ADS系统中少系核素的灵敏度分析

为了确定加速器驱动系统中少系核素的核数据的不确定性对此系统中积分参数的影响,选择一铅 铋冷却次临界系统来进行分析,选择不同的微观评价库CENDL 3.0,JENDL 3.2及ENDF/B VI.6,并只考虑20MeV以下的中子产额,来计算次临界系统的积分参数.以ENDF/B VI.2库的计算结果作为参考结果,依次用CENDL 3.0库及JENDL 3.2库中相应的系核素替换ENDF/B VI.6库中的系核素,计算了次临界系统的Keff值及反应率,并就变换主要系核素时对所计算的Keff值及反应率的影响进行了分析、比对.在堆芯处,当改变系核素时Keff值总的变化为30%,CENDL 3.0的238Pu的弹性散射反应率比ENDF/B VI库的结果高出10%,ENDF/B VI库的242Am的裂变率超过了CENDL 3.0库的15%.     

国际评价中子核数据库

介绍了当今世界上五大评价中子数据库 ,即美国的 ENDF/B- 6、日本的 JENDL- 3.2、欧洲的 JEF- 2 .2、中国的 CENDL- 2 .1及俄罗斯的 BROND- 2 ,的现况 ,以及国际评价中子数据库的发展趋势和用户如何获得与使用这些数据.The present status of the five major evaluated neutron nuclear data libraries in the world, i.e. ENDF/B 6(America), JENDL 3.2(Japan), JEF 2.2(Europe), CENDL 2.1(China), BROND2(Russia) are introduced. The developing trend of the international neutron evaluated nuclear data library is discussed. How to get and use these data for the domestic users is alro presented.     

WIMS多群截面库更新时NJOY输入参数研究

以最新的微观评价库CENDL-2.1为基础,用先进的群常数制作程序系统——NJOY来制作新的WIMS69群截面库,研究不同的 NJOY输入参数对 WIMS程序所计算的积分量的影响,并给出了详细的参数研究结果,同时还分析和讨论了计算结果与基准实验结果的比对.WIMS multi group constant library is the associated working library of WIMS/D4 lattice code, and it was created by using rather old and obsolete data based on ENDF/B3 (1972). Recently, the new evaluated data files such as ENDF/B 6.5, JEF 2.2, CENDL 2.1 and JENDL 3.2 were released. It s necessary to update the old library by the new evaluated data. The parameter study is performed to investigate the sensitivity of the integral parameters calculated with WIMS/D4 on the selection...     

ENDF/B-VIII.β与ENDF/B-VII.1中水的热中子散射截面数据对比分析

以常见的慢化剂材料轻水为例,分析了ENDF/B评价库中两个版本ENDF/B-VⅢ.β与之前ENDF/BVⅡ.1中热中子散射截面数据的差异。由于ENDF/B-VⅢ.β与ENDF/B-VⅡ.1等两个版本中热中子散射截面的温度不尽相同,为定量分析同一温度下二者的差异,通过插值方法得到了任意温度下的热中子散射截面数据。对比发现,二者在同一温度下的热中子散射数据存在差异。相对于ENDF/B-VⅡ.1评价库版本而言,基于ENDF/B-VⅢ.β评价库版本加工得到的束缚在水中的氢的热中子散射截面与ENDF/B-VⅡ.1存在较为明显的差异,两个版本给出的低温情况下的中子与靶核碰撞后的平均散射角余弦在热区的低能段出现极小的负值。此外,当温度高于室温时,在热区的低能段ENDF/B-VⅢ.β给出的平均次级中子能量要稍大于ENDF/B-VⅡ.1给出平均次级中子能量。This work takes the traditional moderator material H₂O for example, the differences between ENDF/B-VⅢ.β and ENDF/B-VⅡ.1 versions from the ENDF/B nuclear evaluated library are analyzed. In order to compare the thermal neutron scattering data between ENDF/B-VⅢ.β and ENDF/B-VⅡ.1 versions under the same temperature, the interpolation method is used to obtain the thermal neutron scattering data under arbitrary temperatures. There are some differences between ENDF/B-VⅢ.β and ENDF/B-VⅡ.1 versions, the thermal neutron scattering cross sections of H bound in H₂O within ENDF/B-VⅢ.β is different with that of ENDF/B-VⅡ.1, the scattering angle cosines of the neutron and target nuclei from the two libraries are slightly smaller than zero in lower energy range within thermal range. Moreover, the second scattering energy spectrum from ENDF/B-VⅢ.β is slightly harder than that from the ENDF/B-VⅡ.1 in higher energy range within thermal range when the temperature is higher than room temperature.     

混合评价核数据库HENDL1.0/MG/MC研制

根据世界几个主要核评价数据库,如ENDF/B 6(美国)、JEF 2.2(欧盟)、JENDL 3.2(日本)、BROND 2.2(俄罗斯)、CENDL 2.1(中国)和FENDL 2(IAEA/NDS),兼顾聚变、裂变以及聚变 裂变次临界混合堆设计研究的多种需要,经过甄别、筛选,最后集成为包含213个核素的基本评价文件,名为HENDL1.0/E的核评价数据库.在此基础上,利用目前流行的群常数加工程序系统NJOY和输运截面制备程序TRANSX制作两套用于中子或/和光子输运计算的输运截面工作库:①参考Vitamin J能群结构制作了175群中子和42群光子、中子 光子耦合多群工作数据库HENDL1.0/MG,可用于离散纵标Sn法程序计算;②连续能群结构、紧凑ENDF(ACE)格式中子截面库HENDL1.0/MC,可用于蒙特卡罗方法输运计算,如MCNP.另外还制作了可用于燃耗(嬗变)计算的燃耗库BURNUP.LIB和响应函数库RESPONSE.LIB两个专用数据库.同时,也对HENDL1.0综合评价核数据库的有效性进行了抽样测试、基准检验和初步确认. A Hybrid Evaluated Nuclear Data Library(HENDL) named as HENDL1.0 has been developed by Fusion Design Study (FDS) team of Institute of Plasma Physics, Academia Sinica (ASIPP) to take into account the requirements in design and research relevant to fusion, fission and fusion-fission sub-critical hybrid reactor. HENDL1.0 contains one basic evaluated sub-library naming HENDL1.0/E and two processed working sub-libraries naming HENDL1.0/MG and HENDL1.0/MC, respectively. Through carefully comparing...     

Study of shape isomer yields of ~（240）Am in the framework of a dynamical-statistical model

A dynamical statistical model is used to analyze the experimental shape isomer yields data in the reaction d＋240Pu at E=20–29 MeV.The possibility of determining the nuclear dissipation is discussed.Comparison of the experimental data with the calculations leads to a value of the reduced dissipation coeffcient β=0.45×1021 s-1 for the Am isotopes.     

Measurements of effective delayed neutron fraction in a fast neutron reactor using the perturbation method

A perturbation method is proposed to obtain the effective delayed neutron fraction β_(eff) of a cylindrical highly enriched uranium reactor.Based on reactivity measurements with and without a sample at a specified position using the positive period technique,the reactor reactivity perturbation △ρ of the sample in β_(eff) units is measured.Simulations of the perturbation experiments are performed using the MCNP program.The PERT card is used to provide the difference dκ of effective neutron multiplication factors with and without the sample inside the reactor.Based on the relationship between the effective multiplication factor and the reactivity,the equation β~(eff)=dκ/△ρ is derived.In this paper,the reactivity perturbations of 13 metal samples at the designable position of the reactor are measured and calculated.The average β_(eff) value of the reactor is given as 0.00645,and the standard uncertainty is 3.0%.Additionally,the perturbation experiments for β_(eff) can be used to evaluate the reliabilities of the delayed neutron parameters.This work shows that the delayed neutron data of ~(235)U and ~(238)U from G.R.Keepin's publication are more reliable than those from ENDF-B6.0.ENDF-B7.0,JENDL3.3 and CENDL2.2.     

Measurement and analysis of the 238U(n, 2n) reactionrate in depleted uranium/polyethylene shells

In order to check the conceptual design of the subcritical blanket in a fusion-fission hybrid reactor, a depleted uranium/polyethylene simulation device with alternate shells has been established. The measurement of the ²³⁸U(n, 2n) reaction rate was carried out using an activation technique, by measuring the 208 keVγ rays emitted from ²³⁷U. The self-absorption of depleted uranium foils with different thicknesses was experimentally corrected. The distribution of the ²³⁸U(n, 2n) reaction rate at 90° to the incident D⁺ beam was obtained, with uncertainty between 5.3% and 6.0%. The experiment was analyzed using MCNP5 code with the ENDF/BVI library, and the calculated results are all about 5% higher than the measured results.     

Influence of thermal and resonance neutron on fast neutron flux measurement by ~(239)Pu fission chamber

The 239Pu fission chambers are widely used to measure fission spectrum neutron flux due to a flat response to fast neutrons. However,in the meantime the resonance and thermal neutrons can cause a significant influence on the measurement if they are moderated,which could be eliminated by using10 B and Cd covers. At a column enriched uranium fast neutron critical assembly,the fission reaction rates of 239 Pu are measured as 1.791×10-16,2.350×10-16 and 1.385×10-15 per second for 15 mm thick10 B cover,0.5 mm thick Cd cover,and no cover respectively,while the fission reaction rate of 239 Pu is rapidly increased to 2.569×10-14 for a 20 mm thick polythene covering fission chamber. The average 239 Pu fission cross-section of thermal and resonance neutrons is calculated to be 500 b and24.95 b with the assumption of 1/v and 1/E spectra respectively,then thermal,resonance and fast neutron flux are achieved to be 2.30×106,2.24×106 and 1.04×108cm-2·s-1.     

ENDF-VI库次级中子数据的能量守恒问题

在评价核数据库ENDF VI版的出射粒子能量分布数据(即文档5)中,一些核素在阈能附近的次级中子能量分布大多采用简单的近似方法描述,例如三角形分布或矩形分布.这些简单的近似描述,将导致产生散射矩阵的"双峰"现象,或使平均次级中子能量过大而产生负KERMA(ki netic energy release in materials)因子.针对这两个现象,通过细致的分析,讨论了这些现象的产生原因,并试对相关数据做一定的修正,以改善次级中子分布的能量守恒问题. In file 5 of ENDF VI, the arbitrary tabulated function (L=1) was applied to some threshold reactions of some nuclides. The energy distribution of secondary spectrum was given approximately as a triangular shape or rectangular shape at the threshold energy for some nuclides. The triangular shape distributions would cause two peaked energy distribution near the threshold energy, which must affect the group transfer matrix. The rectangular shape make the E′n too large, that may lead to...     

核物理数据智能化管理和网络共享的研究

通过对主要的国际核物理基本数据资源收集、遴选和研究分析，建立了各类数据集合的完整数据字典，以关系模型理论和面向对象编程技术为指导，研究设计了不同种类的关系型核物理数据库的管理和检索工具，实现了由核物理基础数据、关系型数据库管理系统、数据应用服务软件和网络与数据库安全保障等部分组成的核物理数据智能化管理与共享信息系统.基于Linux+Apache+SSL+PHP+MySQL的系统配置，将在线计算技术和数据可视化研究相结合，实现了三大类型的9个关系型的国际核物理数据库的网上发布，向用户提供包括在线检索、模糊查 关键词： 核物理基础数据 关系型数据模型 数据可视化 在线数据服务     

西安脉冲堆127I嬗变计算方法与实验验证

为了开展129I的热中子嬗变的研究,在西安脉冲堆上开展了127I靶件辐照实验。以探索实验条件,对127I靶件的嬗变率进行了蒙特卡罗计算,并与实验测量值进行了比对。利用NJOY程序,以ENDF/B VII.0库为基础,制作了127I在西安脉冲堆堆芯辐照温度下的MCNP格式截面库,与MCNP自带库(ENDF/B VI.2)同温度下截面库进行了比较,在不可分辨共振区做了改进,使用新制的截面库,利用MCNP程序对ORIGEN2数据库中的127I辐射俘获截面进行了修正,结合ORIGEN2程序分析了127I靶件在西安脉冲堆实际辐照后的嬗变率和核素的变化,研究了中子能谱和辐照时间对靶件嬗变计算的影响。使用MCNPX自带的燃耗模块CINDER90对127I靶件的嬗变情况进行模拟,结果与ORIGEN2基本一致,与实验数值有2%~3%的偏差,主要原因是MCNP计算中子通量密度存在误差。     

173Lu和174gLu衰变数据的测量

175Lu（n;2n）174m,gLu（n;2n）173Lu反应在高能中子通量监测中具有重要作用,其产物的放射性活度通常用HPGe探测器测量,这就需要174m,gLu和173Lu的半衰期和γ射线发射率等衰变数据必须准确无误。在主要的评价核数据库中,只有ENDF/B 7.1和JEFF 3.1.1数据库给出了173Lu和174Lu的半衰期和γ射线的发射几率等衰变数据,其中173Lu的射线发射几率ENDF/B 7.1库比JEFF 3.1.1库总体偏高,其他评价数据两个数据库非常一致。在中国原子能研究院的串列加速器上用20 MeV质子辐照金属Yb靶生产了含有173Lu,174m,gLu的放射性溶液,用激光共振电离质谱（LRIMS）和热表面电离质谱（TIMS）两种同位素稀释法测定了该溶液中173Lu和174m,gLu核素浓度,然后制备测量源在HPGe探测器上进行了近7年的跟踪测量,发现173Lu的半衰期为1.45 a,比目前评价的数据1.37 a高6.1%,636.1 keV射线发射几率的偏差最大,比ENDF/B 7.1的评价数据偏高7%。174gLu半衰期的测量结果3.37 a175Lu（n;2n）174m,gLu（n;2n）173Lu反应在高能中子通量监测中具有重要作用,其产物的放射性活度通常用HPGe探测器测量,这就需要174m,gLu和173Lu的半衰期和γ射线发射率等衰变数据必须准确无误。在主要的评价核数据库中,只有ENDF/B 7.1和JEFF 3.1.1数据库给出了173Lu和174Lu的半衰期和射线的发射几率等衰变数据,其中173Lu的射线发射几率ENDF/B 7.1库比JEFF 3.1.1库总体偏高,其他评价数据两个数据库非常一致。在中国原子能研究院的串列加速器上用20 MeV质子辐照金属Yb靶生产了含有173Lu,174m,gLu的放射性溶液,用激光共振电离质谱（LRIMS）和热表面电离质谱（TIMS）两种同位素稀释法测定了该溶液中173Lu和174m,gLu核素浓度,然后制备测量源在HPGe探测器上进行了近7年的跟踪测量,发现173Lu的半衰期为1.45 a,比目前评价的数据1.37 a高6.1%,636.1 keV γ射线发射几率的偏差最大,比ENDF/B 7.1的评价数据偏高7%。174gLu半衰期的测量结果3.37 a,比评价数据3.31 a高约1.8%,174gLu 76.5 keV和1 241.8 keV γ射线的发射几率比ENDF/B 7.1的评价数据分别低1.87%和12.8%。175Lu(n,2n)174m,gLu(n,2n)173Lu are important neutron reactions to monitor the high energy neutron fluxes and the residual nuclei, 174m,gLu and 173Lu, are usually measured by a HPGe spectrometer conveniently so that the decay data such as half lives and gamma-ray emission probabilities of 174m,gLu and 173Lu must be accurate. There are evaluated decay data only in ENDF/B 7.1 and JEFF 3.1.1 among the major evaluated nuclear data libraries, where the most data are almost same besides the gamma-ray emission probabilities of 173Lu are higher in ENDF/B 7.1 than that in JEFF 3.1.1. Yb metal as a target was irradiated by 20 MeV proton beams on a tandem accelerator in CIAE to produce 174m,gLu and 173Lu, and 176Lu isotopic dilution methods based upon a laser resonance ionization mass spectrometer (LRIMS) and a thermal surface ionization mass spectrometer (TIMS) were employed to determine the numbers of nuclides of 174Lu and 173Lu in a solution containing the irradiated target. Several radioactive sources made from the solution had been measured by a HPGe detector during the past 7 years and the peak intensities of the characteristic gamma-rays from 174gLu and 173Lu were analyzed to determine the half lives and gamma-ray emission probabilities. According to the measurements, the half life of 173Lu is 1.45 a, which is about 6.1% longer than the evaluated 1.37 a, and the gamma-ray emission probabilities of 173Lu is also different from the evaluated data, especially the emission probability of 636.1 keV gamma-ray of 173Lu is 7% higher than the value in ENDF/B 7.1. The half life of 174gLu is 3.37 a, about 1.8% longer than the evaluated 3.31 a, and the emission probabilities of 76.5 keV and 1 241.8 keV gamma-rays of 174Lu are 1.87% and 12.8% lower than the evaluated data in ENDF/B 7.1, respectively.     

Measurements of the 10B (n,He) reaction rate in zppr

Direct measurements of the ¹⁰B(n, He) reaction rate have been made in ZPPR. The measurements were made by isotope-dilution mass spectrometric analysis of stainless-steel encapsulated samples of enriched ¹⁰B. The results demonstrate that direct measurements of the ¹⁰B reaction rate in ZPPR are obtainable to an accuracy of ～1%. Comparisons with calculations give a predicted-to-experiment value of ～0.93. This value is consistent with earlier Rockwell measurements in benchmark neutron fields and further supports a need for revision in the ENDF/B cross section for ¹⁰B above ～0.1 MeV.     

一维贫铀/聚乙烯交替系统中D-T中子诱发的~(232)Th(n,γ)反应率的测定与分析

开展了钍样品装置内钍核参数的积分中子学基础研究.参考混合堆概念设计搭建了内部放置了钍样品的一维贫铀/聚乙烯交替系统装置,采用加速器D-T中子源模拟聚变堆芯,利用前期开发的离线伽马测量方法测定了不同位置、不同中子谱情况下的232Th(n,γ)反应率,不确定度约为5%.结果显示,聚乙烯对14.1 MeV中子的慢化作用可有效提升钍俘获率,且贫铀对钍俘获率也有显著提升作用.实验结果与主流核数据库计算结果的对比显示,ENDF/B-VI.6和JENDL-3.3数据库的计算值比实验值平均约大6%,而较新的ENDF/B-VII.0数据库的计算值比实验值平均约大4%.因此,相比于之前数据库的钍核数据,ENDF/B-VII.0的计算值与实验结果匹配得较好,可作为相关概念设计的推荐核数据库.     

钍铀循环关键核素的多群数据库加工与初步验证

钍基先进核能系统的发展对核数据提出了新的需求。钍铀循环关键核素的多群数据库是堆芯物理设计与分析计算的基础。因此,基于权威核数据处理程序NJOY对ENDF/B-VII.1评价核数据库进行处理得到WIMS格式多群截面数据,利用专用更新处理程序WILLIE得到WIMS格式数据库,并利用多群堆芯计算程序WIMSD-5B对数据库进行了一系列临界基准检验,结果表明:基于ENDF/B-VII.1加工的WIMS库与WIMS库更新计划（WLUP）的网站上发布的最新版本WIMS库的计算结果基本一致,并且,在16个钍铀循环基准题检验中,新加工的WIMS库计算平均误差要比WIMSD5B程序的自带WIMS库的计算结果小0.225 3%,精度更高,可靠性更好。