Measurement of neutron cross sections and resonance parameters of ^169Tm below 100 eV

The neutron total cross-sections of thulium （^169Tm） were measured in the neutron energy region from 0.01 eV to 100 eV by using the time-of-flight method at the Pohang Neutron Facility, which consists of an electron linac, a water-cooled tantalum target with a water moderator, and a 12 m time of flight path, Two thulium plates with different thicknesses were used for the neutron transmission measurement. The background level was determined by using a notch-filter of Co, In, and Cd sheets. The present measurement was compared with the previous ones, and a new set of resonance parameters of ^169Tm isotope was obtained from the transmission rate by using the SAMMY code, with a comparison with the recommended parameters by Mughabghab.     

钍基熔盐堆用238群中子-48群光子耦合多群常数库的屏蔽基准验证

采用OKTAVIAN脉冲球实验对钍基熔盐堆用AMPX主库格式238群中子-48群光子耦合多群常数库进行了屏蔽基准验证,重点检验了该库中的F,Li,Be,C、Al,Si,Cr,Ni,Zr,Co,Cu,Mn,Mo,Nb,Ti,W,Pb同位素/元素的数据。采用SCALE 5.1程序系统中的XSDRN-PM程序进行一维屏蔽问题计算,将计算结果与实验测量数据及MCNP程序计算结果进行比较,发现中子泄漏谱的符合程度较好,而光子泄漏谱检验中发现大多数核素都出现了不同程度的高估。通过对GENDF格式到AMPX格式的转换程序MILER-4进行修正,解决了这一问题。通过对多群常数库的屏蔽基准验证,进一步证明了该库的可靠性。OKTAVIAN pulsed sphere experiment was used for Shielding Benchmarks of the AMPX formatted multi-group (238n-48γ) coupled neutron-gamma cross-section library for Thorium Molten Salt Reactor, of which the following isotopes/elements were checked-F, Li, Be, C, Al, Si, Cr, Ni, Zr, Co, Cu, Mn, Mo, Nb, Ti, W. One dimension shielding problem was calculated using XSDRN-PM program of SCALE 5.1 code system and results were compared with experiment results and MCNP calculated results, which shows that neutron leakage spectra agree well. Calculated results of photon leakage spectra of most facilities compared with MCNP results and experiment data are over-rated. MILER-4 code which is used for converting GENDF files produced by NJOY to the AMPX master library format is revised to solve this problem. The shielding benchmark verifications confirm the reliability of this new library.     

ISOL靶－源中铀靶的放射性活度研究

用3种方法计算在线同位素分离器(ISOL)靶－源中铀靶在100MeV, 200μA强流质子束照射下产生的放射性核素活度及γ射线强度随时间的变化, 为铀靶的设计、更换及后期处理提供一定的设计依据.     

Measurement of neutron cross sections and resonance parameters of ~(169)Tm below 100 eV

The neutron total cross-sections of thulium(169Tm) were measured in the neutron energy region from 0.01 eV to 100 eV by using the time-of-flight method at the Pohang Neutron Facility,which consists of an electron linac,a water-cooled tantalum target with a water moderator,and a 12 m time of flight path.Two thulium plates with different thicknesses were used for the neutron transmission measurement.The background level was determined by using a notch-filter of Co,In,and Cd sheets.The present measurement was ...     

Neutron activation cross section data library

To satisfy the requirements of nuclear reaction cross sections in nuclear engineering applications and nuclear physics studies, the Neutron Activation Cross Section Data Library has been established. 818 target nuclei including unstable target or isomeric target nuclei are considered in this library. The induced neutron energy range region is between 10^-5 eV and 20 MeV. The standard ENDF-6 format is adopted, including general information, reaction cross sections, multiplicities, and so on. The recommended reaction cross sections were obtained using UNF code system and FDRR nuclear model codes or systematic analysis based on available experimental data. The full evaluated dataset containing the evaluated activation cross sections is openly available at http://doi.org/10.57760/sciencedb.j00113.00024.     

232Th 中子诱发裂变产额数据比对评价

232Th 中子诱发裂变产额在Th/U 燃料循环的核能开发中具有重要的意义,在现有国际评价产额数据库基础上,结合最新的实验数据,对产物核的质量分布和反应堆计算所需的重点产物核进行了比对。这些产物核包括95Mo,99Tc,101Ru,103Rh,109Ag,133Cs,147Sm,149Sm,150Sm,151Sm,152Sm,143Nd,145Nd 和153Eu 等,对有分歧的133Cs,152Sm 和153Eu 累积产额数据进行了评价调整。最后推荐给出了n+232Th 裂变的产额数据库,作为CENDL 库的组成部分。The yields of n+232Th fission are important in study of the Th/U fuel cycle. Based upon the evaluated nuclear data libraries and recent experimental data,comparison and evaluation for yield mass distribution and cumulativeyield were performed for those products are needed and important in reactor calculations,including 95Mo,99Tc,101Ru,103Rh,109Ag,133Cs,147Sm,149Sm,150Sm,151Sm,152Sm,143Nd,145Nd adn 153Eu etc. Adjustments were made for 133Cs, 152Sm and 153En which have discrepancies among the libraries. Database for n+233Th fission yield is recommended as a part of CENDL update.     

一维快装置中心反应率模拟计算的灵敏度和不确定度分析

采用广义微扰理论和离散纵标输运计算方法,推导了中心反应率比值对重要核反应参数的灵敏度的计算公式。以一维快基准装置Jezebel-Pu239为例,计算了其中心反应率比值对关键核数据的灵敏度,并结合多群协方差数据,分析了其中心反应率比值计算中由于核数据的不确定度所引入的不确定度及其来源。根据计算结果以及评价数据的协方差信息,建议对Pu-239的非弹截面数据进行更精细的评价,以期降低反应率比的模拟计算不确定度。On the base of the generalized perturbation theory and SN method, the theoretical model for calculating the sensitivity and uncertainty of reaction rates ratio (RRR) versus nuclear data is developed. Taking Jezebel-Pu239 critical benchmarking facility as an example, the sensitivity of F49/F25 ratio versus important nuclear data is calculated, as well as the uncertainty arising from the nuclear data. Basing on the sensitivity and uncertainty analysis as well as the covariance information of input evaluated data, refined evaluation of the inelastic data of Pu-239 is considered as a feasible measure to reduce the uncertainty of reaction rates ratios.     

新一代堆芯多群常数库制作系统的研制

随着新一代组件计算程序计算精度要求的不断提高,组件计算程序对配套多群常数库提出了更高的要求,例如:要求能群结构更加精细、共振参数更加多样、燃耗数据更加精确等。为满足新一代组件计算程序的这一系列需求研发了一套制作多群常数库的系统NPLC-3。NPLC-3系统以NJOY程序为中心包含了输入参数库、驱动程序、主库制作程序、并群、并库、删除、添加、进制转换、数据自动检验等功能模块,针对各功能模块采用不同方法初步验证了系统的正确性。相对传统的数据制作方法,NPLC-3系统在燃耗链设计、燃耗数据计算以及共振数据制作等方面增加新的特性,而且系统采用独立的I/O接口、全自动读取参数库并生成输入卡,相对过去制作数据库的手段有了很大的改进。With the unceasing enhancement of the calculation accuracy requirement on the new generation of lattice code, higher requirements on multi-group constants library were put forward. For example, fine energy group structures are required, more types of data are need for the resonance processing, more accuracy burn-up data is required. To meet this series of demand, a multi-group constants production system NPLC-3 was developed. The NPLC-3 system which mainly based on the NJOY program contains an input parameter database and a series of functional codes such as driver code, main library production code, energy group collapsing code, work library production code and so on. Recently, different methods were adopted to validate these codes respectively according to their functions. Compared to the traditional methods in multi-group constants production, NPLC-3 system adopts several new methods in the design of burn up chain, calculation of burn up data and resonance parameters. What's more, the NPLC-3 system has an independent I/O interface, and can fully automatic generate input cards from the input parameters database. Relative to the past means of production library, NPLC-3 system has great improvement.     

20MeV以下快中子与56Fe非弹性散射截面的分歧研究

56Fe的非弹性散射截面在核装置中子输运计算中扮演着重要的角色,但无论从实验数据还是从评价数据,非弹性散射截面都存在很大分歧,它的数据直接影响到核装置的设计、建造与运行维护。本工作从实验数据本身出发,深入分析了不同实验室测得的847 keV的γ产生截面的分歧,经转化后补充非弹性散射截面的实验空白能区,并同时利用满足全截面、去弹截面等截面自洽关系的评价方法推荐了高精度的快中子与56Fe的非弹反应截面结果。积分检验表明,新的非弹截面的改进使得评价数据与积分实验结果一致,较CENDL-3.1的评价数据结果有显著改善。Knowledge about the inelastic scattering cross section of 56Fe is very important in neutron transportation calculation. However there are great discrepancies not only between experimental data but also between evaluated data. More detail analysis was performed for inelastic scattering cross section in the fast range up to 20 MeV where there are significant differences among the main evaluated libraries, mainly caused by the different inelastic scattering cross section measurements. The large discrepancies on 56Fe(n, n₁'γ) cross section which could fill the neutron energy blank of the 56Fe(n,inl) were clarified and were converted to the inelastic scattering cross section of 56Fe. And the high-quality results were evaluated by using the unitarity constrain among total cross section, noelastic reaction and other reactions. The integral experiment result indicates that the new evaluated result of inelastic cross section brings greater improvement than that of CENDL-3.1.