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。     

不同固体慢化剂对“花”型快谱超临界水冷堆中子学性能的影响分析

不同于一般采用氢化锆作固体慢化剂的反应堆,快谱超临界水冷堆工作在严酷的高温高压条件下,高氢平衡压以及停开堆造成的热冲击都会导致氢化锆中氢的大量损失,事故工况下甚至会引发氢的无控释放。本文通过分析对比多种材料的有效增殖系数、转换比、慢化剂温度反应性、燃料Doppler反应性、空泡反应性等参数的变化,发现氧化铍、碳化硅是中子学综合性能相对较好的"花"型快谱超临界水冷堆固体慢化剂材料,并且对燃料Doppler反应性系数影响不大。     

~7LiH材料热中子散射数据研究

基于热中子散射理论,编制热中子散射数据计算程序Sirius。采用基于密度泛函理论的第一原理冻结声子方法,得到7Li H中7Li和H的声子态密度。利用Sirius程序和声子态密度,获取7Li H的热中子散射数据。理论分析表明,获得的数据合理。     

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。     

氢化锂热化效应机理研究

氢化锂（LiH）以其低密度、高熔点、较高的H原子份额等良好的热物性,被用作空间核热推进反应堆的慢化剂和屏蔽材料。考虑到低能区LiH热中子数据的缺失使得数值模拟结果与实际相差很大,本文对LiH热化效应机理进行初步研究,基于第一性原理方法计算了LiH的声子谱,采用GASKET和NJOY程序建立LiH热散射律和散射矩阵的计算模型,制作成MCNP的ACE格式的LiH热中子截面数据库。对比文献结果和ZrH2热散射截面,分析差异的原因,采用Debye模型的抛物线效应修正了次级能量分布情况。该截面值可为下一步高温粒子球床堆物理建模提供必要的数据。     

铀氢锆反应堆脉冲参数分析

介绍了铀氢锆反应堆计算模型和程序，分析了反应性引入速率，反应性引入量，燃料瞬发负温度系数，燃料热容和瞬发中子寿命对脉冲参数的影响。计算结果表明，脉冲峰功率与反应性引入量的平方成正比；一次脉冲释放能量与反应性引入量成正比；燃料元件热容随燃料温度变化，脉冲峰功率和释放能量随燃料热容增大而增大。     

21 热离子反应堆电源中子物理模拟实验样机环境影响

热离子反应堆电源中子物理模拟实验样机（SPEM，简称模拟实验样机），是一座采用铀-235富集度90％的金属铀块作核燃料、氢化锆作慢化剂、金属铍作反射层、转动控制鼓控制反应性的临界装置。     

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     

SiO2热中子散射截面在空间堆事故分析中的应用

SiO₂热化效应可能对核废料地质储存库分析和空间反应堆坠落湿沙情况下的临界安全造成一定影响。本文结合最新的ENDF/B Ⅶ-1的TSL库,制作了ACE格式的SiO₂热中子截面数据库。分析了不同温度对SiO₂热中子散射截面的影响,比较了采用声子谱模型的SiO₂热中子散射截面数据与采用自由气体模型的SiO₂热中子散射截面数据的差异,并采用本文制作的截面库,对空间堆坠落湿沙情况下的临界安全特性进行了分析,给出了反应堆最易重返临界的湿沙成分比例。     

棒状氢化锆慢化钍基熔盐堆燃料组件稳态核热耦合程序开发

基于蒙特卡罗粒子输运程序MCNP与自主开发的子通道热工水力学程序SubTH,开发了棒状氢化锆慢化钍基熔盐堆燃料组件稳态核热耦合程序MCNP-SubTH,解决核热耦合程序因网格类型不同难以耦合的问题,程序具有普适性。MCNP-SubTH通过外耦合的方式进行MCNP和SubTH之间的数据交换,将MCNP计算得到的功率场加载到SubTH的求解文件中,然后将SubTH计算得到的密度和温度场更新到MCNP的输入卡中,实现程序迭代计算。分模块验证了MCNP-SubTH的准确性,并用MCNP-SubTH对棒状氢化锆慢化钍基熔盐堆燃料组件进行了稳态核热耦合计算,验证了核热耦合方法的有效性。     

Cold neutron scattering in imperfect deuterium crystals

The cross sections for cold neutron scattering in mosaic deuterium crystals have been calculated for various target structures. The theoretical results are compared with the recent experimental data obtained at the Paul Scherrer Institute (PSI) with the use of FUNSPIN beamline. It is shown that various patterns of the Bragg cross sections, which were observed in different solid D₂ samples at the same temperature, can be explained by contributions from a limited number of monocrystal orientations and the mosaic spread of about 3 degrees.     

Double differential neutron scattering cross sections of materials for ultra high temperature reactors

The five materials with the highest melting point are hafnium, tantalum, niobium and zirconium (ZrC) carbides and graphite (that sublimes). Graphite is the material of choice for very high temperature reactors (VHTR); ultra high temperature reactors (UHTR), like the thermal nuclear propulsion reactor NERVA use a dispersion of ZrC and UC in graphite as the material in the reactor core. Presently there are neither inelastic nor elastic double differential scattering data available that describe the thermalization process in ZrC. We therefore, calculated coherent elastic and incoherent inelastic cross sections for the ZrC crystal which has a face centered cubic (fcc) lattice. The phonon spectrum for the ZrC lattice was calculated with the computer code PHONON using the Hellman–Feynman forces computed with ab-initio methods [Jochyn, P.T., Parlinski, K., 2000. Ab initio lattice dynamics and elastic constants of ZrC. Eur. Phys. J. B 15, 265–268]. This phonon spectrum was then used to compute the S(α, β, T) matrices for the inelastic scattering cross sections for C and Zr in the ZrC lattice using modified versions of the computer codes GASKET, HEXSCAT and NJOY. The results were applied to calculate, with the proper S(α, β, T), criticality and reactivity coefficients of temperate of reactor systems containing ZrC and UC. For comparisons, these parameters were also calculated with approximations of S(α, β, T), i.e. the gas or the graphite scattering kernels. Depending on the degree of thermalization, k_eff is underestimated between 0.6% and 1%, and the values and the shape of the reactivity coefficients as a function of temperature change by substantial amounts.     

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。     

核数据处理程序NECP-Atlas的开发与验证

本文开发了自主化的核数据处理程序NECP-Atlas,该程序将不同的核数据处理功能封装为不同的程序模块,可将评价核数据经过共振重构及线性化、多普勒展宽计算、不可分辨共振区处理、热中子散射计算、多群截面计算等过程,处理为WIMS-D/E格式多群数据库。采用WLUP（WIMSD library update project）基准题、国际临界安全基准题ICSBEP（international criticality safety benchmark evaluation project）等对NECP-Atlas加工产生的核数据进行验证,结果显示NECP-Atlas和NJOY-2016程序精度相当。     

Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.     

Calculation and evaluations for n+Cu reactions

All cross sections of neutron induced reactions, angular distributions, energy spectra and double differential cross sections are consistently calculated and analyzed for n+^63,65,nat.Cu reactions at incident neutron energies below 200 MeV based on the nuclear theoretical models. The optical model, preequilibrium and equilibrium reaction theories, the distorted wave Born approximation theory are used. Theoretical calculated results are compared with existing experimental data and the evaluated results in ENDF/B-VII and JENDL-3 libraries. The optical model potential parameters are obtained according to the experimental data of total, nonelastic scattering cross sections and elastic scattering angular distributions.     

CENACE热散射中子文档的研制

为模拟计算相关中子学问题,中国核数据中心研制了ACE格式的多温度连续能量点截面库CENACE。其中,为计算热中子相关问题,采用NJOY99程序,将ENDF/B-Ⅶ.1库中18种材料的热散射率数据制成ACE格式的热中子散射数据。为验证热中子ACE文档的完整性和可用性,对加工得到的数据进行绘图测试,并将热散射截面的计算结果与实验测量值进行比较。测试结果表明,所有ACE文档数据准确可靠,不存在异常或不合理等现象;对于常见反应堆慢化剂材料,新制作的热散射数据与实验值符合良好,个别材料的热散射率评价数据有待进一步改进。