Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
*
JNC TJ9440 99-010, 42 Pages, 1999/03
This investigation on comprehensive-evaluation-methods for nuclear fuel cycle has been performed through open-literature search. As the results, no proper comprehensive-evaluation-method has been found which integrate several factors to be considered into only one factor. In the evaluation of future advanced nuclear energy systems, it is required to evaluate from both view points of natural resources and natural environment, in addition to the other factors such as safety, economy, and proliferation resistance. It is recommended that clarification of specific items or targets to be evaluated is most important as the first thing to be done. Second, methodology for the evaluation should be discussed.
*; *; *
JNC TJ9410 2001-001, 180 Pages, 1999/03
JNC-TJ9410-2001-001.pdf:5.82MB
The iventoly of the fission products (FPs) retained in the fuel of the demonstration Fast Breeder Reactor (FBR) will be increased due to the high burn-up capability with large-scale core size. In the commercialized FBR, the amount of FPs accumulated in the core system will be more increased because of the development of multiple utilization of FBRs for the miner actinide (MA) burning and incinerations of long-lived FPs(LLFP). In order to minimize risks due to core disruptive accidents (CDAs) and to perform effective accident managements, it is required to evaluate the distribution of FPs released from the degraded core to the primary coolant system during the CDAs. The preparation and development of the accident analysis codes, describing the FP transfer mechanisms eliminating the excess conservatisms, are thus very important. The major objective of this work is to prepare and to develop the FP transfer analysis code applicable to the primary coolant system of the demonstration and commercialized FBRs, TRACER code, based on the current experimental knowledge of the source term and on the systematic analyses of the real plant under the CDA condition. In this study, the FP tansfer analysis using TRACER code was applied to the disrupted core state and primary system of the typical demonstration FBR, where the power transient of the core was calculated for the Unprotected Loss-of Flow accident (ULOF) by employing the accident analysis code ARG0. Based on the investigations on the FP distributions along the primary coolant system including the core, hot and cold sodium pools, IHX, and cover gas space, the physical models to be improved in the TRACER code were clalified. Especially the radiation heat transfer model will be required to describe the thermal consequcnce of the cover gas space heated up by the decay heat of FPs released to the cover gas from the failed fuel through the coolant sodium. It was also emphasizd that the FP transfer models should be ...
*; *
PNC TJ9263 97-001, 108 Pages, 1997/03
Along the line of work of preparing unified cross-section library for demonstration fast breeder reactors, Monju criticality and control rod worth experimental data and their analysis were reviewed to evaluate experimental and analysis errors to apply the data in the cross-section adjustment work. The sensitivity coefficients of each cross-section for criticality and control rod worth were calculated by use of the generalized perturbation code SAGEP. The C/E values, experimental and analysis errors, and sensitivity coefficients are utilized in the adjustment of JENDL 3.2 in the other part of this series of work.