可显式处理燃料组件轴向非均匀性的粗网节块方法

针对传统节块方法无法妥善处理节块内由于燃料组件定位格架或部分插入的控制棒而引入的非均匀性这一弱点,提出2种可显式处理这类非均匀性的方法。以国际原子能机构(IAEA)公布的三维基准问题及实际电厂反应堆问题为例对这2种方法进行计算验证,计算验证表明,本文建议的方法可以实现在粗网节块法的框架内直接显式处理燃料组件的轴向非均匀性。     

超级均匀化方法在压水堆堆芯Pin-by-pin计算中的应用与研究

在压水堆堆芯Pin-by-pin计算中,采用超级均匀化（SPH）方法作为均匀化技术,对燃料组件传统SPH因子进行计算,生成了Pin-by-pin等效均匀化参数。针对存在中子泄漏现象的反射层组件,研究了与空间泄漏相关的SPH方法,在保证反应率守恒的基础上,同时保证各栅元各能群的中子泄漏率守恒,解决了存在中子泄漏时SPH因子迭代计算的不收敛问题,生成了反射层组件的等效均匀化参数。基于KAIST基准题,分析了压水堆堆芯Pin-by-pin计算中应用SPH因子的堆芯计算精度。数值结果表明,与传统组件均匀化计算方法相比,应用SPH方法的压水堆堆芯Pin-by-pin计算的计算精度更高。     

基于BEAVRS基准题高保真建模的OpenMC程序和NECP-X程序的对比验证

基于BEAVRS2.0.1基准题进行高保真建模,构建了含有193个燃料组件的压水堆和含有21个燃料组件的压水堆堆芯模型。应用确定论一步法程序NECP-X和概率论蒙特卡罗程序OpenMC分别对两种模型进行建模,计算热态零功率条件下堆芯的有效增殖因子、组件功率的分布以及各个控制棒组的控制棒价值。对比验证计算结果表明：热态零功率状态下有效增殖因子偏差在1.40×10^-3以内,不同控制棒组插入状态下有效增殖因子偏差低于5.9×10^-4,控制棒价值偏差均在4.9×10^-4以内;不同控制棒组情况下堆芯功率分布的平均相对偏差均在0.6%以内。初步验证了两个程序对复杂堆芯精细建模计算的可行性和准确性,对程序的应用及完善具有参考意义。     

燃料组件弯曲的高保真全堆芯中子学计算方法及应用

为了研究燃料组件弯曲变形对堆芯功率分布的影响，提出了一种等效模拟压水堆堆芯内燃料组件弯曲的方法，即根据弯曲前后燃料组件四周的水隙材料的原子数目守恒原则，通过保持弯曲前后的水隙宽度不变，改变弯曲后水隙内所有核素的原子核密度，近似等效燃料组件弯曲后四周水隙的变化。通过蒙特卡罗程序NECP-MCX和确定论数值反应堆程序NECP-X对其正确性进行验证，并基于NECP-X程序对欧洲先进压水堆（EPR）全堆芯的燃料组件弯曲工况进行了模拟分析，计算结果表明：由于局部慢化效应变化，燃料组件小幅弯曲对堆芯功率分布影响相对较大，全堆芯问题中最大的偏移量在2 mm左右时可使组件功率的相对变化达到5%左右。     

压水堆堆芯Pin-by-pin均匀化计算中非均匀泄漏修正模型研究

在压水堆堆芯Pin-by-pin均匀化计算中采用均匀泄漏修正模型及非均匀泄漏修正模型对组件计算的中子能谱进行修正,本文研究了Pin-by-pin均匀化计算中均匀泄漏修正模型及非均匀泄漏修正模型的实现方式,提出了非均匀泄漏修正模型和栅元均匀化方法的联合实现方式,并分析比较了不同栅元均匀化扩散系数产生方式的计算效果。数值结果表明,非均匀泄漏修正模型及由其产生的中子泄漏系数能有效提高压水堆堆芯Pin-by-pin计算的精度。     

德国推出压水堆新燃料组件

【德国《原子经济》1991年第4期第160页报道】西门子公司能源部(即 KWU)将压水堆的新一代燃料组件,即 FOCUS 燃料组件投放市场。这种型号的燃料组件的定位格架使用锆合金。新的定位格架能更有效地利用堆芯中子、提高负荷转换灵活性和进一步     

NHR200-II燃料组件定位格架简化建模方法研究

提出了一种基于NHR200-II供热堆燃料组件定位格架的简化模型。简化建模方法包括2方面:将定位格架上的内刚凸及三弯弹簧用非线性连接器代替;使用梁单元代替实际燃料棒。结合前期关于NHR200-II定位格架的研究成果,确定了非线性连接器的刚度,并通过有限元软件建立了燃料组件简化前后的1×2局部子模型,分析了其固有频率与碰撞特性,证明了简化建模方法的有效性。随后,该简化方法被应用于全尺寸的9×9定位格架模型,研究了格架夹持能力对动力学特性的影响,结果表明,该简化方法可以有效地模拟不同夹紧程度下格架的地震谱响应。综上,从有限元建模角度来看,本文提出的基于NHR200-II燃料组件定位格架的方法是有效的。      

苏联压水堆核电站燃料组件的科研与生产概况

本文介绍苏联压水堆核电站燃料组件及其结构材料的科研、生产概况.苏联 BBэP-1000压水堆燃料组件采用带有中心孔的二氧化铀陶瓷芯块、Zr+1%Nb 合金包壳,每个组件装有312根燃料棒、18个导向管和16层不锈钢定位格架,燃料棒呈六角形排列。这种堆有较高的堆芯平均功率密度和燃料比功率,并已有10座堆在运行发电,1987年其平均负荷因子为65.7%。由此可见,该燃料组件有较高的安全性和可靠性。     

不同混合堆芯方案对压水堆CIPS风险的影响研究

为提升核电站的经济效益，越来越多的机构开始研发更加高效的先进核燃料组件。当堆芯引入一种新型燃料组件时，堆芯中的两种组件就会形成混合堆芯。垢致轴向功率偏移（Crud Induced Power Shift,CIPS）作为影响核电安全运行的关键因素之一，其在混合堆芯中的研究也有着重要意义。本文针对某压水堆不同混合堆芯方案下的CIPS风险进行了分析评估，计算结果表明，当混合堆芯引入新型燃料组件较少时，CIPS风险变化较小；当引入较多新型燃料组件时，CIPS风险有所降低。研究结果为新型燃料组件入堆的安全评估提供了数据支撑，同时也为混合堆芯中CIPS风险评估提供了思路和参考。     

5×5全长棒束组件通道过冷沸腾工况下均匀与非均匀轴向功率分布对比分析

压水堆燃料组件结构采用正方形排列的棒束形式,本文采用计算流体力学（CFD）方法对5×5全长棒束中过冷沸腾传条件下的均匀轴向功率分布（U-APD）和非均匀轴向功率分布（Non-U-APD）工况进行了热工水力性能对比分析。分析结果表明,所采用的壁面沸腾模型、相间作用力界面力模型和气泡尺寸分布模型能够较好地预测5×5全长棒束组件通道过冷沸腾工况的传热过程。通过对比发现Non-U-APD工况下,棒束通道内平均空泡份额起始点较均匀加热工况提前,增长速度较U-APD工况更快。在子通道平均值方面,Non-U-APD工况下角通道末端平均空泡份额要高于U-APD工况,而中心通道基本相同。Non-U-APD工况下,在第5个和第6个搅混格架（MVG）下游,文中所分析的角通道和中心通道的液相质量流速逐渐低于U-APD工况。      

多群核数据不确定性对堆芯物理计算的影响

核数据不确定性是造成反应堆物理计算结果不确定性的重要因素之一。基于所需抽样核数据的协方差矩阵开发了随机抽样模块(Stochastic Sampling,SAMP),在此基础上利用SCALE(Standardized Computer Analyses for Licensing Evaluation)软件包实现了混合法和随机抽样法两种不确定性分析方法,以研究多群核数据不确定性对堆芯物理计算的影响。以3×3假想堆芯为对象,对两种方法进行了验证,然后应用于国际原子能机构(International Atomic Energy Agency,IAEA)燃料管理基准题中的Almaraz核电厂首循环堆芯。分析结果表明,两种方法结果符合良好,Almaraz核电厂堆芯keff不确定性约为0.5%,堆芯径向和轴向功率的最大不确定性分别为1.9%和0.45%。     

Effects of various spacer grid modeling on the neutronic parameters of the VVER-1000 reactor

The main objective of this paper is to study the effects of various spacer grid models on the neutronic parameters of a VVER-1000 reactor. Specifically, the data of the nuclear power plant at the Bushehr site, which is of a VVER-1000 type, will be studied. Three models, representing the spacer grids along the fuel assemblies are presented. These three models are the homogeneous and the heterogeneous local spacer grid models and the shroud spacer grid model. In the homogeneous and the heterogeneous models, the spacer grids are considered at their actual locations in the axial direction. The only difference between the two models is that in the homogeneous model, the spacer grids are homogenized with the coolant while in the heterogeneous model, the spacer grids are modeled around the fuel cells at their exact axial positions. In the shroud model, the spacer grids are modeled in the shroud region containing the coolant and are not necessarily placed at their appropriate axial positions.     

Preliminary study of parameter uncertainty influence on Pressurized Water Reactor core design

Sub-channel analysis can improve the accuracy of reactor core thermal design. However, the important initial parameters contain various uncertainties during reactor operation. In this work, the Sub-channel Analysis Code of Supercritical reactor (SACOS) code, which is also applicable for Pressurized Water Reactor (PWR), was used to study the coolant flow characteristic and fuel rod heat transfer characteristic of 1/8 assembly which has the maximum linear power density in 300 MWe PWR core firstly. Then the Wilks' method and Response Surface Method (RSM) were utilized to determine the influence of sub-channel input parameters uncertainties on the highest temperature of reactor core fuel rod and Minimum Departure from Nucleate Boiling Ratio (MDNBR). The results show that in the most conservative conditions, the maximum temperature of the fuel rod and MDNBR were 2167.4 °C and 1.08, respectively. Considering the uncertainties of assembly inlet flow rate, inlet coolant temperature and system pressure, the 95% probability values (with 95% confidence) of fuel rod maximum and MDNBR calculated using response surface methodology were 2144.0 °C and 1.6, while they were 2137 °C and 1.74 calculated by Wilks' approach. Results show that the uncertainty analysis methods can provide larger reactor design criteria margin to improve the economy of reactor. Furthermore, the code was developed to have the capacity to perform the uncertainty study of sub-channel calculation.     

从第五不稳定核素系的特性论核燃料的有效利用

以第五不稳定核素系在生长时期的耗裂转化比不断增长的特性以及不同堆型中它的任－衍生核素的饱和含量比值并不相同的特性作为论述的依据，得出了目前从压水堆中取出的废燃料并未获得有效充分利用的论断。认为只须对废燃料经过去除裂变产物的后处理去污流程，即要重新作为动力堆的核燃料使用，避免了使铀钚分离以及＾２８５Ｕ再度浓集的流程。并对如何使用此再制的核燃料提出两种方案，分别适用于压水堆和以天然铀为燃料的坎杜重水堆     

环形燃料反应堆通量密度分布测量

相对中子通量密度分布是反应堆的重要物理参数之一,测量环形燃料零功率反应堆堆芯相对中子通量密度分布对了解环形燃料堆芯反应堆物理特性及开展安全分析具有指导意义。本文在环形燃料堆芯多边形装载下,采用箔活化法对辐照后燃料元件外表面不同位置金箔的γ活度进行测量,得到不同位置燃料元件轴向、径向的相对中子通量密度分布,并将测量值与蒙特卡罗理论计算值进行比对。结果表明：实验测量值与理论计算值最大相对偏差在12%以内,相对中子通量密度分布测量结果符合实验设计预期,现有蒙特卡罗分析手段可较好地分析堆内元件轴向通量密度分布情况。本文结果可为环形燃料的工程化应用提供重要的数据支撑。     

Thermal-hydraulic transient analysis of light water reactors using moving boundaries

This paper examines the applicability of a mathematical dynamic model developed here for the simulation of the thermal-hydraulic transient analysis for light water reactors (LWRs). The thermal-hydraulic dynamic modeling of the fuel pin and adjacent coolant channel in LWRs is based on the moving boundary concept. The fuel pin model (FUELPIN) with moving boundaries is developed to accommodate the core thermal-hydraulic model, with detailed thermal conduction in fuel elements. Some results from transient calculations are examined for the first application of the thermal-hydraulic model and the fuel pin model with moving boundaries in a boiling water reactor (BWR). An accurate minimum departure from nucleate boiling ratio (MDNBR) and its axial MDNBR boundary versus time within the fuel channel are predicted during transients. Transient analysis using a known thermal-hydraulic code, COBRA and FUELPIN linked with a PWR systems analysis code show that the thermal margin gains more by a transient MDNBR approach than the traditional quasi-steady methodology for a pressurized water reactor (PWR). The studies of the overall nuclear reactor system show that moving boundary formulation provides an efficient and suitable tool for thermal transient analysis of LWRs.     

快堆核燃料循环模式的经济性评价

以快堆核电厂的核燃料循环过程及核燃料循环模型为基础,利用注销法对2种核燃料循环方式进行经济性计算和分析;同时,也将快堆燃料循环经济性与压水堆(PWR)燃料"一次通过"的经济性进行对比。按目前价格水平计算,PWR"一次通过"的核燃料循环方式比快堆核燃料循环模式的经济性好,但随着天然铀价格的上涨以及燃料后处理技术水平的进步,快堆核燃料循环费用有望达到或低于PWR"一次通过"的核燃料循环费用。     

压水堆燃料辐照性能多物理耦合并行分析程序开发

核燃料元件是反应堆的核心部件,其性能影响反应堆的安全性与经济性,利用燃料元件性能分析程序开展燃料堆内稳态辐照性能分析对于燃料设计及安全评价具有重要意义。通过开发燃料温度分布、变形计算、裂变气体释放及内压等模型,结合燃料元件热工-力学多物理耦合计算分析耦合方案,基于先进并行计算方法构建了高性能并行化燃料性能分析程序Athena。利用典型商用压水堆核电站数据及同类程序计算结果进行了程序初步验证,结果表明Athena程序计算结果合理可靠。通过定义堆芯功率及热工水力边界条件,程序能够并行开展压水堆全堆芯燃料辐照性能分析,提高燃料辐照性能分析效率,是数值反应堆原型系统（CVR1.0）的重要组成。     

SHTV,as a technique for core calculation using spatial homogenization and temperature variation

The accuracy of static neutronic parameters in the nuclear reactors depends upon the determination of group constants of the diffusion equation in the desired geometry. Although several methods have been proposed for calculating these parameters, there is still the need for more reliable methods. In this paper a powerful and innovative method based on Spatial Homogenization and Temperature Variation (SHTV) of physical properties of a WWER-1000 nuclear reactor core for calculating the relative power distribution of Fuel Assemblies (FA) and the hot fuel rod, is presented. The method is based on replacing the heterogeneous lattices of materials with different properties by an equivalent homogeneous mixture of these material for determining the few group constants, while the effect of temperature variation in the fuel and coolant density along the axial core direction is considered. All calculations are performed using WIMS and CITATION codes. The obtained results are compared with the results of Final Safety Analysis Report (FSAR) prepared by the designer, and good agreement between the two results is shown.     

Fourier expansion as an aid in the solution of the diffusion equation for three-dimensional power distributions in water reactors

The problem of predicting axial power peaking factors in water moderated reactors is not adequately solved by so called coarse mesh methods for the solution of the neutron diffusion equation. The Fourier Expansion method, briefly described, gives an essentially continuous representation of axial power shapes and therefor a higher precision in the peaking factors.—It does this with a precision equivalent to fine mesh 3D methods. Yet, it is shown to require a factor 5–10 less numerical work than fine mesh.Applications to 3D core power calculations for different types of water reactor (the HWR, BWR and PWR) are illustrated by a range of measured and calculated axial power distributions. These applications have been collected from 10 years of experience with the method. The comparisons show that the Fourier Expansion method is well suited to LWR applications.