K. G. DeWall’s scientific contributions

This report describes four sets of tests sponsored by the U.S. Nuclear Regulatory Commission and conducted by the Idaho National Engineering and Environmental Laboratory. The tests support research addressing the need to provide assurance that motor-operated valves are able to perform their intended safety function, usually to open or close against specified (design basis) flow and pressure loads. One of the parameters that affects a gate valve's operability is the friction between the disc seats and the valve body seats. In most gate valves, these surfaces are hardfaced with Stellite 6, a cobalt-based alloy. The tests described in this report investigate the changes that occur in the friction as the Stellite 6 surfaces develop an oxide film as they age. Stellite 6 specimens were aged in a corrosion autoclave, the oxide films were examined and characterized, and the specimens were subjected to friction testing in a friction autoclave. A very thin oxide film formed after only a fe w days of natural aging. Even a very thin oxide film caused an increase in friction. The surface structure of the oxide film was dominated by a hard crystalline structure, such that the friction response was analogous to rubbing two pieces of sandpaper together. In the limited data provided by naturally aged specimens (78 days maximum exposure, very thin oxide films), the friction increased with greater aging time, approaching an as-yet-undetermined plateau. Although the thickness of the oxide film increased with greater aging time, the mechanical properties of the oxide film (larger granules with greater aging time) appeared to play a greater role in the friction response. Friction testing of specimens subjected to simulated in-service testing strokes at intervals during the aging process showed only a slight decrease in friction, compared to other specimens. Results from specimens subjected to accelerated aging were inconclusive, because of differences in the structure and comp osition of the oxide films, compared to naturally aged specimens. For the naturally aged specimens, the highest friction occurred on the first stroke. The first stroke smeared the oxide film and dislodged some of the granules, so that subsequent strokes saw lower friction values and less variation in the friction. This result underscores the importance of planning in-plant tests so that data are collected from the first stroke following a period of inactivity.

For the past several years, researchers at the Idaho National Engineering and Environmental Laboratory, under the sponsorship of the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, have been investigating the performance of motor-operated valves subjected to design basis flow and pressure loads. Part of this research addresses the friction that occurs at the interface between the valve disc and the valve body seats during operation of a gate valve. In most gate valves, these surfaces are hardfaced with Stellite 6, a cobalt-based alloy. Analytical methods exist for predicting the thrust needed to operate these valves at specific pressure conditions. To produce accurate valve thrust predictions, the analyst must have a reasonably accurate, though conservative, estimate of the coefficient of friction at the disc-to-seat interface. One of the questions that remains to be answered is whether, and to what extent, aging of the disc and seat surfaces effects the disc-to-seat coefficient of friction. Specifically, does the environment in a nuclear plants piping system cause the accumulation of an oxide film on these surfaces that increases the coefficient of friction; and if so, how great is the increase? This paper presents results of specimen tests addressing this issue, with emphasis on the following: (1) the characteristics and thickness of the oxide film that develops on Stellite 6 as it ages; (2) the change in the friction coefficient of Stellite 6 as it ages, including the question of whether the friction coefficient eventually reaches a plateau; and (3) the effect in-service cycling has on the characteristics and thickness of the oxide film and on the friction coefficient.

The US Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, is funding the Idaho National Engineering and Environmental Laboratory (INEEL) in performing research investigating the performance of gate valves subjected to pressure locking and thermal binding conditions. Pressure locking and thermal binding are phenomena that make a closed gate valve difficult to open. Pressure locking can occur when operating sequences or temperature changes cause the pressure of the fluid in the bonnet (and, in most gate valves, between the discs) to be higher than the pressure on the upstream and downstream sides of the disc assembly. Thermal binding can occur when thermal expansion/contraction effects cause the disc to be squeezed between the valve body seats. If the loads associated with pressure locking or thermal binding are very high, the actuator might not have the capacity to open the valve. The authors tested a flexible-wedge gate valve and a double-disc gate valve under pressure locking and thermal binding conditions. The results show that these valves are susceptible to pressure locking; however, they are not significantly affected by thermal binding. For the flexible-wedge gate valve, pressure locking loads (in terms of stem thrust) were higher than corresponding hydrostatic opening loads by a factor of 1.1 to 1.5. For the parallel disc gate valve, pressure locking loads were higher by a factor of 2.05 to 2.4. The results also show that seat leakage affects the bonnet pressurization rate when the valve is subjected to thermally induced pressure locking conditions.

Researchers at the Idaho National Engineering and Environmental Laboratory tested the performance of electric motors and actuator gearboxes typical of the equipment installed on motor-operated valves used in nuclear power plants. Using a test stand that simulates valve closure loads against flow and pressure, the authors tested five electric motors (four ac and one dc) and three gearboxes at conditions a motor might experience in a power plant, including such off-normal conditions as operation at high temperature and reduced voltage. They also monitored the efficiency of the actuator gearbox. All five motors operated at or above their rated starting torque during tests at normal voltages and temperatures. For all five motors, actual torque losses due to voltage degradation were greater than the losses calculated by methods typically used for predicting motor torque at degraded voltage conditions. For the dc motor the actual torque losses due to elevated operating temperatures were greater than the losses calculated by the typical predictive method. The actual efficiencies of the actuator gearboxes were generally lower than the running efficiencies published by the manufacturer and were generally nearer the published pull-out efficiencies. Operation of the gearbox at elevated temperature did not affect the operating efficiency.

This report provides an update on the valve research being sponsored by the US Nuclear Regulatory Commission (NRC) and conducted at the Idaho National Engineering Laboratory (INEL). The research addresses the need to provide assurance that motor-operated valves can perform their intended safety function, usually to open or close against specified (design basis) flow and pressure loads. This report describes several important developments: Two methods for estimating or bounding the design basis stem factor (in rising-stem valves), using data from tests less severe than design basis tests; a new correlation for evaluating the opening responses of gate valves and for predicting opening requirements; an extrapolation method that uses the results of a best effort flow test to estimate the design basis closing requirements of a gate valve that exhibits atypical responses (peak force occurs before flow isolation); and the extension of the original INEL closing correlation to include low- flow and low-pressure loads. The report also includes a general approach, presented in step-by-step format, for determining operating margins for rising-stem valves (gate valves and globe valves) as well as quarter-turn valves (ball valves and butterfly valves).

This report documents the results of the main projects undertaken under the Environmental and Dynamic Equipment Qualification Research Program (EDQP) sponsored by the U.S. Nuclear Regulatory Commission (NRC) under FIN A6322. Lasting from fiscal year 1983 to 1987, the program dealt with environmental and dynamic (including seismic) equipment qualification issues for mechanical and electromechanical components and systems used in nuclear power plants. The research results have since been used by both the NRC and industry. The program included seven major research projects that addressed the following issues: (a) containment purge and vent valves performing under design basis loss of coolant accident loads, (b) containment piping penetrations and isolation valves performing under seismic loadings and design basis and severe accident containment wall displacements, (c) shaft seals for primary coolant pumps performing under station blackout conditions, (d) electrical cabinet internals responding to in-structure generated motion (rattling), and (e) in situ piping and valves responding to seismic loadings. Another project investigating whether certain containment isolation valves will close under design basis conditions was also started under this program. This report includes eight main section, each of which provides a brief description of one of the projects, a summary of the findings, and an overview of the application of the results. A bibliography lists the journal articles, papers, and reports that document the research.

This report presents the results of the containment purge and vent valve test program, conducted under the sponsorship of the United States Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research. The test program investigated butterfly valve operability and leak integrity under light-water-reactor design basis and severe accident conditions. Three nuclear-designed butterfly valves typical of those used in domestic nuclear power plant containment purge and vent applications were tested. For a comparison of response, two valve of the same size with differing internal designs were tested. For extrapolation insights, a larger-sized valve similar to one of the smaller valves was also tested. Dynamic flow tests were performed over the range of design basis accident pressures. Leak integrity testing was also performed at both design basis and severe accident temperatures and pressures. The valve experiments were performed with various piping configurations and valve orientations to the flow to simulate the various installation options in field applications. Testing was also performed in a standard ANSI test section.

This report presents the results of the survey of Aged Nuclear Power Plant Facilities conducted for the USNRC Office of Nuclear Regulatory Research. Results recommend methods to help formulate a comprehensive research program that will systematically identify aging and service wear effects which are likely to affect plant safety. The survey centered on safety related plant systems with regard to component failures from operating histories. The age related failure information gathered from the plant histories was analyzed for reoccurring failure patterns. Emphasis was on identification of specific equipment with high failure rates and of failure mechanism relationships. The data would not support specific equipment identification. It did imply a direct relationship between failure and failure mechanism. 70% of the failures reported were due to four failure mechanisms. In addition there appeared to be a strong correlation between cause of failure and the system in which the component operates. This is verified by detailed study of several plant systems and corroborated by personnel interviews. This survey indicates identification and elimination of system level cause of component failure is a viable approach to prevent and mitigate major reported aging effects.

Researchers at the Idaho National Engineering Laboratory recently conducted tests investigating the operating efficiency of the power train (gearbox) in motor-operators typically used in nuclear power plants to power motor-operated valves. Actual efficiency ratios were determined from in-line measurements of electric motor torque (input to the operator gearbox) and valve stem torque (output from the gearbox) while the operators were subjected to gradually increasing loads until the electric motor stalled. The testing included parametric studies under reduced voltage and elevated temperature conditions. As part of the analysis of the results, we compared efficiency values determined from testing to the values published by the operator manufacturer and typically used by the industry in calculations for estimating motor-operator capabilities. The operators we tested under load ran at efficiencies lower than the running efficiency (typically 50%) published by the operator manufacturer.

This report provides an update on the valve research sponsored by the US Nuclear Regulatory Commission (NRC) that is being conducted at the Idaho National Engineering Laboratory. The update focuses on the information applicable to the following requests from the NRC staff: Examine the use of in situ tests results to estimate the response of a valve at design-basis conditions; Examine the methods used by industry to predict required valve stem forces or torques; Identify guidelines for satisfactory performance of motor-operated valve diagnostics systems; and participate in writing a performance standard or guidance document for acceptable design-basis tests. The authors have reviewed past, current, and ongoing research programs to provide the information available to address these items.