数值反应堆堆芯通道级三维热工水力程序CorTAF开发及初步验证

堆芯是核动力系统的核心部件,其完整性是反应堆安全运行的重要前提。传统核反应堆堆芯热工水力分析方法无法满足未来先进核动力系统的高精度模拟需求。本文依托开源CFD平台OpenFOAM,针对压水堆堆芯棒束结构特点建立了冷却剂流动换热模型、燃料棒导热模型和耦合换热模型,开发了一套基于有限体积法的压水堆全堆芯通道级热工水力特性分析程序CorTAF。选取GE3×3、Weiss和PNL2×6燃料组件流动换热实验开展模型验证,计算结果与实验数据基本符合,表明该程序适用于棒束燃料组件内冷却剂流动换热特性预测。本工作对压水堆堆芯安全分析工具开发具有参考和借鉴意义。     

CSR1000堆芯流量分配的研究

以中国百万千瓦级超临界水冷堆(CSR1000)堆芯为研究对象,建立热工水力计算模型,计算出冷却剂和慢化剂温度分布、堆芯功率分布、燃料组件出口压力及流量分配等参数。计算结果表明,适当增加堆芯内部燃料组件流量比例,可以有利于径向功率展平,内外燃料组件通道出口压降,呈现"N"型变化,增大内部燃料组件的堆芯入口功率,内部组件内的流量分配也将减少,而外部燃料组件通道中的流量将增加,适当调整堆芯入口流量初始分配比例,可以使各通道功率分布展平。     

Concept of a pressurized water reactor cooled with supercritical water in the primary loop

A novel concept of a pressurized water reactor with a primary loop cooled with supercritical water is introduced and analyzed in this work. A steam cycle analysis has been performed to illustrate the advantages of such a nuclear power plant with respect to specific power and thermal efficiency. Moreover, a reactor pressure vessel concept including its internals and a suitable core and fuel assembly design are presented overcoming the problems, which arise due to the high heat up of the coolant and the density change involved with it. The core power and coolant density distributions are predicted with coupled neutronic and thermal-hydraulic analyses. The method features the definition of inlet orifices for coolant mass flow adjustment within the core as well as an additional tool for the interpolation of local pin power data. The latter one has been used for a successive sub-channel analysis of the hottest fuel assembly of the core, which provides a more detailed spatial resolution and thus predicts peak cladding temperatures, the maximum linear pin power of fuel pins, and maximum fuel temperatures. It can be shown that maximum temperatures of claddings and fuel are well below the material limits. Thanks to an average core exit temperature below the pseudo-critical temperature, the core concept leaves enough margin for additional uncertainties and allowances for operation.     

超临界水冷堆环形燃料组件核热耦合分析

在超临界水冷堆预概念设计中,组件设计是十分重要的,将影响堆芯性能。超临界水冷堆中水密度变化剧烈的特性要求必须进行核热耦合分析。从中子学及热工性能角度,使用三维核热耦合程序对环形燃料组件进行了优化设计。应用中子学计算程序FENNEL-N对环形燃料组件进行三维扩散计算,可得到组件内单棒功率分布,应用热工计算程序SUBSC对组件进行子通道分析。在计算过程中,分析了燃料棒间距及燃料棒与组件壁盒之间的间隙对组件性能的影响。计算结果显示,增大棒间距和棒壁间隙能提高组件k_inf,但会增大组件内功率峰因子;子通道受热不均匀性对组件热工性能影响较大,通过加入定位格架的方式能展平冷却剂出口温度,降低最大包壳温度。对环形燃料组件的安全分析表明,从中子学角度该组件是安全的。     

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.     

International Solvent Extraction Conference ISEC ′90

The supercritical-water-cooled power reactor (SCPR) is expected to reduce power costs compared with those of current LWRs because of its high thermal efficiency and simple reactor system. The high thermal efficiency is obtained by supercritical pressure water cooling. The fuel cladding surface temperature increases locally due to a synergistic effect from the increased coolant temperature, the expanded flow deflection due to coolant density change and the decreased heat transfer coefficient, if the coolant flow distribution is non-uniform in the fuel assembly. Therefore, the SCPR fuel assembly is designed using a subchannel analysis code based on the SILFEED code for BWRs.

The SCPR fuel assembly has many square-shaped water rods. The fuel rods are arranged around these water rods. The fuel rod pitch and diameter are 11.2 mm and 10.2 mm, respectively. Since coolant flow distribution in the fuel assembly strongly depends on the gap width between the fuel rod and the water rod, the proper gap width is examined. Subchannel analysis shows that the coolant flow distribution becomes uniform when the gap width is 1.0 mm. The maximum fuel cladding surface temperature is lower than 600°C and the temperature margin of the fuel cladding is increased in the design.     

超临界水冷堆双排燃料组件子通道分析

研究基于Cobra-IV程序,开发了适用于超临界水冷堆燃料组件分析的子通道程序.针对超临界水冷堆慢谱双排组件,进行了稳态计算,获取了相关组件热工水力参数.在此基础上,针对单一通道进行了瞬态计算,分析了燃料棒线功率变化和冷却剂流量变化条件下,超临界水冷堆燃料组件的流动和传热的动态响应,为超临界水冷堆组件的优化设计提供了参考.     

超临界水冷混合堆快谱区组件物理——热工分析

提出了超临界水冷混合堆快谱区多层燃料组件设计方案.应用MCNP程序为该组件建立计算模型,并进行了相应的物理计算;同时运用子通道分析程序STAFAS对多层燃料组件子通道进行了初步的稳态热工分析.计算结果表明:超临界水冷混合堆快谱区多层燃料组件燃料转换比超过1.0,并且获得负的冷却剂空泡反应性系数;燃料包壳表面最高温度约为595℃,低于设计准则规定的上限值,同时组件各子通道出口冷却剂温度均匀性较好.通过对燃料棒径敏感性分析可知,较大棒径组件燃料转换比较大,但也会导致热通道包壳表面温度峰值升高.     

Modeling and simulation of coupled nuclear heat energy deposition and transfer in the fuel assembly of the Ghana Research Reactor-1 (GHARR-1)

Monte Carlo N-Particle (MCNP) code coupled with PLTEMP/ANL code were used to model and simulate the heat transfer problems in the fuel elements assembly of the Ghana Research Reactor-1 (GHARR-1) by solving Boltzmann transport approximation to the heat conduction equation. Coupled neutron radiation-thermal codes were used to determine the spatial variations of thermal energy in the fuel channels, the heat energy distribution in the radial and axial segments of the fuel assembly and the convective heat transfer processes in the entire core of the reactor. The thermal energy at maximum reactivity load of 4 mk, reactor power of 30 kW and inlet system pressure of 101.3 kPa were found to be 8.896 × 10⁻¹⁶ J for a single fuel pin, and 1.104 × 10⁻¹⁵ J and 7.376 × 10⁻¹⁶ J, for the radial and axial sectioning of the core respectively. Using the PLTEMP/ANL V4.0 code and given that the inlet coolant temperature was 30 °C, the maximum outlet coolant temperature was 51 °C. The energy values were obtained using the following thermodynamic parameters as maximum pressure drop of 0.7 MPa and mass flow rate of 0.4 kg/s. Neutronics point kinetics model and Safety Analysis Report used to validate the results confirmed that the heat distribution in the core did not exceed 100 °C. The heat energy profiles based on the data suggested no nucleate boiling at the simulated energies, and since the melting point of U–Al alloy fuel material is 640 °C, the reactor was considered to be inherently safe during normal or steady state operations.     

双环路压水堆非对称入口条件下物理-热工特性研究

双环路压水堆存在反应堆入口流量、温度不对称的非正常运行工况。本文建立了基于CFD方法的反应堆整体三维流场模型,并耦合中子动力学计算程序和RELAP5程序,对这种非对称入口条件下的反应堆物理-热工特性进行了数值模拟。结果表明：反应堆入口流量不对称会加剧堆芯入口流量分配的不均匀性,并进一步导致局部功率变化,对反应堆安全不利;在入口温度不对称的条件下,冷却剂在下腔室的混合非常不充分,并导致堆芯入口温度分布不均匀,引起局部功率变化较大,对反应堆安全不利。     

压水堆燃料组件附加质量仿真研究

为了准确探究反应堆冷却剂与燃料组件间存在流固耦合行为对燃料组件振动特性的影响，本文采用计算流体动力学（CFD）软件Fluent平台，运用其中的动态网格技术，以压水堆燃料组件为研究对象，通过建立燃料组件模拟棒束、堆芯围板以及冷却剂模型，实现燃料组件与堆芯围板分别单独运动工况的燃料组件附加质量计算。结果显示：燃料组件运动工况下，燃料组件附加质量系数均值为2.4712；围板运动工况下，燃料组件附加质量系数均值为–3.4713，均与文献值偏差小于5%。叠加附加质量后，燃料组件振动频率计算值与水中振动试验测试结果偏差小于5%，验证了分析方法的合理性。本研究建立的仿真计算方法能够用于压水堆燃料组件附加质量计算。     

Thermal-hydraulics analysis during fuel element failure in an operating PWR

The fuel element failure in an operating pressurized water reactor (PWR), including fuel element breaks, has an effect on the operation safety of PWR. In this paper, the RELAP5 model of the fuel element failure is established for the safety analysis. The RELAP5 time step sensitivity analyses for the element pre-break steady and post-break transient simulation are carried out. And the variations of main thermal-hydraulics parameters related to the fuel element break are quantitatively studied, which include the internal gap pressure and the maximum fuel pellet temperature as well as the releasement of noncondensables in the gap. It is found that (1) the results by the RELAP5 code is very sensitive to the time step in a volume system with the noncondensables, and the time step sensitivity analysis is necessary if the effective time step range is unknown, (2) the larger the break area is, the more quickly the gap pressure increases and the maximum pellet temperature reaches to the stable value, (3) when the gap pressure increases and reaches to the coolant pressure, at the break the liquid inflow from coolant to gap will be turned to the vapor outflow from gap to coolant, (4) during the failure transient, the gap thermal conductivity experiences a sharp decrease in the break instant, which results in the decrease of heat transferred to cladding and the sharp decrease of cladding temperature as well as the sharp increase of minimum departure from nucleate boiling ratio (MDNBR). These conclusions can provide the basic for the operation safety analysis of PWR during the fuel element failure.     

Design analysis of core assemblies for supercritical pressure conditions

The increase of steam parameters to supercritical conditions could reduce the power generating costs of light water reactors significantly [Proceedings of SCR-2000 (2000) 1]. Core assemblies, however, will differ from current BWR or PWR design. In this context, this paper summarizes the main results related to a thermal-hydraulic design analysis of applicable fuel assemblies. Starting from a thorough literature survey on heat transfer of supercritical fluids, the current status indicates a large deficiency in the prediction of the heat transfer coefficient under reactor prototypical conditions. For the thermal-hydraulic design of such fuel assemblies the sub-channel analysis code Sub-channel Thermal-hydraulic Analysis in Fuel Assemblies under Supercritical conditions (STAFAS) has been developed, which will have a higher numerical efficiency compared to the conventional sub-channel analysis codes. The effect of several design parameters on the thermal-hydraulic behaviour in sub-channels has been investigated. Based on the results achieved so far, two fuel assembly configurations are recommended for further design analysis, i.e. a tight square lattice and a semi-tight hexagonal lattice.     

Design and analysis of a thermal core for a high performance light water reactor

The High Performance Light Water Reactor is a Generation IV light water reactor concept, operated at a supercritical pressure of 25 MPa with a core outlet temperature of 500 °C. A thermal core design for this reactor has been worked out by a consortium of Euratom member states within the 6th European Framework Program. Aiming at peak cladding temperatures of less than 630 °C, including uncertainties and allowances for operation, the coolant is heated up in three steps with intermediate coolant mixing to eliminate hot streaks. Different from conventional reactors, the radial power profile is intended to be non-uniform, with the highest power in the first heat-up step in the core center and the lowest power in the second superheater step to result in the same peak cladding temperatures in each region. The concept has been studied with neutronic, thermal-hydraulic and structural analyses to assess its feasibility. Coupled neutronic/thermal-hydraulic analyses are defining the initial distribution of enrichment, control rod positions and the use of burnable poisons. Sub-channel analyses predict the coolant mixing inside assemblies, and a porous media approach simulates the flow of moderator water between assembly boxes. Finally, structural analyses of the assembly boxes are needed to minimize deformations during operation. Even though the core design cannot yet considered to be final, this state of the art review shall summarize the progress achieved so far and outline the remaining challenges.     

Improved core design of the high temperature supercritical-pressure light water reactor

A new coolant flow scheme has been devised to raise the average coolant core outlet temperature of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H). A new equilibrium core is designed with this flow scheme to show the feasibility of an SCLWR-H core with an average coolant core outlet temperature of 530 °C.In previous studies, the average coolant core outlet temperature was limited by the relatively low temperature outlet coolant from the core periphery. In order to achieve an average coolant core outlet temperature of 500 °C, each fuel assembly had to be horizontally divided into four sub-assemblies by coolant flow separation plates, and coolant flow rate had to be adjusted for each sub-assembly by an inlet orifice. However, the difficulty of raising the outlet coolant temperature from the core periphery remained.In this study, a new coolant flow scheme is devised, in which the fuel assemblies loaded on the core periphery are cooled by a descending flow. The new flow scheme has eliminated the need for raising the outlet coolant temperature from the core periphery and removed the coolant flow separation plates from the fuel assemblies.     

Improved lumped parameter formulation for simplified LWR thermohydraulic analysis

Transient heat transfer in a nuclear fuel rod is modelled by an improved lumped parameter approach. Hermite approximation for integration is used to obtain the average fuel and cladding temperatures in the radial direction. Thermohydraulic behaviour of a pressurized water reactor (PWR) during reactivity insertion and partial loss-of-flow is simulated by using a simplified mathematical model of reactor core and primary coolant. Transient temperature response of fuel, cladding and coolant is analysed.     

基于MOOSE框架的五方程两相流分析程序开发

基于多物理场耦合框架MOOSE,采用五方程两相流模型开发了模块化程序ZEBRA,实现了高阶时间、空间离散格式两相流动传热问题的求解。采用Bartolomei开展的垂直圆管过冷沸腾实验对ZEBRA进行验证,在不同热流密度、质量流密度、压力工况下,将程序计算值与实验值进行了数值验证和计算分析。结果表明：ZEBRA中五方程模型预测值与实验值符合良好,沸腾起始点和空泡份额的预测合理,表明ZEBRA初步具备了处理两相流问题的能力。     

中国实验快堆热工参数的自适应BP神经网络预测方法研究

中国实验快堆（CEFR）堆芯的热工参数是否超出限值是评价反应堆安全运行的标准。本文针对燃料包壳最高温度预测问题，通过堆芯子通道分析程序COBRA生成数据样本后，开发基于BP神经网络自适应算法的智能预测程序，对于特定的单盒组件，仅需给出堆芯进口功率和流量，即可实现燃料包壳最高温度的快速准确预测。结果表明，与COBRA相比，在大规模重复性计算的场景下，自开发程序能节约大量计算时间和算力，提高燃料包壳设计和CEFR运行时的操作效率。实验分析得出BP神经网络方法的最大相对误差不超过6%，平均预测相对误差不超过3%，计算效率提升至原程序的300倍，网络模型的预测精度高，且易推广至实验快堆其他参数预测，具有很大的应用前景。     

Research on Adaptive BP Neural Network Prediction Method for Thermal Parameters of China Experimental Fast Reactor

Whether the thermal-hydraulic parameters of China Experimental Fast Reactor (CEFR) core exceed the limit is the standard for evaluating the safe operation of the reactor. For the maximum temperature prediction problem of fuel cladding, after generating the data samples by the core sub-channel analysis code COBRA, an intelligent prediction code based on adaptive BP neural network algorithm was developed in the paper. For a specific single-box component, only the core inlet power and mass flow rate were required to achieve fast and accurate prediction of the fuel cladding maximum temperature. Compared with COBRA, in the scenario of large-scale repetitive calculation, self development code can save a lot of calculation time and rescource, and improve the operating efficiency of fuel cladding design and CEFR operation. The experimental analysis shows that the maximum relative error of BP neural network method is less than 6%, the average prediction relative error is less than 3%, and the calculation efficiency is improved to 300 times of the original code. So the prediction accuracy of the network model is high, and self development code is easy to apply to other parameter predictions of the experimental fast reactor.     

Sub-channel analysis with F-COBRA-TF – Code validation and approaches to CHF prediction

Different simulation scales are applied in AREVA NP's thermal hydraulics department to predict the flow conditions in nuclear reactor cores: there are codes which regard one (or more) fuel assemblies as one lumped sub-channel. Sub-channel codes resolve each individual channel by one radial node. CFD codes use a much finer discretization in order to resolve the local flow conditions within a sub-channel, e.g. behind a spacer grid.The main focus of the present paper lies on the advanced sub-channel code F-COBRA-TF which has been established as a new part in the company's range of thermal hydraulic codes and customized to the industrial needs recently. Furthermore, an impression of the application of multiphase CFD within the course of fuel assembly optimization will be given.