A study of the characteristics of the acoustic emission signals for condition monitoring of check valves in nuclear power plants

The objective of this study is to demonstrate that a condition-monitoring system based on acoustic emission (AE) detection can provide timely detection of check valve degradation and service aging so that maintenance or replacement can be preformed prior to the loss of safety function. This research is focused on the investigation and understanding of the capability of the acoustic emission technique to provide diagnostic information on check valve failures.AE testing for a check valve under controlled flow loop conditions was performed to detect and valve degradation such as wear and leakage due to foreign object interference. It is clearly demonstrated that the distinction of different types of failure were successful by systematically analyzing the characteristics of various AE parameters.     

Fatigue analysis for analytically overloaded piping components and valves in nuclear power plants

Lately, in connection with life extension aspects of power plants, an increasingly accurate determination of the lifetime of components in nuclear stations is being required. In order to assess reliably current fatigue levels in piping systems, variables such as pressure, temperature, and resultant force and moment transients as well as analytical methods which take into account the real operational history must be considered. This paper presents a method for analyzing the transient heat transfer between fluid and pipe wall in order to investigate effects which until now have been assumed conservatively to be caused by a sudden jump in temperature. Further, an example is given showing that the K_e factor approach in current design codes for performing simplified elastic-plastic fatigue analyses is conservative.     

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.     

Offshore nuclear power plants

Offshore Power Systems, a joint enterprise of Westinghouse and Tenneco, has been formed to manufacture floating nuclear power plants. Commitments for the first two offshore plants have been received from the Public Service Electric and Gas Company. This paper describes the floating nuclear plant concept with special reference to its advantages and its novel features. The novel features are a consequence of the floating aspect and include the design of the platform, the safety analysis and also the analysis and specification of plant motions due to environmental effects such as wind, waves and earthquakes. Site-related aspects such as the breakwater and mooring systems are discussed. The nuclear power plants will be manufactured in a central facility and this manufacturing concept is described.     

Mixture loss coefficient of safety valves used in nuclear plants

Theoretical and experimental investigations on the loss coefficient of gas–liquid mixture across safety relief valves have been carried out. Experiments were performed for three different types of safety valves and under different flow conditions. Using the Darcy equation and based on the presented experimental results, a new empirical correlation has been developed to calculate the loss coefficient and hence pressure loss. By consideration of flow contraction, high viscous fluids, Reynolds number and safety valve geometry, the model includes therefore the relevant primary influencing parameters. The reproductive accuracy of the proposed model and the statistical comparison, based on about 2000 measured data in the literature, demonstrated that the proposed model is the best overall agreement with the data. The standard deviation of the data is less than 27%. The model fits the data well and is sufficiently accurate for engineering purposes. The reported results of the tested safety relief valve are very important to improve the practical and safety design of the nuclear plants.     

Managing aging in nuclear power plants

Aging degradation in nuclear power plants must be controlled to prevent safety margins from declining below limits provided in plant design bases. The NPAR Program and other aging-related programs conducted under the auspices of the NRC Office of Research are developing needed technical guidance for control of aging. Results from these programs, together with relevant information developed by industry and elsewhere, are implemented through various ongoing NRC and industry programs and initiatives as well as by means of conventional regulatory instruments. The aging control process central to these efforts consists of three key elements: (1) selection of components, systems, and structures (CSS) in which aging must be controlled, (2) understanding of the mechanisms and rates of degradation in these CSS, and (3) managing degradation through effective surveillance and maintenance. These elements are addressed in Recommended Practices Guidance that integrates information developed under NPAR and other studies of aging into a systems-oriented format that tracks directly with the Safety Analysis Reports and with the NRC Standard Review Plan (NUREG-0800).     

Cosmic thermoemissional nuclear power plants

Translated from Atomnaya Énergiya, Vol. 69, No. 4, pp. 215–219, October, 1990.     

Replacement of safety-related valves in Philippsburg 1 nuclear power plant

The Philippsburg nuclear power plant Unit 1, initially commissioned in 1979 is presently still partly equipped with safety-related isolation valves which correspond to the technical standard of the 1970s. Inspection of these valves according to today's technical rules revealed that a number of these isolation valves can be upgraded in such a way that compliance with the present state of the art is achieved. However, the advanced state of wear of frequently operated valves suggested that a program should be set up for the gradual replacement of these valves, with a view to this upgrading. At present, the replacement program relates to a period from 1993–1997 with the selection of valves to be replaced depending on the signs of wear and the operational possibilities. Within this period, it is planned to replace 50% of the safety related valves with those that have been in compliance with the latest technical rules, and that have been improved with regard to that resistance to wear and ease of maintenance.     

Risk assessment of nuclear power plants

The objective of the present work is to develop recommendations for controlling the safety of nuclear power plants on the basis of risk assessments and safety certification of nuclear power plants. The Kursk nuclear power plant is considered as an example of a nuclear power plant with an RBMK reactor. The concept of risk assessment of a nuclear power plant consists in constructing a set of scenarios of the appearance and development of possible accidents followed by an evaluation of the realization frequency and determination of the scales of the consequences of each one. The result of an analysis is an evaluation of a system of risk indicators in accordance with the requirements of the safety compliance certificate of the nuclear power plant as well as the development of recommendations for increasing plant safety. In risk assessment, the consequences are divided into categories of the seriousness of the damage, for which their probability is evaluated separately. The graphical interpretation of risk due to any dangerous object consists of frequency–consequences curves. Recommendations are developed on the basis of the results of risk analysis.     

Infrastructure of low-capacity nuclear power plants

Conceptual questions concerning the development of a system of low-capacity nuclear power plants are discussed. The basic properties which the nuclear power facilities of such plants must have are formulated. Questions concerning personnel training, the control characteristics of a system of low-capacity plants, decommissioning, and the requirements for physical protection are examined. The need to develop special normative documentation for low-capacity nuclear power plants is substantiated. Questions concerning the ecological effects of low-capacity power generation are touched upon.     

Cost-effectiveness of low-power nuclear power plants

OKMB. Translated from Atomnaya Énergiya, Vol. 75, No. 5, pp. 333-336, November, 1993.     

On-line monitoring applications in nuclear power plants

The nuclear power industry is working to reduce generation costs by adopting condition-based maintenance strategies and automating testing activities. These developments have stimulated great interest in on-line monitoring (OLM) technologies and new diagnostic and prognostic methods to anticipate, identify, and resolve equipment and process problems and ensure plant safety, efficiency, and immunity to accidents. This paper provides examples of these technologies with particular emphasis on eight key OLM applications: detecting sensing-line blockages, testing the response time of pressure transmitters, monitoring the calibration of pressure transmitters on-line, cross-calibrating temperature sensors in situ, assessing equipment condition, performing predictive maintenance of reactor internals, monitoring fluid flow, and extending the life of neutron detectors. These applications are discussed in the following sections. Emphasis is placed on the principles of a core OLM method - noise analysis - and the technical requirements for an integrated OLM system are summarized.     

Preventive fire protection in nuclear power plants

Fire risk considerations in nuclear power plants and questions of preventive fire protection have so far not been dealt with sufficient attention. For this reason a research program was proposed and financed by the government of the Federal Republic of Germany in order to clarify these questions and to optimise preventive fire protection measures especially in nuclear power plants.     

Flood hazards for nuclear power plants

Flooding hazards for nuclear power plants may be caused by various external geophysical events. In this paper the hydrologic hazards from flash floods, river floods and heavy rain at the plant site are considered. Depending on the mode of analysis, two types of hazard evaluation are identified: (a) design hazard which is the probability of flooding over an expected service period, and (b) operational hazard which deals with real-time forecasting of the probability of flooding of an incoming event. Hazard evaluation techniques using flood frequency analysis can only be used for type (a) design hazard. Evaluation techniques using rainfall-runoff simulation or multi-station correlation can be used for both types of hazard prediction.     

Fuzzy identification of transients in nuclear power plants

The identification of transients is of fundamental importance for the timely monitoring of nuclear plants operation. The main target is detecting the occurrence of the onset of a transient, in its early stages, and identifying its kind so as to be able to readily act to fix its causes. Given the safety and economical importance of the problem, various approaches have been investigated and applied for transient identification, and many efforts are still devoted to the improvement of the results so far obtained.In this paper, a fuzzy-logic based method for the identification of transients is proposed. The method is ‘model-free’ in that the if-then rules, which constitute the heart of the approach, are inferred only from the available input-output signal data. The method is tested on an example of identification of reactor transients generated by four forcing functions of different nature. The necessary data for the identification have been simulated by the QUAndry-based Reactor Kinetics code (QUARK, distributed freely by NEA) configured so as to model the operations of the Westinghouse Advanced Pressurized water reactor, AP600.     

Occurrence rates of fires in nuclear power plants

The time histories of fires in nuclear power plants are examined. Occurrence rates and expected number of fires vs. time are obtained. The time dependence analysis is modelled as a non-homogeneous Poisson process with Weibull occurrence rate. Confidence bounds on the model parameters are determined and the appropriateness of the model is evaluated using various statistics. The results are expected to be useful for application to risk studies.     

Welding consumables for nuclear power plants

In nuclear power plants, submerged arc welding and covered arc welding have long been employed especially for main weld seams, including the core region of RPV.This paper investigates the mechanical properties of several welding consumables we have developed for industrial plants — that is, welding consumables which lower the phosphorus and copper content of the welded metal, those for plates possessing particularly high tensile strength and those for the narrow gap welding method.Recent data derived from irradiation embrittlement tests show that these welded metals using a non-copper coating are highly effective in minimizing shifts in the transition curve.Welding consumables for A533B C1.2, A543 C1.1 or A508 C1.4 steels have a higher tensile strength than those for A533B C1.1 or A508 C1.3.We have developed submerged arc and covered arc welding consumables to be used with these kinds of steels, and it was confirmed that these consumables possess excellent tensile strength and notch toughness.Our tests also confirmed that the narrow gap SAW and MIG welds are more efficient than the conventional ones. Moreover, the mechanical properties of the welded metals are also excellent.