组件计算程序PANDA研发及初步验证

PANDA是上海核工程研究设计院研发的压水堆组件计算程序。该程序采用基于特征线方法（MOC）的一步化计算流程,即在不引入能群压缩和栅元均匀化的情况下直接进行组件层面的两维非均匀输运计算。多群数据库采用基于ENDF/B-VI制作的70群结构中子数据库,基于ENDF/B-VII的新版数据库也正在开发中。共振自屏计算采用了空间相关丹可夫方法（SDDM）,既具备燃料芯块分区计算的能力,又保留了传统Stammler方法的计算效率。多群非均匀输运计算采用二维模块化MOC方法,并辅以双重粗网有限差分(CMFD)加速技术,具有良好的计算精度和效率。对传统线性子链解析（TTA）方法以及多种矩阵指数方法进行了研究,选取了适合PANDA程序燃耗链的燃耗方程求解技术。基于以上基本模型开发了PANDA程序,并从程序模块、总体集成和核设计程序系统确认等三个层面,初步验证了PANDA程序的计算性能,表明了PANDA程序的工程设计计算能力。     

温度相关核截面数据库在MCNP计算中的必要性研究

MCNP程序由于其几何模拟和核数据上的优越性,现在在反应堆的研究分析中已经得到较多应用。通过基准题的计算,定量地说明MCNP通过其自带的常温（294K）下的核素截面数据库不能够对反应堆进行非常准确的计算（由于反应堆内各种材料／位置的温度不同）,而且,它也不能够计算反应堆中与温度相关的量,如反应性温度系数。选用了一个带有不同温度下核素截面数据的MCNP输入格式的数据库,使用MCNP-4C对基准题进行了计算,发现计算结果与基准值符合得非常好。这说明通过使用不同温度下的核素截面数据库,MCNP可以准确计算温度系数和增殖系数等,从而说明在反应堆设计计算中制作不同温度下的核素截面库的必要性。Due to the advantage of geometry simulation and nuclear data, the code MCNP is now widely used in the reactor analysis. Based on our calculation of the fuel temperature reactivity coefficient benchmark, it is quantificationally proved that MCNP with its own cross section library can＇ t be used to simulate the reactor accurately and to calculate the temperature reactivity coefficient. Furthermore, we use MCNP- 4C with a database that contains temperature dependent nuclear cross sections to calculate the benchmark. The results are well agreement with benchmark results. This means that, with the temperature dependent nuclear cross sections library, MCNP can calculate the temperature reactivity coefficient and reactor multiplication factor accurately. So the temperature dependent nuclear cross section library should be processed to meet the requirement of reactor calculation.     

快堆中子学计算程序NECP-SARAX的改进与验证

NECP-SARAX是西安交通大学NECP团队开发的用于快中子反应堆的中子学程序系统。为准确处理快中子反应堆中中等质量核素散射共振以及非弹性散射导致的复杂的中子慢化效应,SARAX程序最初采用连续能量的蒙特卡罗方法计算中子能谱从而获得堆芯计算使用的有效多群截面。由于蒙特卡罗程序计算效率低,且在低能量段统计偏差较大,提出采用基于点截面的超细群方法计算中子慢化能谱,避免了蒙特卡罗方法产生参数时存在的缺陷。堆芯计算采用多群中子输运,通过优化简化几何建模,改进了程序的实用性。采用多种微扰方法计算堆芯各种反应性系数,提出了基于中子输运微扰理论的虚拟密度方法以计算堆内组件变形导致的反应性变化。在进行堆芯瞬态计算时,采用了点堆和改进准静态两种方法,可用于一般快堆和快谱ADS的典型事故分析。OECD发布的一系列快堆基准题测试表明,SARAX程序在快堆计算中具有良好的精度,达到了与国外著名快堆程序相当的水平。有效增殖因子与连续能量的蒙卡计算结果相比偏差在300 pcm以内。同时,由于引入了虚拟密度理论和三维时空动力学模型,程序功能更加完善,可以更好地满足快堆工程设计的需求。     

Dragon程序在金属燃料铅铋快堆堆芯计算中的应用与偏差分析

铅铋合金或铅冷却快堆（LFR）是具有良好应用前景的第四代先进核能系统之一。针对环形芯体金属燃料（UZr，UPuZr）LFR的燃料组件与堆芯，利用Dragon/Donjon程序开展中子学计算，获得了基于ENDF/B 8.0库的172群和295群多群中子数据库、输运方法（SP3）和扩散方法（MCFD）的结果及其与蒙卡程序RMC的偏差。采用SP3算法针对UZr燃料得到的k_eff偏差小于550×10?5；对于UPuZr燃料采用MCFD算法得到的k_eff偏差小于?700×10?5。控制棒组件价值的偏差小于7.6%；172群和295群库的结果基本无差异。应用SP3算法的燃料组件功率偏差小于±6.0%；SP3算法的偏差小于MCFD的。结果证明，Dragon/Donjon程序在金属燃料铅铋快堆物理分析中具有可行性。     

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...     

基于VERA基准题的cosRMC程序验证

为了验证反应堆物理软件和方法的计算能力,美国CASL (Consortium for Advanced Simulation of LWRs) 项目提出了VERA (Virtual Environment for Reactor Application) 堆芯物理基准题。该基准题以Watts Bar初始堆芯为模型,涵盖从二维单栅元到三维全堆芯的燃耗及换料的十个基准问题。针对VERA基准题模型,利用COSINE软件包中的反应堆蒙特卡罗分析程序cosRMC进行临界计算,得到了有效增殖因子、组件功率分布、控制棒微积分价值和反应性系数等结果。通过与基准题中提供的KENO结果对比,两种蒙特卡罗程序的计算结果吻合良好。这表明cosRMC程序具有从组件到堆芯的计算能力,其临界计算精度基本与KENO程序相当。     

TREND code for reconstruction of neutron reaction cross sections and calculation of the Doppler broadening in the region of resolved resonances

The TREND code is developed for reconstruction of neutron cross sections in the region of resolved resonances from the data of the international evaluated nuclear data libraries using the Reich-Moore, Adler-Adler, and Breit-Wigner formalisms and for calculation of the Doppler broadening of the resulting cross sections within the classical approximation. The TREND code is incorporated into the MCU code package for Monte Carlo reactor calculations. The TREND and MCU codes have been used to evaluate the data of the ROSFOND data bank on uranium and plutonium isotopes in comparison with the data of the ICSBEP data bank.     

氮化铀热中子截面的第一性原理计算

氮化铀（UN）因其较好的热物性和耐事故容错性成为先进动力堆的候选燃料,但目前热能区缺少可靠的UN热中子截面数据,这对于热中子反应堆物理计算是很不利的.本文基于量子力学的第一性原理,利用VASP/PHONON软件模拟计算了UN的声子态密度,以此为积分得到UN的定容比热容,并基于新制作的声子态密度,采用核截面处理程序NJOY/LEAPR,利用热中子散射理论,得到UN的S（α,β）数据,进而研究UN的热中子散射截面,并与传统压水堆的二氧化铀（UO₂）进行对比.结果表明：优化的晶格参数与数据库符合较好,UN声子态密度的声子项和光子项较UO₂的分隔更加明显,定容比热容计算结果与实验值一致,基于该声子态密度计算得到的UN中²³⁸U的非弹性散射和弹性散射截面比相同温度下UO₂中²³⁸U小,UN中N仅考虑了非相干散射部分,随着温度升高,UN弹性散射截面变小,非弹性散射变大,并在高能段趋于自由核散射截面.本文的研究结果填补了UN热中子截面数据的缺失,为下一步系统研究UN燃料在轻水堆中的中子学性能奠定了基础.     

Measurement of differential cross sections of neutron-induced deuteron production reactions on carbon from 25 to 52 MeV

The angle-differential cross sections of neutron-induced deuteron production from carbon were measured at six neutron energies from 25 to 52 MeV relative to those of n-p elastic scattering at the China Spallation Neutron Source(CSNS) Back-n white neutron source.By employing the ΔE-E telescopes of the Light-charged Particle Detector Array(LPD A) system at 15.1° to 55.0° in the laboratory system,ratios of the angle-differential cross sections of the ~(12)C(n,xd) reactions to those of the n-p scattering were measured,and then,the angle-differential cross sections of the ~(12)C(n,xd) reactions were obtained using the angle-differential cross sections of the n-p elastic scattering from the JENDL-4.0/HE-2015 library as the standard.The obtained results are compared with data from previous measurements,all of which are based on mono-energic neutrons,the evaluated data from the JENDL-4.0/HE-2015 library and the ENDF-B/Ⅷ.0 library,and those from theoretical calculations based on INCA code and Talys-1.9 code.Being the first white-neutron-source-based systematic measurement of the angle-differential cross sections of neutron-induced deuteron production reactions on carbon in several tens of MeV,the present work can provide a reference to the data library considering the lack of experimental data.     

基于随机抽样法的多群核数据不确定性影响分析

基于随机抽样方法,研究多群核数据不确定性对反应堆物理计算的影响。首先利用SCALE软件包中核数据协方差矩阵和自主开发的随机抽样模块SAMP,得到多群微观截面等核数据的抽样值,之后分别使用SCALE/TRITON和PARCS程序进行组件计算及堆芯稳态计算,最后通过统计分析得到组件和堆芯计算结果的不确定度。以Almaraz压水堆核电厂装载的燃料组件和首循环堆芯为对象,研究了不同燃耗下有效增殖因子、动力学参数、核素浓度和双群均匀化宏观截面等组件计算结果,以及堆芯功率分布等堆芯计算结果的不确定度。分析结果表明:组件计算结果不确定度多随燃耗变化,快群宏观截面不确定度总体高于热群;堆芯计算结果受核数据不确定性影响显著,其中稳态径向功率分布的最大不确定度为1.9%左右。     

基于非线性迭代的压水堆瞬态计算程序开发

采用目前工业上成熟的非线性迭代计算策略,基于两群粗网有限差分方法和多群UNM节块方法,开发了针对压水堆工况的三维瞬态扩散计算程序。UNM方法采用解析基函数作为基函数,通过方程变换解决了解析节块法在临界节块计算不稳定的问题,提高了计算精度。热工计算采用单通道模型和燃料棒一维导热模型,相比目前堆芯在用的经验关系式方法,该模型可以更加准确地计算燃料棒温度分布。采用基于横向积分方程的三节块方法,可以有效减轻控制棒尖齿效应对瞬态计算的影响。为测试程序性能,采用NEACRP等基准算例对程序进行了校验。数值结果表明,开发的程序计算结果正确,适用于压水堆堆芯瞬态过程的模拟。     

Elastic scattering of metastable He(21,3 S) atoms by Na(32 S) atoms at 68 meV

The spectroscopy of metastable states is used to make the first experimental measurement of the total differential elastic-scattering cross sections of metastable helium atoms in 2³ S and 2¹ S states by sodium atoms in the ground state at interaction energy 68 meV in the center-of-mass system. To analyze the experimental data, the partial scattering phases are calculated using the method of phase functions in the optical potential approximation. The analysis makes it possible to give a more detailed interpretation of the structure of the differential cross section. The computed integral cross sections, specifically, Penning ionization, diffusion, viscosity, and spin exchange are discussed.     

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.     

基于线性核素链的燃耗算法与蒙特卡罗程序耦合计算

根据线性核素链原理和反应堆燃耗的特点,建立相应核数据库并采用回溯算法生成自适应核素链,完成核素的遍历和计算,形成多群点燃耗计算程序,能够独立进行燃耗计算并具有完整的输出结果形式。同时将所开发的燃耗计算程序与蒙特卡罗程序进行耦合,完成接口模块的设计,形成可用于研究堆和核电站的燃耗-输运耦合计算工具。将所开发的输运-耦合计算程序应用于中国实验快堆首炉堆芯燃耗的计算,将计算结果与现有设计数据进行比较。经过初步分析表明:新的燃耗计算程序能够精确计算锕系核素含量,包括易裂变核素和生成量很小的次锕系核素。对某些锕系核素如Pu241的计算结果还存在较大偏差,这需要对计算结果进一步分析,来确认偏差是来自计算过程还是相关的截面数据。整个燃耗-输运耦合计算系统对裂变产物的处理和反应性变化的计算也与现有的设计数据符合良好。     

The discrete-dipole-approximation code ADDA: Capabilities and known limitations

The open-source code ADDA is described, which implements the discrete dipole approximation (DDA), a method to simulate light scattering by finite 3D objects of arbitrary shape and composition. Besides standard sequential execution, ADDA can run on a multiprocessor distributed-memory system, parallelizing a single DDA calculation. Hence the size parameter of the scatterer is in principle limited only by total available memory and computational speed. ADDA is written in C99 and is highly portable. It provides full control over the scattering geometry (particle morphology and orientation, and incident beam) and allows one to calculate a wide variety of integral and angle-resolved scattering quantities (cross sections, the Mueller matrix, etc.). Moreover, ADDA incorporates a range of state-of-the-art DDA improvements, aimed at increasing the accuracy and computational speed of the method. We discuss both physical and computational aspects of the DDA simulations and provide a practical introduction into performing such simulations with the ADDA code. We also present several simulation results, in particular, for a sphere with size parameter 320 (100-wavelength diameter) and refractive index 1.05.     

Angular distributions of scattering intensities with the SAP code

Compton and Rayleigh scattering peak intensities and their ratio are used in reflection and transmission experiments to obtain information about the density of the investigated specimen. The ratio is preferred because it allows the reduction of the errors due to attenuation and geometry. In all cases it is fundamental to predict their angular distributions in order to design the optimal experiment for a given material. The code SAP (Scattering Angular distribution Plot) is a graphical tool to compute and plot Rayleigh and Compton differential cross‐sections (atomic and electronic), form factors and incoherent scattering functions. In this work, the code is improved by adding the computation of Rayleigh and Compton first‐order peak fluxes and intensities, and the Rayleigh‐to‐Compton peak ratio, in both, reflection and transmission geometries, for single elements, compounds and mixture of compounds, for monochromatic excitation in the range of 1–1000 keV. The new characteristics of the code are illustrated with some examples. Copyright © 2011 John Wiley & Sons, Ltd.     

西安脉冲堆堆芯燃料管理正交优化计算

根据西安脉冲堆工程的实际需要,研制了脉冲堆堆芯燃料管理计算软件包HEX ICFM,并建立了正交优化模型,研制了换料优化计算软件包HEX ORTH.软件包中栅元计算采用了WIMS D 4程序,堆芯扩散计算程序采用了二维六角形节块程序SIXTUS 2.应用HEX ICFM对西安脉冲堆首循环堆芯参数进行了计算,并用HEX ORTH对第3循环末堆芯燃料装载和30根备用新燃料元件进行了堆芯优化分析,得到了目标函数为max(K_eff^BOC)的最佳堆芯倒换料方案.     

Extension of the Monte-Carlo code MOCADI to fusion-evaporation reactions

The Monte-Carlo code MOCADI has been extended, besides to fission and fragmentation reactions, also to fusion-evaporation reactions. The updated version includes the cross sections and the reaction kinematics for the production of evaporation residues calculated according to the statistical model code PACE2, while the atomic interaction with matter can be treated either with the code ATIMA or with the code TRIM. The present version of MOCADI can now be employed world-wide for design studies of new low-energy nuclear physics facilities and for the optimization of experiments involving fusion-evaporation reactions, e.g. mass measurements, laser spectroscopy, super-heavy synthesis.     

中子时空动力学计算程序CORCA-K的进展

主要介绍了自主化中子时空动力学程序CORCA-K的相关研发进展。CORCA-K程序是中核集团自主研发的NESTOR软件包中的三维瞬态中子学计算软件。其核心功能是采用数值计算方法,求解三维瞬态中子扩散方程。其空间离散采用第二类边界条件节块格林函数方法,与三维稳态扩散计算程序CORCA-3D保持自洽性;时间离散采用对角线隐式龙格库塔格式,可在向后欧拉格式的基础上叠加实现,且具备时步自适应功能。以三维LMW基准问题、三维LRA基准问题和秦山核电厂3号机组第5循环为计算对象,开展了瞬态计算验证。结果表明,CORCA-K计算结果与参考程序符合较好,且在保证计算精度的条件下,具有更高的计算效率。     

Precise predictions for W+4-jet production at the large hadron collider

We present the next-to-leading order (NLO) QCD results for W+4-jet production at hadron colliders. This is the first hadron-collider process with five final-state objects to be computed at NLO. It represents an important background to many searches for new physics at the energy frontier. Total cross sections, as well as distributions in the jet transverse momenta, are provided for the initial LHC energy of √s = 7 TeV. We use a leading-color approximation, known to be accurate to 3% for W production with fewer jets. The calculation uses the BLACKHAT library along with the SHERPA package.