Design approach of the ARIES-AT power core and vacuum vessel cost assessment

The complexity of fusion power plants require the integration of many diverse and important system requirements to achieve a design approach that is viewed as a commercially viable electric plant. The ARIES-AT power core design builds upon a history of fusion power core designs that evolve along with physics and engineering advances. The baseline design point is optimized for maximum performance and minimum capital cost based upon the ARIES systems code results, along with physics and engineering analyses. The ARIES-AT power core is designed to be quick and easily maintainable to achieve high plant availability. A key element to achieve the high availability is the integration of the core elements with the design of the vacuum vessel. The vacuum vessel design is developed in more detail to assure the key assembly and maintenance features could be realized at an affordable cost.     

Degradation assessment of structures and passive components at nuclear power plants

This paper describes a multi-year research program to assess age-related degradation of structures and passive components important to the safe operation of nuclear power plants (NPPs). The purpose of the research effort is to develop the technical basis for the validation and improvement of analytical methods and acceptance criteria which can be used to make risk-informed decisions and to address technical issues related to degradation of structures and passive components. The approach adopted for this research program consists of two phases. In Phase I, specific degradation occurrences at plants were collected and evaluated, existing technical information on aging was reviewed, and a scoping study was performed to identify which structures and components should be studied in the subsequent phases of the research program. Based on the results of the Phase I effort, selected structures and passive components are evaluated in Phase II to assess the effects of age-related degradation using existing and enhanced analytical methods. Fragility analyses are performed for undegraded and degraded structures and passive components. These results can then be used to assess the potential impact of degradation on overall plant risk. The Phase II effort also utilizes the results of the analyses to develop probabilistic degradation acceptance criteria for the structures and passive components studied. These research activities provide useful tools to support the current goals of developing risk-informed and performance-based regulation in the nuclear industry.     

Potential application of NRC aging research to license renewal

A strong technical base, when developed and implemented, to manage aging in plant safety related systems, support systems, structures, and components will give confidence to all of us in the nuclear community with regard to our ability to maintain continuous safe operation of nuclear power plants of all ages. This technical base for managing aging must be built over the foundation of reviews and analyses of original designs, operating experience of over 20 years, experts' opinions, and the utilization of research results. The key elements of managing aging are the understanding of risk significance of aging phenomena and the licensee program(s) for inspection, surveillance, condition monitoring, trending, record-keeping and maintenance to mitigate the influence and effects of aging.An overview is provided of the intended application of plant aging research results. In the process it delineates some key steps recommended for the development of appropriate regulatory guides and review procedures involving the technical issues for license renewal considerations. This application of aging research results is based primarily upon the NPAR theme of “Understanding Aging-A Key to Ensuring Safety, and Managing Aging-A Necessity to Ensuring Safety.”     

A systematic methodology approach for selecting preferable and alternative sites for the first NPP project in Yemen

The purpose of this paper is to present the results of the first siting study that was performed to investigate the availability of sites for a 1000 MWt nuclear power plant (NPP) project in Yemen. Only about thirty eight percent of the population in Yemen has access to electric power. This low access has negative impacts to quality of life as well as broad economic implications. Although the political and security situation in Yemen is still unstable, the government of Yemen looks forward to the future and has been developing plans to replace aging electric power infrastructure, to increase the grid capacity and to possibly add a nuclear power plant (NPP) of the 1000 MWt class to its current energy mix. This paper presents the results of a siting study for NPP in Yemen suggesting a PWR type (Pressurized Water Reactor). From eight coastal provinces twelve potential sites were proposed and then evaluated against nine avoidance criteria to screen out some candidate sites, then these sites evaluated again under other nine suitability criteria to choose the preferable and alternative sites. To perform this study in systematic approach using the Analytic Hierarchy Process (AHP) was applied to support the decision making analysis. Results showed that the southern coast of Yemen has the capability to accommodate nuclear power plants due to its large area and low population density. The siting study recommends three high preference sites and another three acceptable alternative sites.     

Maintenance duration estimate for a DEMO fusion power plant,based on the EFDA WP12 pre-conceptual studies

The erosion and high neutron flux in a fusion power plant results in the need for frequent remote replacement of the plasma facing components. This is a complex and time consuming remote handling operation and its duration directly affects the availability and therefore the commercial viability of the power plant.A tool is needed to allow the maintenance duration to be determined so that developments in component design can be assessed in terms of their effect on the maintenance duration. This allows the correct balance to be drawn between component cost and performance on the one hand and the remote handling cost and plant availability on the other.The work to develop this tool has begun with an estimate of the maintenance duration for a fusion power plant based on the EFDA DEMO WP12 pre-conceptual design studies [1]. The estimate can be readily adjusted for changes to the remote maintenance process resulting from design changes. The estimate uses data extrapolated from recorded times and operational experience from remote maintenance activities on the JET tokamak and other nuclear facilities.The Power Plant Conceptual Study from 2005 [2] proposes that commercial viability of a power plant would require an availability of 75% or above. Results from the maintenance estimate described in this paper suggest that this level of availability could be achieved for the planned maintenance using a highly developed and tested remote maintenance system, with a large element of parallel working and challenging but feasible operation times.     

Incorporation of passive components aging into PRAs

The probabilistic risk assessments being developed at most nuclear power plants to calculate the risk of core damage generally focus on the possible failure of active components. The possible failure of passive components is given little consideration. We are developing a method for selecting risk-significant passive components and including them in probabilistic risk assessments. We demonstrated the method by selecting a weld in the auxiliary feedwater system. The selection of this component was based on expert judgement of the likelihood of failure and on an estimate of the consequence of component failure to plant safety. We then used the PRAISE computer code to perform a probabilistic structural analysis to calculate the probability that crack growth due to aging would cause the weld to fail. The calculation included the effects of mechanical loads and thermal transients considered in the design and the effects of thermal cycling caused by a leaking check valve. We modified an existing probabilistic risk assessment (NUREG-1150 plant) to include the possible failure of the auxiliary feedwater weld, and then we used the weld failure probability as input to the modified probabilistic risk assessment to calculate the change in plant risk with time. The results showed that if the failure probability of the selected weld is high, the effect on plant risk is significant. However, this particular calculation showed a very low weld failure probability and no change in plant risk for the 48 years of service analyzed. The success of this demonstration shows that this method could be applied to nuclear power plants.     

Understanding aging in containment cooling systems

A study has been performed to assess the effects of aging in nuclear power plant containment cooling systems. Failure records from national databases as well as plant-specific data were reviewed and analyzed to identify aging characteristics for this system. The predominant aging mechanisms were determined, along with the most frequently failed components and their associated failure modes. This paper discusses the aging mechanisms present in the containment spray system and the containment fan cooler system, which are two systems used to provide the containment cooling function. The failure modes, along with the relative frequency of each, are also discussed.     

ARIES-AT maintenance system definition and analysis

A fusion power plant must have a high availability to be competitive in the electrical generation market. Attaining high plant availability is difficult because the fusion power core has a limited service lifetime. Moreover, the core components are radioactive and very large. To assess these issues, the maintainability of the ARIES fusion power core is analyzed and integrated into the early power core design process, which results in a maintainability approach capable of attaining a relatively short refurbishment time. The developed timelines are presented for the scheduled maintenance of the power core. The short core refurbishment time coupled, with evolutionary improvements in the maintainability of the reactor plant equipment and the balance-of-plant equipment, infer an attractive plant availability in the range of 90%.     

Neutral beam deployment on DEMO and its influence on design

The demands on the neutral beam heating and current drive system of a DEMO device exceed those of existing fusion experiments by several orders of magnitude. By predicting possible power waveforms it is possible to analyse the technological advances necessary to achieve a system relevant to deployment on a power plant. Achieving the necessary efficiency will require simultaneous improvements in beam current density, neutralization efficiency and beam transmission. Considering the deployment on the tokamak vessel shows no major disruption to the tritium breeder blanket and no requirement to reach a high packing density of injectors. The thermal management of components subjected to low heat flux for many hours is considered and it is shown that radiation cooling can be exploited to control the temperature of such items.     

如何发现和解决大亚湾和岭澳核电厂存在的加速老化现象

从振动、腐蚀和有机材料材质三大方面对国内外核电厂设备加速老化现象进行了阐述。根据大亚湾和岭澳核电厂的特点,提出了目前核电厂常规岛可能存在的加速老化问题和主要影响因素,并对如何发现和准确评估已存在的老化现象,以及如何分析和研究加速老化现象和具体的解决措施进行了详细介绍。     

秦山核电厂的老化及寿期管理

介绍了核电厂老化及寿期管理的相关背景以及国外核电厂在延寿方面采取的两种主要模式,即执照更新模式和PSR模式。结合目前秦山核电厂开展的主要老化管理工作,提出了秦山核电厂延寿的设想,并对核电厂寿期管理中存在的问题进行了讨论。     

Aging management of instrumentation & control sensors in nuclear power plants

Pressure to improve plant efficiency and maximize safety and the increasing age of existing NPPs are forcing the global nuclear power industry to confront the challenges of aging - caused by stressors such as temperature, humidity, radiation, electricity, and vibration - in key instrument & control (I&C) components like pressure transmitters, temperature sensors, neutron detectors, and cables. Traditional aging management methods, such as equipment replacement, required the process to be shut down. Recent aging management technologies, collectively known as online monitoring (OLM), enable plants to monitor the condition and aging of their installed I&C while the plant is operating. Developed through R&D initiatives worldwide, such OLM techniques include low- and high-frequency methods that use existing sensors, such as noise analysis; methods based on test or diagnostic sensors, such as for vibration-measuring accelerometers; and methods, such as the power interrupt (PI) test, based on active measurements made by injecting a test signal into the component under test. A review of these aging management methods, their effectiveness, and their interrelation provides a foundation for understanding the next stage in the evolution of OLM: truly integrated hybrid OLM systems capable of robust condition monitoring in both novel and familiar operating conditions.     

Risk-Based,genetic algorithm approach to optimize outage maintenance schedule

A huge number of components are typically scheduled for maintenance when a nuclear power plant is shut down for its planned outage. Among these components, a number of them are risk significant so that their operability as well as reliability is of prime concern. Lack of proper maintenance for such components during the outage would impose substantial risk on the nuclear power plant (NPP) operation.     

PRAAGE88: An interactive IBM-PC code for aging analysis of NUREG-1150 systems

The effects of time dependent failure rates caused by the aging of components are becoming increasingly important in probabilistic risk assessment (PRA) and reliability analyses of nuclear power plant systems. In the NRC Nuclear Plant Aging Research (NPAR) program, the effects of aging in nuclear systems are being evaluated through the use of time varying failure rates that are determined as a function of the age of the system. These analyses involve complex systems and include various sensitivity studies; thus, the PRAAGE88 computer code was developed to facilitate these calculations. PRAAGE88 is an IBM PC based code that computes system unavailability, component unavailability, and various importance measures for use in evaluating the effect of aging on reactor systems. This paper describes the methodology utilized in the code, its capabilities and areas of application.     

Significance of material technology for plant life management in nuclear power plants

The life-limiting mechanisms for components and systems are physical aging and wear. Both of them are related to changes of microstructure in the bulk material or at the phase boundaries medium/material and material/material. They are triggered during operation by factors such as temperature, mechanical load, and environment. Thus, to achieve an utmost effective aging management it is necessary, to understand the underlying aging and wear mechanisms such as neutron irradiation, fatigue, corrosion, fretting, etc. Definition and qualification of suitable corrective and preventive actions against accelerated aging, requires precise knowledge of the aging processes and life-limiting situations and thresholds. It is obvious, then, that materials engineering plays a large part in effective and economical plant life management. Within this paper, the role of materials science and technology in plant aging management during the various stages within a whole life cycle of a power plant is described: (1) the correct choice of materials as part of a well-based materials concept in the design stage is very important for later plant operation. As an example steam generator materials are presented. (2) The parameters of the individual manufacturing processes during erection of components and systems must be optimally selected in order to guarantee long-term operation. As an example the reasons for core shroud cracking in a BWR NPP are discussed. (3) Aging mechanisms must be accounted for in operation of components and systems, and their effects have to be counteracted in order to prevent service-life limiting situations. Details are described with respect of corrosion and neutron irradiation. Demanding future tasks for materials science and technology are presented, which are necessary to continue to contribute to an optimized plant life management and to cost-effective operation of nuclear power plants at high safety levels.     

Quantitative modeling of digital reactor protection system using Markov state-transition model

Recently, digital instrumentation and control systems have been increasingly installed for important safety functions in nuclear power plants such as the reactor protection system (RPS) and the actuation system of the engineered safety features. Since digital devices consist of not only electronic hardware but also software that can control microprocessors, the functions specific to digital equipment such as self-diagnostic functions have been becoming available. These functions were not realized with conventional electric components. On the other hand, it has been found that it is difficult to model the digital equipment reliability in probabilistic risk assessment (PRA) using conventional fault tree analysis technique. OECD/NEA CSNI Working Group of Risk Assessment (WGRisk) set up the task group DIGREL to develop the basis of reliability analysis method of the digital safety system and is now discussing about several issues including quantitative dynamic modeling. This paper shows that, taking account of the relationship among the RPS failures, demand after the initiating event, detection of the RPS fault by self-diagnostic or surveillance tests, repair of the RPS components and plant shutdown operation by the plant operators as a stochastic process, the anticipated transient without scram (ATWS) event can be modeled by the event logic fault tree and Markov state-transition diagrams assuming the hypothetical 1-out-of-2 digital RPS.     

Summary and conclusions of a program addressing aging of nuclear power plant concrete structures

Research has been conducted by the Oak Ridge National Laboratory to address aging management of nuclear power plant concrete structures. The purpose was to identify potential structural safety issues and acceptance criteria for use in continued service assessments. The focus of this program was on structural integrity rather than on leaktightness or pressure retention of concrete structures. Primary program accomplishments include formulation of a Structural Materials Information Center that contains data and information on the time variation of material properties under the influence of pertinent environmental stressors and aging factors for 144 materials, an aging assessment methodology to identify critical structures and degradation factors that can potentially impact their performance, guidelines and evaluation criteria for use in condition assessments of reinforced concrete structures, and a reliability-based methodology for current condition assessments and estimations of future performance of reinforced concrete nuclear power plant structures. In addition, in-depth evaluations were conducted of several nondestructive evaluation and repair-related technologies to develop guidance on their applicability.     

进口民用核安全设备安全检验工作中的设备审查

在进口核安全设备的安全检验中,对这些设备进行审查可以避免有缺陷的或不能证明其满足相关标准的设备用于我国核电厂,从而保障核电厂的安全运行。本文介绍了安检工作的目的、流程、内容和审查范围,重点介绍了对安检工作中设备文件的审查依据,提出了安检工作中设备文件的审查要点。     

Development of novel tungsten processing technologies for electro-chemical machining (ECM) of plasma facing components

Plasma facing components for fusion applications must exhibit long-term stability under extreme conditions, and therefore material imperfections cannot be tolerated due to a high risk of technical failures. To prevent or abolish defects in refractory metals components during the manufacturing process, some methods of electro-chemical machining as S-ECM and C-ECM were developed, enabling both the processing of smooth plain defect-free surfaces of different geometry and the removal of bulk material for the shaping of three-dimensional structures, also without cracks. It is discussed, that tungsten ablation with accurate electro-chemical molding is very sensitive to the kind of electric current, and therefore current investigations focused also on the effects of frequency profiles on the sharpness of edge rounding.     

Seismic fragilities for nuclear power plant risk studies

Seismic fragilities of critical structures and equipment are developed as families of conditional failure frequency curves plotted against peak ground acceleration. The procedure is based on available data combined with judicious extrapolation of design information on plant structures and equipment. Representative values of fragility parameters for typical modern nuclear power plants are provided. Based on the fragility evaluation for about a dozen nuclear power plants, it is proposed that unnecessary conservatism existing in current seismic design practice could be removed by properly accounting for inelastic energy absorption capabilities of structures. The paper discusses the key contributors to seismic risk and the significance of possible correlation between component failures and potential design and construction errors.