS, which was introduced in MARBLE2, has been newly reworked as the KINETICS solver in MARBLE3.

Journal Articles

The Development of a Multiphysics Coupled Solver for Studying the Effect of Dynamic Heterogeneous Configuration on Particulate Debris Bed Criticality and Cooling Characteristics

Li, C.-Y.; Wang, K.*; Uchibori, Akihiro; Okano, Yasushi; Pellegrini, M.*; Erkan, N.*; Takata, Takashi*; Okamoto, Koji*

Applied Sciences (Internet), 13(13), p.7705_1 - 7705_29, 2023/07

 Times Cited Count:1 Percentile:33.61(Chemistry, Multidisciplinary)

JAEA Reports

Investigation of the core neutronics analysis conditions for evaluation of burn-up nuclear characteristics of the next-generation fast reactors

Takino, Kazuo; Oki, Shigeo

JAEA-Data/Code 2023-003, 26 Pages, 2023/05

JAEA-Data-Code-2023-003.pdf:1.66MB

Since next-generation fast reactors aim to achieve a higher core discharge burn-up than conventional reactors do, core neutronics design methods must be refined. Therefore, a suitable analysis condition is required for the analysis of burn-up nuclear characteristics to accomplish sufficient estimation accuracy while maintaining a low computational cost. We investigated the effect of the analysis conditions on the accuracy of estimation of the burn-up nuclear characteristics of next-generation fast reactors in terms of neutron energy groups, neutron transport theory, and spatial mesh. This study treated the following burn-up nuclear characteristics: criticality, burn-up reactivity, control rod worth, breeding ratio, assembly-wise power distribution, maximum linear heat rate, sodium void reactivity, and Doppler coefficient for the equilibrium operation cycle. As a result, it was found that the following conditions were the most suitable: 18-energy-group structure, 6 spatial meshes per assembly with diffusion approximation. Additionally, these conditions should apply to correction factors for energy group structure, spatial mesh and transport effects.

Journal Articles

Evaluation of fuel reactivity worth measurement in the prototype fast reactor Monju

Ohgama, Kazuya; Takegoshi, Atsushi*; Katagiri, Hiroki; Hazama, Taira

Nuclear Technology, 208(10), p.1619 - 1633, 2022/10

 Times Cited Count:4 Percentile:63.92(Nuclear Science & Technology)

Journal Articles

Integral experiment of $$^{129}$$I(n, $$gamma$$) using fast neutron source in the "YAYOI" reactor

Nakamura, Shoji; Toh, Yosuke; Kimura, Atsushi; Hatsukawa, Yuichi*; Harada, Hideo

Journal of Nuclear Science and Technology, 59(7), p.851 - 865, 2022/07

 Times Cited Count:1 Percentile:12.48(Nuclear Science & Technology)

The present study performed integral experiments of $$^{129}$$I using a fast-neutron source reactor "YAYOI" of the University of Tokyo to validate evaluated nuclear data libraries. The iodine-129 sample and flux monitors were irradiated by fast neutrons in the Glory hole of the YAYOI reactor. Reaction rates of $$^{129}$$I were obtained by measurement of decay gamma-rays emitted from $$^{130}$$I. The validity of the fast-neutron flux spectrum in the Glory hole was confirmed by the ${it C/E}$ ratios of the reaction rates of flux monitors. The experimental reaction rate of $$^{129}$$I was compared with that calculated with both the fast-neutron flux spectrum and evaluated nuclear data libraries. The present study revealed that the evaluated nuclear data of $$^{129}$$I cited in JENDL-4.0 should be reduced as much as 18% in neutron energies ranging from 10 keV to 3 MeV, and supported the reported data by Noguere ${it et al.}$ below 100 keV.

JAEA Reports

Development of the unified cross-section set ADJ2017R

Yokoyama, Kenji; Maruyama, Shuhei; Taninaka, Hiroshi; Oki, Shigeo

JAEA-Data/Code 2021-019, 115 Pages, 2022/03

JAEA-Data-Code-2021-019.pdf:6.21MB
JAEA-Data-Code-2021-019-appendix(CD-ROM).zip:435.94MB

In JAEA, several versions of unified cross-section set for fast reactors have been developed so far; we have developed a new unified cross-section set ADJ2017R, which is an improved version of the unified cross-section setADJ2017 for fast reactors. The unified cross-section set is used for reflecting information of C/E values (analysis / experiment values) obtained by integral experiment analyses in reactor core design via the cross-section adjustment methodology; the values are stored in the standard database for FBR core design. In the methodology, the cross-section set is adjusted by integrating the information such as uncertainty (covariance) of nuclear data, uncertainty of integral experiment / analysis, sensitivity of integral experiment with respect to nuclear data. ADJ2017R basically has the same performance as ADJ2017, but we conducted an additional investigation on ADJ2017 and revised the following two points. The first is to unify the evaluation method of the correlation coefficient of uncertainty caused by experiments (hereinafter referred to as the experimental correlation coefficient). Because it was found that the common uncertainty used in the evaluation of the experimental correlation coefficient was evaluated by two different methods, the experimental correlation coefficients were revised for all experimental data, and the evaluation method was unified. The second is the review of the integral experiment data used for the cross-section adjustment calculation. It was found that one of the experimental values of composition ratio after irradiation of the Am-243 sample has a problem in uncertainty evaluation because its experimental uncertainty is extremely small compared to the others. The cross-section adjustment calculation was, therefore, redone by excluding the experimental value. In the creation of ADJ2017, a total of 719 data sets were analyzed and evaluated, and eventually adopted 620 integral experimental data sets. In contrast, a total of 61

Journal Articles

Evaluation of fixed absorber reactivity measurement in the prototype fast reactor Monju

Ohgama, Kazuya; Katagiri, Hiroki; Takegoshi, Atsushi*; Hazama, Taira

Nuclear Technology, 207(12), p.1810 - 1820, 2021/12

 Times Cited Count:4 Percentile:47.47(Nuclear Science & Technology)

Journal Articles

Irradiation growth behavior of improved Zr-based alloys for fuel cladding

Amaya, Masaki; Kakiuchi, Kazuo; Mihara, Takeshi

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1048 - 1056, 2019/09

Journal Articles

Study of the neutron multiplication effect in an active neutron method

Komeda, Masao; Ozu, Akira; Mori, Takamasa; Nakatsuka, Yoshiaki; Maeda, Makoto; Kureta, Masatoshi; Toh, Yosuke

Journal of Nuclear Science and Technology, 54(11), p.1233 - 1239, 2017/11

 Times Cited Count:9 Percentile:58.69(Nuclear Science & Technology)

The previous active neutron method cannot remove the influence of the multiplication effect of neutrons produced by second- and subsequent fission reactions, and it might overestimate the amount of nuclear material if an item contains large amounts. In this paper, we discussed the correction method for the neutron multiplication effect on the measured data in the fast neutron direct interrogation (FNDI) method, one of the active neutron methods, supposing that the neutron multiplication effect is caused mainly by third-generation neutrons from the second-fission reactions under the condition that the forth-generation neutrons are much fewer. This paper proposed a correction method for the neutron multiplication effect in the measured data. Moreover we have shown a possibility that this correction method gives rough estimates of the effective neutron multiplication factor and the subcriticality.

Journal Articles

Model verification and validation procedure for a neutronics design methodology of next generation fast reactors

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

Journal Articles

Design study of a 750 MWe Japan sodium-cooled fast reactor with metal fuel

Ohgama, Kazuya; Ota, Hirokazu*; Ikusawa, Yoshihisa; Oki, Shigeo; Ogata, Takanari*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04

Journal Articles

Tradeoff analysis of metal-fueled fast reactor design concepts

Stauff, N. E.*; Ohgama, Kazuya; Aliberti, G.*; Oki, Shigeo; Kim, T. K.*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

Journal Articles

An Improved fast neutron radiography quantitative measurement method

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:36.07(Instruments & Instrumentation)

The validity of a fast neutron radiography quantification method, the $$Sigma$$-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 $$Sigma$$-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 $$Sigma$$-scaling method could be successfully adopted for quantitative measurements in fast neutron radiography.

Journal Articles

Development of a fast neutron radiography converter using wavelength-shifting fibers

Matsubayashi, Masahito; Hibiki, Takashi*; Mishima, Kaichiro*; Yoshii, Koji*; Okamoto, Koji*

Nuclear Instruments and Methods in Physics Research A, 510(3), p.325 - 333, 2003/09

 Times Cited Count:9 Percentile:52.81(Instruments & Instrumentation)

A fluorescent converter for fast neutron radiography (FNR) comprising a scintillator and hydrogen-rich resin has been developed and applied to electronic imaging. The rate of the reaction between fast neutrons and the converter is increased by thickening the converter, but its opaqueness attenuates emitted light photons before they reach its surface. To improve the luminosity of a fluorescent converter for FNR, a novel type of converter was designed in which wavelength-shifting fibers were adopted to transport radiated light to the observation end face. The performance of the converter was compared with that of a polypropylene-based fluorescent converter in an experiment conducted at the fast-neutron-source reactor YAYOI in the University of Tokyo.

Journal Articles

Doppler effect measurement on resonance materials for rock-like oxide fuels in an intermediate neutron spectrum

Ando, Masaki; Nakano, Yoshihiro; Okajima, Shigeaki; Kawasaki, Kenji

Journal of Nuclear Materials, 319, p.126 - 130, 2003/06

 Times Cited Count:0 Percentile:0.00(Materials Science, Multidisciplinary)

Doppler effect experiments on resonance materials for ROX fuels were carried out to examine the calculation accuracy in the intermediate neutron spectrum using Fast Critical Assembly (FCA) at Japan Atomic Energy Research Institute. This study is the second phase of a series of the Doppler effect experiments on the resonance materials, which is following the measurements in the fast neutron spectrum. The Doppler effect was measured as the sample reactivity change between the heated and unheated samples. The cylindrical samples of the resonance materials such as erbium (Er), tungsten (W) and thorium (ThO$$_{2}$$) were used. The sample was heated up to 800$$^{circ}$$C at the center of the FCA core. The Doppler effect measurements were analyzed using the SRAC 95 code system with the use of JENDL 3.2. The calculated values agreed with the experiment within the experimental error for the W and ThO$$_{2}$$ samples, while the calculation overestimated the experiment for the Er sample about 10 %.

Journal Articles

An Application to Intraoperative BNCT using epithermal neutron of new JRR-4 mode "Epi-12"

Matsushita, Akira*; Yamamoto, Tetsuya*; Matsumura, Akira*; Nose, Tadao*; Yamamoto, Kazuyoshi; Kumada, Hiroaki; Torii, Yoshiya; Kashimura, Takanori*; Otake, Shinichi*

Research and Development in Neutron Capture Therapy, p.141 - 143, 2002/09

A thermal-epithermal mixed beam "Thermal Neutron Beam Mode I" was used in the eleven sessions of boron neutron capture therapy which have been performed at JRR-4 from 1998. We are planning to use an epithermal beam for the treatment of deeper tumors in the next trial of the intraoperative BNCT. In this study, "Epi-12" which was made by putting up a cadmium shutter of "Thermal Neutron Beam Mode I" was investigated for the clinical benefits and safety by epithermal beams. Decrease of fast neutron contamination ratio in Epi-12 mode is the advantage for BNCT, particular in the intraoperative BNCT. Because fast neutron on the brain surface is one of the critical factors in the intraoperative BNCT in which the plain beam directly interacts the normal structures. Furthermore a mixture of mode Epi-12 and Th-12 will provide various dose distribution designs. It may be used as a new method to control the best distribution for individual tumors.

JAEA Reports

Acceleration irradiation test of first-loading fuel of High Temperature Engineering Test Reactor (HTTR) up to high burnup (Joint research)

Sawa, Kazuhiro; Sumita, Junya; Ueta, Shohei; Takahashi, Masashi; Tobita, Tsutomu*; Hayashi, Kimio; Saito, Takashi; Suzuki, Shuichi*; Yoshimuta, Shigeharu*; Kato, Shigeru*

JAERI-Research 2002-012, 39 Pages, 2002/06

JAERI-Research-2002-012.pdf:3.12MB

no abstracts in English

Journal Articles

RAPID-L highly automated fast reactor concept without any control rods, 2; Critical experiment of lithium-6 used in LEM and LIM

Tsunoda, Hirokazu*; Sato, Osamu*; Okajima, Shigeaki; Yamane, Tsuyoshi; Iijima, Susumu; Kobe, Mitsuru*

Proceedings of International Congress on Advanced Nuclear Power Plants (ICAPP) (CD-ROM), 6 Pages, 2002/00

no abstracts in English

Journal Articles

Preliminary examination of the applicability of imaging plates to fast neutron radiography

Matsubayashi, Masahito; Hibiki, Takashi*; Mishima, Kaichiro*; Yoshii, Koji*; Okamoto, Koji*

Nuclear Instruments and Methods in Physics Research A, 463(1-2), p.324 - 330, 2001/05

 Times Cited Count:17 Percentile:74.43(Instruments & Instrumentation)

no abstracts in English

Journal Articles

A Note on the diven factor in fast systems

Okajima, Shigeaki; Yamane, Yoshihiro*; Takemoto, Yoshinari*; Sakurai, Takeshi

Journal of Nuclear Science and Technology, 37(8), p.720 - 723, 2000/08

no abstracts in English

67 (Records 1-20 displayed on this page)