United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 96-03: Main Steam Safety Valve Setpoint Variation as a Result of Thermal Effects

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                          WASHINGTON, DC  20555-0001

                                January 5, 1996

                               RESULT OF THERMAL EFFECTS


All holders of operating licenses or construction permits for nuclear power


The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to a possible source of variation in the setpoints
of various safety valves as a result of changes in temperature in and around
the valves.  It is expected that recipients will review the 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.

Description of Circumstances

On September 22, 1995, with the reactor at approximately 83 percent power in a
coast-down before a scheduled refueling outage, the licensee for Arkansas
Nuclear One, Unit 2 (ANO-2), began testing the ANO-2 main steam safety valves
(MSSVs) in place using a setpoint testing assist device.  The first valve was
tested up to a simulated pressure of 6.1 percent above the nominal setpoint,
but did not lift.  The licensee stopped testing and reviewed the procedures
before resuming further testing.  The next valve tested lifted at 4.3 percent
above its nominal setpoint, and the subsequent valve would not lift at 5.9
percent above its nominal setpoint.  The licensee stopped in situ testing,
and, following a reactor cooldown, all 10 MSSVs were removed and shipped to
Wyle Laboratories for testing and/or refurbishment.  The ANO-2 MSSVs are model
HA-65-FN manufactured by Crosby Valve and Guage Company.

On September 30, 1995, Wyle tested one of the valves with full-pressure steam
which the licensee had been unable to lift with the in situ testing.  The
valve lifted at 0.97 percent above its nominal setpoint, which is within the 
-3 percent, +1 percent tolerance required by the plant Technical
Specifications.  The licensee personnel on site at Wyle investigated the
discrepancy between the two test results.  The licensee determined that the
setpoint difference between the two methods appeared to be caused by the 

9512290299.                                                            IN 96-03
                                                            January 5, 1996
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difference in the thermal environments in which the valve was tested.  The in
situ test at the plant was performed with the valve uninsulated with an
ambient air temperature of approximately 35 �C (95 �F).  The Wyle test was
performed with the valve insulated in an environmental box at an ambient
environment of 60 �C (140 �F).  When the valve was retested at Wyle under a
simulated ANO-2 environment (including lack of insulation), the safety valve
lifted at 5.7 percent above the nominal setpoint, which was similar to the
results observed during the in situ testing.  Wyle then proceeded to test all
of the MSSVs using simulated ANO-2 thermal environmental conditions.  Five of
the 10 valves failed to meet the setpoint tolerance required by Technical
Specifications including two valve setpoints that exceeded +6 percent of their
nominal setpoints.  Since the MSSV setpoints exceeded the allowable tolerance
in the Technical Specifications, the licensee performed an analysis and
determined that using actual as-found setpoint values, the MSSVs could have
provided adequate overpressure protection during all design-basis events to
prevent the peak primary and secondary system pressures from exceeding 
110 percent of the system design pressures.

Additional information regarding the MSSV test results discussed above is
provided in Licensee Event Report (LER) 95-005 submitted for the ANO-2


In discussions with the NRC, testing personnel at Wyle indicated that they use
a default set of thermal environmental conditions unless the customer provides
other guidance.  They also indicated that the licensees approve the test
procedures before they are implemented.  Wyle further indicated that,
historically, most plants have not provided requirements for simulating the
plant thermal environmental conditions, but that recently more plants have
provided detailed thermal environmental requirements.  Many licensees are
committed to American National Standards Institute/American Society of
Mechanical Engineers (ANSI/ASME) OM-1987 Part 1, "Requirements for In-service
Performance Testing of Nuclear Power Plant Pressure Relief Devices," which
requires that the ambient temperature of the operating environment shall be
simulated during the set pressure test.

The testing that was performed for the ANO-2 MSSVs indicates that safety valve
setpoints can vary significantly in response to thermal environmental
conditions and that the magnitude of the setpoint effects can vary from valve
to valve.  

Related Generic Communications

Thermal environmental effects on safety valve setpoints have been discussed in
the following NRC generic communications:

.     NRC INFORMATION NOTICE 93-02:  "Malfunction of a Pressurizer Code Safety
.                                                            IN 96-03
                                                            January 5, 1996
                                                            Page 3 of 3

.     NRC Information Notice 89-90, Supplement 1:  "Pressurizer Safety Valve
      Lift Setpoint Shift"

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 Office of
Nuclear Reactor Regulation (NRR) project manager. 

                                          signed by

                                    Dennis M. Crutchfield, Director
                                    Division of Reactor Program Management
                                    Office of Nuclear Reactor Regulation

Technical Contacts:  Charles G. Hammer, NRR           
                     (301) 415-2791

                     Paula A. Goldberg, Region IV
                     (817) 860-8168

                     John R. Tappert, NRR
                     (301) 415-1167

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