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ASME O&M Scope

2017;():V001T01A001. doi:10.1115/PVS2017-3505.
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Throughout the world, IST Programs are usually required by the Regulatory Body that holds authority over the site. IST programs have several sources. Typically, these include IST Codes and Standards, Plant Technical Specifications, Final Safety Analysis Report and, should the plant have developed it, the Probability Risk Assessment. Rulemaking clarifications, modifications and requirements play a key role connecting all applicable documentation. In Spain, the Spanish Regulatory Body, CSN (Consejo de Seguridad Nuclear) requires all NPPs to develop and implement an IST Program according to the Codes and Standards of the country of design origin. As a result, all Spanish NPPs that have been designed in the US follow a 10CFR50.55a and ASME OM IST-based approach. In order to be able to operate, Spanish NPPs must have an official document called “MISI”, which stands for Manual of In-Service Inspection. The scope of this manual is wide: at the very least MISIs include in their scope ASME OM, ASME Section XI and Appendix J requirements. In this presentation, we explain how we intertwine NRC Regulations with our Regulatory Body’s Regulations, and applicable Codes and Standards specially focusing on ASME OM ISTA requirements.

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Commentary by Dr. Valentin Fuster
2017;():V001T01A002. doi:10.1115/PVS2017-3508.
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The Terry turbine is a small, single-stage, compound-velocity impulse turbine originally designed and manufactured by the Terry Steam Turbine Company purchased by Ingersoll-Rand in 1974. Terry turbines are currently manufactured and marketed by Dresser-Rand. Terry turbines were principally designed for waste-steam applications. Terry turbopumps are ubiquitous to the US nuclear fleet as a steam driven turbopump in either the reactor core isolation cooling system (RCIC) and high pressure coolant injection systems for boiling water reactors (BWRs) or in the auxiliary feedwater system (AFW) system for pressurized water reactors (PWRs).

Prior to the accidents at Fukushima Daiichi, assumptions and modeling of the performance of Terry turbopumps were based mostly on generic vendor use of NEMA SM23 Steam Turbine for Mechanical Drive Service guidance [1]. However, the RCIC/AFW system performance (i.e., the Terry turbopump) under beyond design basis event (BDBE) conditions is poorly known and largely based on conservative assumptions used in probabilistic risk assessment (PRA) applications. For example, common PRA practice holds that battery power (DC) is required for RCIC operation to control the vessel water level, and that loss of DC power results in RCIC flooding of the steam lines and an assumed subsequent failure of the RCIC turbopump system. This assumption for PRA implies that RCIC operation should terminate on battery depletion which can range from 4 to 12 hours. In contrast, real-world observation from Fukushima Daiichi Unit 2 shows that RCIC function was not terminated by uncontrolled steam line flooding or loss of control power, and in fact provided coolant injection for nearly three days [2].

There is a current effort being undertaken by the US industry, the US Department of Energy (DOE), and the Government of Japan to investigate the true operating band of the Terry turbopump for BDBE conditions. This paper provides a summary of the experimental and modeling efforts to date.

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Commentary by Dr. Valentin Fuster
2017;():V001T01A003. doi:10.1115/PVS2017-3533.
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The O&M Code was developed when it was decided to move Pump and Valve Inservice Testing (IST) Requirements from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI to a standalone Code. The Code review process structure at the time was quite small and generally consisted of changing Section XI Subsections IWP and IWV into OM language. At the same time, new testing techniques were being developed that included check valve condition monitoring and current trace testing of motor actuated valves. This necessitated adding groups that were specific to these new initiatives.

Although that was several decades ago, these groups remained and, over the years, it was identified that actions, such as Inquiries, were taking much too long to process. This became abundantly clear with the development of the newly published Mandatory Appendix IV for Air Operated Valve Testing.

This paper discusses how the Code Committee became the organization that it is and how a new realignment will streamline the Code process and make it more efficient and responsive to the industry/regulatory needs.

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Commentary by Dr. Valentin Fuster
2017;():V001T01A004. doi:10.1115/PVS2017-3534.
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The O&M Code was developed when it was decided to move Pump and Valve In-Service Testing Requirements from ASME Boiler and Pressure Vessel Code Section XI to a standalone Code. IST for Pumps was originally is ASME Section XI IWP and for Valves IWV. Safety and Relief Valves were a Power Test Code and not in the scope of the ASME Boiler and Pressure Vessel Code. IWP and IWV were developed after plants had been designed and built. The desire was that no back-fits were to be required to comply with IST requirements. After the 1986 Edition, IWP and IWV requirements were moved into O&M. Appendix 1 of OM was what used to be the Power Test Code.

While this was going on, the NRC issued what has been called “the Richardson Letter”. Among other things, that letter required that IST for pumps better asses the condition of the pumps by putting higher accuracy instrumentation on the test pipe. For many plants, this was the minimum recirculation pipe. Over the course of time, the committee was able to get agreement that if a centrifugal pump were tested “back on its curve” increased instrument accuracy would be meaningless. This was the genesis of what we now call comprehensive Pump testing. Additionally, there were several alternative methods for valve testing that had been developed. It became clear, that simple periodic stroke timing of a power operated valve was simply not adequate for detecting degrading performance.

This presentation will discuss how Pump and Valve In Service Testing evolved to what it is today and discuss what might be alternatives in the future.

I want to thank Robert Parry, who provided some insights into this presentation specifically where my memory needed a bit of jogging.

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Commentary by Dr. Valentin Fuster
2017;():V001T01A005. doi:10.1115/PVS2017-3535.
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In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs.

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Commentary by Dr. Valentin Fuster

Pumps

2017;():V001T02A001. doi:10.1115/PVS2017-3507.
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Smooth Running Pumps have been an Industry issue since 1988. This caused many nuclear plants to obtain a Relief Request to use alternative requirements than those specified in the tables in subsections ISTB or ISTF of the ASME OM Code. Code Case OMN-22, Smooth Running Pumps, was approved in January 2017 for pumps with very low reference value vibration levels. This Code Case specifies the alternative requirements that may be used in lieu of the applicable Code subsections.

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Topics: Pumps
Commentary by Dr. Valentin Fuster
2017;():V001T02A002. doi:10.1115/PVS2017-3509.
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This presentation provides an overview of condition monitoring for rotating equipment in nuclear power plants. Specific condition monitoring technologies addressed include vibration analysis, lube oil analysis, thermography, and motor current signature analysis. Plant and equipment parameters, such as motor electrical and plant process parameters, useful for evaluating equipment condition, are also identified. The technologies are examined based on availability, cost effectiveness, and importance to a condition monitoring program. Although vibration analysis and oil analysis are the primary emphasis for performing condition monitoring, the inter-relationships between the technologies, techniques and other readily available plant data explored here demonstrate how a more complete and accurate diagnosis of the condition of a machine set can be determined. A discussion of each technology will include the various machine set faults which the technology will identify, as well as, how the overlapping technologies improve the effectiveness of a condition monitoring program.

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Commentary by Dr. Valentin Fuster
2017;():V001T02A003. doi:10.1115/PVS2017-3512.
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What appears to be a simple question is often quite difficult to answer depending on the quantity of flow; and size, type, and location of piping. Even the reason for asking the question can be varied and complex — ranging from environmental regulation, investment decisions, aging infrastructure improvement planning, and new equipment evaluation. Absolute field performance testing of power plant equipment yields valuable data that can be used in a variety of ways.

National and International codes list several methods to measure water flow in a performance application and provide realistic uncertainty estimates. Codes and standards exist for equipment evaluation and contractual performance tests. These Codes, though, are sometimes viewed as costly or perceived to impose additional risk on suppliers. Herein, we will present how to obtain performance test data and how that data can be used.

In many rehabilitation or regulation driven projects, an accurate representation of the state of the existing power plant is desired. Pump curves typically do not represent an accurate depiction of flow due to equipment degradation, changes in system components/geometry, and/or bio-fouling. While the testing may be considered costly, it can often be justified as part of a rehabilitation project. Absolute testing provides a lower uncertainty that can yield more definitive estimates of return on investment to justify projects that might be otherwise considered marginal.

Case studies will be discussed that illustrate these points, including:

• Flow measurement feasibility and site testing at a nuclear thermal plant

• In-situ flow testing to calibrate existing ultrasonic flow meters at a biomass thermal plant

• Condenser performance testing at a nuclear thermal plant

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Commentary by Dr. Valentin Fuster
2017;():V001T02A004. doi:10.1115/PVS2017-3514.
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Members of ISTB and ISTA have developed a list of issues regarding ISTB Pump Implementation that have been identified during IST Program Reviews, Day to Day operation of IST Programs, and site assessments of IST Programs (including issues found during updates). Implementation of the ISTB requirements for pumps in commercial US plants has presented challenges over the last few years with all the changes in ISTB since the issuance of the 1995 Edition of the ASME OM Code through the 2006 Addenda. This paper will discuss issues related to IST implementation. Good practices will be discussed to help IST engineers improve their programs. Some projected issues with implementation of 2012 Edition Appendix V will also be discussed.

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Topics: Pumps
Commentary by Dr. Valentin Fuster
2017;():V001T02A005. doi:10.1115/PVS2017-3529.
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Three (3) Component Cooling Water (CCW) pumps, IR 8X18SE, appeared to have been operated beyond their original manufacturer pump curves without proper justification and analysis. The testing was performed to determine exactly where the pumps were operating relative to the original head-capacity curve, at different system demands. Additionally, the results would be compared to the customers flow and pressure measuring capabilities.

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Topics: Cooling , Pumps , Water
Commentary by Dr. Valentin Fuster

Motor-Operated Valves (MOVs) and Air-Operated Valves (AOV)

2017;():V001T03A001. doi:10.1115/PVS2017-3503.
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Calculating margin for valve operation under design basis conditions requires evaluation of the stem loads required to operate the valve and the load capability of the actuator. These evaluations require justified and validated methodologies with verified inputs to implement the methodologies. The lack of validated methodologies in the past led to plant events and issues that prompted three NRC generic letters for MOVs and numerous generic correspondence documents from the NRC on AOV and MOV performance. Over the past 25 years, EPRI has performed extensive research to better understand the performance of valves and power operators. This research has been used to develop predictive methods for the evaluation of valve required operating loads and actuator output capability.

This paper summarizes EPRI’s research related to the development of predictive methodologies for valves and power operators and methods that are available, specifically methods for:

• Predicting required operating loads under design basis conditions,

• Predicting actuator output capability,

• Addressing thermal binding of gate valves, and

• Addressing the rate-of-loading phenomenon for MOVs.

This paper also describes a recent project to develop and validate a method for predicting the required thrust to overcome friction between the valve disk and body due to disk side-loading in cage-guided balanced disk globe valves.

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Commentary by Dr. Valentin Fuster
2017;():V001T03A002. doi:10.1115/PVS2017-3504.
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ASME OM Code Mandatory Appendix III (Appendix III) [1] for inservice testing of motor-operated valves (MOVs) contains prerequisites for a design basis verification test (DBVT) and preservice test prior to initiating inservice testing. The DBVT has specific requirements that depend on valve type and operational experience and the preservice test must adequately bridge the DBVT and inservice test. In addition, certain replacement, repair, or maintenance activities require an evaluation to determine what aspects (if any) of the DBVT or preservice test require repeat testing and/or engineering analysis to either confirm existing reference values or establish new reference values. Finally, existing testing performed under legacy NRC Generic Letter (GL) 89-10/96-05 MOV Programs or ASME QME-1 functional qualification standard may be credited to satisfy all or a portion of the DBVT and preservice test.

The purpose of this paper is to describe, by valve type:

• the specific requirements for the DBVT and preservice test,

• the use of previous qualification testing (e.g. GL 89-10/96-05 and ASME QME-1) to satisfy the DBVT and preservice test requirements,

• what activities may require analysis and/or repeating portions of the DBVT and preservice testing, and

• applicability to legacy MOV programs.

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Topics: Design , Testing
Commentary by Dr. Valentin Fuster
2017;():V001T03A003. doi:10.1115/PVS2017-3513.
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Electric Power Research Institute (EPRI) contracted Kalsi Engineering, Inc (KEI) to perform flow loop testing, computational fluid dynamic (CFD) analyses, and methodology development to more accurately predict flow-induced forces in balanced globe valves. The flow loop test conditions included single-phase and two-phase water flow, straight pipe and upstream-flow-disturbance pipe configurations, and two 4-inch balanced disk globe valves test specimens with a combination of quick opening and linear trim. CFD predictions were performed with a commercial grade dedicated version of ANSYS CFX 16.0 software.

The methodology was developed to utilize key dimensional characteristics of the disk and cage to determine the effective area through the stroke. The methodology accounts for trim characteristics, flow orientation, disk style, maximum valve DP and maximum flow rate. The model is validated for fluid temperatures between 70 °F and 160 °F, flow velocities up to 45 ft/sec. The methodology was validated against flow loop test data over a range of flow conditions, disk styles, and trim characteristics.

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Topics: Valves , Disks
Commentary by Dr. Valentin Fuster
2017;():V001T03A004. doi:10.1115/PVS2017-3518.
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Stem friction in an operating valve is a function of the dynamic interaction of a number of variables — packing material of construction, number of packing rings, compressive load, lubrication, stem surface finish, temperature, cycling, etc. Forces due to friction can be reduced by modifying these factors. Attaining low actuation force and good sealing requires a balanced approach. Packing manufacturers have their own procedures for determining the frictional properties of different packing materials. This paper will show one such procedure and how varying materials and packing set configurations affect actuation force. The focus will be on linear reciprocating valve stems.

The equation F = π × d × H × GS × μ × Y can be used to calculate the force of the packing on the valve stem: Where F - Force needed to overcome packing friction; d - Stem diameter; H - Packing set height; GS - Compressive stress on the packing; μ - Packing coefficient of friction; Y - Ratio of radial to axial load transference, commonly equal to 0.50. Knowing the force, F, by test allows the calculation of the packing set’s frictional characteristics. . This knowledge can guide valve designers and builders to properly size actuating units for consistent and reliable valve performance.

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Commentary by Dr. Valentin Fuster
2017;():V001T03A005. doi:10.1115/PVS2017-3523.
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Motor-Operated Valves (MOVs) in nuclear plants have an equipment qualification (EQ) life of 2000 cycles. Historically, nuclear generating stations have not counted actuator cycles. As nuclear plants age beyond 30 years and licenses are extended from 40 to 60 years, the accumulated actuator operational cycles have become an issue. Without hard data, plants are not able to “prove” that actuator fatigue cycle count is less than 2000 cycles. The only viable option to address this issue is actuator replacement. Actuator replacement is expensive and frequently require long lead times for procurement. Nuclear plants generally have fifty to one hundred qualified actuators and replacing all qualified actuators is prohibitively expensive.

The questions is simply how to deliver the nuclear promise without wholesale replacement of actuators and expending significant capital or O&M dollars.

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Topics: Fatigue , Actuators , Cycles
Commentary by Dr. Valentin Fuster
2017;():V001T03A006. doi:10.1115/PVS2017-3525.
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This paper is about Curtiss Wright’s electro-hydraulic actuator environmental and seismic qualification for Main Steam Isolation Valve (MSIV) and Main Feedwater Isolation Valve (MFIV) applications. The qualification was performed in compliance with IEEE-382 and RCC-E requirements qualifying the actuator for US, Chinese and European power plant designs. The qualification entailed several challenges and application of analytical and test methodologies. The weight of the actuator/yoke assembly made seismic qualification one of the most challenging steps in the program. The seismic qualification was performed jointly with the Areva US Technical Center in Lynchburg, VA. The qualification program was designed to envelope the requirements power plant designs in US, China and Europe.

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Commentary by Dr. Valentin Fuster
2017;():V001T03A007. doi:10.1115/PVS2017-3526.
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To prepare for implementation of ASME OM Code [1] Mandatory Appendix III (Appendix III) for inservice testing of motor-operated valves (MOVs), Tennessee Valley Authority (TVA) performed a comprehensive assessment at all three of their nuclear sites to identify gaps between their legacy IST and MOV programs and an IST program that meets the requirements of Appendix III. This assessment reviewed each paragraph of Appendix III and T VA governing documents to determine how the requirements are already being met or are missing in the legacy MOV program(s). Secondly, the assessment performed a high level overview of TVA’s MOV programs in response to NRC Generic Letters 89-10 [6] and 96-05 [7] and identifies areas for improvement for TVA consideration. This paper presents the assessment purpose and objectives, scope, approach and methods, references, summary of significant gaps, and proposed actions to resolve these gaps prior to Appendix III implementation.

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Commentary by Dr. Valentin Fuster
2017;():V001T03A008. doi:10.1115/PVS2017-3537.
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GE contracted Kalsi Engineering, Inc. (KEI) to perform actuator testing to determine the effective diaphragm area for the Model 37/38 actuator line and to develop a bounding effective diaphragm area tolerance to account for measurement uncertainties and manufacturing tolerances.

The GE sponsored test matrix includes Model 37/38 Sizes 9, 11, 13, 15, 18, and 24 actuators. The test matrix was primary defined to provide EDA data for actuators used in US nuclear power plants. The test matrix was primarily designed to facilitate the evaluation of the effects of stroke position, pressure, diaphragm materials, and measurement uncertainty. The test matrix also included with and without spring test configurations, two spring options for the same actuator size and model, and two diaphragm materials: Nitrile Elastomer and Silicone.

The test program provides reliable data for AOV design basis evaluations as required by the NRC RIS 2000-03. This paper presents the results for the Masoneilan Model 38 Size 11 diaphragm actuator, which show that EDA is strongly position dependent and weakly pressure-dependent.

As part of the project, a method for determining the required EDA tolerance to account for manufacturing variations was developed, which allows EDA determined by testing to be used across the product line.

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Commentary by Dr. Valentin Fuster

Valves

2017;():V001T04A001. doi:10.1115/PVS2017-3501.
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In verifying PRV setpoints, it is important to distinguish if there is any differential (±) between the measured SP (Set Pressure) of a PRV when tested on water versus testing on other fluids such as Diesel Fuel or Lubricating Oil. It is also important to recognize the standard test medium used by the PRV industry for liquid service testing is water. SP testing with other fluids involves issues such as possible serious health and safety as well as equipment cross contamination.

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Commentary by Dr. Valentin Fuster
2017;():V001T04A002. doi:10.1115/PVS2017-3536.
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A previous paper [1] describes the key features of an innovative gate valve design that was developed to overcome seat leakage problems, high maintenance costs as well as issues identified in the Nuclear Regulatory Commission (NRC) Generic Letters 89-10, 95-07 and 96-05 with conventional gate valves [2,3,4]. The earlier paper was published within a year after the new design valves were installed at the Pilgrim Nuclear Plant — the plant that took the initiative to form a teaming arrangement as described in [1] which facilitated this innovative development. The current paper documents the successful performance history of 22 years at the Pilgrim plant, as well as performance history at several other nuclear power plants where these valves have been installed for many years in containment isolation service that requires operation under pipe rupture conditions and require tight shut-off in both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The performance history of the new valve has shown to provide significant performance advantage by eliminating the chronic leakage problems and high maintenance costs in these critical service applications. This paper includes a summary of the design, analysis and separate effects testing described in detail in the earlier paper. Flow loop testing was performed on these valves under normal plant operation, various thermal binding and pressure locking scenarios, and accident/pipe rupture conditions. The valve was designed, analyzed and tested to satisfy the requirements of ANSI B16.41 [9]; it also satisfies the requirements of ASME QME 1-2012 [10]. The results of the long-term performance history including any degradation observed and its root cause are summarized in the paper.

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Commentary by Dr. Valentin Fuster
2017;():V001T04A003. doi:10.1115/PVS2017-3539.
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This report describes the tools employed by the Braidwood Generating Station Main Steam System Engineer to identify Main Steam Safety Valves (MSSVs) which may require refurbishment. These methods include in-service Trevitesting results, visual identification of steam leaks past the valve disc, external temperatures readings on the body and tailpipe flange of the valve, thermography, and risk rank charts. Utilizing these methods, Braidwood Generation Station will begin the transition from preventative maintenance of the MSSVs to a more cost effective predictive maintenance, in which the valves are rebuilt or refurbished on an as-needed basis.

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Topics: Safety , Valves , Steam
Commentary by Dr. Valentin Fuster

Snubbers

2017;():V001T05A001. doi:10.1115/PVS2017-3519.
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The ASME OM Code, Subsection ISTD, is the required code for conducting preservice and inservice examination and testing of dynamic restraints (Snubbers). The latest approved edition of this OM Code is now, or soon to be the 2012 Edition. With the publication of the 2006 addenda to Section XI of the ASME Boiler & Pressure Vessel Code, the snubber requirements, which were previously located in Article IWF-5000, were deleted. When the requirements of IWF-5000 were deleted, the requirements for examination and testing of snubbers as required by 10CFR50.55a were required to be in accordance with the ASME OM Code, Subsection ISTD.

When Owners prepare their ten-year ISI program updates that incorporate the 2006 Addenda and later of the Section XI Inspection Code, the snubber requirements will be required to be in accordance with the ASME OM Code, Subsection ISTD, 2004 Edition with Addenda through 2006 or later approved editions. This edition of the ASME OM Code has been referenced in the NRC Regulations dated June 21, 2011. Since that time, Owners are required to meet the requirements of the latest approved edition of the ASME OM Code, for snubber examination and testing requirements when snubber programs are updated.

With the transition of the snubber program from the ISI program and ASME Section XI Code requirements to the IST program and the ASME OM Code, there is sometimes confusion and implementation gaps where regulatory program requirements could be missed.

This paper will address the transition and identify potential pitfalls and how to mitigate them.

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Commentary by Dr. Valentin Fuster
2017;():V001T05A002. doi:10.1115/PVS2017-3520.
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Delivering cost reductions via the Nuclear Promise can appear to be at odds with the safe operation and maintenance of nuclear facilities. However, In-Service Examination (ISE) and In-Service Testing (IST) programs can deliver significant gains in efficiency and effectiveness with proper application of the ASME O&M code. Along with scheduled maintenance prescribed by the manufacturer, Dynamic Restraints (snubbers) require periodic visual inspection and testing to ensure the installed population will perform its safety function during seismic events or dynamic operational transients. Methods prescribed in the ASME O&M Code Subsection ISTD are effective in identifying bad actors and verifying the operational readiness of the population, but can come at a significant cost when not properly utilized, especially when the penalty for a failed test or inspection is applied to the ISE or IST campaign. The Nuclear Promise can be realized in a snubber ISE or IST program with a thorough understanding of the intent of the prescribed testing and the mechanics of the safety functions to be verified. With this understanding, legacy requirements that were grandfathered into a program can be examined as to their relevance, and procurement specifications and testing procedures can be written that are pertinent and current to industry best practices.

This paper, through the lens of a snubber manufacturer and ASME certificate holder, examines some common and uncommon examples found in industry that add significant cost, time, or dose to a snubber ISE/IST program, and the basis for eliminating them. The methodology used to evaluate an ISE/IST program requirement and determine its effectiveness in verifying a snubber’s safety function while satisfying the O&M code could be used for other components under the jurisdiction of the O&M code. In this manner, the Nuclear Promise can be safely delivered in an ISE/IST program that does not compromise the intent or integrity of Code requirements.

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Commentary by Dr. Valentin Fuster
2017;():V001T05A003. doi:10.1115/PVS2017-3522.
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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.

Commentary by Dr. Valentin Fuster

Risk Insight

2017;():V001T06A001. doi:10.1115/PVS2017-3527.
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This paper will review three options for applying risk insights to the In-service Testing (IST) Program for pumps and valves. Current regulatory framework allows for risk-informing pump and valve testing through the implementation of 10CFR50.69 or by submittal to the NRC per 10CFR50.55a for risk-informed testing in accordance with the OM Code; either using Code Case OMN-3 and the risk-related Code Cases or Subsection ISTE. This paper will offer a third option which involves the combination of the first two options. Each of these IST risk-informed program options will be explored by presenting a general discussion of each option’s risk ranking process and anticipated risk ranking results. The risk ranking review will be followed by a discussion of the implementation processes and finally a look at plant impacts and potential benefits for each option.

IST program scope and testing requirements will be identified for each of these risk-informed program options. References for the implementation processes will be provided and used for the basis of this discussion. The intent of this paper is not to provide a “how to” for each of these options, but rather to provide information to the reader to allow further detailed review of each option. It is expected that through further investigation of these options and discussions with plant management each site may find the option/process that best suits their regulatory and plant safety culture.

Paper published with permission.

Commentary by Dr. Valentin Fuster
2017;():V001T06A002. doi:10.1115/PVS2017-3538.
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Recent consequences analyses of potential station blackout (SBO) accidents at nuclear power plants have shown that an important uncertainty in accident progression and radionuclide release is the probability that a safety valve (SV) will fail-to-close after it has opened to relieve pressure [1]. The U.S. Nuclear Regulatory Commission’s (NRC’s) State-of-the-Art Reactor Consequence Analyses (SOARCA) and associated uncertainty analyses for SBOs at a pressurized-water reactor (PWR) indicated that SV behavior is an important determinant of whether an induced steam-generator tube rupture (an undesirable bypass event) may develop [2], and an important determinant of whether a PWR with an ice condenser containment may experience an early containment failure [3]. Given the importance of SV failure-to-close probabilities in these accidents, available information was reviewed to help develop better estimates of the probability for a SV’s failure-to-close on demand. The SVs of interest in the SOARCA PWR analyses are the PWR code SVs, designated SVVs in a study of SVs published in 2007 (NUREG/CR-7037) [4]. There are two sets of failure probabilities reported in NUREG/CR-7037: failure probabilities based on behavior after reactor scrams i.e., after actual operating events, and failure probabilities based on tests. Information is included for both the secondary-side, main steam system (MSS) valves, as well as reactor coolant system (RCS) valves.

The NUREG/CR-7037 failure probabilities based on actual operating events differ markedly from the failure probabilities based on tests. Further inquiries on valve testing and review of testing requirements show that the focus of testing is to demonstrate that the valves will open to relieve pressure during design-basis accidents to prevent overpressure events. The reseating or closing capability is not tested under severe accident conditions, in other words, the valve’s repeated full-stroking and passing steam. As such, the testing data was not considered applicable for severe accident modeling purposes. Furthermore, the assumption was made that MSS data was representative of RCS valve failures too during severe accident scenarios, as it is judged that they are similar enough in weighing the difference between the valves against the lack of operational data on the RCS SVs (only four data points, and one of two failures having a cause of failure now-defunct in the majority of operating PWRs in the U.S.). Lastly, recovered valve function, e.g., a previously stuck-open valve closing when pressure reduces, was not considered as a successful valve operation based on a review of licensee event reports.

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Commentary by Dr. Valentin Fuster
2017;():V001T06A003. doi:10.1115/PVS2017-3545.
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Subsection ISTE provides mandatory requirements for owners’ who voluntarily elect to implement a risk-informed inservice testing (IST) Program. The Subsection was originally prepared by combining the component categorization requirements and methodology from Code Case OMN-3 with component specific testing requirements developed, or under development, by the component-specific subgroups. Many of these requirements were based on the existing risk-informed Code Cases.

The original publication of ISTE was not endorsed by the NRC. The OM Subcommittee on Risk-Informed Activities has revised the subsection over the last four years and it is now expected to satisfy NRC concerns. This paper presents the upcoming proposed requirements for categorizing plant pumps and valves as either High Safety Significant Components or Low Safety Significant Components in accordance with ISTE and presents examples.

Paper published with permission.

Commentary by Dr. Valentin Fuster

New Reactors

2017;():V001T07A001. doi:10.1115/PVS2017-3506.
FREE TO VIEW

The American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants, Division 1, “OM Code: Section IST,” 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 following January 1, 2000.

The NuScale advanced small modular reactor plant is passive, pressurize water reactor, designed such that from one (1) to twelve (12) nuclear power plant modules (NPM) can operate within a single Reactor Building. Each NuScale Power Module (NPM) consist of a reactor core, two steam generator tube bundles, and a pressurizer contained within a single reactor vessel, along with the containment vessel that immediately surrounds the reactor vessel and is rated at 160 MWt.

The ASME Operation and Maintenance (OM) Code was written considering single-unit reactor plants, not multi-modular SMRs. The ASME Sub-Committee, New Reactors is developing a new Sub-Section, ISTG to address inservice testing (IST) of valves for all new and advanced reactor types.

This paper reviews the unique aspects and programmatic solutions for preservice testing (PST) and IST specific to the NuScale small modular reactor (SMR). The functional design, qualification provisions and IST program is described. The PST and preservice test period will be discussed as some preservice testing may be completed in the factory prior to shipping the reactor module to the site. Additionally, methods eliminate overlap, redundancy and excessive testing between the preservice testing (PST) and IST program plans will be explored.

The intent of these solutions is to provide reasonable assurance that the ASME BPV Code, Section III Class 1, Class 2, Class 3, non-safety-related and non-ASME valves that have an important function will operate when needed. The program considers both deterministic and risk insights in its evaluation of PST and IST and meets the requirements of the ASME OM Code as endorsed by 10 CFR 50.55a.

Paper published with permission.

Commentary by Dr. Valentin Fuster
2017;():V001T07A002. doi:10.1115/PVS2017-3511.
FREE TO VIEW

Curtiss Wright introduced the first Normally Open NozzleCheck valves to the nuclear power industry nearly 20 years ago. This passive valve design was developed to address reoccurring maintenance and reliability issues often experienced by various check valve types due to low flow conditions. Specifically, premature wear on the hinge pins, bushings and severe seat impact damage had been discovered in several applications while the systems were in steady state operating conditions.

Over the last two decades, Curtiss Wright has continued to improve upon the design of the valve, with the latest generation coming most recently in support of the Westinghouse AP1000 design and similar Generation III+ passive reactor requirements. This entirely new valve is designed with minimal stroke, ensuring quick closure under specified reverse flow conditions which no other check valve design could support. Additionally, features such as first in kind test ports, visual inspection points, and the ability to stroke the valve manually or with system fluid in line have resolved many of the shortcomings of previous inline welded flow check valves.

Most importantly, advanced test based methodologies and models developed by Curtiss Wright, allow for accurate prediction of NozzleCheck valve performance. This paper presents the development of Curtiss Wright’s advanced Normally Open NozzleCheck Valve for Generation III and III+ nuclear reactor designs. The Valve performance was initially determined by using verified and validated computational fluid dynamic (CFD) methods. The results obtained from the CFD model were then compared to the data gathered from a prototype valve that was built and tested to confirm the performance predictions. Curtiss Wright has fully tested and qualified the Normally Open NozzleCheck valve, which is specifically designed for applications that require a high capacity in the forward flow direction and a closure at low flow rates with short stroke to minimize the hydraulic impact on the system.

Paper published with permission.

Commentary by Dr. Valentin Fuster

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