PT-SCWR燃料组件的物理热工耦合分析

利用蒙特卡罗程序(MCNP)和子通道程序ATHAS对压力管式超临界水堆(PT-SCWR)燃料组件进行物理热工耦合分析;这种耦合方式是合理有效的。分析结果表明:PT-SCWR组件中燃料富集度的分布对燃料组件的径向功率分布有很大影响,通过调节各圈棒束的燃料富集度,可以有效地改善径向功率分布;慢化剂厚度对棒束轴向功率分布有明显影响,当慢化剂厚度为25 cm时,轴向功率分布最接近余弦形状。     

压水堆功率分布控制与运行

一、概述理论分析计算表明:一个无干扰圆柱形裸堆轴向功率成余弦分布,径向功率成贝塞尔函数分布。在无棒条件下,压水堆轴向功率分布成近似余弦分布,而径向功率分布可以通过燃料的不同浓度分区布置、可燃毒物棒和控制棒的径向对称布置、最佳控制棒分组和提(插)棒程序等设计措施来展平,并可以精确地预测。所以,研究堆功率分布主要是研究堆轴向功率分布。堆功率分布在运行中是变化的。慢化剂温度效应、可燃毒物效应、多卜勒效应和功率水平效应等     

超临界水冷堆MOX燃料特性分析

针对超临界水冷堆组件,采用不同Pu含量的MOX燃料进行组件计算,得到不同燃料条件下的燃耗深度、功率分布因子、慢化剂温度反应性系数等结果,并对比分析在超临界水冷堆中应用MOX燃料与应用UO2燃料对组件性能的影响,以及不同Pu含量MOX燃料间的性能区别。分析结果表明,在超临界水冷堆设计中,应用MOX燃料与应用UO2燃料有相似的功率分布,应用MOX燃料可以增加燃耗深度,并有良好的慢化剂温度反应性系数。经过合理设计的MOX燃料可较好应用于超临界水冷堆中,且产生更好的性能。     

超临界六角形双排燃料组件性能分析

利用物理-热工水力耦合计算程序系统(MCATHAS)分析2种六角形双排超临界燃料组件,充分考虑了超临界水冷堆(SCWR)中冷却剂、慢化剂轴向温度、密度的剧烈变化和功率分布的相互影响。计算结果表明,双排六角形组件具有均匀慢化和充分慢化性能,文中提出的D6-1型组件在仅采用一种燃料成分、不添加可燃毒物的情形下,其径向功率峰值因子低于1.10。另外,研究表明,由于组件间隙具有较大热周和较小流通面积,需要在实际工程应用中增加隔热涂层以降低组件外盒壁的导热率。     

SCWR堆芯稳态性能分析程序计算偏差分析

针对剧烈传热情况下超临界水堆堆芯稳态性能分析程序SNTA与SRAC堆芯轴向功率分布计算结果偏差较大的问题,分析偏差产生的主要原因。逐一排查影响因素,确认轴向功率分布偏差主要源于截面反馈作用不同。SRAC程序与SNTA程序采用的截面数据库和能群结构不同,SRAC程序计算的反应性密度系数相对较大,密度分布与功率分布的反馈作用更为显著,轴向功率分布曲线更为陡峭。相较于SRAC程序,采用精细能群结构的SNTA程序更适用于具备强核热耦合特性且中子能谱偏硬的超临界水堆堆芯的耦合计算与性能分析。     

超临界水堆的燃料组件设计研究

与现有的轻水堆相比,欧洲高性能轻水概念堆（HPLWR）不但具有更高的系统压力（超过水临界点）,而且具有更高的堆芯冷却剂温升和堆芯出口温度,因此,发电厂汽轮机的发电功率和热效率也更高。在HPLWR中,有7种以上的因素会导致堆芯冷却剂密度发生强烈变化,因此需要为其开发新型燃料组件。系统的设计研究表明：在减少结构材料、优化慢化剂一燃料比和展平燃料棒功率等方面,布置有两排燃料棒及一个中心位置的慢化剂盒的方型燃料组件是最佳的。利用中子学和热力学分析,已完成了HPLWR燃料组件的详细力学设计。此外,提出了上管座、下管座、蒸汽腔室、下部搅混腔室以及下堆芯板等概念设计,组成HPLWR特殊的流体通道。这种设计不仅实现了慢化剂与冷却剂相向流动时的防漏,而且实现了不同介质流的均匀混合。燃料组件设计概念可作为关键部件,用于所有HPLWR的先进堆芯设计。     

改进Flower型超临界水冷快堆初步增殖研究

超临界水冷快堆集快堆和轻水堆两种特性。整个堆芯冷却剂流量仅为现BWR的1/8,中子能谱硬于普通PWR,故有一定的核燃料增殖能力。本文建立不同Flower型超临界水冷快堆堆芯物理模型,研究堆芯分区布置、冷却剂密度分层、seed及blanket组件P/D值设计、MOX燃料设计、燃料富集度分区分层布置、blanket内部通道采用贫铀冷却等方案,分析堆芯的空泡反应性、功率分布及增殖比。通过比较,得到了超临界水冷快堆的优化设计方案。     

CSR1000控制棒落入堆芯三维瞬态分析

针对超临界水堆(SCWR)控制棒落入堆芯事件特点,采用堆芯三维瞬态性能分析方法,利用开发的SCWR堆芯三维瞬态物理-热工水力耦合程序STTA,建立SCWR堆芯落棒瞬态三维计算模型和分析流程,研究分析超临界水堆CSR1000在控制棒落入堆芯瞬态过程中的堆芯性能,分析评价落棒瞬态下CSR1000堆芯的安全性能。堆芯三维落棒瞬态分析表明,当落入堆芯棒束价值较高时,落棒初期堆芯功率下降较快,之后由于水密度的反应性反馈,堆芯功率缓慢回升至新的平衡,堆芯功率下降速率超过了停堆信号整定值,将触发保护停堆;当落入堆芯棒束价值较低时,由于水密度的反应性反馈,堆芯功率下降缓慢,堆芯功率下降速率未能达到停堆信号整定值,不能触发保护停堆。控制棒落入堆芯对堆芯轴向功率分布影响很小,高价值落棒导致的落棒区域燃料组件功率坍塌相对低价值落棒更明显。无论是高价值落棒还是低价值落棒,瞬态过程中最大包壳壁面温度均低于瞬态安全限值850℃。水密度的显著反应性反馈及必要的保护停堆措施能保证CSR1000堆芯在控制棒落入堆芯过程中的安全性能。     

超临界水堆子通道分析

超临界水堆作为6种第4代未来堆型中唯一的水冷堆,具有一些独特的特点,受到了广泛重视。本工作以上海核工程研究设计院的常规压水堆子通道程序为基础,开发编制了适用于超临界水堆的子通道程序,并对典型带有慢化剂水棒的超临界水堆燃料组件进行了模拟计算,得到了堆芯子通道内的温度、燃料棒包壳温度、表面传热系数等参数的分布规律。此外,研究了不同超临界流体换热关系式对计算结果的影响,结果显示,各传热关系式的计算结果存在一定差异。     

自主化堆芯三维核设计软件COCO研发

中国广东核电集团正在开发的三维堆芯核设计软件COCO将具备堆内功率分布计算、精细功率分布计算、临界硼浓度搜索、控制棒临界搜索、核子密度计算等基本功能。COCO采用格林函数节块方法作为求解器计算堆芯的功率分布,采用单通道模型和棒传热模型来计算慢化剂的密度和燃料温度。COCO已实现从寿期初到寿期末的燃耗计算能力。通过与参考程序的数值比较发现,COCO采用的理论模型和耦合流程正确,计算精度可满足工程设计的需要。     

Core design study on CANDU-SCWR with 3D neutronics/thermal-hydraulics coupling

Using separated heavy water as moderator and supercritical water (SCW) as coolant introduces challenge for CANDU-SCWR to get a negative coolant void reactivity (CVR), due to which the moderator thickness of the fuel channel is optimized in this paper. When SCW flows through the core, there is a rapid variation in SCW density, which is directly related to the neutron spectrum and subsequently to the power distribution, so the 3D core neutronics/thermal-hydraulics coupling is needed to accurately evaluate the core coolant density and power distribution. In this paper, the neutronics calculation is computed with 3D fine mesh diffusion code while the thermal-hydraulic calculation is based on single channel model, they are coupled with each other automatically by a link code. Further, the in-core fuel management can be simulated by the link code to search the equilibrium cycle. Based on these calculation models, a CANDU-SCWR equilibrium core is designed with a thermal power of 2540 MW, the core equivalent diameter is 4.30 m and the active length is 5.94 m. A 3-batch fuel management scheme with a cycle length of 350 EFPD is used. The numerical results show that a high average outlet coolant temperature of 625 °C is achieved with a maximum cladding surface temperature less than 850 °C. The maximum linear heat generation rate is 50.6 kW/m, the average discharged burnup is 38.1 GWd/tU, and the CVR is negative throughout the cycle.     

Effects of Non-uniform Moderator Temperature on Core Power Distribution in Accelerator Driven Reactors

In a thermal reactor with moderators at different temperatures, a difference arises in the average speeds of thermalized neutrons between the high temperature part and the low temperature part of the moderator, and the non-uniformity of the average speed of thermalized neutrons may effect changes in the spatial dependence of the thermal neutron flux in a core. To investigate the thermal neutron flux in the case the average speed of thermalized neutrons is dependent on the position within a core, time-dependent two-group diffusion equations were applied. The influence of a nonuniform moderator temperature on the core power distribution was investigated about a graphite-moderated subcritical reactor driven by periodic injections of pulsed fast neutrons. The cylindrical reactor model by which a high temperature part of a core that has a spallation target at the center is enclosed by a low temperature part of a core was used. Changes in the core power distribution were calculated. It turned out that the momentary increases of a thermal power density caused by periodic injections of pulsed fast neutrons increase as the difference in the average speeds of thermalized neutrons in the high temperature part and the low temperature part of a core increases.     

Optimization of Axial Enrichment and Gadolinia Distributions for BWR Fuel under Control Rod Programming

The axial enrichment and gadolinia distributions of BWR (boiling water reactor) fuel are optimized under control rod programming. The objective of the problem is to minimize the average enrichment required to reach a planned EOC (end-of-cycle) with criticality condition and axial power peaking constraint.

A method of approximation programming is employed as the basis for the solution method. Resulting linear programming problem at each iteration step is solved by means of goal programming algorithm. The method is applied to the initial fuel for a typical BWR/5 represented by an axial one-dimensional core model

Two-region analysis leads to the conclusion that the core bottom should be depleted during the cycle so that the power shifts to the core top at EOC. The enrichment and gadolinia distributions are determined to maximize EOC power peaking within a limit. The optimal solution of a 24-region fuel with a power peaking limit of 1.4 saves 10.6% in uranium ore compared with a uniform fuel depleted with a Haling power shape. Half the saving comes from an optimal natural uranium blanket implementation.     

On certain aspects of pressure transient without scram in a BWR

This paper examines the effects of gamma and neutron heating and pressure wave propagation on the core response during an instantaneous loss of condenser vacuum transient without scram (ATWS) in a BWR. By incorporating the gamma and neutron heating, which contribute about 3% of the total power to the moderator, into the transient thermal-hydraulic analysis, the peak power was found to be 35% lower compared with the case with no direct heating. The incorporation of the two-dimensional radial and axial variations of reactivity feedbacks and moderator density into the transient analysis led to a lower power prediction than the one-dimensional model. The pressure surge was examined by a computer program based on the method of characteristics. The pressure rise calculated by this new code was found to be in good agreement with experimental data, while the results of similar calculations done by computer code RELAP4, which is based on finite differencing of the flow equations, were 50% lower.     

基于热-流-固耦合的燃料元件性能分析方法研究

中空棱柱形燃料元件形式和运行工况特殊,没有现成的燃料性能分析软件能够满足计算要求,需要建立新的分析方法。本研究基于COMSOL软件二次开发,采用颗粒增强复合材料的等效物性模型和共轭传热技术实现中空六棱柱形燃料的三维热-流-固耦合计算,通过与美国通用电气公司数据的对比证明了该分析方法的有效性。采用该方法计算了多种燃料元件尺寸和不同轴向功率分布下的热应力和温度,结果表明侧棱处温度最高而内壁面壁厚最薄处热应力最大,壁厚越薄、长度越长,燃料元件的最大热应力和温度越小,展平入口段的轴向功率分布也能够略微降低最大热应力和温度。以上分析方法可以用于新型中空棱柱形燃料元件的优化设计。      

Stability analysis of parallel-channel systems with forced flows under supercritical pressure

The flow in the core of supercritical water reactors (SCWRs) experiences drastic change in its thermodynamic properties and transport properties near the pseudo-critical temperature, thus the core flow may be susceptible to density wave oscillation instability, which is a challenge to the system safety and must be studied carefully. This paper studies the dynamic stability characteristics of the fast-spectrum zone of a newly designed mixed-spectrum SCWR (SCWR-M), which is characterized as a parallel-channel system. A frequency-domain model has been developed for linear stability analysis, and marginal stability boundaries under several conditions for the parallel-channel system are generated, which indicate that the system normal operational condition is in the stable region. The stability of parallel-channel systems is dominated by the hottest channel. The higher the power density of the hottest channel is, the less stable the system will be. Increasing mass flow is beneficial for the system stability. Systems with uniformly axial power distribution are less stable than those with cosine-shaped or stair-shaped axial power distributions. In the time-domain, a single-phase one-dimensional model has been developed for non-linear analysis, and several perturbation transients are calculated. The results of analysis show a good agreement with that of the frequency-domain analyses, and the existence of transitional stable region has been verified.     

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.     

Ultra-fine-group resonance treatment using equivalent Dancoff-factor cell model in lattice physics code GALAXY

In order to achieve highly accurate resonance calculations with short computation time , a new ultra-fine-group resonance calculation method is developed. The ultra-fine-group method has a limitation in practical design applications of large and complicated geometries in fuel assembly level due to its long computation time. Therefore, we developed an enhanced one-dimensional (1D) cylindrical pin-cell model to achieve both high calculation accuracy and short computation time. In the enhanced 1D cylindrical pin-cell modeling, moderator radius is adjusted to preserve each fuel pellet's Dancoff factor obtained in the exact 2D fuel lattice arrangement. We call this model the ‘equivalent Dancoff-factor’ cell model. This model can accurately consider heterogeneity effects in PWR fuel assemblies and can represent effective cross sections obtained by the ultra-fine-group calculations in the complicated 2D square lattice arrangements. The present method is implemented with Mitsubishi Heavy Industries, Ltd. lattice physics code GALAXY. From the comparisons of neutron multiplication factors and pin power distributions between GALAXY and a continuous-energy Monte Carlo code, applicability of the present method to lattice physics calculations is confirmed. Application of GALAXY with the present method achieves high accuracy with short computation time in normal operations and accident conditions including low moderator density conditions.     

压水堆核电厂堆芯功率能力验证分析

介绍了压水堆核电厂换料堆芯功率能力验证分析的原理和方法。利用中子学计算程序对换料堆芯正常运行工况（一类工况或工况I）和中等频率事故工况（二类工况或工况II）中可能的堆芯功率分别进行模拟。从反应堆物理和热工水力学的角度论证反映一、二类工况堆芯安全性的线功率密度裕量和偏离泡核沸腾比（DNBR）裕量。从而验证一类工况反应堆运行区域和二类工况超漏、超功率保护限值。本文还给出了大亚湾核电站18个月换料堆芯功率能力验证分析的结果。