CFD investigations of helical wire-wrap fuel pin bundle and its comparison with straight wire bundle

Computational Fluid Dynamics (CFD) investigations of a fast reactor fuel pin bundle wrapped with helical and straight spacer wires have been carried out and the advantages of using helical spacer wire have been assessed. The flow and temperature distributions in the fuel pin bundle are obtained by solving the statistically averaged 3-Dimensional conservation equations of mass, momentum and energy along with high Reynolds number k-ε turbulence model using a customized CFD code CFDEXPERT. It is seen that due to the helical wire-wrap spacer, the coolant sodium not only flows in axial direction in the fuel pin bundle but also in a transverse direction. This transverse flow enhances mixing of coolant among the sub channels and due to this, the friction factor and heat transfer coefficient of the coolant increase. Estimation of friction factor, Nusselt number, sodium temperature uniformity at the outlet of the bundle and clad hot spot factor which are measures of the extent of coolant mixing and non-homogeneity in heat transfer coefficient around fuel pin are paid critical attention. It is seen that the friction factor and Nusselt number are higher (by 25% and 15% respectively) for the helical wire wrap pin bundle compared to straight wire bundle. It is seen that for 217 fuel pin bundle the maximum clad temperature is 750 K for straight wire case and the same for helical wire is 720 K due to the presence of transverse flow. The maximum temperature occurs at the location of the gap between pin and wire. The ΔT between the bulk sodium in the central sub-channel and peripheral sub-channel is 30 K for straight wire and the same for helical wire is 18 K due to the presence of secondary transverse flow which makes the outlet temperature more uniform. The hotspot factor and the hot channel factors predicted by CFD simulation are 10% lower than that used in conventional safety analysis indicating the conservatism in the safety analysis.     

Reactivity insertion limits in a typical pool-type research reactor cooled by natural circulation

The course of reactivity insertion in a pool type research reactor, with scram disabled under natural circulation condition is numerically investigated. The analyses were performed by a coupled kinetic–thermal–hydraulic computer code developed specifically for this task. The 10-MW IAEA MTR research reactor was subjected to unprotected reactivity insertion (step and ramp) for both low and high-enriched fuel with continuous reactivity feedback due to coolant and fuel temperature effects. In general, it was found that the power, core mass flow rate and clad temperature under fully established natural circulation are higher for high-enriched fuel than for low enriched fuel. This is unlike the case of decay heat removal, where equal clad temperatures are reported for both fuels. The analysis of reactivity represented by the maximum insertion of positive reactivity ($0.73) demonstrated the high inherent safety features of MTR-type research reactor. Even in the case of total excess reactivity without scram, the high reactivity feedbacks of fuel and moderator temperatures limit the power excursion and avoid consequently escalation of clad temperature to the level of onset of nucleate boiling and sub-cooled void formation. The code can also be modified to provide an accurate capability for the analyses of research reactor transients under forced convection.     

Calculating decay heat removal rates via natural circulation along variable heat flux reactor fuel plates

A recently developed integral technique is applied to natural convection cooling along test reactor fuel plates. The technique is demonstrated for water and air flow. In the case of air flow, the process is characterized by a large temperature rise along the fuel channel, thereby rendering the commonly applied Boussinesq approximation invalid. This case is a heat transfer problem of particular interest in accident analyses such as determining the level of decay heat dissipation possible, without exceeding the melting temperature of the fuel, subsequent to a hypothetical loss of primary coolant.     

Numerical Simulation of Fluid Flow and Heat Transfer of Supercritical Fluids in Fuel Bundles

A supercritical water-cooled reactor (SCWR) was proposed as a kind of generation IV reactor in order to improve the efficiency of nuclear reactors. Although investigations on the thermal-hydraulic behavior in SCWR have attracted much attention, there is still a lack of CFD study on the heat transfer of supercritical water in fuel channels. In order to understand the thermal-hydraulic behavior of supercritical fluids in nuclear reactors, the local fluid flow and heat transfer of supercritical water in a 37-element fuel bundle has been studied numerically in this work. Results show that secondary flow appears and the cladding surface temperature (CST) is very nonuniform in the fuel bundle. The maximum cladding surface temperature (MaxCST), which is an important design parameter for SCWR, can be predicted and analyzed using the CFD method. Due to a very large circumferential temperature gradient in cladding surfaces of the fuel bundle, the precise cladding temperature distributions using the CFD method is highly recommended.     

CFD investigation of flow inversion in typical MTR research reactor undergoing thermal–hydraulic transients

Three dimensional CFD full simulations of the fast loss of flow accident (FLOFA) of the IAEA 10 MW generic MTR research reactor are conducted. In this system the flow is initially downward. The transient scenario starts when the pump coasts down exponentially with a time constant of 1 s. As a result the temperatures of the heating element, the clad, and the coolant rise. When the flow reaches 85% of its nominal value the control rod system scrams and the power drops sharply resulting in the temperatures of the different components to drop. As the coolant flow continues to drop, the decay heat causes the temperatures to increase at a slower rate in the beginning. When the flow becomes laminar, the rate of temperature increase becomes larger and when the pumps completely stop a flow inversion occurs because of natural convection. The temperature will continue to rise at even higher rates until natural convection is established, that is when the temperatures settle off. The interesting 3D patterns of the flow during the inversion process are shown and investigated. The temperature history is also reported and is compared with those estimated by one-dimensional codes. Generally, very good agreement is achieved which provides confidence in the modeling approach.     

QM-400 CFD自然对流模型研究及验证

秦山第三核电厂乏燃料干式贮存模块QM-400是我国第一座投入商业运行的干式贮存设施，模块内的热量交换主要包括自然对流、热传导、耦合传热和辐射换热等。本文精确计算了典型环境温度下每个燃料篮的衰变热，运用商用计算流体动力学（CFD）软件FLUENT 14.0开展了网格敏感性分析，并建立了QM-400存储模块的自然对流CFD分析模型。结果表明，模块顶面、侧面以及贮存筒表面压力和温度分布符合自然对流规律，计算的测点温度与现场的实测温度符合良好，测点温度随环境温度的变化趋势也与实测趋势符合良好，证明了建立的CFD自然对流计算方法的正确性。本文结果为后续采用CFD方法进行取消绝热板后的温度场计算奠定了基础。     

Heat transfer from a partially uncovered reactor core

This paper documents a model which has been developed for predicting the temperature distribution along a “flow channel” of a pressurized water reactor during simulated, uncovered core conditions. In the model, heat conduction along the fuel element, convection from the surface to the coolant, radiation exchange between the clad surface and steam, and surface exchange between adjacent fuel rods are considered. Variations of the thermophysical properties of the fuel road and of the coolant with temperature are accounted for, but oxidation of Zircaloy is not modeled. Extensive sensitivity studies on the effects of heat generation in the core, steam velocity, pressure level, uncovered core height, presence of hydrogen gas in the coolant, power skew, clad emissivity, and convective heat transfer correlations have been examined. The results show that the importance of radiation in comparison with convection increases with an increase in the fuel rod temperature, pressure, and clad emissivity.     

CFD analysis of fast loss of flow accident in typical MTR reactor undergoing partial and full blockage: The average channel scenario

CFD investigation of loss of flow accident (LOFA) in typical MTR reactor undergoing partial and full blockage under the average channel condition is considered. The blockage scenarios considered in this work describe changes in the geometrical configuration of the flow channels as a result of thermal stresses or any other reason. That is the fuel plates of the average channel are assumed to buckle inwards along the plate height. As a result, the flow area decreases along the height of the channel until it achieves minimum in the middle. Three adjacent channels are simulated. With the area of the blocked channel decreases, that of the adjacent channel increases while the third channel remains unaltered. Blockage ratios considered in this work includes 0%, 20%, 40%, 50%, 60%, 80%, and full blockage. As a result of the change in the geometrical configuration of the flow channels, the hydraulic resistance also changes resulting in flow and heat transfer load to redistribute among the three channels. During the course of LOFA, the decay heat load is taken up by natural convection. While under the hot channel conditions, previous work showed that boiling is inevitable for even small blockage ratios. In this work maximum clad temperature is found to be under the boiling temperature at the operating pressure up to approximately 80% blockage ratio. For blockage ratio larger than 80%, the maximum clad temperature exceeds the boiling temperature indicating that boiling may occur.     

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。     

CFD simulation of the IAEA 10 MW generic MTR reactor under loss of flow transient

Three-dimensional simulation of the IAEA 10 MW generic reactor under loss of flow transient is introduced using the CFD code, Fluent. The IAEA reactor calculation is a safety-related benchmark problem for an idealized material testing reactor (MTR) pool type specified in order to compare calculational methods used in various research centers. The flow transients considered include fast loss of flow accidents (FLOFA) and slow loss of flow accidents (SLOFA) modeled with exponential flow decay and time constants of 1 and 25 s, respectively. The transients were initiated from a power of 12 MW with a flow trip point at 85% nominal flow and a 200 ms time delay. The simulation shows comparable results as those published by other research groups. However, interesting 3D patterns are shown that are usually lost based on the one-dimensional simulations that other research groups have introduced. In addition, information about the maximum clad surface temperature, the maximum fuel element temperature as well as the location of hot spots in fuel channel is also reported.     

Large scale experiments with a 5 MW sodium/air heat exchanger for decay heat removal

Sodium experiments in the large scale test facility ILONA were performed to demonstrate proper operation of a passive decay heat removal system for LMFBRs based on pure natural convection flow. Temperature and flow distributions on the sodium and the air side of a 5 MW sodium/air heat exchanger in a natural draught stack were measured during steady state and transient operation in good agreement with calculations using a two dimensional computer code ATTICA/DIANA.     

The inherently-safe power reactor DYONISOS : Dynamic nuclear inherently-safe reactor operating with spheres

A philosophy of inherent safety is formulated and an inherently-safe thermal power reactor is presented. Solid fuel in the form of spheres a few centimetres in diameter is suspended under the hydrodynamic pressure of molten lead coolant in vertical channels within the graphite moderator. Loss of main pump pressure, or a loss-of-coolant accident (LOCA), results in immediate removal of the fuel to rigid sieves below the core, with consequent subcriticality. Residual and decay heat are carried away by thermal conduction through the coolant or, in the case of a LOCA, by a combination of radiation and natural convection of cover gas or incoming air from the fuel to the reactor vessel and convection of air between the vessel and steel containment wall. All decay heat removal systems are passive, though actively initiated external spray cooling of the containment can be used to reduce wall temperature. This, however, is only necessary in the case of a LOCA and after a period of 24 h.     

Thermal hydraulic and safety analysis for core conversion (HEU–LEU) of Syrian Miniature Neutron Source Reactor

The paper presents the behavior and properties analysis of the low enriched uranium fuel compared with the original high enriched uranium fuel. The MNSR reactor core was modeled with both fuel materials and the reactor behavior was studied during the steady state and abnormal conditions. The MERSAT code was used in the analysis. The steady state thermal hydraulic analysis results were compared with that obtained from the experimental results hold during commissioning the Syrian MNSR. Comparison with experimental data shows that the steady-state behavior of the HEU core was accurately predicted by the MERSAT code calculations. The validated model was then used to analyze LEU cores with two proposed UO₂ fuel pin designs. With each LEU core, the steady state and 3.77 mk rod withdrawal transient were run and the results were compared with the available published data in the literatures for the low enriched uranium fuel core. The results reveal that the low enriched uranium fuel showed a good behavior and the peak clad temperatures remain well below the clad melting temperature during reactivity insertion accident.     

Parametric study on rewetting velocities obtained with a two-dimensional heat conduction code

A comprehensive parametric study has been performed to quantify the effect of different variables on the rewetting velocity in a light water reactor following a loss-of-coolant accident. To this purpose, a numerical solution of the general two-dimensional (axial and radial) heat conduction equation in cylindrical geometry has been obtained. The method used is the alternating-direction implicit procedure developed by Peaceman and Rachford. The model accounts for decay heat generation in the fuel, coolant subcooling, different wall temperatures and different heat transfer coefficients across the gap and at the clad surface. The two-dimensional model can be reduced to a one-dimensional model by setting the heat conduction in either the radial or axial direction to zero. Results with the new model agree with previous models and with experimental data.The variables studied were: axial and/or radial heat conduction, clad temperature, quench temperature, coolant temperature, temperature for the onset of nucleate boiling, heat transfer coefficients, stored and decay heats, clad material and clad thickness. The critical thickness (clad thickness for which the calculated rewetting velocity remains constant) was also determined and found to be larger than the clad thickness of light water reactor fuel pins under usual reflood conditions. According to these calculations, the stored and decay heats affect the rewetting velocity significantly.     

快堆组件数值模拟研究

为详细研究快堆组件棒束中的流动换热特性,本工作采用Fluent程序对169棒束快堆燃料组件进行三维数值模拟。结果表明,在流量为10.92~18.67 kg/s时,计算得到的压降与已公开发表文献结果的相对偏差小于3.41%。内子通道的相对温度升高,呈现出周期为1/3螺距的波动,内子通道的局部温度比子通道程序SUPERENERGY计算的结果更高。根据模拟计算结果可更为准确地预测棒束通道内的流动换热情况,为今后组件棒束热工水力学设计提供参考。     

脉动流下棒束通道内相位差及瞬态流场研究

事故条件及海洋条件下反应堆处于非稳态工况,堆芯燃料组件内热工水力行为具有瞬变及多因素耦合特性,对反应堆的安全提出更高挑战,因此有必要对燃料组件内瞬态特性进行研究。本文通过测量棒状燃料组件内压降和流量之间延迟时间开展棒束通道脉动流条件下相位差研究,对比了相位差在不同振幅、不同流动状态下的变化特性,并分析了定位格架对脉动流相位差的作用特点。另外,基于粒子图像测速（PIV）技术开展了脉动流条件下棒束通道内流场分布特性研究,对比了相同流量条件下稳态工况与瞬态工况下流场分布差异,分析了主流具备不同加速度时棒束通道内流场分布特征。实验结果表明：定位格架可减小脉动流下棒束通道内相位差;棒束通道内流场演化滞后于主流量变化。实验结果有助于揭示燃料组件在非稳态条件下瞬态特性,并为燃料组件的设计和优化奠定基础。     

TACR热传输系统热工水力瞬态分析

钍燃料的利用对于缓解核燃料资源短缺具有重要意义,坎杜型反应堆(Canadian Deuterium Uranium,CANDU)在堆芯布置、中子利用效率及先进燃料循环方面具有较高的灵活性,使得其在CANDU反应堆中引入钍燃料循环更具现实意义。CANDU型反应堆中钍基燃料应用关键基础技术研究是加拿大与我国正在开展的合作课题,其中开发自主的CANDU堆堆芯热工水力设计和安全分析程序是钍基燃料应用必不可少的设计工作之一。本文针对CANDU型反应堆热传输系统结构特点,采用FORTRAN程序设计语言开发了适用于CANDU型反应堆热传输系统的热工水力瞬态分析程序CANTHAC(CANDU Thermal-Hydraulic Analysis Code)。利用CANTHAC对钍基先进CANDU堆(Thorium-based Advanced CANDU Reactor,TACR)进行了瞬态分析,计算工况包括满功率稳态、无保护蒸汽发生器(Steam Generator,SG)二次侧给水温度降低事故及完全失流事故。其中,满功率稳态计算结果与清华大学设计的钍基先进CANDU堆TACR设计值吻合较好,相对误差不超过2%,在可接受范围内;无保护SG二次侧给水温度降低事故及完全失流事故在计算条件下所得的燃料温度及系统压力等关键热工水力参数均在安全限值内,满足安全准则要求。程序为模块化编程,便于移植和改进,具有一定的通用性,为进一步研究工作奠定了基础。     

Improved Solution of the Neutron Diffusion Equation Using Wavelet Theory

The development of the fast reactor (FR) cycle is being advanced to utilize plutonium and transuranium (TRU) in Japan. In the fabrication process, it is considered that a fuel pin spirally wrapped with a thin wire is laid horizontally. Then, cooling air flows vertically from the bottom side into the gap of the pin bundle so as to suppress the temperature increase due to decay heat. From the viewpoint of safety assessment during the fabrication, a thermal hydraulic analysis method plays an important role in investigating the maximum temperature and the temperature distribution of the fuel pins. In the present paper, a subchannel analysis tool has been developed. Using the developed tool, the benchmark analysis of the mocked up experiment has been carried out, as well as the numerical investigation of a multidimensional effect of fuel cladding thermal conductivity on the maximum temperature. It is demonstrated that the multidimensional effect of the cladding thermal conductivity is not negligible in the analysis. A good agreement is achieved in the case of a comparatively large clearance size between the side wall and the pin bundle when one considers a natural convection heat transfer at the outermost boundary with a comparatively low computational cost.