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Surface dose rates of the spent fuel dry storage cask system at the first nuclear power plant in Taiwan   HUANG Yu-Shiang1, LAI Po-Chen2, and SHEU Rong-Jiun3   1. Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Sec. 2, Kuang-Fu Road, Hsinchu, Taiwan, R.O.C. (yushiang@mx.nthu.edu.tw) 2. Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Sec. 2, Kuang-Fu Road, Hsinchu, Taiwan, R.O.C.  (bertieq231@yahoo.com.tw) 3. Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Sec. 2, Kuang-Fu Road, Hsinchu, Taiwan, R.O.C.  (rjsheu@mx.nthu.edu.tw)   Abstract: Source terms and surface dose rates of a spent fuel dry storage cask system were evaluated using TRITON and MAVRIC in the SCALE 6.1 code package. The cask system consists of four major components, called the transportable storage canister, transfer cask, vertical concrete cask, and add-on shield. Based on advanced fuel depletion and radiation transport methodologies, source characteristics of the design basis spent fuel and detailed dose rate distributions over the entire surface of the cask system were obtained. The results confirm the appropriateness of the original shielding analysis and demonstrate great advantages of using this approach. In addition, the comprehensive dose rate distributions can provide useful information in preparing associated health physics programs during the transportation and dry storage of spent fuels. Keyword: spent nuclear fuel; dry storage cask; source term; radiation shielding; Monte Carlo
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