A study of non-Boussinesq integral boundary layer analysis for an MTR research reactor

In the present work, a non-Boussinesq (variable physical properties) integral boundary layer analysis is accomplished. The model analyzes laminar free convection between nuclear fuel plates having large fuel plate length to gap between plate ratio. The coolant channels are undergoing to a uniform, symmetric, heat flux and varying fluid properties. In the present study the flow is assumed to be fully developed. This is a good assumption for channels with large fuel plate length to gap between plate ratios. To describe the velocity and temperature distributions of the coolant the non-Boussinesq approximation is introduced into the integral boundary layer equations of flow between parallel plates. The fuel plate temperature is related to the adjacent coolant fluid temperature by a principle in conduction heat transfer. Fluids considered here are air and water. The obtained results show that the present heat transfer problem encountered in nuclear research reactor such Tehran nuclear research reactor (TRR) is characterized by high temperature ratios and thereby rendering the commonly applied Boussinesq approximation invalid. As a result, the use of the Boussinesq approximation (constant fluid properties) for high temperature ratios is not suggested.     

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.     

Laminar flow induced deflections of stacked plates

Some research and power reactors such as the Engineering Test Reactor (ETR), the Materials Test Reactor (MTR) and the Shippingport Reactor have core designs which consist of parallel, flat or curved plate fuel assemblies. The fuel is contained in the thin plates which are separated by narrow channels through which coolant flows to remove heat generated within the plates. Since the plates are flexible, the coolant flowing through the channels causes the plates to deflect. At high coolant velocities large deflections have been observed causing the plates to deform plastically leading to structural failure or plate collapse. This work examines a single plate bounded by two channels and determines the static plate deflection as a function of plate, channel and flow parameters. The deflection is due to differences in pressure and flow velocity in the channels bounding the plate and also due to different channel dimensions caused by tolerance effects. The classical thin plate equations are used with a nonlinear hydrodynamic loading function expressing the external fluid forces on the plate surfaces.     

Study on Two-phase Flow in Fluid Channel of Lead-based Fast Reactor Based on OpenFOAM

In order to study the thermal-hydraulic characteristics of two-phase flow caused by the special thermal properties of lead/lead-bismuth in lead-based fast reactors, the influence of bubble in fluid channel on the heat transfer capacity and safety of the core was simulated. In this paper the open source CFD calculation software OpenFOAM was adopted, and the numerical simulation was applied based on VOF method to construct a common triangular channel model in lead-based fast reactor. By simulating the two-phase flow of the coolant channel, it is found that as the flow rate increases, the outlet temperature of the coolant decreases. In the flow process of the gas-liquid two-phase flow in the channel, it can be found that the gas phase basically flows inside the channel. In the simulation of the fuel assembly, the corner channel is an area with a large amount of bubbles, which will cause the local heat transfer to deteriorate and cause the fuel assembly to burn out.     

基于OpenFOAM的铅基快堆流体通道内两相流研究

为研究铅基快堆中铅/铅铋的特殊热物性导致的在两相流情况下的热工水力特性,模拟流体通道中空泡存在对堆芯的输热能力以及安全性的影响,本文采用开源的CFD计算软件OpenFOAM,应用基于VOF方法的数值模拟,构建了铅基快堆中常见的三角形通道模型,通过与子通道程序的验证和单相条件下实验的校核,检验了所用代码的准确性,并对堆内冷却剂通道的两相流进行了模拟。模拟结果表明：随着两相流流速的增大,冷却剂出口温度降低。气液两相流在内通道流动过程中,气相基本在通道内部流动。随着轴向高度的升高,气泡会在内通道的中心区域聚合;燃料组件的角通道是气泡含量多的区域,会造成局部传热恶化,导致组件烧毁。     

流量分配比对环形燃料芯块传热特性影响数值模拟研究

环形燃料具有两条冷却通道，外通道与内通道的冷却水流量分配比（φ）的变化可能会对芯块传热特性产生影响。本文建立了环形燃料单棒流固耦合CFD计算模型，在4种不同的流量分配比工况下，通过计算3个反映芯块传热特性的评价指标，研究了流量分配比变化对环形燃料芯块传热特性的影响。由分析计算结果可知，流量分配比变化不会对有间隙结构的环形燃料的芯块传热特性产生显著影响。     

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。     

ATWS Characteristics of Super LWR with/without Alternative Action

Anticipated-transient-without-scram (ATWS) of the supercritical-pressure light water cooled thermal reactor with downward-flow water rods (Super LWR) is analyzed to clarify its safety characteristics. At loss-of-flow, heat-up of the fuel cladding is mitigated by the water rods removing heat from the fuel channels by heat conduction and supplying their coolant inventory to the fuel channels by volume expansion. The average coolant density is not sensitive to the pressure due to the small density difference between “steam” and “water” at supercritical-pressure. Closure of the coolant outlet of the once-through system causes flow stagnation that suppresses an increase in the coolant density due to an increase in the temperature. Therefore, the increase in power is small for pressurization events. The coolant density and Doppler feedbacks provide good self-controllability of the power against loss-of-flow and reactivity insertion. An alternative action is not needed either to satisfy the safety criteria or to achieve a high-temperature stable condition for all ATWS events. Initiating the automatic depressurization system is a good alternative action that induces a strong core coolant flow and inserts a negative reactivity. It provides an additional safety margin for the ATWS events. Even the high core power rating of the Super LWR has excellent ATWS characteristics, providing a key reactor design advantage.     

Investigation of thermohydraulic parameters during natural convection cooling of TRIGA reactor

Important steady-state thermohydraulic parameters of the TRIGA research reactor operating under natural convection mode of coolant flow were investigated using NCTRIGA computer code. Neutronic parameters used in preparing the input of NCTRIGA were taken from the analysis performed by 3-D Monte Carlo code MCNP4C. Benchmarking of the NCTRIGA calculated results were performed against the experimental data measured by the thermocouples in the instrumented fuel element (IFE) during the steady state operation of the reactor under natural convection mode of coolant flow. Various thermohydraulic parameters like the coolant velocity, flow rate and mass flow rate were generated for the hot channel as well as for the two channels comprising instrumented fuels. Calculated peak fuel temperatures at different power levels were compared with the measured values and also with the calculations performed by PARET code. Axial temperature profile at the fuel centreline, fuel surface and coolant in the hot channel were generated. Fuel surface heat flux, heat transfer coefficient and Reynolds’s number for the hot channel were also calculated. The effect of fuel-cladding gap and the influence of fuel rod spacing were investigated to validate the performance of NCTRIGA code. The investigated results were found to be in good agreement with the experimental values, which indicates that the NCTRIGA code can be used with confidence for TRIGA reactor analysis.     

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.     

Analysis of accidents leading to local flow stagnation in parallel fuel channels of RBMK-type reactors

The RBMK-type nuclear power reactors, still operating in Russia, are graphite-moderated with vertical fuel channels, using low-enriched nuclear fuel. The main challenge, which leads to the overheating of the fuel assemblies, fuel channels and other core components in channel type nuclear reactors, is a misbalance between heat generation in core structures and heat sink, which can appear due to the loss of coolant accident. In this accidental case, the emergency core cooling system ensures the core cooling. In RBMK-type reactors this system consists of hydro-accumulators and a number of pumps, taking water from large water reservoirs. This equipment injects water into the fuel channels through the group distribution headers at high pressure. However, the direct supply of cold water from emergency core cooling system into fuel channels is possible only if check valves on group distribution headers are closed properly. If these check valves failed, the part of water would be lost through the break, the flow stagnation in channels could occur, which might lead to overheating of fuel assemblies in the fuel channels. This paper presents the results of deterministic safety analysis, performed using system thermal hydraulic code RELAP5. Using this code the reactor cooling system of RBMK-1500 was modelled and analyses of loss of coolant accidents with failure of few check valves in group distribution headers were performed. The results of the calculations are used for the development of symptom-based emergency operating procedures for RBMK-1500. The basic principles that describe the complex distribution of water flows in vertical forced circulation circuit with parallel fuel channels can be adjusted for the RBMK-1000 reactors, still operating in Russia.     

Effects of Coolant Flow on Light Water Reactor Fuel Behaviors during Reactivity Initiated Accident

This paper describes the in-pile experimental results to study the influences of coolant flow on fuel behaviors under reactivity initiated accident (RIA) conditions performed in the Nuclear Safety Research Reactor (NSRR). A single PWR type test fuel rod was irradiated by a large neutron pulse in the NSRR to simulate a prompt power excursion of RIA's. The effects of coolant flow were studied at a coolant flow velocity of 0.3~1.8m/s and a coolant temperature of 20~90°C under the atmospheric pressure. It was found that the cooling conditions had considerable influences on fuel thermal behaviors under prompt heat-up. The increase of coolant flow velocity and subcooling enhanced heat transfer coefficient at cladding surface during film boiling, which resulted in large decrease of maximum cladding temperature and film boiling duration, and consequently in the increase of fuel failure threshold energy. The data tendencies are summarized and the influences of coolant flow are discussed with some computer analyses.     

Experimental and analytical dynamic analysis of coolant channel and its internals of Indian 540 MWe PHWR reactor

The horizontal coolant channel is one of the important parts of primary heat transport system in PHWR type of reactors. There are in all 392 channels in the core of Indian 540 MWe reactor. Each channel houses 13 natural uranium fuel bundles and shielding and sealing plugs one each on either side of the channel. The heavy water coolant flows through the coolant channel and carry the nuclear heat to outside the core for steam generation and power production in the turbo-generator. India has commissioned one 540 MWe PHWR reactor in September 2005 and another similar unit will be in operation very shortly. For a complete dynamic study of the channel and its internals under the influence of high coolant flow, experimental and analytical study has been carried out. The operating life of a typical coolant channel typically range from 10 to 15 full power years. Towards the end of its operating life, its health monitoring becomes an important activity. Vibration diagnosis plays an important role as a tool for life management of coolant. A good correlation has been achieved between the results of experimental and analytical models. Through the study dynamic characteristics of the coolant channel under simulated loading condition, an attempt has been made to develop a diagnostics to monitor the health of the coolant channel over its operating life. Experimental study has been also carried out to characterize PHWR fuel vibration under different flow conditions. Such results are published for the first time.     

Three-dimensional analysis of flow blockage accident in Tehran MTR research reactor core using CFD

In the present study, a 3D simulation of flow blockage accident which may occur in the coolant channels of a fuel assembly of Tehran research reactor (TRR) is investigated using CFD code. Consideration is given to the scenario in which partial blockage of hot channel occurs due to buckling of its fuel plates towards the inside. Governing conservation laws are solved using Control volume approach and pressure field is coupled to the velocity field through the SIMPLE algorithm. Flow convergence is considered when the residual for all flow variables are less than 10⁻⁵. The simulation is performed under four different obstruction levels of the nominal flow area, i.e., 0%, 20%, 50% and 70%. By solving momentum and energy equation in three channels with their fuel plates, it is found that heat transfer is substantially affected by channels flow field. In the blockage accident, decrease in flow rate of the obstructed channel decreases cooling capacity of the obstructed channel as a result of hydraulic resistance augmentation. The obtained results show that above the 50% blockage, critical phenomena will appear which may compromise the clad integrity. Moreover, in the 70% blockage scenario, the clad temperature in the obstructed channel reaches the value associated with nucleate boiling temperature at the operative pressure.     

Flow-induced plastic collapse of stacked fuel plates

Flow-induced plastic collapse of stacked fuel plate assemblies was first noted in experimental nuclear reactors such as the Oak Ridge National Laboratory High Flux Reactor Assembly and the Engineering Test Reactor (ETR). The ETR assembly is a stack of 19 thin flat rectangular fuel plates separated by narrow channels through which a coolant flows to remove the heat generated by the nuclear fission of the fuel within the plates. The uranium alloyed plates have been noted to buckle laterally and plastically collapse at the system design coolant flow rate of 10.7 m/s, thus restricting the coolant flow through adjacent channels. In this paper a methodology and criterion are developed for predicting the plastic collapse of ETR fuel plates. The criterion is compared to some experimental results and the Miller critical velocity theory.     

Development of a thermal–hydraulic analysis code for research reactors with plate fuels

The thermal–hydraulic analysis code THAC-PRR has been developed with Visual Fortran 6.5 for the investigation of plate type fuel reactors. It is based on the fundamental conservation of mass, momentum and energy, and proper constitutive correlations for flow friction factor, heat transfer and thermophysical properties. Moreover, a simple and improved lumped-differential method has been adopted to analyze the conjugate heat transfer between the fuel plate and the coolant. The Reactivity Insertion Accident (RIA) and Loss Of Flow Accident (LOFA), which have been defined in the IAEA 10 MW MTR Benchmark program, were analyzed with this developed program for the code-to-code validation. Good agreement was achieved. Furthermore, the accidents due to the partial (95%) and total (100%) blockage of one channel in the IAEA 10 MW MTR were investigated with THAC-PRR. The results showed that if the blockage occurred in the average channel, there was no boiling occurred even the channel was totally obstructed. The reason was that the heat was transferred to the adjacent channels by conduction through the fuel plates which formed the obstructed channel. However, if the blockage occurred in the hot channel, boiling did occur. This indicated that it is very important to consider the interaction between the blocked channel and the adjacent channels in this type of transient.     

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.     

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.