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。     

基于不同ENDF格式数据库的MCNP温度相关中子截面库研究

使用截面加工程序NJOY生成以针对最新释放的ENDF/B-VII和CENDL-3.1评价核数据截面库为基础库的2个ACE格式的温度相关中子截面库。使用压水堆多普勒数值基准题对生成的2个ACE格式截面库进行基准验算。验算结果表明,所生成的2个温度截面库在有效增殖系数、多普勒反应性亏损、多普勒反应性系数方面均与原基准题吻合良好,说明评价核数据截面库ENDF/B-VII和CENDL-3.1能很好地应用于ACE格式的截面库的制作。     

基于蒙特卡罗抽样的活化中子能谱测量不确定度分析

不确定度分析是活化法测量中子能谱的关键环节。本文针对SAND-Ⅱ活化中子解谱过程,给出了一种基于先验谱、活化率和截面协方差的中子能谱测量不确定度蒙特卡罗分析方法。首先,建立了基于线性变换的截面协方差抽样方法;然后,利用MCNP计算了误差,使用迭代方法估计了先验谱不确定度;最后,结合活化率的测量不确定度,利用蒙特卡罗抽样算法计算了中子能谱的不确定度。利用锎源自发裂变谱对该方法进行了验证,与传统方法相比,不确定度分析结果更为准确。对西安脉冲堆某次中子能谱测量结果进行了测量不确定度分析,结果表明该方法更具保守性。     

基于启明星Ⅱ号的铅基堆核数据宏观检验

为提高铅基堆中子学模拟的可靠性,基于启明星Ⅱ号铅基零功率反应堆,开展铅基堆相关核数据的入堆宏观基准检验研究。采用周期法测量堆芯反应性,进而获得有效增殖因数keff为1001 14±0000 07。采用MCNP程序对铅基堆进行精细化建模,结合不同数据库内的中子评价核数据,计算实验燃料棒装载下的铅基堆芯的keff。比较结果可知,4种截面库计算的铅基堆keff模拟结果与实验结果吻合较好,最大相对偏差小于1%,其中,ENDF/B Ⅶ.1库的模拟结果与实验结果吻合最好,相对偏差和绝对偏差分别为025%和251 pcm。通过计算关键材料元素核数据引起keff的变化量,可知铅元素核数据引起的堆芯keff结果的波动量最大,在CENDL 31和JENDL 40中的铅元素引起keff的波动值分别为219 pcm和166 pcm。     

~(238)Np全套中子数据更新评价

裂变核全套中子评价数据为反应堆设计和安全运行、乏燃料次锕系核素嬗变、嬗变系统及高燃耗反应堆设计提供重要的基础数据。本文以一套全新的n+238 Np的中子光学模型势参数为基础进行理论分析,并根据Np各同位素反应截面系统变化规律,对模型势参数进行了调整,最后完成了全套中子数据的更新评价,与CENDL-3.1评价结果相比有较明显的改进。     

基于D-T中子源的板状铌样品的评价数据基准检验实验

利用中国原子能科学研究院核数据国家重点实验室的脉冲化氘氚聚变中子源产生的145 MeV单能中子,通过飞行时间法,测量了5、10、15 cm厚度板状铌（Nb）样品在与60°和120°两个方向上的泄漏中子飞行时间谱。利用蒙特卡罗模拟软件MCNP 4C进行了泄漏中子飞行时间谱的模拟计算,分别获得了CENDL 31、ENDF/B Ⅷ0和JENDL 40 3个数据库中Nb评价数据的模拟结果。通过各数据库不同能区的模拟结果与实验结果的比值（C/E）,对3个数据库中93Nb与145 MeV中子作用的角分布和双微分截面等相关评价数据进行了检验,重点分析了CENDL 31库的数据。结果表明,CENDL 31数据库的模拟结果在弹性散射能区、非弹性散射能区以及(n,2n)反应能区与实验结果均存在一定的偏差。而JENDL 40数据库除在120°弹性散射能区有高估现象,其他能区的模拟结果与实验结果均符合较好。ENDF/B Ⅷ0数据库的模拟结果除在60°方向弹性散射峰偏低外,其他能量范围的模拟结果均高于实验。     

板状Bi样品泄漏中子和γ谱实验测量与模拟计算

通过飞行时间法,测量了氘氘脉冲中子与不同厚度209Bi样品作用后61°和119°方向的泄漏中子飞行时间谱和泄漏γ能谱,样品尺寸分别为30 cm×30 cm×5 cm、30 cm×30 cm×10 cm和30 cm×30 cm×15 cm。采用BC501A液体闪烁体探测器测量0.8～3.2 MeV能区的泄漏中子飞行时间谱,钾冰晶石探测器（CLYC）测量0.2～0.8 MeV的泄漏中子飞行时间谱和泄漏γ能谱。用MCNP-4C程序对泄漏中子飞行时间谱和泄漏γ能谱进行了模拟计算,²⁰⁹Bi的评价中子核数据分别采用了CENDL-3.1库、ENDF/B-Ⅷ.0库、JENDL-4.0库以及JEFF-3.3库中的数据,模拟结果分别与实验结果进行比较分析,研究结果表明,泄漏中子谱CENDL-3.1库的模拟结果在119°方向弹性峰位置有较严重的低估现象,JENDL-4.0库在1.5 MeV附近（第二非弹能区）有一定高估,而在低能区有明显低估;泄漏γ能谱JENDL-4.0库和JEFF-3.3库的模拟结果与实验结果偏差明显,而CENDL-3.1库符合较好。     

Theoretical spectrum-threshold detector data correlation in EBR-II

An experimental and computer program to further examine the neutron environment in the Experimental Breeder Reactor-II (EBR-II) has been completed. Monte Carlo and S₄ Transport methods were used to determine the neutron spectrum at various positions in the EBR-II core and blanket regions. Response functions for the threshold detectors ⁵⁸Ni (n, p) ⁵⁸Co and ⁵⁴Fe (n, p) ⁵⁴Mn were determined for each position and the corresponding predicted induced activities are compared with experimental results. Based on combinations of calculated neutron spectra, experimental detector responses, and cross section end points an empirical differential cross section was determined for the ⁴⁶Ti (n, p) ⁴⁶Sc threshold reaction. Spectrum averaged cross sections for the three threshold reactions which have been determined at various positions in this facility suggest that significant errors in fast neutron fluences will result if the usual fission spectrum averaged cross sections are used.     

启明星Ⅱ快中子能谱区裂变率分布研究

使用MCNP程序对启明星Ⅱ进行了裂变率分布的详细计算分析。根据理论计算的分布规律,优化了实验测量裂变率分布方案,合理布局了探测器位置。用固体核径迹探测器开展了启明星Ⅱ快中子能谱区裂变率分布的实验测量研究,确定了快中子能谱区的裂变率分布。测量结果显示：快中子能谱区裂变率分布与理论计算结果基本符合。测量结果对ADS次临界反应堆确定堆芯裂变功率提供了数据参考。     

Preliminary Experiments on Accelerator-Driven Subcritical Reactor with Pulsed Neutron Generator in Kyoto University Critical Assembly

A series of preliminary experiments on an accelerator-driven subcritical reactor (ADSR) with 14 MeV neutrons were conducted at Kyoto University Critical Assembly (KUCA) with the prospect of establishing a new neutron source for research. A critical assembly of a solid-moderated and -reflected core was combined with a Cockcroft-Walton-type accelerator. A neutron shield and a beam duct were installed in the reflector region for directing as large a number as possible of the high-energy 14MeV neutrons generated by deuteron-tritium (D-T) reactions to the fuel region, since the tritium target is located outside the core. And then, neutrons (14MeV) were injected into a subcritical system through a polyethylene reflector. The objectives of this paper are to investigate the neutron design accuracy of the ADSR with 14MeV neutrons and to examine experimentally the neutronic properties of the ADSR with 14MeV neutrons at KUCA. The reaction rate distribution and the neutron spectrum were measured by the foil activation method for investigating the neutronic properties of the ADSR with 14 MeV neutrons. The eigenvalue and fixed-source calculations were executed using a continuous-energy Monte Carlo calculation code MCNP-4C3 with ENDF/B-VI.2 for the subcriticality and the reaction rate distribution, respectively; the unfolding calculation was done using the SAND-II code coupled with JENDL Activation Cross Section File 96 for the neutron spectrum. The values of the calculated subcriticality and the reaction rate distribution were in good agreement with those of the experiments. The results of the experiments and the calculations demonstrated that the installation of the neutron shield and the beam duct was experimentally valid and that the MCNP-4C3 calculations were accurately carried out for analyzing the neutronic properties of the ADSR with 14MeV neutrons at KUCA.     

Measurements of Neutron Capture Cross Section of 237Np for Fast Neutrons

The neutron capture cross section of ²³⁷Np has been measured for fast neutrons supplied at the center of the core in the Yayoi reactor. The activation method was used for the measurement, in which the amount of the product ²³⁸Np was determined by γ-ray spectroscopy using a Ge detector. The neutron flux at the center of the core calculated by the Monte Carlo simulation code MCNP was renormalized by using the activity of a gold activation foil irradiated simultaneously. The new convention is proposed in this paper to make possible a definite comparison of the integral measurement by the activation method using fast reactor neutrons with differential measurements using accelerator-based neutrons. “Representative neutron energy” is defined in the convention at which the cross section deduced by the activation measurement has a high sensitivity. The capture cross section of ²³⁷Np corresponding to the representative neutron energy was deduced as 0:80 ± 0:04b at 214 ± 9 keV from the measured reaction rate and the energy dependence of the cross section in the nuclear data library ENDF/B-VII.0. The deduced cross section of ²³⁷Np at the representative neutron energy agrees with the evaluated data of ENDF/B-VII.0, but is 15% higher than that of JENDL-3.3 and 13% higher than that of JENDL/AC-2008.     

反应堆临界-燃耗耦合蒙特卡罗计算

基于连续点截面MCNP程序 ,研制了三维多群P3 中子输运蒙特卡罗程序MCMG ,并与栅元均匀化程序WIMS耦合 ,实现了临界 燃耗耦合计算。采用WIMS产生的 69群共振、自屏宏观中子截面和BUGLE 80u47群微观中子截面 ,分别计算了简单反应堆和临界实验堆问题 ,计算结果与其它输运方法的计算结果和试验结果一致。在相同计算精度下 ,MCMG的计算时间较MCNP的计算时间少     

快堆用蒙特卡罗多群数据库制作方法研究

Based on the discrete angle method, a Monte Carlo multi-group cross section generation program MGXSMC was developed. This program can read the cross section data from an input file or read the cross section from a library in a specified format to generate the multi-group cross section for MCNP or RMC. The corresponding index file list can be automatically generated. The two-dimensional two-group IAEA pressurized water reactor (PWR) benchmark and lead-based fast reactor (RBEC-M) benchmark were used to verify the cross section library generated by the MGXSMC program. The calculation results show that the difference between the calculated result of the P5 order approximate multigroup section and the continuous point cross section is 24 pcm (1pcm = 10-5), and the difference of the keff result calculated by the P0 order approximate multigroup section and the continuous point section is large. This shows that the method and the program developed for the Monte Carlo Group Section Library are correct. At the same time, the neutron anisotropic scattering has a large impact on the calculation results of the lead-based fast reactor. Therefore, when the Monte Carlo Group Section library is produced, the neutron scattering angle data should be added.     

快堆用蒙特卡罗多群数据库制作方法研究

基于离散角方法,开发了蒙特卡罗多群数据库生成程序MGXSMC,该程序可以实现从输入文件读取截面数据或者从指定格式的截面库中读取截面,产生可供蒙特卡罗程序MCNP或RMC计算的数据库,并且可自动生成相应的索引文件列表。采用二维两群不带反射层的国际原子能机构(IAEA)压水堆(PWR)基准题和铅基快堆(RBEC-M)基准题对MGXSMC程序加工产生的核数据进行验证,计算结果表明,采用P₅阶近似多群截面与连续点截面计算的有效增殖系数(k_eff)结果相差24 pcm(1pcm=10^-5),而采用P₀阶近似多群截面与连续点截面计算的k_eff结果相差较大。由此说明蒙特卡罗多群数据库的制作方法和所开发的程序是正确的,同时,中子各向异性散射对铅基快堆计算结果影响较大,故制作蒙特卡罗多群数据库时应加入中子散射角数据。     

The integral experiment on beryllium with D-T neutrons for verification of tritium breeding

A clean benchmark experiment on beryllium was performed with D-T neutrons at the FNS facility of the Japan Atomic Energy Agency. The main objective was to verify the integral data related to the tritium production on lithium isotopes. Tritium production rates, as well as activation reaction rates were measured inside the beryllium assembly that was shaped as a pseudo-cylindrical slab with an area-equivalent diameter of 628 mm and a thickness of 355 mm. Experimental results were analyzed with a three-dimensional Monte Carlo transport code MCNP-4C and FENDL/MC-2.0, JENDL-3.2/3.3 neutron transport libraries. Evaluation of reaction rates was based on the cross section data taken from the JENDL Dosimetry File and ENDF B-VI data libraries. Analysis shows that all calculation combinations (transport and activation cross section libraries) used for evaluation of reaction rates give data that is agreeable with measured values within 10%.     

Acceleration of source convergence in Monte Carlo k-eigenvalue problems via anchoring with a p-CMFD deterministic method

The Monte Carlo method is widely used in neutron transport calculations, especially in complex geometry and continuous-energy problems. However, an extended application of the Monte Carlo method to large realistic eigenvalue problems remains a challenge due to its slow convergence and large fluctuations in the fission source distribution. In this paper, a deterministic partial current-based Coarse-Mesh Finite Difference (p-CMFD) method is proposed that achieves fast convergence in fission source distribution in Monte Carlo k-eigenvalue problems. In this method, the high-order Monte Carlo method provides homogenized and condensed cross section parameters while the low-order deterministic p-CMFD method provides anchoring of the fission source distribution. The proposed method is implemented in the MCNP5 code (version 1.30) and tested on realistic one- and two-dimensional heterogeneous continuous-energy large core problems, with encouraging results.     

Determination of the neutron flux and energy spectrum in the low-temperature fast-neutron facility in CP-5, calculations of primary-recoil and damage-energy distributions,and comparisons with experiment

We have determined the absolute differential neutron-energy spectrum for the low-temperature fast-neutron irradiation facility in the CP-5 reactor by means of a 20-foil activation technique. This technique employs the most recent version of the SAND-II computer code, which iteratively unfolds the neutron spectrum by fitting the foil activities. A Monte Carlo routine was also employed to calculate standard-deviation errors in each neutron-energy group. Using this differential neutron spectrum we have calculated, for numerous elements, total recoil cross sections, detailed primary-recoil group distributions, total damage-energy cross sections, damage-energy distributions, and an error analysis based on the uncertainties in the neutron spectrum. The significance of this information with respect to the interpretation of various neutron radiation-damage experiments, including sputtering, disordering of ordered alloys, and changes in critical current of A-15 compound superconductors is discussed. A detailed comparison is made among initial resistivity-damage rates for five widely different (well characterized) neutron sources, fission fragments, and heavy ions.     

~(115)In中子非弹性散射截面的实验测量及蒙特卡罗修正

~(115)In是一种重要的活化材料,准确测量它的中子非弹性散射截面数据对中子注量监测具有重要意义。在四川大学原子核科学技术研究所2.5 MV静电质子加速器上,利用核反应D(d,n)~3He产生的单能中子,以~(197)Au作为标准,采用活化法测量了2.95 Me V、3.94 Me V、5.24 Me V能点的~(115)In中子非弹性散射截面。用Monte Carlo程序MCNPX(Monte Carlo N-Particle eXtended)对靶头材料、冷却水层和样品的包层材料等引起的多次散射效应及注量率衰减效应等进行了修正计算,得到最终结果与Loevestam的计算值符合较好,并且实验中可通过减小靶管、靶底衬、水层及样品的包层材料等厚度来减小多次散射效应和自屏蔽效应的影响。     

Neutron Self-Shielding Factor of Scattering-Predominant Resonance Foils

The neutron self-shielding factor of ⁵⁹Co resonance foil as an example of foils whose scattering cross section predominate over their absorption cross sections was obtained by both Monte Carlo method (analog) and the collision probability method for various thicknesses of the foil. Also, the transmission and reflection probabilities of neutrons which have various energies near the resonance energy were obtained, and the effects of multiple scattering of neutrons on the neutron self-shielding factor are discussed.

The neutron self-shielding factors obtained by the Monte Carlo method and by the collision probability method agreed well with each other in the cases Σ t ～ 4.0, in which the Monte Carlo method requires considerably longer machine time. Although for the cases of large Σ t (～4.0) the agreement is not always good because of the flat flux approximation in the collision probability method, the calculation time by Monte Carlo is conveniently short. A combination of both methods is useful in obtaining the neutron self-shielding factor of resonance foils.     

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.