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HomeNRC LibraryDocument CollectionsGeneric CommunicationsInformation Notices1992

# Information Notice No. 92-60: Valve Stem Failure Caused by Embrittlement

```
                                 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 Friday, May 22, 2015*

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