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| | issue date = 06/12/1987 | | | issue date = 06/12/1987 |
| | title = LER 87-011-00:on 870516,low Flow Reactor Trip Occurred. Caused by Hot Leg Loop Stop Valve Stem Failure Permitting Disc to Drop,Blocking Loop Flow.Procedure Revised to Reduce Stress on Valve stem.W/870612 Ltr | | | title = LER 87-011-00:on 870516,low Flow Reactor Trip Occurred. Caused by Hot Leg Loop Stop Valve Stem Failure Permitting Disc to Drop,Blocking Loop Flow.Procedure Revised to Reduce Stress on Valve stem.W/870612 Ltr |
| | author name = SAUNDERS R F | | | author name = Saunders R |
| | author affiliation = VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) | | | author affiliation = VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| | addressee name = | | | addressee name = |
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| {{#Wiki_filter:POW 28-06-01 *.* -NIIC Form 391} U.S. NUCLEAR REGULATORY COMMISSION 19-83) APPROVED OMII NO. 3150-0104 LICENSEE EVENT REPORT (LER) EXPIRES: 8/31185 FACILITY NAME 11) I DOCKET NUMBER 121 I PAG~ 131 Surrv Power Station. Unit 1 o I 5 I o I o I o I 2 I R In 1 OF n I -:i TITLE 1'1 Reactor Trip on Low RCS. Flow Due to Failure of Loop Stop Valve. EVENT DATE 16) LER NUMBER 161 REPORT DATE 171 OTHER FACILITIES INVOLVED (Ill MONTH* DAY YEAR YEAR /{ SEQUENTIAL | | {{#Wiki_filter:POW 28-06-01 U.S. NUCLEAR REGULATORY COMMISSION NIIC Form 391} |
| () REVSSION MONTH DAY YEAR FACILITY NAMES DOCKET NUMBERISl NUMBER NUMBER *.-01s1010101 I I o Is , I 6 011 -ol ,1 , -o I o ol6 112 sl1 0151010101 I I 8 7 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Ch.c:k on, or mor, of th1 following) | | 19-83) APPROVED OMII NO. 3150-0104 EXPIRES: 8/31185 LICENSEE EVENT REPORT (LER) |
| (111 MODE (II) N 20.402(bl 20.40ll(cl X II0.73(o)(21Civl 73.7,Cbl -I----POWER 20.406(0)(1 )Ii) II0.31Cc)l11 II0.73(o)(2llvl 73.7,Ccl LEVEL ----(10) ,, 0 I 0 20.406(0)(1 | | FACILITY NAME 11) IoDOCKET NUMBER 121 I 5 I o I o I o I 2 I R In I |
| )(Iii 60.311(c) 121 60.73(oll211vlll OTHER (S~ify in Abstr,ct -----b1tow ,nd in Text. NRC Form ., ...... :*.**:*. *.*.:;.* .*.;.;.; 20.40llloll1 | | 1 OF PAG~ 131 nI -:i Surrv Power Station. Unit 1 TITLE 1'1 Reactor Trip on Low RCS. Flow Due to Failure of Loop Stop Valve. |
| )(Ill) 50.73(01121111 II0.7311112llwiiillAI 366A) 1*.*.*.* *.*.*.;. *:-::*:*:*:* | | EVENT DATE 16) LER NUMBER 161 REPORT DATE 171 OTHER FACILITIES INVOLVED (Ill DOCKET NUMBERISl MONTH* DAY YEAR YEAR /{ SEQUENTIAL NUMBER () REVSSION NUMBER MONTH DAY YEAR FACILITY NAMES 01s1010101 I I o Is , I 6 8 7 011 |
| .*.*-*. :*.*.*. :*.*.*:* .*.*-*.; , ... , .... --:*:*:*:*:*:
| | - ol ,1 , - o I o ol6 112 sl1 0151010101 I I THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Ch.c:k on, or mor, of th1 following) (111 OPERATING MODE (II) |
| .... ,. *.* .. *.* .*.-*.;.*.;.*
| | N 20.402(bl 20.40ll(cl X II0.73(o)(21Civl 73.7,Cbl I-POWER LEVEL 20.406(0)(1 )Ii) II0.31Cc)l11 |
| 20.40ll(o)(1 lflvl li0.73(111211ill II0.731oll211wlllllBI , ........... | | -- II0.73(o)(2llvl 73.7,Ccl OTHER (S~ify in Abstr,ct (10) 0 I0 20.406(0)(1 )(Iii 60.311(c) 121 60.73(oll211vlll b1tow ,nd in Text. NRC Form 1*.*.*.* |
| ...... * *:*:* .-.*.*: ----, ...... , .*.*.*.**.-.
| | 20.40llloll1 )(Ill) 50.73(01121111 |
| *:*:*::-.*
| | - II0.7311112llwiiillAI 366A) 20.40ll(o)(1 lflvl 20.40llloll1 llvl li0.73(111211ill |
| *.;.*.-.;.-.
| | &0.73(oll2lliiil |
| :*:* 20.40llloll1 llvl &0.73(oll2lliiil 60.73(oll211xl LICENSEE CONTACT FOR THIS LER (121 NAME TELEPHONE NUMBER AREA CODE R. F. Saunders, Station Manager 0 1 o, q 3 I 51 7 I -13 I 1 I 8 14 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (131 MANUFAC-REPORTABLE . ;.;;.;. .*.**.; . ;,;.;.*.*
| | - II0.731oll211wlllllBI 60.73(oll211xl LICENSEE CONTACT FOR THIS LER (121 NAME TELEPHONE NUMBER AREA CODE R. F. Saunders, Station Manager 0 1 o, q 3 I 51 7 I - 13 I 1 I 8 14 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (131 REPORTABLE .;.;;.;. .*.**.;. ;,;.;.*.* ;.*..*.* MANUFAC-MANUFAC- |
| ;.* .. *.* MANUFAC-R ~~o~;:~iE | | .*.**.*. .*.*,*.*. *.*.:*.* SYSTEM COMPONENT CAUSE SYSTEM COMPONENT TURER TO NPRDS CAUSE TURER R ~~o~;:~iE ':J::1111:1::11:1:11111:1::1111:J:f i!iiliil!i:iiiji |
| ':J::1111:1::11:1:11111:1::1111:J:f i!iiliil!i:iiiji CAUSE SYSTEM COMPONENT TURER TO NPRDS .*.**.* . . *.*,*.*. *.*.:*.* CAUSE SYSTEM COMPONENT TURER *.* .. *.* *.*:*.*.* | | *=*=**=- :-:**.*. .....*.*.* *:*::*: |
| *.*::-:-*:*
| | y .....,.; ;.;.,;.; ......,.;..,.,*.*. ,* *:*: .;.; .;. |
| *=*=**=-:-:**.*. ..... *.*.* *:*::*: ;.;.-.* . .;,; :-:,::,:* | | X AIB I 1S1V I A1319 11 I I I I I I I I I I I I I .. |
| .;,::,:-:,:,:, ... ;,; .. ;,; :; ;.;._.;. ;.;,;, X AIB I 1S1V I A1319 11 y ..... ,.; ;.;.,;.; ...... ,.; . . ,.,*.*. I I I I I I ,* *:*: . ;.; .; . *.* .. *.-*.* .. *.* *.*. ........ *.*.*:*:**:*:
| | ; ,;,;*.;.*.*. I I I I I I I ;.-. :*:**:*: .... .*:*.*:*: :,:-:-:,: |
| *:* :*:*:*:,:
| | SUPPLEMENTAL REPORT EXPECTED 114) MONTH DAY YEAR EXPECTED lxl YES (If yos, compl*r* EXPECTED SUBMISSION DATE/ |
| ,;,;*,;. I I I I I I I *:*:*: *:*::*:* .... ;.;,; :-:**:* I I I I I I I *:* *:; .. *.* *.*: ,;.; .. ;,; ;., .... ; .-.*,;.:, ,;,;*.;.*.*.
| | ABSTRACT (Limit to 1400 SPIJCSs. i,tJ,, *PfJroxim,ttJly fifts6n singts-spacs typswrirrsn lim,s} (16) h NO SUBMISSION DATE 1151 018 310 817 On May 16, 1987 at 0824 hours, with Unit 1 at 100% power, a low flow reactor trip occurred when "A"loop reactor coolant system (RCS) {EI IS-AB} flow decreased to 47%. Following the reactor trip, the source range channels {EIIS-DET} did not automatically reinstate. All other protection and control systems functioned properly. Operators followed appropriate plant procedures and stabilized the plant following the reactor trip. |
| ;.-. :*:**:*: .... . *:*.*:*: :,:-:-:,: | | This event occurred when the. II A" hot leg loop stop valve (EIIS-ISV} stem failed, permitting the disc to drop, partially blocking loop flow. A detailed metallurgical analysis is being performed to determine the failure mode and mechanism of the valve stem. The preliminary report indicates that failure was due to stress or fatigue. To reduce the stress on the valve stem, the operating procedure is being revised to normally operate the valves*off the backseat. The failure of source range channels to reinstate was due to the under compensation of the intermediate range channel NI-36. The source range channels were manually reinstated, and technicians readjusted the intermediate range compensating voltage. |
| SUPPLEMENTAL REPORT EXPECTED 114) MONTH DAY YEAR EXPECTED h NO SUBMISSION lxl YES (If yos, compl*r* EXPECTED SUBMISSION DATE/ DATE 1151 018 310 817 ABSTRACT (Limit to 1400 SPIJCSs. i,tJ,, *PfJroxim,ttJly fifts6n singts-spacs typswrirrsn lim,s} (16) On May 16, 1987 at 0824 hours, with Unit 1 at 100% power, a low flow reactor trip occurred when "A"loop reactor coolant system (RCS) {EI IS-AB} flow decreased to 47%. Following the reactor trip, the source range channels {EIIS-DET} | | 8706230353 870612 Itl'- |
| did not automatically reinstate. | | PDR ADOCK 05000280 \ (\ |
| All other protection and control systems functioned properly. | | s PDR NRC Form 386 19-83) |
| Operators followed appropriate plant procedures and stabilized the plant following the reactor trip. This event occurred when the. II A" hot leg loop stop valve (EIIS-ISV} | |
| stem failed, permitting the disc to drop, partially blocking loop flow. A detailed metallurgical analysis is being performed to determine the failure mode and mechanism of the valve stem. The preliminary report indicates that failure was due to stress or fatigue. To reduce the stress on the valve stem, the operating procedure is being revised to normally operate the valves*off the backseat. | |
| The failure of source range channels to reinstate was due to the under compensation of the intermediate range channel NI-36. The source range channels were manually reinstated, and technicians readjusted the intermediate range compensating voltage. Itl'-,..,;.., 8706230353 870612 \ ( \ PDR ADOCK 05000280 s PDR NRC Form 386 19-83) | |
| NRC l'orm 366A * (9-83) FACILITY NAME (1) POW 28-06-01 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION U.S. NUCLEAR REGULATORY COMMISSION APPROVED 0MB NO. 3150--0104 EXPIRES: 8/31/88 DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) YEAR ::::::::::
| |
| SEQUENTIAL
| |
| ::::::::::*
| |
| REVISION *::::::::::
| |
| NUMBER .'."."."." NUMBER Surry Power Station, Unit 1 o 1s Io Io Io I 2 I 81 O 81 7 -011 I 1 -0 I O Ol 2 OF O I 3 TEXT '" more IP"C9 i& "'flUired, U&tl *dditiona/
| |
| NRC Form 36flA '&/ (17) NAC FOAM 366A (9-83) 1.0 Description of the Event On May 16, 1987 at 0824 hours, with Unit 1 at 100% power, a low flow reactor trip occurred when "A" loop reactor coolant system {EIIS-AB} (RCS) flow decreased to 47%. Following the reactor trip, the source range channels {EIIS-DET}
| |
| did not automatically reinstate.
| |
| All other protection and control systems functioned properly.
| |
| Operators followed appropriate plant procedures and stabilized the plant following the reactor trip. 2.0 Safety Consequences and Implications The low flow reactor trip automatically trips the reactor to maintain sufficient margin above a DNBR of 1.3 with a loss of RCS flow. The complete loss of flow in one loop from a reactor power of 100% (2441 MWt) with three loops operating is an analyzed event. During the event, "A" loop flow was maintained at approximately 47% and total core flow remained at 84%. A confirmatory analysis performed by Nuclear Engineering concluded that DNBR was maintained above the accident analysis value of 1.3. In addition, all other safety related systems remained operable during the event, and plant parameters remained well within the bounds of the accident analysis.
| |
| Therefore this event did not constitute
| |
| *an unreviewed safety ques*tion and the health and safety of the public were not affected.
| |
|
| |
|
| ===3.0 Cause===
| | POW 28-06-01 NRC l'orm 366A |
| The cause of this event was a failure of the "A" RCS hot leg loop stop valve (MOV-1590)
| | * U.S. NUCLEAR REGULATORY COMMISSION (9-83) |
| {EIIS-ISV}.
| | LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED 0MB NO. 3150--0104 EXPIRES: 8/31/88 FACILITY NAME (1) DOCKET NUMBER (2) PAGE (3) |
| The valve stern failed, permitting the disc to drop, partially blocking the loop flow. The valve stern is being analyzed for failure mode and mechanism.
| | LER NUMBER (6) |
| A supplemental LER will be submitted when this analysis is completed.
| | YEAR :::::::::: SEQUENTIAL ::::::::::* REVISION |
| The preliminary report indicates that failure was due to fatigue or stress. The failure of the source range channels to reinstate was due to the under compensation of the intermediate range channel NI-36.
| | *:::::::::: NUMBER .'."."."." NUMBER Surry Power Station, Unit 1 o 1s Io Io Io I 2 I 81 O 81 7 - 011 I1 - 0 I O Ol 2 OF OI 3 TEXT '" more IP"C9 i& "'flUired, U&tl *dditiona/ NRC Form 36flA '&/ (17) 1.0 Description of the Event On May 16, 1987 at 0824 hours, with Unit 1 at 100% power, a low flow reactor trip occurred when "A" loop reactor coolant system {EIIS-AB} (RCS) flow decreased to 47%. Following the reactor trip, the source range channels {EIIS-DET} did not automatically reinstate. |
| NRC Form 366A. (9-831 FACILITY NAME (1) POW 28-06-01 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION U.S. NUCLEAR REGULATORY COMMISSION APPROVED 0MB NO. 3150-0104 EXPIRES: 8/31 /88 Dq<:KET NUMBER (2) YEAR LER NUMBER (6) ::::::::::
| | All other protection and control systems functioned properly. Operators followed appropriate plant procedures and stabilized the plant following the reactor trip. |
| SEQUENTIAL | | 2.0 Safety Consequences and Implications The low flow reactor trip automatically trips the reactor to maintain sufficient margin above a DNBR of 1.3 with a loss of RCS flow. The complete loss of flow in one loop from a reactor power of 100% (2441 MWt) with three loops operating is an analyzed event. |
| :::::::::::
| | During the event, "A" loop flow was maintained at approximately 47% and total core flow remained at 84%. |
| REVISION ::::::::::
| | A confirmatory analysis performed by Nuclear Engineering concluded that DNBR was maintained above the accident analysis value of 1.3. In addition, all other safety related systems remained operable during the event, and plant parameters remained well within the bounds of the accident analysis. Therefore this event did not constitute *an unreviewed safety ques*tion and the health and safety of the public were not affected. |
| NUMBER :::::::::: | | 3.0 Cause The cause of this event was a failure of the "A" RCS hot leg loop stop valve (MOV-1590) {EIIS-ISV}. The valve stern failed, permitting the disc to drop, partially blocking the loop flow. |
| NUMBER PAGE (3) Surry Power Station, Unit 1 0 1s101010121a10 817 -01111-olo oh °Fo 1~ TEXT /ff more /J/NJCl1 is n,quired, U/ltl 11dditionsl NRC Form 3lil!iA '*) (17) NRC FORM 366A (9-83) 4.0 Inunediate Corrective Action The Operators performed all appropriate emergency procedures and function restoration procedures to ensure the plant was returned to a stable condition. | | The valve stern is being analyzed for failure mode and mechanism. A supplemental LER will be submitted when this analysis is completed. The preliminary report indicates that failure was due to fatigue or stress. |
| This included manually reinstating the source range channels.
| | The failure of the source range channels to reinstate was due to the under compensation of the intermediate range channel NI-36. |
| Also, the STA performed the critical safety function status tree review to ensure specific plant parameters were noted and that those parameters remained within safe bounds. 5.0 Additional Corrective Actions The unit was placed in the cold shutdown condition and the stem of the hot leg loop stop valve was replaced.
| | NAC FOAM 366A (9-83) |
| The stems of the other five Unit 1 loop stop *valves were ultrasonically tested and found to be satisfactory. | |
| Technicians readjusted the intermediate range compensating voltage. 6. O Actions Ta-ken to Prevent Recurrence A detailed metallurgical analysis is being performed to determine exact cause of the valve stem failure. To reduce the stress on the valve stem, the operating procedure is being revised to normally operate the valves off the backseat.
| |
|
| |
|
| ===7.0 Similar===
| | POW 28-06-01 NRC Form 366A. U.S. NUCLEAR REGULATORY COMMISSION (9-831 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED 0MB NO. 3150-0104 EXPIRES: 8/31 /88 FACILITY NAME (1) Dq<:KET NUMBER (2) PAGE (3) |
| Events A similar failure occurred on the Unit 1 "B" loop Hot Leg Isolation Valve on December 1, 1973. 8.0 Manufacturer/Model Number Anchor Darling/Drawing nos. 95-11778 and 95-11779. | | LER NUMBER (6) |
| * June 12, 1987 U.S. Nuclear Regulatory Commission Document Control Desk Oi6 Phillips Building Washington, D.C. 20555 Gentlemen: | | YEAR :::::::::: SEQUENTIAL ::::::::::: REVISION |
| e VIRGINIA ELECTRIC AND POWER COMPANY Surry Power Station P. 0. Box 315 Surry, Virginia 23883 Serial No.: Docket No.: Licensee No.: 87-013 50-280 DPR-32 Pursuant to Surry Power Station Technical Specifications, Virginia Electric and Power Company hereby submits the following Licensee Event Report for Surry Unit 1. REPORT NUMBER 87-011-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be reviewed by Safety Evaluation and Control. Very truly yours, R. F. Saunders Station Manager Enclosure cc: Dr. J. Nelson Grace Regional Administrator Suite 2900 101 Marietta Street, NW Atlanta, Georgia 30323 _._J}} | | :::::::::: NUMBER :::::::::: NUMBER Surry Power Station, Unit 1 0 1s101010121a10 817 -01111-olo oh °Fo 1~ |
| | TEXT /ff more /J/NJCl1 is n,quired, U/ltl 11dditionsl NRC Form 3lil!iA '*) (17) 4.0 Inunediate Corrective Action The Operators performed all appropriate emergency procedures and function restoration procedures to ensure the plant was returned to a stable condition. This included manually reinstating the source range channels. |
| | Also, the STA performed the critical safety function status tree review to ensure specific plant parameters were noted and that those parameters remained within safe bounds. |
| | 5.0 Additional Corrective Actions The unit was placed in the cold shutdown condition and the stem of the hot leg loop stop valve was replaced. The stems of the other five Unit 1 loop stop |
| | *valves were ultrasonically tested and found to be satisfactory. |
| | Technicians readjusted the intermediate range compensating voltage. |
| | : 6. O Actions Ta-ken to Prevent Recurrence A detailed metallurgical analysis is being performed to determine exact cause of the valve stem failure. |
| | To reduce the stress on the valve stem, the operating procedure is being revised to normally operate the valves off the backseat. |
| | 7.0 Similar Events A similar failure occurred on the Unit 1 "B" loop Hot Leg Isolation Valve on December 1, 1973. |
| | 8.0 Manufacturer/Model Number Anchor Darling/Drawing nos. 95-11778 and 95-11779. |
| | NRC FORM 366A (9-83) |
| | * e VIRGINIA ELECTRIC AND POWER COMPANY Surry Power Station P. 0. Box 315 Surry, Virginia 23883 June 12, 1987 U.S. Nuclear Regulatory Commission Serial No.: 87-013 Document Control Desk Docket No.: 50-280 Oi6 Phillips Building Licensee No.: DPR-32 Washington, D.C. 20555 Gentlemen: |
| | Pursuant to Surry Power Station Technical Specifications, Virginia Electric and Power Company hereby submits the following Licensee Event Report for Surry Unit 1. |
| | REPORT NUMBER 87-011-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be reviewed by Safety Evaluation and Control. |
| | Very truly yours, R. F. Saunders Station Manager Enclosure cc: Dr. J. Nelson Grace Regional Administrator Suite 2900 101 Marietta Street, NW Atlanta, Georgia 30323 |
| | _._J}} |
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ML18153A1921997-11-25025 November 1997
LER 97-010-00:on 971028,discovered Missed Fire Protection Surveillance Pt.Caused by Personnel Error.Satisfactorily Completed PT Procedure 0-OPT-FP-009 & Diesel Driven Fire Pump 1-FP-P-2 Declared operable.W/971125 Ltr
ML18153A1831997-11-12012 November 1997
LER 97-009-00:on 971014,declared Intake Canal Level Probes Inoperable Due to Marine Growth.Cause Indeterminate.Divers Inspected,Cleaned & Returned Probes to Operable Status & Initiated Interdepartmental Team to Investigate Cause
ML18153A1791997-11-0707 November 1997
LER 97-008-00:on 971011,invalid Actuation of ESF Occurred. Caused by Personnel Errors.Main CR Bottled Air Sys Isolated & Containment Hydrogen Analyzer Heat Tracing Actuation Signal Reset
ML18153A1721997-10-30030 October 1997
LER 97-007-00:on 970930,determined That Plant Was Outside App R Design Basis Due to Vital Bus Isolation Issue.Caused by Personnel Error.Installed Circuit Protective Device During Oct 1997 Refueling Outage
ML18153A1421997-06-10010 June 1997
LER 97-001-01:on 970123,shutdown Occurred Due to Drain Line Weld Leak.Inspected & Tested Turbine Trip Actuation circuitry.W/970610 Ltr
ML18153A1391997-05-28028 May 1997
LER 97-005-00:on 970502,Unit 1 Power Range Nuclear Instrumentation Was Inoperable Due to Personnel Error.Sro & STA That Were Involved in Event Were Counseled
ML18153A1291997-04-18018 April 1997
LER 97-006-00:on 970320,loss of Refueling Integrity Due to Inadequate Containment Closure Process & Verification.Fuel Movement Stopped IAW Action Statement Requirements of TS 3.10.B.W/970418 Ltr
ML18153A1281997-04-15015 April 1997
LER 97-004-00:on 970317,main Steam Safety Valve Was Outside as Found Setpoint Tolerance.Specific Cause Unknown,However, Minor Setpoint Drift Can Be Expected.No Immediate Corrective Actions performed.W/970415 Ltr
ML18153A1241997-04-0808 April 1997
LER 97-002-01:on 970116,one Train of Auxiliary Ventilation Sys Was Inoperable Outside of Ts.Caused by Personnel Error. Submitted Deviation Rept Re Reverse Rotation of Fan & Work Request to Adjust linkage.W/970408 Ltr
ML18153A1191997-03-19019 March 1997
LER 97-001-00:on 970218,manual Reactor Trip & ESF Actuation Occurred Due to Loss of EHC Control Power.Caused by Momentary Short.Relay Card Was replaced.W/970319 Ltr
ML18153A1201997-03-19019 March 1997
LER 97-003-00:on 970219,loss of Pressurizer Heaters Resulted in Manual U1 Trip & U2 ESF Actuation.Caused by Loss of Group C Pressurizer Proportional Heaters.Reactor Trip Breakers Were Verified open.W/970319 Ltr
ML18153A1131997-02-20020 February 1997
LER 97-001-00:on 970123,shutdown Occurred Due to Steam Drain Line Weld Leak.Management Was Notified & Shift Supervisor Invoked Requirements of TS 4.15.C.1.W/undtd Ltr
ML18153A1101997-02-13013 February 1997
LER 97-002-00:on 970116,one Train of Auxiliary Ventilation Sys Declared Inoperable.Caused by Personnel Error.Properly Adjusted Damper 1-VS-MOD-58B & Exited Seven Day LCO on 970116.W/970214 Ltr
ML18153A0951997-01-0202 January 1997
LER 97-002-00:on 961213,automatic Reactor Trip Occurred During Planned Shutdown.Caused by Steam Flow/Feedwater Flow Mismatch.Rps Functioned as Designed & Plant Placed in Hot Shutdown
ML18153A0931996-12-12012 December 1996
LER 96-008-00:on 961112,water Gas Decay Tank Oxygen Analyzer Pressure Sensors Inoperable Due to Vendor Supplied Equipment Not Meeting Procurement specifications.Post-implementation Procedures Revised & Transducers replaced.W/961212 Ltr
ML18153A0691996-09-19019 September 1996
LER 96-007-00:on 960821,failed to Complete Fire Detection Zone Inspections within Required Time Period.Caused by Personnel Error.Counseled Personnel Re Fire Detection Zone Inspections & Revised Fire Watch training.W/960920 Ltr
ML18153A0481996-08-26026 August 1996
LER 96-005-00:on 960803,manual Reactor Trip.Caused by Loss of Electro Hydraulic Control Pressure.Repaired Two Compression Fitting Union Connections on Leaking Fitting & Performed Evaluations on Other tubing.W/960826 Ltr
ML18153A0521996-08-20020 August 1996
LER 96-004-01:on 960510,discovered Hydrogen Analyzers Inoperable.Caused by Procedural Deficiencies.Implemented Permanent Changes to Hydrogen Analyzer Instrument Calibr Procedures.W/960820 Ltr
ML18153A0321996-07-30030 July 1996
LER 96-006-01:on 960618,anti-corrosion Coating Had Not Been Reapplied to Station Battery 2B.Caused by Procedural Error in That Verbatim TS Compliance Not Reflected in Procedures. Coating Was Applied to batteries.W/960730 Ltr
ML18153A0281996-07-17017 July 1996
LER 96-006-00:on 960618,failed to Apply anti-corrosion Coating to Station Battery 2B.Caused by Procedural Error. Applied anti-corrosion Coating to Batteries & Revised TS 4.6.C.1.f Re Battery Coating requirements.W/960717 Ltr
ML18153A0141996-07-0202 July 1996
LER 96-004-00:on 960606,turbine/reactor Trip Occurred.Caused by High Level in Steam Generator B.Placed Plant in Hot Shutdown Condition,Calculated Shutdown Margin & Monitored Critical Safety Function Status trees.W/960702 Ltr
1999-08-27
[Table view]
Category:RO)
MONTHYEARML18152B4411999-08-27027 August 1999
LER 99-005-00:on 990731,effluent Radiation Monitors Were Declared Inoperable.Caused by Degraded Heat Trace Circuits for Monitors Sample Suction Line.Degraded Heat Trace Circuit Was Replaced & Addl Heat Trace Is Being Installed
ML18152B4421999-08-27027 August 1999
LER 99-006-00:on 990802,determined That Plant Was Outside of App R Design Basis Due to Fire Barrier Deficiencies. Caused by Original Plant Design Deficiencies.Fire Watches Were Established & Mods Have Been Completed.With 990827 Ltr
ML18152B3771999-08-13013 August 1999
LER 99-004-00:on 990714,TS Violation Due to non-safety Related Fans Effect on CR Boundary Was Noted.Cause of Event Has Not Yet Been Determined.Cable Spreading Room Doors Were Operned to Reduce Pressure in Rooms
ML18152B4181999-05-18018 May 1999
LER 99-002-00:on 990425,MSSVs Tested Out of Tolerance for as Found Setpoint.Caused by Minor Setpoint Drift.No Immediate Action Required.Deviation Rept Submitted for Each Valve.With 990518 Ltr
ML18152B4111999-04-28028 April 1999
LER 99-003-00:on 990331,potential Loss of Charging Pumps Was Noted.Caused by Main CR Fire.Station Deviation Was Issued on 990331.With 990428 Ltr
ML18153A2741999-03-29029 March 1999
LER 99-002-00:on 990301,prematurely Released Fire Watches Resulted in Violation of TS 3.21.B.7.Caused by Inadequate Procedure.Procedure for Opening & Sealing Fire Stops Was Revised on 990212
ML18153A2681999-03-19019 March 1999
LER 98-013-01:on 981122,turbine/reactor Tripped on High Due to Short Circuit in Summator for MSL C Loop Channel III Flow Transmitter.Replaced 1-MS-FT1494 Summator & Module Repair Procedure Revised.With 9903190 Ltr
ML18152B7261999-01-21021 January 1999
LER 99-001-00:on 981222,auxiliary Feedwater Pipe Support Missed Surveillance.Caused by Personnel Error.Station Deviation Rept Was Submitted.Two Supports in Question Received Required Code Insp & Were Found Acceptable
ML18152B5811998-12-16016 December 1998
LER 98-013-00:on 981122,turbine/reactor Trip on High SG Level Occurred.Caused by Instrument Failure.Control Room Operators Placed Unit in Safe,Shutdown Condition
ML18152B5781998-12-16016 December 1998
LER 98-014-00:on 981126,manual Reactor Trip in Response to Main Feedwater Regulating Valve Failure Occurred.Caused by Dislocation of Retaining Clip in Positioner.Control Room Operators Placed Unit in Safe,Shutdown Condition
ML18152B7121998-12-0404 December 1998
LER 98-S01-00:on 981105,noted Failure to Deactivate Station Access Badge.Caused by Human Error.Licensee Will Now Deactivate Station Badges Before Clearance Is Revoked & Process for Badge Deactivations Have Been Strengthened
ML18152B7041998-12-0101 December 1998
LER 98-012-00:on 981102,noted That EDGs Were Concurrently Inoperable.Caused by Required Testing Per TS 3.16.B.1.a.2. Redundant EDG Was Returned to Svc within Two Hour Period, Following Satisfactory Testing.With 981201 Ltr
ML18152B6161998-11-0606 November 1998
LER 98-011-00:on 981008,diesel Driven Fire Pump Failed to Start During Performance of Monthly Operability Test.Caused by Faulty Overspeed Trip Device Failure.Diesel Driven Fire Pump Declared Inoperable
ML18152B6081998-10-23023 October 1998
LER 98-010-01:on 980715,intake Canal Level Probes Were Inoperable Due to Marine Growth.Caused by Design of Canal Level Instrumentation.Canal Level Probes Will Continue to Be Monitored More Closely
ML18152B7811998-07-31031 July 1998
LER 98-010-00:on 980715,low Intake Canal Level Instrument Channel I Was Declared Inoperable to Allow Testing of Intake Canal Level Probe 1-CW-LE-102.Subject Probe Was Cleaned by Diver,Tested & Channel I Was Returned to Operable Status
ML18153A2581998-06-0303 June 1998
LER 98-009-00:on 980509,nonisolable Leak of Reactor Coolant Pump Seal Injection Line Weld,Was Discovered.Caused by Lack of Fusion or Thermal Fatigue Coupled W/Vibration Stress Due to Loose Rod Hanger.Rcp Seal Injection Line Removed
ML18152B8241998-05-22022 May 1998
LER 98-008-00:on 980228,auxiliary Ventilation Fans Were Noted in Condition Outside of Design Basis.Caused by Failure to Recognize Potential Impact of Certain Design Basis Accident Scenarios.No Corrective Actions Needed
ML18152B7951998-04-29029 April 1998
LER 98-007-00:on 980330,radiation Monitors Were Declared Inoperable.Caused by Change in Operating Temperature Range. Preplanned Alternate Method of Monitoring Was Initiated IAW TS Table 3.7-6
ML18153A2521998-04-22022 April 1998
LER 98-005-01:on 980212,fire Watch Insp Exceeded One Hour. Caused by Lack of Attention to Detail by Individual Involved.Individual Involved Was Coached on Requirement to Perform Fire Watch Patrols within Required Time Frame
ML18153A2511998-04-22022 April 1998
LER 98-006-00:on 980324,unisolable Through Wall Leak of RCP Thermowell Was Noted.Cause of Leak Is Unknown.Rtd Will Be Replaced
ML18153A2391998-03-13013 March 1998
LER 98-005-00:on 980212,fire Watch Insp Frequency Exceeded One H Occurred.Category 2 Root Cause Evaluation Being Conducted to Determine Cause of Event.Station Deviation Issued
ML18153A2341998-03-0909 March 1998
LER 98-003-00:on 980226,no Procedural Guidance for Maintaining EDG Minimum Fuel Supply During Loop,Was Identified.Caused by Absence of Procedural Instructions. Deviation Rept Submitted to Document Deviating Condition
ML18153A2301998-03-0606 March 1998
LER 98-004-00:on 980206,fire Watch Was Released Prematurely Resulting in Violation of Ts.Caused by Inadequate Planning of Repair Activity.Work Orders Will Include Ref to Applicable Procedures Developed to Assist in Repairs
ML18153A2251998-03-0404 March 1998
LER 98-002-00:on 980202,automatic Turbine Trip Resulted in Automatic Reactor Trip.Caused Degraded Generator Voltage Regulator sub-component Failure.Placed Plant in Safe Hot SD & Replaced Intermittent Relay & Relay Socket
ML18153A2201998-02-0606 February 1998
LER 98-001-00:on 980108,deficient Test Due to Faulty Test Equipment Resulted in TS Violation.Caused by Faulty Vibration Analyzer Cable or Loose Connection.Station Deviation Rept Was submitted.W/980206 Ltr
ML18153A2071998-01-13013 January 1998
LER 97-012-01:on 971028,loss of Power to Latching Mechanism on Several Doors Occurred.Caused by Tripping of Two Breakers in Security Distribution Panel.Reset Affected Breakers Which Restored Power to Security Systems & Affected Doors
ML18153A2101998-01-13013 January 1998
LER 97-009-01:on 971014,declared Intake Canal Level Probes Inoperable Due to Marine Growth.Caused by Inadequate Maint of Intake Canal Level Probes.Subject Probes Were Cleaned, Tested Satisfactorily & Returned to Operable Status
ML18153A1911997-11-26026 November 1997
LER 97-011-00:on 971030,determined That Periodic Test Procedures for Testing Reactor Trip Bypass Breakers Did Not Test Manual Undervoltage Trip.Caused by mis-interpretation of Term in-service. Procedures Revised
ML18153A1971997-11-26026 November 1997
LER 97-012-00:on 971028,loss of Power to Latching Mechanism on Several Doors Occurred.Caused by Tripping of Breaker in Security Distribution Panel in Central Alarm Station (CAS) Panel.Breakers in Affected CAS Panel Reset
ML18153A1921997-11-25025 November 1997
LER 97-010-00:on 971028,discovered Missed Fire Protection Surveillance Pt.Caused by Personnel Error.Satisfactorily Completed PT Procedure 0-OPT-FP-009 & Diesel Driven Fire Pump 1-FP-P-2 Declared operable.W/971125 Ltr
ML18153A1831997-11-12012 November 1997
LER 97-009-00:on 971014,declared Intake Canal Level Probes Inoperable Due to Marine Growth.Cause Indeterminate.Divers Inspected,Cleaned & Returned Probes to Operable Status & Initiated Interdepartmental Team to Investigate Cause
ML18153A1791997-11-0707 November 1997
LER 97-008-00:on 971011,invalid Actuation of ESF Occurred. Caused by Personnel Errors.Main CR Bottled Air Sys Isolated & Containment Hydrogen Analyzer Heat Tracing Actuation Signal Reset
ML18153A1721997-10-30030 October 1997
LER 97-007-00:on 970930,determined That Plant Was Outside App R Design Basis Due to Vital Bus Isolation Issue.Caused by Personnel Error.Installed Circuit Protective Device During Oct 1997 Refueling Outage
ML18153A1421997-06-10010 June 1997
LER 97-001-01:on 970123,shutdown Occurred Due to Drain Line Weld Leak.Inspected & Tested Turbine Trip Actuation circuitry.W/970610 Ltr
ML18153A1391997-05-28028 May 1997
LER 97-005-00:on 970502,Unit 1 Power Range Nuclear Instrumentation Was Inoperable Due to Personnel Error.Sro & STA That Were Involved in Event Were Counseled
ML18153A1291997-04-18018 April 1997
LER 97-006-00:on 970320,loss of Refueling Integrity Due to Inadequate Containment Closure Process & Verification.Fuel Movement Stopped IAW Action Statement Requirements of TS 3.10.B.W/970418 Ltr
ML18153A1281997-04-15015 April 1997
LER 97-004-00:on 970317,main Steam Safety Valve Was Outside as Found Setpoint Tolerance.Specific Cause Unknown,However, Minor Setpoint Drift Can Be Expected.No Immediate Corrective Actions performed.W/970415 Ltr
ML18153A1241997-04-0808 April 1997
LER 97-002-01:on 970116,one Train of Auxiliary Ventilation Sys Was Inoperable Outside of Ts.Caused by Personnel Error. Submitted Deviation Rept Re Reverse Rotation of Fan & Work Request to Adjust linkage.W/970408 Ltr
ML18153A1191997-03-19019 March 1997
LER 97-001-00:on 970218,manual Reactor Trip & ESF Actuation Occurred Due to Loss of EHC Control Power.Caused by Momentary Short.Relay Card Was replaced.W/970319 Ltr
ML18153A1201997-03-19019 March 1997
LER 97-003-00:on 970219,loss of Pressurizer Heaters Resulted in Manual U1 Trip & U2 ESF Actuation.Caused by Loss of Group C Pressurizer Proportional Heaters.Reactor Trip Breakers Were Verified open.W/970319 Ltr
ML18153A1131997-02-20020 February 1997
LER 97-001-00:on 970123,shutdown Occurred Due to Steam Drain Line Weld Leak.Management Was Notified & Shift Supervisor Invoked Requirements of TS 4.15.C.1.W/undtd Ltr
ML18153A1101997-02-13013 February 1997
LER 97-002-00:on 970116,one Train of Auxiliary Ventilation Sys Declared Inoperable.Caused by Personnel Error.Properly Adjusted Damper 1-VS-MOD-58B & Exited Seven Day LCO on 970116.W/970214 Ltr
ML18153A0951997-01-0202 January 1997
LER 97-002-00:on 961213,automatic Reactor Trip Occurred During Planned Shutdown.Caused by Steam Flow/Feedwater Flow Mismatch.Rps Functioned as Designed & Plant Placed in Hot Shutdown
ML18153A0931996-12-12012 December 1996
LER 96-008-00:on 961112,water Gas Decay Tank Oxygen Analyzer Pressure Sensors Inoperable Due to Vendor Supplied Equipment Not Meeting Procurement specifications.Post-implementation Procedures Revised & Transducers replaced.W/961212 Ltr
ML18153A0691996-09-19019 September 1996
LER 96-007-00:on 960821,failed to Complete Fire Detection Zone Inspections within Required Time Period.Caused by Personnel Error.Counseled Personnel Re Fire Detection Zone Inspections & Revised Fire Watch training.W/960920 Ltr
ML18153A0481996-08-26026 August 1996
LER 96-005-00:on 960803,manual Reactor Trip.Caused by Loss of Electro Hydraulic Control Pressure.Repaired Two Compression Fitting Union Connections on Leaking Fitting & Performed Evaluations on Other tubing.W/960826 Ltr
ML18153A0521996-08-20020 August 1996
LER 96-004-01:on 960510,discovered Hydrogen Analyzers Inoperable.Caused by Procedural Deficiencies.Implemented Permanent Changes to Hydrogen Analyzer Instrument Calibr Procedures.W/960820 Ltr
ML18153A0321996-07-30030 July 1996
LER 96-006-01:on 960618,anti-corrosion Coating Had Not Been Reapplied to Station Battery 2B.Caused by Procedural Error in That Verbatim TS Compliance Not Reflected in Procedures. Coating Was Applied to batteries.W/960730 Ltr
ML18153A0281996-07-17017 July 1996
LER 96-006-00:on 960618,failed to Apply anti-corrosion Coating to Station Battery 2B.Caused by Procedural Error. Applied anti-corrosion Coating to Batteries & Revised TS 4.6.C.1.f Re Battery Coating requirements.W/960717 Ltr
ML18153A0141996-07-0202 July 1996
LER 96-004-00:on 960606,turbine/reactor Trip Occurred.Caused by High Level in Steam Generator B.Placed Plant in Hot Shutdown Condition,Calculated Shutdown Margin & Monitored Critical Safety Function Status trees.W/960702 Ltr
1999-08-27
[Table view]
Category:TEXT-SAFETY REPORT
MONTHYEARML18152A2811999-10-12012 October 1999
Technical Basis for Elimination of Nozzle Inner Radius Insps (for Nozzles Other than Reactor Vessel),Technical Basis for ASME Section XI Code Case N-619.
ML18152B3531999-09-30030 September 1999
Monthly Operating Repts for Sept 1999 for Surry Power Station,Units 1 & 2.With 991012 Ltr
ML18152B6651999-08-31031 August 1999
Monthly Operating Repts for Aug 1999 for Surry Power Station Units 1 & 2.With 990915 Ltr
ML18152B4421999-08-27027 August 1999
LER 99-006-00:on 990802,determined That Plant Was Outside of App R Design Basis Due to Fire Barrier Deficiencies. Caused by Original Plant Design Deficiencies.Fire Watches Were Established & Mods Have Been Completed.With 990827 Ltr
ML18152B4411999-08-27027 August 1999
LER 99-005-00:on 990731,effluent Radiation Monitors Were Declared Inoperable.Caused by Degraded Heat Trace Circuits for Monitors Sample Suction Line.Degraded Heat Trace Circuit Was Replaced & Addl Heat Trace Is Being Installed
ML18151A3981999-08-13013 August 1999
SPS Unit 2 ISI Summary Rept for 1999 Refueling Outage.
ML18152B3771999-08-13013 August 1999
LER 99-004-00:on 990714,TS Violation Due to non-safety Related Fans Effect on CR Boundary Was Noted.Cause of Event Has Not Yet Been Determined.Cable Spreading Room Doors Were Operned to Reduce Pressure in Rooms
ML18152B3791999-07-31031 July 1999
Monthly Operating Repts for July 1999 for Surry Power Station,Units 1 & 2.With 990811 Ltr
ML18152B3911999-06-30030 June 1999
Monthly Operating Repts for June 1999 for Surry Power Station,Units 1 & 2.With 990713 Ltr
ML18152B4341999-05-31031 May 1999
Monthly Operating Repts for May 1999 for Surry Power Station,Units 1 & 2.With 990614 Ltr
ML20195E2401999-05-31031 May 1999
Rev 2 to COLR for SPS Unit 2 Cycle 16 Pattern Ag
ML18152B4181999-05-18018 May 1999
LER 99-002-00:on 990425,MSSVs Tested Out of Tolerance for as Found Setpoint.Caused by Minor Setpoint Drift.No Immediate Action Required.Deviation Rept Submitted for Each Valve.With 990518 Ltr
ML18152B4161999-04-30030 April 1999
Monthly Operating Repts for Apr 1999 for Surry Power Station Units 1 & 2.With 990512 Ltr
ML18152B4111999-04-28028 April 1999
LER 99-003-00:on 990331,potential Loss of Charging Pumps Was Noted.Caused by Main CR Fire.Station Deviation Was Issued on 990331.With 990428 Ltr
ML18152B6511999-03-31031 March 1999
Monthly Operating Repts for Mar 1999 for Surry Power Station Units 1 & 2
ML18153A2741999-03-29029 March 1999
LER 99-002-00:on 990301,prematurely Released Fire Watches Resulted in Violation of TS 3.21.B.7.Caused by Inadequate Procedure.Procedure for Opening & Sealing Fire Stops Was Revised on 990212
ML18153A2681999-03-19019 March 1999
LER 98-013-01:on 981122,turbine/reactor Tripped on High Due to Short Circuit in Summator for MSL C Loop Channel III Flow Transmitter.Replaced 1-MS-FT1494 Summator & Module Repair Procedure Revised.With 9903190 Ltr
ML18152B7331999-02-28028 February 1999
Monthly Operating Repts for Feb 1999 for Surry Power Station,Units 1 & 2.With 990310 Ltr
ML18152B5421999-01-31031 January 1999
Monthly Operating Repts for Jan 1999 for Surry Power Station,Units 1 & 2.With 990210 Ltr
ML18151A3031999-01-29029 January 1999
ISI Summary Rept for 1998 Refueling Outage,Including Form NIS-1, Owners Rept for ISIs & Form NIS-2, Owners Rept for Repairs & Replacements.
ML18152B7261999-01-21021 January 1999
LER 99-001-00:on 981222,auxiliary Feedwater Pipe Support Missed Surveillance.Caused by Personnel Error.Station Deviation Rept Was Submitted.Two Supports in Question Received Required Code Insp & Were Found Acceptable
ML18152B6011998-12-31031 December 1998
Monthly Operating Repts for Dec 1998 for Surry Power Station,Units 1 & 2.With 990115 Ltr
ML18152B5781998-12-16016 December 1998
LER 98-014-00:on 981126,manual Reactor Trip in Response to Main Feedwater Regulating Valve Failure Occurred.Caused by Dislocation of Retaining Clip in Positioner.Control Room Operators Placed Unit in Safe,Shutdown Condition
ML18152B5811998-12-16016 December 1998
LER 98-013-00:on 981122,turbine/reactor Trip on High SG Level Occurred.Caused by Instrument Failure.Control Room Operators Placed Unit in Safe,Shutdown Condition
ML18152B7121998-12-0404 December 1998
LER 98-S01-00:on 981105,noted Failure to Deactivate Station Access Badge.Caused by Human Error.Licensee Will Now Deactivate Station Badges Before Clearance Is Revoked & Process for Badge Deactivations Have Been Strengthened
ML18152B7041998-12-0101 December 1998
LER 98-012-00:on 981102,noted That EDGs Were Concurrently Inoperable.Caused by Required Testing Per TS 3.16.B.1.a.2. Redundant EDG Was Returned to Svc within Two Hour Period, Following Satisfactory Testing.With 981201 Ltr
ML18152B7081998-11-30030 November 1998
Rev 0 to COLR for Surry 1 Cycle 16,Pattern Un.
ML18152B5721998-11-30030 November 1998
Monthly Operating Repts for Nov 1998 for Surry Power Station,Units 1 & 2.With 981214 Ltr
ML18152B6161998-11-0606 November 1998
LER 98-011-00:on 981008,diesel Driven Fire Pump Failed to Start During Performance of Monthly Operability Test.Caused by Faulty Overspeed Trip Device Failure.Diesel Driven Fire Pump Declared Inoperable
ML18152B6241998-10-31031 October 1998
Monthly Operating Repts for Oct 1998 for Surry Power Station Units 1 & 2.With 981111 Ltr
ML18152B6081998-10-23023 October 1998
LER 98-010-01:on 980715,intake Canal Level Probes Were Inoperable Due to Marine Growth.Caused by Design of Canal Level Instrumentation.Canal Level Probes Will Continue to Be Monitored More Closely
ML18152B6881998-09-30030 September 1998
Monthly Operating Repts for Sept 1998 for Surry Power Station Units 1 & 2.With 981012 Ltr
ML18153A3271998-08-31031 August 1998
Monthly Operating Repts for Aug 1998 for Surry Power Station,Units 1 & 2
ML18152B7811998-07-31031 July 1998
LER 98-010-00:on 980715,low Intake Canal Level Instrument Channel I Was Declared Inoperable to Allow Testing of Intake Canal Level Probe 1-CW-LE-102.Subject Probe Was Cleaned by Diver,Tested & Channel I Was Returned to Operable Status
ML18153A3161998-07-31031 July 1998
Monthly Operating Repts for July 1998 for Surry Power Station Units 1 & 2.W/980807 Ltr
ML18152B7621998-06-30030 June 1998
Monthly Operating Repts for June 1998 for Surry Power Station,Units 1 & 2.W/980707 Ltr
ML18153A2581998-06-0303 June 1998
LER 98-009-00:on 980509,nonisolable Leak of Reactor Coolant Pump Seal Injection Line Weld,Was Discovered.Caused by Lack of Fusion or Thermal Fatigue Coupled W/Vibration Stress Due to Loose Rod Hanger.Rcp Seal Injection Line Removed
ML20248F7441998-05-31031 May 1998
Reactor Vessel Working Group,Response to RAI Regarding Reactor Pressure Vessel Integrity
ML18153A3141998-05-31031 May 1998
Monthly Operating Repts for May 1998 for Surry Power Station,Units 1 & 2.W/980610
ML18152B8241998-05-22022 May 1998
LER 98-008-00:on 980228,auxiliary Ventilation Fans Were Noted in Condition Outside of Design Basis.Caused by Failure to Recognize Potential Impact of Certain Design Basis Accident Scenarios.No Corrective Actions Needed
ML18152B8161998-04-30030 April 1998
Monthly Operating Repts for Apr 1998 for Surry Power Station Units 1 & 2.W/980508 Ltr
ML18152B7951998-04-29029 April 1998
LER 98-007-00:on 980330,radiation Monitors Were Declared Inoperable.Caused by Change in Operating Temperature Range. Preplanned Alternate Method of Monitoring Was Initiated IAW TS Table 3.7-6
ML18153A2511998-04-22022 April 1998
LER 98-006-00:on 980324,unisolable Through Wall Leak of RCP Thermowell Was Noted.Cause of Leak Is Unknown.Rtd Will Be Replaced
ML18153A2521998-04-22022 April 1998
LER 98-005-01:on 980212,fire Watch Insp Exceeded One Hour. Caused by Lack of Attention to Detail by Individual Involved.Individual Involved Was Coached on Requirement to Perform Fire Watch Patrols within Required Time Frame
ML20217P9941998-04-0707 April 1998
Safety Evaluation Granting Licensee Third 10-yr Inservice Insp Program Relief Requests SR-018 - Sr-024
ML18153A2951998-03-31031 March 1998
Monthly Operating Repts for Mar 1998 for Sps,Units 1 & 2.W/ 980408 Ltr
ML18153A2391998-03-13013 March 1998
LER 98-005-00:on 980212,fire Watch Insp Frequency Exceeded One H Occurred.Category 2 Root Cause Evaluation Being Conducted to Determine Cause of Event.Station Deviation Issued
ML18153A2341998-03-0909 March 1998
LER 98-003-00:on 980226,no Procedural Guidance for Maintaining EDG Minimum Fuel Supply During Loop,Was Identified.Caused by Absence of Procedural Instructions. Deviation Rept Submitted to Document Deviating Condition
ML18153A2301998-03-0606 March 1998
LER 98-004-00:on 980206,fire Watch Was Released Prematurely Resulting in Violation of Ts.Caused by Inadequate Planning of Repair Activity.Work Orders Will Include Ref to Applicable Procedures Developed to Assist in Repairs
ML18153A2251998-03-0404 March 1998
LER 98-002-00:on 980202,automatic Turbine Trip Resulted in Automatic Reactor Trip.Caused Degraded Generator Voltage Regulator sub-component Failure.Placed Plant in Safe Hot SD & Replaced Intermittent Relay & Relay Socket
1999-09-30
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POW 28-06-01 U.S. NUCLEAR REGULATORY COMMISSION NIIC Form 391}
19-83) APPROVED OMII NO. 3150-0104 EXPIRES: 8/31185 LICENSEE EVENT REPORT (LER)
FACILITY NAME 11) IoDOCKET NUMBER 121 I 5 I o I o I o I 2 I R In I
1 OF PAG~ 131 nI -:i Surrv Power Station. Unit 1 TITLE 1'1 Reactor Trip on Low RCS. Flow Due to Failure of Loop Stop Valve.
EVENT DATE 16) LER NUMBER 161 REPORT DATE 171 OTHER FACILITIES INVOLVED (Ill DOCKET NUMBERISl MONTH* DAY YEAR YEAR /{ SEQUENTIAL NUMBER () REVSSION NUMBER MONTH DAY YEAR FACILITY NAMES 01s1010101 I I o Is , I 6 8 7 011
- ol ,1 , - o I o ol6 112 sl1 0151010101 I I THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Ch.c:k on, or mor, of th1 following) (111 OPERATING MODE (II)
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- II0.731oll211wlllllBI 60.73(oll211xl LICENSEE CONTACT FOR THIS LER (121 NAME TELEPHONE NUMBER AREA CODE R. F. Saunders, Station Manager 0 1 o, q 3 I 51 7 I - 13 I 1 I 8 14 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (131 REPORTABLE .;.;;.;. .*.**.;. ;,;.;.*.* ;.*..*.* MANUFAC-MANUFAC-
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SUPPLEMENTAL REPORT EXPECTED 114) MONTH DAY YEAR EXPECTED lxl YES (If yos, compl*r* EXPECTED SUBMISSION DATE/
ABSTRACT (Limit to 1400 SPIJCSs. i,tJ,, *PfJroxim,ttJly fifts6n singts-spacs typswrirrsn lim,s} (16) h NO SUBMISSION DATE 1151 018 310 817 On May 16, 1987 at 0824 hours0.00954 days <br />0.229 hours <br />0.00136 weeks <br />3.13532e-4 months <br />, with Unit 1 at 100% power, a low flow reactor trip occurred when "A"loop reactor coolant system (RCS) {EI IS-AB} flow decreased to 47%. Following the reactor trip, the source range channels {EIIS-DET} did not automatically reinstate. All other protection and control systems functioned properly. Operators followed appropriate plant procedures and stabilized the plant following the reactor trip.
This event occurred when the. II A" hot leg loop stop valve (EIIS-ISV} stem failed, permitting the disc to drop, partially blocking loop flow. A detailed metallurgical analysis is being performed to determine the failure mode and mechanism of the valve stem. The preliminary report indicates that failure was due to stress or fatigue. To reduce the stress on the valve stem, the operating procedure is being revised to normally operate the valves*off the backseat. The failure of source range channels to reinstate was due to the under compensation of the intermediate range channel NI-36. The source range channels were manually reinstated, and technicians readjusted the intermediate range compensating voltage.
8706230353 870612 Itl'-
PDR ADOCK 05000280 \ (\
s PDR NRC Form 386 19-83)
POW 28-06-01 NRC l'orm 366A
- U.S. NUCLEAR REGULATORY COMMISSION (9-83)
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED 0MB NO. 3150--0104 EXPIRES: 8/31/88 FACILITY NAME (1) DOCKET NUMBER (2) PAGE (3)
LER NUMBER (6)
YEAR :::::::::: SEQUENTIAL ::::::::::* REVISION
- NUMBER .'."."."." NUMBER Surry Power Station, Unit 1 o 1s Io Io Io I 2 I 81 O 81 7 - 011 I1 - 0 I O Ol 2 OF OI 3 TEXT '" more IP"C9 i& "'flUired, U&tl *dditiona/ NRC Form 36flA '&/ (17) 1.0 Description of the Event On May 16, 1987 at 0824 hours0.00954 days <br />0.229 hours <br />0.00136 weeks <br />3.13532e-4 months <br />, with Unit 1 at 100% power, a low flow reactor trip occurred when "A" loop reactor coolant system {EIIS-AB} (RCS) flow decreased to 47%. Following the reactor trip, the source range channels {EIIS-DET} did not automatically reinstate.
All other protection and control systems functioned properly. Operators followed appropriate plant procedures and stabilized the plant following the reactor trip.
2.0 Safety Consequences and Implications The low flow reactor trip automatically trips the reactor to maintain sufficient margin above a DNBR of 1.3 with a loss of RCS flow. The complete loss of flow in one loop from a reactor power of 100% (2441 MWt) with three loops operating is an analyzed event.
During the event, "A" loop flow was maintained at approximately 47% and total core flow remained at 84%.
A confirmatory analysis performed by Nuclear Engineering concluded that DNBR was maintained above the accident analysis value of 1.3. In addition, all other safety related systems remained operable during the event, and plant parameters remained well within the bounds of the accident analysis. Therefore this event did not constitute *an unreviewed safety ques*tion and the health and safety of the public were not affected.
3.0 Cause The cause of this event was a failure of the "A" RCS hot leg loop stop valve (MOV-1590) {EIIS-ISV}. The valve stern failed, permitting the disc to drop, partially blocking the loop flow.
The valve stern is being analyzed for failure mode and mechanism. A supplemental LER will be submitted when this analysis is completed. The preliminary report indicates that failure was due to fatigue or stress.
The failure of the source range channels to reinstate was due to the under compensation of the intermediate range channel NI-36.
NAC FOAM 366A (9-83)
POW 28-06-01 NRC Form 366A. U.S. NUCLEAR REGULATORY COMMISSION (9-831 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED 0MB NO. 3150-0104 EXPIRES: 8/31 /88 FACILITY NAME (1) Dq<:KET NUMBER (2) PAGE (3)
LER NUMBER (6)
YEAR :::::::::: SEQUENTIAL ::::::::::: REVISION
- NUMBER :::::::::: NUMBER Surry Power Station, Unit 1 0 1s101010121a10 817 -01111-olo oh °Fo 1~
TEXT /ff more /J/NJCl1 is n,quired, U/ltl 11dditionsl NRC Form 3lil!iA '*) (17) 4.0 Inunediate Corrective Action The Operators performed all appropriate emergency procedures and function restoration procedures to ensure the plant was returned to a stable condition. This included manually reinstating the source range channels.
Also, the STA performed the critical safety function status tree review to ensure specific plant parameters were noted and that those parameters remained within safe bounds.
5.0 Additional Corrective Actions The unit was placed in the cold shutdown condition and the stem of the hot leg loop stop valve was replaced. The stems of the other five Unit 1 loop stop
- valves were ultrasonically tested and found to be satisfactory.
Technicians readjusted the intermediate range compensating voltage.
- 6. O Actions Ta-ken to Prevent Recurrence A detailed metallurgical analysis is being performed to determine exact cause of the valve stem failure.
To reduce the stress on the valve stem, the operating procedure is being revised to normally operate the valves off the backseat.
7.0 Similar Events A similar failure occurred on the Unit 1 "B" loop Hot Leg Isolation Valve on December 1, 1973.
8.0 Manufacturer/Model Number Anchor Darling/Drawing nos. 95-11778 and 95-11779.
NRC FORM 366A (9-83)
- e VIRGINIA ELECTRIC AND POWER COMPANY Surry Power Station P. 0. Box 315 Surry, Virginia 23883 June 12, 1987 U.S. Nuclear Regulatory Commission Serial No.: 87-013 Document Control Desk Docket No.: 50-280 Oi6 Phillips Building Licensee No.: DPR-32 Washington, D.C. 20555 Gentlemen:
Pursuant to Surry Power Station Technical Specifications, Virginia Electric and Power Company hereby submits the following Licensee Event Report for Surry Unit 1.
REPORT NUMBER 87-011-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be reviewed by Safety Evaluation and Control.
Very truly yours, R. F. Saunders Station Manager Enclosure cc: Dr. J. Nelson Grace Regional Administrator Suite 2900 101 Marietta Street, NW Atlanta, Georgia 30323
_._J
