液态金属冷却快堆子通道分析软件SACOS-LMR研发与工程应用

子通道分析方法是反应堆堆芯设计和热工水力分析的重要手段之一,对于我国提出的压水堆-快堆-聚变堆三步走核能发展战略,开发适用于液态金属冷却快堆热工安全分析的子通道分析程序具有重要意义。本文基于西安交通大学热工水力研究室自主开发的压水堆子通道程序SACOS,通过添加液态金属快堆特有的模型,如绕丝模型、盒间流模型、液态金属对流换热模型等,扩展至适用于液态金属快堆的子通道分析程序SACOS-LMR,该程序具备对液态金属快堆组件开展稳态和瞬态热工水力分析的功能。结合卡尔斯鲁厄开展的37棒钠冷瞬态实验,完成了SACOS-LMR程序的瞬态功能验证。基于验证后的SACOS-LMR程序,对欧洲铅冷快堆(ALFRED)堆芯开展了稳态工况和瞬态事故工况下的热工安全特性分析,计算结果合理,且与同类程序保持一致,表明SACOS-LMR程序可用于液态金属快堆的堆芯设计和热工水力分析研究。     

基于双区域模型的钠冷快堆组件子通道分析程序的开发与验证

子通道分析程序是钠冷快堆堆芯热工水力设计和安全分析的重要工具。本文为计算和分析钠冷快堆组件在径向均匀与倾斜功率分布工况下的热工水力特性,利用双区域绕丝交混模型开发了一款适用于钠冷快堆组件分析的子通道程序SPLICA,并与FFM2A 19棒束实验数据与WARD 61棒束实验数据进行了对比验证。由于本文开发的子通道分析程序SPLICA使用了详细的绕丝交混模型,与经过二次开发后的COBRA程序的计算结果相比,对于FFM2A实验SPLICA程序计算得到的结果与实验结果符合得更好。这两个实验数据的验证结果证明了本文开发的子通道分析程序的准确性以及对高流量工况和低流量工况均具有良好的适用性。本程序能为钠冷快堆组件热工水力分析提供有效的设计和研究手段。     

铅铋冷却快堆含绕丝燃料组件子通道程序开发与验证

由于铅铋冷却剂流动传热现象的复杂性,准确计算铅铋冷却含绕丝燃料组件的冷却剂和包壳温度是液态金属冷却快堆燃料组件热工分析的重点。本文基于集总参数法对守恒方程进行求解,开发了适用于铅铋冷却快堆的子通道分析程序,对液态铅铋在棒束燃料组件中的摩擦阻力模型、湍流交混模型和对流换热模型进行了适用性分析,并对7棒束大涡模拟和19棒束含绕丝传热实验进行了对比验证。结果表明：包壳和冷却剂温度的最大相对误差低于5%。程序能较好完成铅铋冷却含绕丝燃料组件的热工水力计算,可为铅铋冷却快堆设计提供支持。     

钠冷快堆自然循环组件子通道程序初步研发

本文为计算和分析钠冷快堆自然循环组件的热工水力特性,开发了钠冷快堆堆芯自然循环冷却组件子通道分析程序。基于61棒单组件模型,通过将本程序的结果与COBRA程序进行比较,验证了钠冷快堆堆芯自然循环冷却组件子通道分析程序对自然循环冷却组件的适用性。基于多盒组件模型,初步验证了本程序具备自然循环冷却组件的流量分配和盒间换热计算的功能。本程序能为池式快堆自然循环冷却组件提供有效的设计和分析工具。     

氯盐冷却快堆子通道程序开发及初步验证

氯盐冷却快堆属于第四代先进反应堆的熔盐堆类型之一,采用高温氯盐作为冷却剂,具备高温、常压、较硬的中子能谱等特点,在固有安全性以及经济性上具有极大的优势和潜力。为计算和分析氯盐冷却快堆热工水力特性,自主开发了氯盐快堆子通道分析程序。基于7棒束几何结构模型,通过与Fluent计算结果相比较,重点分析和验证了氯盐快堆子通道分析程序采用的压降模型以及湍流交混模型。结果表明:氯盐快堆子通道程序所采用的Cheng-Todreas压降模型和Rogers-Rosehart湍流交混模型与Fluent在稳态与瞬态情况下的计算结果吻合很好,初步验证了所选模型的正确性和适用性。氯盐快堆子通道分析程序的开发和初步验证,将为新型氯盐冷却快堆提供有效的设计和分析工具。     

钠冷快堆乏燃料组件热工水力分析程序开发

为了对示范快堆乏燃料组件的热工水力特性进行分析,自主研发了钠冷快堆乏燃料组件热工水力分析程序SPATANS。该程序基于子通道分析方法,采用适用于低流量下的流动换热和交混关系式。针对乏燃料组件棒束区进行计算,得到组件不同高度处各子通道的温度、压力等热工参数,并将计算结果与三维计算流体力学FLUENT程序的结果进行对比分析。结果表明：自主研发程序的计算结果与FLUENT程序的计算结果较为吻合,偏差在工程可接受范围内,且其计算效率明显高于FLUENT程序。初步表明SPATANS程序可用于钠冷快堆乏燃料组件热工水力分析,并具有良好的应用前景。     

钠冷快堆乏燃料组件热工水力分析程序开发

为了对示范快堆乏燃料组件的热工水力特性进行分析,自主研发了钠冷快堆乏燃料组件热工水力分析程序SPATANS。该程序基于子通道分析方法,采用适用于低流量下的流动换热和交混关系式。针对乏燃料组件棒束区进行计算,得到组件不同高度处各子通道的温度、压力等热工参数,并将计算结果与三维计算流体力学FLUENT程序的结果进行对比分析。结果表明:自主研发程序的计算结果与FLUENT程序的计算结果较为吻合,偏差在工程可接受范围内,且其计算效率明显高于FLUENT程序。初步表明SPATANS程序可用于钠冷快堆乏燃料组件热工水力分析,并具有良好的应用前景。     

超临界水冷堆燃料棒性能分析程序适用性研究

根据超临界水冷堆(SCWR)燃料棒的热工水力特点,基于压水堆(PWR)燃料棒性能分析程序的理论模型和计算方法研究燃料包壳的物性模型和超临界水(SCW)与燃料包壳的传热模型,建立适用于SCWR燃料棒的性能分析程序——SCWRFPA。采用SCWRFPA和可分析SCWR的热工水力子通道程序ATHAS分别对1/8欧洲超临界轻水堆(HPLWR)燃料组件燃料棒进行计算,其计算结果基本一致。     

铅铋冷却快堆子通道热工水力初步数值分析

为探索铅铋冷却快堆子通道的热工水力特性,自主研发了SACOS-PB子通道程序。本工作以矩形通道9根棒束组件为例,使用SACOS-PB程序对铅铋冷却快堆子通道的温度场进行了模拟分析,并用CFX软件进行验证。结果显示,SACOS-PB程序计算结果与文献值比较符合,与CFX软件计算结果符合度也较高。使用SACOS-PB程序分析比较了3种组件结构,表明在铅铋冷却快堆中更适宜使用六边形通道,为进一步对铅铋冷却快堆子通道进行热工水力特性分析奠定了基础。     

海洋条件下反应堆热工水力参数的子通道计算

针对海洋条件下反应堆的子通道热工水力分析,建立了海洋运动附加力模型和瞬态入口边界,将起伏、摇摆及复合运动的附加力关系式用于子通道模型的轴向和横向动量方程,并应用到COBRAⅢC程序将其改造为适应海洋条件的反应堆子通道分析程序。作为验证,计算了加热实验通道和"奥陆"堆在起伏运动情况下热通道的临界热流密度比(CHFR)、出口空泡份额和冷却剂流量,并与文献结果对比。还详细计算了"奥陆"堆在起伏、不同摇摆中心和复合运动情况下,热通道的CHFR和不同位置子通道出口的热工水力参数。研究表明:海洋条件下反应堆的子通道热工水力参数随运动呈周期性变化;起伏运动对子通道的压降影响较大,摇摆运动对子通道冷却剂的流量和温度影响较大。     

Diana — A fast and high capacity computer code for interchannel coolant mixing in rod arrays

A mathematical model and digital computer program are presented for the subchannel thermal and hydraulic analysis of sodium-cooled fast reactor fuel assemblies. The newly developed FORTRAN-IV computer code ‘DIANA’ is much more useful than many other subchannel mixing analysis codes, especially for large size fuel assemblies which contain more than about 80 subchannels, and for assemblies undergoing swelling and thermal bowing which cause deformed coolant flow ducts, because of high computing speed, reduction of necessary core memory and accurate solution by momentum conservation. Numerical solutions are presented for a deformed rod bundle which contains 179 subchannels.     

Subchannel analysis of fuel assemblies of a lead–alloy cooled fast reactor

Lead–alloy cooled fast reactor is one of the six Gen-IV reactors. It has many attractive features such as excellent natural circulation performance, better shielding against gamma rays or energetic neutrons and potentially reduced capital costs. A natural circulation lead–alloy cooled fast reactor with 10 MWth is under design in China (hereafter called LFR-10MW). Fuel assemblies thermal hydraulic analysis is of vital importance for a successful design. A subchannel analysis code with flow distribution model was used to carry out the thermal hydraulic analysis. This work briefly gave the thermal-hydraulic design for the LFR-10MW and analyzed the thermal-hydraulic characteristics under steady-state condition using the subchannel analysis code. Whole core analysis was performed to locate the hottest fuel assembly using the code. The hottest fuel assembly was analyzed to obtain the cladding temperature, fuel temperature and coolant velocity. The maximum cladding temperature, the maximum fuel center temperature and the maximum coolant velocity are all below the design constraints. These results imply that the thermal-hydraulic design of LFR-10MW is feasible.     

On the use of RELAP5-3D as a subchannel analysis code

The capabilities of the RELAP5-3D code to perform subchannel analyses in sodium-cooled fuel assemblies were evaluated. The motivation was the desire to analyze fuel assemblies with traditional (solid pins) as well as non-traditional (e.g., annular pins with internal cooling, bottle-shape) geometries. Since no current subchannel codes can handle such fuel assembly designs, a new flexible RELAP5-based subchannel model was developed. It was shown that subchannel analysis of sodium-cooled fuel assemblies is indeed possible through the use of control variables in RELAP5. The subchannel model performance was then verified and validated in code-to-code and code-to-experiment analyses, respectively. First, the model was compared to the SUPERENERGY II code for solid fuel pins in a conventional hexagonal lattice. It was shown that the temperature predictions from the two codes agreed within 2% (<3.5 °C). Second, the model was applied to the Oak Ridge 19-pin test, and it was found that the measured outlet temperature distribution could be predicted with a maximum error of 8% (<7 °C). Furthermore, the use of semicircular ribs on the duct wall to flatten the temperature distribution in a traditional hexagonal assembly was explored by means of the newly developed RELAP5-3D subchannel model; the results are reported here as an example of the model capabilities.     

Experimental investigations of the hydraulic field in wire-wrapped LMFBR core assemblies

An investigation of the hydraulic behavior of wire-wrapped fuel and blanket assemblies was conducted in an air flow test facility. The test section was a large scale sector (slightly more than one-sixth) of prototypic fuel and blanket assemblies of the Clinch River Breeder Reactor Plant; the scale factor was approximately 11:1 and 5:1 for the fuel and blanket, respectively, thus allowing a very large number of measurements within each subchannel.The purpose of these experiments is discussed along with a brief state of the art review; also discussed is the role of these tests on the core thermal-hydraulic design through calibration and verification of the analytical codes employed in the design. The test section and experimental procedures are illustrated. Experimental results are discussed in detail: static pressure gradients; local and average cross flow through the gap spacing between rods as a function of the wire wrap position and at all typical locations in the assembly; detailed axial velocity mappings in the inboard and peripheral channels. The physical significance of the results is interpreted and the fundamental difference in the hydraulic behavior of fuel and blanket assemblies is pointed out, discussed and explained in terms of fundamental geometric parameters. The application of the fuel assembly data to calibration/verification of subchannel analysis and distributed parameter codes is presented in detail. A quantitative model of the cross flow driving forces is elaborated as the starting point for a comprehensive phenomenological modeling of the hydraulic behavior of wire-wrapped assemblies.     

A subchannel analysis code MATRA-LMR for wire wrapped LMR subassembly

In sodium cooled liquid metal reactors design limits are imposed on the maximum temperatures of the cladding and fuel pins. Thus an accurate prediction of the core coolant/fuel temperature distribution is essential to LMR core thermal hydraulic design. The detailed subchannel thermal hydraulic analysis code MATRA-LMR is being developed for LMFBR core design and analysis based on COBRA-IV-I and MATRA. The major modifications and improvements implemented in MATRA-LMR are as follows: sodium property calculation subprogram, sodium coolant heat transfer correlations, and most recent pressure drop correlations. To assess the development status of this code, benchmark calculations were performed with the ORNL 19 pin tests and EBR-II seven-assembly SLTHEN calculation results. The calculation results of MATRA-LMR were compared to the measurements and to the SABRE4 and SLTHEN code calculation results, respectively. Finally, the major technical results of the conceptual design for the KALIMER U-10%Zr binary alloy fueled core have been compared with the calculations of the MATRA-LMR, SABRE4 and SLTHEN codes.     

基于CFD方法格架局部阻力系数计算模型研究

为了研制高性能燃料组件,定位格架的阻力特性直接关系到燃料组件的热工性能和水力相容性。本文针对5×5规模的定位格架,从流动阻力的基本原理出发,利用CFD方法研究并建立了格架局部阻力特性的理论计算模型,并对计算结果进行验证。结果表明,基于计算模型获得的格架局部阻力系数与直接模拟结果基本一致,验证了计算模型的准确性。     

Channel deformation analysis for fast reactor fuel assemblies undergoing swelling and thermal bowing

Differential thermal expansion and swelling of fuel pins, hexagonal flow ducts and fuel spin spacers, fuel pin bowing between the spacers due to subchannel temperature differentials, and fuel pin bundle bowing due to the cambered hexagonal wrapper tube were analyzed for sodium-cooled fast reactor fuel assemblies.     

SIMPLE-2: A computer code for calculation of steady-state thermal behavior of rod bundles with flow sweeping

A computer code has been developed for use in making single-phase thermal hydraulic calculations in rod bundle arrays with flow sweeping due to spiral wraps as the predominant crossflow mixing effect. This code, called SIMPLE-2, makes the assumption that the axial pressure gradient is identical for each subchannel over a given axial increment, and is unique in that no empirical coefficients must be specified for its use. Results from this code have been favorably compared with experimental data for both uniform and highly nonuniform power distributions. Typical calculations for various bundle sizes applicable to the LMFBR program are also included.     

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。