From Chooz to the Ling'ao NPP:The Technology Transfer of Pressurized Water Reactor Technology from France to China

The transfer of pressurized water reactor(PWR)technology from France to China is an important event in the history of Sino-French scientific and technological relations.China has gradually achieved self-reliance in the field of PWR technology through the introduction and subsequent absorption of France's 900 MW reactors.Compared with the process of introducing and absorbing similar technology from the United States by France,China's experience has been more complicated.This circumstance reflects the differences in the nuclear power technology systems between the two countries.France's industrial strength and early acquisition of nuclear power technology laid a solid foundation for mastering PWR technology.On the other hand,although China established a weak foundation through the implementation of the"728 Project,"and tried hard to negotiate with France,the substantive content of the technology transfer was very limited.By way of the policy transition from"unhooking of technology and trade"to"integration of technology and trade,"China ultimately accomplished the absorption and innovation of PWR technology through the Ling'ao NPP.

Control system design for a pressure-tube-type supercritical water-cooled nuclear reactor via a higher order sliding mode method

Nuclear power plants exhibit non-linear and time-variable dynamics.Therefore,designing a control system that sets the reactor power and forces it to follow the desired load is complicated.A supercritical water reactor(SCWR)is a fourth-generation conceptual reactor.In an SCWR,the non-linear dynamics of the reactor require a controller capable of control-ling the nonlinearities.In this study,a pressure-tube-type SCWR was controlled during reactor power maneuvering with a higher order sliding mode,and the reactor outgoing steam temperature and pressure were controlled simultaneously.In an SCWR,the temperature,pressure,and power must be maintained at a setpoint(desired value)during power maneuvering.Reactor point kinetics equations with three groups of delayed neutrons were used in the simulation.Higher-order and classic sliding mode controllers were separately manufactured to control the plant and were compared with the PI controllers speci-fied in previous studies.The controlled parameters were reactor power,steam temperature,and pressure.Notably,for these parameters,the PI controller had certain instabilities in the presence of disturbances.The classic sliding mode controller had a higher accuracy and stability;however its main drawback was the chattering phenomenon.HOSMC was highly accurate and stable and had a small computational cost.In reality,it followed the desired values without oscillations and chattering.

Optimization of the fuel rod＇s arrangement cooled by turbulentnanofluids flow in pressurized water reactor (PWR)

In this paper, response surface methodology(RSM) based on central composite design(CCD) is applied to obtain an optimization design for the fuel rod's diameter and distance cooled by turbulent Al_2O_3–water nanofluid for a typical pressurized water reactor(PWR). Fuel rods and nanofluid flow between them are simulated 3D using computational fluid dynamics(CFD) by ANSYS-FLUNET package software. The RNG k–ε model is used to simulate turbulent nanofluid flow between the rods. The effect of different nanoparticles concentration is also investigated on the Nusselt number from heat transfer efficiency view point. Results reveal that when distance parameter(a) is in the minimum level and diameter parameter(r) is in the maximum possible level, cooling the rods will be better due to higher Nusselt number in this situation. Also, using the different nanoparticles on the cooling process confirms that Al_2O_3 averagely 17% and TiO_2 10% improve the Nusselt numbers.

Neutronic analysis of silicon carbide cladding accident-tolerant fuel assemblies in pressurized water reactors

In resonance with the Fukushima Daiichi Nuclear Power Plant accident lesson, a novel fuel design to enhance safety regarding severe accident scenarios has become increasingly appreciated in the nuclear power industry. This research focuses on analysis of the neutronic properties of a silicon carbide(SiC) cladding fuel assembly, which provides a greater safety margin as a type of accident-tolerant fuel for pressurized water reactors. The general physical performance of SiC cladding is explored to ascertain its neutronic performance. The neutron spectrum, accumulation of ^(239)Pu, physical characteristics,temperature reactivity coefficient, and power distribution are analyzed. Furthermore, the influences of a burnable poison rod and enrichment are explored. SiC cladding assemblies show a softer neutron spectrum and flatter power distribution than conventional Zr alloy cladding fuel assemblies. Lower enrichment fuel is required when SiC cladding is adopted. However, the positive reactivity coefficient associated with the SiC material remains to be offset. The results reveal that SiC cladding assemblies show broad agreement with the neutronic performance of conventional Zr alloy cladding fuel. In the meantime, its unique physical characteristics can lead to improved safety and economy.

Development and Application of Maintenance Template in Pressurized Water Reactor Nuclear Power Plant

Good practices of maintenance optimization in nuclear power field need to be effectively consolidated and inherited,and maintenance optimization can provide technology support to create a long-term reliable and economic operation for nuclear power plants( NPPs) especially for a large number of nuclear powers under construction. Based on the development and application of maintenance template in developed countries,and combining with reliability-centered maintenance( RCM) analysis results and maintenance experience data over the past ten years in domestic NPPs, the development process of maintenance template was presented for Chinese pressurized water reactor( PWR) NPP,and the application of maintenance template to maintenance program development and maintenance optimization combined with cases were demonstrated. A shortcut was provided for improving the efficiency of maintenance optimization in domestic PWR NPP,and help to realize a safe,reliable,and economic operation for domestic NPPs.

Theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor

This article presents a theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor. Through numerically solving the one-dimensional steady-state single-phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the steam generator, the natural circulation characteristics were studied. On the basis of the pre- liminary calculation analysis, it was found that natural circulation mass flow rate was proportional to the exponential function of the power and that the value of the exponent is related to the operating conditions of the secondary side of the steam generator. The higher the outlet pressure of the secondary side of the steam generator, the higher the pri- mary natural circulation mass flow rate. The larger height difference between the core center and the steam generator center is favorable for the heat removal capacity of the natural circulation.

Passive Cooldown Performance of Integral Pressurized Water Reactor

The design of an integral pressurized water reactor (IPWR) focuses on enhancing the safety and reliability of the reactor by incorporating a number of inherent safety features and engineered safety features. However, the characteristics of passive safety systems for the marine reactors are quiet different from those for the land nuclear power plant because of the more formidable and dangerous operation environments of them. This paper presents results of marine black out accident analyses. In the case of a transient, the passive residual heat removal system (PRHRS) is designed to cool the reactor coolant system (RCS) from a normal operation condition to a hot shutdown condition by a natural circulation, and the shutdown cooling system (SCS) is designed to cool the primary system from a hot shutdown condition to a refueling condition by a forced circulation. A realistic calculation has been carried out by using the RELAP5/MOD3.4 code and a sensitivity analysis has been performed to evaluate a passive cooldown capability. The results of the accident analyses show that the reactor coolant system and the passive residual heat removal system adequately remove the core decay heat by a natural circulation.

Non-integer Order Control Scheme for Pressurized Water Reactor Core Power

Tracking load changes in a pressurized water reactor(PWR)with the help of an efficient core power control scheme in a nuclear power station is very important.The reason is that it is challenging to maintain a stable core power according to the reference value within an acceptable tolerance for the safety of PWR.To overcome the uncertainties,a non-integer-based fractional order control method is demonstrated to control the core power of PWR.The available dynamic model of the reactor core is used in this analysis.Core power is controlled using a modified state feedback approach with a non-integer integral scheme through two different approximations,CRONE(Commande Robuste d’Ordre Non Entier,meaning Non-integer orderRobust Control)and FOMCON(non-integer order modeling and control).Simulation results are produced using MATLAB■program.Both non-integer results are compared with an integer order PI(Proportional Integral)algorithm to justify the effectiveness of the proposed scheme.Sate-spacemodel Core power control Non-integer control Pressurized water reactor PI controller CRONE FOMCON.

Investigation on two－phase critical flow for loss－of－coolant accident of pressurized water reactor

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Response characteristics of PWR primary circuit under SBLOCAs considering steam bypass discharging

Small-break superposed station blackout(SBO)accidents are the basic design accidents of nuclear power plants.Under the condition of a small break in the cold leg,SBO further increases the severity of the accident,and the steam bypass discharg-ing system(GCT)in the second circuit can play an important role in guaranteeing core safety.To explore the influence of the GCT on the thermal-hydraulic characteristics of the primary circuit,RELAP5 software was used to establish a numerical model based on a typical pressurized water reactor nuclear power plant.Five different small breaks in the cold-leg super-posed SBO were selected,and the impact of the GCT operation on the transient response characteristics of the primary and secondary circuit systems was analyzed.The results show that the GCT plays an indispensable role in core heat removal during an accident;otherwise,core safety cannot be guaranteed.The GCT was used in conjunction with the primary safety injection system during the placement process.When the break diameter was greater than a certain critical value,the core cooling rate could not be guaranteed to be less than 100 K/h;however,the core remained in a safe state.

Preliminary study of the tight lattice pressured heavy water reactor loaded with Pu/U and Th/U mixed fuels

To improve nuclear fuel utilization efficiency and prolong fuel cycle burn-up, a tight pitch lattice pressured heavy water reactor was investigated as an alternative of next generation of power reactors. It is shown that the high conversion ratio and negative coolant void reactivity coefficient are challenges in the reactor core physics designs. Various techniques were proposed to solve these problems. In this work, a tight pitch lattice and mixed fuel assemblies pressured heavy water reactor concept was investigated. By utilizing numerical simulation technique, it is demonstrated that reactor core mixed with Pu/U and Th/U assemblies can achieve high conversion ratio (0.98), long burn-up (60 GWD/t) and negative void reactivity coefficients.

A Novel Computerized Water Level Control System of PWR Steam Generator of Nuclear Power Plant

This paper presents a novel method to solve old problem of water level control system of pressurized water reactor (PWR) steam generator (SG) of nuclear power plant (NPP) .The level control system of SG plays an important role which effects the reliablity,safty,cost of SG and its mathematical models have been solved.A model of the conventional controller is presented and the existing problems are discussed. A novel rule based realtime control technique is designed with a computerized water level control (CWLC) system for SG of PWR NPP.The performance of this is evaluated for full power reactor operating conditions by applying different transient conditions of SG′s data of Qinshan Nuclear Power Plant (QNPP).

Flow Instability in Parallel Channels with Water at Supercritical Pressure: A Review

Research into flow instability at both subcritical and supercritical pressures has attracted attention in recent years because of its potential of occurrence in industrial heat transfer systems. Flow instability has the potential to affect the safety of design and operation of heat transfer equipment. Flow instability is therefore undesirable and should be avoided?in the design and operation of industrial equipment. Rahman?et al. reviewed studies on supercritical water heat transfer with the aim of providing references for SCWR researchers. It was found out that most of the CFD studies and experimental studies were performed with single tube geometry due to the complexity of parallel channel geometry. Because studies performed with parallel channel geometry could provide detailed information to the design of the SCWR core, they called for more studies in parallel channel geometry at supercritical pressures in the future. In order to help understand how flow instability investigations are carried out and also highlight the need to understand flow instability phenomenon and equip the designers and operators of industrial heat transfer equipment with the needed knowledge on flow instability, this study carried out a review of flow instability in parallel channels with water at supercritical pressures.

Effect of water injection on hydrogen generation during severe accident in PWR

Effect of water injection on hydrogen generation during severe accident in a 1000 MWe pressurized water reactor was studied. The analyses were carried out with different water injection rates at different core damage stages. The core can be quenched and accident progression can be terminated by water injection at the time before cohesive core debris is formed at lower core region. Hydrogen generation rate decreases with water injection into the core at the peak core temperature of 1700 K, because the core is quenched and reflooded quickly. The water injection at the peak core temperature of 1900 K, the hydrogen generation rate increases at low injection rates of the water, as the core is quenched slowly and the core remains in uncovered condition at high temperatures for a longer time than the situation of high injection rate. At peak core temperature of 2100–2300 K, the Hydrogen generation rate increases by water injection because of the steam serving to the high temperature steam-starved core. Hydrogen generation rate increases significantly after water injection into the core at peak core temperature of 2500 K because of the steam serving to the relocating Zr-U-O mixture. Almost no hydrogen generation can be seen in base case after formation of the molten pool at the lower core region. However, hydrogen is generated if water is injected into the molten pool, because steam serves to the crust supporting the molten pool. Reactor coolant system (RCS) depressurization by opening power operated relief valves has important effect on hydrogen generation. Special attention should be paid to hydrogen generation enhancement caused by RCS depressurization.

Computational analysis for prediction of pressure of PWR presurizer under transient conditions

A computer model has been developed for prediction of the pressure in the pressurizer under transient conditions. In the model three separate thermodynamic regions which are not required to be in thermal equilibrium have been considered. The mathematical model derived from the general conservation equations includes all of the important thermal-hydraulics phenomena occurring in the pressurizer, i.e., stratification of the hot water and incoming cold water, bulk flashing and condensation, wall condensation, and interfacial heat and mass transfer, etc. The bubble rising and rain-out models are developed to describe bulk flashing and condensation, respectively. To obtain the wall condensation rate, a one-dimensional heat conduction equation is solved by the pivoting method. The presented model will predict the pressure-time behavior of a PWR pressurizer during a variety of transients. The results obtained from the proposed mathematical model are in good agreement with available data on the CHASHMA nuclear power plant’8 pressurizer performance.

乏燃料棒M5锆合金包壳的透射电镜分析

压水堆燃料元件的锆合金包壳,在服役期间会经受高中子注量辐照,其微观组织将发生很大变化,从而影响其宏观性能,因此锆合金包壳的中子辐照行为研究一直是核领域的研究重点。但由于材料经中子辐照后具有较强的放射性,相关的实验必须在热室内进行,因此针对辐照后燃料包壳微观组织的研究也一直是工作的难点。本文在中国原子能科学研究院热室设施上,通过透射电镜分析手段,研究了M5锆合金包壳材料中子辐照后的微观组织。样品来源于国内商业压水堆AFA3G型乏燃料棒,其燃耗分别为14 GW·d/tU和41 GW·d/tU。从燃料棒上截取长度约10 mm的包壳样品,在热室内完成去芯块与化学清洗,获得空包壳样品,然后通过机械制样方法,制备出?3 mm薄片状包壳基体样品,最后采用电解双喷减薄方法,制备出包壳透射电镜观察分析样品。另外,为对比锆包壳辐照后的组织变化,采用同样方法制备了相同材料的冷态观察分析样品。冷态样品与辐照样品的观察分析结果表明:冷态Zr合金包壳基体组织内部存在原生的第二相粒子,基体内部整体较为干净,纳米析出相稀少,未观察到明显的位错结构;辐照后,基体内原生的第二相粒子尺寸和分布与冷态样品差异不明显,但出现了明显的纳米析出相和高密度位错组织;随着燃耗的增加,纳米析出相尺寸有增加的现象;低燃耗与高燃耗样品位错组织具有相似性,表明在14 GW·d/tU燃耗下,锆合金包壳内由辐照产生的位错组织已基本趋于饱和状态;电子选取衍射结果表明,辐照后,基体内原生的第二相粒子虽存在一些非晶组织,但仍以bcc晶体结构为主,表明在41 GW·d/tU燃耗下,第二相粒子保持了一定的辐照稳定性;另外,第二相的EDS结果表明,随着燃耗的增加,Nb元素的含量有贫化趋势;分析认为,Zr合金经中子辐照,第二相粒子中的Nb原子扩展至Zr基体内,将促进Nb元素以纳米富Nb相形式在Zr基体中析出。

国和一号关键核安全技术研发

基于当前各类能源技术基础和核电技术发展水平判断,核能是社会清洁低碳转型与可持续发展的重要支撑,高安全高可靠性的大型先进压水堆核电机组是未来30年内的主力机型。本文围绕大型先进压水堆核电站国家科技重大专项所面临的重大挑战,主要阐述了通过解决“高功率核燃料冷却难”“超高温熔融物滞留难”和“高温高压高放射性包容难”三大关键技术难题,来保证从设计上消除大规模放射性释放可能性或进一步降低核电批量化建设的核安全风险。

压水堆主泵及液态金属泵转子动力学研究进展

核主泵是核电站的关键设备之一,也是反应堆冷却系统的唯一旋转机械设备,其稳定运转对整个反应堆的正常工作至关重要,因此,针对核反应堆主泵开展转子动力学研究,探究主泵转子部件的模态振型、固有频率和支撑系统的刚度阻尼、液膜厚度十分必要。以国内外有关压水堆主泵及液态金属泵的转子动力学研究为重点,围绕压水堆主泵、钠冷快堆主泵、熔盐堆主泵、铅冷快堆主泵4种核主泵类型,从核主泵及其转子部件的结构特点出发,对现阶段主泵导轴承润滑性能和主泵转子结构固有频率、模态分析、临界转速等转子动力学特性的研究进展进行综述和展望,以期对有关核主泵转子动力学特性的计算分析起到一定的借鉴和指导作用。

压水堆滞流分支管热分层现象的数值模拟

在核电系统热力管道内,热分层现象较为常见,会造成应力集中并引起管道结构变形,进而带来安全隐患。滞流分支管与冷却剂主管道相连接,管内流体与一回路主管道冷却剂存在较大温差,受到湍流渗透和阀门泄漏等因素的影响,分支管内容易发生热分层现象。对滞流分支管热分层的温度变化特性和流动特性进行研究分析,为后续的实验研究和应力分析提供理论依据。建立了滞流分支管模型,在泄漏流量为0.062 kg·s^(-1)、泄漏温度为488.15 K、泄漏压力为6 MPa条件下,采用SST k-ω模型(Shear Stress Transport k-ωmodel)对滞流分支管热分层现象展开三维数值模拟研究。模拟结果表明:热分层现象容易出现在水平管段,无保温措施及大管径会加剧热分层现象,而弯管段能有效降低截面温差;滞流分支管的水平管段内存在回流现象,而大小头管段结构导致管内流场出现二次回流,回流现象不利于管道内冷热流体的混合,使热分层的影响时间更长。滞流管分支管的热分层现象与等截面管道存在明显区别。

压水堆燃料元件腐蚀产物沉积行为试验研究进展

压水堆(PWR)燃料元件表面腐蚀产物沉积(CRUD)对系统稳定可靠运行产生诸多不利影响,因此研究压水堆燃料元件腐蚀产物沉积的关键影响因素,并针对性地开发CRUD缓解技术,对压水堆的安全和高质量发展至关重要。归纳了压水堆燃料元件表面CRUD试验研究的重要结果,总结了包壳表面CRUD对系统运行的危害,回顾了CRUD的分析表征结果,介绍了不同研究机构建立的堆外动水回路试验方法,重点分析了材料性质、热工水力特性、冷却剂化学条件及腐蚀产物特性对燃料元件表面CRUD的影响规律,调研了CRUD缓解技术及相关应用,并对后续试验研究方向提出了建议。