C.B. Oland

Report on aging of nuclear power plant reinforced concrete structures

The Structural Aging Program provides the US Nuclear Regulatory Commission with potential structural safety issues and acceptance criteria for use in continued service assessments of nuclear power plant safety-related concrete structures. The program was organized under four task areas: Program Management, Materials Property Data Base, Structural Component Assessment/Repair Technology, and Quantitative Methodology for Continued Service Determinations. Under these tasks, over 90 papers and reports were prepared addressing pertinent aspects associated with aging management of nuclear power plant reinforced concrete structures. Contained in this report is a summary of program results in the form of information related to longevity of nuclear power plant reinforced concrete structures, a Structural Materials Information Center presenting data and information on the time variation of concrete materials under the influence of environmental stressors and aging factors, in-service inspection and condition assessments techniques, repair materials and methods, evaluation of nuclear power plant reinforced concrete structures, and a reliability-based methodology for current and future condition assessments. Recommendations for future activities are also provided. 308 refs., 61 figs., 50 tabs.

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.

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.

An overview of the ORNL/NRC program to address aging of concrete structures in nuclear power plants

Abstract The Structural Aging (SAG) Program is being conducted at the Oak Ridge National Laboratory (ORNL) for the Nuclear Regulatory Commission (NRC). The SAG Program is addressing the aging management of safety-related concrete structures in nuclear power plants for the purpose of providing improved technical bases for their continued service. The program is organized into four tasks: Program Management, Materials Property Data Base, Structural Component Assessment/Repair Technologies, and Quantitative Methodology for Continued Service Determinations. Objectives and a summary of accomplishments under each of these tasks are presented.

Ageing management of safety-related concrete structures to provide improved bases for continuing the service of nuclear power plants

ResumeOn décrit les structures en béton se rapportant à la sécurité des installations nucléaires, et on détermine les facteurs de dégradation susceptibles de compromettre la conformité de ces structures aux critères de fonctionnement et de performance qui leur sont assignés. On donne ici, de façon résumée, les exigences d’inspection en service. On examine l’évaluation des performances des éléments en béton dans les installations nucléaires. Enfin, on trace les grandes lignes d’un programme pour gérer le vieillissement des structures en béton en rapport avec la sécurité des installations nucléaires, et améliorer leur capacité de service. On résume également les différents aspects du programme.

Structural aging program status report

Research is being conducted at the Oak Ridge National Laboratory under Nuclear Regulatory Commission (USNRC) sponsorship to address aging management of safety-related concrete 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. Program accomplishments have included development of the 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, performance assessments of reinforced concrete structures in several United Kingdom nuclear power facilities, evaluation of European and North American repair practices for concrete, an evaluation of factors affecting the corrosion of metals embedded in concrete, and application of the time-dependent reliability methodology to reinforced concrete flexure and shear structural elements to investigate the role of in-service inspection and repair on their probability of failure.

Research and development at the structural materials information center

Abstract The U.S. Nuclear Regulatory Commission has initiated a Structural Aging Program at the Oak Ridge National Laboratory to identify potential structural safety issues related to continued service of nuclear power plants and to establish criteria for evaluating and resolving these issues. One of the tasks in this program focuses on the establishment of a Structural Materials Information Center where long-term and environment-dependent properties of concretes and other structural materials are being collected and assembled into a data base. These properties will be used to evaluate the current condition of critical structural components in nuclear power plants and to estimate their future performance during the continued service period.

Flaw behavior in mechanically loaded clad plates

Abstract A small crack near the inner surface of clad nuclear reactor pressure vessels is an important consideration in the safety assessment of the structural integrity of the vessel. Four-point bend tests on large plate specimens, six clad and two unclad, were performed to determine the effect of stainless steel cladding upon the propagation of small surface cracks subjected to stress states similar to those produced by pressurized thermal shock conditions. Test results have shown that the tough surface layer composed of cladding and/or heat-affected zone has enhanced the load-bearing capacity of plates under conditions where unclad plates have ruptured. The results are interpreted in terms of fracture mechanics. The behavior of flaws in clad reactor pressure vessels is examined in the light of the test results.