MCNP温度相关中子截面库制作方法

在截面库研制过程中,着重考虑了在反应堆设计与运行温度范围的温度点;使用NJOY软件将ENDF格式的中子截面文件处理为ACE (A Compact ENDF) 格式的点截面文件,供MCNP程序使用.验证过程应用了3种不同类型的临界基准题:简单的球形几何基准题、板式燃料元件实验装置和带有可燃毒物的堆芯.结果表明,3种临界基准题所得到的验证结果均较为理想,在精确度方面也达到了要求.证明了使用NJOY制作截面库方法的正确性.     

MCNP温度相关中子截面库的研制及基准验证

本文在使用NJOY软件由ENDF格式的中子截面文件处理生成ACE (a compact ENDF) 格式的温度相关中子截面库的方法研究的基础上,开展温度相关中子截面库的研制及验证.研制过程中,选择了在反应堆设计和运行温度范围内的16个温度点.在温度相关中子截面库的验证过程中应用了4个基准题:带可燃毒物的轻水堆芯临界基准题、反应性多普勒系数基准题、标准CANDU组件燃料温度系数基准题和VHTRC温度系数基准题.验证计算结果表明,该温度相关中子截面库可运用于反应堆物理的计算分析中.     

基于OpenMC的多群截面库制作及有效性验证

OpenMC是麻省理工大学计算反应堆物理组开发的开源蒙特卡罗程序,能够方便地制作适用于特定堆芯中子能谱分布的多群反应截面及高阶勒让德散射截面以用于离散坐标输运程序ANISN的计算。本文基于ENDF/B-Ⅶ.1和CENDL-3.1评价数据库,利用OpenMC计算制作了ANSIN格式的多群截面并通过基准题的计算验证计算结果的准确性。通过截面转换程序的编写,将OpenMC给出的堆芯各阶勒让德散射分量,堆芯中子能谱分布,散射、吸收反应率以及裂变中子产生速率等信息转换为ANISN程序可读取的截面库格式。采用制作的截面库利用ANINS计算有效中子增殖因子及堆芯中子通量分布。结果表明,ANISN确定论的计算结果与OpenMC给出的蒙特卡罗计算结果相吻合,验证了这种方法可有效地为ANISN提供截面数据,将来可推广应用于二维、三维确定论中子输运计算。     

基于ENDF/B-VII.0评价库的多群参数库MUSE1.0的开发与初步验证

采用NJOY程序研制了基于ENDF/B-VII.0评价库的172群中子-42群光子多群截面库(MUSE1.0),该库的权重谱采用Vitanim-e谱,角分布采用勒让德P6近似;热散射数据由自由气体模型产生,共振自屏修正选择了10组背景截面。该库含有293、600、800、900 K等温度下的截面数据;采用GENDF、MATXS和ACE多群3种格式存储。采用MCNP程序,从临界计算和屏蔽计算两个方面对该库进行较全面检验。结果表明,MUSE1.0在临界计算以及屏蔽计算方面具有较强的通用性,对于热散射效应以及共振自屏效应具有较好地描述能力,可以满足超临界水堆概念设计研究方面的应用要求。     

CENDL-3.2与ENDF/B-Ⅷ.0的56Fe评价截面对屏蔽计算影响研究

CENDL-3.2评价库对56Fe非弹性散射截面进行了更新,为了验证其与ENDF/B-Ⅷ.0评价库中截面以及屏蔽计算能力的差异,通过NJOY2016程序对56Fe共振重造后的非弹性散射、总截面等微观截面进行了比较;并制作了多群截面,在56Fe非弹性散射能量范围对以56Fe为主要核素的3个系列屏蔽基准题ILL-Fe、OKTAVIAN-Fe、IPPE-Fe进行了屏蔽计算性能的比较。结果表明,CENDL-3.2评价库的非弹性散射截面在4~12 MeV能量范围内低于ENDF/B-Ⅷ.0评价库的结果;多群截面基准题验证表明,CENDL-3.2评价库计算结果与实验值总体符合较好;对于OKTAVIAN-Fe基准题,在0.1~1 MeV能量范围内两评价库计算结果吻合较好。此外,所有基准题验证结果都有共同的现象,即在56Fe非弹性散射截面占主要贡献的1~10 MeV能量范围内,CENDL-3.2的计算结果比ENDF/B-Ⅷ.0的计算结果偏高。      

基于CENDL-3.2的多群截面库含Be快临界基准分析

中国评价核数据库最新版CENDL-3.2(Chinese Evaluated Nuclear Data Library)已于2020年6月发布,对包括核工程计算中常用的235U、238U、239Pu、56Fe等134个核素的中子反应数据重新进行了评价和计算,与CENDL-3.1相比,CENDL-3.2数据种类和数据质量均有大幅提高。Be由于其散射截面大、吸收截面小,常被用作熔盐堆燃料载体盐成分之一,其反应截面数据的准确性在熔盐堆设计中不容忽视。基于CENDL-3.2评价核数据库,采用NJOY制作了199群中子、42群光子的MATXS格式多群截面库,挑选了35个含Be快临界基准对其进行检验分析,并与基于ENDF/B-7.1和JENDL-4.0的多群截面库计算结果进行对比。分析表明：基于CENDL-3.2多群截面库计算的26个基准题（74.29%）的结果与实验值偏差在0.5%以内,整体上优于ENDF/B-7.1和JENDL-4.0。表明CENDL-3.2中的Be数据和基于CENDL-3.2的多群截面库及其制作方法是可靠的,能够用于熔盐堆相关设计计算。     

基于ENDF/B-Ⅶ.1库的AHD1.0加工与基准验证

核数据库是中子输运计算的基础。基于ENDF/B-Ⅶ.1评价库,采用NJOY制作了用于MCNP(Monte Carlo N-Particle Transport Code)程序的AHD1.0(Advanced hybrid database1.0)库,并从国际临界核安全手册(International Handbook of Evaluated Criticality Safety Benchmark Experiments,ICSBEP)中分别选取了高浓铀、中浓铀、低浓铀的快谱、中间谱及热谱的部分基准装置,用MCNP程序调用该数据库进行了临界基准验证,验证结果显示:调用该库的计算值与实验值符合较好,误差在0.5%以内,具有较高的精确度,满足核设计对数据库精度的要求。但对于部分含有W、Fe、Gd等结构材料、吸收材料的基准检验中,存在较大的偏差,造成这些偏差的主要原因是计算过程中核素的处理及评价数据库的来源,需要进一步的研究验证。     

TMSR-SF高温下多群核数据库的研究

基于最新发布的评价核数据库ENDF/B-VII.1,简要介绍了利用标准程序NJOY加工固态燃料钍基熔盐堆(Thorium Molten Salt Reactor-Solid Fuel,TMSR-SF)中子能谱测量所需温度下多群截面库的过程。详细分析了两个典型的核素加工所得核反应道的多群截面与温度的关系,并将不同温度下的截面库用于中子能谱测量,分析了中子能谱测量结果的误差与温度所引起截面库变化的关系。结果表明,不同类型核反应道的截面所受温度影响不同,特别是核素对超热中子的截面存在共振峰问题受温度影响最大,这是由于多普勒效应影响,所以中子能谱测量结果受核反应道选择的影响符合物理规律,加工所得873 K下的核截面库可用于TMSR-SF相关中子能谱测量。     

基于ENDF/B-Ⅶ.0核评价库的ACE格式参数制作与初步检验

2006年发布的ENDF/B-Ⅶ.0评价库较2001年发布的ENDF/B-Ⅵ.8有许多改进之处,与临界积分检验装置实验结果符合得更好.采用NJOY程序及ENDF/B-Ⅶ.0评价库中子入射子库,制作了MCNP系列程序使用的ACE格式参数.阐述了新评价库的特点和参数库的制作过程,重点叙述了对参数库的检验.结果表明,制作的参数库是正确的,可供用户使用.     

ASCFR1.0/MC的研制及其在ASCFR固体慢化剂温度效应计算中的初步应用

为研制出适用于改进型超临界水冷快堆的精确截面库(ASCFR1.0/MC),并将其应用于温度效应的计算中,首先使用快中子临界基准题JEZEBEL对截面加工程序(NJOY)中各模块的重要参数进行敏感度分析,详细比较不同输入参数对NJOY加工时间以及MCNP程序计算精度的影响,从而选择最为合理的输入参数。在此基础上,以2011年9月发布的ENDF/B-VII.1为基础库研制ASCFR1.0/MC,并针对该库应用多普勒反应性系数基准题进行基准验证。验证结果表明,ASCFR1.0/MC库的计算精度非常理想。最后针对改进型超临界水冷快堆(ASCFR)的固体慢化剂进行温度反应性系数的初步计算,发现ASCFR呈现正的慢化剂温度效应。     

Bamboo-Lattice程序多群截面数据库的改进

本文基于ENDF/B-Ⅶ.0核评价数据库,利用核数据加工处理程序NJOY及LATTICE_PRE为Bamboo-Lattice程序研制了一套改进后的多群截面数据库NECL2.0。基于基准题和数值分析的结果表明：采用NECL2.0数据库计算得到的燃料组件的k_inf、裂变率分布、少群均匀化截面与参考解均吻合很好;考虑银铟镉共振对k_inf的计算精度可提高近1000 pcm,与参考解相比最大裂变率相对偏差从-0.97%降低到-0.53%;考虑包壳锆的共振对k_inf的计算精度可提高约60 pcm。     

不同核评价数据库中Pb和Bi的截面对铅基快堆k_(eff)的影响

铅基快堆是GIF官方发布的第四代核能系统堆型之一,不同的核评价数据库中铅截面的较大差别会影响铅基快堆物理设计计算的准确性。本文利用国际上最新发布的核评价数据库JENDL-4.0、JEFF-3.2、ENDF/B-Ⅶ.0和BROND-3.1,制作了关键核素Pb、Bi的连续点截面,利用国际基准题评价手册中的PMF035和国际原子能机构发布的铅基快堆RBEC-M基准题以及cosRMC程序,对Pb和Bi的截面对系统有效增殖因数的影响进行了详细研究。对于PMF035带Pb反射层的临界基准题,上述所有核数据库的新版本较旧版本的计算偏差均有所减小,其中BROND的改善最为明显。对于RBEC-M基准题,使用ENDF/B-Ⅶ.0核数据库的计算结果与基准报告中结果符合较好;使用上述其他新版本数据库中截面数据替换计算结果表明,采用不同库中的Pb、Bi截面数据会使计算结果出现不同的偏差,其中,JENDL-4.0中Pb截面对计算结果的影响较Bi截面的影响大。     

Spectrum-Averaged One-Group Cross Sections of Actinides Based on JENDL-3

When uranium vapor is generated with an electron beam evaporator, a uranium plasma is formed on the evaporating surface. This plasma rises and expands with the vapor. Propagation behavior of this plasma was investigated by measuring plasma parameters, drift energy of ions and vapor flux along the propagation path. Over the range of 20-50 cm from the evaporation surface, the plasma density decreased from 3 × 10⁹ cm^?3 to 3 × 10⁸ cm^?3, while the electron temperature had a constant value of 0.29 eV. When the space potential was lowered from 1.48 to 0.80 V, the plasma ions were accelerated to increase the drift energy from 1.50 to 2.14 eV. Validity of the Boltzmann electron distribution was checked by comparing the space potential distribution with the plasma density distribution, and also the floating potential distribution with the ion flux distribution. These results confirm that the ambipolar diffusion governs the plasma propagation behavior. The change in the plasma density during its propagation occurred not only by an increase of plasma volume, but by the ion acceleration toward the propagation direction as well.     

Study of Neutron Cross Talk Rejection Based on Testing Experiment and Simulation

Experimental data analysis and simulation calculations were performed in order to evaluate the cross-talk rejection performance of a typical neutron detection array. For very closely packed scintillation bars, the CT rejection may rely on the position relation between the two signals. The criteria |△x|≤ 15 cm and |△y|≤12 cm are currently proposed for a rejection rate higher than 90%. For signals coming from distanced bars, the energy conservation relationship can be applied to reject the CT events with a similar performance. In both cases the results of simulation agree very well with the experimental data, assuring their applicability to other detection systems and physics problems.     

Neutron Capture Cross Section Measurement of Praseodymium in the Energy Region from 0.003 eV to 140 keV

The neutron capture cross section of praseodymium (¹⁴¹Pr) has been measured relative to the ¹⁰B(n,αγ) standard cross section in the energy region from 0.003 eV to 140 keV by the neutron time-of-flight (TOF) method with a 46-MeV electron linear accelerator (linac) of the Research Reactor Institute, Kyoto University (KURRI). An assembly of Bi₄Ge₃O₁₂ (BGO) scintillators was used for the capture cross section measurement. In addition, the thermal neutron cross section (2,200 m/s value) of the ¹⁴¹Pr(n, γ)¹⁴²Pr reaction has been also measured by an activation method at the heavy water thermal neutron facility of the Kyoto University Reactor (KUR). The thermal neutron flux was monitored with the ¹⁹⁷Au(n, γ)¹⁹⁸Au standard cross section. The above TOF measurement has been normalized to the current activation data (11.6±1.3 b) at 0.0253 eV.

The evaluated data in JENDL-3.3, ENDF/B-VI, and JEF-2.2 have been in general agreement with the current result, except that the JENDL-3.3 and the JEF-2.2 values are clearly lower than the measurement in the cross section minimum region from about 10 to 500 eV.     

Assessment of Some Cross Section Libraries through Analyses of Coupled-Core Critical Experiments by Monte Carlo Calculations

An experiment was performed on the natural circulation test loop HRTL-5, which simulates the geometry and system design of the 5 MW full power natural circulation nuclear heating reactor. Different flow modes, including density wave oscillation and flow excursion et al., were observed in a wide range of inlet sub-cooling at 1.5MPa. By means of self-developed computational codes, the bifurcation chart has been obtained. Consequently the flow excursion boundary has been determined. Through the analysis on the excursion boundary, the method to avoid the flow excursion during startup has been presented. Analytical results show: (1) with the decreasing heat flux or the increasing system pressure, the static flow excursion occurs at higher inlet temperature and its range in the instability maps becomes narrower correspondingly; (2) to decrease the outlet two-phase resistance or increase the inlet single-phase resistance is beneficial to avoid the flow excursion; (3) by means of increasing the system pressure to start up the reactor with low heat flux, the flow excursion and low steam quality density wave oscillation can be successfully avoided. This investigation is meaningful to the reactor safety and the design of the nuclear heating reactors.     

Measurement of Leakage Neutron Spectra from a Spherical Pile of Zirconium Irradiated with 14MeV Neutrons and Validation of Its Nuclear Data

In order to make a benchmark validation of the nuclear data for Zr, the leakage neutron spectrum from a Zr sphere of a 61-cm diameter was measured between 0.1 and 16MeV using a time-of-flight technique with a 14MeV neutron source facility, OKTAVIAN. The result was compared with the calculation using the Monte Carlo code MCNP-4A. To investigate the spectrum dependence on the individual neutron reactions, test calculations were carried out with the MCNP-4A code using the JENDL-3.2-based libraries, in which partial cross section values were reduced from the original values. From the comparison between the measured and the calculated spectra, it was found that each of the results could predict well the experiment in general. However, in detail, both ENDF/B-VI and EFF-2.4 gave considerable overestimation above 1 MeV. The JENDL-3.2 predicts the spectrum almost satisfactorily except below 0.8 MeV and around 10 MeV. The discrepancy found in JENDL-3.2 calculation is considered due to the cross section values of the (n, 2n) reaction and its secondary energy distributions (SED). The modified JENDL-3.2 library with the reduced (n, 2n) reaction values and the lower SED below 1 MeV reproduced the experiment with better agreement over the whole energy range.     

Trend in criticality calculation results of borated-water-moderation experimental cores relating to 10B(n, α) cross section

ABSTRACT

In connection with the accuracy of the ¹⁰B(n, α) cross section in the thermal- and epithermal-neutron energy regions, criticality calculation results were examined for six benchmark sets of light-water-moderation critical experiments of UO₂ and MOX fuel lattice cores with un-borated and borated water. Two of the benchmark sets were those implemented in the Tank-Type Critical Assembly (TCA). The others were taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP), and the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP). The enrichments of the UO₂ fuel range from 1.9 wt% to 2.6 wt%, and the Pu contents of the MOX fuel do from 2.0 to 6.6 wt%. The boron concentrations in water are up to 1511 ppm. The effective neutron multiplication factors (k_eff ) were taken from the published documents. They were calculated with continuous-energy Monte Carlo calculation codes in combination with JENDL-4.0, and other evaluated nuclear data libraries. It was confirmed that the k_eff values of the critical cores increased with the boron concentrations, which indicates that the ¹⁰B(n, α) cross section in the thermal- and epithermal-neutron energy regions should be larger than those in JENDL-4.0 and other libraries.