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
8808160451
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August 22, 1988
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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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August 22, 1988
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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
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