针对裂变产额和半衰期的燃耗计算灵敏度和不确定度分析方法

基于广义微扰理论推导了裂变产额和半衰期的燃耗灵敏度系数理论模型,该模型考虑了原子核密度和中子通量的相互影响,并开发了燃耗计算中有效增殖因数和原子核密度等响应参数对核数据的灵敏度和不确定度分析程序。基于评价核数据中裂变产物独立产额的标准差数据,产生了针对压缩燃耗数据库的裂变产额协方差矩阵,以提高不确定度的计算精度。基于ENDF/B-Ⅶ.1数据库量化了UAM基准题TMI-1栅元无限增殖因数及重要裂变产物和重核的原子核密度由裂变产额和半衰期引入的不确定度。数值结果表明,对于栅元无限增殖因数,裂变产额和半衰期引入的不确定度很小;对于部分裂变产物的原子核密度,裂变产额和半衰期会引入较大的不确定度。     

基于抽样方法的燃耗计算核素积存量不确定度分析

基于抽样基本原理研究了应用于燃耗计算的不确定度分析方法,并开发了燃耗计算不确定度分析程序。基于评价核数据库ENDF/B-Ⅷ.0的裂变产额标准差和衰变常量标准差计算得到了衰变常量协方差矩阵和带相关性的裂变产额协方差矩阵,并结合SCALE6.2程序包的56群反应截面协方差数据库,对Takahama-3压水堆组件基准题中SF95-4样品进行不确定度分析。计算了反应截面、衰变常量和裂变产额不确定度引起的核素积存量的不确定度。计算结果表明,反应截面的不确定度是锕系核素积存量不确定度的主要来源,裂变产额和衰变常量的不确定度对部分裂变产物的积存量会引入较大的不确定度。但考虑裂变产额相关性后,裂变产额引起的不确定度显著降低。     

~(235,238)U和~(239)Pu中子诱发裂变独立产额的系统学研究和数据评价

裂变产额在核工程中有重要的应用,例如反应堆工程中衰变热、裂变毒物的计算。独立产额是裂变产额数据的重要组成部分。本工作评价给出235,238U、239Pu的热能点、裂变谱中子和14MeV中子裂变的所有质量链(A=66～172)的独立产额。100多个质量链1000多种产物核的独立产额是不可能完     

53 ^235，238U和^239Pu中子诱发裂变独立产额的系统学研究和数据评价

裂变产额在核工程中有重要的应用，例如反应堆工程中衰变热、裂变毒物的计算。独立产额是裂变产额数据的重要组成部分。本工作评价给出^235，238U和^239Pu的热能点、裂变谱中子和14MeV中子裂变的所有质量链（A=66～172）的独立产额。     

基于贝叶斯机器学习对中子诱发235U裂变的产额-能量关系的研究

核裂变碎片产额是核能发展和核技术应用领域的重要基础数据,但在实验和理论上获得精确且完整的能量依赖的裂变产额到目前为止都非常困难。本文提出采用贝叶斯机器学习方法对所有收集到的中子诱发235U裂变产额实验数据进行了数据融合学习和评价。基于该评价方法对关键裂变碎片的产额 能量关系进行推断,并得到了二维的碎片累积产额分布随入射中子能量的变化关系。所得的二维产额分布能合理地反映裂变模式随能量增加的演化,但目前结果的不确定度较大,有待进一步改进。     

锕系核裂变数据关键科学问题研究

正1概述本研究建立了用于中子俘获截面测量的C6D6探测系统,经过在中国散裂中子源上开展的首批物理实验测试,证明该系统能正常工作,获得了197Au和169Tm的初步结果。采用粒子能量速度关联技术和瞬发伽马符合技术分别对初级裂变产物粒子的质量和电荷进行了鉴别。完成了中子诱发235U裂变重点产额的评价以及燃耗计算灵敏度和不确定度对于裂     

压水堆燃耗数据库的制作与验证

基于评价数据库ENDF/B-Ⅷ.0和EAF-2010研制了一套适用于CINDER90程序的压水堆用燃耗数据库,该数据库包含中子反应截面、衰变数据和裂变产额数据3部分。中子反应截面的加工分为两步,首先采用Inverted Stack算法和CRECTJ6程序将EAF 2010库的截面分支比融入ENDF/B Ⅷ0库全套中子评价数据,然后用NJOY2016程序处理成63群截面。衰变数据和裂变产额数据分别由MF8/MT457和MF8/MT454数据加工得到,裂变产额数据共包含36个裂变核的60组产额数据。以SFCOMPO 20中Takahama 3压水堆燃料组件为基准题,对研制的燃耗数据库进行了验证。结果表明,本文制作的燃耗数据库的方法是正确的,对于某些核素,如242Amm,制作的数据库比自带库的计算结果更接近实验值。     

压水堆核电厂废液放射性计算程序配套核数据库的适用性评价

PWR-GALE是美国核管会编制并使用的压水堆核电厂气液态流出物源项计算程序,现有的配套核数据库已有长达四十年之久未进行更新,无法确定是否能够满足先进压水堆的计算和审评需求,需要通过基于最新版本的核评价数据库制作新的配套数据库对其进行适用性的评价。因此,本文基于核评价数据库ENDF/B-VII.0,提取衰变子库中相关信息,根据直接裂变产额、衰变信息以及保留的裂变产物核素得到更新的沿衰变链归并的产额数据,通过中子学-燃耗耦合计算获得了更新的中子微观反应截面数据;并与现有的配套数据库进行了对比分析;然后,通过计算一系列面向不同机型的算例进行了整体的对比验证与分析。结果表明:现有的PWR-GALE配套核数据可以满足先进压水堆的计算和评审需求。     

强中子场裂变产物中子核反应网络方程研究

本文建立了考虑中子参加反应的裂变产物中子反应及衰变的网络方程,选用求解一阶线性刚性微分方程组的Gear方法,开发了可计算任意裂变产物核数量在不同中子场强度和中子谱下随时间变化的核反应网络方程计算系统FIRENEQ,并配套了裂变产物产额和衰变数据库FPYDDL及裂变产物核中子反应截面数据库FPNCDL。检验结果表明,计算结果正确,程序可靠。利用该程序系统,研究了裂变产物核数量在不同中子场、不同诱发中子能量下随时间的变化。     

~(252)Cf自发裂变中子诱发~(235)U裂变时几个核素裂变产额的绝对测定

引言 在燃耗测量工作中,常应用质量产额曲线峰上的裂变产物核素作为监测体。一般说来,这些核素的裂变产额随入射中子能量的变化不灵敏。但是~(99)Mo的情况比较复杂。文献中~(235)U热中子裂变的产额数据相互符合得较好,但是裂变谱或快堆谱诱发裂变时,数据比较混乱,     

Estimation of 239Pu independent and cumulative fission product yields from the chain yield data using a Bayesian technique

The independent and cumulative fission product yields (FPYs) are obtained by using the Bayesian technique based on the evaluated mass chain yield, where required constraints such as the normalization can be straightforwardly included. We apply this technique to the ²³⁹Pu FPY data at neutron incident energies of 0.5, 2.0, and 14 MeV, where the most updated mass chain yield ENDF/B-VII.1 data are available. The obtained yield data are compared with the evaluated values by England and Rider in ENDF/B-VI, and differences from their values are investigated. We show that the modern decay data used, such as branching ratios to ground and metastable states, cause differences in the evaluated individual and cumulative fission yields.     

燃耗信任制下燃耗计算对临界计算的偏差及不确定度的研究

为量化燃耗信任制中燃耗计算传递给临界计算的不确定度,本文基于参数统计法对燃耗计算的核素偏差及偏差不确定度展开分析,并以蒙特卡罗（MC）抽样方法计算的kinf不确定度为基准,比较不同抽样方法对临界计算不确定度的影响。结果表明,核素偏差与偏差不确定度是随样品燃耗变化的分段函数。对于临界计算,拉丁超立方抽样（LHS）方法与MC抽样方法的kinf不确定度计算结果吻合较好,且LHS方法可考虑参数间的相关性,计算结果更真实,可进一步提升电厂的经济性。     

Fission Product Yields for 14 MeV Neutrons on U, U and Pu

We report cumulative fission product yields (FPY) measured at Los Alamos for 14 MeV neutrons on ²³⁵U, ²³⁸U and ²³⁹Pu. The results are from historical measurements made in the 1950s–1970s, not previously available in the peer reviewed literature, although an early version of the data was reported in the Ford and Norris review. The results are compared with other measurements and with the ENDF/B-VI England and Rider evaluation. Compared to the Laurec (CEA) data and to ENDF/B-VI evaluation, good agreement is seen for ²³⁵U and ²³⁸U, but our FPYs are generally higher for ²³⁹Pu. The reason for the higher plutonium FPYs compared to earlier Los Alamos assessments reported by Ford and Norris is that we update the measured values to use modern nuclear data, and in particular the 14 MeV ²³⁹Pu fission cross section is now known to be 15–20% lower than the value assumed in the 1950s, and therefore our assessed number of fissions in the plutonium sample is correspondingly lower. Our results are in excellent agreement with absolute FPY measurements by Nethaway (1971), although Nethaway later renormalized his data down by 9% having hypothesized that he had a normalization error. The new ENDF/B-VII.1 14 MeV FPY evaluation is in good agreement with our data.     

Fission Product Yields from Fission Spectrum n+Pu for ENDF/B-VII.1

We describe a new cumulated fission product yield (FPY) evaluation for fission spectrum neutrons on plutonium that updates the ENDF/B-VI evaluation by England and Rider, for the forthcoming ENDF/B-VII.1 database release.¹ We focus on FPs that are needed for high accuracy burnup assessments; that is, for inferring the number of fissions in a neutron environment. Los Alamos conducted an experiment in the 1970s in the Bigten fast critical assembly to determine fission product yields as part of the Interlaboratory Reaction Rate (ILRR) collaboration, and this has defined the Laboratory's fission standard to this day. Our evaluation includes use of the LANL-ILRR measurements (not previously available to evaluators) as well as other Laboratory FPY measurements published in the literature, especially the high-accuracy mass spectrometry data from Maeck and others. Because the measurement database for some of the FPs is small — especially for ⁹⁹Mo — we use a meta-analysis that incorporates insights from other accurately-measured benchmark FP data, using R-value ratio measurements. The meta-analysis supports the FP measurements from the LANL-ILRR experiment. Differences between our new evaluations and ENDF/B-VI are small for some FPs (less than 1-2%-relative for ⁹⁵Zr, ¹⁴⁰Ba, ¹⁴⁴Ce), but are larger for ⁹⁹Mo (4%-relative) and ¹⁴⁷Nd (5%-relative, at 1.5 MeV) respectively. We present evidence for an incident neutron energy dependence to the ¹⁴⁷Nd fission product yield that accounts for observed differences in the FPY at a few-hundred keV average energy in fast reactors versus measurements made at higher average neutron energies in Los Alamos' fast critical assemblies. Accounting for such FPY neutron energy dependencies is important if one wants to reach a goal of determining the number of fissions to accuracies of 1-2%. An evaluation of the energy-dependence of fission product yields is given for all A values based on systematical trends in the measured data, with a focus on the energy dependence over the fast neutron energy range from 0.2-2 MeV. Based on these trends, we present an evaluation of the FPY data at 0.5 and 2.0 MeV average incident neutron energies. This new set of ENDF/B-VII data will enable users to linearly interpolate between the pooled FPY data at ∼0.5 MeV and our new data at 2 MeV to obtain FPYs at other energies.     

Energy Dependence of Plutonium Fission-Product Yields

A method is developed for interpolating between and/or extrapolating from two pre-neutron-emission first-chance mass-asymmetric fission-product yield curves. Measured ²⁴⁰Pu spontaneous fission and thermal-neutron-induced fission of ²³⁹Pu fission-product yields (FPY) are extrapolated to give predictions for the energy dependence of the n + ²³⁹Pu FPY for incident neutron energies from 0 to 16 MeV. After the inclusion of corrections associated with mass-symmetric fission, prompt-neutron emission, and multi-chance fission, model calculated FPY are compared to data and the ENDF/B-VII.1 evaluation. The ability of the model to reproduce the energy dependence of the ENDF/B-VII.1 evaluation suggests that plutonium fission mass distributions are not locked in near the fission barrier region, but are instead determined by the temperature and nuclear potential-energy surface at larger deformation.     

Burnup Calculation Sensitivity and Uncertainty Analysis Method of Fission Yield and Decay Half-life

The calculation model of sensitivity coefficient for decay half-life and fission product yield in burnup calculation was derived based on generalized perturbation theory, which considered the interaction between nuclear concentration and neutron flux. A code was developed to calculate sensitivity and uncertainty of effective neutron multiplication factors and nuclide concentration caused by nuclear data. Covariance matrix of fission yield for a simplified burnup library was generated based on standard deviation data of independent fission yield in evaluated nuclear data library to improve the accuracy of uncertainty quantification. Uncertainties induced by decay half-life and fission yield on infinite neutron multiplication factors and nuclide concentration for TMI-1 pin-cell in the UAM burnup benchmark were quantified based on ENDF/B-Ⅶ.1. The numerical results show that the uncertainty of infinite neutron multiplication factors induced by decay half-lives and fission yields is low, while the uncertainty of concentration of some fission product nuclide is high.     

Effect of fission products yield library on the calculation of the nuclide composition of the fuel in sodium-cooled fast reactors

The calculation of the composition of irradiated fuel for different degrees of burnup is a basic problem in the analysis of nuclear-radiological safety of objects holding spent fuel assemblies. The yield of fission products is one of the important initial indicators in burnup calculations. Methods for compiling libraries of fission products yield on the basis of the ENDF/B up-to-date evaluated nuclear data files are described. The nuclide composition of uranium oxide and uranium-plutonium-zirconium metal fuel in sodium-cooled fast reactors is analyzed by means of high-precision calculations performed with different fission product yields libraries using different computer codes MONTEBURNS–MCNP5–ORIGEN2 and the results are presented.