200MW 核供热堆的非线性动态模型

本文为２００ＭＷ核供热堆建立了一个用于大功率运行范围控制系统仿真的非线性动态模型。模型除了采用点中子动态方程、集中参数的慢化剂温度和燃料温度负反馈等压水堆控制系统常用的建模方法之外，为了使模型适用于大功率运行范围，还重点考虑了主回路自然循环对堆芯内冷却剂和燃料棒之间的传热系数、主换热器换热系数、主回路时间常数的影响，以及二回路流量变化引入的非线性。仿真结果表明，模型具有较高的精度，可用于控制系统仿真。     

Natural circulation characteristics of lead-based reactor under long-term decay heat removal

The natural circulation of primary coolant plays an important role in removing the decay heat in Station-Black-out (SBO) accident from reactor core to decay heat removal systems, such as RVACS and PHXS cooling, for lead-based reactor. In order to study the natural circulation characteristics of primary coolant under Reactor Vessel Air Cooling System (RVACS) and primary heat exchangers (PHXs) cooling, which are crucial to the safety of lead-based reactors. A three-dimensional CFD model for the China Lead-based Research Reactor (CLEAR-I) has been built to analyze the thermal-hydraulics characteristics of primary coolant system and the cooling capability of the two systems. The abilities of the two cooling systems with different decay heat powers were discussed as well. The results demonstrated that the decay heat could be removed effectively only relying on either of the two systems. However, RVACS appeared the obvious thermal stratification phenomenon in the cold pool. Besides, with the increase in decay heat power, the natural circulation capacity of primary coolant between the two systems had a significant difference. The PHXs cooling system was stronger than the RVACS, with respect to the mass flow of primary coolant and the average temperature difference between cold pool and hot pool.     

After heat removal by a closed secondary cooler circuit on the example of the advanced high temperature reactor AHTR 500 with central graphite column

The application of natural convection in connection with an after heat removal concept in general supports the claim for an inherent safety concept for advanced high temperature reactors (HTR). The effectivity of such an after heat removal (AHR) concept will be explored exemplarily for the process-heat reactor AHTR 500 with central graphite column by a thermohydraulic simulation of a secondary cooler circuit which is thermally connected with the primary circuit. This coupling is undertaken by an AHR-cooler located in the upper part of the graphite column. The heat removal from the secondary circuit is taking place outside of the reactor by a secondary heat exchanger under the assumption that the latter is cooled by a water capacity flow on an ambient temperature level. The developed calculation model determines iteratively the dynamic and thermal positions of equilibrium in the primary and secondary circuit which in the after heat removal mode of operation are exclusively run by natural convection. Different types of design for the central column heat exchanger (coaxial tube, U-tube and helically coiled tube heat exchanger) have been compared. For the secondary heat exchanger a parallel tube design has been supposed. The choice of the secondary flow medium as well as the most important limiting quantities influencing the transmission of heat via the secondary circuit during the after heat removal mode of operation are subject of a parameter study.     

Fundamental Experiment of Potassium Heat Exchanger Using Principle of Heat Pipe

Abstract

In order to provide compact and reliable sodium equipments including a steam generator, performance tests are conducted with a potassium heat exchanger, which is featured by the separate construction of primary and secondary coolant systems. A small amount of potassium plays a role as an intermediate media of heat transportation between these two coolant systems. Heat is transfered by evaporation and condensation of potassium on the surfaces of the primary and the secondary coolant pipings, respectively. The tests are performed in the temperature range of 200-300°C and the maximum heat transfer reaches 1.3 kW (heat transfer rate at the primary heating source: 8.6 W/cm² at 300°C). The experimental results are analyzed by using Langmuir's and Schrage's equations and close agreement between experiment and theory is obtained.     

Experiments on the Heat Transfer and Natural Circulation Characteristics of the Passive Residual Heat Removal System for an Advanced Integral Type Reactor

Experiments on the heat transfer characteristics and natural circulation performance of the passive residual heat removal system (PRHRS) for the SMART-P have been performed by using the high temperature/high pressure thermal-hydraulic test facility (VISTA). The VISTA facility consists of the primary loop, the secondary loop, the PRHRS loop, and the auxiliary systems to simulate the SMART-P, a pilot plant of the SMART. The primary loop is composed of the steam generator (SG) primary side, a simulated core, a main coolant pump, and the loop piping, and the PRHRS loop consists of the SG secondary side, a PRHRS heat exchanger, and the loop piping. The natural circulation performance of the PRHRS, the heat transfer characteristics of the PRHRS heat exchangers and the emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are intensively investigated. The experimental results show that the coolant flows steadily into the PRHRS loop and that the heat transfers through the PRHRS heat exchanger and the emergency cooldown tank are sufficient enough to enable a natural circulation of the coolant. The results also show that the core decay heat can be sufficiently removed from the primary loop with an operation of the PRHRS.     

EBR-Ⅱ余热排出实验及非能动余热排出系统性能分析

本文基于SAC-CFR事故分析程序,在国际原子能机构联合研究项目（IAEA CRP）框架下,对美国EBR-Ⅱ快堆余热排出实验（SHRT-17、SHRT-45R）进行了分析,计算了事故余热排出系统（DRACS）的响应、衰变热功率、关键部件的冷却剂温度、一回路的质量流量等关键参数。将计算参数与实验数据进行了对比,对程序的有效性进行了验证。计算结果表明,在SHRT-17工况下,随DRACS风门的打开,每台事故热交换器可带走330 406.4 W的堆芯余热,DRACS具有长期带走衰变热的能力。     

Analytical one-dimensional frequency response and stability model for PWR nuclear power plants

A dynamic model for PWR nuclear power plants is presented. The plant is assumed to consist of a one-dimensional single-channel core, a counterflow once-through steam generator (represented by two nodes according to the non-boiling and boiling region) and the necessary connecting coolant lines. The model describes analytically the frequency response behaviour of important parameters of such a plant with respect to perturbations in reactivity, subcooling or mass flow (both at the entrances to the reactor core and/or the secondary steam generator side), and perturbations in steam load or system pressure (on the secondary side of the steam generator). From corresponding ‘open’ loop considerations, it can then be concluded - by applying the Nyquist criterion - upon the degree of the stability behaviour of the underlying system. Based on this theoretical model, a computer code named ADYPMO has been established.From the knowledge of the frequency response behaviour of such a system, the corresponding transient behaviour with respect to a stepwise or any other perturbation signal can also be calculated by applying an appropriate retransformation method, e.g. by using the digital code FRETI. To demonstrate this procedure, a transient experimental curve measured during the pre-operational test period at the PWR nuclear power plant KKS Stade was recalculated using the combination ADYPMO-FRETI. Good agreement between theoretical calculations and experimental results give an insight into the validity and efficiency of the underlying theoretical model and the applied retransformation method.     

Influence of cyclic temperature transients on the creep behaviour of heat exchanger tubes

Due to thermal fluxes hot streaks exist in the coolant media of heat exchanger components. They cause alternating cyclically secondary stresses in the component walls which superpose on the primary stresses due to internal pressure or bending. Experimentally it was shown that hot streaks at high temperatures influence the creep behaviour very strongly. Dependent on the ratio of primary and secondary stresses the creep rate of the components is higher than the creep behaviour at the highest cycle temperature under primary stresses only.     

不同运行工况下蒸汽发生器热工水力稳态特性数值研究

基于相似模化理论建立了蒸汽发生器一、二回路流体及传热管流 固耦合传热的单元管三维物理模型,对大亚湾核电厂蒸汽发生器不同工况下的热工水力稳态特性进行了数值模拟研究。采用热相变模型描述二回路汽液两相流动与换热、流-固耦合模型描述一回路冷却剂借助U型管与二回路流体换热。数值计算结果表明：满负荷运行时,传热管内壁温度变化趋势与一次侧流体基本一致,外壁温度与二次侧流体温度变化趋势相同;截面平均含汽率沿传热管高度的升高呈上升趋势,出口质量含汽率与大亚湾核电厂实际运行参数相符;随负荷降低一回路出口温度基本不变,二回路出口温度升高,质量含汽率及传热系数下降,平均传热系数与Rohsenow经验关联式的计算结果基本吻合。     

Preliminary issues associated with the next generation nuclear plant intermediate heat exchanger design

The Next Generation Nuclear Plant, with emphasis on production of both electricity and hydrogen, involves helium as the coolant and a closed-cycle gas turbine for power generation with a core outlet/gas turbine inlet temperature of 850-950 °C. In this concept, an intermediate heat exchanger is used to transfer the heat from primary helium from the core to the secondary fluid, which can be helium, a nitrogen/helium mixture, or a molten salt. This paper assesses the issues pertaining to shell-and-tube and compact heat exchangers. A detailed thermal-hydraulic analysis was performed to calculate heat transfer, temperature distribution, and pressure drop inside both printed circuit and shell-and-tube heat exchangers. The analysis included evaluation of the role of key process parameters, geometrical factors in heat exchanger designs, and material properties of structural alloys. Calculations were performed for helium-to-helium, helium-to-helium/nitrogen, and helium-to-salt heat exchangers.     

A specialized code for operation transient analysis and its application in fluoride salt-cooled high-temperature reactors

Fluoride salt-cooled high-temperature reactors (FHRs) include many attractive features,such as high temperature,large heat capacity,low pressure and strong inherent safety.Transient characteristics of FHR are particularly important for evaluating its operation performance.Thus,a specialized code OCFHR (operation and control analysis code of FHR) issued to study an experimental FHR's operation behaviors.The geometric modeling of OCFHR is based on one-dimensional lumped parameter method,and some simplifications are taken into consideration during simulation due to the existence of complex structures such as pebble bed,intermediate heat exchanger (IHX),air radiator (AR) and multiply channels.A point neutron kinetics model is developed,and neutron physics calculation is needed to provide some key inputs including axial power density distribution,reactivity coefficients and parameters about delayed neutron precursors.For analyzing the operational performance,five disturbed transients are simulated,involving reactivity step insertion,variations of coolant mass flow rate of primary loop and intermediate loop,adjustment of air inlet temperature and mass flow rate of air cooling system.Simulation results indicate that inherent self-stability of FHR restrains severe consequences under above transients,and some dynamic features are observed,such as large negative temperature feedbacks,remarkable thermal inertia and high response delay.     

运行条件对蒸汽发生器热工参数影响的仿真研究

以减轻蒸汽发生器破管事故及考察核电站电力升级为目的,参考大亚湾核电站蒸汽发生器的运行参数,基于分布参数法建立了核动力蒸汽发生器一维数学模型,开发了基于MATLAB的动态仿真程序,进行了改变运行条件时蒸汽发生器热工参数仿真计算。计算结果表明:与满负荷正常运行条件相比,在降低二回路运行温度或增加二回路流量时,二回路预热段变短,出口焓大幅升高;质量含汽率在降低温度时提高54%,增加流量时提高28%;一、二回路及管壁整体温度降低;一回路和内壁温降增大。该计算结果揭示了蒸汽发生器的内在传热规律,可为缓解U形管恶化及提升电力的相关操作提供一定理论依据。     

Stability analysis of a liquid-metal reactor and its primary heat transport system

Part of the reactor design process is the assessment of the impact of different design changes on pre-defined performance criteria including stability of the reactor system under different conditions. This work focuses on the stability analysis of a combined liquid-metal reactor and primary heat transport system where system parameters are free to vary, with particular interest in low reactor power, low reactor coolant flow conditions. Such conditions might be encountered, for example, after a loss of flow without scram in some passively safe reactor designs. Linear-stability-analysis-based methods are developed to find the stability regions, stability boundary surface in system parameter space, and frequency of oscillation at oscillatory instability boundaries. Models are developed for the reactor, detailed thermal hydraulic reactivity feedback associated with coolant outlet and inlet temperatures, decay heat and primary system. The developed stability analysis tools are applied to the system model. The system parameters include integral reactivity parameters, decay heat, primary system mass, coolant flow and natural circulation flow. The resulting stability boundary surface and its associated frequency of oscillation surface in multidimensional system parameter space show the effect of system parameter changes. By adopting model parameters from liquid-metal reactor designs, a stability prediction procedure is illustrated.     

Incorporation of lithium lead eutectic as a working fluid in RELAP5 and preliminary safety assessment of LLCS

The current work involves thermal hydraulic calculation of Lithium Lead Cooling System (LLCS) for the Indian test blanket module (TBM) for testing in International Thermonuclear Experimental reactor (ITER). It uses the RELAP portion of RELAP/SCDAPSIM/MOD4.0. Lithium-lead eutectic (LLE) has been used as multiplier, breeder and coolant in TBM. Thermodynamic and transport properties of the LLE have been incorporated into the code. The main focus of this study is to check the heat transfer capability of LLE as coolant for TBM system for steady state and the considered anticipated operational occurrences (AOO's), namely, loss of heat source, loss of primary flow and loss of secondary flow. The six heat transfer correlation (reported for liquid metals in the literature) has been tested for steady state analysis of LLCS loop and results are roughly same for all of them. A good agreement has been observed between the operating conditions of LLCS with those of RELAP5 calculations. Results from transient calculations show that a maximum temperature of 875 K is attained during a 300 s loss of primary flow (LLE).     

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-hydraulic performance of heavy liquid metal in straight-tube and U-tube heat exchangers

Motivated by an increased interest in heavy liquid metal (lead or lead alloy) cooled fast reactors (LFR) and accelerator-driven system (ADS), the present paper presents a study on resistance characteristics and heat transfer performance of liquid lead bismuth eutectic (LBE) flow through a straight-tube heat exchanger and a U-tube heat exchanger. The investigation is performed on the TALL test facility at KTH. The heat exchangers have counter-current flow arrangement, and are made from a pair of 1-m-long concentric ducts, with the LBE flowing in the inner tube of 10 mm I.D. and the secondary coolant flowing in the annulus. The inlet temperature of LBE into the heat exchangers is from 200 °C to 450 °C with temperature drops from 0 °C to 100 °C within the LBE flow range of Re = 10⁴-10⁵. Analysis of the experimental results obtained provides a basic understanding and quantification of the regimes of lead-bismuth flow and heat transfer through a straight tube and a U-shaped tube. The unique data base also serves as benchmark and improvement for system thermal-hydraulic codes (e.g. RELAP, TRAC/AAA) whose development and testing were dominantly driven by applications in water-cooled systems. Lessons and insights learnt from the study and recommendations for the heat exchanger selection are discussed.     

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.     

一种低温堆核电站的动态数学模型

针对研发的采用一体化布置、全功率自然循环的低温核反应堆电站,建立了一个可用于大功率运行范围控制系统仿真的动态数学模型.模型采用了六组缓发中子动态方程(考虑了慢化剂温度和燃料温度反应性负反馈)、集中参数的堆芯传热模型以及自然循环流动模型,重点考虑了主回路自然循环对堆芯内冷却剂和燃料棒之间的传热系数、主换热器换热系数、主回路时间常数的影响.仿真结果表明,模型能够正确反映低温堆核电站的主要动态特性,可用于电站控制系统仿真.     

Experimental study on the natural circulation of a passive residual heat removal system for an integral reactor following a safety related event

Natural circulation characteristics of an integral type reactor during the operation of a passive residual heat removal system (PRHRS) following a safety related event has been experimentally investigated by using the VISTA facility. A PRHRS actuation trip signal is generated by a high power trip signal following a steam flow increasing event. The experimental results show that the single-phase coolant flows steadily in the primary loop by a natural convection process and that it effectively removes the decay heat from the core through a steam generator during the PRHRS operation. The heat transfers through the PRHRS heat exchanger and the emergency cooldown tank (ECT) are sufficient enough to enable a two-phase natural circulation of the coolant in the PRHRS loop.     

Concept of a pressurized water reactor cooled with supercritical water in the primary loop

A novel concept of a pressurized water reactor with a primary loop cooled with supercritical water is introduced and analyzed in this work. A steam cycle analysis has been performed to illustrate the advantages of such a nuclear power plant with respect to specific power and thermal efficiency. Moreover, a reactor pressure vessel concept including its internals and a suitable core and fuel assembly design are presented overcoming the problems, which arise due to the high heat up of the coolant and the density change involved with it. The core power and coolant density distributions are predicted with coupled neutronic and thermal-hydraulic analyses. The method features the definition of inlet orifices for coolant mass flow adjustment within the core as well as an additional tool for the interpolation of local pin power data. The latter one has been used for a successive sub-channel analysis of the hottest fuel assembly of the core, which provides a more detailed spatial resolution and thus predicts peak cladding temperatures, the maximum linear pin power of fuel pins, and maximum fuel temperatures. It can be shown that maximum temperatures of claddings and fuel are well below the material limits. Thanks to an average core exit temperature below the pseudo-critical temperature, the core concept leaves enough margin for additional uncertainties and allowances for operation.