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JNC TN8410 2000-011, 185 Pages, 2000/05
JNC-TN8410-2000-011.pdf:4.67MB
This report describes the neutronic design calculational methods used in Fuel Design and Evaluation Group in order to inform other related sections of FBR core analysis technology and hand down the technology. Especially we show the neutronics calculation procedures used for the conceptual design study of the advanced core with 127 pin bundle for MONJU that has been carried out in our group. The topics include effective cross section preparation calculations, two-dimensional depletion calculations, three-dimensional diffusion calculations, reactivity coefficient calculations, and control rod worth calculations. The calculational methods shown in this report are the standard neutronics calculation methods employed in our group at the moment. However, the improvement of calculation codes, the reduction of correction factors and uncertainties for design using the nuclear data obtained in the start-up test for MONJU and so on, and the update of nuclear data file will be planned in order to improve evaluation accuracies. Those may change the neutronic design calculational methods, but we decided to describe the present standard calculational methods in our group from the viewpoint of sharing information in JNC.
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JNC TJ9400 2000-004, 109 Pages, 2000/02
JNC-TJ9400-2000-004.pdf:4.96MB
We estimated covariances of the JENDL-3.2 data on the nuclides and reactions needed to analyze fast-reactor cores for the past three years, and produced covariance files. The present work was undertaken to re-examine the covariance files and to make some improvements. The covariances improved are the ones for the inelastic scattering cross section of 
O, the total cross section of 
Na, the fission cross section of 
U, the capture cross section of 
U, and the resolved resonance parameters for U. Moreover, the covariances of 
U data were newly estimated by the present work. The covariances obtained were compiled in the ENDF-6 format.
Ijima, Takashi; ; Matsumoto, Mitsuo; *
JNC TN3410 2000-002, 93 Pages, 2000/01
JNC-TN3410-2000-002.pdf:2.54MB
Fugen Nuclear Power Station ("Fugen") is a prototype Advanced Thermal Reactor (ATR), it has been demonstrated the plutonium utilization by loading many Mixed Oxide Fuels (MOX) since the reactor start up March 1979, and no fuel defect had been occurred, The MOX fuel assemblies has the high reliability and has been loaded more than 700 fuel assemblies. This is the largest in the world as a thermal neutron reactor. However, "Fugen" is planning to stop its operation in the year 2003, because the role of the Fugen almost finished. Therefore, we are going to summarize the ATR project including the Plutonium utilization experience. This paper is summarized as part of the experience.
Nagaya, Yasunobu
no journal, ,
We review Monte Carlo capabilities developed recently and Monte Carlo applications to reactor analysis and reactor design. Finally, we describe future prospects of the Monte Carlo methodological development and the applicability of the Monte Carlo methods to reactor design.
Sugawara, Takanori; Kunieda, Satoshi
no journal, ,
no abstracts in English