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。     

CENDL库中Fe、W评价数据的改进对屏蔽积分实验计算结果的影响

利用中国原子能科学研究院核数据重点实验室中子积分实验装置,分别完成了氘氚中子与不同尺寸Fe、W样品作用的泄漏中子飞行时间谱实验测量。利用MCNP 4C程序开展了泄漏中子飞行时间谱的模拟计算,Fe和W的评价数据分别采用CENDL 32库及CENDL 31库的数据,并将两数据库模拟结果与实验结果进行对比分析,重点分析了CENDL 32库中Fe和W的数据的改进与不足。结果表明：对Fe中子评价数据,CENDL 32库在弹性散射能区、连续能级非弹性散射能区及分立能级非弹性散射能区,模拟结果均与实验结果符合较好,较CENDL 31库有明显改善;对W中子评价数据,CENDL 32库在非弹性散射能区的模拟结果与实验结果符合较好,较CENDL 31库有明显改善,但在弹性散射能区模拟结果高于实验结果,在(n,2n)反应能区模拟结果低于实验结果。CENDL 32库关于天然W的中子评价数据有待进一步改善。     

中子与^181Ta反应的理论计算

本工作运用TALYS程序对中国评价核数据库中重要结构材料核Ta的全套中子数据进行了模型计算。根据中子与^181Ta反应的全截面、弹性散射截面和弹性散射角分布的实验数据，对Koning和Delaroche给出的中子普适光学势参数进行调节，得到一组适用于能量范围在0．1～30MeV之间、中子与^181Ta反应的光学模型势参数。     

CENDL-3.1铜的检验和改进

介绍了利用屏蔽基准实验OKTAVIAN以及核临界安全手册(ICSBEP)中的临界基准实验对CENDL-3.1铜的伞套中子评价数据进行的宏观检验.在屏蔽基准检验中,除了中子和γ泄漏谱上发现了由非弹性散射截面造成的与实验测量结果的分歧,计算结果与实验符合相当好.在快中子谱临界基准检验中,装置HMF072、HMF073和PMF013的keff的计算结果高出实验值大约2%,严重偏离实验结果.针对HMF072装置的灵敏度分析显示,该分歧的产生主要是由于全截面在0.1～1.3 MeV能区的评价不当引起的.在对0.1～1.3 MeV的全截面进行修正后,临界检验的结果获得了明显改善.     

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。     

钒球14 MeV中子的泄漏能谱测量

建立了厚度为 1 0 5cm的金属钒球基准实验装置。钒的纯度为 99 9%。用NE 2 1 3谱仪测量了d T中子的 0 75～ 1 5MeV泄漏中子能谱 ,能量大于 0 75MeV的中子的穿透率为 0 84± 0 0 3 ,中子能谱实验误差为 5 %～ 7%。用MCNP/ 4AMonte Carlo程序和FENDL 2库核数据进行了模拟计算 ,并与实验结果进行了比较     

~(209)Bi中子反应双微分截面的理论计算和分析

根据209Bi与中子反应的总截面、弹性散射截面、去弹性散射截面和弹性散射角分布的实验数据,应用自动调整光学模型势参数程序,得到了一组中子的光学模型势参数;使用这组参数和中子能量在20 MeV以下的核反应理论计算程序并考虑了中子直接非弹性散射的贡献,计算了209Bi与中子反应的所有截面、角分布和能谱,特别是发射中子、质子、氘、氚和α 粒子的双微分截面,γ产生截面和γ产生谱。理论计算结果与实验数据和评价库的结果进行了比较和分析,结果表明:无论是反应截面,还是能谱,现在的结果比ENDF/B-6和JENDL-3评价库中的结果与实验数据符合的更好、更合理。理论计算结果以ENDF/B-6格式推荐并提供使用。     

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。      

n＋^28Si核反应20MeV以下数据的理论研究

在对CEDNL-3．1库^28Si的中子评价数据的泄漏谱积分检验基础之上，先采用中国核数据中心研制的程序APMN调节了光学势参数，得到了一套比较符合角分布实验数据的光参，然后利用DWUCK计算了激发态直接非弹性散射的数据，最后使用UNF程序计算了全套数据，再对新计算的数据进行了同样的泄漏谱积分检验。经过数次循环之后，目前的研究结果表明，     

准直D-T中子束穿过大块板状聚乙烯样品积分实验

本文介绍了中国原子能科学研究院建立的准直中子束积分实验装置。该装置利用T(d,n)⁴He反应产生14.8 MeV脉冲中子束,经1.1 m厚重水泥屏蔽墙上的准直孔道后与样品作用,用飞行时间法测量样品不同方向的泄漏中子谱。首次测量了样品厚度分别为4.5、9、18和27 cm的大块板状聚乙烯样品在30°和50°方向的泄漏中子谱;考虑靶结构、源中子能谱和角分布、脉冲束宽度及探测器效率,利用MCNP程序模拟计算了相同实验条件下的泄漏中子飞行时间谱。实验结果与模拟结果符合较好。     

Benchmarking of JEFF-3.2, FENDL-3.0 and TENDL-2014 evaluated data for tungsten with 14.8 MeV neutrons

Integral experiments on tungsten slab samples were carried out on the D-T neutron source facility at China Institute of Atomic Energy.Leakage neutron spectra from the irradiated tungsten target were measured by the time-of-flight technique.Accuracy of the nuclear data for tungsten was examined by comparing the measured neutron spectra with the leakage neutron spectra simulated using the MCNP-4C code with evaluated nuclear data of the JEFF-3.2,FENDL-3.0 and TENDL-2014 libraries.The results show that the calculations with JEFF-3.2 agree well with the measurements in the whole energy range and all angles,whereas the spectra calculated with FENDL-3.0 and TENDL-2014 have some discrepancies with the experimental data.     

Benchmarking of JEFF-3.2, FENDL-3.0 and TENDL-2014 evaluated data for tungsten with 14.8 MeV neutrons

Integral experiments on tungsten slab samples were carried out on the D-T neutron source facility at China Institute of Atomic Energy.Leakage neutron spectra from the irradiated tungsten target were measured by the time-of-flight technique.Accuracy of the nuclear data for tungsten was examined by comparing the measured neutron spectra with the leakage neutron spectra simulated using the MCNP-4C code with evaluated nuclear data of the JEFF-3.2,FENDL-3.0 and TENDL-2014 libraries.The results show that the calculations with JEFF-3.2 agree well with the measurements in the whole energy range and all angles,whereas the spectra calculated with FENDL-3.0 and TENDL-2014 have some discrepancies with the experimental data.     

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.     

Neutron Total Cross Section for Iron around Resonance Valley near 24 keV

A neutron leakage spectrum from a nickel sphere surrounding a 14-MeV neutron source is measured and analyzed in order to verify the accuracy of nickel cross sections. Measurement is done by means of a time-of-flight technique in the range of 2–15 MeV using an NE213 scintillator, and compared with calculations carried out with MCNP, a continuous energy Monte Carlo transport code, using JENDL-3PR1 and ENDFIB-IV neutron libraries.

In spite of an overestimation of neutron flux near 13 MeV, the calculated result employing JENDL-3PR1 shows generally better agreement with the measured spectrum. In ENDF/B-IV usage, there is disagreement between measured and calculated spectra between 5 and 12 MeV. Problems in evaluated nuclear data for nickel are also described.     

Measurements and Calculations of Angular Flux Spectra Emitted from Lithium and Graphite Slabs with D-T Neutron Source

Angular dependent flux spectra from slab assemblies (lithium and graphite) were measured to test nuclear data and calculational methods for D-T fusion reactor neutronics. The collimated 14 MeV neutron source could be applied by the use of an associated particle method and the neutron spectra from 14 to 2 MeV were observed with TOF technique. The measured spectral pattern was dependent on the anisotropy of secondary neutrons emitted from both the elastic and the non-elastic scattering for 14 MeV neutrons. As for the numerical calculations, one-dimensional discrete ordinates transport codes (ANISN and NITRAN) were used. The multigroup cross sections processed with SPTG4Z from ENDF/B-IV were used as common nuclear data base. The problems of calculational methods and nuclear data were discussed in comparison with the experimental data and it was clarified that sufficient nuclear data of angular dependent cross sections for the non-elastic scattering have not been available in ENDF/B-IV and that the anisotropy of the scattering could not be calculated with ANISN which utilized the scattering kernel generated by incorrect treatment of scattering kinematics in the processing code. However, good agreement between the measurements and calculations was obtained by the use of NITRAN system with the appropriate processing codes of inelastic scattering anisotropies. It was shown that the NITRAN system was useful for anisotropic neutron transport calculations.     

Feedback on neutron capture cross sections of 238Pu and 241Am from analyses of measured isotopic compositions of irradiated LWR fuels and MOX core physics experiments

The atomic fractions of ²³⁸Pu and ²⁴¹Am in MOX fuels recycled in light water reactors are 1% to 2% and not significant compared with those of major Pu isotopes. On the other hand, recent evaluated nuclear data libraries, such as JENDL-4.0 and JEFF-3.2, give noticeably different thermal and epithermal neutron capture cross sections for ²³⁸Pu and ²⁴¹Am. The thermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am in JEFF-3.2 are 31% and 9% larger than those of JENDL-4.0, respectively. This paper shows the effect of the differences in the neutron cross sections on analysis results of two different integral experiments. The first is the isotopic compositions of ²³⁸Pu on UO₂ and MOX fuels irradiated in BWR and PWR, and the second is the critical experiments of the water moderated cores fully loaded with MOX fuels. The former was analyzed by using the continuous energy Monte Carlo burnup calculation code MVP-BURN and the latter by the continuous energy Monte Carlo calculation code MVP. The comparisons between the calculated and measured results indicate that the most likely thermal and epithermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am should be around at the middle between those of JEFF-3.2 and JENDL-4.0.     

Performance of the CENDL-3.2 and other major neutron data libraries for criticality calculations

Nuclear data are the cornerstones of reactor physics and shielding calculations.Recently,China released CENDL-3.2 in 2020,and the US released ENDF/B-Ⅷ.0 in 2018.Therefore,it is necessary to comprehensively evaluate the criticality computing performance of these newly released evaluated nuclear libraries.In this study,we used the NJOY2016 code to generate ACE format libraries based on the latest neutron data libraries（including CENDL-3.2,JEFF3.3,ENDF/B-Ⅷ.0,and JENDL4.0）.The MCNP code was used to ...