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.     

超临界水冷动力堆燃料组件的子通道研究分析

与目前的轻水堆相比较，由于超临界水冷动力反应堆(SCPR)的热效率高、反应堆系统简单，预计将降低发电成本高热效率通过超临界压力水冷却来获得、如果冷却剂流体在燃料组件中的分布是非均匀的．由于冷却剂温度提高、冷却剂密度的变化而出现大的流量偏移和传热系数降低的复合效应，燃料包壳的表面温度会局部升高：因此，SCPR燃料组件设计采用基于沸水堆的SILFEED的子通道分析程序SCPR燃料组件具有许多正方形水棒、燃料棒被布置在这些水捧周围。燃料棒的间距和直径分别为11．2nun和10．2mm。由于冷却剂流体在燃料组件内的分布主要取决于燃料棒和水棒之间的间隙宽度、对适当的间隙宽度进行了研究。子通道分析表明，在间隙宽度为1．0mm时，冷却剂流量分布是均匀的，最高的燃料包壳表面温度低于600℃、在设计中提高了燃料包壳的温度裕度。     

快堆燃料组件热工流体力学计算研究

对于钠冷快堆,在燃料和包壳最高温度等设计限值下,为获得较高的堆芯出口温度,需深入分析燃料组件内的热工流体力学问题,准确预测组件内的冷却剂温度分布。本文在CRT模型和F.C.Engel等人工作的基础上,提出了ICRT压降关系式,用以计算冷却剂在湍流区、过渡流区和层流区的棒束压降;引入CRT模型和WEST对流传热模型,改进了SUPERENERGY子通道分析程序,并将改进程序与原程序计算结果进行了对比,结果表明：最热子通道出口温度略有降低,液膜温压略有增加;并用计算流体力学软件CFX对中国实验快堆单盒燃料组件活性段进行了三维数值模拟,将计算结果用CRT模型、ICRT压降关系式及改进后的SUPERENERGY子通道分析程序进行了验证,相互符合较好。     

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。     

钠冷行波堆和驻波堆堆芯特性对比分析

采用自开发的MCNP-ORIGEN耦合程序MCORE对所设计的钠冷行波堆和驻波堆开展中子学和燃耗分析;基于MCORE获得的功率分布,采用自开发的钠冷快堆堆芯稳态热工水力分析程序SAST对钠冷行波堆和驻波堆堆芯开展热工水力分析。对比钠冷行波堆和驻波堆的堆芯物理特性和热工水力特性,结果表明：驻波堆在燃耗、最高包壳和燃料芯块温度方面具有优势,而行波堆在反应性波动和堆芯冷却剂出口温度均匀性方面具有优势。     

基于CFD方法的行波堆燃料组件燃烧区热工流体特性研究

采用计算流体力学（CFD）方法对行波堆燃料组件7棒束、19棒束及37棒束模型进行计算分析,发现行波堆燃料组件内冷却剂温度随轴向高度增加逐渐升高的同时具有逐渐向中心区域聚集的效应,组件出口区域垂直于流动方向的截面冷却剂温度分布差别很大,对边距约为26 cm的组件中心区域与外围区域最大温差超过100 ℃。组件内较大的冷却剂温度梯度主要出现在组件最外两圈燃料棒及组件盒之间的区域,而其他区域温度梯度较小,该结论可初步推广到有217根燃料棒的行波堆燃料组件。现有行波堆燃料组件结构需进一步优化。     

钠冷行波堆TP-1瞬态安全分析

钠冷行波堆作为一种具有潜力的新堆型,正处于概念研究阶段。本工作根据TerraPower公司最新设计的钠冷行波堆TP-1的具体结构和运行工况方案,建立其一回路主要部件的物理数学模型,用Fortran语言初步开发了钠冷行波堆瞬态安全分析程序TAST,并对钠冷行波堆稳态进行计算,表明系统程序运行稳定可靠。采用TAST对失流事故和反应性引入事故进行计算,得到关键参数的瞬态变化,初步验证了钠冷行波堆在这两个事故工况下的安全性。     

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。     

超临界水冷堆双排棒组件子通道分析

首先利用先进子通道分析程序(ATHAS)对超临界水冷堆(CGN-SCWR)的双排棒组件进行子通道分析,以考察燃料棒包壳温度等热工参数是否达到安全要求。根据分析结果结合子通道水力直径和冷却剂出口温度,选取一些典型子通道的热工参数结果做详细比对,了解组件中不同类型子通道内的热工参数变化对组件性能的影响。另外,对子通道计算采用的湍流交混系数、轴向摩擦系数和传热关系式进行敏感性分析,以了解经验关系式对计算结果的影响。结果显示:所有热工参数结果均达到设计要求,包壳最高温度为685.3℃,且不同传热关系式的选择对包壳温度的影响明显,最大温差达到了41.3℃。     

快堆燃料组件的子通道分析

开发了ＴＨＡＳ－ＰＣ２子通道分析微机程序，用于计算稳态和瞬态工况下快堆燃料组件的流量、温度和压力等参量分布。对ＥＦＲ燃料组件的稳态和瞬态计算结果如下：堆芯出口平均温度和温长分别为５５７℃和１５７℃，最高包壳表面温度为６０１℃，它发生在中心燃料棒上，最大冷却剂温度为５９３℃；主泵断电二次停堆事故作为瞬态计算，算得的最高冷却剂温度和最高包壳表面温度分别为６３０℃和６３７℃（当ｔ＝２ｓ时），它们都远低于     

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.     

Subchannel analysis of a fast reactor cooled by supercritical light water

The concept of a high temperature fast reactor cooled by supercritical water (SCFR-H) was developed for achieving high thermal efficiency and a compact reactor system. The core characteristics were obtained from single channel thermal-hydraulic analysis. Thus, it is necessary to carry out subchannel analysis to estimate the effect of local power peaking and cross flows. For this purpose, a subchannel analysis code is developed. It is verified by comparing the results with experimental data of High Conversion Pressurized Water Reactor (HCPWR). Sensitivities of the outlet coolant and cladding temperature to the subchannel flow area and local power peaking are high. One of the reasons is that the ratio of the coolant flow rate of SCFR-H to the power is smaller than that of LWR. Another reason is that, temperature of supercritical water is more sensitive to the enthalpy change above 450°C. The outlet coolant temperature distribution can be flattened by reducing the area of the peripheral subchannels and by enhancing the mixing between the subchannels.     

Thermal-hydraulic behavior of a marine reactor during oscillations

The effect of ship motion, such as heaving and rolling, on the thermal-hydraulic behavior of marine reactors was investigated. The COBRA-IV-I CODE was modified to analyse the thermal-hydraulic performance on the critical heat flux under oscillating acceleration conditions. The critical heat flux in the code was verified experimentally using freon as a comparison. The Critical Heat Flux Ratio (CHFR) at the hottest channel of the PWR subchannel was analysed using the same code. A system code RETRAN-02/MOD2-GRAV was developed by improving RETRAN-02/MOD2 to simulate the thermal hydraulic transient under ship motion. It was verified by comparison using the experimental results of both two-phase natural circulation flow under heaving motion and single-phase natural circulation flow at an inclined attitude. The code was used to analyse reactor plant behavior in the nuclear ship Mutsu. Natural circulation flow during rolling motion was investigated experimentally. The characteristics of loop flow and core flow rates were clarified. The core flow rate correlated well with the Reynolds number of rolling motion.     

Sabena: Subassembly boiling evolution numerical analysis

This paper describes the computer code SABENA that has been used in subassembly sodium boiling evolution numerical analysis as a contribution to fast breeder reactor safety analysis. SABENA is a two-fluid model subchannel code system to calculate coolant boiling and two-phase flow in a rod bundle together with external loop characteristics which affects the overall boiling behavior in the bundle section. With the use of relatively simple but reasonable constitutive models, the SABENA code has been applied to and validated against many multi-pin sodium boiling problems. The results have shown excellent agreement with the experiments. The numerical methods and models employed in the code have proven to be robust and efficient in light of the extreme severity of the conditions characterizing low-pressure sodium boiling.     

Analysis of three different types of blockage in a sodium flow path with the MATRA-LMR-FB code

A local blockage in a subassembly of an LMR is of particular importance because the local temperature of the coolant increases at the downstream of a blockage and the integrity of the fuel clad can be threatened when an obstacle or a blockage is formed in a flow path. To analyze a flow blockage in a nuclear reactor core, Korea Atomic Energy Research Institute developed a subchannel analysis computer code MATRA-LMR-FB. This code adopts several enhanced modeling features such as a distributed resistance model, state-of-the-art turbulent mixing models, a hybrid difference scheme, and a porous body pressure drop model, therefore, it is applicable to a flow path with a plate-type or a porous-type blockage. The effect of each model has been evaluated through an analysis for the THORS experiment, in which a plate-type blockage was located in a flow channel with wire-wrapped fuel rods. The overall capability of the code has also been evaluated for the KNS experiment with a plate-type blockage in a grid-spaced flow channel, and for the SCARLET-II experiments with a porous blockage in channels formed by wire-wrapped fuel rods. The code shows good predictions for the experiments with a wire-wrapped flow path with a plate-type or a porous type blockage. The analyses for the KNS experiments reveal that the code requires a precise blockage model related to a grid spacer model.     

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。     

Development of a thermal-hydraulic subchannel analysis code for motion conditions

Since the conventional subchannel analysis codes are designed for the land-based reactor core, a thermal-hydraulic subchannel analysis code was developed to evaluate thermal-hydraulic characteristics of the reactor core under motion conditions. The verification of the code was performed with experimental data and commercial codes. The ISPRA 16-rod tests were used to evaluate the steady-state prediction performance of the code, and the simulation results agree well with the test data. COBRA-EN code was applied to check the transient prediction performance of the code, and there is a good agreement between the predictions with both codes. An additional forces model for motion conditions was proposed in the code, and CFX-14.0 code was applied to verify the model. The results show that the code can be used in the thermal-hydraulic analysis of the reactor core under motion conditions. To illustrate the capabilities of the code, a fuel bundle under a complex motion condition was simulated, and the results are reasonable.     

Analysis of temperature noise propagation in a large fuel subassembly of FBR

We have developed a method to calculate the three-dimensional distribution of root-mean-square (RMS) values of temperature noise in the single phase flow in a fast reactor fuel subassembly with a local flow blockage. Employed are the subchannel method in a pin bundle region and the finite difference method in the region downstream of the bundle. We have compared the calculated RMS values of temperature noise with experimental data for a sodium loop test using a wire-spacered 91-pin-bundle fuel sub-assembly with a local blockage. We have investigated the possibility of detection of the blockage by temperature noise by taking into account the influence of structures in the upper part of the subassembly.