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Saito, Kosuke; Maeda, Seiichiro; Higuchi, Masashi*; Takano, Mitsuhiro*; Nakazawa, Hiroaki
JAEA-Technology 2006-035, 76 Pages, 2006/06
JAEA-Technology-2006-035.pdf:5.25MB
Because of the revision on the standardized strength of the ODS steel, the previous design study of MONJU demonstrative core has been obliged to be reconsidered. For economical advantages, only a 127 pins-bundle core was selected to be redesigned. For the sake of cladding endurance, the ratio of cladding thickness to outer diameter was reset incrementally followed by the determination of the basic specification of a pin. Notwithstanding some deterioration thanks to the reduction of a fuel volume fraction, the prospect in neutronics was obtained. Coolant flow distribution design which was based on power distribution was successfully carried out without overheating cladding. Average burn-up of 150 GWd/t and 380 days-long operational period per cycle are to be attained, and the designed core can thermally afford to receive test fuels. The study has necessity to be advanced extensively for the purpose of materialization according to the circumstances of MONJU in future.
Maeda, Seiichiro; Togashi, Nobuhito*; Higuchi, Masashi*; Takano, Mitsuhiro*; Abe, Tomoyuki
JNC TN8400 2003-028, 135 Pages, 2003/12
JNC-TN8400-2003-028.pdf:85.68MB
The Monju advanced core concept to demonstrate a practical LMFBR core with 150GWd/t (average discharged burnup) was embodied in this design study. A high performance fuel with annular pellets of a large diameter filled in ODS (oxide dispersion strengthen ferritic steel) claddings was applied in the advanced core. This enables improvement of an internal conversion ratio in combination with increase of effective fuel volume fraction, achievement of high burnup up to 150GWd/t and a long operation period beyond 1 year in Monju. The core in which the practical high burnup lessens a burden for a fuel cycle system including fuel fabrication and reprocessing can be demonstrated. In the first step, constraints in the existing plant and requirements to demonstrate the practical LMFBR were clarified. The core and fuel specifications were surveyed with parameters of a number of fuel pins in an assembly and so on. Two types of cores with 127-pin-bundle and 91-pin-bundle were selected as candidates. In the second step, performances of these core options were specified in this design study. It was shown that major parameters in neutronic design, hydraulic design and fuel design would meet criteria. The application of the high performance fuel significantly contributes the enhancement of economical efficiency of Monju itself. The net operation cost will be greatly reduced by increase of the annual electricity generated caused by a boost of the plant operating rates and by saving of the annual discharged fuel assemblies up to 1/2 or 1/3. The deliberate margin for thermal limits ensures the irradiation field to develop new type fuels and core materials and to demonstrate a low decontaminated fuel with miner actinides as a candidate of advanced fuel cycle. The results in this study may become a technically credible guideline to make the future management plan of Monju.
Maeda, Seiichiro; Takashita, Hirofumi; Okawa, Tsuyoshi; Higuchi, Masashi*; Abe, Tomoyuki
JNC TN8400 2003-019, 185 Pages, 2003/08
JNC-TN8400-2003-019.pdf:11.78MB
We are studying on an in-core breeding concept as a candidate for a practical FBR fuel cycle system attainable in an early stage on the premise that sodium coolant and mixed oxide fuel should be adopted, since the technical issues with these combination are most advanced and common with the fuel cycle for a LWR-MOX system. An enhancement of fuel volume fraction using thick fuel pins enables the in-core breeding. The fuel material flow can be greatly lessened by minimizing amount of the blanket with the in-core breeding core. The low material flow leads to significant reduction of the fuel cycle cost. We investigated a 3500 MWth large-scale core adjusting several conditions presented in JNC's feasibility study program for a commercialized FBR system in this study. These were shown in this study that a discharged burnup averaged over the core and the blanket could reach approximately 130 GWd/t (core averaged about 150 GWd/t) within the maximum fast neutron fluence about 5
10
/cm
, that the small reactivity loss with burnup easily enabled long operation and that stable power distribution during operation significantly improved hydraulic property in this type core. We investigated measures to reduce sodium void reactivity, because core height enlargement to enhance neutron efficiency caused the increase of sodium void reactivity.We also investigated feasibility of a high breeding type core with low burnup considering a variety of FBR introducing scenarios and a trade-off correlation between breeding performance and burnup extension. The performance in this core design at core disruption accidents is not revealed enough. Further investigation should be made in detail to confirm that the in-core breeding concept could be accepted in a safety aspect.
; *; *; *
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.