Plant Performance History of an Innovative Gate Valve in Critical Service Applications

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. Paper published with permission.

Expectations for Inservice Testing Programs at New Nuclear Power Plants

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. Paper published with permission.

New Validated Methodology for the Required Force to Operate Balanced Disk Globe Valves

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. Paper published with permission.

Effective Diaphragm Area Test Program for Air Operated Valve Actuators

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. Paper published with permission.

ASME OM Code Subsection ISTE: A Discussion of the Upcoming Subsection

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.

Risk-Informed In-Service Testing Programs

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.

TVA OM Code Mandatory Appendix III Readiness Assessment

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. Paper published with permission.

Evaluating AOV and MOV Performance

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. Paper published with permission.

Oil/Water Correlation for PRV (Pressure Relief Valves): Is the Measured Set Pressure the Same on Water Compared to Diesel Fuel, Lubricating Oil, etc.?

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. Paper published with permission.

Gauging the Force Effects of Valve Stem Packing on Valve Stem Actuation

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. Paper published with permission.