[en] NRC has initiated a rulemaking to 10CFR50.55a that would allow Owners to voluntarily update their pump and valve inservice testing programs to the 1995 Edition of the OM Code. The 1992 and 1994 Addenda, and 1995 Edition of the OM Code offers many improvements, eg, clarifications and relaxations, to the 1989 Edition of Section XI or the 1990 Edition of the OM Code. This paper reviews the code changes that may be advantageous for Owners to use, and discusses their related requirements. Additionally, code improvements in the newly issued 1996 Addenda of the OM Code are discussed, as they may be proposed under 10CFR50.55a(a)(3)(i)

[en] The effect of aging upon the Babcock ampersand Wilcox (B ampersand W) and Combustion Engineering (CE) Control Rod Drive (CRD) systems has been evaluated as part of the US NRC Nuclear Plant Aging Research (NPAR) program. Operating experience data for the 1980--1990 time period was reviewed to identify predominant failure modes, causes, and effects. These results, in conjunction with an assessment of component materials and operating environment, conclude that both systems are susceptible to age degradation. System failures have resulted in significant plant effects, including power reductions, plant shutdowns, scrams, and Engineered Safety Feature (ESF) actuation. Current industry inspection and maintenance practices were assessed. Some of these practices effectively address aging, while others do not

[en] The effect of aging upon the Babcock ampersand Wilcox (B ampersand W) and Combustion Engineering (CE) Control Rod Drive (CRD) systems has been evaluated as part of the USNRC Nuclear Plant Aging Research (NPAR) program. Operating experience data for the 1980--1990 time period was reviewed to identify predominant failure modes, causes, and effects. These results, in conjunction with an assessment of component materials and operating environment, conclude that both systems are susceptible to age degradation. System failures have resulted in significant plant effects, including power reductions, plant shutdowns, scrams, and Engineered Safety Feature (ESF) actuation. Current industry inspection and maintenance practices were assessed. Some of these practices effectively address aging, while others do not

[en] A preliminary review of inservice testing (IST) effectiveness for Class 1, 2, and 3 pumps at nuclear power plants was performed. IST requirements are specified by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, and the Operations and Maintenance Standard (OM Part 6). The Institute of Nuclear Power Operations Nuclear Plant Reliability Data System was used to provide failure reports for these components from 1988 to 1992. This time frame coincides with the issuance of Generic Letter 89-04, which resulted in a more consistent application of the requirements by the licensees. For this time, 2,585 pump failures were reported. A review of these failures indicated that the majority (71.6%) resulted from external leakage. These events were excluded from the study because the typically do not affect pump operability and are not detected by the measurement of IST parameters. The remaining 733 events were reviewed to identify the primary failure causes, failure modes, and method of detection. Plant testing programs, consisting of IST, surveillance testing, detected approximately 40% of these occurrences. Others were detected through operational abnormalities, routine and incidental observations, alarms, and while performing maintenance. This paper discusses the results of the study

[en] Aging of systems and components in nuclear power plants is a well known occurrence that must be managed to ensure the continued safe operation of these plants. Much of the degradation due to aging is controlled through periodic maintenance and/or component replacement. However, there are components that do not receive periodic maintenance or monitoring once they are installed; electric cables are such a component. To provide a means of monitoring the condition of electric cables, research is ongoing to evaluate promising condition monitoring (CM) techniques that can be used in situ to monitor cable condition and predict remaining life. While several techniques are promising, each has limitations that must be considered in its application. This paper discusses the theory behind several of the promising cable CM techniques being studied, along with their effectiveness for monitoring aging degradation in typical cable insulation materials, such as cross-linked polyethylene and ethylene propylene rubber. Successes and limitations of each technique are also presented

[en] There are a variety of environmental stressors in nuclear power plants that can influence the aging rate of components; these include elevated temperatures, high radiation fields, and humid conditions. Exposure to these stressors over long periods of time can cause degradation of components that may go undetected unless the aging mechanisms are identified and monitored. In some cases the degradation may be mitigated by maintenance or replacement. However, some components receive neither and are thus more susceptible to aging degradation, which might lead to failure. One class of components that falls in this category is electric cables. Cables are very often overlooked in aging analyses since they are passive components that require no maintenance. However, they are very important components since they provide power to safety related equipment and transmit signals to and from instruments and controls. This paper will look at the various aging mechanisms and failure modes associated with electric cables. Condition monitoring techniques that may be useful for monitoring degradation of cables will also be discussed

[en] A preliminary review of inservice testing effectiveness for Code Class 1,2, and 3 valves at nuclear power plants was performed. These requirements are specified by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI, and the Operations and Maintenance Standard. The Institute of Nuclear Power Operations Nuclear Plant Reliability Data System (NPRDS) database was used to provide failure reports for these components for 1988 to 1992. This time period coincides with the issuance of Generic Letter 89-04, which resulted in a more consistent application of the requirements by the licensees. For this time period, 8,593 valve failures were identified. From the review of the NPRDS database, the primary failure causes and failure modes for motor-operated valves (MOV), air-operated valves (AOV), and check valves (CV) were identified. Solenoid-operated valves were not reviewed in this study. Plant testing programs were effective in identifying approximately 60% of the CV failures, 46% of the AOV failures, and 44% of the MOV failures

[en] A preliminary review of Inservice Testing (IST) effectiveness for Code Class 1, 2, and 3 pumps at nuclear power plants was performed. IST requirements are specified by ASME Section XI, and the Operations and Maintenance Standard (OM Part 6). The INPO NPRDS database was used to provide failure reports for these components for 1988 to 1992. This time frame coincides with the issuance of Generic Letter 89-04, which resulted in a more consistent application of the requirements by the licensees. For this time period, 2585 pump failures were reported. A review of these failures indicated that the majority (71.6%) were due to external leakage, and were excluded from this study since these events typically do not affect pump operability and are not detected by the measurement of IST parameters. Of the remaining 733 events, a review was performed to identify the primary failure causes, failure modes, and method of detection. Plant testing programs, consisting of IST, surveillance testing, and special testing, detected approximately 40% of these occurrences. Others were detected through operational abnormalities, routine and incidental observations, alarms, and while performing maintenance. This paper provides a discussion of the results of the study

[en] An aging assessment of PWR control rod drive (CRD) systems has been completed as part of the US NRC Nuclear Plant Aging Research (NPAR) Program. The design, construction, maintenance, and operation of the Babcock ampersand Wilcox (B ampersand W), Combustion Engineering (CE), and Westinghouse (W) systems were evaluated to determine the potential for degradation as each system ages. Operating experience data were evaluated to identify the predominant failure modes, causes, and effects. This, coupled with an assessment of the materials of construction and operating environment, demonstrate that each design is subject to degradation, which if left unchecked, could affect its safety function as the plant ages. An industry survey, conducted with the assistance of EPRI and NUMARC, identified current CRD system maintenance and inspection practices. The results of this survey indicate that some plants have performed system modifications, replaced components, or augmented existing preventive maintenance practices in response to system aging. The survey results also supported the operating experience data, which concluded that the timely replacement of degraded components, prior to failure, was not always possible using existing condition monitoring techniques. The recommendations presented in this study also include a discussion of more advanced monitoring techniques, which provide trendable results capable of detecting aging

[en] A study was performed to assess the effects of aging on the Containment Isolation (CI) functions of Westinghouse Pressurized Water Reactors and General Electric Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research (NPAR) program, sponsored by the U.S. Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. Failure data from two national databases, Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Reports (LERs), as well as plant specific data were reviewed and analyzed to understand the effects of aging on the CI functions. This study provided information on the effects of aging on component failure frequency, failure modes, and failure causes. Current inspection, surveillance, and monitoring practices were also reviewed