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Inservice Testing of Dynamic Restraints

2014;():3-9. doi:10.1115/NRC2014-5033.
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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.

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Commentary by Dr. Valentin Fuster
2014;():10-18. doi:10.1115/NRC2014-5011.
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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.

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Commentary by Dr. Valentin Fuster
2014;():19-26. doi:10.1115/NRC2014-5029.
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Subsection ISTD of ASME’s Operation and Maintenance of Nuclear Power Plants (OM Code) is the required code for preservice and inservice examination and testing of dynamic restraints (snubbers). This code replaced the inspection requirements of Article IWF-5000, “Inservice Inspection Requirements for Snubbers,” in Section XI, “Inservice Inspection of Nuclear Power Plant Components,” of the ASME Boiler and Pressure Vessel Code after the publication of the 2006 addenda to Section XI, which deleted Article IWF-5000. When the requirements of IWF-5000 were deleted, the requirements for examination and testing of snubbers, as required by Section 50.55a, “Codes and Standards,” of Title 10, “Energy,” of the Code of Federal Regulations (10 CFR 50.55a) became those specified by Subsection ISTD of the ASME OM Code. Therefore, when nuclear power plant owners prepare their ten-year inservice testing (IST)/inservice inspection (ISI) program updates that incorporate the 2006 (or later) addenda to Section XI, the snubber requirements will be required to be in accordance with those of Subsection ISTD of the latest approved edition and addenda of the ASME OM Code (2004 Edition with Addenda through 2006). This edition of the ASME OM Code is cited in the NRC Rulemaking which was published on June 21, 2011.

Because this is a change in requirements, owners should be asking some of the following questions: What is the difference between our existing program requirements and those included in Subsection ISTD of the ASME OM Code? How will this change our existing program or the way the current snubber examination and testing program is implemented? How much effort will be required to implement this program change? This paper will provide some specific guidance for the implementation of the ISTD Code and will identify typical areas where changes may be required to existing snubber examination and testing programs. It will also describe some approaches to satisfy the requirements of ISTD-6000, “Service Life Monitoring,” which might not have been included in the previous requirements under Section XI.

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Commentary by Dr. Valentin Fuster
2014;():27-41. doi:10.1115/NRC2014-5017.
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Most power plant piping systems experience some level of vibration during operation. In some cases, the addition of restraints is required to keep operational vibration levels within acceptable limits. Vibration levels may need to be controlled to limit stresses in the piping or attached components, minimize wear on supports or other components, or prevent impacts with other piping or components. This paper discusses the use of dynamic restraints to effectively mitigate piping-system vibrations. Topics covered include: The effectiveness of using dynamic restraints for various sources of vibration; the effectiveness and applicability of various types of dynamic restraints; design considerations, including sizing and placement of dynamic restraints; and maintenance and testing considerations for dynamic restraints.

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Commentary by Dr. Valentin Fuster
2014;():43-52. doi:10.1115/NRC2014-5023.
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A number of industry and U.S. Nuclear Regulatory Commission (NRC) requirements exist for the quantification and qualification of piping-system vibrations. An ASME Operating and Maintenance (OM) Standard was written to provide methods for obtaining piping vibration measurements and to define acceptance criteria for the evaluation and qualification of the vibrations. Described herein is an overview of this standard, ASME OM-3, “Vibration Testing of Piping Systems,” along with discussions of the acceptance criteria (Reference 1).

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Commentary by Dr. Valentin Fuster

Valves I

2014;():55-62. doi:10.1115/NRC2014-5001.
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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.

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Commentary by Dr. Valentin Fuster
2014;():63-84. doi:10.1115/NRC2014-5016.
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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.

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Commentary by Dr. Valentin Fuster
2014;():85-92. doi:10.1115/NRC2014-5038.
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Excessive stem nut thread wear represents a potential common-cause failure mode that could impact all rising-stem valves. The consequence of unexpected failure of stem nuts emphasizes the importance of improving condition monitoring and maintenance practice activities by identifying, quantifying and minimizing stem nut thread wear.

In the nuclear industry, motor-operated valve (MOV) diagnostics estimate stem nut thread wear on safety-related valves using the stem-to-stem nut transition time (zero plateau). But the stem-to-stem nut transition time could also be affected by other variables that would lead to an inaccurate calculation of wear. Using stem-to-stem nut transition time to estimate wear, coupled with generally erring on the conservative side, usually indicates wear that is more severe than actual. This method, combined with all of the unknown variables, results in nuclear plants using valuable outage resources and dose to pull good stem nuts unnecessarily. This white paper identifies some of the variables that can be mistakenly construed as wear and offers a method that more accurately and efficiently validates this wear.

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Topics: Wear , Testing
Commentary by Dr. Valentin Fuster
2014;():93-103. doi:10.1115/NRC2014-5041.
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The “Consejo de Seguridad Nuclear” (Nuclear Safety Council), or CSN, is the nuclear regulatory body of Spain.

U.S. Nuclear Regulatory Commission (NRC) regulations and standards have been primarily used in the past up to the present. However, there is a process such that regulations recently generated in Spain replace or complement regulations coming from other countries. This is not the case with the evaluation and control of motor-operated valves (MOVs), which are mainly monitored using the process described in Generic Letters 89-10 and 96-05.

During a nine-month assignment from April to December 1989 at NRC offices in King of Prussia, PA, the author gained knowledge of NRC Bulletin 85-03 and of Generic Letter 89-10, which was issued in June 1989. The author realized the importance of these communications for improving the safety of the plants in the future if the issues they describe are adequately managed and solved.

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Commentary by Dr. Valentin Fuster

Regulatory Interactive Session

2014;():107-121. doi:10.1115/NRC2014-5002.
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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.

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Commentary by Dr. Valentin Fuster
2014;():122-130. doi:10.1115/NRC2014-5015.
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Motor-operated valves (MOVs) play an important role in the safe and reliable operation of today’s nuclear power plants. The purpose and scope of this paper is to review recent MOV operational experience events. The paper will discuss current findings and trends that relate to operation, maintenance, and surveillance testing of MOVs.

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Topics: Engines , Motors , Valves
Commentary by Dr. Valentin Fuster
2014;():131-137. doi:10.1115/NRC2014-5022.
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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.

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Commentary by Dr. Valentin Fuster
2014;():138-153. doi:10.1115/NRC2014-5030.
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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.

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Commentary by Dr. Valentin Fuster

Pumps and Valves II

2014;():157-176. doi:10.1115/NRC2014-5039.
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The reliability of check valves is paramount to the safe operation of plant systems. This paper provides a description of the benefits of applying advanced phased array techniques to establish the operational readiness of swing-check valves in the static or dynamic operational modes. In addition, a utility model perspective is described explaining how the Phased Array Sectorial Scanning (PASS) data assessments can be used to support operational predictive maintenance decisions. A collaborative effort between Arizona Public Service Co., the Palo Verde plants, and IHI Southwest Technologies, Inc. (IHI), was realized when IHI personnel applied Phased Array Sequence Scanning techniques to swing-check valves operating in the closed static position, as well as to three identical valves operating in the dynamic open positions in Palo Verde Units 1, 2 and 3 respectively.

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Topics: Valves
Commentary by Dr. Valentin Fuster
2014;():177-180. doi:10.1115/NRC2014-5018.
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This paper presents a discussion of the activities ongoing within the ASME (formerly the American Society of Mechanical Engineers) Operation and Maintenance of Nuclear Power Plants (OM) Code Subgroup on Air Operated Valves (SG-AOV), along with an overview of Revision 0 of Mandatory Appendix IV, “Preservice and Inservice Testing of Active Pneumatically Operated Valve Assemblies in Light-Water Reactor Power Plants.”

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Commentary by Dr. Valentin Fuster
2014;():181-191. doi:10.1115/NRC2014-5028.
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Kalsi Engineering, Inc. (KEI), initiated an independent test program that includes a number of actuator manufacturers, models, and sizes based on a survey of United States (U.S.) nuclear power plants. The test matrix includes evaluation of the effect of the key parameters on the effective diaphragm area (EDA) throughout the stroke. These parameters include stroke position, pressure, materials, measurement uncertainty, and manufacturing tolerances. Because of differences in the test data obtained by different sources for the same actuator type and size, systematic test procedures have been developed by KEI to address differences in the testing methods and test configurations, including testing of a balanced actuator (no spring in the actuator) vs. a spring-return actuator of the same diaphragm size. The effect of elevated temperature and aging may also be included later by testing a selected number of actuators based on industry input. The benefit of this program is to provide reliable data for air-operated valve (AOV) design-basis evaluations as required by U.S. Nuclear Regulatory Commission (NRC) Regulatory Issue Summary (RIS) 2000–03. This paper presents the results for the Masoneilan Model 38 Size 18 diaphragm actuator, which show that EDA is both position- and pressure-dependent.

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Commentary by Dr. Valentin Fuster
2014;():192-207. doi:10.1115/NRC2014-5031.
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Common quarter-turn (QT) mechanisms used in nuclear plant air-operated valves (AOVs) include scotch yoke, lever, and link-and-lever mechanisms coupled to diaphragms and pistons. QT mechanism efficiency varies as a function of valve position and is a critical design input used to determine AOV margin. Because of the lack of publicly available data of a quality commensurate with “nuclear QA [quality assurance],” Kalsi Engineering, Inc. (KEI), initiated an independent QT-mechanism efficiency test program that includes a number of commonly used actuator manufacturers, models, and sizes based on a survey of U.S. nuclear power plants.

The first test specimen was a diaphragm actuator with a lever QT mechanism. The diaphragm rod of the test specimen was instrumented with strain gauges so that a direct measurement of the net actuator force transmitted to the QT mechanism could be measured. In addition to the net thrust, the output torque, diaphragm pressure, and actuator position were measured. Measuring the net thrust, diaphragm pressure, and position allowed the spring rate, spring preload, and effective diaphragm area to be quantified.

This test specimen was tested using two different types of bearings at the actuator shaft-to-lever connection. Needle bearings were used to provide torque results for a nearly frictionless QT mechanism, and bronze bearings were used to simulate a more realistic QT-mechanism configuration. Predictions made using the first-principles efficiency model are compared to efficiencies extracted from test. The predicted efficiency using a realistic range for the friction coefficient of the bronze bearings is in good agreement with the extracted efficiencies.

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Commentary by Dr. Valentin Fuster
2014;():208-222. doi:10.1115/NRC2014-5024.
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A high-temperature molten salt pump, described as mechanical, free-surface, centrifugal, vertical-shaft, sump type, working at 500°C [932°F] to 600°C [1112°F], has been developed for the Thorium-Based Molten Salt Reactor (TMSR). Flow passage components of the pump are made of Hastelloy C-276 to ensure sufficient strength and corrosion resistance at high temperature. Also, a heat shield plug with air-cooled channels was designed to separate the drive motor, seal elements, and bearings from intense radioactivity and to keep the temperature of the flange seal below 150°C [302°F] and the temperature of the bearing below 80°C [176°F]. A dry gas seal was used so that there is zero leakage. Furthermore, some analysis of hydraulics characteristics, temperature field, thermal stress, and strain was performed to research the pump’s performance, and then the temperature field and the hydraulics were measured to validate the analysis results. The results show that the hydraulics, thermal stress, and strain meet the design value very well. The pump has been successfully operated on a LiF-NaF-KF test loop for over 250 hr. at temperatures of 500°C [932°F] to 600°C [1112°F], speeds of 1050 to 1450rpm, and flows of 15 m3/h [66 gpm] to 25 m3/h [110 gpm].

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Commentary by Dr. Valentin Fuster
2014;():223-237. doi:10.1115/NRC2014-5037.
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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.

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Commentary by Dr. Valentin Fuster

Inservice Testing Software

2014;():241-257. doi:10.1115/NRC2014-5003.
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The nuclear utility Inservice Testing (IST) Program Engineer is faced with a number of challenges on a daily basis. The burden of additional responsibility has been increasing dramatically in recent years because of personnel departure without staff replacement. This condition has increased the reliance on software to assist the program engineer in performing their assigned duties.

IST Program administration and implementation software can be the IST Program Engineer’s “best friend” when properly used. Newer Web-based software such as the Engineering Programs EP-Plus ENGAGE™ IST Software is designed to meet and exceed the requirements of ASME (formerly the American Society of Mechanical Engineers) Operations and Maintenance of Nuclear Power Plants (OM Code) and regulatory expectations while simultaneously improving the IST Program Engineer’s “quality of life.”

Web-based, fully supported, readily accessible IST software is one of the IST Program Engineer’s most important tools. The EP-Plus ENGAGE Software Suite is aligned with corporate strategies and objectives by using the latest technology to improve the way that engineering programs are managed. The standardized common platform can be cost-effectively expanded to include any engineering program at the unit, station, or fleet level.

The EP-Plus ENGAGE Suite covers the full range of engineering programs and provides the opportunity to realize benefits in an expedited manner while also allowing improvement in efficiency and cost reduction.

A series of figures and associated narrative text is used to provide the IST Program Engineer with information pertaining to the features and functionality of today’s IST software and associated engineering-program software solutions.

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Commentary by Dr. Valentin Fuster
2014;():258-268. doi:10.1115/NRC2014-5013.
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Trending, standard deviation, forecasting, linear regression, anomaly detection, and normalization are all techniques that IST Program Engineers use as tools to help them analyze and evaluate test results. But does the program engineer remember what these terms mean? Are they being used correctly? Are the techniques being used when they need to be? Are their benefits being fully used? This paper briefly goes over what each of these terms mean and looks at examples of each. Knowing how to use these tools is integral to a quality IST Program. Some software, such as True North Consulting’s EP-Plus Engage IST Software, incorporates these “value added” techniques. A brief overview of how they are used as tools for the IST Engineer is provided. Also included is a discussion on how these tools can be improved on and leveraged to decipher the program engineer’s IST Program data.

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Commentary by Dr. Valentin Fuster
2014;():269-278. doi:10.1115/NRC2014-5014.
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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 (IST) and examination of certain components to assess their operational readiness in light-water reactor nuclear power plants. The Code of Federal Regulations (CFR) endorses the use of the ASME OM Code in 10 CFR 50.55a(b)(3) . This paper focuses on applicable regulatory requirements and regulatory perspectives associated with the use of IST software in the nuclear industry.

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Commentary by Dr. Valentin Fuster
2014;():279-294. doi:10.1115/NRC2014-5027.
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This paper focuses on Inservice Testing (IST) software, in particular software developed by Kalsi Engineering, Inc. (KEI), to assist plant personnel in implementing IST requirements. The requirements in Mandatory Appendices II, III, and IV to ASME’s Operation and Maintenance of Nuclear Power Plants (OM Code) for check-valve condition monitoring, motor-operated valve (MOV) inservice testing, and air-operated valve (AOV) inservice testing, respectively, are identified. Each requirement in Appendix II is mapped to specific functionality in the condition-monitoring software. In addition, methods used in the design-analysis software for design-basis verification, trending of diagnostic test data, and functional margin determination to satisfy the requirements in Appendices III and IV are also described and mapped to specific requirements.

Conditioning-monitoring management database software is designed to comprehensively meet the documentation and trending requirements of Appendix II. The software addresses all program aspects, including valve grouping, program analysis, development of optimization and performance-improvement activities, evaluation of test and inspection intervals, trending and feedback, and corrective maintenance. To achieve this functionality, the condition-monitoring software includes check-valve design information, condition-monitoring activity setup and test history, trending of test parameters, a repository for miscellaneous data, and documentation of expert panel reviews. Database query tools and hard-copy reports are also provided.

The benefit is that this software provides a standardized, central collection point from which plant engineers can effectively manage their Appendix II program without having to develop and maintain a multitude of non-standardized personal spreadsheets or databases. In addition, the software assists in succession planning and minimizes the transition time for new check-valve program owners.

The MOV and AOV design-analysis software determines initial design margin and then uses field test data and associated uncertainties to determine actual setup functional margin. In addition, trending of diagnostic in-service test data is performed to verify that design values are conservative and that the inservice test intervals are appropriate based on projected degradation rate. The benefit of this software is that it assists utility engineers in satisfying the design-basis verification, preservice, and inservice testing requirements in Appendices III and IV in a standardized and comprehensive platform.

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Commentary by Dr. Valentin Fuster

New Reactors

2014;():297-304. doi:10.1115/NRC2014-5004.
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Several new reactors are currently under construction in the USA. Based on current construction schedules, Watts Bar 2 will be the first new reactor to go online for commercial generation since Watts Bar 1 was issued its operating license in 1996. New engineering programs will be going online with new reactors like Watts Bar 2. The startup of these new engineering programs is not without its own set of challenges. One of the programs has undergone a significant transformation since the last nuclear power plant started commercial operation in terms of industry implementation methods and regulatory requirements.

In 1996, the NRC issued Generic Letter 96-05 to communicate issues related to periodic verification (PV) of motor-operated valves (MOVs) and to request action by operating commercial power reactors to establish an MOV PV program. Subsequently, the regulations were revised to include a requirement to have an MOV PV program in Title 10, “Energy,” of the Code of Federal Regulations (10 CFR) 50.55a(b)(3)(ii). Generic Letters 89-10 (on MOV surveillance and testing) and 96-05 have been closed and today stand as historical references. Their provisions do not directly apply to new reactors, but there are many lessons available from MOV PV programs at operating sites in terms of safety, implementation, and cost.

There is only one consensus standard available to describe the requirements for an acceptable MOV PV program. This is contained in the ASME’s Operation and Maintenance of Nuclear Power Plants (OM Code) as Mandatory Appendix III. The U.S. Nuclear Regulatory Commission (NRC) previously endorsed this approach as a Code Case and is preparing a proposed change to 10 CFR 50.55a to incorporate by reference the ASME OM Code edition that includes Appendix III. This paper conveys the technical complexities and financial concerns faced by plant staff in making the right technical decisions for new program implementation at a new reactor in the USA.

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Topics: Engines , Motors , Valves
Commentary by Dr. Valentin Fuster
2014;():305-315. doi:10.1115/NRC2014-5010.
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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.

Commentary by Dr. Valentin Fuster
2014;():316-324. doi:10.1115/NRC2014-5012.
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The ASME (formerly the American Society of Mechanical Engineers) code titled Operation and Maintenance of Nuclear Power Plants (OM Code), Division 1, “Section IST: Rules for Inservice Testing of Light-Water Reactor Power Plants,” defines a post-2000 plant as a nuclear power plant that was issued (or will be issued) its construction permit or its combined license for construction and operation by the applicable regulatory authority on or after January 1, 2000. The New Reactors OM Code (NROMC) Task Group (TG) of the ASME OM Code Committee 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. The NROMC TG has prepared updated guidance for pumps and pyrotechnically operated (squib) valves in new reactors that has been incorporated in the ASME OM Code. Currently, the NROMC TG is preparing proposed guidance for surveillance of safety-significant pumps, valves, and dynamic restraints in nonsafety systems at post-2000 plants that employ passive safety-related post-accident heat-removal systems (referred to as passive post-2000 reactors).

The NROMC TG is also evaluating pump and valve surveillance provisions that would ensure that safety-significant components in small modular reactor (SMR) designs are verified to be operationally ready while providing flexibility to accommodate potentially extended refueling cycles associated with these post-2000 plants. There are several other changes and evaluations being performed by the NROMC TG to provide reasonable assurance of the operational readiness of safety-significant pumps, valves, and dynamic restraints while still weighing the cost-effectiveness of the requirements. The NROMC TG also considers risk insights in its evaluation of PST and IST provisions for post-2000 plants. This paper discusses the NROMC TG activities to develop the recent and planned improvements to the PST and IST provisions in the ASME OM Code to address pumps, valves, and dynamic restraints in new reactors.

Paper published with permission.

Commentary by Dr. Valentin Fuster
2014;():325-341. doi:10.1115/NRC2014-5025.
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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.

Commentary by Dr. Valentin Fuster
2014;():342-355. doi:10.1115/NRC2014-5032.
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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.

Commentary by Dr. Valentin Fuster
2014;():356-373. doi:10.1115/NRC2014-5040.
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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.

Commentary by Dr. Valentin Fuster

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