Study on improvement of reactor physics analysis method for FBRs with various core concept

*; Kitada, Takanori*; Tagawa, Akihiro; *; Takeda, Toshikazu*

JNC TJ9400 2000-006, 272 Pages, 2000/02

JNC-TJ9400-2000-006.pdf:9.69MB

Investigation was made on the follwing three themes as a part of the improvement of reactor physics analysis method for FBR with various core concept. Part 1: Investigation of Error Estimation of Neutron Spectra in FBR and Suggestions to Improve the Accuracy. In order to improve the spectrum unfolding method used in fast experimental reactor JOYO, a trial was made to evaluate the error in the estimated neutron spectrum, cause by cause. And the evaluated errors were summed up to obtain the most probable and reasonable error as possible. The summed up error was found relatively small compared to the error caused by the uncertainty of cross section data: most of the error in the spectrum unfolding method can be attributed to the error in cross sections. It was also found that the error due to the fission spectrum causes a considerable error in the high energy neutron spectrum which is over several MeV. Part 2: Study on Reactor Physics Analysis Method for Gas-Cooled FBR. In gas-cooled FBR, the portion of coolant channels in core volume is larger than sodium-cooled FBR. This leads to strong neutron streaming effects. For sodium-cooled FBR, several methods were proposed to evaluate the neutron streaming effect, however, these methods can not be used directly to gas-cooled reactor because the direction dependent diffusion coefficient becomes infinitive along the direction pararel to the coolant chammel. In this study, a new method is proposed to evaluate the neutron streaming effect, based on the method taking the axial buckling into consideration, which method was originally proposed by Khler. Part 3: Study on Reactor Physics Analysis Method for Water-Cooled FBR An investigation was made on low-moderated water-cooled FBR, on the point that the ordinary used analysis method for FBR may give considerable difference in results in such core. In light water reactors, it is well known that the space dependence of self-shielding effect of heavy nuclides are considerably ...

Precise determination of for water-moderated U and U-Pu cores by a method using buckling coefficient of reactivity

Suzaki, Takenori; Sakurai, Kiyoshi; Nakajima, Ken; Horiki, Oichiro*

Proceedings of 6th International Conference on Nuclear Criticality Safety (ICNC '99), 1, p.386 - 394, 1999/00

no abstracts in English

Boundary element analysis of the geometric buckling of the solution fuel with a slant surface in a rectangular tank (contract research)

Yamane, Yuichi; Miyoshi, Yoshinori; *

JAERI-Tech 97-039, 16 Pages, 1997/08

JAERI-Tech-97-039.pdf:0.68MB

no abstracts in English

Boundary element analysis of the geometric buckling of the liquid fuel with a slant surface in a cylindrical tank

Yamane, Yuichi; ; *

JAERI-Tech 96-018, 14 Pages, 1996/05

JAERI-Tech-96-018.pdf:0.6MB

no abstracts in English

Evaluation of critical bucklings of light-water moderated low enriched UO cores by the variable loading method

Nakajima, Ken;

Journal of Nuclear Science and Technology, 30(11), p.1175 - 1179, 1993/11

https://doi.org/10.3327/jnst.30.1175

no abstracts in English

Geometric buckling expression for regular polygons, I; Measurements in low-enriched UO-HO lattices

Miyoshi, Yoshinori; ; ; Hirose, Hideyuki;

Nuclear Technology, 103, p.380 - 391, 1993/09

https://doi.org/10.13182/NT93-A34858

no abstracts in English

Evaluation of effect of solution-density formula on criticality parameters

Miyoshi, Yoshinori; Hirose, Hideyuki; ; Sakurai, Satoshi

JAERI-M 93-031, 32 Pages, 1993/03

JAERI-M-93-031.pdf:0.98MB

no abstracts in English

Measurement of infinite multiplication factor by using central cell reactivity worth on FCA XIV-1 core simulating high conversion light water reactor

Sakurai, Takeshi; Okajima, Shigeaki; Osugi, Toshitaka

JAERI-M 91-014, 25 Pages, 1991/02

JAERI-M-91-014.pdf:0.61MB

no abstracts in English

Evoluation of Two-Demensional Numerical Calculation Used in Fast Reactor Neutronics

; ;

JAERI-M 5307, 52 Pages, 1973/06

JAERI-M-5307.pdf:1.38MB

no abstracts in English