Home > NRC Library > Document Collections > General Communications > Information Notices > 1992 > IN 92-60
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555 August 20, 1992 NRC INFORMATION NOTICE 92-60: VALVE STEM FAILURE CAUSED BY EMBRITTLEMENT Addressees All holders of operating licenses or construction permits for pressurized water reactors (PWRs). Purpose The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert addressees to problems in which valve stems manufactured from American Society of Mechanical Engineers (ASME) SA 564, Type 630, H900 through H1150 age treatment condition (17-4 PH) stainless steel could become brittle and fail if used in environments that exceed 600 �F. It is expected that recipients will review this information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. However, suggestions contained in this information notice are not NRC requirements; therefore, no specific action or written response is required. Background Power-operated relief valves (PORVs) connected to the pressurizer are designed for the valve stems to operate at saturated steam temperatures. The valve stem thermal history is affected by thermal conduction through the valve disk and by direct contact to the fluid discharge as a result of valve actuation. The normal operating saturation temperature in a PWR is approximately 650 -F. The maximum design temperatures are higher. Data obtained by the Duke Power Company from testing on 17-4 PH material in the age treated condition indicate that this material will, after several thousand hours at 600 -F, exhibit an increase in tensile strength with an accompanying large decrease in ductility (secondary aging). The secondary aging mechanisms are the continued precipitation of the intermetallic compounds and the precipitation of chromium in ferrite (885 -F embrittlement). After secondary aging occurs, the material with low ductility will have an increased susceptibility for fracture, especially when subjected to high torque from a power actuator. Description of Circumstances On December 9, 1991, the Catawba Nuclear Station, Unit 2, was in Mode 5 (hot shutdown) in the final stages of a refueling outage. The pressure in the reactor coolant system was approximately 177 psig. 9208140032. IN 92-60 August 20, 1992 Page 2 of 3 Operations personnel had noted some unexpected perturbations in the pressurizer relief tank earlier during the normal reactor coolant system fill- and-vent process which led them to suspect that either pressurizer PORV 2NC-32B or its associated block valve 2NC-31B may not be opening properly. Both valves appeared to stroke when observed locally. To verify that the valves were functioning properly, operations personnel attempted to depressurize the reactor coolant system by opening the PORV from the control room. When the PORV was opened, and the block valve 2NC-31B indicated an open condition, the reactor coolant system pressure remained stable. Operations personnel suspected that the PORV block valve was stuck in the closed position. The PORV block valve is a motor operated 3-inch Rockwell International (now Edward Valve Company) Equiwedge gate valve. The licensee reviewed the data on a test performed on the block valve actuator on November 25, 1991, which suggested that the stem of the valve had separated from the gate assembly because the stem pullout force was much lower than normal. The licensee replaced the PORV block valve and verified that the stem had failed. The failure occurred in an area where the valve stem attached to the gate assembly. The licensee replaced the valve with a different type of valve, and radiographically tested the other PORV block valves on both units to verify that they were intact and open. Discussion The licensee performed a metallurgical analysis of the fractured stem and found that the material had lost ductility at the point of fracture. Apparently, the stem end attached to the disk was exposed to pressurizer temperatures above 600 -F for several thousand hours. The high torque applied by the power operator was sufficient to shear the stem. The Edward Valve Company indicated that valve stems made of this material, ASME SA 564, Type 630 (17-4 PH, with an aging treatment of H1100), can become embrittled in as little as 5,000 hours when exposed to temperatures greater than 600 -F. Preliminary information indicates that this material is widely used in the stems of PORVs and PORV block valves. It is important to note that a valve with a severed stem may pass a surveillance test. For example, if the safety function of the valve is to close, and the valve is tested in the closed direction, an acceptable closing force can be obtained even with a severed stem. This was, in fact, the case with the 2NC-31B valve. Determining the valve disk position is an operability concern for those systems with valves that are normally opened during operations but are closed during accident conditions. . IN 92-60 August 20, 1992 Page 3 of 3 This information notice requires no specific action or written response. If you have any questions about the information in this notice, please contact one of the technical contacts listed below or the appropriate project manager in the Office of Nuclear Reactor Regulation (NRR). ORIGINAL SIGNED BY Charles E. Rossi, Director Division of Operational Events Assessment Office of Nuclear Reactor Regulation Technical contacts: William Orders (803) 831-2963 Donald Naujock (301) 504-2767 Attachment: List of Recently Issued NRC Information Notices .
Page Last Reviewed/Updated Thursday, March 29, 2012