本研究針對台電公司核一乾貯熱測試TSC01護箱進行嚴謹的表面輻射劑量分析，採用SCALE6.1程式集中的MAVRIC運算序列，它是一套應用CADIS理論且能自動化地結合DENOVO決定論法程式及MONACO蒙地卡羅法程式之運算序列，主要機制在於藉由一次近似的決定論法伴隨函數建構出具有一致性的權重射源及重要性地圖，大輻加速蒙地卡羅法的計算效率以解決困難的屏蔽問題。
本研究的最終目標是為了與熱測試的護箱表面劑量量測實驗直接比較，用來測試目前最佳輻射屏蔽分析技術之準確程度。針對TSC01護箱的內容物，本研究考慮了56根用過燃料束的獨立射源，每根燃料束分別考量三種射源項（燃料中子、燃料光子及結構光子）的能譜和軸向分布，同時將護箱細部幾何模型納入計算模型，包含密封鋼筒內部56根燃料束的擺放位置與幾何形狀，以及鋼筒外部進出氣孔位置、不銹鋼層、混凝土屏蔽的模擬。蒙地卡羅計算應用FW-CADIS理論，將伴隨射源設定為護箱側邊和頂部之表層，有效地迫使大量追蹤粒子往外遷移以增加護箱表面粒子計數的機會。本研究已獲得最可靠詳細的TSC01護箱表面劑量的分布圖，未來量測規劃預計藉由三種高效率的偵檢器（He-3比例計數器、HPIC、及TLD）對TSC01護箱作表面劑量掃描測量。同時，藉由已建置之屏蔽模型，本研究進行了大量的參數靈敏度分析以確認不同目的下的最佳輻射屏蔽模型，研究成果有助於未來國內類似案例的安全分析與相關研究。

The first nuclear power plant in Taiwan has been commercially operated for more than 34 years. The capacity of its spent fuel storage pool has been nearly exhausted. In order to solve the impending shortage problem of spent nuclear fuel storage, Taiwan Power Company is currently constructing an independent spent fuel storage installation (ISFSI) at the plant site to maintain normal plant operation. Focusing on the first storage cask TSC01, this study models the source terms and cask geometry as detailed as possible in order to predict a realistic dose rate distribution over the cask surfaces. Neutron and gamma-ray source terms of the first batch of 56 spent fuels were determined one by one according to their specifications, burnup histories, and cooling times. Energy spectra, axial distributions, and prescribed loading pattern of 56 spent fuels in the canister were included in the source description. The geometry of storage cask was modeled in detail including all the air inlets/outlets of the cask to evaluate possible radiation streaming. Monte Carlo method is the only choice for performing such complicated simulations. Effective variance reduction techniques are indispensable to making it computationally practical. MAVRIC in the SCALE 6.1 code system was used for calculating the dose rate distribution over the entire cask surface as functions of radial, axial, and azimuthal dependences. Based on the CADIS methodology and deterministic adjoint solution, MAVRIC can perform efficient radiation transport calculations for fixed-source shielding problems through consistent use of source biasing and weight window techniques. In addition to verify the conservativeness of results in the safety analysis report, the main purpose for such comprehensive and detailed modeling aims to predict a dose rate distribution as accurate as possible for making a high-quality comparison with field measurements. High-efficiency neutron proportional counter, HPIC, TLD, and integrated scanning frame were in preparation and test for measuring dose rates on the cask surfaces.

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