United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 88-68: Setpoint Testing of Pressurizer Safety Valves with Filled Loop Seals Using Hydraulic Assist Devices

                                  UNITED STATES
                          NUCLEAR REGULATORY COMMISSION
                      OFFICE OF NUCLEAR REACTOR REGULATION
                             WASHINGTON, D.C.  20555

                                 August 22, 1988


Information Notice No. 88-68:  SETPOINT TESTING OF PRESSURIZER SAFETY 
                                   VALVES WITH FILLED LOOP SEALS USING 
                                   HYDRAULIC ASSIST DEVICES 


Addressees: 

All holders of operating licenses or construction permits for nuclear power 
reactors. 

Purpose: 

This information notice is being provided to alert addressees to potentially 
generic problems that have occurred during testing of pressurizer safety 
valves with filled loop seals using a hydraulic assist device.  Use of 
hydraulic assist devices may result in inaccurate results for safety valve 
setpoint testing when the valves are subjected to water or two-phase flow.  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 notice do not constitute NRC 
require-ments; therefore, no specific action or written response is required. 

Description of Circumstances: 

On August 30, 1986, the licensee of Diablo Canyon 1 tested the three pressur-
izer safety valves with the loop seal filled using a hydraulic assist device.  
The initial lift points were reported as 2747.8, 3028.0, and 2661.0 psig for 
valves number RCS-1-8010A, B, and C, respectively.  The required setpoint for 
the valves was 2485�1% psig.  The test method monitored hydraulic pressure on 
the test rig for an indication of valve stem displacement to infer lift point.  
The licensee concluded that the inferred lift point was not accurate on the 
first lift because the loop seal was not drained.  Water moving through the 
seat area produced little valve stem displacement because of the different 
physical properties of steam and water.  Steam then entered the valves at an 
elevated hydraulic pressure and caused a larger displacement, resulting in the 
prediction of an inaccurately high lift point.  After the loop seal was 
drained, the lift points of the valves were measured again and found to be 
within technical specification (TS) limits (2485�1%) at 2464, 2493, and 2503 
psig (LER 50/275-86/018). 

On April 2, 1988, the licensee of Sequoyah 2 tested the setpoints of the 
pressurizer safety valves to determine if low setpoints could be the cause of 
the leakage that the valves had been experiencing.  A hydraulic assist device 



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was used with reactor pressure between 1700 psig and normal operating 
pressure, and with a water seal at the valve seat maintained at an elevated 
temperature by external heaters.  The lifting force required to open the valve 
was measured by a load cell and recorded on a strip chart to provide data 
necessary to calculate the valve setpoint.  Disc lift was determined by a 
change in slope on the load cell trace and confirmed by test personnel who 
listened for audible passage of flow through the discharge piping.  With valve 
lift assumed known, the lift force was used to calculate an equivalent 
pressure.  This pressure was added to the system static pressure to infer the 
valve setpoint.  Initial lift results were 2634 psig for valve 2-68-563, 2678 
psig for 2-68-564, and 2660 psig for 2-68-565.  The required setpoint was 
2485�1% psig for each valve.  Setpoints were readjusted and the valves were 
retested in situ, after reestablishing the water seal. 

On April 8, 1988, with the unit in cold shutdown, the Sequoyah 2 pressurizer 
safety valves were sent to Wyle Laboratories for bench testing and seat refur-
bishment.  The valve lifts were performed using water heated to 120�F and 
pressurized by nitrogen.  Valve stem lift was directly measured by a linear 
voltage differential transformer (LVDT) mounted on the valve stem, and the 
stem displacement was recorded on a strip chart.  The lift setting was 
indicated by a clear peak on the pressure strip chart and confirmed by spindle 
displacement measured by the LVDT.  Two of the valves had some internal parts 
replaced prior to the tests.  Lift points were 2435 and 2384 psig for valve 
2-68-563, 2430 and 2432 psig for 2-68-564, and 2390 psig for 2-68-565. 

The different lift points of the Sequoyah 2 valves were attributed to differ-
ences in determining the time at which the disc began to lift.  The LVDT used 
at Wyle was accurate to 0.001 inch.  The licensee's in situ method used a 
change in the slope of a trace of lift force on a strip chart, backed up by 
technicians' confirmation of audible flow.  

The licensee found that the setpoint adjustments made as a result of the 
Sequoyah 2 in situ tests brought the setpoint of the valves outside the limits 
of the TS.  This situation could have resulted in the premature lifting of the 
safety valves (LER 50/328-88/016). 

The licensee for Diablo Canyon has decided to drain the pressurizer safety 
valve loop seal before testing.  The licensee for Sequoyah is planning to send 
pressurizer safety valves to Wyle Laboratories for future testing. 

Hydraulic assist devices have been shown to be accurate in testing spring-
actuated safety valve setpoints with saturated steam as the lift medium.  They 
have not been shown to be accurate for safety valve setpoint testing using 
water or two-phase flow.  The results of the Diablo Canyon 1 and Sequoyah 2 
tests appear to show that these devices produce inaccurate results when 
testing pressurizer safety valves with filled loop seals.  If the inaccurate 
results are believed and the valves are reset, a situation can occur in which 
the setpoints of the valves are low, resulting in valve leakage, and/or 
premature lift. 

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                                                            Page 3 of 3 


No specific action or written response is required by this information notice.
If you have any questions about this matter, please contact one of the techni-
cal contacts listed below or the Regional Administrator of the appropriate NRC 
regional office. 




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


Technical Contacts:  Mary S. Wegner, AEOD
                     (301) 492-7818

                     Charles G. Hammer, NRR
                     (301) 492-0919


Attachment:  List of Recently Issued NRC Information Notices
.                                                            Attachment
                                                            IN 88-68 
                                                            August 22, 1988 
                                                            Page 1 of 1

                             LIST OF RECENTLY ISSUED
                            NRC INFORMATION NOTICES 
_____________________________________________________________________________
Information                                  Date of 
Notice No._____Subject_______________________Issuance_______Issued to________

88-67          PWR Auxiliary Feedwater Pump  8/22/88        All holders of OLs
               Turbine Overspeed Trip                       or CPs for nuclear
               Failure                                      power reactors. 

88-66          Industrial Radiography        8/22/88        All NRC industrial
               Inspection and Enforcement                   radiography 
                                                            licensees. 

88-65          Inadvertent Drainages of      8/18/88        All holders of OLs
               Spent Fuel Pools                             or CPs for nuclear
                                                            power reactors and
                                                            fuel storage 
                                                            facilities. 

88-64          Reporting Fires in Nuclear    8/18/88        All holders of OLs
               Process Systems at Nuclear                   or CPs for nuclear
               Power Plants                                 power reactors. 

88-63          High Radiation Hazards        8/15/88        All holders of OLs
               from Irradiated Incore                       or CPs for nuclear
               Detectors and Cables                         power reactors, 
                                                            research reactors 
                                                            and test reactors.

88-62          Recent Findings Concerning    8/12/88        All holders of NRC
               Implementation of Quality                    quality assurance 
               Assurance Programs by                        program approval 
               Suppliers of Transport                       for radioactive 
               Packages                                     material packages.

88-61          Control Room Habitability -   8/11/88        All holders of OLs
               Recent Reviews of Operating                  or CPs for nuclear
               Experience                                   power reactors. 

88-60          Inadequate Design and         8/11/88        All holders of OLs
               Installation of Watertight                   or CPs for nuclear
               Penetration Seals                            power reactors. 

88-04,         Inadequate Qualification      8/9/88         All holders of OLs
Supplement 1   and Documentation of Fire                    or CPs for nuclear
               Barrier Penetration Seals                    power reactors. 

88-59          Main Steam Isolation Valve    8/9/88         All holders of OLs
               Guide Rail Failure at                        or CPs for nuclear
               Waterford Unit 3                             power reactors. 
_____________________________________________________________________________
OL = Operating License
CP = Construction Permit 
Page Last Reviewed/Updated Tuesday, November 12, 2013