混合式輻射遷移計算可結合決定論法與蒙地卡羅法的優點，是目前公認對於困難輻射屏蔽問題最佳的解決方案。本研究著重於探討混合式遷移計算在用過核子燃料乾式貯存屏蔽分析的應用，一系列的研究獲致二項重要成果：(1)針對二套主流的混合式遷移程式MAVRIC序列與ADVANTG/MCNP耦合計算，證明二者的整體表現是一致的，不論是準確度或是其計算效率；(2)針對複雜的輻射屏蔽問題，建構一套可有效區分不同輻射遷移途徑之貢獻的計算方法。第一項成果透過一個詳細的乾貯護箱表面劑量計算來呈現，第二項成果則透過一個大型複雜的室內乾貯設施周遭輻射場特性分析來呈現。
第一部分：針對核二廠的MAGNASTOR護箱案例，本研究有系統地檢視MAVRIC與ADVANTG/MCNP二套程式在護箱表面劑量率計算的準確度與效率，結果顯示兩套程式皆能有效地大幅加速整體計算，特別是在加馬射線的深穿透計算上，而且二者可達到的加速幅度相當接近。另外，MAVRIC (CE)與ADVANTG/MCNP雖然皆使用同樣來源的ENDF連續能量截面，前者仍低估護箱側邊中子劑量率約30 %，此一差異經深入探討證實是來自兩程式預設對於混凝土中的氫採用之S(,)處理不一致所致。修正此一差異後，本研究證實MAVRIC與ADVANTG/MCNP二者在計算準確度與效率接近等價，澄清文獻中許多混淆不清的二者差異討論。
第二部分：利用MAVRIC程式探討大型集中式貯存設施周遭輻射場的特性，參考並改進文獻中針對高能加速器引發之中子天空散射的計算模式，本研究建構了一套可有效區分下列不同輻射遷移途徑的分析方法：直接穿透、輻射滲流、天空散射、地面散射、多次散射。此一方法擴展了傳統蒙地卡羅計算的限制，不同輻射途徑的劑量貢獻可直接指出屏蔽加強的最有效方向與其最佳化。此一區分不同輻射遷移途徑分析方法的應用，並不僅限於用過核子燃料乾貯設施，亦可應用於其他輻射設施的屏蔽分析。

Hybrid transport calculations combine the advantages of deterministic and Monte Carlo methods and can make a challenging Monte Carlo shielding simulation computationally feasible and practical. Based on the consistent adjoint driven importance sampling (CADIS) methodology, ADVANTG/MCNP and MAVRIC are two widely used code systems in this area. This study applied the hybrid method to investigate the radiation field around spent fuel dry cask storage facilities, which involve various computational challenges, including multiple sources, complicated geometries, deep penetration, radiation streaming, and skyshine. The presentation was divided into two parts.
The first part described a real-world cask shielding problem and presented a systematic comparison of the two hybrid code systems that were used to solve it. Compared with the continuous-energy ADVANTG/MCNP calculations, the coarse-group MAVRIC calculations underestimated the neutron dose rates on the cask’s side surface by an approximate factor of two and slightly overestimated the dose rates on the cask’s top and side surfaces for fuel gamma and hardware gamma sources because of the impact of multigroup approximation. The fine-group MAVRIC calculations improved to a certain extent and the addition of continuous-energy treatment to the Monte Carlo code in the latest MAVRIC sequence greatly reduced these discrepancies. For the two continuous-energy calculations of ADVANTG/MCNP and MAVRIC, a remaining difference of approximately 30% between the neutron dose rates on the cask’s side surface resulted from inconsistent use of thermal scattering treatment of hydrogen in concrete. After this inconsistency had been solved, the ADVANTG/MCNP and MAVRIC exhibited excellent agreement in predicting dose rates on the cask’s surfaces and similar performance in terms of computational efficiency.
The second part dealt with how to separate neutron and gamma-ray contributions from multiple transport pathways including direct, streaming, skyshine, groundshine, and multishine. A modified version of the method that was originally developed by Oh et al. (2016) for the evaluation of neutron skyshine from a high-energy electron accelerator was proposed. The application of the methodology was demonstrated in this study and the flux/dose contributions of individual pathways were examined and compared. The results provided additional insight into how the radiation propagated from the source to off-site locations. The modified method for separating five transport pathways can provide valuable information for shielding optimization during the design phase and is generally applicable to Monte Carlo shielding analyses of other nuclear facilities.

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