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Category:REPORTABLE OCCURRENCE REPORT (SEE ALSO AO LER)
MONTHYEARML20202F9161998-01-29029 January 1998 Special Rept:On 980128,two of Three Fire Protection Pumps Were Removed from Svc as Result of Sys Configuration Necessary to Support Planned Maintenance.Fire Pumps Were Returned to Svc on 980128 ML20138F5801997-04-29029 April 1997 Special Rept:On 970402,declared Reactor Bldg Wide Range Gas Monitor Inoperable.Caused by Ruptured Pump Diaphragm. Initiated Work Order,Installed Replacement Pump & Declared Pump Operable ML20137Y2171997-04-15015 April 1997 Special Rept:On 961220,util Rail Car Was Released to Burlington Northern Railroad from Monticello Nuclear Plant. Rail Car Was Delayed in Chicago as Result of Problems W/Bill of Lading on Computer Sys.Rail Car Was Delivered on 970131 ML20101E8321996-03-15015 March 1996 Special Rept:On 960314,two of Three Fire Pumps Removed from Svc for Planned Mod Work.Returned Both Fire Pumps to Svc Same Day ML20094E4371995-10-31031 October 1995 Special Rept:On 951010,sys Engineer & Technicians Reintiated Work on B Channel of Reactor Bldg Vent Radiation Monitor, After Providing Proper Notification to Main Control Room. Caused by Incorrect Channels Removed from Svc ML20082R6601995-04-19019 April 1995 Special Rept:On 950404,discovered That Manual Isolation Valves for 'A' Channel in Closed Position.Cause of Failure to Correctly Position Valves Will Be Assessed.Valves Placed in Required Position & Demonstrated Operability ML20082E5661995-04-0606 April 1995 Special Rept:On 950326,pump Discharge Relief Valve for Electric Screen Wash/Fire Pump Became Inoperable.Valve Had to Be Secured to Permit Installation of an Isolation Device. Pump Returned to Operational Status After Device Installed ML20080G4731995-02-0101 February 1995 Special Rept:On 950119,two of Three Fire Protection Pumps Out of Svc for Period of Approx 15 H.Action Taken for Planned Preventative Maint on Two Check Valves & Drain Valve in Fire Suppression Sys.Roving Fire Watch Established ML20067D1351994-02-23023 February 1994 Special Rept:On 940126,fire Door 105 Declared Inoperable Due to Inability to Close Properly.Caused by Higher than Normal Differential Pressure Between Plant Administration Bldg & Turbine Bldg.Cause of Failure Corrected ML20117A5801992-11-24024 November 1992 Special Rept:On 921020,determined That RHR Sample Supply Loops a & B Excess Flow Check Valves Not Included in ASME Section XI Program.Caused by Lack of Proven Test Procedure. Valves Incorporated Into Third 10-yr ASME Program ML20082G9261991-08-15015 August 1991 Special Rept:On 910719,potential for Inoperable Penetration Fire Barrier Identified.Caused by Use of Less Conservative, But Technically Acceptable,Design Alternatives in Plant Const.Vents Will Be Rerouted ML20244E5941989-06-16016 June 1989 Special Rept:On 890603,two Out of Three Diesel Fire Pumps Were Inoperable for Less than 12 H.Caused by Diesel Fire Pump Day Tank Outlet Valve Being Closed Instead of Fill Valve During Performance Test 1158.Test Revised ML20081G9561983-10-25025 October 1983 Ro:On 831024,trip Coil for Reactor Recirculation Motor Generator Set 11 Drive Breaker Failed to Trip Automatically. Investigation Ongoing.Trip Coil Replaced ML20066E0131982-11-0101 November 1982 Ro:On 821031,thru Wall Defect Found on C 12-inch Recirculation safe-end to Pipe Weld Joint.Resolution Under Investigation ML20064E5141982-10-21021 October 1982 Ro:On 821020,thru Wall Defect Found on E 12-inch Recirculation Sys Safe End to Pipe Weld Joint.Indications Will Be Documented as Revision to RO 82-16 Reported on 820928.Resolution Being Investigated ML20064E4761982-10-11011 October 1982 Ro:On 821009,linear Indications Confirmed to Exist on Three Addl Welds in Recirculation Sys.Indications Will Be Documented as Revision to RO 82-16 Reported on 820928. Resolution Being Investigated ML20071N4361982-09-28028 September 1982 Ro:On 820928,crack Indication in End Cap of a Recirculation Riser Confirmed by Radiography.Indications Initially Detected by Ultrasonic Exam During Normal Inservice Insp. Remedial Measures Under Investigation ML20071L8261982-09-16016 September 1982 Ro:On 820915,leak Discovered on Primary Containment Suppression Chamber Nitrogen Control Sys Inboard Isolation Valve (AO-2378).Leakage Less than Tech Spec Allowable Leakage ML20053E8071982-06-0202 June 1982 RO Iii:On 810224,failure of Discharge Valve on Instrument & Svc Air Compressor Resulted in Loss of Instrument Air Sys Pressure Causing Plant Scram.Failed Check Valve Replaced W/Similar Unit ML20058L3431978-07-31031 July 1978 Ro:On 780728,control Rod Drive 30-47 Delayed Approx 1.4 Before Scramming.Buna N Disk in Plunger Broken in Pieces.Plungers on 242 Scram Pilot Valves & Two Backup Scram Pilot Valves Removed & Replaced ML20090L8311978-06-0202 June 1978 Ro:On 780505,steam Leak Noted in RCIC Inlet Steam Line & Drain Line to Condenser.Caused by Pinhole Failure on Weld on 300 Lb Socket Weld.Hole Temporarily Patched ML20091A9941978-03-10010 March 1978 Ro:On 780309,smear Survey of Chem Nuclear Sys Inc Model 4-45 Shipping Cask Disclosed Max Surface Contamination of 25,900 Disintegrations Per Min Per 100 Square Centimeters.Cask Will Be Decontaminated ML20127G6171978-02-0202 February 1978 Advises That on 780202,during Surveillance Test,One Suppression Chamber to Drywell Vacuum Breaker Failed to Reclose Following Exercise.Plant Shutdown Initiated ML20086D5141978-01-0606 January 1978 Telecopy Ro:On 780105,discovered That One of Two Nuclear Engineering Co Model B3-1 Shipping Casks Provided w/1-inch Diameter Lid Bolts Rather than 1-1/4 Inch Size Specified in Certificate of Compliance 6058 ML20086D5201977-12-14014 December 1977 Telecopy Ro:Heating Steam Coil Leak in Reactor Bldg Ventilation Supply Unit V-AH-4A Resulted in Freezing of Condensate at Inlet to Unit & Subsequent Inoperability of Associated Secondary Containment Isolation Dampers ML20086D5371977-10-14014 October 1977 Telecopy Ro:Insp of Internal Torus Catwalk Support Structure Revealed That Catwalk Mitered Sections Not All Attached to Horizontal Catwalk Support Plates in Some Manner.Attachment Locations Will Be Upgraded ML20086D5501977-10-13013 October 1977 Telecopy Ro:On 771012,local Leak Rate Testing of MSIV AO-2-80A & Nitrogen Instrument Air Sys Isolation Valve CV-7436 Indicated That Leakage Exceeded Tech Spec Acceptance Criteria.Cause Under Investigation ML20086D5541977-10-0505 October 1977 Telecopy Ro:On 771004,local Leak Rate Testing of HPCI Sys Discharge Isolation Check Valve HPCI-9 Indicated That Leakage Exceeded Tech Spec Acceptance Criteria.Cause Under Investigation ML20086D5571977-09-29029 September 1977 Telecopy Ro:On 770929,local Leak Rate Testing of Core Spray Inboard Isolation Check Valve AO-14-13A Indicated That Leakage Exceeded Tech Spec Acceptance Criteria.Cause Under Investigation ML20086D5611977-09-14014 September 1977 Telecopy Ro:On 770913,local Leak Rate Testing of Core Spray Inboard Isolation Check Valve AO 14-13B Indicated That Leakage Exceeded Tech Spec Acceptance Criteria.Related Event on 770914 ML20086D5671977-09-12012 September 1977 Telecopy Ro:Main Steam Drain Isolation Valve MD 2373 & Main Steam Outboard Isolation Valve AO 2-86A Exceeded Tech Spec Acceptance Criteria for Local Leak Rate Tests ML20086D6421977-07-11011 July 1977 Telecopy Ro:On 770709,determined That Recombiner Sys a Offgas Flow Control Valve PCV 7489A Could Be Opened W/ Controlling Solenoid Valve Deenergized.Valve Held from Svc Pending Investigation ML20058K6111977-03-0909 March 1977 Ro:On 770223,discovered Withdrawal of in-sequence Control Rod Resulted in Period Less than 5-s & IRM Scram ML20086D6911977-03-0202 March 1977 Telecopy Ro:On 770301,torus Water Vol Found to Be Slightly Below Min Vol Established by Tech Specs.Water Vol Restored to Normal Operating Level ML20086D7031977-02-24024 February 1977 Telecopy Ro:On 770223,while Withdrawing in-sequence Control Rod to Bring Reactor Critical,Reactor Period of Less than 5 Obtained ML20086D7881976-09-10010 September 1976 Telecopy Ro:On 760909,discovered That Torus Water Vol Several Hundred Cubic Ft Below 68,000 Cubic Ft Min Required by Tech Specs.Caused by Failure to Correct for Vol Vs Level Correlation ML20056B8541976-05-21021 May 1976 Informs of 760520 Event Involving TIP Bell Valve Which Failed to Properly Close During Operation of TIP Sys.Valve Cleaned,Relubricated & Closure Spring Tension Increased ML20086D8231976-05-0505 May 1976 Telecopy Ro:On 760504,during Routine Surveillance Test, Primary Containment Oxygen Concentration Found to Be in Excess of Tech Spec Limit.Caused by Leakage from Drywell Instrument Air Sys Into Containment Nitrogen Supply Line ML20127L2611975-09-22022 September 1975 Informs of 750921 Incident Re Countrate Decreasing Below 3 Counts Per While Performing Refueling Core Alterations. Countrate Increased & Refueling Operations Recommenced. Further Investigation Continuing ML20127L2751975-09-22022 September 1975 Informs of 750920 Incident Re Drywell Equipment Drain Sump Isolation Valves AO 2561A & B Exceeding TS Acceptance Criteria.Further Investigation Continuing ML20127L2551975-09-19019 September 1975 Informs That on 750919,during Local Leak Rate Testing of MSIVs 2-86A & 2-88A,combined Leakage of Valves Exceeded Acceptance Criteria.Further Investigation Continuing ML20127L2391975-09-15015 September 1975 Informs of 750914 Incident Re Discovery That Hydraulic Shock Suppressor Located Inside Primary Containment on Loop B HPCI Line Inoperable Due to Loss of Oil.Reactor in Cold Shutdown for Refueling.Further Investigation Underway ML20127L1741975-08-21021 August 1975 Discusses 750818 Incident Re Broken Seal Cooling Line at Threaded Connection on Reactor Water Cleanup Pump 12.Pump Immediately Isolated & Line Repaired.Investigation Into Cause of Failure Continuing ML20127L2651975-07-28028 July 1975 Ro:On 750727,operator Motor Failed & Stalled Due to Stalled Rotor Current Overheating Windows.Investigation Pending ML20127L2521975-07-14014 July 1975 Ro:On 750713,small Leak Developed on Local Pressure Gauge Monitoring Reactor Pressure.Root Valve for Pressure Gauge Closed,Allowing Ref Leg to Refill & Water Level Indication to Return to Normal ML20058L1761975-07-11011 July 1975 Ro:On 750624 & 25,10 Initial Core Fuel Assemblies Which Had Been Identified as Leakers & Removed from Core at End of Cycles 2 & 3 Inspected Using Underwater Television ML20058K6751975-07-0303 July 1975 Notifies That on 750703,during Routine Surveillance Testing, Reactor High Pressure Scram Switches Found Inoperable ML20127B7371975-05-27027 May 1975 Ro:On 750525 Core Spray Sys Motor Operated Valve Failed to Reopen by Means of Motorized Operator.Valve Was Returned to Normal Open Position by Local Handwheel ML20058K6951975-05-0505 May 1975 Ro:On 750505,discovered That Control Circuit for Valve Steam Supply to RCIC Turbine Was Deenergized by Motor Control Unit Undervoltage Relay Coil Opening.Coil Replaced & Remote Operability of Valve Demonstrated ML20127L1471975-05-0101 May 1975 RO 75-12:on 750429,surveillance Test for Exercising Unit 2 Containment Vacuum Breaker Isolation Valve CV-31628 Showed Valve to Be Inoperable 1998-01-29
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217D1261999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Monticello Nuclear Generating Plant.With ML20216E7031999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Monticello Nuclear Generating Plant.With ML20210Q0521999-08-0404 August 1999 Safety Evaluation Approving Relief Request 10 to License DPR-22 Per 10CFR50.55a(g)(6)(i).Inservice Exam for Relief Request 10,Parts A,B,C,D & E Impractical & Reasonable Assurance of Structural Integrity Provided ML20210Q6611999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Monticello Nuclear Generating Plant.With ML20209F7901999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Monticello Nuclear Generating Plant.With ML20195H0351999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Monticello Nuclear Generatintg Plant.With ML20206N1721999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Monticello Nuclear Generating Plant.With ML20205N0861999-04-12012 April 1999 Safety Evaluation Re Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves ML20205P5701999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Monticello Nuclear Generating Plant.With ML20204H4951999-03-18018 March 1999 SER Concluding That Licensee Established Acceptable Program to Verify Periodically design-basis Capability of safety-related MOVs at Monticello.Therefore Staff Concludes Licensee Adequately Addressed Action Required in GL 96-05 ML20205G7391999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Monticello Nuclear Generating Plant.With ML20202F7901999-01-25025 January 1999 1999 Four Year Simulator Certification Rept for MNGP Simulation Facility ML20199E4871999-01-0606 January 1999 SER Accepting Licensee 951116,960214 & 0524 Responses to NRC Bulletin 95-002, Unexpected Clogging of Residual Heat Removal Pump Strainer While Operating in Suppression Pool Cooling Mode ML20199F6211998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Mngp.With ML20205H0561998-12-31031 December 1998 Northern States Power Co 1998 Annual Rept. with ML20198P0691998-12-28028 December 1998 Safety Evaluation Concluding That NSP Proposed Alternative to Paragraph III-3411 of App III to 1986 Edition of Section XI of ASME Code Provides Acceptable Level of Quality & Safety.Alternative Authorized ML20198D0751998-12-10010 December 1998 Safety Evaluation Supporting NSP Proposed Change to EOPs to Use 2/3 Core Height as Potential Entry Condition Into Containment Flooding ML20198B2531998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Monticello Nuclear Generating Plant.With ML20195E3691998-11-12012 November 1998 Safety Evaluation Concluding That Licensee USI A-46 Implementation Has Met Purpose & Intent of Criteria in GIP-2 & Staff Sser 2 for Resolution of USI A-46 ML20195D2381998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Monticello Nuclear Generating Plant.With ML20198J4451998-10-22022 October 1998 Rev 2 to SIR-97-003, Review of Test Results of Two Surveillance Capsules & Recommendations for Matls Properties & Pressure-Temp Curves to Be Used for Monticello Rpv ML20154L3471998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Monticello Nuclear Generating Plant.With ML20153F0511998-09-21021 September 1998 Rev 2 to MNGP Colr,Cycle 19 ML20153E9361998-09-0808 September 1998 Rev 1 to MNGP Colr,Cycle 19 ML20153B0861998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Monticello Nuclear Generating Plant.With ML20237B8461998-07-31031 July 1998 Monthly Operating Rept for July 1998 for Monticello Nuclear Generating Plant ML20236W5041998-07-21021 July 1998 ISI Exam Summary Rept - Refueling Outage 19 ML20236R1941998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Monticello Nuclear Generating Plant ML20249A5861998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Monticello Nuclear Generating Plant ML20247K3971998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Monticello Nuclear Generating Plant ML20217D8731998-04-13013 April 1998 Rev 0 to MNGP Colr,Cycle 19 ML20217F6431998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Monticello Nuclear Generating Plant ML20216D1041998-03-0505 March 1998 Rev 21 to Operational QA Plan ML20216H6481998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for Monticello Nuclear Generating Plant ML20203G1431998-02-10010 February 1998 Rev 2 to Inservice Insp Exam Plan,Third Interval,920601- 020531 ML20203B2821998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for Monticello Nuclear Generating Station ML20202F9161998-01-29029 January 1998 Special Rept:On 980128,two of Three Fire Protection Pumps Were Removed from Svc as Result of Sys Configuration Necessary to Support Planned Maintenance.Fire Pumps Were Returned to Svc on 980128 ML20216D2071997-12-31031 December 1997 1997 Annual Rept for Northern States Power Co ML20197J8131997-12-31031 December 1997 Revised Evacuation Time Estimates for Plume Exposure Pathway Emergency Planning Zone at Monticello Nuclear Power Plant. W/One Oversize Drawing ML20198P2201997-12-31031 December 1997 Monthly Operating Rept for Dec 1997 for Monticello Nuclear Generating Plant ML20203J7131997-11-30030 November 1997 Monthly Operating Rept for Nov 1997 for Monticello Nuclear Generating Plant ML20199G7051997-11-19019 November 1997 Safety Evaluation Authorizing Relief Request 8 of Third 10 Yr Inservice Insp Interval ML20199H8181997-10-31031 October 1997 Monthly Operating Rept for Oct 1997 for Monticello Nuclear Generating Plant ML20217K2081997-09-30030 September 1997 Monthly Operating Rept for Sept 1997 for Monticello Nuclear Generating Plant ML20216H7771997-08-31031 August 1997 Monthly Operating Rept for Aug 1997 for Monticello Nuclear Generating Plant ML20217K2741997-07-31031 July 1997 Monthly Operating Rept for Jul 1997 for Monticello Nuclear Generating Plant ML20196H1081997-07-0808 July 1997 Rev 20 to Operational QA Plan ML20141B9271997-06-30030 June 1997 LOCA Containment Analyses for Use in Evaluation of NPSH for RHR & Core Spray Pumps ML20149E2921997-06-30030 June 1997 Monthly Operating Rept for June 1997 for MNGP ML20148S6341997-06-23023 June 1997 NPSH - Rept of Sulzer Bingham Pump 1999-09-30
[Table view] |
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Northern States Power Company 414 Necollet Mall Minneapohs, M:nnesota 55401 Telephone (612) 330-5500 June 2, 1982 Regional Administrator Region III U S Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, IL 60137
!!ONTICELLO NUCLEAR CENERATING PLANT Docket No. 50-263 License No. DPR-22 Supplemental Information on Event Involving Loss of
_ _ _Instrum_en_t_ Air _ C_ompressp_r_ an_d_ Subsequen_t_ _ Events At the NRC/NSP meeting on Octobr 8, 1981 covering the Systematic Assessment of Licensee Performance, NSP re;.esentatives discussed the above event. It was further agreed that NSP would provide Region III with a written summary of the event when the investigation had been completed, since the event was ultimately classified as non-reportable under the Monticello Technical Specifications.
Attached is a report that has been prepared for your information. A complete package containing all details of the investigation is available at the site for further review by NRC personnel, ua David Musolf Acting Head-Nuclear Support Services D5Dt/bd Attachment cc: Resident Inspector, NRC NRR Project Manager, NRC C Charnoff JUN 71987 8206100111 820602 PDR ADOCK 05000263 S pyg
ATTACHMENT TO LETTER DATED JUNE 2, 1982 TO R0-III Supplemental Information On Event Involving Loss of Instrument Air Compressor and Subsequent Events
- 1. Occurrence Date: February 24, 1981
- 2. Identification of the Occurrence:
Failure of a discharge check valve on an instrument and service air compressor resulted in the loss of instrument air system pressure, which led to a plant scram. -
- 3. Conditions Prior to Occurrence:
Steady-State Power - Plant operating at 99% power.
- 4. Method of Discovery:
Operational Event - Failure of the check valve was discovered upon shutdown of #13 air compressor. The subsequent drop in. instrument air pressure due to the leakage of air from the intake of the #13 air compressor indicated that the check valve had failed to function properly.
- 5. Description of Occurrence:
The #12 compressor was taken out of service for maintenance on Februa ry 23, 1981. This left two compressors (#11 & 13) available for supplying the air requirements of the plant. Immediately preceding the loss of air pressure the operators' noticed that the #13 air compressor was running hot and was noisy, so the #11 compressor was started up and the
- 13 compressor was shut down in preparation for maintenance of this unit.
The failure of the discharge check valve (AS-1-3) on #13 compressor, coupled with the failure of the channel valves (reed type) located in the compressor head, provided a direct path for bleed-down of the air system.
Plant air pressure commenced to drop immediately following shutdown.of #13 compressor.
The following is a chronological listing of the sequence of events immediately following this equipment failure.
Time 043144 Instrument air header alarmed low at 85 psig.
043305 Condensate Demin System trouble alarmed. This resulted from high system dp caused by the effluent valves
_ = _ _-
-2
, failing closed due to non-essential instrument air isolation at 80 psig.
Time 1
043315 Both feedwater pumps tripped on low suction pressure. A low suction pressure trip occurs at 85 psig and would result from closure of the condensate demin effluent valves.
043326 The reactor scramed from low water level (+10").
043417 Low-low reactor level (-48") interlocks properly tripped the reactor recirc pumps and started RCIC and the emer-gency diesel generators (HPCI was being started manually at the time this trip occurred). The MSIV's also properly closed to less than 90% open at this time.
043447 to The four low-low water level switches reset as water 043450 level inventory recovered. A graphical analysis (Attachment 1) indicates that the water level didn't drop much below the low-low water level trip setpoint. This occurred due to the manual initiation of HPCI concurrent with low-low level trip. This would indicate that the lowest water level during the scram was 6'-6" above the top of the core.
043753 The low level switches reset.
043858 The operators manually transferred power and tripped the turbine.
A chart of the instrument air pressure (Attachment 2) shows that the pressure dropped to as low as 18 psig before recovery occurred. Recovery was brought about by the isolation of the #13 compressor. Normal instrument .
air pressure was restored approximately 20 minutes after the initial blowdown occurred.
t
- 6. Description of Apparent Cause of the Occurrence:
Procedure - Although this event was initiated by a component failure, there are procedural changes which could have eliminated the scram. If the operators would have immediately closed the manual discharge valve when the compressor was taken out of service, the air system would not have bled down through the defective check valve and the scram would not have occurred. This isolation valve would have been shut in the normal course of events - a WRA (Work Request Authorization) would have been issued and the compressor would have been isc 'ed in accordance'with the appropriate OCD (Operations Controlling Document). This process does not provide for the immediate isolation of the compressor, however, as was required in this situatioc Also, these check valves are not inspected on a regular e -
4 - - , c-_m, _ . -
-3 .
basis. Routice inspection of these check valves will detect signs of degradation which could lead to eventual failure.
Component Failure - fhe check valves failed when one of the butterfly flappers of the valve broke. The apparent cause of failure was fatigue of this component due to the many cycles this valve was subject to over the history of the p. ant.
The channel valves failed due to wear and fatigue.
The reeds in these valves are flexed with each cycle of the air compressor - which is 450 rpm.
- 7. Analysis of Occurrence:
The instrument air system is not safety-related and the operation of this system is not required for the safe shutdown of the plant. All air-operated safety-related valves are designed for proper operation on loss of air.
These valves all operated properly during this event, and as a result, this event had no effect on the health or safety of the public.
A similar incident, where the loss of instrument air resulted in a plant scram, occurred on February 21, 1973 (Unusual Occurrence Report No. 39).
The cause of the instrument air loss was different, however - the failure of the #12 compressor in conjunction with the inadvertent isolation of the
- 11 compressor resulted in this particular occurrence (the plant had only two air compressors at that time).
- 8. Corrective Action:
The failed check valve was replaced with a similar unit. The reed valves on the #13 air compressor were repaired and the compressor was placed back in service. The procedure for shutdown of the air compressors (located in the Instrument and Service Air Operations Manual B.8.4.1) was modified to add a step which requires the closure of the manual discharge valve on the compressor to be taken out of service.
- 9. Failure Data:
The check valve that failed was a 3" Mission Duo Check Valve Figure K-15-SMF, Model A, 285 psi rating.
The channel valves that failed are the A-43 type supplied with the Ingersoll-Rand Model ESV-NL compressor part assembly Number 13172-D1.
- 10. License Considerations:
There were no violations of any federal or state regulations as a result of the failure of the instrument air system which is not a safety-related system. There are no Technical Specifications which apply to the instrument air system.
This event was reported (by phone) to the NRC Operations Center as an NRC Significant Event in accordance with 10CFR50.72. The resident NRC inspector was also notified immediately. This item was not reported to the NRC as a Reportable Occurrence.
_10CFR2J _ Repp_r_t_ing_ Status Reporting pursuant to 10CFR21 is not required because the occurrence does not involve a Defect or Failure to Comply.
- 11. _0pera tiona_1_ Cons _ide ra tion _s :
Following the scram, the recirculation pump discharge valves were lef t closed, with the result that they thermally clamped and could not be re-opened. The primary containment had to be de-inerted to provide access to the valves so that they could be opened manually. Recommended fixes to prevent this thermal clamping are discussed in General Electric SIL #368, dated December, 1981. Recommendations in the SIL are being pursued via the SIL followup process. Another item of operational concern is the fact that the recirculation pumps were not restarted as soon as possible following the scram, to prevent temperature stratification. Operations Manual Section C.4 was revised to assure the recirc pumps are restarted as soon as possible to avoid temperature stratification and to assure that valves are opened as soon as possible to prevent thermal clamping.
Following the scram, a recirc loop temperature transient occurred when placing the shutdown cooling system in operation. This portion of the incident is being investigated as a seperate event.
- 12. Recommenda_ tion _ fo_r_ P_revention_ of_ Similar_ Occurrences _:
In addition to the procedure revisions already made (Operations Manual B.8.4.1, Instrument Air, and C.4, Abnormal Procedures), the event investi- '
gator further recommended that the P.M. procedure for the air compressors (PM-4160) be modified to require the inspection of the discharge check valves whenever maintenance is performed on the associated air compressor.
The event investigator also recommended that a check valve manufactured by a dif ferent company be considered for replacement of the existing units.
Center Line, a unit of Ibrk Controls Corporation, manufactures a butterfly check valve which has a resilient lined insert which prevents the metal-to-metal contact between the valve body and the two hinged plates. The metal-to-metal contact between these items is suspected as having lead to the failure of the check valve discussed in this report. Prairie Island uses this type of check valve in their systems and highly recommends their use.
The use of two check valves in series could be considered, but the event investigator feels that with the replacement of this check valve with a more reliable unit, along with the procedural changes previously recommended, that this would not be required at this time.
The present P.M. schedule calls for an annual overhaul of the plant's air compressors. The Ingersoll-Rand Manufacturer Representative recommends
==_ ..
i that this maintenance be performed on a 6-month basis for compressors that operate continuously. The last time this P.M. had been performed on #13 compressor (preceding this incident) was May 14, 1980, so there was approximately 9 months operating time on #13 compressor preceding the channel (reed) valve failure. Since the #13 compressor is used on an almost continuous basis and the #11 and #12 compressors are used much less, the event investigator also recommended that the frequency of the P.M. for all compressors be increased to a 6-month schedule and that the units receive a complete overhaul, including the channel valves, when six months operating time has expired since the last overhaul (hour meters are l already installed on these compressors).
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