Acoustic and Ultrasonic Signals as Diagnostic Tools for Check Valves

1993 ◽  
Vol 115 (2) ◽  
pp. 135-141 ◽  
Author(s):  
M. K. Au-Yang
Keyword(s):  
Nuclear Power ◽  
Structural Integrity ◽  
Operating Experience ◽  
Check Valve ◽  
Nuclear Regulatory Commission ◽  
Nuclear Plant ◽  
Diagnostic Tools ◽  
Piping System ◽  
Proper Training ◽  
Regulatory Commission

A typical nuclear plant has between 60 and 115 safety-related check valves ranging from 2 to 30 in. The majority of these valves control water flow. Recent studies done by the Institute of Nuclear Power Operations (INPO), Electric Power Research Institute (EPRI) and the US Nuclear Regulatory Commission (NRC) found that many of these safety-related valves were not functioning properly. Typical problems found in these valves included disk flutter, backstop tapping, flow leakage, disk pin and hinge pin wear, or even missing disks. These findings led to INPO’s Significant Operating Experience Report (SOER, 1986), and finally, NRC generic letter 89-04, which requires that all safety-related check valves in a nuclear plant be regularly monitored. In response to this need, the industry has developed various diagnostic equipment to monitor and test check valves, using technologies ranging from acoustics and ultrasonics to magnetic—even radiography has been considered. Of these, systems that depend on a combination of acoustic and ultrasonic techniques (Au-Yang et al., 1991) are among the most promising for two reasons: these two technologies supplement each other, making diagnosis of the check valves much more certain than any single technology, and this approach can be made nonintrusive. The nonintrusive feature allows the check valves to be monitored and diagnosed without being disassembled or removed from the piping system. This paper shows that by carefully studying the acoustic and ultrasonic signatures acquired from a check valve, either individually or in combination, an individual with the proper training and experience in acoustic and ultrasonic signature analyses can deduce the structural integrity of the check valve with good confidence. Most of the conclusions are derived from controlled experiments in the laboratory where the diagnosis can be verified. Other conclusions were based on test data obtained in the field.

Overview of Boiling Water Reactor Steam Dryer Alternating Stress Assessment Procedures

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Stephen A. Hambric ◽  
Samir Ziada ◽  
Richard J. Morante
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
High Cycle Fatigue ◽  
Structural Integrity ◽  
Fatigue Cracking ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Boiling Water ◽  
Cycle Fatigue ◽  
Regulatory Commission

The United States Nuclear Regulatory Commission (USNRC) has approved several extended power uprates (EPU) for Boiling Water Reactors (BWRs). In some of the BWRs, operating at the higher EPU power levels and flow rates led to high-cycle fatigue damage of Steam Dryers, including the generation of loose parts. Since those failures occurred, all BWR owners proposing EPUs have been required by the USNRC to ensure that the steam dryers would not experience high-cycle fatigue cracking. This paper provides an overview of BWR steam dryer design; the fatigue failures that occurred at the Quad Cities (QC) nuclear power plants and their root causes; a brief history of BWR EPUs; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluation methods (static and alternating stress).

scholarly journals Small Modular Reactor Control Room Workstation Demonstration

2019 ◽  
Vol 63 (1) ◽  
pp. 479-480
Author(s):  
Jessica Stevens ◽  
Kevin LaFerriere ◽  
Ryan Flamand NuScale
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Operating Experience ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Agile Development ◽  
Control Room ◽  
Small Modular Reactor ◽  
Regulatory Commission

A control room simulator was designed to model the operation of a NuScale small modular reactor (SMR) nuclear power plant and provide enough fidelity to perform staffing validation studies for Nuscale’s Nuclear Regulatory Commission Design Certification Application. The simulator serves as a simulated control room with work stations to mimic the operation of an SMR module, turbine generator, and support systems using a proprietary human system interface (HSI) software package. The simulator, which includes all HSI screens, was designed by a team of Human Factors and Plant Operations staff to capitalize on best practices, lessons learned, and operating experience using the Agile development process. Finally, the design process included the development of plant operating procedures and training material as well as a training platform for future plant operators at an SMR nuclear power plant.

Review of the Applicability of Nuclear Regulatory Commission Information Notice (IN) 2012-14, “Motor Operated Valve Inoperable Because of Stem-Disc Separation,” to Constellation Energy Nuclear Group’s Ginna Station

2014 ◽  
Author(s):  
David Garofoli ◽  
Gregg Joss
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Operating Experience ◽  
Nuclear Regulatory Commission ◽  
Corrective Action ◽  
Level Of Confidence ◽  
Valve Stem ◽  
The Subject ◽  
Regulatory Commission

U.S. Nuclear Regulatory Commission (NRC) Information Notice (IN) 2012-14, “Motor-Operated Valve Inoperable Because of Stem-Disc Separation”, was issued to inform nuclear power-plant licensees of recent operating experience involving a motor-operated valve (MOV) that failed at the connection between the valve stem and disc. The NRC expectation was that recipients would review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. Additional regulatory suggestions and insights contained in the IN are not NRC requirements. On closer examination of the events involved, it became apparent that the undetected stem-disc separation observed with the subject MOV was not necessarily limited to that type or style of valve. In fact, the vast majority of inservice testing (IST) valves, and the manner in which they are tested, could also be susceptible to loss of functionality going undetected. The intent of the compliance project performed at the R.E. Ginna Station nuclear power plant was to examine the current testing performed on each IST program valve and determine the level of confidence that stem-disc separation would be detected. If the level of confidence was deemed less than acceptable for a subject valve, one or more augmented actions, as deemed both practicable and viable, were recommended for implementation. The purpose of this presentation paper is to describe the systematic methodology that was employed to validate the effectiveness of the current periodic IST valve testing conducted at the R.E. Ginna Station and the corrective-action recommendations that were made as deemed appropriate. The corrective action(s) were designed to preclude the occurrence of future stem-disc separation issues going undetected, which could result in the loss of valve and potentially the loss of the associated accident-mitigation system’s operational readiness condition. Paper published with permission.

Evolution of Section XI of the ASME Boiler and Pressure Vessel Code

2000 ◽  
Vol 122 (3) ◽  
pp. 234-241 ◽  
Author(s):  
Owen F. Hedden
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
Operating Conditions ◽  
Nondestructive Examination ◽  
The Us ◽  
Regulatory Commission ◽  
Technical Basis ◽  
Year 2000

This article will describe the development of Section XI from a pamphlet-sized document to the lengthy and complex set of requirements, interpretations, and Code Cases that it has become by the year 2000. Section XI began as a set of rules for inservice inspection of the primary pressure boundary system of nuclear power plants. It has evolved to include other aspects of maintaining the structural integrity of safety class pressure boundaries. These include procedures for component repair/replacement activities, analysis of revised and new plant operating conditions, and specialized provisions for nondestructive examination of components and piping. It has also increased in scope to cover other Section III construction: Class 2, Class 3 and containment structures. First, to provide a context for the discussions to follow, the differences in administration and enforcement between Section XI and the other Code Sections will be explained, including its dependence on the US Nuclear Regulatory Commission. The importance of interpretations and Code Cases then will be discussed. The development of general requirements and requirements for each class of structure will be traced. The movement of Section XI toward a new philosophy, risk-informed inspection, will also be discussed. Finally, an annotated bibliography of papers describing the philosophy and technical basis behind Section XI will be provided. [S0094-9930(00)01703-0]

scholarly journals Inservice Examination and Testing Issues for Dynamic Restraints (Snubbers) in Nuclear Power Plants

2014 ◽  
Author(s):  
Gurjendra S. Bedi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Experience ◽  
Nuclear Regulatory Commission ◽  
Regulatory Commission ◽  
The U.S

This paper discusses recent issues related to inservice examination and testing of dynamic restraints (snubbers) at U.S. nuclear power plants. These issues were identified during the U.S. Nuclear Regulatory Commission (NRC) staff review of snubber examination and testing programs, relief requests, and applicable operating experience. This discussion includes information that could have generic applicability in the implementation of effective snubber programs at U.S. nuclear power plants. Paper published with permission.

Inservice Examination and Testing Issues for Dynamic Restraints (Snubbers) in Nuclear Power Plants

2017 ◽  
Author(s):  
Gurjendra S. Bedi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Experience ◽  
Nuclear Regulatory Commission ◽  
Regulatory Commission ◽  
The U.S

This paper discusses recent issues related to the inservice examination and testing of dynamic restraints (snubbers) at U.S. nuclear power plants. The U.S. Nuclear Regulatory Commission (NRC) staff identified these issues during its review of examination and testing snubber programs and relief requests, as well as operating experience. This discussion includes information that could apply generically to the implementation of effective snubber programs at U.S. nuclear power plants. Paper published with permission.

scholarly journals Demonstrating Structural Adequacy of Nuclear Power Plant Containment Structures for Beyond Design-Basis Pressure Loadings

2010 ◽  
Author(s):  
Joseph Braverman ◽  
Richard Morante ◽  
Charles Hofmayer ◽  
Robert Roche-Rivera ◽  
Jose Pires
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
Gas Generation ◽  
Design Basis ◽  
And Performance ◽  
Regulatory Commission ◽  
Technical Basis

Demonstrating the structural integrity of U.S. nuclear power plant (NPP) containment structures, for beyond design-basis internal pressure loadings, is necessary to satisfy Nuclear Regulatory Commission (NRC) requirements and performance goals. This paper discusses methods for demonstrating the structural adequacy of the containment for beyond design-basis pressure loadings. Three distinct evaluations are addressed: (1) estimating the ultimate pressure capacity of the containment structure (10 CFR 50 [1] and US NRC Standard Review Plan, Section 3.8) [2]; (2) demonstrating the structural adequacy of the containment subjected to pressure loadings associated with combustible gas generation (10 CFR 52 [3] and 10 CFR 50 [1]); and (3) demonstrating the containment structural integrity for severe accidents (10 CFR 52 [3] as well as SECY 90–016 [4], SECY 93–087 [5], and related NRC staff requirements memoranda (SRMs)). The paper describes the technical basis for specific aspects of the methods presented. It also presents examples of past issues identified in licensing activities related to these evaluations.

Uncertainties in NDE Reliability and Assessing the Impact on RI-ISI

2010 ◽  
Author(s):  
Steven R. Doctor ◽  
Michael T. Anderson
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Flaw Detection ◽  
Operating Experience ◽  
Nuclear Regulatory Commission ◽  
Regulatory Commission ◽  
The Impact ◽  
Procedure Development ◽  
Service Inspection ◽  
Fast Growth Rate

A major thrust in the past 20 years has been to upgrade nondestructive examinations (NDE) for use in inservice inspection (ISI) programs to more effectively manage degradation at operating nuclear power plants. Risk-informed ISI (RI-ISI) is one of the outcomes of this work, and this approach relies heavily on the reliability of NDE, when properly applied, to detect sources of expected degradation. There have been a number of improvements in the reliability of NDE, specifically in ultrasonic testing (UT), through training of examiners, and improved equipment and procedure development. However, the most significant improvements in UT were derived by moving from prescriptive requirements to performance based requirements. Even with these substantial improvements, NDE contains significant uncertainties and RI-ISI programs need to address and accommodate this factor. As part of the work that PNNL is conducting for the U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, we are examining the impact of these uncertainties on the effectiveness of RI-ISI programs. One of the primary objectives of in-service inspection, including a RI-ISI program, is to manage potential degradation that may occur, but that had not been foreseen through previous operating experience. However, RI-ISI programs in the U.S are primarily based on history, looking back at past failures in the operating fleet. Therefore, RI-ISI may not adequately manage degradation events that are yet to occur, such as those that may have a long incubation (initiation) time, but a potentially fast growth rate. For this reason, RI-ISI will always be reactive to such failure events. Successful ISI needs to determine what NDE is required, when and how frequently it needs to be applied, how effective the NDE must be and where the NDE needs to be applied. Both flaw detection and accurate characterization need to be addressed. This paper will examine the reliability and uncertainties of NDE, and how these may impact RI-ISI.

An Equipment Reliability Strategy and Maintenance Program for Improving the Reliability of Cranes at Commercial Nuclear Power Facilities

2008 ◽  
Author(s):  
Jeffrey B. Kriner ◽  
Bradford P. Lytle ◽  
John C. Lauri
Keyword(s):  
Nuclear Power ◽  
Operating Experience ◽  
Nuclear Regulatory Commission ◽  
Lost Generation ◽  
Equipment Reliability ◽  
Maintenance Program ◽  
Plant Equipment ◽  
Maintenance Activities ◽  
American Society ◽  
Regulatory Commission

Many commercial nuclear power facilities have been in operation well over 20 years, and many facilities have been or will have their original 40 year operating license renewed for an additional 20 years. The anticipated stresses to plant equipment and the longer service life increase the challenge to maintain reliable equipment performance. Establishing equipment maintenance programs that are effective and compliant with applicable regulations is critical to avoid unplanned equipment unavailability and the potential costs of lost generation. An equipment reliability (ER) strategy for commercial nuclear power plant equipment is described that considers the programmatic recommendations of the Institute of Nuclear Power Operations ([1], [2]), Electric Power Research Institute ([3], [4], [5], [6]), Nuclear Energy Institute standard nuclear business model [7], Nuclear Regulatory Commission ([8], [9], [10], [11]), and industry societies and working groups, such as the American Society of Mechanical Engineers ([12], [13]). All ER strategies must properly implement mandatory requirements and commitments ([14], [15], [16]). Additionally, ER strategies should also consider the appropriate manufacturer/vendor recommendations, industry and plant personnel operating experience feedback, equipment operating and maintenance history information, etc. As a result, the ER strategy includes reviewing multiple information sources to inform the decisions to either include or exclude the specific maintenance activities that impact reliability. Ultimately the maintenance program is tailored for each equipment application and implements the necessary maintenance activities while avoiding the cost of performing unnecessary maintenance activities.

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