Information Notice No. 89-90: Pressurizer Safety Valve Lift Setpoint Shift

                                UNITED STATES
                           WASHINGTON, D.C.  20555

                              December 28, 1989



All holders of operating licenses or construction permits for pressurized 
water reactors (PWRs).


This information notice is to alert addressees to potential problems 
resulting from operating pressurizer safety valves (PSVs) in an environment 
different from that used to establish the PSV lift setpoint.  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 do not constitute 
NRC requirements; therefore, no specific action or written response is 

Description of Circumstances:

In October 1989, Westinghouse informed its plant owners of a potential 
deviation of the PSV set pressure from the ASME Code and the plant technical 
specification (TS) requirements for plants having loop seals upstream of the 
PSVs.  Recent plant operating experience and test data indicate that the PSV 
lift pressure changes by more than 1 percent from the original set pressure 
when the valve is operated at conditions different from those used during 
the establishment of the lift setpoint.  Westinghouse observed a shift of 4 
to 8 percent on Crosby PSVs when setpoints were initially established using 
a loop seal with 300 F water, draining the loop seal, and checking the lift 
set pressure with steam.  As ASME Code Section III requires a safety valve 
setting with a tolerance of +1/-1 percent of the set pressure and the plant 
TSs typically specify the PSV lift setting of 2485 psig +1/-1 percent, some 
plants may be operating with PSV setpoints not in compliance with their TSs 
or the ASME Code if they are operating in an environment different from that 
used to establish the valve setpoints.  In addition, some plant TSs have a 
footnote which states, "The lift setting pressure shall correspond to 
ambient conditions of the valve at nominal operating temperature and 

The Westinghouse letter specifically identified a potential safety issue 
with setting the PSV setpoint with steam and operating the valves in a loop 
seal containing water.  Because the actual lift set pressure could be 4 to 8 
percent higher than the 2485 psig +1% set pressure, this increased PSV lift 


                                                       IN 89-90
                                                       December 28, 1989
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could result in primary system overpressurization for certain event 
scenarios.  Thus, plant-specific analyses of those scenarios in which the 
availability of the PSV was used in the licensing basis criteria for the 
facility may show that the appropriate pressure limit is exceeded for the 
pressurizer and associated piping.

The PSV loop seals may be uninsulated or insulated and may or may not have 
heaters.  Water temperatures in the uninsulated and insulated loop seals are 
approximately 140 F and 300 F to 400 F, respectively.  The temperature of 
the loop seal water affects the PSV temperature and the lift pressure.  The 
lift pressure setpoints of the PSVs with loop seals are not established 
consistently by the industry.  A survey showed that the inlet conditions 
under which the PSV lift setpoints were established include steam, nitrogen, 
and actual loop seal water conditions.  As a result, the actual PSV lift 
pressure could be either too high or too low, depending on the method of PSV 
setting and the actual fluid condition upstream of the PSVs.


Inconsistencies in the conditions at which the PSV lift pressure is actually 
established, compared to actual operating conditions of the PSV, have safety 
implications and affect compliance with the plant TSs.  It is essential that 
the PSVs be operable with proper lift pressure settings that are low enough 
for acceptable plant overpressure protection but high enough to prevent 
lifting of the PSVs at a pressure approaching the reactor coolant system 
(RCS) operating conditions.  If the lift pressure for a PSV is set on steam 
and is operated with loop seal water, the actual lift setpoint may be too 
high and result in noncompliance with the TSs as well as possible 
overpressurization of the primary system in excess of the acceptance limit 
of 110 percent of design pressure for certain accidents.  On the other hand, 
if the PSV setpoint is established with loop seal water, a loss of loop seal 
water will result in a lower actual PSV lift pressure.  This situation may 
result in noncompliance with the plant TSs and also may cause the PSV to 
lift at a pressure approaching the RCS operating pressure, increasing the 
probability of a challenge to the PSVs.  NUREG-0737, Item II.K.3.2, 
addresses the need for reducing challenges to PSVs.  Repetitive or frequent 
challenges to the PSVs may prevent the PSVs from reseating, with a potential 
for an unisolable small-break loss-of-coolant accident (LOCA).  The effect 
on the actual lift pressure of a PSV that is set with nitrogen and operated 
with loop seal water has not been determined.

There were two instances (on May 17 and August 25, 1989) at the V. C. Summer 
plant in which the loop seal was lost as a result of the PSV leakage.  
Because the setpoint of the V.C. Summer PSV was established with hot water, 
the actual lift setpoint decreased as a result of the absence of water in 
the loop seal piping.  The PSV opened prematurely, resulting in a partial 
depressurization of the reactor coolant system.  PSV leakage also occurred 
at the Diablo Canyon plant, where the PSV setpoint was also established with 
hot water.  Leakage past the PSV was detected by the PSV tailpipe 
temperature monitoring devices and the acoustic leak monitors and 
subsequently resulted in a plant shutdown.


                                                       IN 89-90
                                                       December 28, 1989
                                                       Page 3 of 3

In the case of the PSVs at Summer, the PSV lift setpoints are still based 
upon a loop seal configuration.  The licensee has taken steps to maintain 
the loop seal.  They have installed thermocouples in the PSV loop seals to 
monitor the PSV body inlet and loop seal temperatures.  If the valve body 
inlet temperature exceeds a predetermined temperature, this is considered 
indicative of an impending loss of the loop seal and the licensee will then
take action to shut down the plant to restore the loop seal.  The licensee 
is considering removing the PSV loop seals at the next refueling outage.

In October 1989, the Surry Station Unit 2 PSVs were shipped to the 
Westinghouse Western Service Center to undergo testing after leakage was 
observed from one of the valves.  The setpoints of the Surry PSVs had been 
established with steam.  The test results showed that the as-found lift 
pressure setpoints differed significantly when tested under steam vs loop 
seal water conditions.  The licensee reset the Unit 2 PSVs under water 
conditions to comply with the TS requirements.  However, during a subsequent 
post-maintenance pressure test, the "C" PSV lifted at an RCS pressure of 
2335 psig and reclosed at 2255 psig, apparently from a loss of loop seal 
water.  As a result of this event, the licensee decided to return to the 
previous method of establishing the PSV lift pressure with steam to avoid 
challenges to the Unit 2 PSVs.  On November 10, 1989, the licensee requested 
a TS change for Units 1 and 2 for the remainder of Cycle 10 operation to 
increase the PSV setpoint tolerance to the value observed in the Unit 2 PSV 
test data.  This TS change request was supported by a safety analysis 
showing that the reactor system pressure remains below the 110-percent 
design pressure limit for the limiting pressurization events if the PSV 
setpoint is increased provided a power-operated relief valve (PORV) is 
operable.  The TS change was approved with the provision that the licensee 
take compensatory measures to ensure operability of at least one of the 
PORVs and also ensure the operability of the direct reactor trip upon a 
turbine trip.

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 NRR project 

                              Charles E. Rossi, Director
                              Division of Operational Events Assessment
                              Office of Nuclear Reactor Regulation

Technical Contacts:  Y. (Gene) Hsii, NRR
                     (301) 492-0887 

                     W. Jensen, NRR 
                     (301) 492-1190 

Attachment:  List of Recently Issued NRC Information Notices

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