中国先进研究堆稳态热工水力计算程序开发

针对中国先进研究堆(CARR)的具体特点开发了堆芯多通道热工水力计算程序ECARR。通过对全堆芯的数值模拟，得到了堆芯流量分配和非对称冷却条件下板状燃料元件的温度场，为进一步分析燃料元件的温差热应力等其它参数提供了所需数据。同时还对堆芯最热通道进行了热工水力计算及相应准则的判定。各参数符合CARR热工水力设计准则要求。     

Evaluation of Hot Spot Factors for Thermal and Hydraulic Design of HTTR

Abstract

High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal power and 950°C in reactor outlet coolant temperature. One of the major items in thermal and hydraulic design of the HTTR is to evaluate the maximum fuel temperature with a sufficient margin from a viewpoint of integrity of coated fuel particles. Hot spot factors are considered in the thermal and hydraulic design to evaluate the fuel temperature not only under the normal operation condition but also under any transient condition conservatively. This report summarizes the items of hot spot factors selected in the thermal and hydraulic design and their estimated values, and also presents evaluation results of the thermal and hydraulic characteristics of the HTTR briefly.     

Nonlinear analysis of nuclear coupled density wave instability in time domain for a boiling water reactor core undergoing core-wide and regional modes of oscillations

The objective of the paper is to develop a nuclear coupled thermal-hydraulic model in order to simulate core-wide (in-phase) and regional (out-of-phase) stability analysis in time domain within the limitation of desktop research facility for a boiling water reactor subjected to operational transients. The integrated numerical tool, which is a combination of thermal-hydraulic, neutronic and fuel heat conduction models, is used to analyze a complete boiling water reactor core taking into account the strong nonlinear coupling between the core neutron dynamics and primary circuit thermal-hydraulics via the void-temperature reactivity feedback effects. The integrated model is validated against standard benchmark and published results. Finally, the model is used for various parametric studies and a number of numerical simulations are carried out to investigate core-wide and regional instabilities of the boiling water reactor core with and without the neutronic feedback effects. Results show that the inclusion of neutronic feedback effects has an adverse effect on boiling water reactor core by augmenting the instability at lower power for same inlet subcooling during core-wide mode of oscillations, whereas the instability is being suppressed during regional mode of oscillations in presence of the neutronic feedback. Dominance of core-wide instability over regional mode of oscillations is established for the present case of simulations which indicates that the preclusion of the former will automatically prevent the latter at the existing working condition.     

CFD在核能系统分析中应用的最新进展

介绍了国内外计算流体力学(CFD)方法在核能系统分析中应用的最新进展.CFD已经可以应用到一些三维单相瞬态流动情况中,其中包括堆芯及组件内的流场模拟,以及堆芯外空腔和其他领域的模拟分析.CFD的应用还需要进行进一步验证和基准化,并需要针对CFD的应用方法进行研究,包括建立最佳实践导则(BPG).CFD程序还被用于与热工水力系统程序的多尺度耦合以及与中子物理程序的耦合.CFD程序在两相流领域的应用还处于初级阶段,在湍流和多相流的模拟中需要更进一步发展.     

压水堆核电站热工水力系统程序的研发现状与趋势

比较分析了目前世界上典型的压水堆核电站热工水力系统程序的研发历程、发展现状、应用范围,着重指出了最佳估算、程序耦合、程序评估在热工水力系统程序研发中的重要作用,阐述了各国热工水力系统程序研发模式对我国自主创新的借鉴意义。     

EAST阻性换热器多物理场耦合模拟计算

阻性换热器是EAST高温超导(HTS)电流引线的重要组成部分,目前有三头螺旋翅片和叠片两种结构形式,为了比较这两种阻性换热器的优劣,对它们的热工水力性能进行了多物理场耦合模拟计算,计算结果表明:两种阻性换热器在换热性能方面基本相当,均可满足快速换热的要求,但叠片换热器的流动阻力远小于三头螺旋翅片换热器的。实际运行过程中,三头螺旋翅片换热器中氮冷却回路的压力控制较为困难,经常需人工调节控制阀阀门,而叠片换热器中氮冷却回路的压力控制则较为简单,不需经常调节。因此,叠片式结构较三头螺旋翅片式结构更适合应用在EAST阻性换热器中。     

Development of a steady thermal-hydraulic analysis code for the China Advanced Research Reactor

A multi-channel model steady-state thermal-hydraulic analysis code was developed for the China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed mass flow distribution in the core was obtained. The result shows that structure size plays the most important role in mass flow distribution, and the influence of core power could be neglected under single-phase flow. The temperature field of the fuel element under unsymmetrical cooling condition was also obtained, which is necessary for further study such as stress analysis, etc. of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of the mean and hot channel was carried out and it is proved that all thermal-hydraulic parameters satisfy the “Safety design regulation of CARR”. Translated from Atomic Energy Science and Technology, 2006, 40(1): 51–55 [译自: 原子能科学技术]     

绕丝结构对氦氙气体流动换热特性影响研究

在开放栅格式气冷空间堆堆芯的设计中,会在燃料棒上采用绕丝结构,这将对工质的流动换热特性产生很大影响。本文采用CFD方法开展了He-Xe混合气体在气冷空间堆典型带绕丝燃料通道内的流动换热特性的数值研究,获得进出口温度、压力、流速及流体密度等参数的空间分布。结果表明引入绕丝使得范宁摩擦因子出现大幅增加,部分绕丝结构会使努塞尔数降低20%～30%,且本文所研究5种绕丝结构热工水力性能比均小于1。研究结果对于气冷空间堆堆芯热设计、提高系统的安全性具有重要意义。     

Safety analysis of reactivity insertion accidents in a heavy water nuclear research reactor core using coupled 3D neutron kinetics thermal-hydraulic system code technique

Nuclear power plant Safety analysis using coupled 3D neutron kinetics/thermal-hydraulic codes technique is increasingly used nowadays. Actually, the use of this technique allows getting less conservatism and more realistic simulations of the physical phenomena. The challenge today is oriented toward the application of this technique to the operating conditions of nuclear research reactors. In the current study, a three-Dimensional Neutron Kinetics and best estimate Thermal-Hydraulic model based upon the coupled PARCS/RELAP5 codes has been developed and applied for a heavy water research reactor. The objective is to perform safety analysis related to design accidents of this reactor types. In the current study two positive reactivity insertion transients are considered, SCRAM protected and self-limiting power excursion cases. The results of the steady state calculations were compared with results obtained from conventional diffusion codes, while transient calculations were assessed using the point kinetic model of the RELAP5 code. Through this study, the applicability and the suitability of using the coupled code technique with respect to the classical models are emphasized and discussed.     

Sub-channel-based multiscale thermal-hydraulic analysis of a medium-power,lead-cooled fast reactor

The lead-cooled fast reactor (LFR) offers enhanced safety and reliability with the fine properties of liquid lead and lead alloy. To study accurately the thermal characteristics of fast reactors, the multiscale thermal-hydraulic coupling simulation is an effective way. Multiscale coupling based on the sub-channel code has evident advantages on the analysis of fuel assemblies. In this study, a multiscale thermal-hydraulic analysis of a forced-circulation, medium-power LFR under steady-state and transient conditions is performed with the system code ATHLET and sub-channel code KMC-SUBtraC which was developed based on the previous version by modifying the pressure drop correlations and adding the assembly-level calculation. The codes are one-way-coupled, with good efficiency and precision. Transient verification of the sub-channel code is conducted with the CFD code. In the steady-state analysis of M²LFR-1000, mass flow and temperature distributions of the assemblies, sub-channels, and fuel rods in the hottest assembly are analyzed and the safety performance is investigated. In the transient analysis, two typical DECs (unprotected overpower transient and ULOF+ULOHS) are simulated and the multiscale thermal-hydraulic characteristics are analyzed. With the negative reactivity feedback, the variations of the temperatures of the coolant and fuel rods are within the safe limits, which shows the inherent safety of the reactor. And the results indicate that the loss of primary flow could increase the risk of cladding corrosion.