臨界安全分析在核燃料循環過程中扮演極為重要的角色，而燃料中的燃耗效應是影響臨界狀態的主要原因。本論文針對用過核燃料乾式貯存設施進行燃耗信用效應（Burnup Credit）對臨界分析影響之探討；並從事第四代鉛冷快中子反應器SSTAR（Small Secure Transportable Autonomous Reactor）的爐心模擬計算，並分析燃耗期間爐心各項物理特性。
本研究採用SCALE（Standardized Computer Analyses for Licensing Evaluation）程式系統版本5.1與6.0，以其中之三維蒙地卡羅臨界計算程式KENO進行複雜問題的幾何模擬，並選用適當截面數據庫，彙同相關截面處理程式模組、燃耗計算程式模組結合成之計算序列進行計算分析。
用過核燃料乾式貯存系統之傳送護箱裝載滿24束PWR燃料，235U濃縮度為4.2 wt%，中子吸收板含有75%的10B。以全新燃料計算，增殖因數（k-eff）結果為0.9227；假設燃料經過45,000 MWD/MTU的燃耗，考慮於系統中陸續加入經燃耗產生的各種錒系元素與分裂產物，結果顯示增殖因數隨之下降，當考慮程式內建所有可得的錒系元素與分裂產物時，k-eff最後降低至0.6270。
SSTAR鉛冷快中子反應器的特點為很長的燃料週期，燃料為氮化超鈾元素，以固定比功率（specific power）21.19 MW/MTHM，相當約150MW的功率運轉，計算結果顯示此反應器可維持20年無須更換燃料。運轉過程中，爐心各區總通率約在5.07×1014neutrons/cm2-sec至1.70×1015 neutrons/cm2-sec間。低濃縮度燃料區的平均能量範圍約為3.35×105eV至4.41.×105eV，趨勢為上升或持平；高濃縮度燃料區則從6.00×105 eV降至4.80×105 eV。燃料中的U-235與U-238量隨時間呈現消耗狀態；Pu-239在低濃縮度區域為上升，在高濃縮度區則是消耗。

During all process of nuclear fuel cycle, criticality safety analysis is indispensable. Burnup effect dominates the result of criticality analysis. In this study, we evaluated the effect of burnup credit on the k-eff of a spent fuel dry storage transfer cask system, and performed core neutronics analysis for a generation IV lead-cooled fast reactor SSTAR(Small Secure Transportable Autonomous Reactor).
The computer software systems SCALE 5.1 and 6.0 were used in this thesis. We modeled complex geometries by taking advantage of the 3D Monte Carlo code KENO, which is a functional module for criticality evaluation in SCALE. Several sequences consisting of cross section processing, depletion modules and KENO were used to complete the assessments.
We consider a transfer cask loaded with 24 PWR fuel assemblies with 4.2 % 235U enrichment. For fresh fuels, k-eff was calculated to be 0.9227. For fuel assemblies with 45,000MWD/MTU burnup, k-eff kept decreasing as actinides and fission products being added into the fuel batch by batch. When all available actinides and fission products are included, k-eff was reduced to 0.6270.
The SSTAR lead-cooled fast reactor is characterized by its long core life. The fuel composition is transuranium nitride, and the specific power is 21.19 MW/MTHM. The calculated result shows that the reactor can last for 20 years without refueling. During operation, the total flux of each core region varies from 5.07×1014 neutrons/cm2-sec to 1.70×1015 neutrons/cm2-sec. The average energy of low-enriched fuel region is about 3.35×105 eV to 4.41×105 eV, and the trend is increasing or fixed; in high-enriched fuel region the average energy decreases from 6.00 × 105 eV to 4.80 × 105 eV. The amount of U-235 and U-238 consumed with time; Pu-239 is bred in low-enriched region but consumed in high-enriched region with time.

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