Repair materials and techniques for concrete structures in nuclear power plants

The paper summaries portions of work of the Structural Aging Program, sponsored by the Nuclear Regulatory Commission (NRC). The paper addresses the assessment and repair of concrete structures in nuclear power plants. It presents the results of a survey of the the nuclear power plants in the United States to identify susceptible concrete components, rates of occurrence of deterioration, and to determine the durability of repairs. The paper describes deterioration mechanisms and discusses their effect. Repair techniques are described. Evaluation techniques and nondestructive test techniques are also discussed.     

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.     

Reliability-based assessment of integrity of bonded prestressed concrete containment structures

The main function of a nuclear containment structure is to prevent the leakage of radioactive materials from the reactor in the event of a serious failure in the process system. To maintain a high level of leak integrity, prestressed concrete is widely utilized in containment construction. In bonded prestressing systems, excessive prestressing losses caused by unexpected material deformations and degradation of tendons could result in the loss of leak integrity under an accident. To safeguard against this, the Canadian Standard, CSA N287.7 (1995), recommends periodic inspection and evaluation of prestressing systems of CANDU containments. As bonded tendons are not amenable to direct inspection, the evaluation is based on the testing of a set of beams with features identical to the containment. The paper presents a quantitative reliability-based approach to evaluate the containment integrity in terms of the condition of bonded prestressing systems. The proposed approach utilizes the results of lift-off, destructive, and flexural tests to update the probability distribution of prestressing force, and to revise the calculated reliability against through-wall cracking of containment elements. An acceptable criterion for the results of beam tests is established on the basis of maintaining adequate reliability throughout the service life of the containment.     

Reliability and service life of VGKK series sealed lead-throughs for nuclear power plants

Evaluation tests of sealed cable lead-throughs for nuclear power plants are examined. It is shown that the currently existing systematic testing for the effect of all factors arising in a large-leak regime does not give a real picture of the resistance of sealed cable lead-throughs. Data on the admissible absorbed dose which are obtained only on the basis of tests for radiation resistance can be erroneous when extended to operating conditions where environmental factors act simultaneously. A testing chamber and a method for performing measurements under conditions simulating a regime with a large leak are described. __________ Translated from Atomnaya énergiya, Vol. 104, No. 3, pp. 161–164, March, 2008.     

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.     

Aging management of containment structures in nuclear power plants

Research is being conducted by Oak Ridge National Laboratory under US Nuclear Regulatory commission (USNRC) sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the USNRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a Structural Materials Information Center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of non-destructive evaluation techniques, assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants.     

Analysis of the failure probability of pipelines and equipment with service life extension of nuclear power plants

A method is proposed for calculating the failure probability of pipelines and equipment that takes account of experience in operating the structures in different regimes and aging. The random quantities used in the method are parameters characterizing the applied loads and defects as well as the strength, mechanical, and thermophysical properties of metal. An example of a calculation of the failure probability for a RBMK emergency recirculation pipeline of emergency feed pumps with service life extension is presented. __________ Translated from Atomnaya énergiya, Vol. 104, No. 5, pp. 284–290, May, 2008.     

Pressure relief structures of multi-unit candu nuclear power plants

Ontario Hydro's three principal nuclear power plants, Pickering, Bruce and Darlington GS incorporate a sequence of pressure relief structures linking all reactors to a vacuum building where accident produced volatile radioactive substances are contained. The paper discusses the design of Pickering's elevated pressure relief duct and its connection to the reactors by rupture panels. The development of the Pickering prototype pressure relief valve is described. At Bruce and Darlington a linear duct cuts through solid rock beneath the reactors, serving both as an access tunnel for their fuelling machines and as part of the pressure relief system. Its continuation, the reinforced concrete manifold, comprises an above-ground toroidal structure partially or fully encircling the vacuum building. The nature of the manfold and its supporting structures is outlined, with emphasis on the special loading conditions at Darlington arising from high seismic forces, hydrostatic uplift, tornado, and explosion.     

Service life extension of power-generating units in nuclear power plants

Questions arising in the analysis of service life extension of power-generating units in nuclear power plants under the conditions of national realities are analyzed. Science and Technology Center for Nuclear Reactor Safety. Translated from Atomnaya énergiya, Vol. 88, No. 1, pp. 14–21, January, 2000.     

Safeguard assessment for life extension in nuclear power plants (NPPs) using a production function

Life extension is investigated as a safeguard assessment for the stability on the operation of the nuclear power plants (NPPs). The Cobb-Douglas function, one of the production functions, is modified for the nuclear safeguard in NPPs, which was developed for the life quality of the social and natural objects. Nuclear Safeguard Estimator Function (NSEF) is developed for the application in NPPs. The cases of NPPs are compared with each other in the aspect of the secure performance. The results are obtained by the standard productivity comparisons with the designed power operations. The range of secure life extension is between 1.008 and 5.353 in 2000 MW_e and the range is between 0.302 and 0.994 in 600 MW_e. So, the successfulness of the power operation increases about 5 times higher than that of the interested power in this study, which means that the safeguard assessment has been performed in the life extension of the NPPs. The technology assessment (TA) is suggested for the safe operation which is an advanced method comparing conventional probabilistic safety assessment (PSA).     

Probabilistic assessment of seismic risk for nuclear power plants

Probabilistic approaches to the design, siting, and safety analysis of nuclear power plants have been proposed by Farmer, Wall, and Garrick. Farmer and Wall established a limit line which delineates between acceptable and unacceptable risks. To implement the method, all accidental chains are systematically analyzed to determine their probability and associated fission product release magnitude; the combination is compared to the limit line. For proper implementation, the seismic risk should be evaluated in a quantified manner. Conceptually, this evaluation is made in two stages: the probability of an earthquake occurrence as a function of its intensity and, given a seismic intensity, the conditional probability of damage. This paper reports on an initial study into the latter aspect.The effect of uncertainty in several parameters which determine the response of a nuclear reactor building to earthquake forces is assessed. Probability distributions for material properties were determined from site measurements and these distributions were utilized for determining the building response and the damage criterion. A subjective probability density function for damping was assigned from the available information and the judgment of experienced engineers. Four accelerograms, El Centro N---S 1940, and three artificial earthquakes were used to represent the variability in the forcing functions. The uncertainty in the model idealization was assessed by evaluating three alternate models. A versatile computer program was developed to compute the response of the mathematical model to the forcing functions using matrix formulation and modal method of analysis. An exact solution, rather than numerical integration, was used to obtain the dynamic response of the system in generalized coordinates.The stresses within the reactor building are similar for different earthquakes considered in this study when they are normalized to ground acceleration, indicating that the shape of the accelerogram and its frequency content are less significant than the magnitude of the maximum ground acceleration for the reactor building. The variation in modulus of elasticity for concrete had a significant effect on the building response. Damping, in general, reduced the response, but in cases where the duration of an earthquake is short the effect was not very significant.A simple failure criteria for ultimate shear stress in shear walls, τ_ult = 4.75 √ƒ′_c, where ƒ′_c is the ultimate compressive strength of concrete, is used to estimate the initiation of cracking in the walls. The normal design of the reactor building is deterministic and is based on a 0.2 g design basis earthquake. Using the results obtained by the parametric study on the variation of response, the probability of damage was estimated by a Monte Carlo analysis. It was estimated that, given the occurrence of a design basis earthquake, there is less than approximately 10⁻³ probability of cracking in the shear walls of the reactor building. The initiation of cracking in the concrete should not lead to a significant release of contained fission products.     

Study on a concrete filled structure for nuclear power plants

The feasibility of a new structural system for nuclear power plant buildings utilizing concrete filled steel structures, termed ‘SC structural system' was studied. SC wall test specimens (1/5 scale) were manufactured and compressive loading tests were carried out to determine how to prevent buckling. Also, bending shear tests were performed using H-section wall specimens to determine the shear and bending characteristics of SC walls. This paper presents an outline of the feasibility study, and the various structural properties resulting from the experiments.     

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.     

Seismic design of nuclear power plants — an assessment

This paper presents a review and evaluation of the design standards and the analytical and experimental methods used in the seismic design of nuclear power plants with emphasis on United States practice. Three major areas were investigated: (a) soils, siting, and seismic ground motion specification; (b) soil-structure interaction; and (c) the response of major nuclear power plant structures and components. The purpose of this review and evaluation program was to prepare an independent assessment of the state-of-the-art of the seismic design of nuclear power plants and to identify seismic analysis and design research areas meriting support by the various organizations comprising the ‘nuclear power industry’. Criteria used for evaluating the relative importance of alternative research areas included the potential research impact on nuclear power plant siting, design, construction, cost, safety, licensing, and regulation.Three methods were used in the study herein. The first involved the review of current literature, focusing primarily on publications dated later than 1970. This review included the results of numerous studies, of which those of Japanese origin and those presented in recent international conferences were predominant. The second method entailed a review of international experience in the dynamic testing of nuclear power plant structures and components, and related experience with scaled and model tests. Included in this experience, in addition to the questions of analysis, design, and measurement of dynamic parameters, are related efforts involving a review of responses obtained during measured earthquake response and investigations into appropriate methods for backfitting or upgrading older nuclear power plants to meet new seismic criteria.The third approach was to obtain the opinions and recommendations of technically knowledgeable individuals in the US ‘nuclear industry’; the survey results are shown in the Appendix.     

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.     

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).     

Overview of the use of prestressed concrete in U.S. nuclear power plants

The extent of the use of prestressed concrete in nuclear power plants is outlined. Evolution of large size prestressing systems and corrosion inhibiting materials is described. A summary of major problems which have been encountered with prestressed concrete construction at nuclear power plant containments in the United States is presented; that is, dome delamination, cracking of anchorheads, settlement of bearing plates, etc. Guidelines for a tendon inservice inspection program are described as well as the effectiveness of these programs. The paper concludes with an assessment of the overall effectiveness of the prestressed concrete containments.