ML20127B140
| ML20127B140 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 08/07/1989 |
| From: | Rosenthal J NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | Berlinger C Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20127B143 | List: |
| References | |
| FOIA-92-87 NUDOCS 8908110030 | |
| Download: ML20127B140 (2) | |
Text
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MEMORAllDUM FOR: Carl H. Derlinger, Chief Generic Communications Branch '
Division of Operational Events Assessment Office of Nuclear Reactor Regulation FROM: Jack E. Rosenthal, Chief Reactor Operations Analysis Branch Division of Safety Programs Office of Analysis and Evaluation of Operational Data
SUBJECT:
TRANSi1ITTAL OF LERs FOR POTENTIAL GENERIC COMMUNICATIONS ACTION Enclosed fcr your information and consideration for an information notice is an LER describing Potter and Brumfiled Relay Problems and an LER reportable pursuant to 10CFR21 on atmospheric dump valve deficiencies. We are providing this information as part of the AE00 function to identify potential adverse events of generic concern and your ability to take action on deficient performance.
No response to this memorandum cr specific action on these events is required. q i
/DI Jack E. Rosenthal, Chief Reactor Operations Analysis Branch Division of Safety Programs Office of Analysis and Evaluation
, of Operational Data
Enclosures:
- 1. 530/89-007-00
- 2. 528/89-005-00 Distribution:
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MEMORANDUM FOR: Carl \)i.Berlinger, Chief Gener'ic Communica tions Branch Division of Operational Events Assessment Office hf Nuclear Reactor Regulation FROM: Jack E. Rbsenthal, Chief Reactor Op' rations Analysis Branch Division o Safety Programs Office of An lysis and Evaluetion of Operatio al Data
SUBJECT:
TRANSMITTAL 0F ERs FOR POTENTIAL GENERIC COMMUNICATIONS A TION
- Enclosed for your information and co sideration are LERs that at least one reviewer in the AE00 screening prograq identified for consideration of an NRC l Information Notice or informative to tle Part 21 program. We are providing this information as part of the AE00 ft. ction to identify potential adverse events of generic concern and your ability to take action on deficient performance.
No response to this memoraadum or specific ction en these events is required.
Jack E. R senthal, Chief Reactor 0p rations Analysis Branch Division o Safety Programs
^
Office of A glysis and Evaluation of Operatid al Data
Enclosures:
- 1. 227/87-028-01
- 2. 530/89-007-00
- 3. 528/89-005-00 i
Distribution:
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, On May 3, 1989 at approximately 0730 MST, Palo Verde Unit 3 was in a refueling
! outage with the core off-loaded when APS determined that deficiencies discovered during the installation of Potter and Brumfield (P&B) relays constituted a reportable condition pursuant to 10CFR21 and 10CFR50.73. The
, P&B relays are utilized in the PVNGS Engineered Safety Features Actuation i Systems and cause safety-related components to actuate when de-energized.
On August 3, 1988, APS reported a deficiency in the P&B MDR series relays
- (Reference LER 528/88-018). As a result, APS and P&B re-designed the relays for installation during the PVNGS Unit 1, 2, and 3 refueling outages. During post installation testing of the relays in Unit 3 on April 24 and 25, 1989 and 4
prior to declaring the relays operable, it was discovered that approximately twenty-five percent of the new model relays malfunctioned.
The cause of the relay malfunctions has been determined to be an inadequate methodology of applying an epoxy material to the relay coils to preclude contamination of the rotor and stator mating surfaces in the relay internals.
The epoxy causes the rotor and stator to bond which results in the relay failing to operate.
i l The complete root cause and corrective action is currently under investigation and will be reported in a supplement to this report.
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This report is also being provided pursuant to the provisions of 10CFR21. The narrative below includes the information requested by 10CFR21.21(b)(3);
however, it is being formatted to report this event in accordance with the requirements of 10CFR50.73.
I
- 1. DESCRIPTION OF WHAT OCCURRED:
A. Initial Conditions:
l The following plant conditions existed when the event described in this LER was determined to be reportable at approximately 0730 MST j on May 3, 1989 Palo Verde Unit 3 was in a refueling outage with the core (AC) off-loaded to the Spent fuel Pool (NO).
B. Reportable Event' Description (Including Dates and Approximate .
Times of Major Occurrences):
Event Classification: Condition which could have prevented the fulfillment of a safety function.
Note: This section includes information requested by 10CFR21 concerning the nature of the defect and dates on which information i
was obtained/ developed.
, On May 3, 1989 at approximately 0730 MST, APS determined that deficiencies discovered during the installation of Potter and
. Brumfield relays (RLY) in Unit 3 constituted a reportable condition pursuant to 10CFR21 and consequently 10CFR50.73.
1 i Prior to the event described in this LER, on August 3, 1988 APS reported a defect in Potter and Brumfield MOR series relays being
, utilized at PVNGS (Reference'LER 528/88-018). As corrective action to prevent recurrence, APS and Potter and Brumfield designed replacement MOR' series relays to be installed during each Unit's refueling outage. The re-designed relays were being installed
. during the Unit 3 first refueling outage. -During the post installation testing of the replacement relays on April 24 and 25, i 1989, several of the relays would not rotate to their de-energized position. Of forty-two (42) relays tested.in Unit 3, ten (10) relays:did not operate properly. Five (5) of the malfunctioning relays seized and the other five (5)' operated slowly. The malfunctioning relays were installed in the "B" Train Nuclear Steam Supply System Engineered Safety Features Actuation System (NSSS ESFAS)(JE).
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1 C. Status of structures, systems, or components that were inoperable at the start of the event that contributed to the event:
Prior to the installation of the replacement relays in Unit 3, the
- "B" Train NSSS ESFAS system was inoperable for the scheduled
. performance of a "B" Train electrical (EB) outage.
t D. Cause of each component or system failure, if known:
Note: This section includes information. requested by 10CFR21 concerning the nature of the defect and dates on which information i was developed.
An extensive investigation of the Potter and Brumfield (P&B) relay failures is being conducted. Personnel from P&B and an independent i
testing laboratory (HI-REL Labs) are assisting APS engineering personnel with the investigation.
The relay failures do not appear to be isolated to a particular
' model number, which would suggest a common mode failure. P&B Engineering and-Quality Control management personnel inspected the
- failed relays at PVNGS while they were installed in the NSSS ESFAS cabinet (CAS). Following the in situ inspection, the failed relays 2
were removed. Five (5) relays were provided to P&B for their failure analysis. HI-REL Labs management inspected several relays i
' at PVNGS. HI-REL was provided two (2) relays for an independent verification of the failure mechanism, i
i
! During the investigation of the cause of the relay malfunctions,
- APS and HI-REL Labs personnel discovered the presence of an epoxy I
material on some of the coil rotor and stator metallic surfaces.
l The epoxy material, which is utilized for coil insulation, was determined by APS and HI-REL Labs personnel to have caused the rotor and stator surfaces to bond together preventing the free rotation of the rotor by spring pressure when the coil is de-energized. (See Section 1.E and I.K for further information concerning the operation of the relays.) The epoxy material was confirmed to be present on-the samples inspected by P&B on April
- -27, 1989. The material was confirmed to be epoxy by HI-REL Labs and P&B on April-28, 1989.
E. Failure mode, mechanism, and effect of each failed component, if knowri:
The MDR relay malfunctions occur when the relays do not change position after they are de-energized. Normally, when the coils are de-energized, the rotor rotates approximately 30 degrees due to spring force. However, during the identified failures, the spring
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_ ..c s mu m force was not able to return the rotor to its de-energized position. The relays were " sticking" in their energized position.
This condition resulted in the relay contacts not properly changing state. The consequence of the. relay failures is that the related safety equipment would not be actuated as required.
- f. For failures of components with multiple functions, list of systems l or secondary functions that were also affected:
i The information concerning the function of the relays is discussed in Section I.K.
G. For failures that rendered a train of a safety system inoperable, estimated time elapsed from the discovery of the failure until the train was returned to service:
j The information required above is not considered appropriate for i the particular event being reported in this LER. However, in l general, it takes approximately 8-12 hours to replace a failed relay and conduct appropriate retests to return safety systems to j full operability.
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H. -Method of discovery of each component or system failure or procedural error:
l
- The relay failures were discovered during post installation testing of the relays as discussed in Section I.B.
I. Cause of Event:
f
} The cause of the relay malfunctions is considered to be improper i manufacturing of the relays by Potter and Brumfield.
l The root cause of the epoxy on the mating surfaces discussed in Section I.D is under investigation by P&B. A supplement to this report will be issued to describe the result of this-investigation.
- = J. Safety System Response
Not applicable - there were no safety system responses and none were necessary.
1 K. Failed Component Information:
Note: This section includes information requested by 10CFR21 concerning the identification of the firm supplying the ' basic component and the number and location of. the relays at Palo Verde.
The malfunctioning relays are manufactured by Potter & Brumfield
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0j0 05 1 os ,0 ;7 4 ru, ,, - - . - m , mu .. e 4 and are used in equipment supplied to Palo Verde by Combustion Engineer.ing (CE) and General Atomics (GA). The relays consist of a rotary actuator mechanism with the contact sections mounted in
, insulating rings on top. The actuator mechanism embodies a stator i assembly on which two relay coils are mounted. The two coils are connected in series inside the relay. When the coils are energized, a rotor turns through an arc of approximately 30 degrees. This operates the contact section on the extension of the rotor shaft. The travel of the rotor is confined to a 30-degree arc between the stator faces and the stop ring. Two springs return the rotor to the stop ring when the coils are de-energized. This
- also returns the contacts to their normal position. Thus, the
- relays provide an energized and a de-energized position. When the
- relay repositions to the de-energized position, various valves (V),
pumps (P), motcrs (MTR), etc. would be actuated.
t l The relays are supplied in a variety of sizes, coil voltage
- ratings, and contact numbers. At Palo Verde, nine (9) different l re-designed relays are being utilized. The relays that failed in
. Unit 3 were Models MDR-7061, 7062, and 7063 in the NSSS EUAS i cabinet. However, due to sne similarities in construction and
- materials, all Potter and Brumfield models could be subject to the
- same failure mechanism. No new model relays have been installed in Palo Verde Units 1 and 2.
l The MOR relays are used in three systems at PVNGS. These systems are:
l i) The Nuclear Steam Supply System Engineered Safety Features Actuation System (NSSS ESFAS)(JE).
ii) The Balance of Plant Engineered Safety Features Actuation System (B0P ESFAS)(JE).
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' iii) The Reactor Trip Switchgear (RTSG)(AA)(JD).
l The NSSS ESFAS uses the MOR relays as actuation relays. They are i
used to control valves and motors and to provide indication. .There is a total of 57 relays used in each NSSS ESFAS train. At two trains per unit, this adds up to a total of 342 relays used in the NSSS ESFAS systems for the three Palo Verde units.
The BOP ESFAS uses the MDR relays as actuation relays to provide t control of motors, valves, dampers (DMP), and emergency diesel generators (EK) (DG):following an actuation signal. Each BOP ESFAS ,
train has 30 MDR relays. -At two trains per unit, this adds up to a
The reactor trip switchgear uses one MDR relay for each reactor trip 3.; . ,.. u..
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breaker. The relay is used to provide an indication signal to the
! Plant Protection System (PPS)(JC) after a reactor trip breaker has opened. Failure of an MOR relay in this application would not
- prevent the reactor trip breaker from performing it's safety function of opening. There are 4 reactor trip breakers in each i unit. This leads to a total of 12 MDR relays used in the reactor
, trip switchgear (RTSG) systems (AA) at PVNGS.
i, II. ASSESSMENT OF THE SAFETY CONSEQUENCES AND IMPLICATIONS OF THIS EVENT:
i Note: This section contains the information requested by 10CFR21 concerning the nature of the safety hazard which is created or could be L created.
j It should be noted that the malfunctioning relays were discovered during 1 post installation testing in Unit 3 prior to their being returned to
- service. There are no new model relays. installed in Palo Verde Units 1 and'2. Therefore, the relays were never relied upon to perform a safety-related function. However, the failure of a relay in the ESF to 2
properly rotate by spring tension upon being de-energized by a valid safety system actuation signal would have prevented the associated valves, pump motors, etc. from operating as required for a safe plant
, shutdown. The failure of the relays in the RTSG to properly rotate i results in erroneous indication of reactor trip breaker (BKR) position
. to the PPS and in the Control Room. There are no other components which
, perform the same function as the relays that would be available during j an event.
III. CORRECTIVE ACTIONS:
i i This section contains the information requested by 10CFR21 concerning the corrective action which has been, is being, and will be taken; the i- organizations responsible for the corrective action; and the length of
- time for accomplishing the corrective action.
. A. Immediate:
As immediate corrective action, replacement of the Potter and Brumfield relays in Unit 3 was stopped in order to investigate the problem.
l B. Action to Prevent Recurrence:
$ Development of corrective action to prevent recurrence is under evaluation and will be based in part upon the results of the Potter and Brumfield determination of the cause of the epoxy material
! being on the relay stator / rotor interface surface. As an interim measure, APS is returning all "new model" replacement relays to P&B l for dissassembly, inspection, and testing. The relays which
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satisfactorily pass the tests and inspections will be re assembled and returned to PVNGS for re-installation and testing. The relays which do not pass the testing and inspection will be re manufactured, appropriately inspected, tested, and returned to PVNGS. A supplement to this report is expected to be issued by July 1, 1989 to describe the final corrective actions being taken.
IV. PREVIOUS SIMILAR EVENTS:
A previous similar event was reported in LER 528/88 018. Since the failure mechanism previously reported was different than the failure mechanism reported in this LER, the previous corrective action would not have prevented this event.
9 l
J Arizona Public Service Company P O SOE $3999 . P 40ENix art 2ONA 86072 3999 4 192-00477-JGH/TDS/DAJ May 8, 1989 4
- U. S. Nuclear Regulatory Commission NRC Document Control Desk Washington, D.C. 20555
Dear Sirs:
Subject:
Palo Verde Nuclear Generating Station _(PVNGS) l Unit 3
' Docket No. STN 50 530 (License No. NPF-74)
Licensee Event Report 89-007 00 File: 89 020 404 4
Attached please find Licensee Event Report (LER) No.89-007 00 prepared and submitted pursuant to 10CFR50.73. In accordance with 10CFR50.73(d), we ar~e herewith forwarding a copy of the LER to the Regional Administrator of the Region V office.
2 This report is also being submitted pursuant to 10CFR21 and includes
- information requested in 10CFR21.21(b)(3). ~In accordance with 10CFR21.2)(b)(2), three copies of this report are being provided to the Director, Office of Nuclear Aeactor Regulation.
! If you have any questions, please contact T. D. Shriver, Compliance Manager at 4
(602) 393-25?l.
Very truly yours, l YW ~4 *Mflf----
J. G. Haynes j Vice President Nuclear Production JGH/TDS/DAJ/kj Attachment cc: D. B. Karner (all w/a)
E. E. Van Brunt, Jr.
T. E. Murley (3 copies)
J. B. Martin T. J. Polich M. J. Davis
! A. C. Gehr e l
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On April 12, 1989 APS completed an evaluation of a deficiency identified by
- the manufacturer of the NNGS Units 1, 2, and 3 Atmospheric Dump Valves (ADV's). The ADV's are manufactured by Control Components Incorporated (CCI). Based upon APS' evaluation, it was determined that the deficiencies reported by CCI constituted a reportable condition pursuant to 10CFR21 and consequently 10CFR50.72 and 73.
On April 4,1989 CCI notified APS that an evaluation had been perfortned and that excessive internal valve leakage could result in the inability to remotely or manually operate the PVNGS ADV's. The cause of the excessive leakage is the result of an internal piston ring which fails to seat.
Excessive leakage by the piston ring results in high internal pressures which would preclude opening of the valve. .
A supplement to this report will be submitted to detail the final corrective e actions developed as a result of APS's ongoing investigation.
No previous similar events have been reported pursuant to 10CFR50.73.
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0l0l 5 - 0j0 0l2 or 1l5 This report is also being provided pursuant to the provisions of 10CFR21. The narrativebelowincludesinformationrequestedby10CFR21(b)(3);however,it is formatted to report this event in accordance with the requirements of 10CFR50.73.
- 1. DESCRIPTION OF WHAT OCCURRED:
A. Initial Conditions:
The following plant conditions existed when the event described in this LER was determined to be reportable at approximately 1254 MST on April- 12, 1989.
Palo Verde Unit I was in Mode 4 (HOT SHUTDOWN) at-approximately 2000 pounds per square inch (psi) and 325 degrees Fahrenheit (F). .
Palo Verde Unit 2 was in Mode 3 (HOT STANDBY) at normal operating temperature and pressure.
Palo Verde Unit 3 was in Mode 6 (REFUELING) at approximately 82 degrees F.
B. Reportable Event Description (Including Dates and Approximate Times of Major Occurrences):
Event Classification: Condition which could have prevented the fulfillment of a safety function.
Note: This section includes information requested by 10CFR21 concerning the nature of the defect and dates for which information was obtained/ developed.
On April 12, 1989 at approximately 1254 MST Arizona Public Service (APS) detennined that deficiencies identified by the manufacturer of the PVNGS Unit 1, 2, and 3 Atmospheric Dump Valves (ADV)(SB)(V) constituted a reportable condition pursuant to 10CFR21 and 10CFRSO.73.
On March 3,1989, a Palo Verde Unit 3 reactor trip occurred from approximately 98 percent power (Reference Unit 3 LER 530/89-001-00). Following the reactor trip, Control Room personnel-
.(utility, licensed and non-licensed) attempted to remove decay heat and control steam generator (AB)(SG) pressure utilizing the Atmospheric Oump Valves (ADV's)(SB)(V). Control Room personnel could not remotely operate the ADV's from the Control Room or Remote Shutdown Panel. Heat -removal was subsequently established by manually opening the ADV's. N'Gt Hver/ Pi/FicMfo t
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010 013 or 1l5 Because 'of the ADV problems encountered during the Unit 3 reactor trip event, APS engineering personnel have been conducting an extensive evaluation of the ADV design and operation. The original equipment manufacturer, Control Components incorporated (CCl), has been assisting during the APS evaluation. On April 4, 1989 CCI sent a letter to APS providing notification that a " potential significant deficiency" existed with the ADV design. Following receipt of this information, APS conducted an evaluation pursuant to 10CFR21 to determine the reportability of the information contained in the CCI notification. Further information was received from CCI on April 10, 1989 informing APS that local manual operation of the ADV's would not be possible if the deficient condition were to occur.
On April 12, 1989, PVNGS Engineering completed the evaluation and determined that the deficiency identified by CCI constituted a reportable condition.
The following discussion is intended to assist the reader in understanding the ADV's principle of operation. The disk stack (Figure 1) permits changes in flow rate while limiting flow velocity through the control element. The disk stack consists of a number of disks into which labyrinth flow passages have been etched to allow a fixed impedence. Impedence in the passages is developed by a series of right-angle turns, with a specific number of turns in each passage to limit the velocity to an acceptable level.
Since each disk has a known flow capacity, flow through the control element can be accurately measured and controlled. The position of the plug within the disk stack bore determines flow by exposing more or fewer disk passages.
With the valve in the closed position, upstream pressure fills the chamber above the plug by way of a controlled k across the piston ring. This provides a seating load squal o the inlet pressure times the full area of the plug.
When a signal to open the valve preceived, the actuator lifts the stem, opening the nJ1M oat ch results in the chamber pressure above the plug @ualizind h the downstrcam pressure. Upstream pressure acts upon the differential plug area and provides an axial biasing force which causes the plug to remain on the main seat.
As the valve stem continues to move in the opening direction, the pilot valve shoulder engages the plug to lift it off the main seat. The axial biasing force causes these opposing faces to remain in contact under all operating conditions.
When the plug is in the modulated mode, biasing force provided by
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l pressure' acting on the differential area overcomes fluctuating pressures from the fluid jets exiting the disk stack.
When a signal to close the valve is received, the actuator moves j the stem in the closing direction. The biasing force on the plug
- causes it to follow the stem until the main seat is contacted. The j actuator then seats the pilot section. Controlled leakage by the
- piston ring then fills the chamber above the plug providing additional seating force.
i C. Status of structures, systems, or components-that were inoperable at the start of the event that contributed to the event:
j Other than the ADV problems discussed in this LER, there were no structures, systems, or components inoperable at the start of the event which contributed to the event.
I D. Cause of each component or system failure, if known:
1 Note: This section includes information requested by 10CFR21 i
concerning the nature of the defect and dates for which information was developed.
As a result of the ADY malfunctions experienced at Unit 3, APS
- engineering contracted with CCI to assist in the root cause 4 investigation. The Unit 1, 2, and 3 ADV's were tested in- ,
accordance with approved test instructions. The purpose of the testing was to determine the force involved in the-operation of the l
ADV's and to characterize the positioner operation at normal i operating temperature and pressure. The results of the testing are i summarized below:
- 1. Test Results ll.!11L_1 i
ADV 184 was the first valve to be tested on March 14, 1989 using nitrogen gas supply at 95 psig. The valve did not stroke when given up to a 50 percent open demand signal. A
.l bonnet pressure tap was not installed at this time which made i- the valve malfunction difficult to analyze, i Following the malfunction of ADV 184, one operable ADV was
- required to allow Unit I to remain in Mode 3 for completion
- of additional testing. ADV 179 was tested on March 16, 1989
-and given 10 percent incremental open demand signals up to 50 percent. Nitrogen was used to stroke the valve with an initial pressure of 93 psig. It stroked very smoothly and
! followed within 6 percent of the demand signal. As a result get .tew wee
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010 015 or 1 1 5 w . . - ..c ma.e of' this test, ADV 179 was confirmed to be operable.
1 j On March 18,1989 ADV 178 was given both incremental and 4 step demand signals initially using nitrogen at 95 psig. As the valve opened through the disk stack transition region (approximately 15 to 20 percent open) it oscillated between
. 20 and 60 percent for several seconds. During this test, a i close signal was given to the valve and the valve closed.
After repeated testing, it was observed that the ADV did-not oscillate, but would stroke relatively smoothly. The testing 1
i j of ADV 178 was repeated using instrument air at normal supply i pressure; all strokes were smooth, and no oscillations were observed.
i ADV 185 experienced substantial oscillations when originally l tested using nitrogen supply. During the first testing on , -
j March 18, 1989, a 20 percent open demand signal was given and i the valve oscillated and closed. During additional testing the valve exhibited damped oscillation. It was observed that
- the more the valve was exercised, the more smoothly it would-
- stoke. The valve was manually stroked and then observed to operate smoothly. ADV 185 testing was repeated using
, instrument air; all cycles were smooth, the valve closely followed the input demand signal.
A second attempt to test ADV 184 was made on March 21, 1989 i using instrument air. This time ADV 184 began to open when
! given a 30 percent demand signal, but quickly shut on its I own. A 40 percent demand signal was then applied. The valve oscillated slightly, then opened 40 percent. The test was repeated several more times to a maximum open signal of 50
- percent. Each time the valve stroked smoothly.
l Unit 2 i
! All Unit 2 ADV's were stroked utilizing nitrogen at normal
- pressure (95 psig) and most utilizing instrument air at approximately 110 psig. A total of 22-tests were performed stroking the ADV's to 20 percent or more. No oscillations were observed and no instances occurred wherein the valves did not open.
{
Unit 3 Unit 3 ADV's with the exception of 179 were stroked utilizing nitrogen after the plant had been cooled down in Mode 5.
- (ADV 179 could not be tested since the actuator was damaged
- following the March 3, 1989 Unit 3 trip.) When ADV 178 was
! given a 10 percent open demand signal, the valve moved to 6 n..
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percent smoothly and the actuator force required to move the valve more than twice the expected force. Additional stroking consistently required excessive force to move the valve. In order to identify the source of the excessive resistance, the packing gland follower was loosened and approximately 50 percent of the packing was removed from the valve. Retesting the valve showed a significant reduction in the actuator force required to open the valve; however, it was still much higher than originally predicted. The actuator was decoupled from the valve. Stroking the actuator alone required approximately twice the predicted force. When the actuator was disassembled, an extra spring was found (two springs are specified by CCI). This explained the excessive force required to stroke the actuator.
- ADV 184 and 185 were both stroked and actuator forces were observed to be on the high end of the predicted range. Both ADV's 184 and 185 experienced a reduction in the opening force when the packing gland follower was loosened. During disassembly of both ADV actuators, a third spring was discovered to be improperly installed in both valves.
Summary I
During the testing described above, APS determined that the
[
i Unit 1 ADV 184 malfunction caused excessive bonnet pressure and, therefore, the force necessary to open the valve to exceed the capability of the actuator when the valve was being operated on the nitrogen gas supply. This discovery led to the development of revised test instructions to be performed on the ADV's in Units 1 and 2. The purpose of the r procedure was to verify all the ADV's would operate on both l the non-Class IE Instrument Air supply and the Class 1E nitrogen gas supply. The valves were stroked using the safety-grade nitrogen system and then repeating the test using the Instrument Air (IA) system. The IA system provides additional force for opening the valve since it is maintained I at 110 psig while the nitrogen system pressure regulator maintains pressure at 95 psig. An abnormally high bonnet pressure was suspected of causing the excessive force holding valve ADV 184 closed. As a result, a bonnet pressure tap was added and appropriate pressure measurements were taken.
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! Since ADV 184 had already been tested using the nitrogen
- accumulator, that portion of the test was deleted and the j
valve was stroked using the normal IA supply. The valve was tested in the following seq % ace and with the following j results:
i 1) A 10 percent demand open signal was given. The valve did j not move in response to the demand signal as expected, t
- 2) A 20 percent demand was then given and the pilot valve opened. This allowed the bonnet pressure to decrease and the condition of the seal ring to be determined. Bonnet pressure decreased _to 60 psig and then slowly increased to 110 psig (this is approximately 6 to 10 times higher thandesign).
- 3) Next, a 30 percent demand signal was given. The pilot valve opened and the bonnet pressure decreased to approximately 42 psig. The valve rapidly opened to 20 percent, the bonnet pressure rapidly increased from 42 psig to 110 psig,.and the valve shut.
- 4) A 40 percent demand signal was given. .The bonnet depressurized to-between 44 and 34 psig and the valve rapidly opened to 38 percent, closed to 6 percent, and then opened smoothly to 40 percent.
- 5) The valve was then given another 40 percent open demand signal. The bonnet depressurized to between 2 and 8 psig, and the valve opened smoothly to 45 percent.
- 6) A 30 percent demand was then repeated. The bonnet depressurized to approximately 2 to 7 psig and the valve stroked smoothly to 32 percent. The-valve was then given an incremental signal from 10 percent to 50 percent pausing at each 10 percent increment to allow the valve to stabilize prior to increasing demand.
The bonnet pressure measured on ADV 184 initially was 110 psig. This would require approximately 14,000 sounds. force (1bf) to open the valve. Based upon the availa)1e Instrument Air (IA) or nitrogen supply pressures, the IA system will not provide enough force to open the valve unless the bonnet :
pressure is less than approximately 80 psig . Also, the nitrogen gas supply will not provide adequate force to open the valve unless_the bonnet pressure'is approximately 60 psig.-
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0l0l 5 0l0 0;8 o, 1l5 CCI believes that the cause of the valve being able to stroke only to the pilot open position is excessive bonnet pressure due to excessive piston ring leakage. To investigate this
- hypothesis, CCI fabricated a fixturc in order to flow test the 12 inch piston ring. The flow test was conducted utilizing air at 1700 psi. CCI tested the design currently installed at PVNGS for 100 open-shut cycles. During one of the tests, excessive leakage resulting in high bonnet pressure w s observed. These tests were performed in late 1986 as a result of erratic performance observed on the non-safety related valves at another nuclear facility. The excessively leaking piston ring condition is random and cannot be predicted.
During further testing, CCI intentionally placed a 0.010 inch 1
high spot on the piston ring to simulate dirt. CCI then measured the leakage flow coefficient (Cv). The measured Cv corresponds to a leak which would be expected to resuit in excessive bonnet pressure.
A second series of tests were performed by CCI to investigate potential problems in the pilot plug area. CCI constructed full size models of the existing pilot plug and also designed a new pilot area. Both models were flow tested on a low pressure air flow system to determine their Cv and develop improvements to the design.
Prior to the malfunctions which occurred at PVNGS, CCI installed pressure taps on numerous valves which had failed to open at other facilities. The valves were always operable after instrumentation was installed. Consequently, CCI did not have any evidence that excessive bonnet pressure was the cause of the failure. The test at PVNGS on SG-HV-184 is the first valve failure during which representative pressure measurements could be taken.
Mechanical binding due to thermal expansion mismatch, hoop deflection due to pressure, and flow and galling due to high piston ring hub forces have also been postulated to be the cause. However, many valves have been disassembled and examined by CCI. No inordinate rubbing has been found and no visible reason for binding has been observed. CCI has performed thermal and stress calculations and did not find any mismatch or fit problems.
- 2. Root Cause CCI has over 200 similarly constructed valves in other nuclear facilities which have been in service for the last s........
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several years. The " stuck at pilot open" problem.has occurred least often with 8 inch plug valves, and most often with 12 inch plug valves. The sticking seems to be most likely when the valve is not stroked over r. period of time.
Based upon previous CCI experience, when a valve exhibits the problem observed at PVNGS, it has been discovered that stroking the valve for 3 to 4 cycles "re seats" the piston ring and the valve operates properly.
The following root causes have been provided by CCI based on their investigation of tb ADV problems experienced at PVNGS.
a) Dirt or foreign. material such as corrosion products.
(magnetite) is building up_on sealing surfaces of the piston ring when the valve is' closed. The piston ring would not be energized due to equal pressures on both -
sides of the piston ring. When the pilot plug is opened during attempted operations,_there is excessive piston ring leakage since the contamination holds the piston ring off the sealing surfaces. Cycling.of the ADV's three (3) or four (4) times allows the contamination to
" wash" away and the piston ring seal operates properly.
b) There is a vertical clearance of approximately 0.005 inch between the piston ring and the upper sealing surface.
CCI believes that, when the pilot valve is opened, the fluid rushing past this 0.005 inch upper clearance
! results in a dynamic pressure holding the piston ring j down, away from its sealing surface. To address this 4 scenario, CCI proposed " wave springs" which hold the
- piston ring in contact with its upper sealing surface at all times. There has been at least one instance of a i valve not opening as required with a wave spring installed to energize the piston ring.
! E. Failure mode, mechanism, and effect of each failed component, if l known:
! The failure mode, mechanism, and effect of potential ADV failures are discussed in Sections I.D and II.
I F. For failures of components with multiple functions, list of. systems or secondary functions that were also affected:
! Not-applicable - the ADV's do not have multiple functions.
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0l0 1l0 or 1;5 nm, - . - , _ . c m ., ,m G. For failures that rendered a train of a safety system inoperable, estimated time elapsed from the discovery of the failure until the train was returned to service:
The information requested by the above is not considered appropriate for the event being described in this LER. There have been no ADV failures at PVNGS wherein the capability to remotely and locally operate the ADV's was lost as a result of the causes described in Section I.D.
H. Method of discovery of each component or system failure or '
procedural error:
The inability to remotely operate the ADV's was originally discovered during the reactor trip event discussed in Section I.B.
Subsequent malfunctions were discovered during testing conducted after the Unit 1 trip. The cause of the ADV malfunctions was identified by CCI and provided to APS on April 4, 1989 as discussed in Section I.B. There have been no procedural errors discovered.
I. Cause of Event:
The cause of the event being reported in this LER has been determined to be an inadequate design by the original equipment manufacturer. Further investigation of the ADV problems is continuing and will be discussed in a supplement to this report expected to be submitted by June 15, 1989.
J. Safety-System Response:
Not applicable - there were no safety system responses and none were necessary.
X. Component Information:
Note: This section includes information requested by 10CFR21 concerning the identification of the firm supplying the basic component and the number and location of the relays at Palo Verde.
The PVNGS design incorporates the use of four (4) ADV's per unit (twelve total) as a means of providing decay heat removal in the event of a loss of offsite power. These valves are located between the steam generator and Main Steam Isolation Valves (SB)(V). The ADV's are manufactured by Control Components, Inc. (CCI) in accordance with Specification 13-JM-601A. They are model number 83G9-10-12P8 31NASI. The valves are pilot operated, pneumatically actuated drag valves. The valves are powered by a double acting, spring to close, pneumatic piston actuator. The actuator area is approximately 111 square inches developing over 10,000 lbf of
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thrust when one side is fully pressurized and the other side is vented to atmosphere. The design relieving capability is 1.47 x 10E06 pounds-mass (lbm) per hour.
- 11. ASSESSMENT OF THE SAFETY CONSEQUENCES AND IMPLICATIONS OF THIS EVENT:
The ADV's are used to remove decay heat from the steam generator in the event that the main condenser (SG) is unavailable for service for any reason including a loss of ac power. The decay heat is dissipated by venting steam to the atmosphere. In this way, the reactor coolant system (RCS)(AB) can either be maintained at hot standby conditions or cooled down. The system instrumentation and controls for the atmospheric dump valves are described below.
Initiating Circuits and Logic There are no automatic initiating circuits for operation of the atmospheric dump valves.
The atmospheric dump valves are positioned manually by a controller (manual loading station) from either the main control room or the remote shutdown panel as part of the 4
capability for emergency shutdown from outside the control room. Each valve has two separate permissive control circuits. Valve position indication is provided at each remote control station. A handwheel is also provided with the atmospheric dump valve for local manual eperation.
Bypasses, Interlocks, ar.d Sequencing No bypasses, interlocks, or sequencing are provided for the atmospheric dump valves.
Redundancy Two (2) redundant, atmospheric dump valves are provided for each steam generator.
The major accident scenarios which credit the use of the ADV's are:
6.3.3.4 - Post Loss of Coolant Accident (LOCA) long Term Cooling 15.1.4 - Inadvertent Opening of a Steam Generator Relief or Safety Valve (HSSV) 15.3.1 - Total Loss of Reactor Coolant Flow
- 15.4.1 - Uncontrolled Control Element Assembly (AA) y,.,.... ...
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Withdrawal from a Suberitical or low Power Condition 15.6.3 - Stearn Generator Tube Rupture In the event that all four (4) ADV's could not be opened upon demand (due to a failure of the pneumatic actuators to provide sufficient opening force by themselves as a result of the reported deficiency),
reactor decay heat will be removed through the Main Steam Safety Valves (MSSV's). The HSSV's will open when pressure in the steam generator reaches the pressure relief setpoints. Steam release will continue until the pressure is reduced to the safety valve reset pressure. The safety valves will continue to cycle in this manner as steam generator pressure increases and decreases. The RCS will remain at hot standby conditions during this pressure relief cycling. Hence, the RCS pressure boundary integrity will be maintained and the safety analysis will bound the consequences of the reported deficiency.
APS has reviewed Chapters 6 and 15 of the Combustion Engineering Standard Safety Analysis Report (CESSAR) and the PVNGS Updated Final i
Safety Analysis Report (UFSAR) and determined that the earliest the ADV's are required for any of the accident scenarios is 30 minutes from the onset of the particular accident. In these acenarios, the ADV's are used to cooldown the plant in the event of a loss of offsite power coincident with the particular accident. APS has reviewed the Chapter 15 CESSAR events and has found several instances wherein manual operation of the ADV's is credited. However, it should be noted that the safety analyses do not make a distir.: tion between " remote manual" or
" local manual" operation of the ADV's. APS considers that remote or local manual operation of the ADV's are equally valid methods of performing the manual operatit.1 discussed in the safety analyses.
APS was informed by the valve manufacturer on April 10, 1989 that neither the pneumatic actuator nor handwheel alone can produce sufficient force to open the valve for valve inlet pressures of 1150
- psia and the worst case piston ring seal leakage is assumed. However,
- CCI has indicated that if the pneumatic actuator is given a signal to
- open (remote manual operation) and the handwheel (local manual operation) is used to open the valve in conjunction with the pneumatic actuator, the combination will provide sufficient opening force to open the valve even with the valve inlet pressure equal to the lowest set MSSV plus accumulation (approximately 1302 psia) and worst case piston ring seal leakage assumed. Although the procedures are in place for remote or local operation of the ADV's, no procedures were in place for the combined remote / local operation of the valve at the time the ADV failed to open remotely at PVNGS. Hence, credit is not taken for the combined remote / local manual operation from a 10CFR21 reportability standpoint.
The loss of the remote and local manual operation (no credit taken for the combined remote / local operation) of the ADY's will not allow the
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Based on the above, the failure of all 4 ADV's to open due to a failure of their pneumatic actuators and handwheel assemblies has been determined to be safety significant. Loss of the remote and local operation of the ADV's adversely affects the ability of the plant to achieve or maintain safe shutdown conditions.
The consequences of the reported deficiency (loss of both remote and local valve operation) will result in the loss of the safety function (i.e., decay heat removal) cf the ADV's to the extent credited in the safety analyses presented in Chapter 6 and 15 of the UFSAR/CESSAR.
!!!. CORRECTIVE ACTIONS:
This section contains the information requested by 10CFR21 concerning the corrective action which has been, is being, and will be taken; the organizations responsible for the corrective action; and the length of time for accomplishing the corrective action.
A. Immediate:
PVNGS initiated an extensive investigation of the ADV malfunctions. As a result of APS concerns regarding the operability of the ADV's, Palo Verde Unit I remained shutdown following a reactor trip on March 5, 1989. Palo Verde Unit 2 was shutdown on March 15, 1989. Palo Verde Unit 3 remained shutdown and began a refueling outage on March 8, 1989.
In order to ensure the continued operability of the Unit 2 ADV's, APS has installed the capability to determine bonnet pressure.
This will enable the detection of excessive piston ring leakage.
APS is developing administrative controls for periodically monitoring for excessive piston ring leakage in the Unit 2 ADV's.
If excessive piston ring leakage is determined to exist during the periodic monitoring, the ADV(s) will be declared inoperable. These administrative controls will be in place and implemented prior to restarting Unit 2.
B. Action to Prevent Recurrence:
CCI has provided the following recommendations to eliminate the valve deficiency.
increase the pilot valve capacity. This requires rework of the plug to enlarge the pilot flow area and a new stem to scal the pilot valve when closed.
Use a two piece wedge style piston ring to ensure a good s.,....
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- E Ch 5i!M
- Phot % A AAl20NA 66072 2034 192-00467-JGH/TDS/0AJ April 17, 1989 U. S. Nuclear Regulatory Commission NRC Document Control Desk Washington, D.C. 20555
Dear Sirs:
Subject:
Palo Verde Nuclear Generating Station (PVNGS)
Unit 1 Docket-No. STN 50-528 (License No. NPF-41)
Licensee Event Report 89 005-00 File: 89 020-404 1-1
- Attached please find Licensee Event Report (LER) No.89-005 00 prepared and .
submitted pursuant to 10CFR50.73. . In accordance with 10CFR50.73(d), we.are
- herewith forwarding ~ a copy- of the LER to the Regional Administrator of the Region V office.
]
l This report is also being submitted to include the-information requested by.
1 10CFR21. In accordance with 10CFR21.21(b)(2), three copies three copies of i this report are being provided to the Director, Office of Nuclear Reactor l Regulation.
If you have any questions, please contact T. D. Shriver, Compliance Manager at I
(602) 393-2521.
i Very truly yours, h /Lt Mk J. G. Hay s
.Vice Pres dent i Nuclear Production-JGH/TOS/DAJ/kj l '.
.i Attachment-cc: D. B. Karner (all w/a)
E. E. Van Brunt, Jr. .
- - T. . E. Murley (3- copies) l J..B.JMartin i F T. J. Polich M. J.-Davis A. C. Gehr- -
> !NPO Records Center- -
~
H. L. Miller l
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