Qualification of Pumps and Valves for the Safety Injection Path of Nuclear Power Plants

The reliability of a nuclear power plant depends on the safe functioning of its components during its lifetime: from design through construction, operation and maintenance. This is valid for new build projects as well as for the current fleet. As plants undergo modifications for increased performance or extended lifetimes, component integrity becomes a critical factor in those efforts, particularly for safety-related plant functions. This paper focuses on the qualification of pumps and valves of the safety-injection path, considering new requirements. Going back to the Barsebäck event in the year 1992, it is known that insulation material may cause clogging. Consequently, the presence of debris material in the water may have an impact on the functioning of pumps and valves. For this purpose, AREVA has built new thermo-hydraulic test loops in its accredited test and inspection body (according to International Organization for Standardization (ISO) 17025 and 17020) to consider this effect as it relates to components qualification (Ref. 1). The main relevant aspects of these tests will be discussed together with corresponding thermal shock tests. Paper published with permission.

Validation of Valve Leak Quantification With Non-Intrusive Acoustic Emission Technology

In the late 1980s and early 90s, several companies tested a range of acoustic devices for monitoring valve leakage during the check-valve diagnostic system research performed at the Utah State Water Research Laboratory as part of two separate nuclear-industry-sponsored initiatives. The acoustic sensor technology and analysis techniques evaluated were found helpful but no progress was made in non-intrusively quantifying the leak rate through the valves tested during these programs. Around that same time, oil & gas companies in the UK were experimenting with detection and quantification of valve leakage using acoustic emission (AE) technology. The AE sensors and signal-processing technology selected for the UK oil & gas effort responded to much higher frequencies compared to the sensors and systems used during the nuclear-utility initiative in the U.S. This research led to new products for detection and quantification of valve leakage in oil & gas applications. Because of minimum leak threshold and accuracy concerns, non-intrusive acoustic valve leak measurement has remained an elusive goal for commercial nuclear power. Various general-purpose acoustic tools have been trialed to detect leakage with mixed results because of complications caused by plant and system acoustic characteristics. Several of today’s moderately successful check-valve diagnostic systems employ acoustic sensors and can detect the most likely event representing flow cutoff when a check-valve disc fully closes, but leak-rate quantification with any of these systems is not possible. Correlation methods and other AE analysis techniques that have been developed to quantify leakage in steam systems have been generalized as small, medium, and large leakage classifications with no clear criteria for these levels. During the last couple of years, nuclear-plant engineers responsible for programs for compliance with Appendix J, “Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors,” to Part 50, “Domestic Licensing of Production and Utilization Facilities,” of Title 10, “Energy,” of the Code of Federal Regulations (Appendix J to 10 CFR 50) have made extensive use of a new acoustic valve leak-detection system known as MIDAS Meter®. Appendix J valve testing (also known as Type C testing) requires that sections of nuclear-plant piping be isolated by closing a number of valves, thereby creating a confined pressure boundary. The isolated piping within the boundary is pressurized with approximately 344.7 kilopascals (kPa) [50 pounds per square inch (psi)] of air and the leak-tightness of the boundary is evaluated. When the isolated piping exhibits excess leakage or cannot maintain the test pressure, the valves creating the boundary are evaluated one by one to find the culprit leaker. The process of finding and correcting the problem valve can take from hours to several days and may become an outage critical-path activity. Appendix J engineers have enjoyed considerable success with their newfound ability to quickly and confidently identify the leaking valves with MIDAS Meter® and remove their test programs from the critical path. MIDAS Meter® is a high-frequency acoustic-emission-based system which includes algorithms that convert the acoustic emission signal to leak rate. The basic algorithms were first developed from the field results obtained during the early development work for UK oil & gas operators and refined over the next 20 years. Though not originally validated under a quality-assurance (QA) program of the 10 CFR 50 type, nuclear plants that own MIDAS Meter® have been eager to go beyond simple troubleshooting and use the leak quantification results for nuclear applications, including safety-related decisionmaking. In order to support owners and avoid improper application of this very successful new tool, Score Atlanta embarked on an extensive validation program consistent with 10 CFR Part 50 requirements. A purpose-built leak-test flow loop and valve simulator apparatus were constructed in the Atlanta facility and testing began in early 2013. To support Appendix J users, the air testing was performed first and completed in July 2013. The water testing followed and should be completed in early 2014. Numerous combinations of leak path, leak-path geometry, and differential pressure were created and evaluated during the air phase of the program. Pressure was limited to 1034 kPa [150 psi] for air testing. The water testing includes pressures up to 8,618 kPa [1,250 psi] and a similar number of varying leak paths and pressure test points. This paper discusses the preliminary results of the test program, including any special limitations required for use of AE-derived valve leak results in nuclear safety-related applications. The full results of the test program and guidance for nuclear safety-related use of the technology are expected to be available ahead of the 2014 ASME-NRC Valve Symposium. Paper published with permission.

Overview of NRC NUREG-1482, Revision 2, Guidelines for Inservice Testing at Nuclear Power Plants: Inservice Testing of Pumps and Valves, and Inservice Examination and Testing of Dynamic Restrains (Snubbers) at Nuclear Power Plants

The U.S. Nuclear Regulatory Commission (NRC) staff issued Revision 2 to NUREG-1482, “Guidelines for Inservice Testing at Nuclear Power Plant,” to assist the nuclear power plant licensees in establishing a basic understanding of the regulatory basis for pump and valve inservice testing (IST) programs and dynamic restraints (snubbers) inservice examination and testing programs. Since the Revision 1 issuance of NUREG-1482, certain tests and measurements required by earlier editions and addenda of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) have been clarified, updated, revised or eliminated. The revision to NUREG-1482 incorporates and addresses those changes, and includes the IST programs guidelines related to new reactors. The revised guidance incorporates lessons learned and experience gained since the last issue. This paper provides an overview of the contents of the NUREG-1482 and those changes and discusses how they affect NRC guidance on implementing pump and valve inservice testing (IST) programs. For the first time, this revision added dynamic restraint (snubber) inservice examination and testing program guidelines along with pump and valve IST programs. This paper highlights important changes to NUREG-1482, but is not intended to provide a complete record of all changes to the document. The NRC intends to continue to develop and improve its guidance on IST methods through active participation in the ASME OM Code consensus process, interactions with various technical organizations, user groups, and through periodic updates of NRC-published guidance and issuance of generic communications as the need arises. Revision 2 to NUREG-1482 incorporates regulatory guidance applicable to the 2004 Edition including 2005 and 2006 Addenda to the ASME OM Code. Revision 0 and Revision 1 to NUREG-1482 are still valid and may continue to be used by those licensees who have not been required to update their IST program to the 2004 Edition including the 2005 and 2006 Addenda (or later Edition) of the ASME OM Code. The guidance provided in many sections herein may be used for requesting relief from or alternatives to ASME OM Code requirements. However, licensees may also request relief or authorization of an alternative that is not in conformance with the guidance. In evaluating such requested relief or alternatives, the NRC uses the guidelines/recommendations of the NUREG, where applicable. The guidelines and recommendations provided in this NUREG and its Appendix A do not supersede the regulatory requirements specified in Title 10 of the Code of Federal Regulations (10 CFR) 10 CFR 50.55a, “Codes and standards”. Further, this NUREG does not authorize the use of alternatives to, grant relief from, the ASME OM Code requirements for inservice testing of pumps and valves, or inservice examination and testing of dynamic restraints (snubbers), incorporated by reference in 10 CFR 50.55a. Paper published with permission.

Alternative Selection of Pump-Test Instrumentation and Variance of Fixed Reference Values

This paper describes actions being taken by the ASME (formerly the American Society of Mechanical Engineers) Operation and Maintenance of Nuclear Power Plants (OM Code) Subgroup on Pumps (ISTB) to revise pump test requirements to be more clearly stated and provide more flexibility in performance of tests. Specifically, this paper addresses two aspects of pump test requirements that are in the process of being changed. • Instrumentation requirements for measurement of hydraulic parameters (approved by ISTB subgroup). • Variance around the fixed reference value for establishing pump-test conditions (approved by ASME and published in the 2012 Edition of the OM Code). The OM Code changes discussed in this paper are currently in various stages of approval and endorsement. Therefore, the information provided in this paper is subject to change as a result of the OM Code approval and U.S. Nuclear Regulatory Commission (NRC) endorsement processes. Paper published with permission.

Preliminary Study of the Use of Freeze-Valves for a Passive Shutdown System in Molten Salt Reactors

The use of passive shutdown systems to enhance safety is one element of next-generation reactor design. The Freeze-Valve has been proposed as a key device in the passive system to stop the chain reaction of the Molten Salt Reactor (MSR), which has been chosen by Generation IV International Forum (GIF) as one of the six Generation IV reactor concepts. During reactor normal operation, the molten salt in the valve is cooled to a solid plug. In the event that the reactor overheats under accident conditions when all other active control systems fail, the plug will melt. The liquid fuel salt will be pulled out from the reactor core by gravity into dump tanks, and criticality will cease because the reaction is no longer moderated by the graphite in the reactor core. The more accurate the Freeze-Valve’s thermal design is, the more efficient the passive shutdown system becomes. In this study, an investigation of the thermal performance of the Freeze-Valve is conducted based on finite element methods verified by experimental data, and some modified designs are presented with recommendations. For further consideration, some innovative governing techniques used to control the Freeze-Valve are discussed in detail. Here, a more critical thermal design is focused on that can make the passive system shut down the nuclear reactor quickly and reliably. The Freeze-Valve can be used in the molten salt loop rather than a mechanical valve, which may become jammed by frozen salt. Paper published with permission.

Regulatory Perspectives on Risk-Informed Inservice Testing

In this paper, we review the various regulatory mechanisms that are available to licensees today for risk-informing their IST programs and that are acceptable to the NRC staff. These mechanisms have all been available for a decade or more, but have seen little interest or use. Paper published with permission.

Regulatory Perspectives on ASME OM Code, Mandatory Appendix IV, Revision 0, “Inservice Testing of Air-Operated Valves in Light-Water Reactor Nuclear Power Plants”

The American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) establishes the requirements for preservice and inservice testing and examination of certain components to assess their operational readiness in light-water reactor nuclear power plants. The Code of Federal Regulations (CFR) endorses and mandates the use of the ASME OM Code for testing air-operated valves in 10 CFR 50.55a(b)(3)(ii) and 10 CFR 50.55a(f)(4), respectively. ASME has recently approved Mandatory Appendix IV, Revision 0. NRC currently anticipates that Mandatory Appendix IV will first appear in the 2014 Edition of the ASME OM Code. Publication of the 2014 Edition of the ASME OM Code begins the NRC rulemaking process to modify 10 CFR 50.55a to incorporate the 2014 Edition of the ASME OM Code by reference. NRC staff has actively participated in the development of Mandatory Appendix IV, Revision 0, through participation in the ASME OM Code Subgroup on Air-Operated Valves (SG-AOV). The purpose of this paper is to provide NRC staff perspectives on the contents and implementation of Mandatory Appendix IV, Revision 0. This paper specifically discusses Mandatory Appendix IV, Sections IV-3100, “Design Review,” IV-3300, “Preservice Test,” IV-3400, “Inservice Test,” IV-3600, “Grouping of AOVs for Inservice Diagnostic Testing,” and IV-3800, “Risk Informed AOV Inservice Testing.” These topics were selected based on input received during NRC staff participation in the SG-AOV and other industry meetings. The goal of this paper is to provide NRC staff perspectives on the topics of most interest to NRC staff and members of the SG-AOV. Paper published with permission.

Implementation of Snubber Service-Life Monitoring Programs

A well-planned and -implemented service-life program which is properly used can reduce the need for extended testing and examination activities and can result in a cost-effective overall program. Service-life monitoring is an essential part of an effective snubber program, yet it is often the least detailed and most overlooked aspect. Because of the historical emphasis on examination and testing requirements, there has been little industry-wide consistency or emphasis on the specifics of service-life monitoring activities. This paper will identify the purpose and basis for snubber service-life requirements, as well as outline key elements of an effective program to both identify service-life values and monitor them over periods of extended plant operation. Included in the discussion will be topics such as: Identifying regulatory and code requirements, determining the scope of the program, establishing original service-life values, monitoring and evaluation, adjusting values, program documentation, and reporting. Identifying pertinent parameters for monitoring, appropriate methods for monitoring and trending, and incorporating condition monitoring and preventive-maintenance activities as alternatives to traditional programs will be discussed. Common challenges to implementing an effective program will be addressed, as well as some pitfalls to be avoided. Paper published with permission.

Regulatory Treatment of Nonsafety Systems in Passive Advanced Light-Water Reactors

Some new nuclear power plants have advanced light-water reactor (ALWR) designs with passive safety systems that rely on natural forces, such as density differences, gravity, and stored energy, to supply safety-injection water and to provide reactor-core and containment cooling. Active systems in such passive ALWR designs are categorized as nonsafety systems with limited exceptions. Active systems in passive ALWR designs provide the first line of defense to reduce challenges to the passive systems in the event of a transient at the nuclear power plant. Active systems that provide a defense-in-depth function in passive ALWR designs need not meet all of the acceptance criteria for safety-related systems. However, there should be a high level of confidence that these active systems will be available and reliable when challenged. Multiple activities will provide confidence in the capability of these active systems to perform their defense-in-depth functions; these are collectively referred to as the Regulatory Treatment of Nonsafety Systems (RTNSS) program. The U.S. Nuclear Regulatory Commission (NRC) addresses policy and technical issues associated with RTNSS equipment in passive ALWRs in several documents. This paper discusses the NRC staff’s review of pumps, valves, and dynamic restraints within the scope of the RTNSS program in passive ALWRs. Paper published with permission.

Inservice Testing Program Improvements for New Reactors: Small Modular Reactors

The nuclear industry is preparing for the licensing and construction of new nuclear power plants in the United States. Several new designs have been developed, including more traditional evolutionary designs, passive reactor designs, and small modular reactors (SMRs). ASME (formerly the American Society of Mechanical Engineers) provides specific codes used to perform inspections and testing, both preservice and inservice, for many of the components used in the new reactor designs. The U.S. Nuclear Regulatory Commission (NRC) reviews information provided by applicants related to inservice testing (IST) programs for design certification (DC) and combined license (COL) applications under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” of Title 10, “Energy,” of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code, Operation and Maintenance of Nuclear Power Plants, defines a post-2000 plant as a nuclear power plant that was issued (or will be issued) its construction permit, or combined license for construction and operation, by the applicable regulatory authority on or after January 1, 2000. The ASME New Reactors OM Code (NROMC) Task Group (TG) is assigned the task of ensuring that the preservice testing (PST) and inservice testing (IST) provisions in the ASME OM Code are adequate to provide reasonable assurance that pumps, valves, and dynamic restraints (snubbers) for post-2000 plants will operate when needed. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in nonsafety systems for passive post-2000 plants, including SMRs. (Note: For purposes of this paper, “post-2000 plant” and “new reactor” are used interchangeably throughout.) Paper published with permission.