Preliminary thermal hydraulic safety analysis of water-cooled ceramic breeder blanket for CFETR

The Chinese fusion engineering test reactor (CFETR) was expected to bridge from the international thermonuclear experimental reactor (ITER) to the demonstration fusion reactor (DEMO). The water-cooled ceramic breeder (WCCB) blanket is one of the blanket candidates for CFETR. In this paper, preliminary thermal hydraulic safety analyses have been carried out using the system safety analysis code RELAP5 originally developed for light water fission reactors. The pulse operation and three typical loss of coolant accidents (LOCAs), namely, in-vessel LOCA, in-box LOCA, and ex-vessel LOCA, were simulated based on steady-state initialization. Simulation results show that important thermal hydraulic parameters, such as pressure and temperature can meet the design criterion which preliminarily verifies the feasibility of the WCCB blanket from the safety point of view.     

Development of a thermal–hydraulic analysis code for the Pebble Bed Water-cooled Reactor

The Pebble Bed Water-cooled Reactor (PBWR) is a water-moderated water-cooled pebble bed reactor in which millions of tristructural-isotropic (TRISO) coated micro-fuel elements (MFE) pile in each assembly. Light water is used as coolant that flows from bottom to top in the assembly while the moderator water flows in the reverse direction out of the assembly.Steady-state thermal–hydraullic analysis code for the PBWR will provide a set of thermal hydraulic parameters of the primary loop so that heat transported out of the core can match with the heat generated by the core for a safe operation of the reactor. The key parameters of the core including the void fraction, pressure drop, heat transfer coefficients, the temperature distribution and the Departure from Nucleate Boiling Ratio (DNBR) is calculated for the core in normal operation. The code can calculate for liquid region, water-steam two phase region and superheated steam region. The results show that the maximum fuel temperature is much lower than the design limitation and the flow distribution can meet the cooling requirement in the reactor core. As a new type of nuclear reactor, the main design features with a sufficient safety margin were also put forward in this paper.     

Safety assessment of a lithium target

This paper addresses the safety assessment of the lithium target of the International Fusion Materials Irradiation Facility (IFMIF) through evaluating the most important risk factors related to system operation and verifying the fulfillment of the safety criteria. The hazard assessment is based on using a well-structured Failure Mode and Effect Analysis (FMEA) procedure by detailing on a component-by-component basis all the possible failure modes and identifying their effects on the plant. Additionally, a systems analysis, applying the fault tree technique, is performed in order to evaluate, from a probabilistic standpoint, all the relevant and possible failures of each component required for safe system operation and assessing the unavailability of the lithium target system. The last task includes the thermal–hydraulic transient analysis of the target lithium loop, including operational and accident transients. A lithium target loop model is developed, using the RELAP5/Mod3.2 thermal–hydraulic code, which has been modified to include specific features of IFMIF itself. The main conclusions are that target safety is fulfilled, the hazards associated with lithium operation are confined within the IFMIF security boundaries, the environmental impact is negligible, and the plant responds to the simulated transients by being able to reach steady conditions in a safety situation.     

A Fail-Safe Reactor Safety System

A highly reliable solid state reactor safety system operated in dynamic mode is described. Direct current signals from detectors for reactor control are converted to pulse signals, which are continuously generated by the circuit until some abnormal condition occurs.

The design and construction of several kinds of logic circuits used for reactor control are explained in detail. The fail-safe property of the circuits have been ascertained by tests, and effects due to temperature variation are negligible in a range 10°~50°C.

Results of reliability analysis show that the reactor safety system described, with redundant dynamic logic circuits, has the necessary qualities to assure highly reliable and safe operation of reactor plants.     

田湾核电厂TXS系统模拟量输入信号交叉比较的设计分析

文章阐述了田湾核电厂TXS系统模拟量输入冗余通道交叉比较标准偏差的数学模型，分析模拟量输入冗余通道的结构设计。运行实践证明：TXS系统模拟量信号交叉比较设计能够较好地完成对模拟量通道的测试和对通道故障趋势的监测任务，系统设计满足核安全法规标准要求。     

Experiment and Verification Test of the Once-through Steam Generator of the 10 MW High-temperature Gas-cooled Reactor Flow Stability of the Once-through Steam Generator

In January 2003, the 10MW High-temperature Gas-cooled Reactor (HTR-10) reached its full power for continuous operation of seventy-two hours in the Institute of Nuclear Energy Technology, Tsinghua University. The reactor operated smoothlyqbthe design parameters were successfully attained.

The once-through steam generator (SG) is one of key equipments of the HTR-10 reactor. The SG includes 30 modular heating helical tube assemblies. There are two thermal hydraulic requirements to be satisfied for the once-through steam generator: (1) enough heat transfer surface; (2) qualified steam can be produced under rated electrical generation power, and water-steam two phase flow un-stability can be avoided. In order to obtain the thermal hydraulic characteristics of the SG reliably, before design, a numerical code was developed for the design, and a full-scale test loop with two heating tubes as model was established, and series experiments had been carried out.

The purpose of this paper is to introduce the design of SG and researches on the stability of small bending radius helical coil-pipe used in HTR-10, for exempla, the effects of outlet steam pressure, inlet water sub-cooling degree, thermal power and inlet throttling degree. Up to now, the SG has experienced full power operation smoothly, and approvingly reached its original design requirements.

In the paper, some operational experimental data of the HTR-10 S.G have been presented.     

核电厂安全系统冗余度研究

本文对核电厂安全系统冗余度的概念进行了澄清,认为不能简单地将安全系列的数量机械地等效于冗余度。N+1的冗余度满足单一故障准则的强制性要求,N+2的冗余度是实现在线维修的可选项。进而介绍了国际上主要核电机型的安全系统配置和冗余度,说明了冗余度与运行灵活性的具体关系。在冗余度研究的基础上,对三环路压水堆的两种安全系统配置方案(两个系列带母管和三个独立系列)进行了分析比较。两种方案均为N+1冗余度,但是对非能动部件(母管)单一故障的考虑有所差异。通过对我国和国际核安全法规、用户要求文件及相关标准的研究发现,非能动部件的单一故障问题不应成为这两个方案选择的决定因素。综合考虑安全性利益及经济性代价,两个系列带母管的方案是更加优化更平衡的设计。     

Designing fault-tolerant real-time computer systems with diversified bus architecture for nuclear power plants

Fault-tolerant real-time computer (FT-RTC) systems are widely used to perform safe operation of nuclear power plants (NPP) and safe shutdown in the event of any untoward situation. Design requirements for such systems need high reliability, availability, computational ability for measurement via sensors, control action via actuators, data communication and human interface via keyboard or display. All these attributes of FT-RTC systems are required to be implemented using best known methods such as redundant system design using diversified bus architecture to avoid common cause failure, fail-safe design to avoid unsafe failure and diagnostic features to validate system operation. In this context, the system designer must select efficient as well as highly reliable diversified bus architecture in order to realize fault-tolerant system design. This paper presents a comparative study between CompactPCI bus and Versa Module Eurocard (VME) bus architecture for designing FT-RTC systems with switch over logic system (SOLS) for NPP.     

200MW核供热站方案设计

200MW核供热示范站反应堆设计中采用了一系列先进技术,如自然循环、一体化布置、自稳压、双层壳结构、控制棒水力驱动系统和非能动式安全系统等,使得供热站更安全、可靠、结构简单、易于建造和维修。本文简要介绍了该站的安全原则、主要设计考虑、总体方案和主要设计特点等。     

PIUS - status and perspectives

Today's environmental concerns show that nuclear energy is an important option for meeting future increases in global energy demand. Significant nuclear expansion will probably require new reactor designs in which safety is ensured by simple, convincing means. PIUS represents such a reactor design. It is a re-configured 600 MWe PWR, in which the primary safety goal, protection of the reactor core integrity, is entrusted to built-in, self-protective, passive features, without reliance on any monitoring, detection or actuation system, nor operator action. Its basic design features a core that is openly connected, in a natural circulation loop, to a large pool of borated water. The pool is enclosed in a prestressed concrete pressure vessel provided with redundant leakage barriers. The reactor coolant pumps are operated in such a way that there is hydraulic balance in the openings between the primary coolant loop and the pool. Thereby, the hot, low boron content primary loop water is kept separated from the pool water, in spite of the always open natural circulation loop. In severe transients this balance is disturbed, and pool water ingress occurs, shutting down the reactor, or restricting the power to a safe level. The decay heat is transferred to the pool by the natural circulation loop, and a passive pool cooling system, utilizing natural circulation and natural draft cooling towers, prevents boiling of the pool water, even in a station blackout situation. Transient analyses have shown that this passive long-term RHR function will be available in all accident situations, even after double-ended cold leg breaks. Such breaks result in a temporary pressurization of the reactor containment, but the releases of radioactivity will be extremely small and the doses at the fence boundary very low. Cost estimates indicate that PIUS will be quite competitive, and evaluation studies are now under way in several countries.     

Testing of the multi-application small light water reactor (MASLWR) passive safety systems

Experimental thermal hydraulic research has been conducted at Oregon State University for the purpose of assessing the performance of a new reactor design concept, the multi-application small light water reactor (MASLWR). The MASLWR is a pressurized light water reactor design with a net output of 35 MWe that uses natural circulation in both normal and transient operation. Due to its small size, portability and modularity, the MASLWR design is well suited to help fill the potential need for grid appropriate reactor designs for smaller electricity grids as may be found in developing or remote regions. The purpose of the OSU MASLWR test facility is to assess the operation of the MASLWR under normal full operating pressure and full temperature conditions and to assess the passive safety systems under transient conditions. The data generated by the testing program will be used to assess computer code calculations and to provide a better understanding of the thermal-hydraulic phenomena in the design of the MASLWR NSSS. During this testing program, four tests were conducted at the OSU MASLWR test facility. These tests included one design basis accident and one beyond design basis accident. During the performance of these tests, plant operations to include start up, normal operation and shut down evolutions were demonstrated successfully.     

European lead fast reactor—ELSY

The conceptual design of the European Lead Fast Reactor is being developed starting from September 2006, in the frame of the EU-FP6-ELSY project. The ELSY (European Lead-cooled System) reference design is a 600 MWe pool-type reactor cooled by pure lead. The ELSY project demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features, while fully complying with the Generation IV goal of sustainability and minor actinide (MA) burning capability. Sustainability was a leading criterion for option selection for core design, focusing on the demonstration of the potential to be self sustaining in plutonium and to burn its own generated MAs. To this end, different core configurations have been studied. Economics was a leading criterion for primary system design and plant layout. The use of a compact and simple primary circuit with the additional objective that all internal components be removable, are among the reactor features intended to assure competitive electric energy generation and long-term investment protection. Low capital cost and construction time are pursued through simplicity and compactness of the reactor building (reduced footprint and height). The reduced plant footprint is one of the benefits coming from the elimination of the Intermediate Cooling System, the low reactor building height is the result of the design approach which foresees the adoption of short-height components and two innovative Decay Heat Removal (DHR) systems. Among the critical issues, the impact of the large mass of lead has been carefully analyzed; it has been demonstrated that the high density of lead can be mitigated by compact solutions and adoption of seismic isolators. Safety has been one of the major focuses all over the ELSY development. In addition to the inherent safety advantages of lead coolant (high boiling point and no exothermic reactions with air or water) a high safety grade of the overall system has been reached. In fact the overall primary system has been conceived in order to minimize pressure drops and, as a consequence, to allow decay heat removal by natural circulation. Moreover two redundant, diverse and passive operated DHR systems have been developed and adopted. The paper presents the overall work performed so far.     

核电机组辅助给水系统水温超运行技术规范限值的改造研究

针对夏季高温天气下，辅助给水系统（ASG）水温超过运行技术规范限值而导致机组后撤的问题，提出了对辅助给水贮水箱（ASG001BA）加装热交换器的改造方案，从工艺设计、仪控修改和运行控制角度进行了详细分析和论证。机组实践表明，在蒸汽发生器冷却正常停堆模式下，本文提出的改造方案保证了ASG001BA的水位和水温在运行技术规范要求的范围内，保证了机组安全经济的运行。本文的研究对机组大修优化、提升机组核安全水平具有参考价值。      

核电站保护屏障的监护神——核电站辐射监测系统(RMS)概述

为确保核电站运行安全,防止核电站正常运行或事故状态下放射性物质泄漏外逸,在核电站的设计和建造中,就考虑到对核电站进行四重保护屏障的设计,而核电站辐射监测系统则是确保四重屏障核安全的重要措施之一。通过对核电站辐射监测系统(RMS)的介绍,使人们对核电站保护屏障的完整性和有效性有一定的了解,对核电安全性的认识进一步提高。     

Protection against fatigue damage with respect to the environmental influence of LWR operating conditions

As nuclear power plant operation proceeds, a certain number of isolated leakages and cracks are found in pipes and components. Considering the extensive design and analysis efforts made, it is at least the early occurrence of the incidents in relation to the design lifetime of the material that is surprising.In the Federal Republic of Germany, a tremendous effort has been directed towards the development of optimized steels with respect to fracture toughness and weldability, the manufacturing of integral forged products (forged-on nozzles, seamless pipes and elbows), and the reduction of peak stresses by an advanced design (basic safety concept). The design of the areas of cold water injection has been improved to prevent temperature stratification. The safety of pipes and components has been considerably improved by these developments. Pipes and components previously affected by material, design and manufacturing problems have been replaced by components built to the new high standard of quality.An area which still deserves attention with respect to the lifetime of a plant is the protection against fatigue damage. The present design fatigue curve (ASME Section III or KTA 3201.2) only allows a generic assessment of the safety against cyclic loadings. In most cases, this concept has proved its worth and is sufficiently safe. In some cases, however, and especially in the presence of a corrosive environment, a detailed investigation is necessary. The design fatigue curve can no longer be applied if cracks form in the protective magnetite layer and, owing to this, strain-induced corrosion cracking occurs.The concept developed in this paper defines a number of correction factors which take into account the effects of a high-temperature water environment as a function of the duration of the strain cycle. It is based on work presented by General Electric. The concept developed allows a more detailed assessment of the local conditions and, consequently, serves as a useful tool for both the designer and the operator of the plant.     

不同初因事件下的两相排放过程热工水力特性研究

本文针对稳压器安全阀开启后的复杂两相热工水力过程进行研究,确定不同初因事件下的稳压器安全阀两相排放特性。采用自主化系统分析程序ARSAC对稳压器安全阀的上下游进行建模分析,选取三种典型的阀门排放过程,包括稳压器安全阀误开启事故、导致一个或多个稳压器安全阀开启的主蒸汽流量完全丧失事故、以及低温超压保护条件下导致的稳压器安全阀间歇性开启的安注泵误启动事故,研究稳压器安全阀开启后水封及蒸汽（或水）排放过程中涉及的复杂两相热工水力特性,结果表明：ARSAC程序能够捕捉两相排放过程中管道内部的流型变化;水封通过下游管道会形成明显的流量峰值,且不同的上游初始条件下排放过程对于下游管道造成的流量峰值及时间特性不同。通过本文的研究可以为载荷分析、安全评价及设计优化提供指导性建议。     

The influence of ocean conditions on thermal-hydraulic characteristics of a passive residual heat removal system

There are many differences between the flow and heat transfer characteristics of nuclear reactors under ocean and land-based conditions for the effects of ocean waves. In this paper, thermal hydraulic characteristics of a passive residual heat removal system (PRHRS) for an integrated pressurized water reactor (IPWR) in ocean environment were investigated theoretically. A series of reasonable theoretical models for a PRHRS in an IPWR were established. These models mainly include the core, once-through steam generator, nitrogen pressurizer, main coolant pump, flow and heat transfer and ocean motion models. The flow and heat transfer models are suitable for the core with plate-type fuel element and the once-through steam generator with annular channel, respectively. A transient analysis code in FORTRAN 90 format has been developed to analyze the thermal–hydraulic characteristics of the PRHRS under ocean conditions. The code was implemented to analyze the effects of different ocean motions on the transient thermal-hydraulic characteristics of PRHRS. It is found that the oscillating amplitudes and periods of the system parameters are determined by those of the ocean motions. The effect of rolling motion is more obvious than that of pitching motion when the amplitudes and periods of rolling and pitching motions are the same. The obtained analysis results are significant to the improvement design of the PRHRS and the safety operation of the IPWR.     

The basic concept to ensure the integrity of components during operation

The integrity of components which are relevant to safety is one of the most important prerequisites for the safe operation of a power plant. A systematic way of proceeding can be achieved by the basic safety concept. In doing so attention has to be paid to the balanced application of the so-called ‘redundant measures’. Using the pressuriser spray system of a nuclear power plant as an example, the relevant factors of influence and the way of proceeding are demonstrated for piping system components which are important to safety.     

200MW低温核供热堆研究进展及产业化发展前景

低温核供热堆技术是我国独立开发的拥有完全自主知识产权的高新技术。200MW壳式核供热堆采用了一体化、自稳压、全功率自然循环、非能动安全系统和水力驱动控制棒等先进技术,具有安全性高、运行可靠、放射性隔离措施完善,可在热用户附近建设等特点。低温核供热堆技术应用领域广泛,其推广应用具有良好的社会效益和经济效益,尤其是核能海水淡化技术的应用,将是解决淡水资源短缺的有效途径之一。本文简要介绍了2200MW低温核供热产业化示范工程的概况、研究进展,总结了核供热堆的主要技术特点,并给出社会经济效益分析和应用前景展望。     

Advantage of redundancy in the controllability of remote handling manipulator

To carry out a variety of remote handling operations inside the ITER divertor a Water Hydraulic MANipulator (WHMAN) and its control system have been designed and developed at Tampere University of Technology. The manipulator is installed on top of Cassette Multifunctional Mover (CMM) to assist during the cassette removal and installation operations. While CMM is designed to carry heavy components such as cassettes through the service ducts relying on positioning accuracy and repeatability, WHMAN is designed to execute a mix of remote handling operations using position trajectories and master-slave telemanipulation. WHMAN is composed of eight joints: six rotational and two translational. Since a manipulator requires only six joints to acquire the desired position and orientation in operational-space, the two additional joints of WHMAN provide the redundant degrees of mobility. This paper presents how this redundancy of WHMAN can be an advantage to optimize the execution of remote handling tasks. The paper also discusses an effective way to practically exploit the redundancy. The results show that the additional degrees of freedom can be utilized to improve the dynamic behavior of the manipulator.