Intelligent system for transient data collection and fatigue monitoring of pressurised water reactors nuclear steam supply system

Electricité de France (EDF), the French national electricity company, is operating 54 standardised pressurised water reactors. This about 500 reactor-years experience in nuclear stations operation and maintenance area has allowed EDF to develop its own strategy for monitoring of age-related degradations of NPP systems and components relevant for plant safety and reliability. After more than fifteen years of experience in regulatory transient data collection and seven years of successful fatigue monitoring prototypes experimentation, EDF decided to design a new system called SYSFAC (acronym for SYstème de Surveillance en FAtigue de la Chaudière) devoted to transient logging and thermal fatigue monitoring of the reactor coolant pressure boundary. The system is fully automatic and directly connected to the on-site data acquisition network without any complementary instrumentation. A functional transient detection module and a mechanical transient detection module are in charge of the general transient data collection. A fatigue monitoring module is aimed towards a precise surveillance of five specific zones particularly sensible to thermal fatigue. After a first step of preliminary studies, the industrial phase of the SYSFAC project is currently going on, with hardware and software tests and implementation. The first SYSFAC system will be delivered to the pilot power plant by the beginning of 1996. The extension to all EDF’s nuclear 900 MW is planned after one more year of feedback experience.     

Residual life assessment, technical improvements and backfitting of components and systems from the materials standpoint

The safety standard which is attained at the time of initial startup by a nuclear power plant built in accordance with state-of-the-art design and engineering principles must be assured throughout the plant's entire lifetime. Based on operating experience a plant's original design life should be systematically reevaluated in the light of new findings and developments in safety. The results of an analysis of this type can also be useful for the purposes of preventive maintenance or to prolong a plant's planned or licensed service life. Moreover, they form the basis of decisions regarding technical upgrades and backfits and are of value in optimizing plant reliability and availability.A concept exists for components in light water reactors which provides for prompt identification and remedying of damage due to the deterioration in service of materials and their properties (e.g. fatigue, local corrosion mechanisms, erosion corrosion, neutron irradiation). It may be necessary or appropriate to implement technical upgrades or backfits where application of more up-to-date safety standards (e.g. break preclusion methodology) demands compliance with more stringent requirements than those originally used in the design basis, or where systematic damage cannot be ruled out (unstabilized austenitic steels in BWR plants), or where it is possible to make substantial reductions in the radiation exposure of operating or maintenance personnel (substitution of cobalt-containing materials) during the plant's lifetime.     

Apparative developments for inservice inspections of reactor pressure vessels

Inservice inspections of primary circuit components are important preventive measures to guarantee nuclear power plant integrity, satisfying simultaneously reactor safety and economy in plant operation. Emphasizing pressurized water reactor pressure vessel (RPV) inspections, recent developments of new generations of automated and mechanized ultrasonic inspection equipment are presented. Starting from general equipment design and inservice implementation criteria, specific examples are given. Main attention is directed to equipment realization of phased array and ALOK inspection techniques, especially in their combination. Refined aspects of subsequent computer processing and evaluation of defect detection data are described. Analytical features and potential for further developments become evident. Remote controlled RPV inspections are stressed by describing a new generation of central mast manipulators, forming an integral part of total inservice inspection system.     

新一代CANDU—9重水堆核电厂简介

ＣＡＮＤＵ－９是电功率为９００ＭＷ级的重水堆核电厂,其设计基于达灵顿和布鲁斯Ｂ多机组核电厂,并融入了一些最新的工程设计和研究成果,除了继续采用成熟的系统和部件外,在安全性,地可靠性和可维护性方面作了重要改进。ＣＡＮＤＵ－９综合考虑了安全审评和执照申请过程中发现的问题,产使其体现在安全设计理念中,特别是对慢化剂系统,端屏蔽冷却系统,系统和应急堆芯冷却系统进行了改进。     

The safety of the HTR 500

Comprehensive construction and operating experience is available in the High-Temperature Reactors constructed by the BBC group, the AVR experimental nuclear power plant and the THTR 300 MW in Hamm-Uentrop, being in its commissioning phase. The main objective pursued in the planning and design of the HTR 500 was to create a marketable and commercial plant, maintaining at the same time high safety standards corresponding to the current state-of-the-art. Thus full use is made of the system-specific safety characteristics of the HTR. In addition, the great number of probabilistic analyses likewise confirming the extremely low risk of the HTR have been used as a basis of the HTR 500 safety concept.     

Use of plant specific data in risk studies

It is shown how the operating experience with specific components in a specific plant can be used to modify generic distributions of failure rates by using Bayes' theorem. These component-specific distributions of failure rates are then employed to evaluate the distribution of the (time averaged) failure probability of the specific safety system (emergency feedwater system in the Gemeinschaftskernkraftwerk Neckar). The calculation shows that even the limited operating experience with specific components in specific plants leads to an appreciable modification of the distribution of the failure probability.     

Proactive approaches to assure the structural integrity of Boiling Water Reactor components

Pressure vessel components in operating Boiling Water Reactor (BWR) plants are subjected to a variety of loading and environmental conditions which could lead to degradation over time. The significant damage mechanisms such as fatigue, stress corrosion cracking (SCC) and irradiation embrittlement are considered in the design basis of the reactor components and thus provide adequate structural margins over the operating life of the plant. Nevertheless, when the design basis assumptions are exceeded, e.g., thermal cycles, vibratory loading or chemistry transients, cracking may occur in pressure boundary components. Several proactive measures are being implemented to address this concern and assure the structural margins in BWR plants. These measures include: (i) control of materials and design to mitigate SCC and improvement of the environmental conditions through the implementation of Hydrogen Water Chemistry, (ii) advances in automated ultrasonic inspection of the BWR pressure vessel and piping, (iii) improved monitoring techniques for tracking fatigue usage and SCC effects in the piping and in the core, and (iv) development and qualification of durable repairs and specialized techniques such as use of high purity materials and temper bead repair. This paper describes current progress in implementing these proactive approaches for Boiling Water Reactors.     

The evolution of the structure and application of U.S. NRC regulations and standards

NRC regulations and standards and their implementation have evolved from early adaptations of conventional engineering practices to a mature, cohesive set of regulations that govern NRC regulation of nuclear power plant safety in the United States.From a simple set of rules and design criteria and from the standards of the professional engineering societies, a hierarchy of practices, standards, guides, rules and goals has developed. Resting on a foundation of industrial practices, this hierarchy rises through levels of national standards, regulatory guides and standard review plans, policy statements and NRC regulations.The licensing process is evolving today toward one that permits both site approval and standard design certification before the plant is constructed. At the present time, NRC is reviewing five standard designs for certification for a period of 15 years. NRC focuses its regulation of operating nuclear plants on inspections conducted from five regional offices. Resident inspectors, specialist inspectors, and multi-disciplinary inspection teams examine specific plant situations. The results of all these inspections are used to develop a complete understanding of a plant's physical condition, its operation, maintenance and management.To improve safe operation of nuclear plants in the U.S., a most important program, the Systematic Assessment of Licensee Performance, measures operational performance, using a broad spectrum of functional areas.     

Probabilistic Service-Life Assessment for Safety Analysis for Preparing a Power-Generating Unit of a Nuclear Power Plant for Service-Life Extension

Information about the actual history of cyclic loading in the actual state of the metal in equipment components must be used for analyzing the safety of power-generating units in operating nuclear power plants, in preparation for service-life extension, and to determine the moment of onset of rapid aging of the metal and the end of the period of stable operation.The properties of the exponential distribution function and probability functions which are constructed for various loading scenarios using the hypothesis of probabilistic summation of fatigue damage for estimating the -percentage residual service life of equipment components are examined.Probabilistic estimates of the service life for operating nuclear power plants make it possible to control effectively the residual service life of the components of a nuclear power plant on the basis of information provided by the diagnostics systems and by the systems monitoring the state of the metal and the data on loading parameters from the control systems.     

Boiling water reactor with innovative safety concept: The Generation III+ SWR-1000

AREVA NP has developed an innovative boiling water reactor (BWR) SWR-1000 in close cooperation with German nuclear utilities and with support from various European partners. This Generation III+ reactor design marks a new era in the successful tradition of BWR and, with a net electrical output of approximately 1250 MWe, is aimed at ensuring competitive power generating costs compared to gas and coal fired stations. It is particularly suitable for countries whose power networks cannot facilitate large power plants. At the same time, the SWR-1000 meets the highest safety standards, including control of core melt accidents. These objectives are met by supplementing active safety systems with passive safety equipment of various designs for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads.The functional capabilities and capacities of all new systems and components were successfully tested under realistic and conservative boundary conditions in large-scale test facilities in Finland, Switzerland and Germany.In general, the SWR-1000 design is based on well-proven analytical codes and design tools validated for BWR applications through recalculation of relevant experiments and independent licensing activities performed by authorities or their experts. The overview of used analytical codes and design tools as well as performed experimental validation programs is presented.Effective implementation of passive safety systems is demonstrated through the numerical simulation of transients and loss of coolant accidents (LOCAs) as well as through analytical simulation of a severe accident associated with the core melt. In the LOCA simulation presented the existing active core flooding systems were not used for emergency control: only passive systems were relevant for the analyses. Despite this - no core heat-up occurred. In the case of reactor core melting numerically is demonstrated that the molten core debris would be retained inside the reactor vessel due to the effective passive external water cooling of the vessel, keeping it completely intact.A short construction period of just 48 months from first concrete to provisional take over, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burn-up all contribute towards meeting economic goals. Realistic average availability for a plant lifetime of 60 years and 12 months cycle is 94.5%. Systems and plant design were reviewed by expert groups of European utilities. With the SWR-1000, AREVA NP has developed a design concept for a BWR plant that is now ready for commercial deployment and which fully meets the most stringent international requirements in terms of nuclear safety and nuclear regulatory.     

Optimisation of the operating conditions of pressurised water reactors limited by the DNB phenomenon

The replacement of steam generators in a pressurised water reactor (PWR) requires an adaptation of the normal operating conditions imposed by modified characteristics of the replaced steam generators (RSGs). To cover the cost of steam generator replacement, plant uprate is often considered. This is made possible by a significant improvement of the heat exchange characteristics of the SGs, and also by modification of the position of the turbine control valve.The present paper discusses the methodology that has been developed by Tractebel to solve this problem for most of the up-to-date PWRs that are limited by the departure from nucleate boiling (DNB) phenomenon (boiling crisis). A code has been developed, validated and applied to two units in Belgium, respectively Doel 3 and Tihange 1, for which it has been decided to replace the steam generators. This code is founded on the basic principle that for most DNB-limiting class-2 accidents, the maximum reactor power and reactor inlet temperature may not violate either the overpower limit (currently 118% of nominal power), or the DNB-ratio criterion in the course of this accident. The code allows the designer to assess the influence on the operating conditions of the various key safety parameters, as well as of the basic assumptions that are necessary to conduct the thermalhydraulic design work. Examples of such investigations are given in this paper. Special emphasis is put on the thermal-hydraulic (TH) design procedure, the plant principal TH key parameters, as well as the features of the new SGs. The development of the DNBOPT code was also motivated by the wish to perform audit and sensitivity calculations, as well as to confirm calculations performed by the SG supplier.     

Design procedure for the primary circuit of the Konvoi plants

The loading data for the design and fatigue analyses of the primary system components are derived from the plant operating transients as specified in the Operational Manual, the accident analyses and from the external event requirements. The methods used, applicable codes and standards, and QA procedures produce a design concept that contains significant safety margins. Measurements conducted in pre-Konvoi plants and during commissioning of the Konvoi plants deliver additional verifications of the analyses.     

Some Aspect of Protective Systems in Nuclear Power Plants

Protective systems for nuclear power reactors have assumed a high degree of development. The primary motive has been the emphasis given to making reactors safe and more recently the desire to offset the site distance requirement through engineered safeguards. Increased reactor operating experience and improved safety analyses have contributed to a better definition of the events and conditions requiring protective systems. The judicious use of such philosophies as redundancy and coincidence in system designs has led to both greater safety and reactor operating continuity. Reliability of protective systems has been enhanced by parallel efforts in system and individual component design. Techniques and procedures for checking and testing protective systems have been developed and adopted at many installations to offset the inherent difficulty of assessing reliability of systems which experience little or no use under actual stressed conditions. Design practices have been effected to provide greater assurance that systems are independent although this remains as one of the outstanding problems. Protective systems in the context of monitoring devices are being developed by the application of noise analyses, digital computer control and the use of transistorized or solid state circuitry. Finally, the actual performance of protective systems is being manifested through analysis of existing operating records.     

Estimate of the Effect of the Time Interval between Technical Inspections on the Strength Reliability of Certain First-Loop Components of a VVER-1000 Reactor

The results of a computational investigation of the possibility of and the safety conditions for switching from a 4- to an 8-yr time interval between technical inspections of the main circulation pipeline and the pressure compensator vessel in VVER-1000 reactors are presented. To this end, calculations of the critical and admissable sizes of defects in the main metal and weld seams in the operating regime, in accident situations, and during earthquakes have been performed. Calculations of the time for a through defect to reach a critical size for different operating periods have been performed. The influence of the hydrotesting pressure and the time interval between such tests on the operational safety and the effect of the time between technical inspections on the reliability of the indicated first-loop components of a VVER-1000 reactor with respect to the criterion of fracture resistance, taking account of the probabilistic nature of the initial data, is analyzed.The calculations were performed using normative and certified procedures geared toward the typical characteristics of steel and structures and the conditions of fabrication, assembly, and operation.__________Translated from Atomnaya Energiya, Vol. 98, No. 4, pp. 267–273, April, 2005.     

ITER operating limit definition criteria

The operating limits and conditions (OLCs) are operating parameters and conditions, chosen among all system/components, which, together, define the domain of the safe operation of ITER in all foreseen ITER states (operation, maintenance, commissioning). At the same time they are selected to guarantee the required operation flexibility which is a critical factor for the success of an experimental machine such as ITER. System and components that are important for personnel or public safety (safety important class, SIC) are identified considering their functional importance in the overall plant safety analysis. SIC classification has to be presented already in the preliminary safety analysis report and approved by the licensing authority before manufacturing and construction.OLCs comprise the safety limits that, if exceeded, could result in a potential safety hazard, the relevant settings that determine the intervention of SIC systems, and the operational limits on equipment which warn against or stop a functional deviation from a planned operational status that could challenge equipment and functions. Some operational conditions, e.g. in-Vacuum Vessel (VV) radioactive inventories, will be controlled through procedures. Operating experience from present tokamaks, in particular JET, and from nuclear plants, is considered to the maximum possible extent.This paper presents the guidelines for the development of the ITER OLCs with particular reference to safety limits.     

Application of advanced ultrasonic test techniques coupled with fatigue and leakage monitoring to assure integrity of BWR feedwater nozzles

As a result of feedwater nozzle cracking observed in Boiling Water Reactor (BWR) plants, several design modifications were implemented to eliminate the thermal cycling that led to crack initiation. BWR plants with these design changes have successfully operated for over ten years without any recurrence of cracking. To provide further assurance of this, the U.S. Nuclear Regulatory Commission (NRC) issued NUREG-0619, which established periodic ultrasonic testing (UT) and liquid penetration testing (PT) requirements. While these inspections are useful in confirming structural integrity, they are time consuming and can lead to significant radiation exposure to plant personnel. In particular, the PT requirement poses problems since it is difficult to perform the inspections with the feedwater sparger in place and also leads to additional personnel exposure. Clearly, an inspection and monitoring program that eliminates the PT examination and still verifies the absence of surface cracking would be extremely valuable in limiting costs as well as radiation exposure. This paper describes a program involving the application of advanced UT techniques coupled with fatigue and leakage monitoring to assure integrity of BWR feedwater nozzles. The inspection methods include: (1) scanning with optimized transducers and techniques from the outside vessel wall surface to inspect the nozzle inner radius region, and (2) scanning from the nozzle forging outside-diameter to inspect the nozzle bore region. Methods of analyzing the data using 3-D graphics displays have been developed that show crack location, size, and maximum depth of penetration into the nozzle inner surface. These techniques have been developed to the point where they are now considered a reliable alternative to the liquid penetrant requirements of NUREG-0619. An important supplement to the UT program is the use of automated fatigue, leakage and crack growth monitoring to verify the absence of cracking. This approach provides for a continuous assessment of the integrity of the nozzle structure by tracking the actual fatigue duty, measuring thermal sleeve bypass leakage and performing crack growth predictions based on actual thermal duty. Collectively, the monitoring and inspection program provides technically sound assurance of nozzle integrity and a firm basis for plant operational planning.     

Ageing management strategies for instrumentation and control equipment in NPPs

This paper aims to explain main ageing stressors and effects on instrumentation and control (I&C) components followed by ageing management strategies through developed design such as using freewheeling diode to eliminate relays contact ageing due to inductive loads, or through traditional practices and new online monitoring where I&C components can be checked and assessed remotely while the plant is operating. It also includes some recommendations for obsolescence management for special I&C components like microcontrollers and programmable logic controllers (PLCs).     

Two new passive safety systems for LWR applications

In a research activity that SIET has been conducting for years about safety systems for light water reactors (LWRs), attention has been paid to developing two passive injection systems representing an innovative solution in mitigating the consequences of loss of coolant accidents. Both systems allow the completely passive injection of cold water into a pressurised vessel. They are triggered by a low-level signal and work on the base of phenomena like natural circulation and condensation. The simplest system, Sistema Iniezione Passiva 1 (SIP-1), injects water contained in a tank into a circuit at the same pressure as the circuit. The most complex system, injection cyclic system (ICS), injects cold water, by filling cyclically a proper tank with the water stored in an atmospheric pressure pool. Thanks to the ENEA sponsorship, this activity has been conducted in three steps: the definition of the conceptual design of the systems; the application of the Relap5 code to simulate their behaviour; and the proposal of their specific applications to pressurised and boiling LWR. In this paper, both systems are presented in their structural and operating characteristics together with the main results of the code application for their simulation. Some proposals of application of SIP-1 and ICS to pressurised water reactors and boiling water reactors are also shown. The developments and reached goals of the prosecution of the research are also summarised here, together with future needs.