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 ...

Irradiation test of plutonium rock-like fuel sample at JRR-3M

Yamada, Tadanori; ; Shirasu, Noriko

JAERI-Tech 97-072, 32 Pages, 1998/01

JAERI-Tech-97-072.pdf:1.22MB

no abstracts in English

Measurements of neutron fluxes and cadmium ratio at equilibrium core in JRR-3M

; Sasajima, Fumio; Ishida, Takuya*; *; *; Shigemoto, Masamitsu; Takahashi, Hidetake

JAERI-M 93-154, 46 Pages, 1993/08

JAERI-M-93-154.pdf:1.5MB

no abstracts in English

Measurement of high energy neutron-flux with foil-activation technique at a neutron-instrument of the J-PARC spallation-neutron-source

Harada, Masahide; Teshigawara, Makoto; Oi, Motoki; Oikawa, Kenichi; Takada, Hiroshi; Ikeda, Yujiro

no journal, , 

no abstracts in English

Benchmark analysis of FFTF unprotected loss of flow without scram test No.13 with fast reactor plant dynamics analysis code Super-COPD

Hamase, Erina; Ohgama, Kazuya; Kawamura, Takumi*; Doda, Norihiro; Yamano, Hidemasa; Tanaka, Masaaki

no journal, , 

Validation of an analysis model for a plant dynamic analysis code named Super-COPD including neutronics calculation of a one-point reactor kinetics model necessitates the further work on the beyond design basis accident. Therefore, JAEA participated in IAEA benchmark for Loss of Flow without Scram (LOFWOS) test No.13 performed at the Fast Flux Test Facility (FFTF), and the transient analysis at the first blind phase considering with major reactivity feedback mechanisms was carried out. It was observed that the whole plant dynamics analysis could follow the measured data. As a future work, the gap conductance model for transient, the upper plenum of reactor vessel with dividing several regions or multi-dimensional modeling, and the core model that can evaluate the radial heat transfer rate more accurately will be refined.