一种基于马氏距离建立简化多元统计模型的方法

提出一种基于样本之间最小马氏距离的样本平均方法 ,从总体正常历史采样数据样本集合中 ,构造新的数据样本集 ,建立简化多元统计模型 .然后通过判断两数据集的质心偏移和协方差的差异程度来检验新的数据样本集对总体样本集的可代表性 ,从而达到用较少的有效样本代表总体样本统计特征的目的 .仿真结果表明用本文提出的简化多元统计模型进行故障诊断的效果与传统模型相同 ,而降低了对系统存储量和计算量的要求     

Modeling representative Gen‐IV molten fuel reactivity effects in the ZEPHYR ZPR ‐ LFR analysis

Studies related to severe core accidents constitute a crucial element in the safety design of Gen‐IV systems. A new experimental program, related to severe core accidents studies, is proposed for the zero‐power experimental physics reactor (ZEPHYR) future reactor. The innovative program aims at studying reactivity effects at high temperature during degradation of Gen‐IV cores by using critical facilities and surrogate models. The current study introduces the European lead‐cooled system (ELSY) as an additional Gen‐IV system into the representativity arsenal of the ZEPHYR, in addition to the sodium‐cooled fast reactors. Furthermore, this study constitutes yet another step towards the ultimate goal of studying severe core accidents on a full core scale. The representation of the various systems is enabled by optimizing the content of plutonium oxide in the ZEPHYR fuel assembly. The study focuses on representing reactivity variation from 900°C at nominal state to 3000°C at a degraded state in both ELSY and Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) cores. The study utilizes the previously developed calculation scheme, which is based on the coupling of stochastic optimization process and Serpent 2 code for sensitivity analysis. Two covariance data are used: the ENDF 175 groups for ELSY and the Covariance Matrix Cadarache (COMAC) 33 groups for ASTRID. The effect of the energy group structure of the covariance data on the representativity process is found to be significant. The results for single degraded ELSY fuel assembly demonstrate high representativity factor (>0.95) for reactivity variation and for the criticality level. Also, it is shown that the finer energy group structure of the covariance matrices results in dramatic improvement in the representation level of reactivity variations.     

试谈影响混凝土试件强度代表性的几个因素

分析了影响砼试件强度代表性的几种原因并提出改进措施。     

Development of a “best representativity” method for experimental data analysis and an application to the critical experiments at the Toshiba NCA facility

For nuclear critical experiments, it is essential to certify similarities of the experiment with the objective of the actual reactor conditions or actual reactor equipment. To judge the applicability of the experimental data, the concept of a “representativity factor” has recently been adopted in the critical experiment field, particularly for fast breeder reactors and future reactor studies. In this study, we extended this concept to the design of a light water reactor system. We developed a new numerical evaluation method and a calculation system. The method is based on a linear combination of the sensitivity coefficient vector of an experiment in which the representativity factor to the target system is maximized to utilize experimental data effectively. Simultaneously, using the measurement data of critical experiments, the method enables us to evaluate calculation errors caused by errors or uncertainties of physical parameters. The derivation of the new calculation method is explained first. We then qualify it with a sample calculation, presenting numerical results for three kinds of critical experiments conducted at the Toshiba Nuclear Critical Assembly facility. Finally, the results are compared with those of an extended bias factor method to clarify the performance of the new method.     

Modeling representative Gen‐IV molten fuel reactivity effects in the ZEPHYR fast/thermal coupled ZPRs. Part I—Assembly level

The comprehension of severe criticality accident is a key issue in Gen‐IV neutronics and safety. Within the future zero‐power experimental physics reactor (ZEPHYR), to be built in Cadarache in the next decade, innovative approaches to reproduce high temperature partially degraded Gen‐IV cores into a critical facility is being investigated. This work presents the first attempt to represent a fuel assembly of sodium‐cooled fast reactor severe criticality accident based on surrogate models. One identified way to construct such representative configuration is to use MASURCA plates stockpile (MOX, UOx, Na, U, and Pu metal) in a fast/thermal coupled core to model a stratified molten assembly. The present study is the first step in a more global approach to full core analysis. The approach is based on a nature‐inspired metaheuristic algorithm, the particle swarm optimization algorithm, to find relevant ZEPHYR configuration at 20°C that exhibits characteristics of (2000‐3000°C) molten MOX assembly in a stratified metal arrangement in a reference sodium‐cooled fast reactor core. Thus, the underlying research question of this study is whether it is possible to represent temperature‐related reactivity effects occurring at fuel meltdown temperatures in a power reactor as density‐related reactivity effects at the operation temperature of a zero‐power reactor, and if so, how should it be done? The calculations are based on a Serpent‐2 Monte Carlo sensitivity and representativity analysis using the Cadarache's cross sections covariance data (COMAC). The single fuel assembly studies show that it is possible to represent the multiplication factor with a representativity factor greater than 0.98. As for reactivity variations, it is possible to achieve a satisfactory representativity factor of above 0.85 in all the presented cases. The representativity process demonstrates that temperature effects could be translated into density effects with good confidence. A complementary analysis on modified nuclear data covariance matrix demonstrates the importance of selecting consistent and robust uncertainties in the particle swarm optimization algorithm. This work provides insights on the behavior of the representativity scheme in different core states and shades some light on the problem in hand.     

铅、锌精矿质量检验技术的研究及应用

对多种散装铅锌精矿取样方法进行了研究，确定改进型货车取样法切实可行，且应用后效果良好。