In this study, the Best Estimate Plus Uncertainty (BEPU) approach is developed for the systematic quantification and propagation of uncertainties in the modelling and simulation of lead-cooled fast reactors (LFRs) and applied to the demonstration LFR (DLFR) initially investigated by Westinghouse. The impact of nuclear data uncertainties based on ENDF/B-VII.0 covariances is quantified on lattice level using the generalized perturbation theory implemented with the Monte Carlo code Serpent and the deterministic code PERSENT of the Argonne Reactor Computational (ARC) suite. The quantities of interest are the main eigenvalue and selected reactivity coefficients such as Doppler, radial expansion, and fuel/clad/coolant density coefficients. These uncertainties are then propagated through safety analysis, carried out using the MiniSAS code, following the stochastic sampling approach in DAKOTA. An unprotected transient overpower (UTOP) scenario is considered to assess the effect of input uncertainties on safety parameters such as peak fuel and clad temperatures. It is found that in steady state, the multiplication factor shows the most sensitivity to perturbations in ²³⁵U fission, ²³⁵U ν, and ²³⁸U capture cross sections. The uncertainties of ²³⁹Pu and ²³⁸U capture cross sections become more significant as the fuel is irradiated. The covariance of various reactivity feedback coefficients is constructed by tracing back to common uncertainty contributors (i.e., nuclide-reaction pairs), including ²³⁸U inelastic, ²³⁸U capture, and ²³⁹Pu capture cross sections. It is also observed that nuclear data uncertainty propagates to uncertainty on peak clad and fuel temperatures of 28.5 K and 70.0 K, respectively. Such uncertainties do not impose per se threat to the integrity of the fuel rod; however, they sum to other sources of uncertainties in verifying the compliance of the assumed safety margins, suggesting the developed BEPU method necessary to provide one of the required insights on the impact of uncertainties on core safety characteristics.

中文翻译：

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铅冷却快堆安全性分析的核数据不确定性量化和传播

在这项研究中，开发了最佳估计加不确定度（BEPU）方法，用于铅冷快堆（LFR）建模和仿真中的系统量化和不确定性传播，并将其应用于西屋公司最初研究的示范LFR（DLFR）。 。使用蒙特卡罗代码Serpent和Argonne反应堆计算（ARC）套件的确定性代码PERSENT实施的广义扰动理论，在格点上量化了基于ENDF / B-VII.0协方差的核数据不确定性的影响。感兴趣的量是主要特征值和选定的反应性系数，例如多普勒，径向膨胀以及燃料/包层/冷却剂密度系数。然后，这些不确定性会通过使用MiniSAS代码进行的安全分析传播，遵循DAKOTA中的随机抽样方法。考虑采用无保护的瞬态过功率（UTOP）方案，以评估输入不确定性对安全参数（例如峰值燃料和复合温度）的影响。发现在稳定状态下，乘法因子显示出对扰动最敏感。²³⁵ ü裂变，²³⁵ ü ν和²³⁸ ü俘获截面。²³⁹ Pu和²³⁸ U捕获截面的不确定性随着燃料的照射而变得更加重要。通过追溯到常见的不确定性因素（即核素反应对）（包括²³⁸ U非弹性，²³⁸ U捕获和²³⁹），构造各种反应性反馈系数的协方差。Pu捕获横截面。还观察到，核数据的不确定性分别蔓延到峰值包覆温度和燃料温度分别为28.5 K和70.0 K的不确定性。这种不确定性本身不会对燃料棒的完整性构成威胁。但是，它们在验证假设的安全余量的合规性时总结了其他不确定性来源，表明开发出的BEPU方法对于提供不确定性对核心安全特性影响的必要见解之一是必要的。