快中子反应堆多群常数程序RQCS

KQCS是用“自屏因子”法制作快中子反应堆多群常数的程序,它输出的群常数有无限稀释截面、自屏因子、P8展开的弹性散射转移矩阵、非弹散射转移几率和转移截面。能提供快堆扩散、S_N和P_N程序使用。本文全面介绍了KQCS计算方法,重点对自屏因子计算方法进行了研究,并对共振重叠效应作了新的考虑。     

基于等价理论的压水堆包壳材料共振计算方法研究

共振自屏效应的处理是影响压水堆组件程序反应性精度的主要因素之一,压水堆锆包壳材料同样具有共振自屏效应,忽略其影响会对反应性造成100~300 pcm（1 pcm=10-5）的偏差。目前,主要通过提供经验上的参考稀释截面与包壳等价理论处理包壳材料的共振自屏效应,但并未对其适用性及精度进行完整的分析。因此,本文采用DRAGON程序,通过一系列压水堆算例对这2种方法进行测试,确定包壳共振自屏效应的主要影响因素以及这2种方法的适用性。结果表明,包壳材料的共振自屏效应仅仅与包壳区的原子核密度、厚度、慢化区的水铀比有关,并且参考稀释截面方法可以满足大部分典型压水堆系统的计算精度,但是对于包壳区尺寸、原子核密度、慢化区水铀比变化较大的系统计算精度较差,而包壳等价理论计算精度和普适性强,可用于不同类型压水堆系统包壳材料的共振自屏计算。      

核燃料溶解器临界计算对截面的要求

用群截面对燃料溶解过程中出现的栅格、燃料双重不均匀和溶液３种系统作临界计算时，需要考虑中子的共振自屏效应。标准自屏公式或经过丹可夫因子修正的自屏公式不适用于燃料双重不均匀系统。ＯＥＣＤ／ＮＥＡ临界工作小组的结果表明，必须用碰撞概率（ＰＩＣ）方法，子群方法或精细慢化方法修正才能得到共振自屏效应的准确结果。用点截面作临界计算时，不会观察到自屏效应，可以准确进行包括燃料双重不均匀系统在内的临界计算。     

群常数制作软件Ruler研发

采用国际公认的群常数制作理论方法，包括共振重造方法、多普勒展宽方法、热散射率处理方法、群截面和散射矩阵计算方法、共振自屏处理方法等，研发了包括主驱动程序、评价数据输入输出模块、公共数学模块、系统公共子程序模块、进制转换模块、截面线性化和共振重造模块、截面温度展宽模块、不可分辨共振自屏模块、热散射截面计算模块、中子多群常数计算模块、WIMS-D格式接口模块等11个模块的群常数制作软件Ruler。采用与国际通用核数据处理程序NJOY99比较的方式对Ruler进行了验证，包括群常数比较和基准检验结果比较。验证结果表明，Ruler的计算精度与NJOY99相当，其计算速度、可维护性、可扩展性优于NJOY99。     

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。     

二维任意几何子群共振计算及加速优化

共振计算是反应堆组件堆芯设计和燃料管理的基础.子群共振计算方法基于共振能群子群截面,调用输运程序作为求解器,对子群中子注量率进行求解并且归并得到有效共振自屏截面,实现任意二维复杂几何的共振计算.由于子群方法在每个共振能群内部需要反复调用输运求解器,因此和等价理论相比速度较慢及本文基于子群方法的理论模型和自主开发的子群共振计算程序,提出并且完成了多群数据库、输运计算源项及多共振核素迭代的优化方案.通过基准题的验证可知,该方案在保持精度的同时提高了子群程序的计算效率,保证了该程序在工程上的实用性.     

多群中子输运计算中的共振自屏问題

在混合堆包层和快堆的中子学设计中,考虑共振自屏效应已成为中子学计算中一个必不可少的环节。本文讨论了在混合堆和快堆的多群中子输运计算中,考虑共振自屏效应的重要性及其考虑方法。     

基于改进缺陷因子方法的FCM燃料有效多群截面计算方法研究

全陶瓷微胶囊封装（FCM）燃料是重要的候选事故容错燃料,与传统燃料相比,FCM燃料的双重非均匀性使得其有效多群截面计算面临较大的挑战。本文提出一种改进的缺陷因子方法来处理FCM燃料在共振能区和非共振能区的自屏效应,实现FCM燃料的等效均匀化。通过颗粒丹可夫因子守恒来构建新的等效模型以克服传统的体积权重等效模型无法考虑燃料棒间自屏的影响;在共振能量段,基于新的等效一维球模型求解超细群慢化方程获得共振能量段的超细群缺陷因子;在非共振能量段,利用新等效模型的特征值计算获得快群和热群的多群缺陷因子;在此基础上实现FCM燃料棒的等效均匀化。本方法已在高保真中子学程序NECP-X上实现,并在一系列工况下进行了测试,与蒙特卡罗程序的比较表明,本方法能处理不同情况下的双重非均匀性,并可获得准确的有效自屏截面。     

反应堆多群常数计算程序的开发、研制(英文)

在CYBER-825计算机上移植、开发了引进的快中子多群常数产生程序FOURACES,热中子多群常数产生程序FLANGE-AE程序,进一步开发了计算共振自屏因子的MINX程序,并研制了产生反应堆多群常数的程序包RMCPP。利用这个程序包产生了63群多群常数工作库MC,将它与核工程临界安全计算MONTE CARLO程序NEMCS相联接,并计算了Pu和U系统的有效增殖因数,取得了初步满意的结果。     

共振自屏计算的子群法与共振干涉效应研究

对以子群法与特征线法相结合的中子共振自屏计算方法进行了研究,编制了共振计算程序(SGMOC);程序采用WIMSD格式的多群数据库。数值验证表明,SGMOC的计算结果与MCNP程序计算结果吻合良好,具有较高的计算精度与几何通用性。以SGMOC为基础,对子群法共振干涉效应修正计算的两种方法进行了研究分析。条件概率法对UO2燃料栅元无限增殖系数(kinf)计算的修正约为0.0003～0.0018;借助NJOY程序的方法对UO2燃料栅元kinf计算的修正约为0.0002～0.0013。     

Performance of advanced self-shielding models in DRAGON Version4 on analysis of a high conversion light water reactor lattice

A high conversion light water reactor lattice has been analysed using the code DRAGON Version4. This analysis was performed to test the performance of the advanced self-shielding models incorporated in DRAGON Version4. The self-shielding models are broadly classified into two groups – “equivalence in dilution” and “subgroup approach”. Under the “equivalence in dilution” approach we have analysed the generalized Stamm’ler model with and without Nordheim model and Riemann integration. These models have been analysed also using the Livolant–Jeanpierre normalization. Under the “subgroup approach”, we have analysed Statistical self-shielding model based on physical probability tables and Ribon extended self-shielding model based on mathematical probability tables. This analysis will help in understanding the performance of advanced self-shielding models for a lattice that is tight and has a large fraction of fissions happening in the resonance region. The nuclear data for the analysis was generated in-house. NJOY99.90 was used for generating libraries in DRAGLIB format for analysis using DRAGON and A Compact ENDF libraries for analysis using MCNP5. The evaluated datafiles were chosen based on the recommendations of the IAEA Co-ordinated Research Project on the WIMS Library Update Project. The reference solution for the problem was obtained using Monte Carlo code MCNP5. It was found that the Ribon extended self-shielding model based on mathematical probability tables using correlation model performed better than all other models.     

复杂几何燃料组件的共振自屏计算

共振参数计算是反应堆堆芯设计计算中的重要内容,传统的共振计算模型只适应于简单几何计算。本工作应用A.Hebert提出的子群共振自屏计算模型研制了复杂几何燃料组件的共振自屏计算程序。该程序能处理含有两种共振核素的复杂几何下的共振自屏。对一系列问题的数值校验计算表明,该模型在低富集度时具有较好的计算精度。     

低能电子轫致辐射的蒙特卡罗模拟

应用低能电子束进行辐照加工需要采用自屏蔽结构来达到辐射防护的要求。自屏蔽防护层的功能是屏蔽由轫致辐射产生的光子。本工作利用蒙特卡罗应用程序EGSnrcMP模拟了低能电子束轰击钢和聚乙烯两种靶材的轫致辐射过程，比较和分析了轫致辐射反射光子的能谱及角分布与靶材料种类、靶材料厚度之间的关系，为优化设计自屏蔽结构提供了基础数据。     

NECP-Atlas中屏蔽数据库制作模块的开发与验证

在核数据处理程序NECP-Atlas中开发了屏蔽数据库制作模块Shield_calc,该模块先利用NECP-Atlas产生问题无关的MATXS格式细群中子、光子截面数据库;然后采用超细群方法、Bondarenko迭代方法进行共振自屏计算,获得有效自屏截面;最后,基于1维反应堆模型采用NECP-Hydra进行输运计算获得应用堆型的典型权重谱,将细群屏蔽数据库归并为宽群屏蔽数据库NECL-SHILED。利用Shield_calc模块,基于与BUGLE-B7相同的评价核数据库ENDF/B-Ⅶ.0,制作了47群中子、20群光子的NECL-SHILED,并与BUGLE-B7进行了对比,数值结果显示NECL-SHILD与BUGLE-B7计算结果吻合较好,验证了Shield_calc模块具有较高的精度。      

Thermal neutron self-shielding correction factors for large sample instrumental neutron activation analysis using the MCNP code

Thermal neutron self-shielding within large samples was studied using the Monte Carlo neutron transport code MCNP. The code enabled a three-dimensional modeling of the actual source and geometry configuration including reactor core, graphite pile and sample. Neutron flux self-shielding correction factors derived for a set of materials of interest for large sample neutron activation analysis are presented and evaluated. Simulations were experimentally verified by measurements performed using activation foils. The results of this study can be applied in order to determine neutron self-shielding factors of unknown samples from the thermal neutron fluxes measured at the surface of the sample.     

Recent activities in the field of nuclear waste management and environmental science

The multiband method has been applied to analyses of critical experiments related to the high-conversion core at the Kyoto University Critical Assembly in order to accurately treat the resonance self-shielding in heterogeneous cells. Three-band parameters were generated using the self-shielding table installed in the SRAC code, and used to calculate the cell-averaged cross sections. The k _∞ values calculated by this method have been compared to those by the VIM Monte-Carlo calculation, the SRAC fine group calculation, Dancoff factor method and/or Tone's method self-shielding calculation. The k∞ values calculated by the present method agree with those by the VIM calculation within 0.3%Δk for all the cases considered.     

The pseudo-resonant-nuclide subgroup method based global–local self-shielding calculation scheme

The pseudo-resonant-nuclide subgroup method (PRNSM) based global–local self-shielding calculation scheme is proposed to simultaneously resolve the local self-shielding effects (including spatial self-shielding effect and the resonance interference effect) for large-scale problems in reactor physics calculations. This method splits self-shielding calculation into global calculations and local calculations. The global calculations obtain the Dancoff correction factor for each pin cell by neutron current method. Then an equivalent one-dimensional (1D) cylindrical problem for each pin cell is isolated from the lattice system by preserving Dancoff correction factor. The local calculation is to perform self-shielding calculations of the equivalent 1D cylindrical problem by the PRNSM. The numerical results show that PRNSM obtains accurate spatial dependent self-shielded cross sections and improves the accuracy of dealing with the resonance interference over the conventional Bondarenko iteration method and the resonance interference factor method. Furthermore, because both global and local calculation is linearly proportional to the size of problems, the global–local calculation scheme could be applied to large-scale problems.     

Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF₃ detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample’s surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required.     

A modified area function in time scale for the transmission data analysis of neutron resonance

A modified area function in time scale has been constructed with the phosphorescent decay of the detector (long-optical modes) and the background fraction included. A coefficient in the modified area function called self-shielding parameter is established to indicate the self-shielding effect in the experiment. The modified area function is independent from the equipment resolution. For the thick sample (nσ₀ ～ 10), the self-shielding effect would greatly improve the sensitivity of the modified area function to the temperature. Theoretically, we discuss the properties of the modified area function and estimate the least temperature deviation for this area method in the temperature measurement. Procedure of temperature measurement from isolated or multiple resonances by using the modified area method is also suggested.