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。     

MeV中子穿透~9Be板状样品评价中子核数据的基准检验

为检验次级中子泄漏谱及其角分布,利用飞行时间法测量了出射角为90°的板状9Be样品的泄漏中子谱。同时,以CENDL3.0、ENDF/B-6、JENDL3.3、JEF3.0/3.1等库作为中子输运计算的数据库,采用MCNP程序对实验装置及条件进行了精确模拟。在3～14MeV能量区间,将实验结果与模拟结果进行了     

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。     

高效慢化材料氢化锆的研制

氢化锆作为一种新兴的屏蔽、慢化材料,由于ZrHx的氢含量高和密度较低,可作为空间核反应堆的中子慢化材料。俄罗斯已将其作为一种新型高效的屏蔽和慢化材料进行研究;日本已将其应用于MUTSU核动力船压力容器顶部和主屏蔽体之间的空隙处,它可以在220℃运行温度下保持良好的屏蔽效果。在我国研制的热离子核反应堆中子物理模拟实验样机中,固态氢化锆中子慢化剂圆盘是物理样机堆芯必不可少的部件,该材料部件的研制成功与否关系到整个热离子反应堆电源系统的发展。开展氢化锆慢化材料的研制不仅具有较高的研究和应用价值,而且具有较高的经济价值,市场应用前景广阔。中国核动力研究设计院已建立了完整的氢化装置及其检测系统,经过多年的研究实验,积累了大量金属锆及锆合金的氢化工艺的数据和经验,建立了一整套氢化锆工艺控制、成品检验的规程和相关企业标准。     

氢化锆-水-铀栅格特性研究

本文的目的在于研究氢化锆在增殖系统中的慢化特性,在N_h/N_5为233—543范围内做了七个临界实验,在N_H/N_5=233,337两个准均匀栅格上做了热谱、通量分布、温度系数、瞬发中子衰减常数、材料的反应性系数等测量。七个临界实验、两个准均匀栅格的热谱与反射层节省等测量结果均与理论计算结果相符合,计算中用的是一维四群扩散理论,氢化锆介质的散射核是用晶格结构的声子谱模型获得的,从而说明这个计算模型是可用的。     

CENDL-3．1硅的宏观检验

利用日本的中子和γ泄漏谱实验OKTAVIAN对CENDL-3．1版硅同位素的全套中子评价数据进行宏观检验。检验工作比较了基于评价核数据库CENDL-2．1、3．1，ENDF／B—VII．0和JENDL-3．3的中子和γ泄漏谱的计算值和实验测量值。检验分析发现，CENDL-3．1的^28Si的分立能级非弹散射截面可能偏小，（n，p）、（n，α）反应的光子产生数据缺失，14Mev的弹性截面偏大等。这些问题有待进一步修正。     

聚乙烯样品泄漏中子谱的测量

聚乙烯（CH2）n只含有C、H两种元素,是理想的中子慢化材料。C、H的评价数据比较精确,因此,通过实验测量聚乙烯样品的泄漏中子谱,不仅可以验证对应的中子输运程序,还可以建立起对泄漏谱实验测量系统的检验方法。在中国原子能科学研究院600kV高压倍加器上,利用D．T反应产生的14MeV单能脉冲中子源,通过飞行时间法,测量了中子通过尺寸为1m×1m,厚度为I／2自由程、1个自由程、2个自由程和3个自由程的大块聚乙烯样品的泄漏中子谱,实验安排由图1所示。     

氢化锆零功率堆热中子通量谱测量

一、引言氢化锆晶体中氢原子处于束缚态。中子热化过程中与氢化锆的氢原子核能量交换具有量子化特征。所以,用水栅中常用的尼尔金散射核计算氢化锆反应堆热中子通量谱,结果比实际谱软。这是因为在氢化锆-轻水混合慢化剂中,能量低于0.14 eV的中子主要被周围轻水慢化,而中子却可以从氢化锆的氢核获得0.14 eV的整数倍能量。特别当温度升高后,氢化锆中氢原子处于高激发态的份额增加,中子被加速几率增加,使中子谱比水堆     

不同235U快裂变缓发中子群参数比较和验证

分析了ENDF/B7、JEFF3.1、JENDL3.3、CENDL2.2及Keepin数据中的²³⁵U快裂变缓发中子群参数的差异,通过CFBR-Ⅱ堆超瞬发临界实验检验了这几套缓发中子数据的准确性。检验结果表明,用于计算反应堆反应性,Keepin缓发中子群参数仍比数据库数据准确;数据库数据中,JEFF3.1的8群结构数据好于其他数据库6群结构数据。     

氘核中子反应截面的法捷耶夫方程理论计算

中子引起的轻核反应是核数据研究的重要内容。当前我国核数据库中氘核中子反应截面的计算结果局限于采用s 波可分离势,且入射能量在20 MeV以下。需要发展三体核反应的法捷耶夫方程理论方法,采用超出s 波的核子 核子相互作用,从而对更高能量范围内氘核全套中子反应截面做出准确的描述。本文介绍了利用法捷耶夫方程计算n+d三核子反应体系的弹性散射微分截面、破裂反应、破裂反应出射中子和质子的双微分截面的理论框架及数值计算结果,同时计算了弹性散射总截面和破裂反应总截面的激发函数。计算结果与实验数据及CENDL 32、ENDF/B Ⅷ.0、JENDL 5、JEFF 33等数据库中的评价数据符合较好。     

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。     

Measurement of ~(232)Th(n,2n)~(231)Th reaction cross-sections at neutron energies of 14.1 MeV and 14.8 MeV using neutron activation method

In this study, the activation cross-sections were measured for ~(232)Th(n,2n)~(231)Th reactions at neutron energies of 14.1 and 14.8 MeV, which were produced by a neutron generator through a T(d,n)~4He reaction. Induced gamma-ray activities were measured using a low background gamma ray spectrometer equipped with a high resolution HPGe detector. In the cross-section calculations, corrections were made regarding the effects of gamma-ray attenuation, dead-time, fluctuation of the neutron flux, and low energy neutrons. The measured cross-sections were compared with the literature data, evaluation data(ENDF-B/VII.1, JENDL-4.0 and CENDL-3.1), and the results of the model calculation(TALYS1.6).     

Direct neutron spectrum measurement to validate Zr(n,2n) reaction cross-section at 14 MeV

Lithium zirconate, Li₂ZrO₃, is known as a candidate blanket material in a fusion reactor. Various neutronics benchmark experiments for zirconium have thus been carried out so far. According to the independent benchmark studies by two parties, the neutron spectrum calculations show fairly large overestimation for most evaluated nuclear data libraries. However, the reason has not yet been made clear up to now. The author's group expects it would be due to a problem of evaluation for the ^natZr(n,2n) reaction cross-section, because the cross-section measurement is basically not possible with the foil activation method for zirconium isotopes except for ⁹⁰Zr.In the present study, two neutrons emitted from ^natZr(n,2n) reaction have been measured directly to investigate the reason for the above overestimation. The measurement was done with our own special technique of detecting angle-correlated neutrons by the coincidence detection technique and the pencil-beam DT neutron source of FNS, JAEA. Angle-correlated energy differential cross-sections for ^natZr(n,2n) reaction were successfully measured. The obtained total cross-section above the emitted neutron energy of 800 keV was fairly larger than the one evaluated in JENDL-3.3. The total cross-section of ^natZr(n,2n) reaction was estimated by extrapolating the spectrum down to zero energy taking into account the nuclear temperature. The estimated cross-section value with the nuclear temperature of 1 MeV, which is larger than the one adopted in JENDL-3.3, was in acceptable agreement with JENDL-3.3. It is suggested from the result that the disagreement pointed out in the previous benchmark studies may be due to inappropriate nuclear temperature used in the evaluation. Further investigation of the nuclear temperature employed in the nuclear data evaluation should thus be carried out once again.     

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 ...     

Discrepancies in Estimates of the Number of Secondary Fission Neutrons for 239Pu

The energy dependence of the total number of secondary neutrons v ²³⁹Pu is analyzed. The results of variations of different national libraries are compared: ENDF/B-5, ENDF/B-6.5, JEF-2.2, JENDL-3.2, CENDL-2, and BROND-2. It is concluded that the required accuracy has not been attained for the entire energy range.     

Measurement of 232Th(n,f) and 232Th(n,γ) reaction rates in thorium powder cylinder bombarded with D-T neutrons†

In order to provide reference for the evaluation of thorium parameters for the conceptual design of fusion–fission hybrid energy reactor, a dedicated integral experiment was carried out in a thorium powder cylinder bombarded with D-T neutrons. Thorium capture and fission rates in the 0° direction to the incident D⁺ beam were obtained using the activation method followed by the off-line gamma-ray technique, experimental uncertainties were ～3.1% for thorium capture rate, and were 5.5%, 8.1%, and 6.3% for thorium fission rates based on fission products ^85mKr, ¹⁴³Ce, and ⁸⁷Kr, respectively. The thorium fission rate based on ^85mKr agreed well with the simulation employing ENDF/B-VII.0 library data. The influence of the oxygen contained in the thorium oxide powder and the scattering neutrons from the experimental hall was also evaluated. MCNP simulations employing ENDF/B-Ⅶ.0, JENDL-4.0 library data agreed with experiment within uncertainties except that employing ENDF/B-Ⅶ.1 (6.0%) and CENDL-3.1 (7.9%) for thorium fission rate, while for thorium capture rate, simulation employing JENDL-4.0 agreed with experiment best. The influence of reaction channels of thorium transport medium employing different library data on the thorium reaction rates could be neglected according to the simulation. The thorium capture to fission ratio demonstrated that the fuel breeding efficiency is quite low and energy production plays a leading role under the neutron spectra in this experiment.     

Measurement and analysis of reflector reactivity worth by replacing stainless steel with zirconium at the fast critical assembly (FCA)

A reflector reactivity worth was measured by replacing stainless steel with zirconium at the FCA. The experimental result of the positive reflector reactivity worth demonstrates the effectiveness of the zirconium reflector compared with the SS reflector in the fast reactor core. This paper also focuses on the validation of standard calculation methods used for fast reactors with JENDL-4.0. As a result, it is confirmed that the standard calculation methods for the reflector reactivity worth show agreement within the experimental error.     

Features of Zr-rich corner of the Zr-N-O ternary system by controlled low-pressure oxidation and thermodynamic analysis

In order to clarify the Zr-rich corner of the Zr-N-O ternary system, a series of nitridation and controlled low-pressure oxidation heatings were made on zirconium metal plates. The experiments were made in separated reaction sequences, which were (1) oxidation of pre-nitrided zirconium, and (2) simultaneous nitridation and oxidation of zirconium. Low-partial pressures of oxygen were obtained by using a redox couple of either Mo/MoO₂ or Cu₂O/CuO. The samples were characterized by X-ray diffraction, optical microscopy, and electron probe micro analysis. Experimental results were coupled with a preliminary thermodynamic analysis with a sublattice formalism to draw a provisional ternary isotherm at 1373 K.     

Self-Shielding Factors for Neutron Capture Reactions of Uranium-238 and Thorium-232 in Energy Range of 1–35 keV

Self-shielding factors for the neutron capture reactions of ²³⁸U and ²³²Th were measured in the resonance energy region of 1–35 keV, using a neutron time-of-flight method with an electron linear accelerator. The self-shielding factors for arbitrary dilution cross sections were obtained from sets of neutron transmission ratios and self-indication ratios measured with several transmission samples of different thicknesses. The maximum experimental errors for ²³⁸U and ²³²Th were about 3 and about 7%, respectively.

The experimental results were compared with calculations based on JENDL-2, JENDL-3 and ENDF/B-IV. For ²³⁸U, an energy dependent structure was observed in the experimental self- shielding factors. The calculations based on JENDL-2 and ENDF/B-IV did not show this structure in the unresolved resonance region and were smaller than the experimental values from 4 to 6 keV. The calculation based on the resolved resonance parameters in JENDL-3 showed better agreement with the experiment from 4 to 6 keV, but discrepancies still remained in other energy ranges.

For ²³²Th, no remarkable discrepancy was observed in the unresolved resonance region, but JENDL-2 and JENDL-3 tended to give smaller values than the experimental self-shielding factors in the resolved resonance region.     

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.