ML20212M468
| ML20212M468 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 08/14/1986 |
| From: | Donohew J Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| IEIN-86-047, IEIN-86-47, NUDOCS 8608260183 | |
| Download: ML20212M468 (6) | |
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August 14, 1986 Docket No. 50-?19 LICENSEES:
GPU Nuclear Corporation Jersey Central Power and Light Company FACILITY:
Oyster Creek Nuclear Generatina Station SUR.1ECT:
JUNE 12, 1986, MEETING WITH GPU NUCLEAR CORPORATION (GPUN) AND OTHER LICENSEES TO DISCUSS THE ERRATIC REHAVIOR OF STATIC-0-RING SWITCHES On Thursday, June 12, 1986, a meeting was held at the request of the NRC staff at NRC, Bethesda, Maryland. This meetino was with GPU Nuclear (the licensee 1, other nuclear power plant licensees, the Static-0-Ring (SOR) switch manufacturer and General Electric to discuss the erratic behavior of Snp differential pressure switches. is the list of individuals who attended the meeting. Attachment 2 is the daily highlight issued internal to the NRC by the staff to announce the meeting with selected utilities and the staff's agenda for the meetina. is the IE Information Notice No. 86-47, " Erratic Behavior of Static 'O' Ring Differential Pressure Switches," issued to all power reactor licensees on June 10, 1986. This Notice discusses the incident at laSalle Unit 2 on June 1, 1986, involvino these SOR switches. Attachment 4 is the material on GPU's experience with SOR differential pressure switches at Oyster Creek which was handed out and discussed at this meeting.
The following is a brief summary of the significant items discussed in this meeting and the decisions, if any, taken or proposed.
The erratic behavior of SOR differential pressure switches has been the subject of the licensee's prompt reporting of loss of operability of the reactor water level (pWL) instrumentation at Oyster Creek and the IE Infonnation Notice No.
86-47, dated June 10, 1986. The loss of operability in the Oyster Creek Operating Cycle 10 for the RWL instrumentation is discussed in the previous meeting summaries dated February 24 and March 14, 1986. Personnel from Oyster Creek attended this meeting along with representatives from the switch manufac-turer, General Electric and other licensees using these switches. The IE Infonnation Notice was issued and this meeting was called in response to an incident at LaSalle Unit 2 where only one of four SOR switches in the RWL low function channels tripped although the level had apparently gone below the setpoints for the channels by several inches and thus a malfunction of the reactor scram system occurred. This event is discussed in more detail in the enclosed information notice.
In this meeting, the information on the experience at Oyster Creek with the SOR switches in the RWL low and low-low functions was discussed. One conclusion of the meetina was that the event at laSalle 2 was not an 8600260103 860814 PDR ADOCK 05000219 O
2-August 14, 1986 indication of a failure of the SOR switch to operate but a failure of the SOR switch to operate near its calibrated setpoint. This conclusion is borne out by the licensee's experience at Oyster Creek with the RWL low function channels. [It should be noted that testing subsequent to this meeting did show that one SOR switch did fail.)
For the staff, the conclusions of this meeting were also that the erratic behavior of the setpoint for the differential pressure where the SOR switch is to operate was affected by the following:
(1) the static pressure in the switch, (2) the number of times the sw tch has recently operated, (3) the method by which the setpoint for the switch is calibrated and (4) the fact that each switch may have unique characteristics affecting this drift. For example, if the switch setpoint is calibrated at zero gauge static pressure as was done at Oyster Creek, raising and holding the pressure in the switch to the reactor operating pressure causes the actual trip point to shift in a non-conservative direction from the calibrated setooint. This erratic behavior and what may affect this behavior does not give confidence in the reliability of these SOR switches for the RWL low-low function which initiates engineered safety feature system actuation in response to an accident.
The staff stated at the conclusion of the meeting that an IE Bulletin may be issued on this subject. This will depend on the information gained by the staff from its study of the incident at laSalle Unit 2.
M
( Jack N. Donohew,'Jr., Project Manager BWR Project Directorate #1 Division of BWR Licensing Attachments:
1.
List of Attendees 2.
Staff's Daily Highlight for the June 17, 1986 Meeting 3.
IE Bulletin 86 47
- 4. Material Handed Out By Licensee cc:
R. Bernero G. Lainas M. Srinivasan G. Holahan J. Beard s
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Oyster Creek Nuclear Generating Station cc:
Ernest L. Blake, Jr.
Resident inspector Shaw, Pittman, Potts and Trowbridge c/o U.S. NRC 1800 M Street, N.W.
Post Office Rox 445 Washington, D.C.
20036 Forked River, New Jersey 08731 J.R. Liberman, Esquire Commissioner Rishop, Liberman, Cook, et al.
New Jersey Department of Energy 1155 Avenue of the Americas 101 Commerce Street New York, New York 10036 Newark, New Jersey 0710?
Eugene Fisher, Assistant Director' Regional Administrator, Region i Division of Environmental Quality U.S. Nuclear Reculatory Commission Departnent of Environmental 631 Park Avenue Protection King of Prussia, Pennsylvania 19406 380 Scotch Road Trenton, New Jersey 08628 BWR Licensing Manager P. R. Fiedler GPU Nuclear Vice President & Director 100 Interpace Parkway Oyster Creek Nuclear Generating Parsippany, New Jersey 07054 Station P. O. Box 388 Forked River, New Jersey 08731 Deputy Attorney General State of New Jersey Department of Law and Public Safety 36 West State Street - CN 11?
Trenton, New Jersey 08625 Mayor Lacey Township 818 West Lacey Road Forked River, New Jersey 08731 D. G. Holland Licensino Manager Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731
4.
N stribution for Meetino Summary dated August 14,1986 Facility: Oyster Creek Nuclear Generating Station
- gDecket'f11e'.52191 NRC PDR Local POR BWD1 Reading OC file RBernero JZwolinski JDonohew EJordan BGrires Glainas MSrinivasan GHolahan JTBeard OGC - Reth (for info only)
ACPS (10)
ABournia
- Copies sent to those persons on facility's service list
ATTACHMENT 1 MEETING TO DISCUSS THE ERRATIC BEHAVIOR OF SOR dP SWITCHES JUNE 12, 1986 NAME ORGANIZATION Wayne A. Priest SOR, Inc.
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Poss E. Johnson SOR Inc.
Harry P. Hartman SOR, Inc.
Jane Peternel SOR, Inc.
Russell C. Engel SOR, Inc.
Tom H. Crawforde Houston Light & Power Kenneth R. Eibon GPU Nuclear Tom Hoatson GPU Nuclear John Rogers GPil Nuclear Jack Donohew HRC Larry Nicholson NRC Bob Sanworth NRC Lamont H. Youngborg GE H. C. Pferrerlen GE Bob Pierson NRC M. Grotenhuis NRC Joe Holonich NRC T. J. Kenyon NRC G. J. Freeman WPPSS Majorie Widmeyer WPPSS Jon McGaw SCE Newell Porter WPPSS John 0. Bradfale NRC D. B. Vassallo NRC James J. Shea NRC Vincent D. Thomas NRC A. W. Dimenick NRC Cornelius T. Caddington Penn. Power & Light N. K. Trehan NRC Theodore C. Daldiaz Penn. Power & Light Martin G. Santic Commonwealth Edison James S. Abel Comenwealth Edison Kenneth L. Gvaesson Commonwealth Edison l
Michael S. Turbak Connonwealth Edison Frank M. Siler TVA William C. Ludwig TVA l
Dave Lynch NRC l
Clarence M. Root INPO Eliot Abolafia Northeast Utilities i
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5 NAME ORGANIZATION J. D. Thonpson TVA J. D. Wolcott TVA M. Srinivasan NDC W. Hodces NRC T. Collins NPC M. Virgilio NRC E. Baker NRC Raymond F. Schallor NRC Michael P. Purphy TED/SMA, Inc.
Daniel R. Muller NDC Donald J. Florek NRC Jim Prell NRC Bob Newlin NRC Sheldon Schwart7 NPC E.L. Jordan NRC G. C. Wright NRC D. Allison NRC R. Woodruff NRC J. T. Beard NRC G. Polahan NRC M. Campagnone NPC e
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Docket Nos.
50-373/374 June 11, 1986 MEMORANDUM FOR:
H. Denton R. W. Houston D. Eisenhut D.' Crutchfield J. Lyons E. Rossi H. Thompson G. Lafaas F. Miraglia T. Spets R. Bernero M. rye, ell T. Novak F. $chroeder THRU:
Gary M. Holahan, Director Operating Reactor Assessment Staff, NRR FROM:
J. T. Beard 0'perating Reactor Assessment Staff, NRR DAILY HIGHLIGHT - FORTHCOMING MEETING WITH SELECTED UTILITIES
SUBJECT:
Time & Date:
June 12, 1986 10 A.M.
Location:
Phillips Building Room P118 7920 Norfolk Avenue Bethesda, Maryland To discuss the current situation with static "0" ring
Purpose:
pressure switches Requested UTILITIES participants:
NRC U. Beard Penn. Power & Light M. Virgilio South. Cal. Edison G. Holahan TVA D. Crutchfield WPPS G. Lainas GPU D. Allison, et al N.E. NUC Comonwealth Edison K. Naidu Regions J. T. Beard Operating Reactor Assessment Staff, NRR
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4 STATIC "0" RING DIFFERENTIAL PRESSURE SWITCHES MEETING AGENDA JUNE 12, 1986 1.
Welcome and Opening Comments -
Holahan/ Jordan 2.
Summaries of Operating & Maintenance Utilities Experiences -
- LaSalle
- Oyster Creek
- Browns Ferry
- Sequoyah
- Others 3.
Safety Significance for BWRs (Design /
GE Licensing Basis) -
- Rx Water Level
- Other Applications 4.
Design and Operation of SOR dP Switches -
SOR, Inc.
5.
Discussion of Possible Future Actions -
Open
- Reporting
- Testing
- Calibration
- Potential Replacement or Modification 6.
Concluding Remarks llolahan/ Jordan
-s ATTACHMENT III SSINS No.: 6835 IN 86-47 UNITED STATES NUCLEAR REGULATORY COMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, DC 20555 June 10, 1986 IE INFORMATION NOTICE NO. 86-47: ERRATIC BEFAVIOR OF STATIC "0" RING DIFFERENTIAL PRESSUPE SWITCHES Addressees:
All boiling water reactor (BWR) and pressurized water reactor (PWR) facilities holding an operating license (OL) or a construction pemit (CP).
Purpose:
This information rotice is intended to advise licensees of erratic behavior of certain differential pressure switches supplied by SOR, Incorporated (fomerly Static "0" Ring Pressure Switch Company) which apparently caused failure of the LaSalle 2 reactor to scram automatically when it was operating with water level below the low level setpoint. Similar switches are also installed in the high pressure core spray system and the residual heat removal system.
It is expected that recipients will review this infomation for applicability to their reactor facilities and consider actions, if appropriate, to preclude the occurrence of a similar problem at their facility. Sugoestions contained in this notice do not constitute NRC requirements. Therefore, no specific action or written response is required.
The NRC evaluation of this incident is continuing. If specific action is determined to be necessary, a separate notification will be issued.
Sumary of Circunstances On June I,1986, LaSalle 2 experienced a feedwater transient that resulted in a low reactor water level. One of the four low level trip channels actuated, resulting in a half scram. The operator recovered level and operation was continued. Subseouent reviews by licensee personnel raised concerns that the level had apparently gone below the scram setpoint and thus a malfunction of the reactor scram system may have occurred. Based on this concern, the licensee declared an " Alert" and shut the plant down. The NRC dispatched an augmented inspection team to the site. Subsequently, the licensee found that the " blind" switches which operate on differential pressure perfom erratically. The licensee also found erratic operation for similar switches in the high pressure core spray system and the residual heat removal system which operate valves in the minimum flow recirculation lines. Based on these results, the licensee declared all emergency core cooling systems in LaSalle 1 and 2 to be inoperable.
Both units are in cold shutdown pending further evaluation of the problem.
-8606090487
. lune 10, 1986 Page 2 of 4 Description of Circumstances:
The following description was constructed from a preliminary sequence of events prepared by the auorented inspection team and from other input by the team.
At 4:PO A.M. on Sunday, June 1,1986, l.aSalle 2 was operating at 93 percent of full power. Both turbine-driven feedwater pumps were operating, with the "A" pump in manual control and the "B" pump in automatic control. The motor-driven feedwater pump was in standby. While a surveillance test was being conducted on feedwater pump "A", the turbine governor valve opened further and caused pump speed and reactor water level to start increasing. At about the same time, the automatic cortrol systems for both turbine-driven pumps incked out. The reactor operator regained control of feedwater pump "A" and ranback feedwater pump speed in an attempt to restore water level to the nominal value (36 inches on the narrow range recorder). A few seconds later when the control system was reset, the "B" feedwater pump controller automatically ranback the pump speed to zero for no apparent reason, Reactor water level started falling at about 2 inches /second.
Subsequently, the reactor protection system responded via separate level switches to the falling reactor water level by reducing recirculation flow to reduce power, and the operator started the motor-driven feedwater pump to increase level. The level continued to fall for a few more seconds before turning around. The minimum reactor scram setpoint required in the technical specification is 11 inches. The level channels are normally set to trip at 13.5 inches, and the operators sre trained to expect reactor scram by the time that the water level reaches 12.5 inches. As the level was falling, one of the four reactor scram level switches (the "D" switch) tripped at approximately 10 inches, causing a
" half scram." As designed, this did not initiate control rod motion. None of the other three level switches tripped during this transient. No reactor scram occurred during this transient, either automatically or manually.
In the BWR scram system logic, which is one-out-of-two-taken-twice, at least one instrument channel in each scram system must trip to generate a scram demand signal and thereby initiate control rod motion. Preliminary results of the investigation indicate that the reactor water level fell to a minimum value of about 4.5 inches on the narrow range instrumentation, which is several inches below the specified scram setpoint but still 13 to 14 feet above the i
top of reactor fuel. The period that the water level was below the specified scram setpoint value was approximately 2 seconds. After feedwater flow turned the transient around, the plant stabilized at a power level of about 45 percent.
The "B" scram system half scram was manually reset about 30 seconds later. The power level was increased to 60 percent about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> later.
Shortly after the subsecuent shift change, the oncoming shift engineer's review was effective in indicating that the reactor water level appeared to have fallen below the scram setpoint and the level switches may not have perfomed properly.
He then requested that an instrumentation technician check the calibration of the switches. The results were that the "A" and "C" switches, which are in the "A" scram system, tripped at 10 and 13.5 inches respectively during the calibration check; the "B" and "D" switches, which are in the "B" scram system, tripped at 11 and 13.5 inches respectively. The switches were read,iusted to
IN 86 47 June 10, 1986 Pape 3 of 4 trip at 13.5 inches. Based on these results, the operating staff believed that a malfunction of the scram system mey have occurred. An orderly shutdown of the plant was initiated at 2:00 P.M. (CDT). At 2:30 P.M., the resident inspector was notified, and at 5:30 P.M., the NRC Operations Center was called via the emergency notification system and informed of this event by the licensee.
At 6:20 P.M., the licensee decided that the "A" scram system had failed to perform during the transient. The "A" scram system was manually tripped providing a half scram on the side that had apparently malfunctioned. The orderly shutdown was continued, and an " Alert" was declared. Vhen all the control rods had been' fully inserted at 9:22 the next mornina, the Alert was terminated.
On Monday, June 2 the NRC determined that the incident warranted a thorough investigation. The NRC pegional Administrator dispatched an augmented inspection team to the plant site.
On Ponday evening, Jure 2, the licensee checked the calibration of the reactor scram water level switches by varying the actual level in the vessel. The results were that the "A" and "C" switches tripped at indicated levels of 9.0 and 6.9 inches respectively and the "B" and "D" switches tripped at 3.9 and 10.2 inches respectively. These data were obtained about 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after the switches had been calibrated accordina to plant procedures and suagest a non-trivial difference. Additional data obtained over the next two days by varyino reactor water level demonstrated continued erratic behavior of switch setpoints.
On Saturday, June 7, after calibrating the Static "0" Ring flow switch which actuates the minimum flow recirculation valve in the M gh pressure core spray systen, the licensee perfomed a different test using actual system flow. The switch actuated when flow was at 530 spm instead of 1000 gom where it had been set to actuate. The licensee found similar performance of flow switches in the residual heat removal system. The licensee now suspects all Static "0" Ring differential pressure switches and has declared all emeroency core cooling systems in both units to be inoperable. Both units remain in cold shutdown.
Discussion:
It appears at present that the water level decreased below the scram setpoint for about two seconds and reached a minimum level of about 4.5 inches. This is based on a recording from the narrow rance water level instrument and records from the startup testing data acquisition system which recorded levels from the same transmitter. Had the reactor operator been aware of this fact before the water level had increased to a level above the setpoint, the reactor operator would have been expected to scram the reactor manually.
The differential pressure switches which provide the water level trip input to the reactor scram system were provided by SOR, Incorporated. These level switches.
are not original equipment; but were installed during replacement of equipment in secondary containment. Affected licensees had determined that the original switches were not qualified to operate in the environment created by an accident.
Operation of the SOR switches has been demonstrated to be erratic with little correlation between the setpoints established during atmospheric pressure
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IN 86-47 June 10, 1986 Page 4 of 4 calibrations and switch actuations under system pressure conditions. Exercising the switches by applying successive differential pressure cycles appears to mask erratic setpoint behavior. Similar problems with SOR differential pressure switches have been reported at Oyster Creek.
Per plant procedure, the switches for reactor water level had been exercised prior to calibration following failure of the reactor to scram automatically.
For this reason, performance of the level switches may have been different during calibration than during the event. Further, none of the level switches in the LaSalle 2 reactor scram system operate in conjunction with individual level transmitters. Therefbre, the calibration and performance of the individual low level trip channels cannot easily be compared to each other. In effect, the operator is blind to switch performance.
The vendor has indicated that those plants identified in Attachment I have similar differential pressure switches. This list of plants includes pressurized water reactors as well as boiling water reactors. NRC intends to meet with representatives of General Electric Company, SOR Incorporated, and interested licensees at 10 A.M. on Thursday, June 12, 1986, in Bethesda, Maryland to discuss experience with the switches.
It is succested that licensees consider advising their reactor operaters of the laSalle incident and providina guidance to them as to how to promptly detect the occurrence of a similar problem at their plants and the proper remedial action to be taken.
No specific action or written response is required by this notice. If you have any cuestions regarding this matter, please contact the Regional Administrator of the appropriate regional office or this office.
Edward L. Jordan, Director Division of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement Technical Contacts:
J. T. Beard. MRR (301) 492-4415 Roger W. Woodruff, IE (301)492-7E07 i
Attachments:
1.
Plants with Similar Differyntial Pressure Switches 2.
List of Recently Issued IE Information Notices
IN 86-47 June 10, 1986 PLANTS WITH SIMILAR DIFFERENTIAL PRESSURE SWITCHES PLANT SOR MODEL NUMRER Penn. Pwr. & Light /Susquehanna 103/R202 So. Cal. Edison / San Onofre 103/R903 TVA/ Brown's Ferry 103/B212 TVA/Sequoyah 103/RB212 103/BR203 103/BB803 WPPS 103/RB203 GPU/0yster Creek 103/B905 103/BB21?
103/B212 103/R202 N.E. Nuc./ Millstone 103/B903 South Texas Projects 103/PB212 103/BBB03 Commonwealth Edison /LaSalle 103/B202 103/B212 103/B203 103/BP203 103/BB212 103/BP205 103/BB202
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TN 86-47 Juve 10,1986 LIST OF RECENTLY ISSUED IE INFORKAT!0N NOTICES Information Date of Notice No.
Sub,iect Issue Issued to 86-46 Inproper Cleaning And Decon-6/12/86 All power reactor tamiration Of Respiratory facilities holding Protection Equipment an OL or CP and fuel fabrication facilities 86-45 Potential Falsification Of 6/10/86 All power reactor Test Reports On Flanges facilities holding Manufactured By Golden Gate an OL or CP and Forge And Flange, Inc.
research and test 4
facilities 86-44 Failure To Follow Procedures 6/10/86 All power reactor k' hen Working In High Radiation facilities holding
' Areas an OL or CP and research and test reactors 86-43 Problems Vith Silver Zeolite 6/10/86 All power reactor Samplino Of Airborne Radio-facilities holding iodine an OL or CP B6-42 Improper Maintenance Of 6/9/86 All power rector Radiation Monitoring Systems facilities holding an OL or CP 86-41 Evaluation Of Questionable 6/9/86 All byproduct l
Exposure Readings Of Licensee
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Personnel Dosimeters l
86-32 Request For Collection Of 6/6/86 All power reactor Sup. 1 Licensee Radioactivity facilities holding Measurements Attributed To an OL or CP The Chernobyl Nuclear Plant Accident 86-40 Degraded Ability To Isolate 6/5/86 All power reactor The Reactor Coolant System facilities holding From Low-Pressure Coolant an OL or CP Systems in BWRS DL = Operating License CP = Construction Permit pp.--
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