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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2014.1
- 수록면
- 547 - 556 (10page)
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초록· 키워드
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Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/orconcrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. Forthe purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is beingdeveloped by Korea Radioactive Waste Management Corporation (KRMC) in Korea.
At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature,pressure, etc. of the moderator, boron, cladding, UO2 pellets) in which the design basis fuel is loaded, as inputdata. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for thefinal shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate.
The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiationshielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases;one for a single cask and the other for a 2x10 cask array. For the single cask, dose rates at the external surface of the metalcask, 1m and 2m away from the cask surface, were evaluated. For the 2x10 cask array, dose rates at the center point of thearray and at the center of the casks’ height were evaluated. The results of the shielding analysis for the single cask show thatdose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) ofthe cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue sinceadditional shielding will be installed at the storage facility. The shielding analysis results for the 2x10 cask array showedexponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280mfrom the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than25mrem/yr, due to the decay of radioactivity of spent fuel in storage.
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참고문헌 (11)
참고문헌 신청
/ / Final Safety Analysis Report for the HI-STAR 100 Cask System. Rev.0
/ / Licensing requirements for the independent storage of spent nuclear fuel and high-level radioactive waste / 10 CFR Part 72
/ / ORNL/TM-2005-39, SCALE A Modular Code for performing Standardized computer Analyses for Licensing Evaluation, Sect. 32 S2 “SAS2:A Coupled 1-D Depletion and Shielding Analysis Modu
/ 1996 / Conversion Coefficients for Use in Radiological Protection against External Radiation
/ 2000 / Standard Review Plan for Spent Fuel Dry Storage Facility
/ / Licensing requirements for the independent storage of spent nuclear fuel and high-level radioactive waste / 10 CFR Part 72
/ / ORNL/TM-2005-39, SCALE A Modular Code for performing Standardized computer Analyses for Licensing Evaluation, Sect. 32 S2 “SAS2:A Coupled 1-D Depletion and Shielding Analysis Modu
/ 1996 / Conversion Coefficients for Use in Radiological Protection against External Radiation
/ 2000 / Standard Review Plan for Spent Fuel Dry Storage Facility
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