The swiss heating reactor (SHR) for district heating of small communities

With fossil fuel running out in a foreseeable future, it is essential to develop substitution strategies. Some 40–50% of the heat demand in industrial countries is below 120°C, for space heating and warm water production, causing a corresponding fraction of air polution by SO₂ and to a lesser extent NO_x if fossil fuels are used. Yet, contemporary LWR technology makes it feasible to supply a district heating network without basically new reactor development. Units in the power range 10–50 MW are most suitable for Switzerland, both in respect of network size and of the democratic decision making structure. A small BWR for heating purpose only is being developed by parts of the Swiss Industry and the Swiss Federal Institute for Reactor Research (EIR). The economic target of 100–120 SFr/MWh heat at the consumer's seems achievable.     

基于CFD方法的高温热管特性研究

高温热管运行特性的分析与预测,对热管设计和应用具有重要意义。为分析高温热管内两相流动传热特性,首先建立钠热管的计算流体力学（CFD）分析模型,并对模型计算值与钠热管稳态实验数据进行对比校核,模拟结果与实验测点温度的绝对误差小于40℃,相对误差在5%以内;其次,利用本文模型与方法对不同传热功率和倾角下的热管内部流场特性进行分析研究。研究表明,均匀加热条件下,蒸气腔内的速度在蒸发段接近线性变化,而在冷凝段,气体流速减小致使压强回升,同时,蒸气的流动压降和速度随加热功率增加呈下降趋势;在热管水平和倾角运行工况,热管内两相流动压降中液相压降均占主导;而气液间剪切效应中,气体流动速度为主导效应。本文模型可为热管堆等高温热管应用领域提供热管设计与分析方法。      

以R134a为工质的乏燃料池非能动冷却热管实验研究

随着分离式热管不断被提出用于核电站非能动余热排出方案中,开展针对大尺度分离式热管的换热性能的实验研究变得日益迫切。为此,本文开展了以R134a为工质的304不锈钢材质的分离式热管传热特性实验研究,获得了热管整体换热性能、蒸发段内部温度分布特性,以及热源温度和冷凝段外风速对热管工作温度、换热量、换热系数和循环流量的影响。热管蒸发段内R134a经历过冷、两相和过热状态,其中两相区域较长,达6.6 m,因而具有较好的换热能力,在所研究的工况下换热量最高达21 kW。参数敏感性分析表明,热源入口温度和冷凝段风速的增大能促进热管的换热性能,特别是热源入口温度的影响更显著。冷凝段风速较小时,其对换热量的影响较为显著,然而随空气速度的增加,影响降低。此外,依据试验数据拟合得到了换热量与冷热源温差的经验关系式,能在工程应用中快速预测热管的性能。     

Water flow rate pulsations with a flicker power spectrum in a commercial sodium steam generator

The spectral characteristics of the water flow rate pulsations in the evaporation module of a steam generator with sodium heating in the BN-600 power-generating unit at nominal power are investigated. The results show that the power spectrum of the flow rate pulsations vary inversely as the frequency (flicker pulsations). The appearance of the high-energy low-frequency pulsations of the water flow rate is attributed to a critical heat transfer regime (nonequilibrium phase transition) which is realized in the evaporation module.__________Translated from Atomnaya Énergiya, Vol. 98, No. 2, pp. 105–110, February, 2005.     

An investigation of transient heat transfer in the region of flow boiling dryout with freon-12 in a heated tube

Power transient experiments using vertical round tube test sections have provided information on the heat transfer characteristics associated with a change from pre-dryout to post-dryout flow boiling conditions. The test sections were heated by passing electric current along the tube wall, and cooled internally by Freon-12 flowing upwards through the tube.Seven steel tubes of various sizes were used (internal diameters in the range 7.1–26.6 mm, wall thicknesses 0.9–2.0 mm, and lengths of 0.9–3.9 m). Data were obtained for coolant mass fluxes in the range 150–3270 kg m⁻² s⁻¹, at a nominal pressure of 1.0 MPa, with exit qualities in the range 0.3–1.0. The transients were initiated by small increases in power input to the test section. Heat transfer characteristics were determined by calculating wall temperature responses as functions of time and comparing these with the corresponding temperature traces recorded in the experiments.In relation to the temperature responses of the tube wall under these transient conditions, the results show that transition boiling has only a slight effect and that film boiling has very significant effects.     

高温锂热管三相耦合数值模拟

为了研究锂热管的传热机理,推动锂热管在小堆中的应用。采用COMSOL Multiphysics软件,建立了管壁、吸液芯和管内蒸气腔室的固液气三相耦合模型,对热管的温度分布、压力分布和速度分布进行计算。结果表明:当蒸发段的热流从13.9 kW增加到20.8 kW时,管壁温度、蒸气温度、蒸气压力以及吸液芯内部的液体压力、液体轴向速度随着加热功率的增加而增加,蒸气轴向速度随着加热功率的增加先增加后降低。在稳态运行时,管壁温度呈现阶梯形下降,而蒸气温度和压力基本保持不变,表明了锂热管具有良好的等温性。      

Performance evaluation of ultra-long lithium heat pipe using an improved lumped parameter model

Lithium heat pipes have broad applications in heat pipe cooling reactors and hypersonic vehicles owing to their ultra-high working temperature.In particular,when the length of the lithium heat pipe is ultra-long,the flow and heat transfer characteristics are more complex.In this study,an improved lumped parameter model that considers the Marangoni effect,bending effect,and different vapor flow patterns and Mach numbers was developed.There-after,the proposed model was verified using the University of New Mexico's Heat Pipe and HTPIPE models.Finally,the verified model was applied to simulate the steady-state operation of an ultra-long lithium heat pipe in a Heat Pipe-Segmented Thermoelectric Module Converters space reactor.Based on the results:(1)Vapor thermal resistance was dominant at low heating power and decreased with increasing heating power.The vapor flow inside the heat pipe developed from the laminar to the turbulent phase,whereas the liquid phase in the heat pipe was always laminar.(2)The vapor pressure drop caused by bending was approximately 22-23％of the total,and the bending effect on the liquid pressure drop could be ignored.(3)The Marangoni effect reduced the capillary limit by hindering the liquid reflux,especially at low vapor temperatures.Without considering the Marangoni effect,the capillary limit of the lithium heat pipe was overestimated by 9％when the vapor temperature was 1400 K.(4)The total thermal resistance of the heat pipe significantly increased with increasing adiabatic length when the vapor tempera-ture was low.Further,the wick dryness increased with increasing adiabatic length at any vapor temperature.Such findings improve on current knowledge for the optimal design and safety analysis of a heat pipe reactor,which adopts ultra-long lithium heat pipes.     

Effect of temperature on graphite oxidation behavior

The temperature dependence of oxidation behavior for the graphite IG-11, used in the HTR-10, was investigated by thermogravimetric analysis in the temperature range of 400–1200 °C. The oxidant was dry air (water content <2 ppm) with a flow rate of 20 ml/min. The oxidation time was 4 h. The oxidation results exhibited three regimes: in the 400–600 °C range, the activation energy was 158.56 kJ/mol and oxidation was controlled by chemical reaction; in the 600–800 °C range, the activation energy was 72.01 kJ/mol and oxidation kinetics were controlled by in-pore diffusion; when the temperature was over 800 °C, the activation energy was very low and oxidation was controlled by the boundary layer. Due to CO production, the oxidation rate increased at high temperatures. The effect of burn-off on activation energy was also investigated. In the 600–800 °C range, the activation energy decreased with burn-off. Results of low temperature tests were very dispersible because the oxidation behavior at low temperatures is sensitive to inhomogeneous distribution of any impurity, and some impurities can catalyse graphite oxidation.     

Experimental study on the sub-atmospheric loop heat pipe passive cooling system for spent fuel pool

As one kind of the natural circulation cooling system, loop heat pipe is promising in improving the safety of the nuclear power station since it is passive and has no electricity driven components. A novel heat pipe cooling system is designed for passively removing the residual heat released by the spent fuel stored in the spent fuel pool (SFP) under the accidental conditions such as the station blackout. This system is characterized by its large-diameter and long-length evaporator. Its working fluid is water and it's sub-atmospheric. To test such system's heat transfer performance and get to know its thermo-fluid dynamics, a test facility for a simplified heat pipe made of one evaporator tube and one condenser has been developed. The heat transfer rate of the simplified heat pipe is obtained in a wide range of conditions covering the potential working conditions in spent fuel pool. Moreover, it's found that heat pipe with such a large-diameter and long-length evaporator is vulnerable to be unstable. The periodic state mode is more likely to happen when the heat source temperature, the air velocity or the volumetric filling ratio is low. Furthermore, the effects of hot water temperature, the air velocity and the filling ratio of the water in the circulation system have been analyzed.     

液态金属高温热管传热极限研究

热管作为一种高效可靠、可进行长距离传热的非能动设备,在核能领域有着广泛的应用。本文针对工质为钠、充液量为158 g与208 g的毛细驱动热管,对其传热极限开展实验和理论研究。实验方面,设计搭建了高温热管传热极限测试分析实验平台,研究了液态金属高温热管在不同水平倾角和不同加热功率下传热功率的变化。理论方面,验证了连续流动极限与夹带极限理论模型的正确性,总结了两种极限的发生规律。研究发现,热管连续流动极限影响热管的启动;由于水平夹角较大时转变温度较高,因此大角度下的热管更容易发生连续流动极限,小角度下经验模型的预测误差在6.58%以内,大角度下误差超过28%。夹带极限发生时热管蒸发段温度骤升且冷凝段温度出现波动,热管倾角越大夹带极限越容易发生,经验模型在不同角度下均存在误差,大角度下误差超过100%。本文总结了连续流动极限与夹带极限的发生规律,为先进核反应堆系统中热管的设计提供参考。     

碱金属高温热管启动运行特性分析

本文研究了碱金属高温热管的启动及运行瞬态,在传统热管冷态启动模型基础上,建立了以蒸气流态转变为划分依据的冷态启动三阶段模型,并依据该模型开发了热管启动瞬态分析程序HPSTAC,程序模拟结果与实验值相对偏差不超过15.7%。使用该程序模拟单根钠钾热管的启动瞬态并进行了敏感性分析,结果表明：启动开始后450、660、1 550 s,热管分别进入启动第2、3阶段与准稳态,蒸气区域趋于一致性温度837 K;启动环境温度主要影响冷凝段升温速度,输入热流密度对热管启动各阶段的影响存在阈值效应。     

Prediction of flooding gas velocity in gas–liquid counter-current two-phase flow in inclined pipes

The purpose of the experimental study is to investigate the effects of pipe inclination, pipe length, pipe diameter and surface tension of the working liquid on the onset of flooding of gas–liquid adiabatic counter-current two-phase flow in inclined pipes. Flooding in inclined pipes were observed by using the combination of visual observation, measurement of discharged liquid flow rate and time variation of liquid hold-up. And it was defined as the maximum air flow rate at which the discharged liquid flow rate is equal to the inlet liquid flow rate. As a result we proposed a correlation to predict the flooding gas velocity in inclined pipes under a given liquid flow rate, and the predictions agreed well with the experimental observations.     

An experimental and analytical study of the effect of axial power profile on CHF

The effect of axial heat flux distribution (AFD) on the critical heat flux (CHF) was investigated. CHF measurements were obtained with HFC-134a cooled vertical tubes having four non-uniform and one uniform AFD profiles. The HFC-134a test conditions covered a pressure range from 1.6 to 2.4 MPa, a mass-flux range from 2.8 to 4.7 Mg m⁻² s⁻¹, and an inlet-quality range from −0.9 to 0. The water-equivalent pressure and mass-flux ranges are 10–14 MPa and 4–6.5 Mg m⁻² s⁻¹, respectively.In general, the observed AFD effect on critical power is small at high inlet subcoolings. At low inlet subcoolings, the critical power for the inlet-peak profile is up to 15% higher than that for the outlet-peak profile. A local conditions analysis showed that the AFD has the strongest effect on CHF at high dryout qualities. CHF values for non-uniform AFDs could be 50% lower than those for the uniform AFD. The AFD effect on CHF becomes diminished with decreasing dryout quality.Four different approaches to account for the effect of AFD on CHF were assessed against the experimental values from the current experiment. The boiling-length-average heat-flux approach with the boiling-length starting point at the onset of annular flow (OAF) provided the best prediction of the critical power and the CHF location.     

Analysis and Optimization Laminar Viscous Flow Through a Channel

Optimizing of laminar viscous flow through a pipe by two dimensionless values is investigated analytically. Dimensionless entropy generation and pumping power to heat transfer rate ratio are used as basis for constant viscous and the temperature dependence on the viscosity. For this matter we calculate entropy generation and pumping power for a fully developed in a pipe subjected to constant wall temperature for either constant viscosity and the variable viscosity. The variation entropy generation increase along the pipe length for viscous fluid is drawn, either the variation summation dimensionless entropy generation and the pumping power to heat transfer rate ratio are varying the fluid inlet temperature for fixed pipe length and are varying pipe length for fixed fluid inlet temperature are drawn. For low heat transfer conditions the entropy generation due to viscosity friction becomes dominant and the dependence of viscosity with the temperature becomes essentially important to be considered.     

Analysis of the capability of system codes to model cavitation water hammers: Simulation of UMSICHT water hammer experiments with TRACE and RELAP5

The capabilities of the nuclear system transient codes TRACE and RELAP5 to model coupled two-phase flow and pressure wave propagations in a pipe are assessed by analyzing the UMSICHT PPP cavitation water hammer experiments 329 and 135 after valve closure. Time-dependent pressure, flow behaviour, and the generation and collapse of vapor bubbles at the valve and the first bridge are discussed. We show that both codes are able to model the flow behaviour of the water hammer for the high pressure and high temperature case 329 (initially 10–13 bar and 420 K), however condensation heat transfer for the base case needed to be increased in order to accurately model the magnitude of the first pressure excursion. The experimental broadening and damping of the subsequent pressure peaks by Fluid-Structure Interaction (FSI) phenomena arising from the interaction of the flow with the vibrations of the piping structure are not considered in the modeling results. For the lower pressure and temperature case 135 (initially 1–4 bar and 294 K), the TRACE code provides a good approximation of the propagation of the pressure wave and the void fraction behaviour, already with base case conditions, while RELAP5 overpredicts the vapor generation along the pipe and, as a result, considerably underpredicts the pressure amplitudes and overpredicts the water hammer frequency.     

Experimental study of heat transfer enhancement in narrow rectangular channel with longitudinal vortex generators

In order to enhance heat transfer in cooling channels of plate-type fuel elements in reactor cores, the experimental research is conducted on the heat transfer and pressure drop in horizontal narrow rectangular channels with mounted longitudinal vortex generators (LVGs) for water flow with Prandtl number Pr = 4–5. The parameters examined were: flow velocity from 0.5 to 3.4 m/s, Reynolds number from 3000 to 20,000, heat flux 43.6 kW/m², maximum system pressure 1.3 atm, and viscosity ratio from 1.05 to 1.2. It is found that the LVGs could greatly improve the heat transfer rate by 10–45%. Thermal performance is compared under three constraints, i.e., identical mass flow rate (IMF), identical pressure drop (IPD) and identical pumping power (IPP). It is found that the heat transfer performance of channel with LVGs on two sides are better than those on one side. Application of LVGs to plate-type fuel element is a potential technique for next generation advanced nuclear reactors concepts.     

Evolution of the structure of a gas–liquid two-phase flow in a large vertical pipe

The evolution of the structure of a gas–liquid flow in a large vertical pipe of 195 mm inner diameter was investigated at the TOPFLOW test facility in Rossendorf. Wire-mesh sensors were used to measure sequences of two-dimensional distributions of local instantaneous gas fraction within the complete pipe cross-section. The sensors own a resolution of 3 mm at a frequency of 2500 Hz. Superficial velocities were varied in a range covering flow regimes from bubbly to churn-turbulent flow. The distance between the gas injection and the sensor position was changed using a so-called variable gas injection system. It consists of six gas injection units, each equipped with three rings of injection orifices in the pipe wall (orifice diameter: 1 and 4 mm), which are fed from ring chambers. The gas flow towards these distributor chambers is individually controlled by valves. Measured bubble-size resolved radial gas fraction profiles reveal differences in the lateral migration of bubbles of different size starting from the injection at the wall. The evolution of bubble-size distributions allows to study bubble coalescence and break-up. The influence of the physical properties of the fluid was studied by comparing cold air–water experiments with steam–water tests at 65 bar.     

加热功率对低压低高差自然循环系统两相流动特性影响研究

对浮动式核电站中一类具有倾斜热管段的低压低高差自然循环系统的两相流动特性进行了实验研究,分析了加热功率对两相流动特性的影响。结果表明,不同功率条件下系统存在两相稳定冷凝和伴随蒸汽冷凝诱发水锤两相振荡2种流动模式,热管段内过冷水倒流和蒸汽与低温过冷水直接接触冷凝是导致2种流动模式的内在机制。此外,蒸汽冷凝诱发水锤的发生会产生较大压力脉冲,并导致过冷水倒流长度显著增加,进而加剧系统流动不稳定。进一步研究表明,加热段出口含气率可以作为流动不稳定判断依据。      

Power up-grading study for the first Egyptian research reactor

In the present work, power up-grading study is performed, for the first Egyptian Research Reactor (ET-RR-1), using the present fuel basket with 4×4 fuel rods, (17.5 mm pitch), and a proposed fuel basket with 5×5 fuel rods, (14.0 mm pitch), without violating the thermal hydraulic safety criteria. These safety criteria are; fuel centerline temperature (fuel melting), clad surface temperature (surface boiling), outlet coolant temperature, and maximum heat flux (critical heat flux ratio). Different thermal reactor powers (2–10 MW) and different core coolant flow rates (450, 900, 1350 m³ h⁻¹) are considered. The thermal hydraulic analysis was performed using the subchannel code COBRA-IIIC for the estimation of temperatures, coolant velocities and critical heat flux. The neutronic calculations were performed using WIMS-D4 code with 5 — group neutron cross section library. These cross sections were adapted to use in the two-dimensional (2-D) diffusion code DIXY for core calculations. The study concluded that ET-RR-1 power can be upgraded safely up to 4 MW with the present 4×4-fuel basket and with the proposed 5×5-fuel basket up to 5 MW with the present coolant flow rate (900 m³ h⁻¹). With the two fuel arrays, the reactor power can be upgraded to 6 MW with coolant flow rate of 1350 m³ h⁻¹ without violating the safety criterion. It is also concluded that, loading the ET-RR-1 core with the proposed fuel basket (5×5) increases the excess reactivity of the reactor core than the present 4×4 fuel matrix with equal U-235 mass load and gave better fuel economy of fuel utilization.     

CATHARE phase separation modeling for small breaks in horizontal pipes with stratified flow

Phase separation is analysed for a two-phase mixture flowing through dividing tee junctions with a small ratio of branch to main pipe diameter. The case of a stratified flow in a horizontal main pipe connected to a horizontal, vertical upward or vertical downward branch is considered. A model is developed on the basis of the available data banks. A wide range of parameters is covered (pressure, 0.2–7 MPa; main pipe diameter, D = 80–284 mm; branch diameter, 4–34 mm). The model has been introduced in the CATHARE Pressurized Water Reactor (PWR) safety code, thereby leading to a considerable improvement of the code predictions.