Dynamic simulation of accidental closure of intermediate heat exchanger isolation valve in a pool type LMFBR

In a pool type liquid metal cooled fast breeder reactor (LMFBR), core and other internals such as pumps, heat exchangers are immersed in a pool of sodium. Heat exchange from primary sodium circuit (pool) to secondary sodium circuit (loop) is through four intermediate heat exchangers (IHX) immersed in primary sodium pool. Each IHX is provided with a sleeve valve at its primary sodium inlet window for the purpose of isolating the shell side of IHX from the sodium pool. With such a provision, an inadvertent partial or complete closure of a sleeve valve of one of the IHX during normal operation of the reactor has been considered as a design basis event for the reactor. One dimensional transient thermal hydraulic models of the primary and secondary sodium circuits have been developed to study the thermal hydraulic consequences of such an event. The main areas of concern in the plant and the availability of safety parameters for the detection of the event have been evaluated.     

A thermal hydraulic calculation method of an intermediate heat exchanger of a loop type FBR

This paper describes a thermal-hydraulic calculation of an intermediate heat exchanger (IHX) with the computational fluid dynamics (CFD) code CFX. The motivation of this paper is to clarify a heat transfer degradation phenomenon in the IHX through three-dimensional calculation. The whole IHX of the “Monju” reactor is modeled with three parts, i.e., the primary side, the secondary side and the heat transfer region. Through a partial calculation using these models, the flows on the primary side and the secondary side are shown to be axisymmetric. Therefore, a sector model is adopted for the calculation model in the heat transfer region. The calculated temperatures in the IHX are compared with the measured results using the IHX in the “Monju” reactor. Good agreement is obtained for the predicted outlet temperatures and temperatures on the shell surface. As a result of the CFD calculation, it is evaluated that a heat transfer in the lower plenum on the secondary side is dominant under the low flow rate conditions. This fact contributes to lower the heat transfer coefficient in the IHX when the standard heat exchanger theory is applied to calculate the overall heat transfer coefficient between the primary and the secondary sides.     

SADHANA facility for simulation of natural convection in the SGDHR system of PFBR

The Prototype Fast Breeder Reactor (PFBR) is a 500 MWe sodium cooled pool type fast reactor being constructed at Kalpakkam, India. PFBR has all the reactor components immersed in the pool of sodium and the fission heat generated in the core, is removed by the sodium circulating in the pool. During normal operation this fission heat is transferred by primary sodium to secondary sodium, which in turn transfers the heat to water in the steam generator for producing steam. The removal of the decay heat generated in the reactor core after the reactor shutdown is also very important to maintain the structural integrity of reactor core components. PFBR employs two independent systems namely, Operational Grade Decay Heat Removal system (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS) for decay heat removal. SGDHR system is a passive system working on natural convection to ensure the core coolability even under station blackout condition. It is very important to study the thermal hydraulic behavior of Safety Grade Decay Heat Removal system of PFBR to ensure its reliable operation. A scaled down model of the circuit, named SADHANA has been modeled, designed, constructed and commissioned for demonstration and evaluation of these systems. The facility has completed around 2000 h of high temperature operation. The performance of the experimental system is satisfactory and it meets all the design requirements. At 550 °C sodium pool temperature in test vessel the secondary sodium loop generated a sodium flow of 6.7 m³/h. These experiments have revealed the adequacy and capability of SGDHR system to remove the decay heat from the fast breeder reactor core after its shutdown.     

Experimental investigation of weir instability in main vessel cooling system of 1/4 FBR model

The 500 MWe Prototype Fast Breeder Reactor (PFBR) is under construction at Kalpakkam, India. The main vessel of this pool type reactor acts as the primary containment in the reactor assembly. In order to keep the main vessel temperature below creep range and to reduce high temperature embrittlement and also to ensure its healthiness for 40 years of reactor life, a small fraction of core flow (0.5 m³/s) is sent through an annular space formed between the main vessel and a cylindrical baffle (primary thermal baffle) to cool the vessel. The sodium after cooling the main vessel overflows the primary baffle (weir shell) and falls into another concentric pool of sodium separated from the cold pool by the secondary thermal baffle and then returned to cold pool. The weir shell, where the overflow of liquid sodium takes place, is a thin shell prone to flow induced vibrations due to instability caused by sloshing and fluid-structure interaction. A similar vibration phenomenon was first observed during the commissioning of Super-Phenix reactor. In order to understand the phenomenon and provide necessary experimental back up to validate the analytical models, weir instability experiments were conducted in a 1:4 scale stainless steel (SS) model installed in a water loop. The experiments were conducted with flow rate and fall height as the varying parameters. The experimental results showed that the instability of the weir shell was caused due to fluid structure interaction. This paper discusses the details of the model, the modeling laws, similitude criteria adopted, analytical prediction, the experimental results and conclusion.     

Thermal hydraulic investigations of intermediate heat exchanger in a pool-type fast breeder reactor

Intermediate heat exchanger (IHX) in a pool-type liquid metal cooled fast breeder reactor is an important heat exchanging component as it forms an intermediate boundary between the radioactive primary sodium in the pool and the non-radioactive secondary sodium in the steam generator (SG). The thermal loads during steady state and transient conditions impose thermal stresses on the heat exchanger tubes and on the shells which hold the tube bundle. Estimation of these thermal loads and achieving uniform temperature distribution in the tubes and shells by having uniform flow distributions are the major tasks of thermal hydraulic investigations of IHX. Through multi-dimensional thermal hydraulic investigations performed using commercially available computer codes such as PHOENICS, the flow and temperature distributions in the tubes and shells and in its secondary sodium inlet and outlet headers are obtained with and with out provisions of flow distribution devices. The effectiveness of these devices in achieving acceptably uniform flow and temperature distributions has been assessed and thermal loads on the tubes and shells for thermo mechanical analysis of the IHX have been defined. The predictions of the computational studies have been validated against simulated experiments.     

Experimental qualification of subassembly design for Prototype Fast Breeder Reactor

The Prototype Fast Breeder Reactor (PFBR) which is under construction at Kalpakkam, India, is a 500 MWe sodium cooled pool type reactor. The core of the PFBR consists of 1758 free standing subassemblies supported on the grid plate. The entire core is divided into 15 different flow zones and the flow rate required through each zone is calculated based on the fission heat generation. The coolant sodium flows from the bottom of the subassembly to top and the design of the subassembly for each flow zone is quite complex. There are 181 fuel subassemblies in PFBR core with 217 fuel pins in each subassembly, vertically held in the form of bundle within a hexagonal wrapper tube. The pins are separated by spacer wires wound around the pins helically. Analytical prediction of subassembly pressure drop, vibration and determination of inception of cavitation for this complex geometry is very difficult. So experiments were conducted extensively to get a more accurate evaluation of the design and for its qualification for the use in PFBR, which is designed for 40 years of operation.Pressure drop and cavitation experiments were carried out in water on full scale (1:1) subassemblies of all flow zones. The overall pressure drop of the subassembly determines the ratings of the pump. Cavitation of the pressure drop devices lead to erosion damage of fuelpins and may also result in reactivity fluctuation due to sodium-void effect. So it is essential to confirm that the subassembly is not cavitating in the operating regime of the reactor. Subassembly can vibrate in cantilever mode due to the turbulence in the flow and can result in reactivity fluctuation, reactor control problem and can even lead to the failure of the fuel pins. So vibration measurements were carried out in water on the maximum rated subassembly. This paper discusses various experiments carried out on PFBR subassembly, the similarity criteria followed, instrumentation, results and conclusion.     

A computational study of flow mal-distribution on the thermal hydraulic performance of an intermediate heat exchanger in LMFBR

The flow and thermal non-uniformities occurring in the intermediate heat exchanger (IHX) of a liquid metal-cooled fast breeder reactor have been characterized through numerical simulations. For modeling the primary and secondary sodium flow through the IHX, an equivalent anisotropic porous medium approach has been used. The pressure drop in the equivalent porous medium is accounted through the inclusion of additional pressure drop terms in the Navier–Stokes equations, with the help of standard correlations for cross flow or parallel flow over tubes. For secondary sodium flow, the effects of a flow distributor device with orifices and baffles at the inlet have also been included, in addition to axial flow through the tubes. The heat exchange between primary and secondary streams is incorporated in the form of a volumetric heat source or sink term, which is corrected iteratively. The resulting flow distributions are in reasonable agreement with available experimental results. The study shows that the temperature of the secondary sodium flow at the exit can be made more uniform by exchanging less heat near the inner wall of IHX, as compared to the region close to the outer wall, using suitable flow distribution devices.     

Thermal Performance Tests on a Sodium-to-Sodium Heat Exchanger

Thermal performance of a 3MW sodium-to-sodium intermediate heat exchanger (IHX) was evaluated under temperature conditions typical of a Fast Breeder Reactor IHX. A regenerative figure of eight loop was used with the heat exchanger at the cross over point, and a 500 kW heat source and an air cooled sink to maintain the desired test conditions. The overall heat transfer coefficient was found to vary from 4.02 to 4.87 kW/m²·K for Peclet numbers varying from 37 to 112.5 on the shell side and 44.4 to 133.5 on the tube side respectively. The Peclet numbers were representative of low turbulent regime in this case. While the overall heat transfer coefficient was found close to predictions using Lubarsky's correlation, it was somewhat lower than that predicted by later correlations of Spukunsky & Borishansky. The reasons for the lower overall heat transfer coefficient have been explained in terms of possible maldistribution of shell side flow in low turbulent regime reducing the effective heat transfer area and increased thermal contact resistance. Based on their findings the authors feel that heat transfer in a sodium-to-sodium heat exchanger at low Peclet numbers is expected to differ from that obtained with large Peclet numbers.     

Hydraulic design of sparger arrangement for long guide tubes of a fast breeder reactor

In liquid metal cooled fast reactors, the core is submerged in sodium pool by ∼5 m below sodium free surface. This necessitates the control and shutdown of reactor be achieved by long overhanging mechanisms housed inside a control plug. These mechanisms are protected by porous guide tubes with a sparger type arrangement for the sodium flow through them. Comprehensive knowledge of flow distribution of sodium through these guide tubes is essential to assess the risks of flow induced vibration of thin thermowell tubes that pass close to these shroud tubes and entrainment of cover gas due to high free surface velocities. Three dimensional hydraulic analysis of single isolated shroud tube and integrated assembly of shroud tubes have been carried out using CFD tools to acquire this knowledge. The predictions of the CFD models have been validated against experimental predictions. These studies have provided important information regarding critical design parameters. Size of holes in the shroud tube, location of holes in the control plug shell and arrangement for breaking sodium jets emanating from shroud tubes have been optimized to reduce free surface velocity.     

Thermal hydraulic study on a high-temperature gas–gas heat exchanger with helically coiled tube bundles

An experimental study was carried out to investigate flow-induced vibration, heat transfer and pressure drop of helically coiled tubes of an intermediate heat exchanger (IHX) for the HTTR, using a full-size partial model and air as the fluid. The test model has 54 helically coiled tubes separated into three layer bundles, surrounding the center pipe. The vibration of the tube bundles was mainly at the center pipe, and the individual vibrations of the tube bundles were not significant under the operation conditions of the IHX. The heat transfer of the tube outside, due to forced convection, was obtained as a function of Re^0.51Pr^0.3, and the friction factor, depending on the tube arrangement, was proportional to Re^−0.14.     

基于固有频率分析的中间热交换器管束支撑设计

在进行示范快堆电站的中间热交换器(IHX)设计时,需考虑流致振动对管束的影响。在计算IHX换热管的固有频率时,空间弯管结构形式是一个技术难点。通过有限元方法,对中国实验快堆(CEFR)的IHX换热管固有频率进行了计算。通过对比分析,得到了合理的计算模型及其适用范围。通过敏感性分析,提出管束支撑方案的设计原则。     

压水堆蒸汽发生器一、二次侧稳态流场耦合分析

蒸汽发生器（SG）在运行过程中主要面临流致振动所导致的传热管破裂事故,而流致振动分析需以SG内的三维两相流场作为输入条件。采用多孔介质模型,对SG二次侧流场进行求解,同时耦合一、二次侧换热,获得SG二次侧速度场、温度场、压力场及流动含气率分布,并获得传热管一维的一、二次侧流体温度和换热系数及传热管温度分布。由于一次侧向二次侧释热极不均匀,SG内流场分布及汽水分离器内的含气率分布极不均匀;汽水分离器内的最大、最小含气率分别为0.62和0.05,该参数可为汽水分离器负载设计提供依据。通过计算还获得弯管区速度分布,该分布可为传热管的流致振动磨损评估提供输入条件。     

螺旋管直流蒸汽发生器一、二次侧耦合传热特性分析

为获得螺旋管直流蒸汽发生器（HCOTSG）螺旋换热管内两相流动换热特征，以国际革新安全反应堆（IRIS）HCOTSG为研究对象建立了HCOTSG一、二次侧耦合热分析模型，分析了稳态工况下，不同二次侧给水流量对HCOTSG热工水力参数产生的影响，并将所建立的HCOTSG一、二次侧耦合热分析模型与计算流体力学软件（CFX）三维流动换热计算相结合，对HCOTSG稳态工况下螺旋管内精细的热工水力参数进行计算。通过HCOTSG一、二次侧耦合热分析模型计算得到HCOTSG稳态工作时沿管程的相关热工水力参数；通过CFX三维模拟发现螺旋管横截面流体流速和温度分布不均匀现象，得到螺旋内侧流体温度高于螺旋外侧，螺旋内侧流体速度低于螺旋外侧，螺旋内侧流体比螺旋外侧流体先开始沸腾的结论。因此，本研究对于HCOTSG稳态运行和螺旋换热管事故分析具有指导作用。      

A feasibility study of ferro-boron as in-core shield material in fast breeder reactors

Shields around core and blankets form major part of reactor assembly in fast reactors as the incident neutron spectrum is hard with negligible thermal component and has anisotropic angular distribution. In this paper, a study is presented on the use of ferro-boron as neutron shield material in pool type fast reactors. The reference case chosen is the Prototype Fast Breeder Reactor (PFBR), a 500 MWe which is sodium cooled, pool type, mixed oxide (MOX) fuelled reactor, which is under construction at Kalpakkam, India. It is shown through 2D transport calculations, carried out using 175 neutron multigroup cross-sections, that this low cost material as shield is capable of satisfying the radiological safety criteria as well as the shields in the reference case. The secondary sodium activity and dose in steam generator building are marginally lower than the reference case. The total shield material weight will be lower by about 50 tonnes and the material cost lower by a factor 5 as compared to PFBR shields comprising of stainless steel and B₄C.     

螺旋管直流蒸汽发生器一、二次侧耦合传热特性分析

To study the flow and heat transfer characteristics of the primary and secondary sides of the helical coil once-through tube steam generator (HCOTSG) under steady state conditions, taking HCOTSG of International Reactor Innovative and Secure (IRIS) as the research object, a primary and secondary sides heat balance calculation model for steady state operation of HCOTSG is established. The influence of different secondary side feed water flow rate on HCOTSG thermal and hydraulic parameters under steady-state condition is analyzed, and the detailed thermal and hydraulic parameters in the helical tube under steady-state condition are calculated by combining the coupled thermal analysis model with the three-dimensional flow and heat exchange calculation of CFX. The relevant thermal and hydraulic parameters along the tube side of HCOTSG during steady-state operation are calculated by the thermal analysis model. The CFX simulation results show that the velocity and temperature distribution of the fluid in the cross section of the helical tube are not uniform. The temperature of the fluid inside the helix is higher than that outside the helix. The velocity of the fluid inside the helix is lower than that outside the helix. The boiling of the fluid inside the helix occurs earlier than that outside the helix. Therefore, this study has a guiding role in the accident analysis for HCOTSG steady-state operation and spiral heat exchange tube.     

螺旋管式直流蒸汽发生器壳侧流场数值模拟方法研究

铅冷快堆（LFR）采用一体化堆芯设计方案,其中的直流蒸汽发生器（OTSG）多采用螺旋管式结构以使整体结构小型紧凑。为研究LFR中螺旋管式OTSG壳侧铅铋冷却剂的流动传热特性,利用FLUENT软件,采用一种分区段计算方法,通过管壁热流密度拟合公式对螺旋管式OTSG壳侧进行了三维数值模拟。最终验证了该分段计算方法的正确性,分析了OTSG壳侧铅铋冷却剂的流动传热特性,获得了其速度、温度以及压力场的计算数据,为下一步OTSG流致振动分析和高温应力计算提供了依据。      

基于多孔介质模型的快堆蒸汽发生器热工水力特性数值研究

蒸汽发生器（SG）作为钠冷快堆一次侧钠与二次侧水的热交换器，其可靠程度直接影响反应堆能否安全运行，因此对SG的一次侧热工水力特性的研究具有重要意义。本研究采用多孔介质模型，对快堆蒸汽发生器一次侧流场进行分析。通过对支撑板模型的计算，获得多孔介质控制方程的阻力源项。一次侧向二次侧的释热量通过系统程序Relap5计算，确定多孔介质控制方程的能量源项。通过用户自定义程序将动量源项与能量源项编译至FLUENT求解器中。通过FLUENT求解器求解控制方程，获得SG一次侧流场、压力场、温度场等信息。并通过对比模拟结果与设计值，验证了计算的准确性。      

Three-dimensional study on steady thermohydraulics characteristics in secondary side of steam generator

Steam generator (SG), as the primary-to-secondary heat exchanger and pressure boundary of primary loop, should be integrated and perform well in heat transfer ability. Flow characteristics of the secondary side fluid of SG are essential to analyze U-tube wastage caused by the flow-induced vibration and thermal stress. In this paper, secondary side two-phase flow was simulated based on the porous media model. Additional momentum and energy source terms were appended to the momentum and energy equations for porous media region, respectively. The additional momentum source contained the resistances of downcomer, tube bundle, support plate and separator. The additional energy source included the heat transfer from primary side to secondary side fluid. Solving the governing equations by ANSYS FLUENT solver yielded the distributions of velocity, temperature, pressure, density and quality, which can be used in the analysis of flow-induced vibration and separators. The thermal-hydraulic characteristics of hot side differed from these of cold side considerably. The minimum flow quality of cold side was 0.07, while the maximum one of hot side was 0.71; the average flow quality of outlet was 0.272. The flow rate in the gap of the hot side was 1.02 times of that of the cold side.     

铅铋快堆螺旋管直流蒸汽发生器热工水力特性数值研究

本研究以铅铋快堆螺旋管直流蒸汽发生器（HOTSG）设计结构为研究对象,采用精细网格与多孔介质相结合的物理建模方法,通过一次侧三维湍流计算与二次侧用户自定义函数（UDF）分区传热计算相耦合的手段,在FLUENT求解器中开展了蒸汽发生器的热工水力特性数值分析研究。研究表明:铅铋入口附近的流量分配孔和腔室对应的直管段区域出现铅铋流速峰值,径向最大速度为0.431 m/s;入口腔室至管束区位置受到阻力突变的影响,压力、横流速度、轴向速度变化较大;热工参数变化符合流动与传热机理,临界热流密度（CHF）点附近一二次侧温差最大为109.61 K,此处最大热流密度为323.55 kW/m2。该研究将为铅铋快堆HOTSG结构设计、流致振动及安全评价提供重要的参考。      

钠冷快堆二回路系统热工水力瞬态分析程序SELTAC的开发与应用

针对钠冷快堆二回路系统的具体结构和运行特点，对中间热交换器、直流蒸汽发生器、钠缓冲罐以及泵、管道等设备和部件建立模型，采用FORTRAN语言自主编制了二回路系统热工水力瞬态分析程序SELTAC。利用中国实验快堆的停堆试验数据对所编制程序进行了初步验证。结果表明，程序计算值与试验值趋势一致，最大相对偏差不超过4.34%，吻合程度较好。将验证后的程序与一回路系统程序耦合，分析了某600 MW钠冷快堆在主热传输系统保持排热能力时的紧急停堆工况，得到了二回路系统的瞬态特性，为大型商用快堆电站的设计提供了参考。