ML19248D016
| ML19248D016 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 07/11/1979 |
| From: | Capra R Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19248D017 | List: |
| References | |
| FOIA-79-98 TAC-11649, NUDOCS 7907300411 | |
| Download: ML19248D016 (4) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION e
a WASHINGTON, D. C. 20f 55
/
JUL 11 1979
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Docket No. 50-346 LICENSEE:
Toledo Edison Company (TECO)
FACILITY:
Davis-Besse Nuclear Power ation nit 1 (DB-1)
SUBJECT:
SUMMARY
OF MEETING H ON JUNE 29, 1979 TO DISCUSS RECENT PRESSURE SW
,7AILURES IN THE AUXILIARY FEEDWATER SYSTEM F DB-l' A meeting was held in Bethesda, Maryland on June 29, 1979 with representatives of TECO to discuss Licensee Event Report (LER)79-062 concerning the in-operability of several pressure switches associated with the DB-1 auxiliary feedwater (AFW) system. A list of attendees is included as Enclosure 1.
Background
In light of the Three Mile Island, Unit 2 accident, the Commission confimed by Order dated May 16,1979, TEC0's. undertaking of a series of actions, both immediate and long-tem, to increase the capability and reliability of the plant to respond to various transient events.
In addition, the Order confirmed that DB-1 would not be restarted until the immediate actions had been accomplished and found acceptable by the Director, Office of Nuclear Reactor Regulation (the Director).
Item (a) of the imediate actions required by the Order directed TEC0 to review all aspects of the safety-grade AFW system for DB-1 to further upgrade components for added reliability and performance. By letter dated May 23, 1979 (L. E. Roe (TECO)to R. W. Reid (NRC)), the licensee forwarded infomation, requested by the NRC staff, on the overall reliability of the AFW system with regard to operating history and design improvements at DB-1.
Based on the infomation presented in the May 23 letter and supplemented by many telephone conservations with the licensee's staff and a visit by members of the NRC staff to the DB-1 facility on June 8,1979, the staff concluded that TECO had increased the reliability of the AFW system by implementing appropriate corrective actions and design modifications to the system, based on previous AFW system component failures. However, on June 28, 1979 the NRC Bulletins & Orders Task Force first became aware of the additional component failures associated with LER 79-062. The Bulletins & Orders Task Force was responsible for assessing the licensee's compliance with the May 16 Order. The failures occurred on May 21, 1979 and were reported to the Office of Inspection and Enforcement, Region III on June 15, 1979. As a result of this new infomation, the staff requested a meeting with TEC0 to discuss the LER and its effect on the staff conclusions concerning the reliability of the 08-1 AFW system.
Enclosure 2 is a copy of TEC0's letter of May 23, 1979 and Enclosure 3 is a copy of LER 79-062.
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2-Discussion TECO pr'esented a discussion of the failures of the six pressure switches reported in LER 79-062. Three of the pressure switches were found inoperable and three others were found to have setpoints which were below the acceptable tolerance. The failures were discovered on May 21,1979 while performing a calibration check of the switches to satisfy the surveillance requirements of Technical Specification 4.7.1.2.d.
Three of the pressure switches are associated with the automatic shifting of water supplies for the AFW pumps from the condensate water storage tank (CWST) to the service water system (SWS).
Two of the switches are associated with automatic shutdown of the AFW pumps in the event suction is lost to the pumps. The remaining pressure switch is associated with the steam supply for the turbine for #1-2 AFW
- pump.
This pressure switch is one of four which will actuate on a loss of steam pressure to the AFW pump turbine and shut the steam supply valve to the turbine.
(The drop in steam pressure would be due to a rupture in the steam line between the steam supply valve and the pump turbine.) Enclosure 4 is a simplified sketch of the DB-1 AFW system showing the location and function of the subject pressure switches.
Based on the logic circuitry involved (as discussed in Enclosure 3) and the specific location and function r:f the failed components, it appears that had the AFW system been called upon to provide water to the steam generators and had no_ operator action been taken at any time during operation of the system, one AFW train would not have automatically shifted to its alternate water supply and its associated AFW pump (AFP #1-1) would have shutdown automatically on low suction pressure. The other AFW train would have shifted automatically to its alternate water supply (at 1.55 psig instead of the minimum allowed 2.8 psig) and continued to operate. The shutdown of AFP #1-1. would not have occurred until several hours into the event when the water supply in the CWST was almost depleted, In addition, only one AFW train is necessary to provide the required design flow rate of 800 gpm to the steam generators. TEC0 stated all of these switches perfomed thcir intended functions properly during the most recent monthly functional test conducted on March 8 and 22, 1979 for train #2 and train #1, respectively.
It was noted however, that a functional test will test the components' ability to perfom their function, but does not show that the components actuate at the proper setpoints.
A matter of concern to the NRC staff was that a detailed investigation of the cause of the failures was not conducted. The pressure switches involved were either recalibrated and returned to service or replaced.
TECO neither contacted the pressure switch vendor (Static-0-Ring) nor disassembled any of the failed switches to help determine the failure mode.
Conclusion In order to further enhance the reliability of the AFW system, with respect to this 559184
. type of ' component failure, the NRC staff concurred in the following comitments made by the licensee:
1.
Modify operating / emergency procedures te provide instructions to operators to manually shift (from the control room) to the alternate supply of water for the AFW pumps, when the level in the CWST drops to three feet (if automatic switchover has not occurred). This procedural change will provide greater assurance that, even with failures of this type, the AFW system will be provided with a source of water for long-tem cooling.
2.
In order to gather infomation to fom a data base from which to detemine the optimum calibration frequency of this type of pressure switch, TECO will perfonn a calibration check of the 16 pressure switches associated with the system once/ week for the first month and then once/ month thereafter until a sufficient data base has been established on the setpoint drift rate that an optimum calibration interval can be specified.
In addition, TEC0 will perfom a bench test calibration check on 2 piessure switches once/ week until the calibration interval has been established.
The calibration checks of the installed pressure switches will not take the place of the normal monthly functional test required by the DB-1 Technical Specifications.
3.
Two switches replaced as a result of the May 21, 1979 checks will be disassembled and inspected to aid in the determination of a failure mode.
4.
TEC0 will establish contact with the vendor to obtain any past reliability infomation.
5.
TEC0 will provide the staff with calculations showing the line losses (pressure drop) between the various pressure switches and the suction of the AFW pumps.
In addition TECO will provide the staff with verification that the AFW pumps installed in DB-1 are designed for a net positive suction head (NPSH) of "O psig". is a copy of the licensee's letter of July 2,1979 (L. E. Roe (TECO) to R. W. Reid (NRC)) documenting its comitment to items 1 through 4 above.
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Robert A. Capra, B&W Project Manager Bulletins & Orders Task Force
Enclosures:
Division of Project Management 1.
List of Attendees 2.
TEC0's ltr of May 23, 1979 3.
LER 79-062 4.
Drawing - DB-1 AFW system 5.
TECO's ltr of July 2, 1979 559195
i ENCLOSURE 1 LIST OF ATTENDEES MEETING HELD JUNE'29, 1979 - AFW SYSTEM DB-1 Toledo Edison Company L. E. Roe Vice President, Facilities Development T. D. Murray Station Superintendent E. C. Novak Gen. Supt. of Power Eng. & Const.
T.
Myers Nuclear Licensing Engineer Shaw, Pittman, Potts & Trowbridge J. E. Silberg Counsel for the licensee
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NRC D. F. Ross Director, Bulletins & Orders Task Force T. H. Novak Deputy Director, B&O Task Force G. S. Vissing DB-1 Project Manager, DOR G. R. Mazetis Section Leader, Systems Group, B&O Task Force D. C. Fisher Section Leader, Aux. System Branch F. S. Ashe Systems Group, BCO Task Force D. F. Thatcher Syst' ems Group, BCO Task Force A. C. Thadani Reactor Systems Branch R. H. Wessman IE Region II G. H. Cunningham Asst. Chief Hearing Counsel, OELD R. A. Capra B&W Project Manager, B&O Task Force 5591 8
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~J ENCLOSURZ 2 TcLEDO
%: EDISON Docket No. 50-346 LowELL E. ROE v.c.mm a.=
License No. NPF-3 F.=w a c a.a==
(4193 2SS-5242 Serial No. 508 May 23, 1979 Director of Nuclear Reactor Regulation Attention:
Mr. Robert N. Reid, Chief Operating Reactors Branch No. 4
~ ____
Division of Operating Reactors United States Nuclear Regulatory Cocmission Washington, D. C.20555
Dear Mr. Reid:
This letter transmits information for the Davis-Besse Nuclear Power Station Unit 1 (DB-1) requested in our meeting with NRC staff members on May 8,1979 as well as one item concitted to in Toledo Edison's letter dated May 4,1979 (Serial No. 500).
V discusses the overall reliability of the auxiliary feedwater (AW) syste= with ragard to operating history and design improvements at Davis-Besse Unit 1. discusses our review to verify the adequacy of the AFW system capacity.
Very truly yours, f
h,/eW pu LER:TJM cc:
J. Zwet::ig Operating Reactors Branch No. 4 Division of Operating Reactors U. S. Nuclear Regulatory Commission Washington, D. C.
20555 b "
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I4 559197 v
6 Il THE TCLECO EC! SON CCMPANY ECISCN PLAZA 2C0 MAC!SCN AVENUE TOLEDO. CHIO 4G552
Docket No, 50-346
~ License No, NPF-3 Serial No, SOS May 23, 1979 Past Auxiliary Feedwater System Reliability Imorevements For the purposes of this discussion, a failure of the auxiliary feedvater system (AFWS) is defined to be the inability of an auxiliary feedwater pu=p (AFP) to deliver water to a steam generator. Most of the failures at Davis-Besse Unit 1 occurred during the early life of the unit, and the failure trend significantly decreased thereaf ter, mainly due to implementation of several design improvements.
All of these failures were reported to the NRC through Licensee Event Reports (LERs).
Attachment A is a list of all these LERs with brief descriptions of the occurrences.
Several design improvements were made to enhance the reliability of the AFWS, as listed on Attachment B.
A design change initiated by a particular failure (lek) is indicated in the description of the LER to facilitate their correlation.
There have been an aggregate of 17 failures of the AFP or auxiliary feed pu=p turbine (AFFT) since entry into Mode 3 to date.
This is based on indivi, dual consideration of each train.
There have been a total of 120 actuations of AFP/AFFT,
since entry into Mode 3 to date; this includes all test actuations and all transients where SFRCS was actuated.
The maximum number of failures occurred during the initial operation and debugging phase of the facility.
It is e=phasized that fourteen (14) of the seventeen (17)
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failures occurred prior co January, 1978.
Consequent to implementing design changes listed on Attachment B, the failure rate has been greatly reduced and the reliability
- of auxiliary feedwater system enhanced.
~
As indicated above, there have been only three failures of the AFWS since January 1978 to date.
There vere a total of 65 actuations of the AFP/AFPT in this time period. Therefore, the failure rate was 0.046.
All these failures were valve ag_.
related. According to the B&W report titled " Eval.uation of. Transient Behavior and Snall Reactor Coolant System Breaks in the 177 Fuel Assembly Plant Volume III, Revision 1. May 19, 1979", the initiation of AFW can be delayed for at least 20 minutes after a loss of feedwater transient or a small break'in the Reactor Coolant System.
This provides enough time for the operator to manually actuate an af fected valve.
Figure 1 provides failure rate curves for the auxiliary feedwater system. For the sake of clarity the f ailures have been broken into four categories based on the cause of failures.
The curves shown in the figure follow an expected trend showing reduction in the final periods caused by i=plementation of several design improve =ents in the early plant life.
559188
.s Dooket No. 50-346 License No. NPF-3 Serial No. SOS May 23, 1979 Attachment A Summary of LERs Reoort #NP-32-77-Il occurred 7/27/77 Plant Mode 3:
During a surveillance test, all speed control and indication was lost to AFPT 1-2 caused by the f ailure of a " lower" speed relay and blown control power fuses.
Correction - the defective " lower" speed relay and blown control power fuses were placed on AFFT 1-2 speed control circuit. The insufficient current breaking capability of the relay leading to this inoperability was re-solved by putting additional contacts in series.
Design improvement: 2 Plant Mode 3: While AFPT 1-2 was inoperable, the motor control center (MCC) E11C was inadvertantly tripped by construction personnel.
The trip of the MCC resulted in a loss of power to steam supply valves (MS 106 and MS 106A) to AFPT 1-1.
This resulted in the inoperability of AFPT 1-1.
Correction - the circuits were inspected and tha breakers were successfully reclosed to re-energiz,e MCC E11C.
Design improvement: Not applicable Recor: #NP-33-77-46 occurred 8/3/77 Plant Mode 3:
The auxiliary feed pump 1-2 f ailed to start on a loss of feedwater trip signal on steam generator 1-1.
An investigation found the governor valve for the turbine had been closed due to construccion crew negligence. Workers were observed standing on this valve during maintenance.
The construction in this area is now over and only limited access is provided to this area.
Correction - the valve was manually opened and all workers were instructed to be careful during maintenance in this area.
Desing improvement: Not applicable Reoort #NP-33-77-53 occurred 8/4/77 Plant Mode 3: Auxiliary feed pump 1-2 was inoperable due to blown fuse in AFPT 1-2 discharge valve (AF 388) control caused by crack in the terminal board, Correction - The f aulty terminal board on AFPT 1-2 was replaced and terminal board on AFPT 1-1 was inspected.
Design improvement: Not applicable 559189 A-1
Docket No. 50-346 License No. NPF-3 Serial No. 508 May 23, 1979 Attachment A (Cont'd)
U Report #NP-33-77-45 occurred S/6/77 Plant Mode 3:
Auxiliary feed pump turbine 1-2 failed to start on SFRCS signal.
The cause was found to be a repositioned disconnect switch located in a high personnel traffic zone.
Correction - All personnel were instructed to use care while passing through this area.
Design improvement:
Not applicable Reoort #NP-33-77-51 occurred 3/6/77 Plant Mode 3:
Auxiliary feed pump turbine 1-2 was inoperable due to speed control relay failure.
Failure found during testing. Correction - Additional relay contacts were added to increase current breaking capability.
Design improvement: 2 Recort #NP-33-77-52 occurred 8/7/77 Plant Mode 3:
Speed control of AFPT l-1 was lost because of failure of relays in the AFFT speed control circuit and blown control power fuse.
Correction -
Additional contacts were added to i= prove the current breaking capabilities.
Design improvement: 2 Reoort #NP-32-77-16 occurred 9/24/77 Plant Mode 1: Power 263 MWT; Load 0 MWE: Turbine generator was of f-line, reactor was at about 9% power.
Due to spurious trip in the steam and feedwater rupture control system, normal feedwater flow was interrupted and auxiliary feedwater initiated.
Auxiliary feed pump No. 2 failed to come up to full speed initially due to binding in the auxiliary feedpump turbine govenor.
Rise in reactor coolant pressure due to interruption of normal feedwater flow caused the pressurizer electromatic relief valve to open. Missing relay in valve control circuit caused valve to rapidly cycle open and close nine times resulting in valve failing open.
Continued relief to quench tank within containment caused rupture disk to blow and discharge steam into containment. The remaining auxiliary feedwater train performed its intended function.
Correction - The binding in the auxiliary feed pump governor was rectified and the design was modified to eliminate binding.
Design improvement: 3 s
559190 A-2
Docket No. 50-346 License No. NPF-3 Sorial No. 508
~ May 23, 1979 Attachment A (Cont'd)
U Report FNP-33-77-80 occurred 10/16/77 Plant Mode 3:
Auxiliary feed pump 1-2 governor valve closed.
Excessive vibra-tion of the governor valve linkage caused by surging of the nearby startup feed pump motor caused the valve to close. Correction - Valve manually opened immediately.
AFP 1-2 was satisfactorily retested.
Design improvement: 9
~
Reoort #NP-33-77-83 occurred 10/15/77 Plant Mode 3: Motor operated auxiliary feedwater valve (AFP 2 discharge-valve to SG 2, AF 3872) inoperable due to defective torque switch. The valve failed to open when it was given an open signal. Failure found during testing.
Correction - Torque switch replaced.
Design improvement: Not applicable Reoort #NP-32-77-20 occurred 11/29/77 Plant Mode 1: Unit was at 40" full power with reactor trip setpoint at 507. power.
Testing resulted in an inadvertant increase in reactor power and a reactor trip at the 50% over power setpoint. ' Operator action opened electrical breakers between the generator and the transmission system. The action precluded autecatic transfer of station power needs to the off-site source. Loss of station power interrupted normal feedwater flow, started emergency diesel generators, and initiated auxiliary feedwater flow.,Three seconds after diesel generator start, one diesel generator tripped on overspeed due to incorrect trip setting; the other diesel generator carried load safely. Due to the overspeed trip of the emergency diesel generator, one auxiliary feed pump could not be started for a short duration until the emergency diesel generator was restarted. Correction - Emergency diesel generator overspeed trip setpoint was readjusted.
Design improvement: Not applicable Report 1NP-33-77-110 occurred 12/11/77 Plant Mode 1:
- 1) The reactor was tripped for 40% reactor trip test.
During trip recovery, SFRCS was actuated on low SG level on both SCs.
AFP 1-1 started properly but AFFT speed control was lost because of binding of governor speed setting shaft.
Correction - 0111te bushings for the speed setting shaf t bearing surface were installed in the governor housings for both AFPTs.
Design improvement: 4 559191 A-3
Docket No. 50-346
~
License No. NPF-3 Serial No. 508
.May 23, 1979 Attachment A (Cont'd)
V 2)
AFP l-2 started but later failed at the minimum speed, indicating loss of speed control. The f ailure was caused by blown control power f uses which also provide power to the governor speed changer motor.
Correction - speed control relays of higher current ratings were installed.
Design improvement: 2 Reoort #NP-33-77-116 occurred 12/28/77 Plant Mode 1: Power 2064 MNT,' Lead 689 MWE - Auxiliary feed pump turbine 1-2 speed control lost due to blown fuse and f ailed relays. Failure occurred during surveillance test.
Correction - Fuse and relays replaced and a modification designed and installed to correct design.
Design improvement: 2 Recort #NP-33-73-06 occurred 1/6/78 Plant Mode 1: Power 100 MWT, Load 0 MWE.
Speed control of AFPT l-1 was lost following SFRCS trip.
The high speed stop on AFPT l-1 was not set at proper speed.
Correction - relays in the AFFT speed control circuit were replaced.
Design improvement: 2
-Recort # NP-3 3-78-3 3 occurred 3/16/78 Plant Mode 1: Power 2102 MWT, Load 736 MWE; auxiliary feed water valve (AFP 1 to SG 1.stop valve, AF 3870) f ailed to open using motor operator. The check valves (AF39 and AF72) leaked causing a high differential pressure across valve AF 3870.
The motor operator limit switches were not initially adjusted properly, which caused the motor operator to torque out before opening. Correction -
The limit switch (which bypasses the torque switch) on the motor operator of AF 3870 was readjusted.
Design improvement: Not applicable Recor: #NP-33-79-03 occurred 1/2/79 Plant Mode 1: Power 2331 MWT, Lead 785 MWE; Steam supply valves (MS 106 and MS 106A) to auxiliary feed pump turbine f ailed to operate properly during testing due to dirt build up on stems.
The dirt buildup was cuased by adverse environmental conditions resulting from construction in this area. Correction - Valves made to cycle several times.
Design i=provement : Not applicable 2k 55919%
A-4
Docket No. 50-346 License No. UPF-3 Serial No. 508 thy 23,1979 Attachment B U
Desien Imorave=ents 1.
Capacitors were added to the four SG level bistable cabinets to filter r.oise off the analog inputs to these cabinets. This change improved the Steam and Feedwater Rupture Control System (SFRCS) performance. This modification was completed by 7/12/77.
2.
Auxiliary Feedwater Pump Turbine (AFPT) speed control raise and lower relays were replaced with relays of higher current ratings to avoid re-occurring failures of the above relays. -This modification resulted in enhanced relia-bility of the auxiliary feedwater system due to improvement in the /JPT speed changer control circuit. This modification was completed by 3/16/78.
3.
AFFT governor design was modified to remove portions of the pneumatic speed setting mechanism to eliminate possibilities of binding of the governor under certain conditions. The modification allows the pilot valve to overtravel allowing the pilot bearing to always remain in contact with the flo.ing lever.
This in turn removes the possibility that the turbine will be prevented from reaching design speed. This modification was completed by 10/14/77.
4.
Self oiling (Oilite) bushings were installed in both AFPTs for the.spe'ed
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setting shaft bearing surface in the governor housing to reduce the probability of AFPT governor binding at a speed other than design speed. This modifica-tion was completed by 12/15/77.
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5.
Two-second latch-in of computer alar =s from SFRCS was provided to track any spurious signals from the SFRCS. This modification was completed by 5/14/78.
6.
"Close" circuits on valves AF 3369 and AF 3871 were revised to be " Torque Seating" circuit to match the torque seating circuit on valves AF 3870 and AF 3872.
This modification was completed by 6/27/78.
7.
Annunciator alarms were provided for the following SFRCS trips to facilitate increased operator awareness.
a) Low steam generator level b)
High SG - main feedwater differential pressure c) Loss of all four reactor coolant pumps This was completed by 7/24/78 8.
Main feedwats.r - steam generator reverse differential pressure switches were relocated to avoid spurious trips on the SFRCS. This modification was com-pleted by 9/28/78.
9.
A spring was added between the upper valve lever and the lower support bracket of the AFPT governor valve. This will cause the valve to stay open wher required, regardless of the presence of vibration.
This modification was completed by 11/23/78.
s 559193 B-1
Docke t No. 50-346 License No. NPF-3 Serial No. 508 May 23, 1979 (Con'd)
~
10.
Dynamic braking resistors of higher current ratings were installed in the
.\\?PT speed changer control circuit to improve AFFT speed control. For d' tails refer to L. E. Roe letter to R. W. Reid, (Serial No. 505), dated May 19, 1979.
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559194 B-2
ENCLOSURE 3 e
e
- ) TOLEDO t
% EDISON L79-506 June 15,1979 FILE: RR 2 (NF-33-79-64)
Docket No. 50-346 License No. NPF-3 Mr. James G. Keppler Regional Director, Region III Of fice of Inspection and Enforcement U. S. Nuclear Regulatory Ccamission 799 Roosevelt Road Glen Ellyn, Illinois 60137
Dear Mr. Keppler:
Reportable Occurrence 79-062 Davis-Besse F
- clear Power Station Unit 1 Date of L _urrence: May 21, 1979 79-062 with a supplemental Enclosed are three copies of Licensee Event Report inf orr.at! 'n sheet which is being submitted in accordance with Technical Specifi-occurrence.
cation 6.9 to provide 30 day written not:.fication cf the subject Yours truly,
~l% 0 Terry D. Murray Station Superintendent Davis-3 esse Nuclear Power Statica TDM/SNB/ljk Enclosure cc:
Dr. Ernst Volgenau, Director Office of Inspection and Enforcement Enc 1:
30 copies LER 79-062 Mr. William G. Mcdonald, Director Office of Management Information and Program Control Enc 1:
3 copies LER 79-062
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559395 THE TCLECO ECISCN CCMPANY ECISCN PLAZA 3CO MaCISCN AVENUE TCLECO, CHr0 43552
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U. S. NUCLEAR REGULATORY cot.tMeSSION
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EVENT DESCRIPTION AND PROBABLE CONSEQUENCES h switches PSL107C. PSL4928B.1 lOn 5/21/79, the auxiliary feedwater ( AW) suction pressure
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I o i)I l and PSL4930A vere found inoperable, and the pressure setooints of PSL4929A. PSL49 ified tolerances. The station 4 J l and PSL4931A were discovered to be outside of the spec C
was in Mode 5 at the time of occurrence and throughout the correc tive action. At no I
o#s caused inocerability of either AW J j o j6 ) I time would the above cochinations of f ailures have There was no danger to the health and safety of the.public or station per-J C 7j l trains.
f ais lsonnel.
(NP-33-79-64)
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i o l Component f ailure and instrument drift are the apparent causes of the inoperability.
iii l Further investigation of the root cause of the failure vill be conducted when the e a u '5-i ment is returned to operation. On 5/22/79, the cressure switches were returned within 7
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TOLEDO EDISON CCMPANY DAVIS-BESSE NUCLEAR POWER STATION UNIT OhT SUPPLEMENTAL INFOR 'ghiFOR LER NP-33-79-6',
DATE OF EVENT: Ma y 21, 1979 FACILITY: Davis-Besse Unit 1 IDESTIFICATION OF OCCURgiNCE:
Failure of Auxiliary Feedwater (AFW) Suction Pres-surc Switches Conditions Prior to Occurrence: The unit was in Mode 5, with Pcwer (MWT) = 0, and Load (Gross MWE) = 0.
Description of occurrence: During performance of Maintenance Uork Order IC-269-79 on May 21, 1979 to satisfy surveillance requiremente. of Technf eal Specification 4.7.1.2.d, it was discovered that the Auxiliary Feedvater (AFY) suction pressure switches PSL-107C, PSL-49283, and PSL-4930A vcald not ac?vate to their desir.ed positions. In addition, during perfor=ance of this same 531ntenance work order, it was discovered that the AFW suction pressure switches TSL-4929A, PSL-49293, and PSL-4931A would actuate at a pressure belew the speciZied tolerances. Specific information on each pressure switch is provided on the attached tabulation. All of the switches were manufactured by Static-0-Ring, model numbers 12V2-E4-ITLL23 and 6V2-E5-TTX4.
The station was in Mode 5 at the time of the occurrence and throughout the correc-tive action. This incident is being reported as documentation of a co=ponent failure.
Designation of Acoarent Cause of Occurrence: The cause of the occurrence is component f ailure possibly due to vibration of the pressure switches. The three switches which would not a c t ua t e, even with 0 psig applied, were removed and a bench calibration was a t t c= p t ed. Although the switches could be adjusted to the required setpoint, subse-quent calibration checks indicated that the actuation setpoint could not be repeated.
The setpoints of the remaining three failed switches had drif ted outside of the speci-fied tolerance.
The root cause of chis occurrence is still under investigation.
Analysis,of__0ccurrence: There was no danger to the health and safety of the publ'c or to station personnel. PSL-107C, AFPT #2 Steam Inlet is ace of four such switches.
The logic behind their actuation is two cut of four, and with this switch inoperable, the icgic would not have been rendered inoperable. Of the remaining five defective sw it c hes, two were associated with AFP #1 sucticn pressure, and three were asscciated with AFP !2 suction pressure. On AFP #1, PSL-4928A&B measure "before strainer suction pressure" and PSL-4930A43 =casure "after strainer suction pressure". Cn PSL-4930A&3 the logic is such that either the "A" cy; "B" switch will provide the LER #79-062 559197
TOLEDO EDISON COMPANY DAVIS-BESSE NUCLEAR POWER STATION UNIT ONE PAGE 2 SUPPLEMENTAL INFORMATION FOR LER NP-33-79-64 Since PSL-4930B actuated within the specified col-intended safety function.
On AFP #2, PSL-4929A&B the equipment would have operated as required.
- erance, measure "before strainer suction pressure" and PSL-4931A&B measure "af ter strainer suc tion pressure". In the case of PSL 4931A5B switches, the "B" switch actuated within its specified tolerance; therefore, the equipment would have per-for=ed the intended safety function. The "before strainer suction" pressure 2.47 psig sw1;ches PSL-4929ALB both did actuate, however, PSL-4929A actuated at 1,55 psig instead instead of the minimum allowed 2.8 psig and PSL-4929B actuated at of the minimum allowed 2.8 psig.
Therefore, AFP l-1 operation could have been af fected by the defective pressure if the normal condensate switches. AFP 1-2 would have operated properly except tank supply to the AFW Pu=p 1-2 had failed, the automatic transfer to the service' water supply would have transferred at 1.55 psig instead of the minimum allowed 2.8 psig.
AFP l-2 would have provided the required supply of feedwater to the steam genera-tors if the normal supply of condensate had f ailed. Also, the monthly performance of ST 5071.01, "AFW Monthly Test", when the unit was in operation, verified that both AFP suctions were automatically transferred upon a loss of nor=al condensate supply.
Corrective Action: On May 22,1979, PSL-4929A, PSL-4929B, and PSL-4931A were The recalibrated to within tolerance under tiaintenance Work Order IC-269-79.
re=aining three pressure switches PSL-4930A, PSL-4928B, and PSL-107C were replaced under Maintenance Work Crder IC-272-79.
Surveillance Test ST 5071.01, " Auxiliary Feedwater Monthly Test" will be perfor=ed to prove operability of the pressure switches prio'r to the unit startup.
A program is being instituted by the Instrument and Control section to investigate the possibility of vibration ef f ects once the unit returns to power cperation.
In until a calibration history on these switches can be determined, the
- addition, frecuency of calibration checks will be increased f rem once every eighteen months to once every three months.
The required calibration frequency will be established frem this historical data.
This is Failure Data:
There have been no previously reported similar occurrences.
time the calibration check has been perfor=ed since the setpoints on these the first svitches were established.
LER #79-062
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