基于多孔介质模型的钾热管数值模拟

为研究钾热管内传热传质机理,对钾热管进行了数值模拟。建立了固液气三相耦合数学模型。其中对吸液芯液体流动区域采用了多孔介质模型,该模型考虑了液体流动对热管传热性能的影响。利用PHOENICS3.6对数学模型进行数值计算,得到了热管内的稳态工作参数。分析模拟结果得到了钾热管内部各相工质传热、传质机理,并与试验数据进行了比较。结果表明,模拟结果与试验数据符合较好。     

基于多孔介质模型的行波堆TP-1堆芯稳态温度场与流场数值模拟

根据TerraPower公司最新设计的钠冷行波堆TP-1的具体结构和运行特点,采用多孔介质模型,使用商用软件CFX对行波堆堆芯的热工水力特性进行数值模拟,得到了TP-1稳态运行条件下堆芯温度场、速度场和压力场分布。结果表明：应用多孔介质模型对行波堆堆芯进行三维热工水力数值模拟的方法直观、快速、有效,将它应用于行波堆堆芯稳态条件下三维流场和温度场分析具有一定的意义。     

CEFR堆芯出口温场和流场的三维数值模拟

池式快堆的温度场、流场是快堆重点研究课题。钠池区域较大,域内影响流动和传热的因素较多,不同区域(轴向和径向),流动形式各不相同,温度分布差异明显,采用常规的计算方法很难得到满意的结论。本文应用大型三维流体力学程序CFX,模拟计算CEFR堆芯出口区域的温度场和流场。通过计     

12 CEFR堆芯出口温场和流场的三维数值模拟

池式快堆的温度场、流场是快堆重点研究课题。钠池区域较大，域内影响流动和传热的因素较多，不同区域（轴向和径向），流动形式各不相同，温度分布差异明显，采用常规的计算方法很难得到满意的结论。本文应用大型三维流体力学程序CFX，模拟计算CEFR堆芯出口区域的温度场和流场。     

不溶性粒状腐蚀产物在压水堆堆芯内沉积的数值模拟

在反应堆运行时,由于燃料棒、堆内构件等部件会受到高压过冷态的冷却剂的腐蚀冲刷的影响,会产生许多不溶性腐蚀产物。利用FLUENT软件模拟不溶性粒状腐蚀产物在堆芯燃料棒流域里沉积分布。对液相采用标准k-ε模型预测通道内流场与近壁面区域的湍流变化,对腐蚀产物颗粒物采用DPM模型(离散相模型)来跟踪颗粒的运动轨迹。研究发现:在堆芯流域腐蚀产物颗粒在对称面附近形成高浓度区域,在入口段腐蚀产物颗粒浓度比出口段高。在包壳入口段表面呈大面积附着沉积,这会改变堆芯中子通量分布和包壳材料的热导率,引起堆芯轴向功率偏移;而在包壳出口段表面呈点状沉积,这会导致包壳出现点蚀现象。点蚀区域会引起传热恶化,破坏包壳完整性。针对腐蚀产物颗粒沉积规律和堆内组件的腐蚀特点,提出定时定点、针对局部强化清理等缓解措施。     

压水堆堆芯双节点模型及其应用

基于节点方法建立双节点反应堆模型,并利用微小摄动理论对模型进行线性化处理,得到双节点线性化反应堆多变量模型;为了更加准确地对棒控系统进行仿真,建立了双节点棒控系统模型。参考AP1000的机械补偿控制策略(MSHIM),在Matlab/Simulink平台上搭建平均温度控制系统和轴向偏移控制系统的仿真程序,并对功率阶跃负荷瞬态工况进行仿真。仿真结果表明,功率偏差始终在目标控制带以内,平均温度和功率水平的实际值与目标值之间存在稳态偏差,偏差处于死区范围之内。因此双节点线性化堆芯模型适用于功率控制系统和功率分布控制系统的仿真和设计。     

TMSR堆芯CFX多孔介质传热建模分析

钍基熔盐堆(TMSR)是一种使用石墨包覆颗粒作为燃料,熔盐作为冷却剂的第4代反应堆。TMSR堆芯区域的球形燃料增加了反应堆热工水力分析的复杂程度,为了分析反应堆在发生丧失强迫循环后堆芯的温度分布情况,需对整个堆芯进行CFD建模模拟。本文对TMSR堆芯进行几何建模和网格划分,并使用ANSYS CFX进行了多孔介质模型的建模模拟。在主要考虑导热换热和浮力影响以及两种不同的保温层厚度情况下,对堆芯稳态运行时的温度分布和发生事故后60s的瞬态温度分布进行了初步分析。研究结果证明了利用CFX及其多孔介质模型对TMSR堆芯进行模拟的可行性,并与REALP5-3D结果进行比较,初步验证了在该简化模型的边界条件下,堆芯熔盐短时间内不会发生沸腾。     

基于多模型策略的压水堆堆芯功率控制与仿真

在建立压水堆堆芯非线性模型的基础上求取5个功率水平处的线性化模型作为堆芯局部模型,以局部模型组合替代堆芯非线性模型。利用基于局部模型全维观测器的全状态反馈法,设计带有鲁棒性能的局部模型控制器作为非线性堆芯局部控制器,用于相应功率水平域内的功率控制。仿真结果表明,所设计的堆芯多模型控制系统能很好地控制堆芯功率。     

压水堆堆芯pin-by-pin计算环境效应处理模型研究

栅格非均匀计算过程中采用的全反射边界条件近似带来的中子射流效应和中子能谱干涉效应等环境效应对栅元均匀化常数具有较大影响。为在全堆芯pin-by-pin计算中处理环境效应带来的影响,本文从两个方面进行了计算分析。首先,基于棋盘式多组件问题对栅元均匀化群常数相对误差及各能群栅元不连续因子相对重要性进行了分析,可发现在等效均匀化常数中,热群不连续因子对全堆芯pin-by-pin计算精度的影响最重要;其次,基于最小二乘法建立了热群栅元不连续因子和堆芯中子学特征量之间的多项式函数关系,利用参数化技术提出了热群常数堆芯在线计算方法,其中堆芯中子学特征量包括扩散系数、移出截面、中子源项、归一化中子通量密度等。采用C5G7基准题和KAIST基准题进行了数值验证,计算结果表明,热群常数堆芯在线计算方法能有效降低全堆芯pin-by-pin计算特征值和棒功率相对误差,对处于不同燃料组件交界面附近的栅元,计算精度提升尤为显著。     

压水堆堆芯熔融物的运动及传热数值分析

针对HPR1000堆型堆芯熔融坍塌问题建立了精确的三维堆芯模型,使用时间推进方法通过求解熔融物的瞬态运动、传热微分方程,确定熔融物在堆芯中的瞬态位置和瞬时温度,以模拟堆芯升温及堆芯熔融进程。研究结果表明：停堆后约2 400 s开始出现熔融现象,熔融物在堆芯活性区域内下落且发生多重相变过程;在4 900 s后,熔融物在堆芯底部形成约1.5 m高的稳定熔池;由于外围组件与低温围栏装置换热,最外围的组件不会发生熔融。本文建立的堆芯熔融物运动与传热分析模型及相关计算结果,可为事故缓解和处理提供技术参考。     

Research on optimization design of PWR flow distribution device based on numerical simulation

ABSTRACT

In the lower chamber of pressurized water reactor (PWR), the flow distribution device is the core module to distribute coolant into the core. It has complex structure and numerous design parameters. Therefore, it has important theoretical and practical significance to optimize the device. The mesh independence verification, turbulence model selection, and data processing all can influence the numerical simulation results of the lower chamber, in order to research the influence, a numerical simulation method based on the original model of CNP1000 reactor lower chamber is proposed in this paper. In the method, an optimization design method of flow distribution device is established based on surrogate model. The main design variables and optimization objectives are determined based on the device’s structure and function characteristics. And then it respectively adopts Kriging algorithm and multi-objective genetic algorithm to establish a surrogate model of flow distribution device and optimize it globally. Finally, the optimal design variables are obtained. Compared with the device’s performance before optimization, the after optimization has smaller total pressure loss and more uniform flow. The effectiveness and practicability of proposed optimization design method can be verified.     

压水堆核电站堆芯集中参数模型的微机仿真

阐述了PWR核电站堆芯的模型化问题,提适用于微机仿真的核电站堆芯的物理数学模型,将核电站堆芯分为三大块分别建立模型,中子动力学模块,反应性反馈模块,堆芯热力学模块,建立系统传递函数,运用MATLA仿真,得到良好结果。     

A prediction model of IASCC initiation stress for bolts in PWR core internals

A model of IASCC initiation stress for bolts of core internals in pressurized water reactors was developed considering differences in material property changes due to irradiation and material conditions. Assuming that IASCC initiation was controlled by grain boundary composition and yield strength, these values for each specimen of post-irradiation IASCC initiation tests were calculated by physical kinetic models considering dose rate, temperature, material composition and surface hardening. Then, correlations of grain boundary composition and yield strength with IASCC initiation stress were determined. The model predicted that the IASCC initiation stress became lower with dose and was lower for higher temperature, lower flux and higher surface hardening level.     

Coupled neutronic thermo-hydraulic analysis of full PWR core with Monte-Carlo based BGCore system

BGCore reactor analysis system was recently developed at Ben-Gurion University for calculating in-core fuel composition and spent fuel emissions following discharge. It couples the Monte Carlo transport code MCNP with an independently developed burnup and decay module SARAF. Most of the existing MCNP based depletion codes (e.g. MOCUP, Monteburns, MCODE) tally directly the one-group fluxes and reaction rates in order to prepare one-group cross sections necessary for the fuel depletion analysis. BGCore, on the other hand, uses a multi-group (MG) approach for generation of one group cross-sections. This coupling approach significantly reduces the code execution time without compromising the accuracy of the results.Substantial reduction in the BGCore code execution time allows consideration of problems with much higher degree of complexity, such as introduction of thermal hydraulic (TH) feedback into the calculation scheme. Recently, a simplified TH feedback module, THERMO, was developed and integrated into the BGCore system. To demonstrate the capabilities of the upgraded BGCore system, a coupled neutronic TH analysis of a full PWR core was performed. The BGCore results were compared with those of the state of the art 3D deterministic nodal diffusion code DYN3D (Grundmann et al., 2000). Very good agreement in major core operational parameters including k-eff eigenvalue, axial and radial power profiles, and temperature distributions between the BGCore and DYN3D results was observed. This agreement confirms the consistency of the implementation of the TH feedback module.Although the upgraded BGCore system is capable of performing both, depletion and TH analyses, the calculations in this study were performed for the beginning of cycle state with pre-generated fuel compositions.     

IIST and LSTF counterpart test on PWR station blackout transient

A reduced-height, reduced-pressure (RHRP) integral system test facility at the Institute of Nuclear Energy Research (INER) has been established for simulating the thermal-hydraulics of a Westinghouse three-loop pressurized water reactor (PWR). To understand whether or not the physical phenomena observed in this RHRP integral system test facility during a station blackout (SB) transient can be reliably extrapolated to those for an actual plant, a counterpart test based on the same scenarios as those of the full-height, full-pressure (FHFP) large-scale test facility (LSTF) test was performed. To see the result of differences in the design, scaling approach and facility operational conditions in the systems, the present study examines their effects on the SB transient, particularly for the tests performed at full and reduced pressures. We also identify the occurrence of key thermal-hydraulic phenomena, as well as their possible distortions. Results of the INER integral system test (ISST) facility and LSTF tests showed the common thermal-hydraulic phenomena, such as the secondary coolant boil-off and the subsequent primary coolant saturation, pressurization, coolant inventory depletion and redistribution, and core uncovery caused by coolant boil-off. The sequence and timing of the significant events during the SB transient studied in the RHRP IIST facility are also consistent (in most cases) with those for the SB transient studied in the FHFP LSTF.     

Numerical simulation of a blanket cooling system for fusion reactor based on PWR conditions

The simulations of a blanket cooling system were presented to address the choice of cooling channel geometry and coolant input data which are related to blanket engineering implementation. This work was performed using computer aided design (CAD) and computational fluid dynamics (CFD) technology. Simulations were carried out for the blanket module with a size of 0.6 m × 0.45 m in toroidal plane, and the nuclear heat was applied on the cooling system at P_n (neutron wall load) of 5 MW/m². The structure factors and input data of hydraulics were investigated to explore the optimal parameters to match the PWR condition. It was found that the inlet velocity of first wall (FW) channel should be within the range of 2.48–3.34 m/s. As a result, the temperature rise (TR) of the coolant in the FW channel would be 24–25 K. This leads to the remaining space for TR within the range of 15 K in the piping circuits. It also indicated that the FW plays an important role in TR (reaches 60% of the whole cooling system) due to its high level of P_n and heat flux in the zones. It was predicted that the nuclear heat inside blanket module could be removed completely by the piping circuits with an acceptable pipe bore and the related input data. Finally, a possible design range of cooling parameters was proposed in view of engineering feasibility and blanket neutronics design.     

An improved porous media model for nuclear reactor analysis

In this study, two modifications are proposed to mitigate drawbacks of the conventional approach of using the ‘‘Porous Media Model'(PMM) for nuclear reactor analysis. In the conventional approach, whole reactor core simplifies to a single porous medium and also, the resistance coefficients that are essential to using this model are constant values. These conditions impose significant errors and restrict the applications of the model for many cases,including accident analysis. In this article, the procedures for calculating the coefficients are modified by introducing a practical algorithm. Using this algorithm will result in obtaining each coefficient as a function of mass flow rate.Furthermore, the method of applying these coefficients to the reactor core is modified by dividing the core into several porous media instead of one. In this method, each porous medium comprises a single fuel assembly. PMM with these two modifications is termed ‘‘multi-region PMM' in this study. Then, the multi-region PMM is introduced to a new CFD-based thermo-hydraulic code that is specifically devised for combining with neutronic codes.The CITVAP code, which solves multi-group diffusion equations, is the selected as the neutronic part for this study. The resulting coupled code is used for simulation of natural circulation in a MTR. A new semi-analytic method,based on steady-state CFD analysis is developed to verify the results of this case. Results demonstrate considerable improvement, compared to the conventional approach.     

Challenge of PWR new core design simulation: A focus on uncertainties due to nuclear data and reflector modelling

This paper presents a reactor core uncertainty analysis in the framework of the OECD/NEA UAM Benchmark. Three types of uncertainties affecting the predictions of power distribution in the core of a nuclear reactor are discussed: the uncertainties of basic nuclear data, the uncertainties resulting from the use of different simulation tools and those due to approximations in reflector modelling. The contribution of nuclear data uncertainty on the power distribution of a UOX and a MOX core is assessed with the XSUSA tool. Overall, the results obtained with different tools in both institutions are in good agreement, showing that the power distribution uncertainty due to the use of different simulation tools is much lower than the one due to nuclear data, which is a large contributor. Lastly, the paper presents preliminary work showing the relevance of reflector modelling on the uncertainty of the power distribution at nominal conditions as well as on an asymmetrical case representative of accidental conditions.     

Three-dimensional transport coefficient model and prediction-correction numerical method for thermal margin analysis of PWR cores

Combustion Engineering Inc. designs its modern PWR reactor cores using open-core thermal-hydraulic methods where the mass, momentum and energy equations are solved in three dimensions (one axial and two lateral directions). The resultant fluid properties are used to compute the minimum Departure from Nucleate Boiling Ratio (DNBR) which utlimately sets the power capability of the core. The on-line digital monitoring and protection systems require a small fast-running algorithm of the design code. This paper presents two techniques used in the development of the on-line DNB algorithmFirst, a three-dimensional transport coefficient model is introduced to radially group the flow subchannel into channels for the thermal-hydraulic fluid properties calculation. Conservation equations of mass, momentum and energy for these channels are derived using transport coefficients to modify the calculation of the radial transport of enthalpy and momentum.Second, a simplified, non-iterative numerical method, called the prediction-correction method, is applied together with the transport coefficient model to reduce the computer execution time in the determination of fluid properties.Comparison of the algorithm and the design thermal-hydraulic code shows agreement to within 0.65% equivalent power at a 95/95 confidence/probability level for all normal operating conditions of the PWR core. This algorithm accuracy is achieved with 1/800th of the computer processing time of its parent design code.