Improvement in understanding of natural circulation phenomena in water cooled nuclear power plants

The IAEA has organized a coordinated research project (CRP) on “Natural Circulation Phenomena, Modelling, and Reliability of Passive Systems That Utilize Natural Circulation.” Specific objectives of CRP were to (i) establish the status of knowledge: reactor start-up and operation, passive system initiation and operation, flow stability, 3-D effects, and scaling laws, (ii) investigate phenomena influencing reliability of passive natural circulation systems, (iii) review experimental databases for the phenomena, (iv) examine the ability of computer codes to predict natural circulation and related phenomena, and (v) apply methodologies for examining the reliability of passive systems. Sixteen institutes from 13 IAEA Member States have participated in this CRP. Twenty reference advanced water cooled reactor designs including evolutionary and innovative designs were selected to examine the use of natural circulation and passive systems in their designs. Twelve phenomena influencing natural circulation were identified and characterized: (1) behaviour in large pools of liquid, (2) effect of non-condensable gases on condensation heat transfer, (3) condensation on the containment structures, (4) behaviour of containment emergency systems, (5) thermo-fluid dynamics and pressure drops in various geometrical configurations, (6) natural circulation in closed loop, (7) steam liquid interaction, (8) gravity driven cooling and accumulator behaviour, (9) liquid temperature stratification, (10) behaviour of emergency heat exchangers and isolation condensers, (11) stratification and mixing of boron, and (12) core make-up tank behaviour. This paper summarizes the achievements within the CRP for the first five phenomena (1-5).     

Accident analysis on supercritical water-cooled reactor during design basis and beyond design-basis condition

Supercritical water-cooled reactor (SCWR) is the only water-cooled reactor among six Generation IV reactor concepts. Safety analysis is one of the most important tasks for SCWR design. A typical thermal spectrum SCWR with passive safety system during design-basis accident (DBA) and beyond design-basis accident (BDBA) is performed. For DBA, reactor system is modeled based on a revised code ATHLET-SC. Loss of coolant accident is chosen to perform safety analysis and sensitive analysis. The results achieved demonstrate the feasibility of proposed passive cooling system to provide sufficient cooling. However, it should be noted that if one of safety systems fails to actuate during loss of coolant accident, although the likelihood is fairly low, there is potential risk of cladding failure. Consequently, the DBA will develop into the BDBA. For BDBA, a postulated severe accident is analyzed after melt pool is formed in the lower plenum. Heat transfer behavior in the melt pool as well as two-dimensional heat transfer effect in the lower head wall is discussed. Then, key parameters are chosen to perform parametric analysis. Results show that the safety margin to critical heat flux is significant. After considering two-dimensional heat conduction effect in the lower head, the safety margin could be further increased.     

Issues affecting advanced passive light-water reactor safety analysis

Next generation commercial reactor designs emphasize enhanced safety by means of improved safety system reliability and performance. These objectives are achieved via safety system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet, the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs will necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing U.S. advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes may require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.     

Prism concept, modular LMR reactors

The Power Reactor Inherently Safe Module (PRISM) is a sodium cooled breeder reactor, conceived by the General Electric Company, which incorporates a number of innovative features with the objective of simultaneously achieving favorable power plant safety and economics. The most important of the innovative features are inherently safe reactor behavior, passive decay heat removal, and small, factory fabricated reactor and equipment modules. The inherently safe reactor behavior is achieved through strong negative reactivity feedbacks and passive, natural circulation decay heat removal. The reactor modules are installed below ground and rest on seismic isolators to mitigate the effects of earthquakes. The standard PRISM power block utilizes three reactor modules supplying steam to a single 415 MWe turbine-generator.     

Thermal hydraulic analysis of SMART for heat removal transients by a secondary system

SMART is an integral type reactor of 330 MW, which enhances its safety by adopting inherent safety design features. Thermal hydraulic characteristics of transients in heat removal by a secondary system for the SMART have been carried out by means of the TASS/SMR and MATRA codes. The primary, secondary, and passive residual heat removal systems RHRS of the SMART were modeled properly. Then, a set of transients for the whole system was investigated. The results of the analyses using the conservative initial and boundary conditions showed that the safety features of the SMART design carried out their functions well and there was a strong moderator temperature coefficient due to the soluble boron free reactor affected by the transient behavior. The natural circulation was well established in the primary and passive residual heat removal systems during the transients and was enough to ensure a stable plant shutdown condition after a reactor trip.     

Experimental study of a passive system for lowering damaging pressure levels in the protective shell of the KLT-40S reactor facility

In the designs of the new-generation reactor facilities, including floating nuclear heat and electricity plants, new passive containment safety systems are used to increase operational safety. The objective of the present work is to validate experimentally the effectiveness of heat removal by a system which lowers damaging pressure levels in the protective shell under conditions of the maximum anticipated accident with loss of coolant in the first loop. The results of the experimental studies on a full-scale model of a secondary loop of the first cooling loop for lowering damaging pressure confirm the validity of the technical decisions made; all prescribed design characteristics of the cooling loop are upheld.     

Development of the FHR advanced natural circulation analysis code and application to FHR safety analysis

The University of California, Berkeley (UCB) is performing thermal hydraulics safety analysis to develop the technical basis for design and licensing of fluoride-salt-cooled, high-temperature reactors (FHRs). FHR designs investigated by UCB use natural circulation for emergency, passive decay heat removal when normal decay heat removal systems fail. The FHR advanced natural circulation analysis (FANCY) code has been developed for assessment of passive decay heat removal capability and safety analysis of these innovative system designs. The FANCY code uses a one-dimensional, semi-implicit scheme to solve for pressure-linked mass, momentum and energy conservation equations. Graph theory is used to automatically generate a staggered mesh for complicated pipe network systems. Heat structure models have been implemented for three types of boundary conditions (Dirichlet, Neumann and Robin boundary conditions). Heat structures can be composed of several layers of different materials, and are used for simulation of heat structure temperature distribution and heat transfer rate. Control models are used to simulate sequences of events or trips of safety systems. A proportional-integral controller is also used to automatically make thermal hydraulic systems reach desired steady state conditions. A point kinetics model is used to model reactor kinetics behavior with temperature reactivity feedback. The underlying large sparse linear systems in these models are efficiently solved by using direct and iterative solvers provided by the SuperLU code on high performance machines. Input interfaces are designed to increase the flexibility of simulation for complicated thermal hydraulic systems. This paper mainly focuses on the methodology used to develop the FANCY code, and safety analysis of the Mark 1 pebble-bed FHR under development at UCB is performed.     

Addressing the challenges posed by advanced reactor passive safety system performance assessment

Passive safety features play an essential role in the development of nuclear technology and within advanced water cooled reactor designs. The assessment of the reliability of such systems in the frame of plant safety and risk studies is still an open issue. This complexity stems from a variety of open points coming out from the efforts conducted so far to address the topic and concern, for instance, the amount of uncertainties affecting the system performance evaluation, including the uncertainties related to the thermal-hydraulic (T-H) codes, as well as the integration within an accident sequence in combination with active systems and human actions. These concerns should be addressed and conveniently worked out, since it is the major goal of the international community (e.g. IAEA) to strive to harmonize the different proposed approaches and to reach a common consensus, in order to add credit to the underlying models and the eventual out coming reliability figures. The main key points that may influence the reliability analysis are presented and discussed and a viable path towards the implementation of the research efforts is delineated, with focus on T-H passive systems.     

Neutronic and thermo-hydraulic analyses of a small,long-life HTGR for passive decay-heat removal

Since the accident at Fukushima Daiichi Nuclear Power Plant in 2011, design concepts for nuclear reactors have been reconsidered with much greater emphasis placed upon passive systems for decay-heat removal. By considering this issue, the design parameter conditions for high temperature gas-cooled reactors (HTGRs) with passive safety features of decay-heat removal were obtained by residual-heat transfer calculation using equations for fundamental heat transfer mechanisms in our previous works. In the present study, the appropriate size of reactor core for a 100 MW_t reactor operating at 1123 K of the initial core temperature was found using the conditions. Consequently, neutronics and thermo-hydraulic analyses for the proposed reactor core were performed and the proper optimizations to control the excess reactivity and flatten the change in power peaking factor during operation were done successfully. By the systematic method to decide the core design which satisfies the condition for passive decay-heat removal, a long-life small HTGR concept whose excess reactivity was small during the operation was shown. The small excess reactivity is a significant advantage from the view point of safety in reactivity accident.     

Safety features of the gas-cooled fast breeder reactor

The safety features of the gas-cooled fast breeder reactor (GCFR) are described in the context of the 300-MW(e) demonstration plant design. They are of two general types, inherent and design-related. The inherent features are principally associated with the helium coolant and the nuclear coefficients. Design-related features influencing safety include shutdown systems, residual heat removal systems, method of core support, and the prestressed concrete reactor vessel (PCRV). This paper discusses the safety-related aspects of each of these. Recently completed residual heat removal system reliability studies are also discussed. The probability of residual heat removal system failure in the GCFR is found to be lower than that described for light water reactors. The safety characteristics of larger plants are examined, and increases in size are found to improve GCFR safety margins.     

Safety approach of DFBR design study in Japan

The goal of the safety design for the demonstration fast breeder reactor is to ensure that the safety level is equivalent to or higher than that of the light water reactors of the same period. The design of the safety features such as reactor shutdown, decay heat removal and confinement systems is of importance to reach the goal. The reactor core is equipped with two independent fast shutdown systems, the primary system and the backup system. In addition, it is planned to strengthen the passive shutdown capability by using self- actuated systems such as a Curie point device for the backup system. The decay heat is removed from the core to the atmosphere through the safety lines of the direct reactor auxiliary cooling system which is composed of four independent lines. Furthermore, under the severe conditions that no active function of the decay heat removal system is available, the heat can be removed by natural convection through the safety lines by taking advantage of the high boiling temperature of sodium. For the confinement function, the reactor vessel is surrounded by a containment vessel and a confinement area.

The design concept of these safety features is described in this paper.     

Flexible conversion ratio fast reactors: Overview

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. The performance achievable by the unity conversion ratio cores of these reactors was compared to an existing supercritical carbon dioxide-cooled (S-CO₂) fast reactor design and an uprated version of an existing sodium-cooled fast reactor. All concepts have cores rated at 2400 MWt. The cores of the liquid-cooled reactors are placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchangers (IHXs) coupling a primary coolant to a compact and efficient supercritical CO₂ Brayton cycle power conversion system. The S-CO₂ reactor is directly coupled to the S-CO₂ Brayton cycle power conversion system. Decay heat is removed passively using an enhanced reactor vessel auxiliary cooling system (RVACS) and a passive secondary auxiliary cooling system (PSACS). The selection of the water-cooled versus air-cooled heat sink for the PSACS as well as the analysis of the probability that the PSACS may fail to complete its mission was performed using risk-informed methodology. In addition to these features, all reactors were designed to be self-controllable. Further, the liquid-cooled reactors utilized common passive decay heat removal systems whereas the S-CO₂ uses reliable battery powered blowers for post-LOCA decay heat removal to provide flow in well defined regimes and to accommodate inadvertent bypass flows. The multiple design limits and challenges which constrained the execution of the four fast reactor concepts are elaborated. These include principally neutronics and materials challenges. The neutronic challenges are the large positive coolant reactivity feedback, small fuel temperature coefficient, small effective delayed neutron fraction, large reactivity swing and the transition between different conversion ratio cores. The burnup, temperature and fluence constraints on fuels, cladding and vessel materials are elaborated for three categories of material - materials currently available, available on a relatively short time scale and available only with significant development effort. The selected fuels are the metallic U-TRU-Zr (10% Zr) for unity conversion ratio and TRU-Zr (75% Zr) for zero conversion ratio. The principal selected cladding and vessel materials are HT-9 and A533 or A508, respectively, for current availability, T-91 and 9Cr-1Mo steel for relatively short-term availability and oxide dispersion strengthened ferritic steel (ODS) available only with significant development.     

Fast reactor decay heat removal: approach to the safety system design in Japan and Europe

In the framework of the cooperation on fast reactor between the European and Japanese electrical utilities, the design companies responsible for the demonstration fast breeder reactor (DFBR) in Japan and the European fast reactor (EFR) have performed a comparative evaluation of the safety qualified decay heat removal systems of the two reactor designs. At the level of overall safety and concept design there is an obvious similarity between the two DHR systems. In both cases heat is removed directly from the reactor vessel primary sodium by systems designed according to a similar deterministic methodology, with a probabilistic assessment performed to demonstrate achievement of the required reliability. Nevertheless, the evaluation revealed a number of differences resulting from different national practices. These include the application of diversity and redundancy philosophy, the extent of passivity taken into account, the consequences of postulated maintenance outage on the design and the decay heat curve.     

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。     

Western and Asian light water reactors at the millennium threshold

This paper presents the most advanced Western and Asian light water reactor (LWR) designs. The following pressurized water reactor (PWR) and boiling water reactor (BWR) designers are covered: Westinghouse ( ), Babcock and Wilcox (B&W—now part of Framatome), Combustion Engineering (CE—now ABB CE), Siemens (PWR), Framatome, Mitsubishi, General Electric (GE), Asea Brown Boveri (ABB), Siemens (BWR), Hitachi and Toshiba. The motivations that led to the design of the next generation of LWRs are discussed. The technical bases for evolutionary and innovative plants are summarized. Important safety features of some of the most complete (in operation, under construction or certified) evolutionary designs are described detail. Analogous implementations of systems into other advanced designs are given.     

我国核供热堆的设计特性和安全概念

根据核供热应用的特点和先进反应堆的发展目标，我国的核供热堆采用新的安全原理和一系列先进技术，其中包括一体化布置、全功率自然循环、自稳压、控制棒动压水力驱动和非能动安全系统等，从而使其达到更高的安全标准，同时做到核供热站系统简化和经济上有竞争力。主要论述核供热堆设计应考虑的主要问题、设计特点和安全概念。还给出一些主要的试验和分析研究结果，以验证核供热堆的安全特性。     

Renewed interest in lead cooled fast reactors

In the last few years a number of compact designs of lead-alloy cooled systems have been promoted. Moreover, in Russia a design effort was started earlier on the pure lead-cooled BREST reactor but this effort does not appear to be strongly funded any more. But now the lead cooled and compact STAR-LM reactor is promoted in the US and in the European Union there is some interest in a mediumsized lead-cooled fast reactor (LFR). It has brought some nuclear industries, a large utility, several research centers and universities together to ask the European Commission for a partial funding of design and safety efforts. A 600 MWe LFR design is proposed which would be useful for base load operation but as a fast system it could also be used for load following. Because of the possible plant simplifications and the use of pure lead, the economics of such a system should be good. Moreover, efficient fuel utilization, the burning of higher actinides and a closed fuel cycle make it a sustainable system. Whether, this larger system has the same inherent / passive safety characteristics as smaller LFRs needs to be examined. In this paper the passive emergency decay heat removal by reactor vessel aircooling of such a larger system is investigated. Moreover an inlet blockage in a subassembly of a low power density LMR is analyzed. Furthermore, the pros and cons of lead vs. lead/bismuth coolants are discussed.     

AP1000非能动余热排出系统瞬态工况分析

非能动余热排出系统（PRHR）作为AP1000非LOCA情况下带走堆芯热量的安全手段,其设备可靠性对电厂安全和经济性极为重要,文章主要介绍PRHR结构上的薄弱部分和在整个寿期的瞬态发生频度,分析了温度瞬态、流量瞬态等情况,为电厂的运行、维修和役检提供参考。     

Analysis of loss of coolant accident in MTR pool type research reactor

In MTR research reactors, heat removal is, safely performed by forced convection during normal operation and by natural convection after reactor shutdown for residual decay heat removal. However, according to the duration time of operation at full power, it may be required to maintain the forced convection, for a certain period of time after the reactor shutdown. This is among the general requirements for the overall safety engineering features of MTR research reactors to ensure a safe residual heat removal. For instance, in safety analysis of research reactors, initiating events that may challenge the safe removal of residual heat must be identified and analyzed.In the present work, it was assumed a total loss of coolant accident in a typical MTR nuclear research reactor with the objective of examining the core behavior and the occurrence of any fuel damage.For this purpose, the IAEA 10 MW benchmark core, which is a representative of medium power pool type MTR research reactors, was chosen herein in order to investigate the evolution of cladding temperature through the use of a best estimate thermalhydraulic system code RELAP5/mod3.2.     

直接注入管线失水事故非能动安全系统运行特性研究

小型模块式反应堆ACP100采用了非能动安全和模块化设计技术,可用于地区集中供暖、海水淡化和核动力商船等多个方面。其中,非能动安全设计主要包括非能动应急堆芯冷却系统、非能动余热排出系统等非能动安全系统和自动卸压等专设措施。针对ACP100非能动安全设计技术特点,在中国核动力研究设计院非能动安全系统综合性能缩比试验装置上开展了大量失水事故系统特性试验研究,根据试验数据分析,获得了非能动安全系统在直接注入管线发生破口后系统的综合响应特性,掌握了系统间的相互影响规律,并初步评估其对堆芯的冷却效果。