Evaluation of Potential Severe Accidents During Low Power And

NUREG/CR-6144 BNL-NUREG-52399 Wol. 2, Part 2

Evaluation of Potential Severe Accidents During Low Power and Shutdown Operations at Surry, Unit 1 : Analysis of Core Damage Frequency from Internal Events During Mid-Loop Operations

Appendices A- D

Manuscript Completed: January 1994 Date Published: June 1994

Prepared by T. L. Chu, Z. Musieki, P. Kohut, D. Bley1,J. Yang, B. Holmes a, G. Bozoki, C. J. Hsu, D. J. Diamond, D. Johnson1, J. Lin1, R. E SuS, V. Danga, D. nberg 4, S. M. Wong, N. Siu3

Brookhaven National Laboratory Upton, NY 11973

Prepared for Division of Safety Issue Resolution Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 NRC FIN L1922

1PLG Inc., 4590 MacArthur Boulevard, Newport Beach, CA 92660-2027 2AEA Technology, Winfrith, Dorchester, Dorset, England, DT2 8DH _:_,_ / _ '_"_. " 3M1T,Cambridge, MA 02139 (N. Siu currently at EG&G, Idaho Falls, ID 83415) _ 17!i_' r, _ ;_ 4Soreq Nuclear Research Center, Yavne 70600, Israel

DISTRIBUTION OF THIS DOCUMENT IS UNLIMITEi_ _(_ _j ABSTRACT

Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk.

During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. ","heprogram includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied.

The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA.

The objective of this volume of the report is to document the approach utilized in the level-1 internal events PRA for the Surryplant, and discuss the results obtained. A phased approach was used in the level-1 program. In phase 1, which was completed in Fall 1991, a coarse screening analysis examining accidents initiated by internal events (including internal fire and flood) was performed for all plant operational states (POSs). The objective of the phase 1 study was to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenarios, and to provide a foundation for a detailed phase 2 analysis.

In phase 2, mid-loop operation was _lected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective, :! the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The scope of the level-1 study includes plant damage state analysis, and uncertainty analysis. Volume 1 summarizes the results of the study. Internal events analysis is documented in Volume 2. It also contains an appendix that documents the part of the phase 1 study that has to do with POSs other than mid- loop operation. Internal fire and internal flood analyses are documented in Volumes 3 and 4. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associates, Inc. Volume 6 documents the accident progTession,source terms, and consequences analysis.

In the phase 2 study, system models applicable for shutdown conditions were developed and supporting thermal hydraulic analysis were performed to determine both the timing of the accidents and success criteria for systems. Initiating events that may occur during mid-loop operations were identified and accident sequence event trees were developed and quantified. In the preliminary quantification of the mid- loop accident sequences, it was found that the decay heat at which the accident initiating event occurs is an important parameter that determines both the success criteria for the mitigating functions and the time available for operator actions. In order to better account for the decay heat, a "time window" approach was

iii NUREG-CR/6144 developecL In this approach, time windows after shutdown were defined based on the success criteria established for the various methods that can be used to mitigate the accident. Within each time window, the decay heat and accident sequence timing are more accurately defined and new event trees developed and quantified accordingly. Statistical analysis of the past outage data was performed to determine the time at which a mid-loop condition is reached, and the duration of the mid-loop operation. Past outage data were used to determine the probabilitythat an accident initiating event occurs in each of the time windows. This probability is used in the quantification of the accident sequences.

The mean core damage frequency of the Surryplant due to internalevents that may take place during mid-loop operations is 5E-06 per year, and the 5th and 95th percentiles are 5E-07 and 2E-05 per year, respectively. This can be compared with the mean core damage frequency from internal events of 4E-05 per year estimated in the NUREG-1150 study for full power operations.

NUREG/CR-6144 iv Contents

Appendix A Data Base of Plant Experience Identified For Initiating Event Quantification

Appendix B Review of U.S. Nuclear Regulatory Commission (NRC) Information Notices, Generic Letters, Bulletins and Circulars

Appendix C System Fault Trees

Appendix D Statistical Analysis of Time to Mid-Loop and Duration of Mid-Loop

v NUREG/CR-6144 FOREWORD

(NUREG/CR-6143 and 6144) Low Power and Shutdown Probabilistic Risk Assessment Program

Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only duringfull power operation. Some previous screening analysis that were performed for otL r modes of operation suggested that risks duringthose modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contn'butors to risk.

During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects performed by Brookhaven National Laboratory(BNL) and Sandia National Laboratories(SNL), with the seimnic analysisperformed by Future Resources Associates. Two plants, Surry(pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied.

The objectives of the program are to assess the risks of severe accidents due to internal events, internal fires, internal floods, and seismic events initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated duringfull power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PR_

The results of the program are documented in two reports, NUREG/CR-6143 and 6144. The reports are organizeasdfollows:

For Grand GulP.

NUREG/CR-6143 - Evaluation of Potential Severe Accidents during Low Power and Shutdown Operations at Grand Gulf

Volume 1: Summaryof Results Volume 2: Analysis of Core Damage Frequency from Internal Events for Operational State 5 During a Refueling Outage Part 1: Main Report Part 1_ Sections 1 - 9 Part 1B: Section 10 Part 1C: Sections 11 - 14 Part 2: Internal Events Appendices A to H Part 3: Internal Events Appendices I and J Part 4: Internal Events Appendices K to M Volume 3: Analysis of Core Damage Frequency from Internal Fire Events for Plant Operational State 5 During a Refueling Outage Volume 4: Analysis of Core Damage Frequency from Internal Flooding Events for Plant Operational State 5 During a Refueling Outage Vo_lme 5: Analysis of Core Damage Frequency from Seismic Events for Plant Operational State 5 During a Refueling Outage

vii NUREG/CR-6144 Foreword (continued)

Volume 6: Evaluation of Severe Accident Risks for Plant Operational State 5 During a Refueling Outage Part 1: Main Report Part 2: Supporting MELCOR Calculations

For Surry:

NUREG/CR-6144- Evaluation of Potential Severe Accidents during Low Power and Shutdown Operations at Surry Unit-1

Volume 1: Summaryof Results

Volume 2: Analysis of Core Damage Frequency from Internal Events during Mid-loop Operations Part 1: Main Report Part 1A: Chapters I - 6 Part 1B: Chapters 7 - 12 Part 2: Internal Events Appendices A to D Par:'3: Internal Events Appendix E Part 3A: Sections E.1 - E.8 Part 3B: Sections E.9 - E.16

I Part 4: Internal Events Appendices F to H Part 5: Internal Events Appendix I Volume 3: Analysis of Core Damage Frequency from Internal Fires during Mid-loop Operations Part 1: Main Report Part 2: Appendices Volume 4: Analysis of Core Damage Frequency from Internal Hoods during Mid-loop Operations Volume 5: Analysis of Core Damage Frequency from Seismic Events during Mid-loop Operations Volume 6: Evaluation of Severe Accident Risks during Miid-loop Operations Part 1: Main Report Part 2: Appendices

NUREG/CR-6144 viii APPENDIX A

DATA BASE OF PLANT EXPERIENCE IDENTIFIED

FOR INITIATING EVENT

QUANTIFICATION

A-! Appendix A

Data Base of Plant Experience Identified for Initiating Event Quantification

A.1 One Line Summary of All Events in the Data Base ...... A-3

A.2 Descriptions of Events Used in the Quantification ...... A-18

A.2.1 Loss of RHR Events (RHR2A, 2B, 3, 4, 5, 6) ...... A-19

A.2.2 Loss of 4 kV Bus (4 kV, 4 kV/P6) ...... A-45

_2.3 Loss of Vital Bus (VITAL) ...... A-53

A.2.4 Loss of Component Cooling Water (CCW) ...... A-60

A.2.5 Inadvertent Safety Feature Actuation (ESFAS, ESFAS/SI) ...... A-62

A.2.6 Loss of Instrument Air ...... A-69

A.2.7 LOCAs (HCVCS, HRHR, JCVCS, JRHR, KRCS) ...... A-75

A.2.8 Transients ...... A-80

A-2 Appendix A.1 One Line Summary of All Events in the Data Base

A-3 Page No. 1 08118/93 PHASE2 - LOSS OF RESIDUALHEATREHOVALWHILE AT SHtiTDOtaiFOR PtmS

INITIAL SEA- PHASE2 SSS EVENT PLANT RECOVERY NSACS2 6ROOK CR5015 G199 LER NUREGAPPLI- CAT CAT PLANTNAME DOCKETNO DATE LER CONOITION TIME CAT. CAT. AEOO CAT CAT SEARCH1410 CADIL.

...... 0...... 0...... o..o....0...... -.. 0.....-- .---..-.---0------0------..------:------

1 Byron 1 5000454 07124185 85-070-1 Hode 1 A-IO HA2 2 Davis Besse 1 5000346 06105178 78-063 Mode 5 A.7 E27 HA2 3 Maine Yankee 5000300 09129182 82-032 Hode 6 A-IO HA2 4 McGuire 1 5000369 02107/81 81-010 Hods 6 A.5 Ell NA1,2 5 PaLisades 5000255 06122184 84-007 Hode 5 A-1 X_.2 6 Rancho Seco 5000312 05/26/83 83-023 Rode 6 A-10 HA2 7 Rancho Seco 5000312 07/25/89 009 Mode 5 1 hr 26 min 98 HA2 8 SaLem 1 5000272 05120189 019 zero po,er 58 min 33 3.51 HA2 9 Three NiLe IsLand 5000320 091081;'7' Mode 4 and 5 A.9 E40 NA1 10 Zion 1 5000295 09110182 82-028 "Node 1 Not actual A-4 HA2 fat Lure 11 Zion 1 5000295 11116182 82-042 Hode 1 Hot actuat A-4 NA2 fai Lure 12 Zion 1 5000295 06109183 83-018 HOt or cord Not 8ctuat A-4 HA2 shut. fai Lure 13 3b 3b Braidwood 2 5000457 03/18/90 002 zero power 10 fain 231 14 3b 3b Csttaway 1 5000483 07117184 84-016 Mode 5 A-3 O.2 15 3b 31) Ginne 5000244 0310718/, 84-003 Hods 5 k-3 6.1 Nk3 16 3b 3b Maine Yankee 5000309 03124182 82-013 Mode 5 A-3 17 3b 3b McGulre 2 5000370 09/05189 010 Mode 5 2 hr 51 sin 179 18 3b 3b Rancho Seco 5000312 04119181 81-024 Node 5 A.3 Do6 19 3b 31) Sequoyah 1 5000327 C2101187 0!3 Node 5 124 20 3b 3b St. Lucia 1 5000335 11/03/78 78-041 Mode 5 3.7 hr A.3 Pbl 21 4k'V 6b Arkansas NucLear 1 5000313 12/05/89 040, 1 of 2 CSO 103 events 22 4KV 6b Al*keneaa NucLear 1 5000313 12/06189 040, 2 of 2 CaD 9 mtn 103 : events 23 4KV 6b Arkansas Nuctear 2 5000368 11/14/89 022 Shutclown 1 rain 171 24 41iV 4 Beaver VaLLey 1 5000334 07103179 79-018 Mode5 or 6 21 sin A.6 E20 25 4KV 5a Beaver VaLLey 1 5000334 06129183 83-020 Mode 6 92 sec A-6 AEOD-A EIO NAI 26 4KV 5b Cstaktm 1 5000413 01/07/89 001 Node 5 193 27 4KV 68 Crystm[ River 3 ' 5000302 08128/89 031 Mode 5 19 mln 77 28 4KV 61) Davis geese 1 5000346 05128178 78-060 Mode 6 1.5 rain A.6 AEOD-C . E15 Page No. 2 08/18/93 PHASE2 - LOSS OF RESIDUALHEATREMOVALNHILE AT SHUTDOWFORN PNRS

INIT|AL SEA- PHASE2 ass EVENT PL_T RECOVERY NSAC52 BROOK CR5015 G199 LER NLIREGAPPLI- CAT CAT PLANTNAME DOCKETNO DATE LER CONDITION TiME CAT. CAT. AEO0 ICAT (:AT SEARCH1410 CABiL.

...... _ ....G...,...... -.-...... --.--...... o.o...... 29 4KV 61) Davis geese 1 5000346 06115178 78-067 Cord shut, 2 mtn (toter of A.6 AEO0-C E16 3 events) 30 4KV 6b Ft. Cethoun 5000285 12/14/85 85-011 Refueting 15 mln E.5 shut. 31 4KV 6b McGutre 1 5000369 1 m/29/88 038 zero poser 174 32 4KV 6b Millstone 2 5000336 12109181 81-043 Mode 5 30-60 min A.6 EE5 33 4KV . 6b North Anna 1 5000338 03123189 006 Mode5 10 aec 150 34 4KV 61) North Anna 1 5000338 04116/89 010 Mode 5 151 35 410/ 6b North Anna 2 5000338 04/16/89 010, Inferred Mode5 151 from LER 36 4KV 1 Rancho Seco 5000312 12/08/86 86-030 "Mode 5 A-1 A.4 37 4KV 2b Sstem 1 5000272. 04124179 79-059 Mode6 2 .tin (second A.6 E18 : event on 05108179) 38 410/ 2b astern 1 5000272 05108179 79-059 Mode6 4 rain (1st A.6 E18 event on 04124179) 39 4KV 10 Setem 1 5000272 03116/82 82-015 45 rain AEOO-A E2 t,O 4KV 1 astern 2 5000311 12/20/83 83-066 Mode5 22 m|n A-1 AEOO-A A7 41 4KV 6b Surry I 5000280 05124/86 86-017 Mode6 A-6 E,3 42 4KV 6b Three Mite Intend 5000289 01109187 001 Refuettng 64 shutdown 43 41(V 6b Uotf Creek 10/16/87 17 min 1.34 44 4KVIP6 4 Cetvert Ctiffs 1 5000317 11/12/80 80-058 Mode 5 or 6 10 mln A.4 E6 45 4KVIP6 1 Olsbto Canyon 1 5000275 01125/85 85-006 Mode 5 2 mln A-1 A.6 46 4KV/P6 61) Fa'rtey 1 5000348 01116/81 81-001 Shutdown 1 mtn A.6 E23 47 4KVIP6 Ob North Anna 2 5000339 04108/83 83-031 Mode 5 A-6 48 4KV/P6 6b Salem 1 5000272 01104183 83-001 Modes 4 and A-6 A12 5 49 4KV/P6 61) Salem 2 5000311 04113/83 83-014, 1 of 2 Mode 5 < 1 rain A-6 events 50 4KV/P6 61) Salem 2 ' 5000311 04118183 83-014, 2 of 2 Mode 5 < 1 rain A-6 events Page He. 3 08118/93 PHASE2 - LOSSOF RESIDUALHEAT REMOVALWHILE AT SHUTDOWHFORI_RS

!.mAL SEA- PHASE2 SSS EVENT PLAN1 RECOVERY NSAC52 BROOK CRS015 G199 LER NUREGAPPLI- CAT CAT PLANTNAME DOCKETNO DATE LER CONDITION TIME CAT. CAT. AEOD CAT CAT SEARCH1410 CABIL.

51 4KV/P6 6b burry 1 5000280 02/04189 005, Unit 2 •Mode 5" 3 mtn 44 event also 52 4KVIP6 6b Surry 1 5000280 04/06/89 010, Unit 2 Mode 5 1 min 47 event also 53 4KV/P6 61) $urry 2 5000280 02104189 005 (descr in Mode5 44 Surry 1) 54 4KV/P6 6b Surry 2 5000280 04106/89 010 (descr in Mode5 1 min 47 Surry 1) 55 4KV/P6 6b Zton2 5000304 01117186 86-005-1 Mo¢_ 5 A-6 E.9 56 AIR Beaver VaLLey 1 5000334 08129185 85-015 Mode 1 10 mtn 12 57 AIR Catlaway 5000483 11105184 84-059 Mode 1 20 min 35 58 AIR CaLvert CLiffs 1 5000317 01/27/87 87-003 Mode 1 20 mtn 2 59 AIR Cata_foa 1 5000413 01/14/85 85-004 Mode 2 15 min 51 60 AIR Cook 1 5000315 11125185 NPE Mode 1 20 mtn 2969 1_61 AIR Davis Besse 1 5000346 12107/87 87-015 Hode 1 31 m'n 62 62 AIR McGulre 1 5000369 11/02/85 85-034 Bode 1 20 min 179 63 AIR MiLLstone 3 5000423 04112187 87-020 Mode 1 20 m|n 1 64 AIR 10 North Anna 2 5000339 04/03/89 89-007 Node6 22 mtn 153 HA 65 AIR Oconee3 5000287 08114184 84-005 Mode I 20 mtn 1 66AIR Prairie Island 1 5000282 05108185 85-009 Mode 1 20 mtn 246 67 AIR Shearon Harris 1 5000400 08104187 87-041 Mode 1 20 min 245 68AIR burry 1 5000280 01/07/86 86-001 Mode 1 10 min 287 69AIR Yankee Rowe 5000029 10104186 86-012 Node 1 10 min 305 70 CCW 10 Bratc_ 1 5000456 01/21/87 011 zero power 14 mtn 226 71CCg 10 Byron 1 5000454 04108/87 012 zero power 17 min 222 72CCW 10 Matne Ysnkee 5000309 06/02/81 81-007 Mode6 30 mtn A.IO E47 73CCW 10 SaLem2 5000311 06123183 83-032 AEOO-A E5 HA 74 CCW 10 Turkey Point 3 5000250 10107183 83-018 AEOD-A E1 75 OIL TRAHSSurry 2 281 10/25/89 015 76ESFAS 61) Arkansas Nuclear 2 5000368 04/23/88 007 Node5 _ min 168" _A3 77 ESFAS 4 CaLvert CLiffs 1 5000317 05107179 79-015 Shutc_ 1_ _n A.4 E5 HA3 70ESFAS 4 CaLvert CLiffs 1 5000317 11/03/80 80-062 - 2 of 3 Mode6 A.4 E7 HA3 events Page No. 4 08/18/93 PHASE2 - LOSSOF RESIDUALHEATREHOVALWHILE AT SHUTDO_ FOR PtJRS

INITIAL SEA- PHASE2 SSS EVENT PLANT RECOVERY NSAC52 BROOK CR5015 G199 LEA: HUREGAPPLI- CAT CAT PLANTNAHE DOCKETNO DATE LER CONDITION TIME CAT. CAT. AEOD CAT CAT SEARCH1410 CABIL.

79 ESFAS 4 CaLvert CLiffs 1 5000317 11103180 80-062 - 1 of 3 Mode6 A.4 E7 HA3 events 80 ESFAS 4 CaLvert CLiffs 1 5000317 11104180 80-062 - 3 of 3 Mode 6 A.4 E7 HA3 + events 81 ESFAS 4 CaLvert CLiffs 2 5000318 09124178 78-033 Hode 5 or 6 23 m_n A.4 E4 HA3 82 ESFAS 4 CaLvert Ctiffs 2 5000318 01107/83 83-005 9 min AEO0-A E8 HA3 83 ESFAS 6b Cook 1 5000315 09/07/85 85-046 Rode 5 2 rain A-1 E.IO 84 ESFAS 1 Davis Besse 5000346 05114177 77-006 AEO0-C 85 ESFA$ 1 Davis 8esse 5000346 05119177 77-007 AEOD-C 86 ESFAS 1 Davis 8esse 5000346 05127/77 77-002 AEOD-C 87 ESFAS 1 Davis Besse 5000346 05128177 77-003 AEO0-C 88 ESFAS 1 Davis Besse 5000346 06112177 77-005 . AEO0-C 89 ESFAS 1 Davis 8esse 1 5000346 04119180 80-029 Mode 6 2 hr 30 min A.1, A.4 AEO0-C All, E1 90 ESFAS 4 Davis gesse 1 5000346 06/14180 80-049 Hode 6 2 mtn A.4 AEO0-C E2 91 ESFAS 1 Davts Besse 1 5000346 07/24/80 80-058, 1 of 2 Mode 5 50 m|n A.1 AEO0-C 414 events 92 ESFAS 1 Davis Besse 1 5000346 07124180 80-058, 2 of 2 Mode 5 2 rain A.1 AEO0-C A15 events 93 ESFAS 1 Davis gesse 1 5000346 08/01/80 80-058, 8/3/80 Mode 5 3 mtn A.1 A16 in clescrtp 94 ESFAS 1 Davis Besse 1 5000346 08/13/80 80-060 Mode 5 5 rain A.1 AEO0-C 417 95 ESFAS 1 Dlabto Canyon 1 5000275 09108186 86-012 Mode S 2 rain A-1 :A.7 96 ESFAS 1 FarLey 1 5000348 09118178 78-069 Node 5 7 rain A.1 A6 97 ESFA$ 1 North Arms 1 5000338 11106179 79-145 Mode 6 5 rain A.1 A9 98 ESFAS 1 Salem 1 5000272 09102176 76-004 Mode 6 19 rain A.1 A1 99 ESFAS 61) SaLem 2 5000311 06/23183 83-031, 2 of 2 Mode 5 A-6 AEO0-A E4 events 100 ESFAS 61) SaLem 2 5000311 06123183 83-031, 1 of 2 Mode 5 A-6 AEOD-A E4 events 101 ESFAS 1 Sequoyah 1 5000327 09116182 82-116 Mode 5 6 rain A-1 AEOD-A A13 102 ESFAS 1 Bummer 5000395 10102184 84-044 Mode 5 immediately A-1 A.18 . 103 ESFAS LTOP Burry 1 280 11113184 023 RefueLing SD HA Page No. S 08118/93 PHASE2 - LOSSOF RESIDUALHEAT REMOVALNHILE AT SHUTDOWFORN P_RS

INITIAL SEA- r. PHASE2 SSS EVENT PLANT RECOVERY NSAC52 BROOK CR5015 G199 LER HUREGAPPLI- CAT CAT PLANTNAME DOCKETNO DATE LER CONO|TIOlt TIME CAT. CAT. AEOD CAT CAT SEARCH1/,10 CA81L.

104 ESFAS 4 Surry 2 281 08/18/89 004 0_. Unit 2 384" " csb 105 ESFAS 1 Trojan 50003/,4 03/20/78 78-010 H(xle 5 A.1 A4 106 ESFAS/SI LTOP Starry 1 280 04/26/80 024 CSD Maintenance " 107 ESFAS/S| TITANSSurry 1 280 03/01/84 005; toss of 2 C_D v|tat bus' 108 ESFAS/S! LTOP burry 1 280 11/16/84 024 Refuet ing SD 109ESFAS/S! LTOP Surry 1 280 05/11/86 014 Shutclmm 110 ESFAS/SI LTOP Surry 1 280 06/05/89 022 CSO 111 ESFASISI LTOP Surry 2 281 06/12/85 007 Refueling SD 112 ESFAS/SI LTOP SUrly 2 281 11/16/86 016 CSO 113 ESFAS/S| 4 burry 2 281 09/08/89 005 0_; Unit 2 38S* CSD 114 HCVCS HCVCSTurkey Point 3 250 01/15/88 002 being cooted doun 115 HRHR HRHR Cetswbe 1 413 06/11/90 013 CSD (Node 5) 116 HRHR HRHR Sequoyah 1 327 02/11/81 021 A.3 Db2 117 HRHR HRHR Sequoyeh 2 328 08/06/81 094 A.3,A.4 E3 NAi 118 JCVCS ,ICVCSRebinson 2 261 01129181 005 119 JCVCS JCVCSt/eterford 3 382 12/16/85 057 pint hestup 120 JCVCS JCl/CSYankee Rowe 029 06/27/86 010 Mode 5. D.1 Melnt Out 121 JRHR JRHR Brstd_ 1 456 03/25/88 008 0"/,power tevet 122 JRHR JRHR 8rsldumod 1 456 12101189 016 OX pouer tevet 123 JRHR JRHR McGuire 2 370 08105/84 017 OZ power tevet 124 JRHR JRHR Summer1 395 05106185 014 CSD (Mode 5) A.20 125 KRCS KRCS North Anna 1 338 06/17/87 014 OR power, refuel. 126 KRCS ICRCS Trojan 344 07/03/81 013 Page No. 6 08/18/93 PHASE2 - LOSSOF RESIDUALHEATREMOVALWHILEAT SHUTDOWFNORPWRS

IN|TIAL SEA- PHASE2 SSS EVENT PLANT RECOVERY HSAC52 BRO01C CR5015 G199 LER HUREGAPPLi- CAT CAT PLANTNAME DOCKETNO DATE LER CONDITION TIME CAT. CAT. AEOD CAT CAT SEARCH1410 CABIL,

127 LOOP Connecticut Yankee 5000213 08124184 84-014 OXpower 20 min E.1 LOOP 128 LOOP Ft. CaLhoun 1 5000285 03/21/87 008 RefueLing 40 min 53 Loop cond. : 129 LOOP indian Point 3 5000286 11116184 84-015 Mode 5 E;6 LOOP 130 LOOP HcGuire 1 5000369 09/16/87 021 zero power 29 mtn 173 LOOP 131 LOOP 6b Oconee 3 5000287 09/11/88 005 Mode 5 15 mfn 62 1.35 132 LOOP 6b SaLem1 5000272 06/02/84 84-013, 1 of 2 Mode6 30 sec A-6 E.2 events 133 LOOP 61) Salem2 5000272 06/02/84 84-013, 2 of 2 Mode5 30 sac A-6 E.2 events : 134 LOOP San Onofre 1 5000206 04/22/80 80-015 Mode S 4 min A.6 E21 LOOP 135 LOOP 2b Vogtte 1 5000424 03/20/90 006 RefueLing 36 mtn 207 outage 136 RHR1 1 Beaver VaLLey 1 5000334 05121/80 80-031 Mode 5 A.1 A12 137 RHR1 1 6raidMood2 5000457 02/23/89 001 zero power 43 min 230 138_HR1 1 CaLvert CLiffs 1 5000317 10/12/83 83-061 Hocle6 30 min A-1 AEOO-A A9 139RHRI 1 CaLvert CLiffs 1 10123183 40 mtn 3.23 140 RHR1 1 CaLvert CLiffs 1 5000317 03/22/86 86-002 Mode 5 1 min A.13 141RHR1 1 CaLvert Ctiffs 2 5000318 10/22179 79-038 Mode5 3 min A.1 A8 142 RHR1 1 Catauba 1 5000413 08115/86 86-044-1 Mode5 15 min A-1 A.21 143 RHR1 1 Cry_tat River 3 5000302 03/06/80 80-015 Mode 5 8 mtn A.I AIO 144 RHR1 1 Crystat River 3 5000302 10123/88 022 Mode5 74 145 RHR1 1 Davis Bease 5000346 07122/77 77-009 (t_o AEOO-C events) 146 RHR1 1 Davis Besse 5000346 08/08/80 80-058 3 min AEO0-C 147 RHR1 1 Davis Besse 1 5000346 05/28/80 80-043 Mode 6 2 min A.1 AEO0-C A!3 148 RHR1 1 Olabto Canyon 1 5000275 09/29/81 84-004 Modes 415/6 5 mtn A-1 149 RHR1 I Oisb[oCanyon 1 5000275 10127183 84-004 Nodes 41516 1 hr A-1 150 RHR1 1 D|abto Canyon 1 5000275 01120185 85-005 Mode 5 9 mtn A-1 A.5 151RHR1 t FarLey 1 5000348 09/18/78 78-069 Mode 5 3 mtn A.1 A6 152 kHR1 1 Farley 1 5000348 11/28/80 80-077 Mode5 4 mtn A.1 A18 153 RHR1 1 FarLey 1 5000348 12/25180 80-080 Mode 6 5 min A.I A19 154 RHR1 1 FarLey 1 5000348 05/06/85 85-008 Mode 5 47 mtn A-1 A.16 Page No. 7 08/18/93 PHASE2 - LOSSOF RESIDUALHEATREMOVALWHILE AT SHUTDOt/NFORPNRS

INITIAL SEA- PHASE2 SSS EVENT PLANT RECOVERY NSAC52 BROOK CR5015 G199 LER HUREGAPP'|- CAT CAT PLANTNAME DOCKETNO DATE LER CONOITIOH TIME CAT. CAT. AEO0 CAT CAT sEARCH1410 CABIL.

...... '...... '...... '...... 155 RHR1 1 Ft. Ca|houri 1 5000285 01/08/89 001 Refueting 5 mln 59 cond. 156 RHRI 1 Indian Point 3 5000286 09/30/76 76-3-36(A) Mode 5 8 mtn A.1 A3 157 RHR1 1 McGuIre 1 5000369 04127/81 81-072 Mode 5 15 mtn A.1 A21 HA1 158 RHR1 1 HcOutre 1 5000369 08/04/81 81-129 Mode 5 15 min A.1 A22 HA1 159 RHR1 1 McGuire 1 5000369 11/18/81 81-185 Mode5 22 min A.1 A23 HA1 160RHR1 1 HcGuire 2 5000370 01/15/84 84-002 Mode5 49 min A-1 AEOO-BA.17 161RHR1 1 North Anna 2 5000339 04/23/80 80-001 Mode6 A.1 Al1. HA1 162 RHRI 1 North Anna 2 5000339 04114/83 83-023 Mcgie6 < 1 min A-1 AEOO-A A17 163 RHR1 1 Rancho Seco 5000312 08/08/85 85-016 Hode 5 A.IO 164 RHR1 1 Rancho Seco 5000312 08/14/85 85-016 Hode 5 A.IO 165 RHR1 1 Rancho Seco 5000312 07115187 038 Modes 40 mtn 93 166 RHR1 1 Rancho Seco 5000312 02/28/89 002 Mode 5 10 mtn 97 167 RHR1 1 Salem 1 5000272 09120176 76-005 • Hode 5 30 mtn A.1 A2 168 RHR1 1 Salem2 5000311 05114/83 83-024, 2 of 2 AE(30-A A5 events

169 RHR1 1 Salem2 5000311 02109184 84-002 Mode 5 17 mtn A-1 AECO-8 A.9 . 170 RHR1 1 Seabrook 1 5000443 10/11/89 012 zero po_er 50 min 217 HAl 171RHR1 1 Sequeyah 1 5000327 05114/85 85-020 Mode5 16 mtn A-1 A.15 172RHRI 1 ShearonHarrls 1 5000400 10/15187 060 Ho_5 5 mtn, 15 mtn 185. 173 RHR1 1 Sheeron Harris 1 5080400 12/10/89 022 Hode 5 3 mtn 186. 174 RHR1 1 St. Lucle 1 5000335 03129/83 83-021 10 min A-1 AEOO-A A14 175 RHR1 1 Summer 5000395 09/16/82 82-002 A-1 176 RHR1 1 Summer 5000395 10/15/82 82-004 Mode 5 1 mtn A-1 177 RHR1 1 Turkey Point 3 5000250 10108183 83-019 Rixle 6 6 rain A-1 AEOO*A A2 178 RHR1 1 Turkey Point 3 5000250 10125/85 85-036 Mode 5 27 rain A-1 A.1 179 RHR1 1 Turkey Potnt 3 5000250 11121/88 029 Cold 4 mtn 16 shutdown 180 RHR1 t Turkey Point 4 _000251 11/28181 81-015, 1 of 2 Mode5 2 _tn A.1 A24 events 181 RHR1 1 Turkey Point 4 5000251 11129181 81-015_ 2 of 2 Mode 5 1 mtn A.1 A25 events 182 RHR1 1 Turkey Point 4 5000251 11130/84 84-027 Node 6 4 mtn A-1 A.2 Page No. 8 08118/93 PHASE2 - LOSSOF RESIDUALHEATREMOVALNHILE AT SHUTDOI_/FORPMRS

INITIAL SEA- PHASE2 ass EVENT PLANT RECOVERY NSAC52 BROOK CR5015 G199 LER NUREGkPI_LI- CAT CAT PLANTNAME DOCKETNO DATE LER CONDITION TIME CAT. CAT. AEOD CAT CAT SEARCH1410 CAOIL.

183 RHP.2A 2a Arkansas Nuctear 2 S000368 08/29/84 84-023 Mode 5 AEO0-B8.5 184 RHR2A 2a Catawba 1 5000413 04122185 85-028 Hode 5 12 mln A-2 B.8 185 RHR2A 2o Cook 2 06/16/88 22 rain 3.46 186 RHR2A 2a Davis Besse 1 5000346 04/18/80 80-030 Mode 5 29 rain A.3 AEO0-C . Do3 187 RHR2A 2a _|rme 5000244 04/12/83 83-015 Mode 6 12 min A-2 AEOO-A C1 188 RHR2A 2a McGutre 1 5000369 03/02/82 82-024 Mode 5 50 m|n A-2 AEOO-A 87 189 RHR2A 2a HcGu4re 1 5000369 04105183 83-017 Mode 6 A-2 AEOD-A 88 190 RHR2A 2a HcGu|re 2 5000370 12/31/83 83-092 Mode 5 43 mln A-2 AEOO-A 99 191 RHR2A 28 McGutre 2 5090370 01109184 84-001 Mode 5 62 rain A-2 AEO0-8 8.6 192 RNR2A 2a North Anna 2 5000339 05/20/82 82-026 8 rain A-2 AEOO-A B5 193 RHR2A 2a North Anna 2 5000339 05/20/82 82-026 26 m|n A-2 AEOO-A 84 194 RNR2A 2a North Anna 2 5000339 05/20/82 82-026 1 hr A-2 AEOO-A 83 195 RHR2A 2a North Anrm 2 5000339 05103/83 83-038 Node 6 A-2 AEOD-A C5 196 RHR2A 2._ North Anna 2 5000339 10/16/84 84-008-1 :Hode 5 2 hr A-2 AEOD-B8.3 197 RHR2A 2a Son Onofre 2 5000361 03/26/86 86-007 Mode 5 49 mln 6.4 198 RHR2A 2a Sequoyah 2 5000328 08106183 83-101 Hode 6 77 rain A-2 AEOO-A C3 Mode 10 199 RHR2A 2a Trojan 5000344 03/25/78 78-011, 1 of 2 Mode 5 10 m|n A.2 B1 events 200 RHR2A 2a Trojan 5000344 03125178 78-011, 2 of 2 Node 5 10 min A.2 82 events 201 RHR2A 2a Trojan 5000344 06126181 81-012 Mode 5 75 m_n A.2 B4 202 RHR2A 28 t/aterford 5000382 03/14/86 86-004 Mode 5 1 hr 19 mtn E.11 203 RHR2A 2e Naterford 5000382 07/14/86 86-015 Mode 5 3 hr 41 rain 8.7 204 RIIR2A 2a Uaterford 3 05/12188 3.45 205 RHR2A 2a Zion 1 5000295 09/14/84 84-031 Mode 5 45 mtn A-2 AEO0-8 8.2 206 RHR26 2b Beaver Vat Lay 1 5000334 09/04/78 78-049 1 hr A.2 C3 207 RHR28 21) Beaver Vattey 1 5000334 01117180 80-002 Mode 5 A.2 C5 208 RHR2B 2I) Beaver Vattey 1 5000334 04108180 80-022 Mode 5 35 m4n A.2 C6 209 RHR2B 21) Beaver Vattey 1 5000334 04111180 80-023 Mode 5 70 mfn A.2 C7 210 RHR2B 2b Beaver VatLey 1 5000334 03/05181 81-019 Mode 5 St, m|n A.2 C8 211 RHR2B 2b Byron 1 , 5000454 09/19/88 007 Mode 6 14 rain 22/, 212 RHR28 2b Cook 2 5000316 05121/84 84-014 Mode 5 25 mfn A-2 AEOD-BC.2 213 RHR28 2b Dial)to Canyon 2 5000323 04/10/87 005 Mode 5 1 hr 28 m|n 116 Pmge No. 9 08/18/93 PHASE2 - LOSSOF RESIDUALHEAT REMOVALWlIILE AT SHUTDOk0FORN PWRS

INITIAL SEA- PHASE2 SSS EVENT PLANT RECOVERY NSAC52 BROOK CILS01S G199 LER"_ NUREGAPPLI- CAT CAT PLANTNAME DOCKETNO DATE LER CONDITION TIME CAT. CAT. AEOD CAT CAT SEARCH1410 CABIL.

214 Rm128 2b Iqlttstone 2 5000336 03114179 79-008 Mode 5 A.2 C4 215 RIIN2B 21) North Anna 1 5000338 10/19/82 82-067 Mode 6 36 rain A-2 AEOO-A B1 216 RHR2B 21) North Anne 1 5000338 10/20/_2 82-067 Mode 6 33 rain A-2 AEOO-A B2 217 RllR28 21) North Anne 2 5000339 07/17/82 82-049 Mode 5 A-2 218 RIIR28 21) North Anna 2 5000339 07/30182 82-049 46 sin AEOD-A 06 219 RHR2B 21) Patludes 5000255 10/15/87 035 Cord 29 mtn 22 3.44 shutdown 220 RNP.211 21) Rlnghals 4 08123184 26 iin 3.30 NA1. 221 RHR_ 2b Sstem 1 5000272 06130179 79-059 Mode6 34 rain A.2 B3 222 RHR2B 21) Sequoyah 1 5000327 10109185 65-040 Mode 5 A-2 C.3 223 RNR2B 2b Sequoyah 1 5000327 01/28/87 012 Mode 5 30 mtn 123 3.40 224 RHR28 21) Sequoyah 1 5000327 05123188 021 Mode 5 134 22S Rift28 21) Surry I 5000280 05/'_7/83 83-024 _cJe 5 J_-2 AEOD-A C2 226 RHR20 21) Trojan 5000344 05/21/77 77-016 Node 5 55 rain A.2 C1 227 RHR2B 2b Trojan 5000344 04117/78 78-011 Node 6 A.2 C2 228 RHR28 21) Trojan 5000344 05/04/84 84-010-1 Hode S 40 sin A-2 AEOO-B 229 RHR2B 21) Zton 2 500030/. 12/14185 85-028 Mode S 75 rain C.1 230 RItR2B 2b Zion 2 5000304 01113186 85-028 Node 5 75 rain A-2 231 RIIR3 3a2 Arkansas 2 5000368 11114179 79-087 Mode 5 3 I,r A.T E28 232 RmL3 9 catvert Ctiffs 2 5000318 10117178 78-035 Mode S 2 hr A.IO E45 233 RNIL3 3a2 ¢atvert Ci|ffs 2 5000318 05/24/87 003 Node 6 15 hr 20 rain 113 23/, RIIR3 3a2 catvort C[iffs 2 5000318 05/07/87 004 Mode S 16 hr 55 rain t14. 235 RHIL3 Cry|tat River 3 5000302 02/12186 Mode S 236 RHR3 Sb Davis Ilesse 1 5000346 07/10/80 150-0S7 Hode 6 A.5 E9 N_3 237 Rm_ 3a2 alabto Canyon 5OOO275 06125188 017 Node 5 34 HA2 238 RHIL3 (hi Fartey 2 500036_ 09128183 8.3-042 Mode6 A-6 AEOO-A Ell 239 RI(1L3 Sb HcGu|re 1 5000369 04/28/81 81-073 Node 5 20 rain A.8 E37 NA1 240 RIIP.3 9 lqcGuSre 1 5000369 06/02/81 IE Circ. 81-10 Made 5 A.9 E42 NA1 07/02/81 241 RIqL3 11 H_;ulre 2 5000370 03/07/83 83-002 Mode 6 -3 hr A-IO Mode 8 242 RIIP3 5b Pmtlucles 5000255 01/08/78 78-003 Mode 5 45 rain A.S E8 243 RIIR3 3a2 PaID Verde 5000528 03/19188 022 Mode 6 247 LEAK 244 Rlflt3 5b |area 2 5000311 04/13/81 81-00S Mode 5 5 hr 22 m|n A.7 £33 Page No. 10 08/18/93 PHASE2 - LOSS OF RESIDUALHEATREMOVALlallLE AT SIRITDCXAFORI PI,/RS

INITIAL SEA- PHASE2 SSS EVENT PLANT RECOVERY NSAC52 BROOK CR5015 Gl_ LER NUREGAPPLI- CAT CAT PLANTHAHE DOCKETHO DATE LER CONglTION TIHE CAT. CAT. AEO0 CAT CAT SEARCH1410 CABIL...... ,,...... °...... ,..,...... °...... °...... ° ...... °...... 245 RHR3 3e2 Zion 1 5000295 01/04/Q0 001 Hods 3 7 hr 47 rain 70 246 RHR4 6o Arkansas 2 5000368 08116178 78-001 Hods 5 2.8 hr A.6 E17 NAi 247 RHR4 Arkansas 2 5000368 01/01/79 79-086 Hods 5 248 RHR4 3al Arkansas 2 5000368 10113179 79-086 Mode 6 A.IO E46 249 Rim4 3el Arkansas 2 5000)68 01/09183 83-003 Hods 5 A-IO 250 RNR4 3el Calvert Cliffs 1 5000317 01/24/78 78-004, 78-011 Mode *) and 5 A.IO E44 251 RHR4 Ga Crystal River 3 5000302 04/25/78 78-020 CoLd shut. A.6 E14 252 RHR4 6a Crystal River 3 5000302 02/02/86 86-003 Hods 5 24 rain A-6 E.8 253 RHR4 3c Farley 2 5000364 11/27/87 008 Refueling 167 outage 254 RHR4 5a HcGulre 1 5000369 02/05/81 81-009 Hods 6 3 hr A.5 EIO HA1 255 RNR4 9 I_Gulre 1 5000369 11/23188 049 Hods 6 39 sin 175 256 RHR4 6m North Anna 1 5000338 06/14/82 82-043 Mode 5 A-6 257 RHR4 6m North Anne 2 08/02/82 3.17 258 RNR4 (hi North Anna 2 5000339 08/16/82 82-050 Hode 5 A-6 259 RHR4 3el North Anna 2 5000339 05/22/83 83-042 Mode 4 30 rain A-IO _A 260 RHR4 Oa Rancho Seco 5000312 05/05/82 82-012 Mode 5 A-6 261 RHR4 2he1 Salem 1 5000272 12/13182 82-089 Mode6 A-6 262 RHR4 3al San Onofre 5000361 08/31/87 014 Mode5 18 hr 16. 263 RHR4 3el Trojan 5000344 04/25/78 78-017 Mode 5 1 rain A.3 On1 264 RNR4 6a Turkey Point 3 5000250 10/23/85 85-034 Mode 5 A-7 265 RHILS 5b Arkansas Nuclear 1 5000313 10/26188 014 Mode6 23 rain 100 3.50 266 RHRS 5b Arkansas Nuclear 1 500G313 12/19188 024 zero pouer 12 mtn lOT 267 RHP.5 5a Beaver VaLLey 1 5000334 05/12182 82-018 Mode 5 2 rain A-6 AEOO-A E9 268 RHR5 3c CaLvert Cliffs 1 5000317 02108180 80-011 Mode 5 1 hr 33 rain A.3 De2 CLosecl 269 RHR5 6b Calvert CLiffs 1 5000317 05117182 82-026 2 rain AEOO-A E6 270 RHR5 5b CaLvert CLiffs 2 5000318 01123181 81-003 Hods 6 5 mtn A.7 E32 271 RHR5 Connecticut Yankee 5000317 03/05/88 Node 5 1 hr 2 rain 272 Rats 6b _ttcut Yankee 5000317 03110/88 007 Hods 5 1 hr 22 rain 6 273 RHR5 6b Cook 2 5000316 12/09/82 82-109 Rode 6 A-6

274 RHR5 6b Crystal River 3t 5000302 07116/80 IE Info Notice ;Mode 5 21 Bin A.3, A.7 DeS, 81-10 E31 275 RHR5 9 Crystal River 3 5000302 04/21/81 IE Circ. 81-10_ Mode 5 Similar event A.9 E41 Page No. 11 08/18/93 PHASE2 - LOSSOF RESIDUALHEATREMOVALUIIILE AT SHUTDOWFORN PURS

INITIAL SEA- PHASE2 SSS EVENT PLANT RECOVERY NSAC52 BRIX)I( CR5015 GI_ LEI_ NUREGAPPL|- CAT CAT PLANTNAME DOCKETNO DATE LER COMDITION TIME CAT. CAT. AEOD CAT CAT SEARCH1410 CABIL.

...... o.....oa,..o _o....d,...... -- ...------. 4....,...... _.....-...... ,...... o..o., ...-_. o...... -....°...... -. 07/02/81 10101/81 276 RIIR5 Crystal River 3 5000302 10/01/81 Mode5 277 RllR5 (2) Davis 8esse 1 5000346 05131180 80-044 Mode6 8 ntn A.7 AE(X)-C E29 278 RIIR5 6b Davts Resse 1 5000346 04118/81 81-024 flode 5 1 rain (2 rain in A.6 AEO0-C E24 AEO0-C) 279 RHRS 6h FarLey 1 5000348 05107183 83-009 Mode6 A-6 280 R1¢15 3c FarLey 1 5000348 ' 11/07/86 86-020 Mode 5 A-3 Hocle 12 281 Rlgt5 FarLey 1 5000348 11115186 Node S 282 RHR5 6b Ft. CaLhoun 5000285 10119/77 77-023 Mode 6 A.7 E26 283 RHIt5 5a Glnne 5000244 05101/83 83-017 Mode 5 A-6 AEOO-A A1 284 RIlIP,5 3a2 lndlam Point 3 5000286 05112/83 83-002 Mode 5 A-IO 285 RHR$ 5b Maine Yankee 5000309 06/10/81 81-008 Mode 6 5 min A.5 E12 286 Rim5 11 MILLstone 3 5000423 06108187 030 Mode S 33 rain 1_

287 RHR5 3a2 North Anna 1 5000338 06/01/80 80-053 Mode 5 34 mln A.3, A.7 De4, E30 288 RHR5 6b North Anna 1 5000338 02118183 83-009 Mode 5 5 mtn A-IO AEOO-A C4 289 RHR5 9 Oconee 1 500026? 06/29/81 Duke ttr to NRC Mode 5 A.9 E43 07/31181 290 RWt5 3a2 Oconee 3 5000287 04/06/87 005 Mode 5 61 291 RNR5 5b Pattude_ 5000255 07/18181 81-030 Mode 5 1 hr 30 mtn A.5 E13 292 RIIRS 4 Rancho Seco 5000312 10/03/86 86-016 Node 5 13 mfn A.11 293 Rills 3a2 Robinson 5000261 06/11/87 014 zero power 27 294 RIgRS 3c SaLem 2 5000311 06/12181 81-041 Node 4 1 hr 5 rain A.3 De7 295 RIIRS. 6b Satan 2 5000311 05/24/83 83-025 Mode 5 A-6 A_OO-A E3 296 RHR5 9 San Onofre 2 500(01 03/14/82 82-002 Mode 6 90 rain A-IO 297 RHR5 _b Summer 5000395 11/06184 IE De! ty Report 7 rain AEOO-B 298 RHR5 10 Surry 1 5000280 03118/89 009 zero power 11 hr 28 rain 46 299 RHR5 10 Surry 2 5000281 02119/86 86-004 Mode 5 10 rosin E.4 300 RHR5 7 Vogtte 1 5000424 03/18;87 055 Mode3 205 . HA2 301 RIIR6 8 Beaver VaLLey 1 5000334 05/03/81 81-048 Mode 4 A.8 E38 302 RHR6 8 Crystal River 3, 5000302 08/15177 77-101 Mode4 A.8 E34 303 RIIR6 8 Crystal River 3 5000302 03104179 79-022 Mode4 A.8 F.35 304 RHR6 8 Crystal River 3 5000302 04125179 79-046 Mode4 A.8 E36 Page No. 12 08/18/93 PHASE2 - LOSS OF RESIDUALHEATREMOVALNIlILE AT SHtJrtl)ONNFOIt Plats

INITIAL SEA- PHASE2 SSS EVENT PLANT RECOVERY NSAC52 BROOiC CIL_015 GI_ LEIt. IAJMEGAPPLI- CAT CAT PLANTNN4E DOCKETNO DATE LER CONDITION ThOlE CAT. CAT. AEOD CAT CAT SEARCH1410 CABIL.

...... ****..,...... ,...... ------, 305 RHR6 8 Davts Besse 1 5000346 01/07/81 81-004 Mode4- 15 ,,in A.6 AE(X)-C E22 306 RHR6 8 Glnrm 5000244 03/03/84 84-002 Mode4 A*8 307 RNR6 8 61rtm 5000244 05/14184 84-005 Mode4 A-8 308 RHR6 8 Oconee 2 5000270 09118181 81-017 Mode 1 A.8 E39 309 Rlflt6 8 Oconee 3 5000287 10115185 8S-003 Mode4 A*8 E.7 310 RNt6 8 Robinson 5000261 07115182 82-009 Rode 6 A-8 311 RHR6 8 Robinson 5000261 07/27182 82-009 Mode 6 A-8 312 RIgt6 8 Sen Onofre 2 5000361 04/26/83 83-038 Mode4 A-8 313 RNR6 8 Turkey Point 4 5000251 05/23/89 004 . Node 4 20 314 _ 12 Surry 2 5000281 03112187 001 Node 5 48 315 TRANS TRAILSSurry 1 280 07/09181 026 Routine startup 316 TITANS TRAILSSteffy 1 280 01/05/82 010 Routine SD 317 TITANS TITANSSurry 1 280 01106182 001 HSO 318 TITANS TRAILSSurry 1 280 04/15/82 048 Routine • ptsrtup 319 TRAILS TRAILSSurry 1 280 10117182 112 Routine startup 320 TITANS TRANSSurry 1 280 04107/84 008 Routine SD 321 TitANS TitANSSurry 1 280 06/19184 016 Routine etartup 322 TITANS TRAilS Surry 1 280 01127185 004 Routine stirrup 323 TRAN$ TITANS$urry 1 280 01/28/85 006 Stertup 324 TRANS TRAILSSurry 1 280 02/07/86 010 Routine stsrtup 325 TRANS TRAILSSurry 1 280 02108186 009 Routine startup 326 TITANS TITANSSurry 1 280 09116/86 022, event on CSO q/20/_ 327 TITANS TITANSSurry 1 280 07118/89 030 IOOX, Unit 2 26_* CSO 328 TRAILS TRANSSurry 1 280 07/23189 031 IOOX, Unit 2 266* Page No. 13 08118/93 PHASE2 " LOSS OF ltESIIXJALNEATREHOVALfailLE AT $1flJTOGalfOIL PLatS

INITIAL SEA- PHASE2 SSS EVENT PLANT RECOVERY NSACS2 BROOK CR5015 GIg9 LER NUREGAPPLI- CAT CAT PLANTNAME DOCKETNO DATE LER CONDITION TIME CAT. CAT. AEO0 CAT CAT SEAR_ll 1410 CASIL.

.. _,....,...... ,_ o.,,,.e _...... o.... ,,...,...... ,...... e d,...,,.,...... , q, ...... ,, a,.b......

329 TRAilS TRAilSSurry 2 281 12119182 075 Not SO 330 TRAILS TITANSSurry 2 281 09/21/83 038 Hot SO 331 TRANS TRANSSurry 2 281 04115184 009 2X power 332 TITANS TRAIn Surry 2 281 04120184 012 Hot SO 333 TRAILS TRAILSSurry 2 281 03116187 002 Hot SO 334 TRANS TRAIn Starry 2 281 09/10/88 022 RefueLing outage 335 TRAILS TRAILSSurry 2 261 09/16/89 007 Subcrtticat 336 TRAILS TITANSSurry 2 281 09118189 009 14X power .337 VITAL 1 CaLvert CLiffs 2 5000318 02/04/81 81-004 Mode6 17 rain A.1 A20 338 VITAL 1 CaLvert CLiffs 2 5000318 11/22/82 82-053 Node 6 4 sin A-1 AEOD-A AIO 339 VITAL 6b CaLvert CLiffs 2 5000318 11124182 82-054 AEOO-A E7 340 VITAL 1 CaLvert CLiffs 2 5000318 12128/82 82-055 AEOO-A All 341 VITAL 1 CaLvert CLiffs 2 5000318 01104183 83-001 15 rain AEOO-A A12 342 VITAL 1 Cataube 1 5000413 11/26/88 026 Mode5 2 sec 191 343 VITAL 21) Coemru:hePeak t 07/18/89 3.52 HA1 344 VITAL 1 Crystal River 3 5000302 08/19/78 78-042 Mode6 15 rain A.1 A5 345 VITAL 1 Davis 8esse 1 5000346 06128179 79-067 Mode 5 18 rain A. I AEOD-C A7 34_ VITAL 1 Dlebto Canyon 2 5000323 01/17/86 86-002 Node 5 13 e|n A-I A.14 347 VITAL 1 Olab[o Canyon 2 5000323 06/29/87 011 Mode4 5 adn 118 348 VITAL 1 14cGulre1" 5000369 06/24182 82-053 6 nln AEOO-A A19 349 VITAL 1 NcGutre 1 5000_'9 07113/82 82-053 A-1 350 VITAL 1 MiLLstone 2 5000336 01106182 82-002 7 rain AEOO-A A15 351 VITAL 1 North Anne 1 5000338 01122183 83-003 Mode6 4 rain A-1 AEOO-A A16 352 VITAL 1 llorth Anne 1 5000338 04/22187 007 Mode 5 146 353 VITAL 1 North Anna 2 5000339 04/29183 83-036 Mode 6 < 1 rain A-1 AEOO-A A18 354 VITAL 1 ParD Verde 2 5000529 01/30186 86-005 Mode 5 E.12 355 VITAL 61) Rancho Seco 5000312 06/24102 82-015 Modes4 and A-6 AEOD-A A8 5 356 VITAL 1 Rancho Seco 5000312 11/15/86 86-024 Mode5 A-1 A.12 357 VITAL 1 SaLem2 5000311 05/14183 83-024,1 of Modes 4 and immediately A-1 AEOO-A A4 2-5115183 LE1t S Page No. l& 08/18/93 PHASE2 - LOSSOF RESIDUALHEATREMOVALUHILE A1 SllUTD¢)UliFORPUllS

IlITIAL SEA- PHASE2 SSS EVENT PLANT RECOVERY NSACS2 BROOK CR501S G199 LER NUREGAPPLi- CAT CAT PLANTNAHE DOCKETNO DATE LER CONDITION TIE CAT. CAT. AEQO CAT CAT SEARCH1410 CABIL. - ...... 0...... 0...... 0...... 0...... 0...0 ...... 0...... ,...... e_.....o...... -...... -.-.-....°. 358 VITAL 1 Salem 2 5000311 " 11128183 83-062 Mode 5 A-1 AEOO-A A6 35Q VITAL 61) South Texas 1 5000498 02112189 008 Mode 5 2A3" 360 VITAL 1 Sumer 5000395 11112/83 83-136 Mode 5 5 rain A-| AEIX)-A AJO 361 VITAL 1 Sumer 5000395 10118/84 84-045 "Mode 5 25 sin A-1 AEOO-BA.19 362 VITAL 1 Turkey Point 4 5000251 03/15/86 86-006 Mode 6 5 gin A-I A.3 363 VITAL 1 Zion 1 $000295 03/17/82 82-011 Mode 6 3 min A-1 AEOD*A A3 364 VITAL 1 Zion 2 5000304 01/03/86 86-001 Mode 5 A-1 A.8

Number of Records 364 file: query and report Fullphs2 Appendix A.2 Descriptions of Events Used in the Quantification

A-18 Appendix A.2.1 Loss of RItR Events (RItR2A, 2B, 3, 4, 5, 6)

A-19 p. 1 LER DATA BA.SE- RHR2A 08/27/93

PHASE2 PLANT NAHE EVENT INITIAL PLANT RECOVERY CATEGORY DATE COND! T!ON T! HIE

RHR2A Arkansas Muctear 2 08/29/84 Node 5 (from AECO): During Rls draindm_l, faulty [eve[ instrumentation ted to air binding of the DHRpump. A tygon manometer configuration was being used - however, the operators did not account for reactor vesset pressurization due to the presence of nitrogen purge gas. RCS temp. went from 140 (leg Fah to 205 (leg esh (approx. 35 min. Loss).

(from LS5015)= Power Level - OZ. On 8/29/84 the plant was in mode 5 and the RCS Level was being malitored by a l:elllporary Level indicator connected to the bottom of the RCS hot Leg and vented to atmosphere. A nitrogen purge of the RCS Mas in progress to "sweel_ hydrogen from the system prior to ma|nt/mence. T_e RLS w being vented via the upper vessel head vent and due to nitrogen flow exceeding vent flow capacity the IllS became slightly pre_suPized. This resulted in a manometer effect arid inaccurate indication of RCS tevat. The Level indication irlac_acy Led to draining of the water in the Ills hot Leg below the minimum Level for adequate Ihutdokm coating pump suction. SDS Loop flow indicltton began oscillating betweer_ 2000 and 4000 _ indicating cavitation of the SDC pump. Conuqusntly the "11" SDC pump was nitrogen purge were secured. Decay heat r _ovaL alignment was Idlifted to the "A- SDC tOOl) and norlmt flow of approx. 3000 gpm uas established. During the period SOC flow Mas off, RCS bulk average temperature increased from approx. 1/4 to 205 des Feb resulting in a Ct_lmge frail lode S to mode 4. To prevent recurrence the tel_rary Level system reference Leg has been changed from venting to atmxqphere to venting to the pressurizer stem space. Changes have been to _t and al=normal operating procedures to improve system and operator response to simi tar events.

RIIR2A cats 1 04/22/85 Node 5 12 nan (from Scabreak): Both trains of Residual Heat Removal (lID) _ere irloperabLe. This was the result of NO Train A being declared inoperable on April 20, 1985, at 1600 hrs for the performance of various lID Train A related work requests and NO _ B being secured on Apri t 22, 1985, at 2039:21 hrs due to Loss of pulp suction. ALso, Tech. Spat. 3.4.1.4.2 m violated on April 22, 1985 at 0522 hrs when reactor coolant (MC) system draining began with lID train A inoperable.

(from CR5015): Pouer Level - OZ. On April 22, 1985, from 2039=21 to 2051=17 hrs, both trains of reliduat heat removal (RHR) were inoperable. This was a result of RHR train A being declared inoperable on April 20, 1985, at 1600 hrs, for the performance of various train A related Mark recFaests, and RHR pump B being secured on April 22, 1985, at 2039:21 hrs due to toss of pump suction. ALso, tesh opec 13.4.1.4.2 was violated on April 22, 1985, at 0522 hrs _hen reactor coolant (RC) system draining began with RHR train A inoperable. Catatd_a Unit 1 was in mode S (cold shutdown) when these incidents occurred. False RC system Level indication apparently contributed to the Loss of RHRpump O suction. However, the cause of the false Level {ndication is not knoun at this time. Math RHRTrain A inoperable, the Limiting conditions for operation of Tech Spec 3.4.1.4.2 were not met. However, prior t_, beginning RC sys?om draining, a decision had been made to allow draining to begin with

RHR2A Cook 2 06116/88 22 rain After art fuel from the reactor vessel had been removed with the reactor cavity flooded, the upper internals were _eptaced. The midotoop Mater Level indicator was vatved out. The water Level was radioed by i=Ullpilrlg Mater to the refueling water storage tank at 2000 gpm with a residual heat removal pulp. When the ol_terved water tevet decreasedto the tel) of the upper inter_ats, the pump began to cavitate. The p_ was stopped and vented. The pump was then started at a reduced flow rate. After q=proximtety 22 minutes of operation with the ol=_erved water Level just below the top of the upper internals, the pump again started to cavitate. The Ill{d-Loop Mater Level indicator was valved in and indicated that the reactor vessel Level wa_ at mid-Loop. Because the punlp flow rate exceeded the Leakage rate ar_ncl the upper internals, a void had been created under the _ internals. The industry MILSinforlmd of this event in INPO Significant Event Report (SER) 36-88.

A-20 p. 2 LER DATA BASE - RHR2A 08127/93

PiiASE 2 PLANT NAME EVENT INITIAL PLANT RECOVERY CATEGORY DATE _ I Tl ON T!NE

- _. _ _.Q D-._ _,. Q'._Q. _ e. m m t _ o _ w o.....,.... _ .. _... Q .... _ .... _T..O_.OO..._.._. _..._ _ ...... _ ...... RHR2A Davis Besse 1 04118180 Node 5 29 rain (from HSAC52, p. A-11): RCS _ater Level decreased from 70- to 37" above the hot Leg piping centerLine, causing erratic flow rate clue to inadequate pump suction conditions. The RHRS pump _as secured five minutes Later at a Level of 35". The inventory Loss mm via a partially open diachsrge valve from RHRS cooler tl2 to the makeup and purification system. Some inventory may have also been Lost via cross ccmnects to RHRS Loop 1, uhich _ms being drained.

(from AEOD)= 29-minute Loss of DHR. Leakage of RCS water through s partially closed valve resulted in inadequate DHR Pulp NPSH and erratic pump ftou operation. The pump was secured anti t the Leak bins stopped lind RCS Level restored. During the event, RCS teleperature rose fl _ll 93 dog Fah to 103 des Fah.

RHREA Girm 04112/83 Node, 6 12 uin (from Seabrosk): Prior to the dilute chemical addition, a small amount of water was being used in the channel head in conjunction u|th air pressure at 30 psig to prqperty seat the dam to minimize Leakage past the clans into RCS. During this process, water drained completely through the Leaky dam in the cold Leg nozzle, thus aLLc_ing air to pass through, resulting in an air bubble formation passing through the RCS and entraining the RHR pump suction thus causing toss of the RHR Located on the hot tag of the same Loop. The RHR Ixalp in operation at the time was manually tripped by the Control Room operator to prevent damage to the ptlp. Prior to this event, the RCS Boron Concentration had been borated to greater than 2400 ppm (2000 ppm required for refueling shutdown mode).

(frmm AEO0): Air binding of RHR pump (12 sinute toss).

RHR2A NcGuire 1 03/02/82 Hode 5 50 m|n (from Seabraok): A Residual Heat Removal (liD) pump tow discharge atom resulted tn liD Pump 1A being stopped due to signs of cavitation. With the redundant Pup 1B out of service for maintenance, no means existed for core residual.

(from AEOD): Low RCS Level due to vessel draining and inaccurate Level indication. Operating RHR pump started to cavitate, the other pump was undergoing mintemmce. (Event tasted 50 minutes - a Licensee armtysis indicated that 4 hrs were available prior to the onset of boit|ng.)

RHR2A 14cGuire 1 04/05/83 Node 6 (from Se_breok): The Residual Heat Removal (lID) Pueps began to cavitate and eventually both the pumps were stopped.

Cfrom AEOD): Lou RCS Level due to vessel draining and valved out Level sensor. Both RHRpumps cavitated. (Dural|on of event unknmm)

RHR2A HcGuire 2 12/31/83 Hede S 43 rain (from Seabreok): Residual Heat Removal (liD) Pump B was observed to have zero discharge fto_ and was subsequently tripped and ItD Train B declared inoperable.

(from AEOD): Lo_ RCS Level due to draining and inadequate Level indication. Running RHR pump had no flow (43 sin. toss).

RKR2A ItcGuire 2 01109/84 Node S 62 rain (from Seal=cook): During draining operators of the Reactor Coolant (He) System Residual Heat Removal (liD) Pump B was observed to have zero discharge fLc_. Pump B motor amperage uas Low and the ND system pressure and Pump B discharge pressure were equal. Based on these factors, lid Pump B was tripped and ND train B was declared inoperable at 1650. The FWSTto ND Pump Isolation valve was twice cycled to A-21 p. 3 LER DATA BASE - RHR2A 08/27/93

PHASE2 PLANT IL4J4E EVENT INZTIAL PLANT RECOVERY CATEGORY DATE CONDI T!ON TI NE

....,...... ,.,...... o...... provide core cooLing and raise NC System Level with water from the Fueling Idater Storage Tank while venting the lID Suction Line and pump B. The core temperature rate of rise decreased after the first water addition, and the second addition resulted in slightly decreased core temperatures. NO Pump B was restarted at 1720 and flow was restored. On January 9, 198/,, operators were again decreasing Level in the Reactor coolant Loops i_hen a c_lputer alarm for Low DID PumpA discharge pressure was received. FLuctuations in lID Pump A discharge pressure was received. FLuctuations in liD Pump A motor amperage were noted and simultaneous fluctuations in discharge pressure and floe also occurred. After the Low lid FLow wlnialciator alarmed, NO Pump A _as tripped at 1246 and lID Train A was therefore inoperable. Operators manually opened the lID system to FWST isolation valve, raising the Reactor Coolant Loop Level with water from the FWST. The suction Line and puny) were vented and the pump was restarted at 1348.

(from AEO0): During draining operations, a procedural deficiency Led to inadequate NPSH/air entraimmnt of the DHR imJPS (1 hr. 2 sin. Loss).

(from Clt5015): Pa_r Level - OL On Dec. 31, 1983 at 1640, during draining operations of the reactor coolant (NC) system, residual heat removal (liD) pump B _es observed to have zero discharge floe. Pump B motor amperage m Low, and the HI) system pressure and lump B discharge pressure were equal. Based on these factors, lID pump 6 us tripped and ND train 6 Has declared inoperable at 1650. The FWSTto ND Ixmp isolation valve ms twice cycled to provide core cooling and raise NC system Level with water from the refueling rater storage tank, while venting the lid suction Line and pump B. The core temperature rate of ri_ decreased after the first rater addition, and the second addition resulted in decreased core temperatures. ND pump g yes restarted at 1720, and floe was restored. On Jan. 9, 1984 operators were decreasing Level in the reactor coolant Loops Nhen a camputer slam for Luw ND pump A discharge preslmre was received. Flucttmtions in I} pump A motor amperage were noted and simultaneous fluctuations in discharge pressure and floe also occurred. After the "Lee lid fLo_ annunciator alarmed, lID pump A m tripped at 1246, and ND train A _ inoperable. Operators manually opened the lID m/stem to FIJST isolation valve, raising the reactor coolant toqp Level with water from the FIdST. The suction Line and pump were vented, and the pump was restarted at 1348. These incidents are due to inadequate guidelines recording the tater Level to be mintained in the reactor coolant Loops during lID operet i on.

RHRZA North Anna 2 05/20/82 1 hr (from Seal)rook): Not found in A-2

(from AEOD): Lost suction to RIIR pumps due to draining of RCS and erroneous Level indication (three events= 8 min, 26 min, 1 hr Losses).

RHR2A North Anna 2 05/20/82 8 rain (from Seabrook): Not found in A-2.

(from AEOD): Lost suction to RHRpumps due to draining of RCS and erroneous Level indication (3 events= 8 rain, 26 rain, 1 hr Losses).

lUlR2A North Anna 2 05/20/82 26 rain (from Seabrook)= Not found in A-2.

(from AEOD): Lost suction to RHRpumps due to draining of RCS and erroneous Level indication (3 events: 8 rain, 26 rain, 1 hr Losses).

A-22 p. 4 LERDATABASE- RHR2A 08/27/93

PHASE2 PLANTNAME EVENT INITIAL PLANT RECOVERY CATEGORY DATE CONDITION TIME

RHR2A North Anna2 05/03/83 Node6 (from Seabrook Sheet 3 of 9): OnMay 3, 1983. suction to the B Residual Heat Removal (RHR) pumpwas Lost uhite transferring rater from the Reactor Coolant system (RCS) to the refueling ueter storage tank (ItWST)via the Refueling Purification (RP) System. The A Pumpwas secured and the B RHRPumpstarted but suction ms not available.

(from AEGO): Inadequate monitoring of RCS Level. Loss of RHRptmp auction. (Duration of event unknoun)

RHR2A North Anna2 10/16/84 NodeS 2 hr (from Seebro_ Sheet 7 of 9): A ¢oq)tete Loss of Residual Heat Removat(RHR) capability occurred uhen both RHRlamPSwere unable to operate due to the introduction of air into the RHRsystem. The incident occurred during the drain dean of the Reactor Coolant System(RCS) uhen the Level of the RCSuas being monitored via a standpipe off the centertine of one of the RCS tocps.

(fr_n AEO0): CLoggingof a standpipe used for RCSLevel monitoring resulted in a 64" error. Upon introduction of air, the operating Imp cavitated. The redundant pumpms started and it also ¢arvttated. Both _ _ eirbound (2 hr Loss).

(frem Clt5015): I_xder Level - OZ. On 10-16-84, with North AnnaUnit 2 in Node S a complete Lossof RHR cap_ittty occurred Lawnboth RIR IXmpe rare unable to qxmrate due to the introduction of air into the RHRsystem. The Incident occurred during the drain dean of the RCS, whenthe Level of the RCSwac being mnttored via a standpipe off the centertine of one of the RCSloops. The isolation valve to uhiclt the standpipe ms attached becameclogged =tiara during the drain dean and falsely indicated 64" above centertine _ in fact the Level uas hetou the Rim suction line (belou contertine). Subeacpontty, tetdoun fran the RCSues isolated and makeupinitiated. RHRcapability was regained 2 hrs after init|ation of the event. RCS Level indication =as wved to on alternate tap off Loop center|lee and indicated settsfactorl ty.

RNR2A San Onofre 2 03/26/86 Node 5 49 ain (frm= CLL5015): Pouer Level - 0_. March 26, 1986 at 2208 uith Unit 2 in cold shutdoun, the shutdoun cooling system (SDCS)experienced a total toss of flow for a period of 49 minutes. This occurred white reactor coolant system (RCS) tevetues being reduced to repair a Leaking cold tng steam generator nozzle damuhich had been installed to aLtou uork in stem generator channel heads. Using the established tevet indication, which ues Later found to he in error, the RCSNas drained to a Level where vortexing occurred at the RCS/SOCSsuction connection causing the SDCS/LPSIpumpsto eventually becomeairbound. The pumpswere stopped and the system vented, reestablishing SDCSftou at 2257. Concurrent uith the restoration of SOCSflea, both gas channels of the fuel handling isolation system actuated on high noble gas a result of the RCSdegesing. The high pressure safety injection system uas used to make-up to the RCSuntil Sl)CSflea returned to • stable state. The cause of the event uas erroneous teveL indication resulting in the operators not recognizing the RCS Lea Level condition prior to ce=ptete toss of SDCSflea. lmediate corrective action uas taken to prevent SOCS/LPSIpumpdamage, restore SOCSftou to a stable state and recatibrate the Level indicators. Changesin plant design, procedurat revisions, formal control of Level indicator instatLation, and operator training will be undertaken.

RNR2A Sequoysh2 08/06/83 Made 6 77 rain (from Seabrook Sheet 4 of 9): At 0838 (C) during pumpdean of the refueling cavity to perfom maintenance on the Loop 4 Reactor Coolant SystemCold Leg Nozzle Inspection PLate, the 8 Train Residual Neat RemovalPumPbegan to cevitate.

(free AEO0): False RCS Level indication by makeshift tygon tube and rubber hose Level instrument. RCS A-23 p. 5 LE_OATAeASE- _H_2A 0a/27/_

PHASE2 PLANT NAME EVENT INITIAL PLANT RECOVERY CATEGORY DATE CONDITION TIME

• , _--- -..D Q - i,.,,_ e- u.,- - .. o,,.- -. p.- - - -...... ,.,...... --.-,------temperature rose frm 103 de9 Fah to 195 (leg Fah in 77 rain. PLant had be_ shut do_ 18°days earlier.

RHR2A Trojan 06/25/81 Node 5 75 rain (from MSAC52, p. A-IO): White reducing RCS Level, in preparation for m|ntenance, the RHRSpump began cavitating and was secured. Inventory reduction was terminated and RCS charging was established. Investigation found a shut pressurizer vent valve Nhich isolated the reactor vessel Level indicating system interrmt reference Leg. The pressurizer vent and reactor vessel head vent were opened, and standpipe Level stabilized approx. 5 feet Lower than the previously indicated Level. (cress to the RHRS suction tap off an RCS hot Leg.) Attallpts to restart the ItHRS I=Uml)failed due to air ontrained in the RHRS suction Line. The ItHRS suction valves vere closed end the Itl_T suction valve was opened in order to provide a positive suction to the RRRS IXaq_. FLoe was restored 75 minutes after event initiation.

RHR2A Waterford 03/14/86 Node 5 1 hr 19 gin (from CltS015): Power Level - OZ. On 3-14-86 Ihmterford Steam ELectric Station Unit 3 teas in Node 5 (cold shutdom) (as a result of a scheduled surveillance/maintenance outage which began on 3-7-86) Mith both Loops of the reactor coolant systm (All) drained. At 1035 his on 3-14-86 operations personnel, in preparation for mintenence on $1-_dMA, locp 2 shutdom cooling return relief valve, started toe pressure safety injection (LPSI) IXmp B (BP). The primry nuclear plant operator observed the floe in the B shutdom cooling (BP) train to be zero (0 gl_) and LPSl pump B rotor current to be 33 ants. The control room supervisor ordered the pump secured, and proceeded to the B safeguards pep room to investigate Local conditions. The |nvastigatton revealed that SI-124B, Lay pressure safety injection ptampB discharge valve, teas closed with a danger tell affixed to the valve. The valve was mistakenly closed during a tag-out vhich was conducted on the 3/13-14/86 midnight shift. The valve was opened and at 1154 hrs on 3-14-86 the 8 LPSI llmp tlas pieced into serv|ce. The valve vas inadvertently closed because plant operators d|d not use the clearance request sheet when they conducted the tag-out. To prevent this from recurring, a revision wilt be asde to procedure UliT-5-OQ3, "CLearance Requests, Approval end Retease," end the operet|ons superintendent will stress the function of clearance sheets tdith operations personnel.

RHR2A Waterford 07/14/86 Mode 5 2; hr 41 rain (from CR5015): Power Level - OX. On duly 14, 1986 Waterford Stem ELectric Station Unit 3 was Jn Mode 5 (cold shutdown) when operations personnel were draining the reactor coolant systm (RCS) (All) to facilitate the replacement of the seat package for reactor coolant pump 2A. The RCS was being mintatned by draining into the refueling tmter storage pool (IK/SP) (via the toe pressure safety injection lxaq_ B mtnf-rectrcutation valves, SI-1208, -121B) and holdup tanks (via the chemical volume control system (Cll) purification ion St-&23). At 0113 his operations perlsonnet secured draining the RCS by closing Sl-4Z3. Hoverer, operations personnel neglected to close S1-1208 and -121B resulting in RCS inventory being IXaq:)ed into the RWSP. ]n addition, because of insufficient nitrogen pressure, local reactor vessel Level indication was suspect. At 0317 his LPSI pump B began cavitating. Operat|ons ilmed|atety secured the pump, terminating shutdown cooling (SOC) (alP). At 0658 his SDC kms restored by 8 process of refilling the RCS and cycling the LPSl pumps to restore floe. (Since the RCS temperature increased to the point of Localized boit|ng, the LPSI pumps were subjected to stem binding). This event was due to simultaneously using more than one method of draining the RCS, and inaccurate tevet indication. These problemswilt be corrected by plant modification end procedural changes.

RHR2A 14stafford 3 05/12/88 (from IAJItEG-1410): The reactor vessel water Level was being towered to mid-Loop to remove the steam generator nozzle dams and to test a new dig|tat reactor vessel water Level indicator. Tygon tubing was being used to monitor the reactor vessel water Level. At an indicated Level of 18 ft on the tygon tubing end 14 ft on the digital instrument, the operating Low pressure safety injection pump started cavitating and was secured. The water Level was raised and the other toe pressure safety injection A-24 p. 6 LERDATABASE- RHR2A 08127193

PHASE2 PLANTNAME EVENT INITIAL PLANT RECOVERY CATEGORY DATE COND;T!ON T!ME

Ixmp tam started. However, • Loop seat Jn the _ tubing was not detected. The reactor vessel Hater Level Has again toasted until the operating Los pressure safety injection pumpstarted cav|tating. The reactor vessel Hater Level was then raised, restoring decay heat removaL. The industry Has informed of this event in INPOSignificant Operating Experience Report (SOER)88-3.

RilP.?.A Zion 1 09/14184 Node 5 45 sin (from Sesbrook Sheet 6 of 9): WhiLe in cold shutcloun, draining the RCSin preparat|on for steam generator primry-secordary Leak testing, the RCS Level dropped beLou the suction Line for the RHR Ixmp.

(fres AEOO): Yhite draining the ItCS in prqxwation for primary-secondary Leak testing, the RCS Level betas the DHRsuction Line. The Liquid Level ms being read from a manometertype arrang_ent. incorrect Level measurementresulted from the fact that the _ter reference tag Has pressurized by nitrogen purge gas. RCStemperature increased from 110 dag Fah to 147 dag Fah (45 rain. Loss).

(from CRS015): Pauer Level - OZ. IdhiLe in cold shutcloon, draining the RCSin preparation for SG Ix'tlmry-seconchlry Leak testing the RCSLevel drqN)ed below the suction Line for the RHRpumpas s result of an improper valve Lineup uhtch gave false indication of the RCSLeveL. The RHRpumpHas staffed uhen it biasnoticed that the motor amperagevim fluctuating. The valve Lineup ues checked and the Lineup error corrected. RCS Level Has increased to normal end the ItHRlump uas restarted. RCS temperature increased from 110 F to'147 F during the 45 minutes the I=_mpuas off. No abnormt conditions devatqped as a result of this event. Station procedures siLL he revised to prohibit simultaneous clrstntng mcl purging operations, a procedure for toss of RHRsiLL be prepared. Retraining ui t t be conducted in proper vaLve t ineup procedures.

A-25 p. 1 LERD_TABASE- RHR_B 08/27/93

PHASE2 PLANTNAME EVENT INITIAL PLANT RECOVERY CATEGORY DATE CONDZT|ON TIHE

RHR28 Beaver Valley 1 01/17/80 Node5 (from NSACS2,p. A-9): The reactor vessel vend eductor mis in service in prq:eret{en for refueling. A Lowfloe atam for RHR$was received and Lo_ fto_ and Lowmotor current were indicated. A seco_l RHRS pumpwas started and was also air bound, Both Ixaq_ mire vented and flaw restored.

RIIR2B Beaver Valley 1 04/08/80 Node5 35 rain (from NSACS2,p. A-9): UhiLe attempting to increase RHRSflow to the Tach Spec value requ!red for dilution (3000 gPm)ethe running MiltS lamp Lost flay. The secondRImSpumpuas also found to be air bound. ICS Level ,as increased, _ vented and fLo_ restored in 3S minutes.

RHR28 Beaver Valley 1 04/11/80 Made5 70 rain (f_ lSAC52, p. A-tO): While attempting to increase RHItSheat exchanger ftoN, the running RiOtSpu_ Lost suction. (Heat exchanger flay ms tncr'eaud by' decreasing bypeos flay, not Increasing lamp flow.) The eirbotmd RHRSptatp ms secured. The other pap m started, becameairbotmcl and mis also secured. Both _ trace vented, and It_ Level ms Increased, FuLL Milts flow t restored in 70 minutes.

ItllR3 Beaver Valley 1 03/05/81 NodeS _ rain (from liSAC52, p. A-IO): Reactor vessel rater Level ms inadvertently Layered. The RMS Los fto_ atam mis actuated, accempenied by Lay current on the q:orating Imp. Systemflay dropped to 500 91m. Smippin9 the operating pumpfor the idle lamp resulted in an indicated flow of zero gpa. Returning to the original Islp restored flay to 500 glm. 600 gallons of coolant were added to the RC$, RlfltSIxaqs mire vented, and flay restored in 5_ minutes. RCStemperature increased 66 F to 168 F. lnvesttgatiort revealed that remote Level Indication in the control roomm 6" higher than actual Level as indicated by a Localstandpipe.

•ItHR28 8yr_ 1 09/19/88 Mode6 l& sin (from LERSearch): On Sept. 19, 1988, the reactor mis fuLLy depressurtzed in the refuet|ng operet|onaL modeat • tenpereture of q=prox. 95F. Reactor cavity rater Level mis approx. 4% A reactor vessel stud hole protective |rmert had vorked free and _s floating on the water surface. At 10S2 the 1B residual heat reaovat (RIIR) pulp ties operated to Lover reactor cavity miter Level to the vessel flange to pemit insert replacement. Visual sighting of cavity miter Level was hetieved to be an accurate and timely indication for the evolution based on pest experience. WhiLe completing the draining evolution, the 1ARHRImp showedsigns of cavitation amclwas stopped by a Licensed reactor operator. U|thin 2 minutes of stopping the pump,the reactor vessel m gravity filled from the refueling water storage tank. Within 14 minutes, shutdouncore cooling ties restored using the 1B RHRIxnP. This report is suba|tted voluntarily. The 1A RHRpulp cavitation was caused by the entretrment of air to the IxmPtS Suction. It iS bet|eyed thgt air mis admitted by a vortex vhen reactor mt water Level Lot,rod belay the top of the reactor coolant hot tega. The cue of the excessive Lowering of vessel rater Level mis a failure to comprehendthe fluid restriction created vhen the q:per internals essmbLy is fully seated on the hold do_n spring.

RHR2B Cook 2 05121/84 Mode5 25 mid (frm Seabrook): Yith the Reactor Coolant Systemat haLf-Locp, the control Roomoperators started a secondresidumt heat removal (RHR) pumpin preparation for remov|ngthe operating RHRpt,llq}fr_l service. Utth both _ training, fto_ becameexcessive for the half Loop condition causing cavitation and air binding of both pullS. Both puq:6 tdere out of service for approximately 25 minutes Nh|te they uere being vented uhlch is ultMn the one hr.

(froa AEOO-B)= Procedural error t_ith a partially drained RCS. Simultaneous operation of two DHRpumps causedvortexing 8t the Loopsuction. Both pumpsbecamea|rbota_cl(25 Din. toss).

A-26 p. 2 LERDATA BASE- RHR2B 08/27/93

PNASE2 PLANTNAME EVENT |NITIAL PLANT REI_)VERY CATEGORY DATE GONDITION TIME

.-.-.....,.,.,..,..,-...... o°...:..o.oo.:.o ...... °...... o ...... ° ...... (frm CRS015): Polter Level - 0_. With the unit tn co|d shutdom (HocleS) and the reactor coolant system st half-loop, the control roomoperators started a second restcluat heat reaovst (RIIR) punp in prq_ret|on for removing the operating RHR_ frca service. W|th both _ running, flow became excelmive for the half-Loop canclttlon causing cavitation end sir btnd|ng of both pmq=e. Both pumps were out of service for q:l:_o 2S mini while they were being vented Nhtch is within the 1 hr action statement tim Limit of Tech q3ec 3.4.1.3. To prevent recurrence the procedure uhich controls the operation of the Rim ptaips has been cher_KI to Include specific initructions to stop the operating pump prior to starting the secondpumpNhtte st half-Loop.

RHItZll Dfabto _ 2 06/10/87 Rode5 1 hr 28 sin (from LBI Search): On April 10, 1987, st 2123 POT, with the Unit in Node5 (cold shutdoun) during a refueling outage, RHRfloe eas interrupted uhen both RHRtrains becameInoperable clueto airbound RHR Imlpe. The 10 CFR50.72 report Is rods aZ 2230 POT. The RC$had been drained to aidtoop Level for t;6 nozzle claninstallation. The toss of ItCSinventory to the reactor cootlmt drain tank due to a Leaking valve causeda clecrease In RI:Seater Level, vortexlng in the plape' suction Line. and sir entrainment tn the Rim IMqs. At 2251 POT, after verification that the SG_ were still installed and after venting of the lNIt IXmlm, the RCS_m flooded frm the refueling eater storage tank and an RHRlamp started. RHItflow eas tnterrq)ted for _prox. 1 hr and 28 minutes. This resulted in m Localized boiling ancl, _ntrary to TS 3.0.6, an irmdvertent entry to Hods6 (hot shutdown) but no c/toga to the core or sl_lfl_ _lotqlcat release _r_. The unit i stale at 0230 POT, April 12, 1987, _d i _tum_ to _nmt Node5 Itdt_p operation. The mam_ a:ti_ taken to prevent recurrence of this event, Including procedure revisions, training, md design changes, ere described in the text of this LBt.

RIIR28 Illttstone 2 03116/79 Mode5 (fram NUTS|, p. A-9): ItCSLevel ms drained to the hot te9 center tins for I;6 eddy current testing. RHRSflow m Lost due to RHRSpumpair binding. Attempts at IXJp venting and transfer to the alternate lamp Here urmuccatmfut, laten RCStemperature increased to 190 F, the RHRSpumpsuction from the RVSTeas opened to prise the suction. This action restored flow, but resulted in RCSftoed_ and q_ittover of N3_ox. 15,000 gallons of eater thrash the open S/6 mnww to containmnt. Temperature reached 208 F during the transient. Contahlent integrity was verified prior to entering Hocle6.

HR28 North Anna1 10/19/82 Hods6 36 rain (from Seebrook): On October 19, 1982, suction to the A and 8 Resicluet Neat Rmovst System (RHR) puq:$ eas Lost for _ 36 minutes. On Oct. 20, 1982, A and B (RXR) pumpsuction was Lost for 33 minutes.

(frm AEO0): RCSdrained to below contertine of hot leg nozzles. RHRsuction was Lost becauseof tow RCS Level and incorrect Level indication. (10119182, 36 rain Loss)

RHR28 North Anna1 10120/82 Hode6 33 rain (from Sesbroak): Not found in A-2

(from AEOI)): RCSdrained to below centertine of hot tag nozzles. RHRsuction was Lost becauseof tow RCS level and incorrect Level indication. (10/20/82, 33 rain. Less)

A-27 p. 3 LERDATABASE- RHR28 08/27/93

PHASE2 PLANTNN4E EVENT |NZTIAL PLANT RECOVERY CATEGORY DATE CONDITION TIRE

,= .=.,-= e..,. o,=,=.o m I,= e. =o m...,i.,..,, o m e...=...,...,.m ..v._ ,. ,.,=... o ..... o.,. o,= * o J**,=*. i*.,e ,=*.. o ,*.. m o*,,.,.,.*,, .* ,,o I o..,=...! e. 0,. e*- Mt2B North Anne 3 07/17/82 Oqodo5 (from Sesbrook): Not found in A-2

ItHR28 North Anna 2 07/30/82 46 sin (fren AECO)I Lest suction to .An RHRpulp due to draining. Diagnosed as a pulp problem. The =g" ues than started and It also betas atrbound (&6 man. toss).

ItHItEB PaLisades 10/15/87 CoLd=hutdoun 29 gin (free LEItSearch): OnOct. 15, 1987, at 1837, Los prNlmre safety injection (LPSX) Ixmp# P-67A (P;P) ms mruitty xcurod from operation duo to erratic discharge pressure and item. The reactor uas in cold mutdmm condition uith the prtury coolant system (PCS) (All) drained to the 6170p Level (osntartlne of the hot and cold Legs Is at 618'2") at the tim of the event. The PGSms dralnect to the osntertlne in ocder to _t stem generator (AB;SG) nozzle clan modifications. At the time of the event, LPSI pulp P-67A ms taking suction free the PCSat the hot tag, discharging through shutdmm cooLing heat exehmwere E-60A and |-608 (BP;NX) and returning ftou to the PCSst the cold Legs. The erratic dloeharge pressure end ftou mere the result of an improperly pteced jumper uhlch r.ma_ a LPS! disobarge valve to cycle o_m/ctosed. The failure to property pLace the jumper m the result of a cbtm t_|tton error chr|ng the ptsnntng phase for "as-Left u valve tNtlno. Shutdooncooling ftos ms isolated for 29 minutes, ulth PCStaq=erature Increasing from 92 to 129 degrees F. Shutd0un cooling flow ms restored after the errantly placed japer was removed. ALL similar valve testing tam immdlatety stqpped and eLL Julpere InstaLLed to sq=port testing removed.

(free NUlEG-1410) The reactor i in cold shutckxmuith the reactor cooLant systemdrained to mid-Loop for steel 9oneretor nozzle dm edification. An errantly placed jumper in the control circuit for the qporatlnB ton prommre esfety Injection IxmP cased the valve to cycle openend close. The valve cycling caused the reactor coolant system Level to rise and faLL, whtch resulted in 1000 9eLLonsof ureter being pulped out through the open steam 9eneretor mnmys. The FLIp ms then tripped by the aporetors. Outing the 29 minutes decw hast romvat ues test, the reactor coolant tamp. Increased from 92 des Fah. to 129 des Fah.

RHP_B SaLem1 06/30/79 Hock6 34 man (from NSACS2,p. A-9): Reactor vessel Level ws touered to q=proxinateLy one inch above the Los operating Loyal for the RHRSIXap. The operating pumpstarted to Lose suction and ues secured. Itmmtor vaamet Level urns tncreued six Inches, and the RHRSIxnp restarted. Ftou ues Lost for 3A minutes.

Itillt28 Sequoyah1 10109185 Node5 (free Seobrook Sheet 8 of 9) At 1807 (:ST during cold Shutdosn, step over from S Train to A Train Residual Heat Renovat (RHR) resulted in both trains becem|ng inoperable due to air injection |nto the suction of the p_ps. This requires both pumpsto be vented and required RCS Level to be raised from 695 ft. 1 in. to 695 ft. 5 in. to prevent a possible recurrence of the vortex problem. Suction for RHR comesfrom the Loop 4 hot Leguhtch has centertine of 693 ft. 5 in.

(from CRS015): Pouer Level - OZ. On 10-9-85 at 1807 CSTduring cold shutdoun, swap over ires 0B' train to 0AOtrain RHRresulted in both trains becoming inoperable due to air injection into the suction of the preps. This required both pumpsto be vented end required RCS ta_0etto be raised iron 695°1" to 69Sa§m to prevent a pmmtbte recurrence of the vortex problem. Suction for RHItcomesfrom the Loop& hot to9 uhtch ham• osntar Line of 6gS05=. The cause for the Los= of ftos cambe attributed to the additional suction caused by placing the standby RHRpump|rmervice coupled uith the tomRCS Level of 69S,1-. Systemoperating instruction (S0X)-74, "RHRSysteln,= is being revised to change the tamer RCSoperating Limit from 695'0" to 69So6" and mill require the operating pumpto be removedfrom A-28 p. 4 LERDATABASE_ ._RHP.28 08/2"//93

PHASE2 PLANTNAME EVENT INITIAL PLANT RECOVERY CATEGORY DATE CONDITION TIM ...... "';''"'°'T ...... "...... "..... """ ...... service prior to •t•rttng the standby p_p. The unit UN in cold shutdoun with only a 0.2 degrees F rise in RCStemperature resulting from the event. Tech Spec3.&.1.4 action (8) says that "... tilth no tint tools in ql:eratlon, suspendeLL qi=erattons involving 8 recks:tim in boron concentration of the ECS.= At the time of this event, the ¢hmtcaL volume control systm iCe"S) (Ktkaup System) ues tawed out of service; therefore, no violations of tach Specs occurred.

ItlW28 sequoyah 1 01/28/87 Mode5 30 mtn (frOM UER_lSr(:h)= Unite 1 Mid 2 were Ih ModeS at O_ poller. Previous to tht• event, the Unit 1 IICS voter Level ms drained doun to elevation 695,6. for SGmaintenance ark. Oue to s faulty tenet indication of 696'6, In the ItCSvoter level indication 8yJtI, the operator i touerlng the voter Level beck to ahem 695'6 =. The Itm lamp 1A-A began to Lose suction IN; indicated by the Itor currant and pumpftov rate leer8 osciLLating and the Ilnlftov valve q=oning. The operator ImMdlatety stopped the pump; entered the action stetI of Tach Spec LCO3.&.1.4; and begunto align lent Imp lI-I for tetdoun to chemical votI control system (CVCS), vent pumpIA-A, end raise voter Level using the charging pump. Shutdoun decay host removal vos toot at 0620 EaT and reinstated at 0750 EJT. ItCSteal). Inoremsed by 20 F, and the detwJbte boron concentration was not reduced. About 500 gaLLonswere =piLLed frm the SG mmmlm •laDe the voter Level uea being Incrmm_ by the charging pump. No permit voro injured, and there ws no equtl=mnt drags. The cause of the faulty Level indication m coLLection of debris in the InLet to the sight item• tdhlch m being monitored. Operatiom initiated voekty ftushJnga of the sight 9tus and instaLLed a redundant tygon tube vhtch wILL be camperad to the sight glass Level an • doily Ixmts.

(from NUREG-I&IO): The reactor coolant system had been reduced to mid-Loop for stem 9anerstor york. Became of a faulty Level indicator, the Nectar coolant ulmtm rater Level m reduced until operating restduet heat remvat lamp •tarted Losing suction. Approx. 500 gaLLonsof voter yore spiLLed frm the stem generator mmays Mum Level ms lncrac6ed using the charging pump. The false Level lad|cation i cawed W the pMmm_ of dd_ls in the inlet to the sl_t 9Lm _1_ ws heing mntto_d. _lng _ _ mim_os _t d_w kat _m_L mm Lost, the reactor cooLmt ti_. I_e_d frm _ dq F_ to 115 dq F_.

RNt28 sequoyah 1 05/23/88 pkxle5 (from LEIt Search): On Nay 23, 1M8, at 1215 EDT, vhite Unit 1 was in cold shutdoun tilth the reactor coolant system (RCS) partiaLLy drained to support eaJntenmce, • Loss of the operating train of the residual heat removal (RHIt) system occurred. The qi= train of NNt was in operation uhon it was decided to peace the aA" train RHRheat exchanger in serv|ce to er_nce plant tmrature controL. To peace the aA= train in service, an ess_stont unit operator (AUO) _ dispatched to open two valves. The AUO, he_r, re|•understood the instruction and wrote deeman incorrect valve number. The incorrect valve _s s mmmL valve (1-NCV-74-34) used to align the discharge of the RHRpimps to the refueling tatter storage tank (RWST). Upon opening valve l-NCV-74-34, the AUOheard unusual ftou noise and subsequently telephoned the control room ((:It) operator for further instructions. The assistant shift operation supervisor (ASOS)in the CR received an RHRmint fide alarm, and noticed RHRpumpamperageosciLLating, unstable ftou indication, and the indicated RCSvoter Level uas off-scaLe tou. The ASOSsubsequently stopped the "Be train RliR _ and entered the applicable action statements of Tach Specs for • Loss of Rim. The RCSms then refiLLed above the top of the tiCS Loopsby 9revtty feed from the RVSTvie the RHRsystem.

RHR2t Surry 1 05/17/83 Mode5 (from Seabrm_ Sheet 3 of 9): The B Rim pumpyes removedfrom service on two occasions due to cavitation. Tht• resulted in tess than ttio operable Rim Loopsend no loops in operation.

(from AEO0): Inaccurate standpipe Level indication - to_ RCS tevet, RHRpumpcavttated. (Duration of A-29 p. 5 LEII,,DATAllAI - thrill 08/27/93

PHASE2 PLANTNAME EVENT IN11TIALPLANT RECOVERY CATEGORY DATE CONOIiT11ON T1114E oeeel eeeee eeeeoo _._IoeqDeoeeeo _ e I eom oeee_oeeee eegee eeeeeeee oo _e_e_e u_oeo oea_eoooeeeee eoeeemo oea_o oee_eelaeee event unll._)

ENI_ Trojan 05/04184 Mode5 &Outn (from IlesbroJ ht 4 of 9)z I)urtng an IICSdrolln clounto =tiprt refueLIq qxrotlcm ot 16S0on Nay &, lg_, Residual Noat bmova11(Hit) CooLing could not be rolnttlatecl for a total of 40 minutes clueto air entrstraant in tho aucttan of the A Rim puq_. Both ElUt_ had been etol_d for tan minutes for mvulL UF aotustlan respmtse tim testing. An eddttllonaL 30 minutes uero reck|red to restore sufflolant lies wtor Inventory to restart an EHRpuq=. Tho II IUEI pap ms then started and the RCS tmperoturo rim ms terminated. The highest Indicated Rill hot Le9 tmperoturo reached abeu¢ 20t doll Fahr_he|t.

(flr_ AEOO)= During RCSclrotndom fauLty Level esesuresmt Led to air binding of the Elm lU=P. Tho IICS ties vented to e_. A tyllen ummester eanflgurat|cm was being used to aeesure RCS LeveL, heuevor, =c_ud bteskage= of the almnetor tap Led to orraneom level aessuresant. ECll tesporeture uant from lOS cb9 Feb to =01 dell Fah (40 mtn. Loss).

ImlP2B Zion 2 1Z/14/85 Node5 75 aln (frm a15015)8 Pouer Level - _1_. On 12-1& at 3825, 211EHEpumpbecau =lrbeund u a result of ==lng. Unit 2 m in estd _ utth _ r_tor head Ir==t_tted but not temlemcl and the It= vented to ntam0eqpl_. 211ItHEpump(ted bean t, _rotlan prcwlding decay heat reno_L ,tth IHII L0t= tn pregrm and 20 che_tng pumpIx_yvlding m=ke-up ft_ to the _. _ heat removal Is Lest for 75 minutes utth • _ chen_ tn INwerature of t5 clegr_s F. Tho unit had been dwtcloun for q_uqux. 100 ciws theroforo the safety elgnlftcan_ uu atnlesL. The cause of _ ever_ ues Identified to be tnedeclu=to preesdures coupled ulth the Lack of Imoutedee of the Level at uhlch tho IiNItpu_e begin to cmrltato. As a ccxttrlbuttng factor, there uero pr_bLau found ulth the Level Indtcatlan. To _, preeedm_ ultt be revleued end chewed roftecttne the teesorm tesrmd. /Trelnlng uttt be cendu=ted an ItCS Level meaurmmt and Loss of ilmt suction. Tho ItCS Level systm uiLL be aodifted tn order to provide reLiabLe remote Level Indication during aLL refueling ¢cmflipaatlons.

RHII2B Ztan 2 01113186 Mode 5 75 rain (Sheet 9 of 9) The 29 ResidUaL Hoat Removal(RHR) puq) becameairbeund am I result of vortexlng, Unit 2 ms in coLd shutdoun (Node 5) utth tho reactor hoad InstaLLed but not taneimed end the reactor coolant system (RC_) vented to amosphere. 28 RHRpulp had _ In c_erotton providing ¢lecw heat raesvut ulth Elm Lotdoun in pregre_ _ 211chargllng pumpprovldllng makeupflow to the RCS. Decay heat r_no_t ues Lost for 7S urintu_ ulth an ECSchance In temp. of 15 deg Fdtranhelt. The unit hid been _lxltcloun for q=prax. 100 clays; therefore the safety significance Is u_nimL.

A-30 p. 1 LERDATABASE_-RHR3 08/18/93

PIIJUIE2 PLANTNNqE EVENT INITIAL PLANT RECOVtRY CATEQOItY DATE COtIDITloll TIME

oeQ a e De i*,oo oDD o ee o ee,i a,_ o eit*_ eo*,o o_*,l o De. e o e e De w *, e o o m _ e.a,, e a e u ,_ *, • _ _ *.Q ,, ,ma eLIDe em Qoal ,e a e o ode _ 0 o o O,, o I*,*,1 o I*ooo O*, l, O ,,me RHIL3 Arkansas 2 11/14/79 Node5 3 hr (f_ _-_2, p. A-23)1 atoLL Leek discovered in socket weld on LPSI fL_ orifice valve. LMI _tm (iNn) i out of _ml_l_ for three hrs for weld r_tr.

i_ CaLvert CLiffs 2 _4/87 Node6 15 hr 20 mtn (f_m_i _m_)i At 1330 onkr_ 24, 1987, ulth Unit 2 tam6 at _r, two In_te tee_ _ dl_mfl_ in the 1_ m diameter inLet pIpl_ to s relief valve (2-RV-&39) in the _nm cooling (_)/tw _ safety tnJ_tt_ (LAX) s_tm h_r. bret_ _ter m s_WI_ frm the s_m _ _pl_m _t_to _,eh_r_zztest the rate of W_. 1 _. At _onb_ 2&, 1_, Mm_mm _tl_ ft_ m _ ktou the T_ _ r_tr_ of 3_ _ wd _lt 2 enter_ _tlm stat_t. At _ _ _ _, 1_7, an aLt_te ¢_tt_ _ f_ t_ _t_ mt hot t_ _te, _ L_I _ _ th_m_ _ _ h_t _mr to t_ _t fat _t _d _ fat tr_fer _mL _d to _ re_tl_ _t m _t_tl_. At _ _ork m _m _ _taci_ t_ L_I_ pl_ mt W _tt_ _tva ut_ s t_pere_ m_,mtcat _lco (I.e., nl_te _ m_) _r pt_ WamPum. Aft_ this mtmmm_ m _Let_, _ m rmmt_tI_ in _ m_mt m st l&_ _ _, t_. _e tim q_R in the action StSt_t (i.e., _t _tl_) mm 15 _ _ ml_. _ _ 31, 1_7, = facility _ _mt (F_) _7-_ me _ to Im_tl_te U_ pipt_ _ _ _l_tlon st _V _9 _ to I_tmmt _t_ _ if _1_.

IUIR3 CAtwlrt CLiffs 2 05107/87 Hode S 16 hr 55 rain (finn LEIt Ilesrch): At 0700 on Nay 7, 1987, with Unit 2 in Hods S lit O_ poser, a Leek tlae discovered in the 0.5 e diameter InLet pipl_ to • relief valve (2-_-439) In the __ cootl_ (_C)/tov _mmm_ safety injection (U_|) syetm header. At 1_ _stmmnt tnt_riW m_ eet_tl_ W pt_ and an alternate cooling _th to the core tm established mi_ hi_ _'_sure safety injection (_!) _ _ and _elmmt _rey _ _1. At 1&10utth no _ Loopsqx_te_ unit 2 _ an _lm etst_t. The cram pl_ spooLtam repLaced and _ Loop_1 ms retm_ to service at 235S on Nay 7_ 1967. The event duratlm tim 16 hours and 55 ainu|as. BeCameof a slat Lar failure of the inlet pipe to this relief vulva on Itsr¢l_ 2t, 1wrr and the sdbmmpJantinstiLLation of a mm InLet ptpl_ _rt; anat_ls to _tm_ the _a_ of faiLure uas _m_t_. _tett_lcat examination _temlm_ t_ mxb of failure to _ fatt_ _ to _ttc torstonat_di_ t_t_. armtysls and redesign of the inlet piping m performd and the new piping configuration installed on 23. 1987. Piping seac©iated uith other reLief valves in the tJ_l and IIPSl system uftt be revlet_l for similar configuration and support problem.

R_ Crystal River 3 02/12/86 _ 5 See description of C_taL alvar 3, 02102/86 event.

RHIL3 FarLey 2 09/28/83 Node6 (from Saabrcok): The 2B RHRpup tripped _hi te the B RHRtc_ mm tn service and the A RHRLoopwas secured.

(from _): _rst|ng ItHRpumpfailed while r_w_mt _ uN secured. (Duration of _mt _)

RH!L3 NcGuire 2 03/07/83 Node6 3 hr UhlLe Iovino • t_8_ t_e _t_tor, the rope used to hold the detector cable feLL Into the reactor vessel and m dram into C Hot Leg. Residual Heat Removal(RHR) Pump2A ms secured from eel|ca for 3 his (uith _ 211also secured) to aLLowrope retrieval efforts.

A-31 p. 2 _|R DATABASE- RHR3 08/18/93

PNAH 2 PLANTNAME EVI_NT |N|TIAL PLANT RECOVERY CATEOORY OATE CON[DTION T1ME

_.km • ,_ e_ o....,,. Q...... ,...... -,,....,.... -- .--,.-..- ... °.., ...... • ... • ItHIt3 SaLem2 04/13/81 H0de 5 5 hr 22 nin (frul NSACS2,p. A-2S): ALL RCPsand RHRSIXaq_ were de-energized to fac|t|.tate the repLacementof 8 relief valve tn the HRS. Tech specs aLLowthese Ixaupsto be de-ermrgtzed for one hour. The mlntenmnce and restoration of IIHiISflow requfred 5 hours and 22 minutes.

A-32 p. 1 LERDATABASE_-RHR4 08/27/93

PIMSE 2 PLANTNNqE EVENT ZN/T|AL PLANT RECO_RY CATEGORY DATE CONOIT 1ON TIME

RHR4 Arkansas 2 01/011_9 Mo_ S* ' See description of Arkansas 2, 01/09/83 event. (Date of this event is dummydate--date not specified in _,_riptlon.)

RHIt4 Arlumeas 2 10/13/79 Mode6 (from NSACS2,p. A-30): An RHRSheat exchanger tube Leakms indicated by an increase tn the heat exchanger radiation mlftor reading. Theheat exchanger wes then isolated on both the service Nater and reactor coolant stdes.

Re& Ar_ 2 01_/_ _ $ An atam ues received on the Servtce Vater (St/) mnitor (WItITS-14b'3)for the shutdouncoo|_n9 (SOC) heel exchangor (H-3SA) uhich Indicated the poea|blttty of • tube Leak. Redlechem|stry eamLtng verified the tube Leak. 2D-SSAul secured and the redundanthaet exchanger (2E-358) ues placed in service. Since this teak frm the SOCr_tm to the W syltl tins a degradation of • system designed to contain radleacttve rotter. A prevlaue occurrence regarding SO¢tube teaks Is reported In LEII-79-OIM. It should be noted that thlo heat exchanger tube bundle had been repLoced during the last Untt 2 re_tl_ outage.

Rt& Crystal River 3 0_ NodeS 24 Iln (frm Seabrm_). The Iteec.tor Coolant system ms vented to the Reactor Btdg ataosphere and drained beL(mthe Level of the rm_tor cmtent pums. At 21_ hrs, decay heat IXm tripped due to • wter _ert_d caused by a _ _hoft fat(ure. Start up of the redundant ptalp ms delayed because an Isolation valve on the su_on side of the pip c_td not be q_ened fr_a the Control Room. The valve t IInullly q:Mwd Mid _tIO operation uas mtorsd at 2212 hrs. On 2/12/&S, the S train of the Decay Neat Remvel Systemms being refilled and reverent of the _ and piping uas noticed. ExaiinuUon of pipe restraints tn the SyStem reveoLad that several pipe hangers mrs Looseor damged. Art dmuged eciutpmnt hu been repaired. Both decay heat lips have been rebuilt.

(free _S015): Pouerlevel - 0_. OnFeb.2, 1t, Crystalliiver Unit 3 .as in Node§ White perfoming _I_ on a _tor _|ant _. The _tor coolant _tl I vented to the _tor bL_ amm_mre md _atm_ _tow the Level of _ reactor coolant _I_. At 21_ hrs, _W heat _ IS tri_ _ to a Iotor overt_ cauIad W a _m _aft fai lure. Stert-up of the r_t _ ms delayed because an Isolation valve on the suction side of the txap could not he openedfrom the control tel. The valve ms mnuatty opened andsystem operation uas restored at 2212 hrs. On Feb. 14, 1t, the "In train of the decay heat reamer system uas being refitted and movementof the p_p and piping I noticed. Exmtnation of pipe restrI|nts in the SystI revealed that .seyeraLpipe hangers were loose or dmaged. All damged equipmenthas been repe|red. Both decay heat pu_s have been rebuilt. Decay heat removal system operating precedures have been revised to address Iiniu required reactor coolant Level and provide I_ fill and vent instructions. N_ breaker and torque s_|tch settings have been established for the isolation valve. Preventative Iaintenance procedures uitl require periodic tubricattm of the valve drive shaft.

RHR& Farley 2 11/27/87 Refuel tng outage (fr_ LER Search): This special report ts being elated in accordance uith Technical Specification 3.4.10.3. At Ot_5 on 11-27-87', during a refueling outeD,, a residual heat removal (Rim) Loopsuction pressure relief verve openeduhenone of two series RHRconta|rment sunp suction valves |n the same train (NOV8812A) uas stroked. Reactor coolant systi (RCS) pressure and pressurizer level decreased. The Lave| and pressure in the pressur|zer relief tank (PRT) (to uhich the RHRrelief valve relieves) Increased and the PRT rupture disk r14ptured. Based on these tndicat|ons, |tuas believed that the relief valve my have stuck Jn the open position. To |solute the relief valve, the operato_-s stopped the RHR_ in the affected train and closed the loop suct|on isolation valves. Th|s event uas apperentty caused by a Localized pressure pulse created uhenNOVM12A _as opened. The section of pipe A-33 p. 2 q.ERDATABASI_- RHR4 08/27'/93

PHASE2 PLANTNAME EVENT INXTIAL PLANT RECOVERY CATEGORY DATE CONDI Tl ON TIME

..._.... I.Q , ....,...... ,,...------..------between the two contatrment suap suction isoLation valves had been drained tO perfom • Local Leak rate test and had not been refitted. Ap_rerttLy, the fitting of this space, Idlich occurred whenNOV8819.4 ms opened, cued a tocatized pressure pulse .hich resutted in the Lifting of the retief valve. To prevent recurrence, the appLicabLe procedures wiLL be changed to ensure that Lines drained for Local Leak rate testing are subsequently refitted.

ItHR& NcQuire 1 11123188 Node6 39 rain (from LERSearch)= On 11/28/88, at 1643, perfonm_e personnet uere perfoming the contaimont spray (NS) systemvatve stroke tilling procedure for valve 1NS-IB, NSsuction frm corttsirllMnt sump. About25 sac after the execution of the valve stroke timing test, RHR(liD) pump1a tost its suction pressure and q:erattons (OPS) mnustty stopped the Ix-,p to prevent clmmge to the pump. Unit 1 uas in mode6, refueLing, Idth train 18 of the lid iq_tMi In qperation. OPS impLementedLoss of liD procedure, and directed the rlmo_or head mtntanance crlm to exit the reactor cavity area. At 1722, OPSafter ensuring that lid train 1Atms properly fiLLed and vented, cross connected train 1A to the train 13 heat exchanger to pressurize the train IA piping, started NOpuap 1A, and put train 1A of the NOsystem in service to provide decay heat rlmovat for the reactor coolant (NO) system. Between 1643 to 1722, the NC system tap. increased from 90 F to 116 F. At 1930, after fitting and venting lid train 18 piping and NOpap IB, OPSstarted H() IXmp lB. At 1950, train 1B of the liD system was put beck in service and at 2108, OPSsecured liD train 1A. This event is assigned a cause design deficiency because inadequate system venting in the horizontal piping between valves 1NS-1Band 1NI-184B, Sl contairtMnt IKJp B train isolation, stLotled air to be trapp_ in the piping and forced the air to the suction of lid pump1B causing the ImP to Ixcom air bound.

XMt& North Anna 1 06/14/82 Node 5 On 6/14/82, sly one of the coolant Loops Listed in T.S. 3.4.1.3 tms operabte clueto the failure of the Nestdust Neat RemovalSubsystem| Pimp (1-1tlI-P-1B).

Illm6 North Arm 2 M/02/82 With the reactor coolant system drained to mid-Loop, one mtdust heat removal _ was operating. Whenftou from the operating residus| heat remvat _ decreased, approx. 200 gattons of _mter was added to the reactor coolant system to restore fto_ to normaL. Since suction to the operating residual heat rmovat ImP was mints|ned, decay heat rmovat cafmbittty Has not test. the _ter Sever betou Eid-toop becauseof a seat Leak on one residual heat remoras IxmP combinedu|th inaccurate reactor coolant system miter revel indication.

RNIt4 North Anna2 08/16/82 Node 5 The 1AItesidumt lint Rut (RHR) pumpus r_ from service thereby Leaving operable only one loop for decay heat removal since the 1B RHRpumpremained avai table to ensure decay heat removaL.

RHR4 RanchoSacs 05/05/82 Mode 5 (Sheet 1 of 11) The A Decay Heat Pumpinboard bearing uas found to be Leaking oiL. Since the unit uas in cold shutdoun, the only consequenceus the changing over to the B Decoy Neat Ixmp,

RNR4 Salem 1 12/13/82 Node6 (Sheet 2 of 11) At 0750 hrs, 12/13/82, due to excessive teakage from the llechanicot seat, No. 12 Residual Heat Removal (RHR) pumpms declared tnoperabte and Action Statealertt 3.9.8.2 was entered. No. 11 RNRIXap tms started to provide RI_ flow.

A-34 p. 3 LERDATABASE- RHR4 08127/03

PHASE2 PLANTNAME EVENT INITZAL PLANT RECOVERY CATEGORY DATE COH1TD l OH TXFIE

l_i gmlo 4 ._ e,qF__. 0 IQ "_ " _/1_ " _ "_W tile" el Sill. ee.! l." I eeelm .lie I wee i lee I Ilia ! e w Ill We ! e I i ! ! ! i el I I e i I U ! I .. I e e i e. el O. ie i e ! RHR4 San Onofre 08/31187 HOd. 5 18 hr (from LERSearch): On 8/31/87, at approx. 1900, idith Unit 2 in Mode5 and the reactor coolant system (RCS) at approx. 350 Paia and 127 degrees F, failure of aLLoy steel packing gland foLLower studs during minuet operation of motor operated shutcloimcooling system (SDCS) suction isolation valve 2HV-_J?8 resulted in Leakageestimated it 100 glm through the packing gland. Operation of the SDCScontinued via a redundant ftou path. RCSinventory Ires nmintained by isolating Letclmmflow and using charging as clepressurizstion and venting of the RCSproceeded. Containmentclosure was promptly restored end there ires no effluent release fren contairnont adder, regulatory tiaits. At 1100 on Sept. 1, a temporary repair ms completed idlich reduced the Leak rate to q3prox. 1/4 gpm, effectively Laminating the event. The cause of stud failure has been attributed to (1) I_=king Leakage resulting in wastage due to boric acid corrosion, (2) decrease in Lubricating characteristics and hardening of packing, and (3) the initial thrust required to open the valve. Corrective actions include reduction of the naxiu thrust necessary to open the valve by instaLLation of a modified pecking gland assembly Less suK_)tibLe to Leakageand hardening of packing, lind replacement of packing gland studs with corrosion resistent materiaL. The health and safety of plant personnel Imclthe public u, re not affected by event.

Rlitlt4 Turkey Point 3 10/23185 Node5 (Shoot 1 of 2) At 0915 on 10/23/85, the 3A Residual Heat Removal (RHR) pumpuas declared out of service (OOS) Nhen tt did not met the seal teakoff acceptance criteria during an operability test. At this time, the B olmrgency diesel gl_erstor (EDG) uas OOSfor maintenance. This placed the unit in a condition where t4pontoss of off-site power# no RHRLoopwould be avaiLabLe for core cooling for qNx'ox. 18 hrs. PLant mmnageunt decided since the 3A RHRIXnp could stilt operate and pumpwater to Leave it tined UP to the RHRsystem until the B EDG_ returned to service. This imutd aLLowfor core cooling in the event that off-site pouer was Lost. TS 3.4.1.E re.ires t_o coolant Loopsbe operable end one coolant Loop in operation tdyonever the Reactor CooLantSystem (RCS) temp. is Less than 350 deg Fahrenheit. The 3A RIIRpumpIDal_ _ exceadsd the requirements of this TS. During this event, Unit 3 ms in cold shutdounwith the 3B ItHRpumpproviding core cooling. The A EDGuas operable and the 3A MiStIxlP us tined UP to the RHRsystem to aLtou for core cooling in the event of a Loss of off-site poller until the B EDGtins placed back in service. No heat UPof the' _._or CooLant System _s d:lerved wh']Le the BEDO was OOS.

A-35 p. 1 _ER DATA BASE - RHR5 08/18/f3

PHASE2 PLANT NAJ4E EVENT INITIAL PLANT RECOVERY CATEGORY DATE COHD]Tl ON TI ICE ...... _.._ .... -.--.-.-:.---.- ..... : ...... - ...... RHR5 Arkansas NucLear 1 10126188 Mode 6 23"rain (frnat LER Search): On 10/26i88, with the reactor in a refueling mode, I&C technicians working on a trend recorder in a control race cabinet removed an unlabeled fuse thought to be in Line with the power supply to the recorder. This resulted in a Loss of pe_r to the controllers for both decay heat removal (DHR) cooler outlet valves CV-1428 and CV-1429. Upon toss of power to their controllers the valves went to the closed position, Nhich mm inconsistent with their design failure mode of open. CLosure of both these valves resulted in the inoperabitity of both trains of DHR, due to unavattabit|ty of a flow path through the DHR coolers. The event was terminated approx. 20 minutes tater Nhen control operators questioned the I&C technicians about their work, and the technicians reinserted the fuse. During the event, the reactor coolant temperatures increased Wax. 18F to a final temperature of approx. 87F. No change in reactor btdg radioLog|cat conditions, source range counts or reactor coolant Level was noted. Investigations after the event revealed that the cooler Outlet valves' electric prmJmtic posit|Drier output ti_ had be4m reversed causing both valves to fail closed on Loss of controller Ixx_r. The reversed output Lines were properly rec_nectad and both valves were verified to open on Loss of power.

(from NUREG-1410) Refueling had been completed and the reactor head re-installed. The reactor coolant system water level was et mid-Loop because the reactor coolant _ seal areas and the steam generator mmmys were open. A technician error caused a toss of power to both decay heat ftau control valve controllers, causing the valves to close. Decay heat removal was Lost for 23 minutes and reactor coolant tap. inormmad from 69 dog F to 87 dog F. By design, in order to prevent the unnecessary Loss of core cooling capability, the d_y heat ranovet flow control valves fail open upon Loss of either control _ or control sir. Investigation revealed that the valves failed close becaur,e the air tubing fr_a the valves' eLectr|c-pnstmtic posttioner had been reversed during installation. The industry Has tnforaied of this event in INPO Significant Event Report (SER) 26-89.

RHR5 Arkansas Nuclear 1 12/19/88 zero Ixx_r 12 intn (from LBt Search): On 12/19/88 the decoy heat meal (DHR) system irboerd suction valve (CV-1050) closed rasutt|ng in • Loss of DHR system ft_. Following indicatlcx_that the DHR suction valve was closing, the plant q=eretor fotto_d the appropriate procedures to secure the operating Dmt i:alp. Actions _ere than taken wh|ch returned the DHR system to operation in approx. 12 minutes. At the time of the event, • contract electrician was performing equipment Inspections in the roca _h|ch contains a panel housing the control relays for CV-1050. This individual inadvertently jarred the panel housing the control relays for CV-1050 at approx, the time of this event. The cause of this event has been determined to be inadvertent opening of the normally closed pemissive contacts of a control relay for CV-1050. As determined during the investigation of this event, the pemissive contacts of this relay are sensitive to mechanical shock. As a result of this event, a caution Label has been placed at this control panel to caution against mechanical agitation of the panel. A plant andification wilt be implemented to replace this relay w|th • shader Less sensitive to mechanical shock. AdditionaLLy, other relays of this type wilt be reviewed for possible safety or operational problems due to susceptibility of these relays to mechanical shock.

RHR5 Dearer VaLLey 1 0S!12/82 Node S 2 rain (Fr¢la Seabrook): An attempt to start RHR pump (RH-P-1B) failed due to a circuit breaker racking mc_lan|m probten. Immediately prior to this attempt, the Ix_r to the bus sq=ptying the q)erating RHR (RH-P-IA) had been reaved in accordance with rrocedure TOP 82-27. This resulted in an interruption of Rim fto_ Lasting 2 minutes.

(from AEO0): Failure to start RHR pump due to circuit breaker problem. RHRpump that had been operating Has erroneously secured prior to attempt to startup idle pump (2 rain. loss). A-36 p. 2 LER DATABASE- RHR5 08/18193

PHASE2 PLANTNAHE EVENT INIT|AL PLANT RECOVERY CATEGORY DATE COND]TION TINE q..... ;'"_''''.''''...... "''''''r ...... ''''" "'..... " .... "..'''''......

Itlilt5 CaLvert CLiffs 1 02/08/80 Hode5 1 hr 33 sin (frog NSAC52,p. A-11): A pressure spike during pumpventing caused an RHRSrelief vstve to tift and fail to reseat. (Required relief setpoint as 315 + or - 8 ps|g, but actual relief setpoint leas 285 pstg, so prrosure spike of approx. 35 psig causedevent.) Pressurizer Level decreased frog l&O" to 90" (approx. 1,200 gaLLonsof coolant) in about 2 minutes, tlhen the RHRStins stopped and isolated from the RCS, hatting the toss of coolant. The relief was gagged and RHRSftos restored 1-1/2 hrs Later. An RCP yes jogged during the Loss of RHRSto circulate coolant.

Rlllt5 CaLvert CLiffs 1 05/17/82 2 rain (from AEOD): Spurious opening of breaker from the operating DHRpump(2 rain. toss).

ItHRS Calvert CLiffs 2 01/23/81 Node6 5 sin (frost IISACS2,p. A-24): VaLve position indication m test for an RHRSsuction valve. Questioning the actual valve position, the operator stopped ftotJ teslx)rar|ty to prevent possible pumpclaw. VaLve position indication and floe sere restored.

MRS Connecticut Yankee 03/05/88 Node5 1 hr 2 rain See description of "C_necticut Yankee, 03/10/88 event.

RIIR5 Connecticut Yankee 05110/88 _ 5 1 hr 22 rain (from LEItSearch): On3/10188 at 1659 Nhtte performing s plant heatq: from NodeS (RCSpressure = 309 Pst9, RCStrap = 108 F) in preparation for s reactor coolant system (RCS) hydrostatic pres_ test, ii control operator completing an operating procedure checklist discovered that the residust heat removal (RIUt) system had.been shutdmm for greater than one hour (1 hr, 22 minutes) resulting in a violation of the plant Tech spec. Fottovlng the discovery of this event, a revise of the Operations tog reveated that a similar violation occurred on 3/5/88, during • plant heatup vhen the operating RHRpumpties shutdoun for 1 hr, 2 minutes tdhite StiLt in NodeS. At the tim of the discovery, the plant had already entered Node4 vhera operation of the RHRsystem is not required. Therefore, no ilmediste corrective action to place the RNRsystem beck in service was necessary. These events are the result of m failure to observe a specific precaution |n the plant operating procedure for plant heatt_. ALL Licensed operators have been madeaware of the violation and changes have been madeto the appropriate operating procedures to prevent recurrence. This event is being ,.sported under 10 CFR 50.73(A)(Z)CI)(B) since it involved s cond|tion prohibited by the plant's Tsch Spec.

RHR5 Cook2 12/09/82 Node6 The vest RHRteas in service supplying core cooling st 3000 9_. A reactor coolant high Level slam uas received and invest|gaLlon of equilment shovedthat the west RHRpumpbreaker had tripped.

RHR5 Crystal R|ver 3 07/16/80 Hods5 21 rain (frog MSAC52p., A-12): Gross packing Leak on makeuppumpdischarge cross connect valve ftecess|tsted securing only operable makeuppump,t_hich in turn necessitated securing RCPsdue to Lossof seal injection. Since the RHRSwas not operating, decay heat rogovet ,,as Lost for 21 minutes during interim valve repair. The RHRSwas then restored to operation to conduct final repairs. During those repairs, an improper vaLve tineq)on the RHRSheat exchangers caused s rapid cooLdotmof the RCSend Loss of pressurizer LeveL. _en recovery was attempted by depressur|ztng the ItCS, and providing RliRSsuction from the IWST, injection ftos could not be established due to reverse seating of RUSTcheck valves from RCSpressure. Inventory Loss due to shrinkage caused pressurizer Level to rawin offscate for al:Prox. 45 mtr_es.

A-37 p. 3 LERDATABASE- RHR_ 08118/93

PHASE2 PLANTIINIE EVENT INITIAL PLANT RECOVERY CATEGORY DATE COk'D! Tl ON TINE

...... _ i _ I .,.. ! o e e t...,, _ I _. I o D I....., ...... _.tt ...... t.... t it...... , t... t ...--...... _...... (freI NSACS2,p. A-24) sameas above.

RHR5 Crystal River 3 04/21/81 NodeS SimiLar event 10/01/81 (from NSACS2,p. A-28): Final stages of plant cooLdmm, RHRSinitiated and RCP• secured •t 270 f and 250 pai9; plant then cooled and dqsrassurized over next 15 hrs to 106 F, 50 pai9. During dapreasurization, the injection of auxiliary spray and cceqx_sating coolant tetdmm created a •Low pressurizer outsurge of greater than 360 F water into the _A' hotteg. Further depressurization flashed the hot later in the mA_hot Leg. Pras•urizer Level increased fr_I about 82. to 180m (equating to a void of q:prex. 300 it3). The A Loophot Leg RTDread approx. 300 F, uhich was above the $0 psig saturation temperature.

RHRS Crystal River :5 10/01181 Node 5 See description of CrystaL River 3, 04/21/81 event

RHRS Davis m 1 05/31/80 14ode6 8 Iin (from NSAGS2,p. A-25): Indicated decay heat flow dropped Lowoff-scaLe. The RHRSpumpwas stopped to prevent possible diage. The other RHRSpumpIs not avat table due to maintenance. The pumpus restarted after cause of Lowindication I determined.

(from _): 8-ii_e tI of Dm flow. T_ operating DHR_ I I_ _ I contrei operator. (An I&C mechanic took s 6HRflow meter out of service to perfom surveillance testing. Control ro4I pers4xlneLwere Ire of this. Upon•eaing that the DHR•_tl flay had drqppad allstate, a control roomoperator stopped the pump.)

RHR5 Omvi•m 1 04118/81 Node 5 1 rain (2 rain in AEOD-C) (frm ISAC52, p. A-21): As • resutt of • mail fire in a 745 kV bus, a rapid bs isolation i perforld. A 13.8 kV bus Has not transferred in sufficient time to prevent • toss of easantiat 4160 V power to the operating RHRS_mP- The eme_ diesel started and flow was restored in q:Twox, one minute.

(from AEO0): 2-minute Lossof DHRflow. in response to utwo burning potential devices" on a bus, the bus leas isolated. An error Has made in the sequence of transferring power end isolating the bus. Power was Lost to the operating DHRpump.

RImS FarLey 1 03/07/83 Node6 The A Train RHRsystem I declared inoperable uhen the 1A RHRpumpvia inadvertently secured.

ItHRS FarLey 1 11107/86 NodeS A ResiduaL Heat Removal(RHR) loop suction pressure relief valve opened to reduce Reactor CooLant System (lieS) pressure. On 11/7/86, the RCS(in the solid condition) pressure I being increased to 400 psig prior to starting A Reactor Coolant P_p (RIP). The operator increased pressure too rapidly 8rid I unable to stop the increase prior to the lI RHllLoop suction pressure relief valve opening. The opening of the relief valve contreLLed _ reduced the RCSpressure. On 11/15/86, the RCSI being Iintainecl in the solid condition at approximately 400 paig with the 1B and 1C RCPSrunning. Depending on plant conditicms, RCSpressure vhiLe soLid can either increase or decrease vhen starting RIP. The operating crew had anticipated a pressure decrease; however, pressure increased idlen the 1A RIP was started. The operator tried to Limit the pressure increase but the 1A RHRLoopsuction A-38 p. 4 .L.._RDATABASE- RHR5 08/18/93

PHASE2 PLANTNANE EVENT INITIAL PLANT RECOVERY CATEGORY DATE CONDT;ION TIIIE •.... o.,..._, oo-._...., ._oo...... -,.._o. -.._. _- -- °,-:. o°,-...... pressure relief vatve opened. The opening of the relief valve controLLed 8rid reck_ed RCSpressure.

RHRS FarLey 1 11115/86 Hods5 See descriptim of Fartey 1, 11/07/86 event.

RI_.5 Ginna 05101183 Node5 (free Sealx-ook): The RWSTrevel had decreased to 20_ revel _hich required by procedure to stop one RHR Ixnp. The A RNRIxnP Igas st_ed and flow drq_ to zero. Control Roomqperator noticed NOV-704B(B RHRPumpsuction valve) closed. The A RHRpumpwas restarted and flow reutabtished and the B RHItpump stopped. Thus for a period of tess than two hours Idhite fiLLing the reactor cav|ty the B RHRIxnp us run vtth its sucttm valve closed. The auxiliary operator checked B RIIRIxnp end found it Imm but no mt Leakage. The IXllp mm tested for ftou end vibration with ¢enditione found normt.

(from AEOD): FiLLing reactor refueling cavity - tou RL_r. Secured "A" Rlilt_ - Suction valve on operating °°ILKIx.nip _ ¢losod. (Duration of event urlkllOl_)

RImS Indian Point 3 0S/12/83 _ 5 A roLL teak ms identified on the RHRminifto, at the Netd joint be|wen valve 1870 and Line 337.

RHRS Nstne ysnkea 06/10/81 Hods6 S sin (from NSACS2,p. A-19): RImS flow ms interrq_ted uhon an outboard stop valve closed. VaLve reqxmed and flow r_establtfJ_ed in five minutes.

IWRS Iqt t Is tone 3 06/08/87 Iqode 5 33 sin (from LER Search): On June 8, 1987 at 1345 hrs and O_ pmder, 8n inadvartont mmdeehlmge fren node 5 (ootd shutdmm) to Hode & (hot a_hutclmmo) ccurred. The Incident occwred during an lrnmstiption by plant pereor_et to determine uhy the train S resiclua| heat renmmt systemsuction tsotatim (fren the reactor coolant system) verve did not stroke during the previous ptmt coot-dram. The root reuse of the inadvertent node change ms operator error. The reactor spare|or faired to monitor reactor cootent system temperature for a period of 33 minutes and did not identify the possibit|ty of a Ixle changedue to the isotation of the heat s|nk for the reactor coolant system. The entry into Node& occurred before the _iate plant technical specifications for the modechangewere satisfied. After the |ncidont ms identified, the immediate operator action ms to return the plant to Node 5 (coLd shutdmm) beto_l the naxiu temperature requirements. As corrective action, qperetore have been briefed of the incident and have been reminded of the importance of monitoring plant respmwes during evolutions. The operator directty immtvtd in the incident has been relieved of License duties until demonstrated def|ciencies in performnce can be resolved by ptant namngment.

RHRS North Arm 1 06/01/80 Mode5 34 sin (from IISAC52, p. A-12 and p. A-24): To check the size of the pecking gland on an RIIRSvalve, the packing gland nut _ms Loosened. The mchanic was not aware that RCSpressure ires being increased. The pecking cane Loose, dislodging the packing nut, and Leaking primary rater into contairment. To reduce RIIRSpressure, the operating RHRSpumpwas secured for 34 nirv.rtes end the RCSpressure reduced uhite the pecking nut was temporarily instaLLed. One RilRSpumpwas then operated for 5& minutes. The Ixanp wos then secured again to ropack the valve. Neither RCPsor RHRSIxam_ were operated for 1 hour amd&3 minutes _hite repacking the vaLve.

A-39 p. 5 I,ERDATABASE- RHRS_ 08/18/93

PIIASE2 PLANTIL4J_ EVENT INITIAL PLANT RECOVERY CATEGORY DATE CONDITION TINE

ItlIR5 North Anna 1 07./18183 Node5 5 mih (fran SeabrooL): On 2/18183, indications of cavttstion Mere observed on B Residual Neat Rmovet (RHR) pumpand Later on A RHRIXmp. The B pumpwas started and returned to service in q:proximtety 5 minutes. An RHRpumpq:mrabItity was subsequently verified.

(frm AEO0): bth RHRixaq=sMere cavitating. Causenot determined (5 rain. Loss).

RiOt5 oconee 1 06/29/81 NodeS (fr_ IISACS2,p. A-29): Final stages of plant cootdo_n: Rims initiated and Last RCPsecured at 225 F, 310 ps|9. Pressurizer Level was then reduced fr_ 250 to 100 inches to fulfill an omitted requ|rMlnt (procedure required t_erlng pressurizer Level to 100" before securing RCPs). This RCS tetd_n moved 4,000-5,000 gallons of stagnant 423 F water frm the pr_ssurizer to the aA" hetteg. The RCSpreamure usS then Lowered. After about 25 hrs of pressurizer cooLdewn, pressure mw tm_d below saturation for the hottest water in the SA" Loop (120 pslg, 350 F). A steam bubble of eppr_. 300 (m. ft. ms f_md |n the top of the SAnLoopaj. Leg. A rapid pressurizer Irma1, occurred as the 300 cu. ft void _ formed.

JtJP.5 Oconea 3 04106187 Hede S (fr_ LERSearch): On April 6, 1987 at approx. 1500 hrs the reactor c_tant (RC) system on Unit 3 Mes cooled to the extant that technical specification Limits in Table 3.1-2 Mere exceeded. Unit 3 ms at cold _utdmm end in prep.ration for stsrtup fall,lag 8 refueling sutage. The |nadvertont ¢_ldmln Mesdue to reecter coolant Leaking through the outlet valve (3LP-12) of 3A to pressure injection cooler uhlte the Matt aide of the cooler m being flushed uith tow pressure service tater. The Leakagethrough 3LP-12 and subsequent c_tdmm occurred When3LP-11 ms opened in the prmm of performing an equipment restoration procedure. The root cause of the event was that the procedure for perf_ming iintenance on 3LP-12 did not provide adequate guidance to direct peMmnnet on h_ to adjust the closing Limit suitch on valve qperators Whichare cont_ttad by a Limit s_itch on the closing stroke. Ira.diet. corrective actions were taken to isolate the Lad pressure injection cooler and secure the RC ft_ by closing valve 3LP-9 I_p6treJl Of the cooler. The teJtng valve 3LP-12 was then rurally closed. Linear elastic fracture mechanicsanalysis of the rest Limiting reactor vessel betttine metal indicated that this event had no effect on the integrity of the pressure vessel.

Rlgt5 Palisades 07/18/81 Node5 1 hr 30 rain (frm NSACS2,p. A-19): A shutdmm cooling heat exchanger outlet verve fatted closed, causing Loss of RNRSflow. Closure of this air-operated valve isolated ft, from both RIIRSheat exchangers to the RCS. Primary coolant temperature increased fr_ appr_. 123 F to 197 F.

RHP.5 RanchoSeco 10/03/86 Node5 13 rain (fr_ C_015): PoMer Level - OZ. I_ite in cold shutdownon Oct. 3, 1986, during irlst_t & control investigation of el:natal indication on panel ll2SFBfor decay heat system (oNe) hE" r_m sumpstack tights, SFAS"P bistabtes tripped causing llV-20002 to close, Nhich trilTml DHS"8" pump. The plant was uith(x_t the use of the nor'matDHSfor approx. 13 minutes. Due to the extended period that the plant has been shut do.n, there us a mall+ but detectable increase of reactor coolant temperature. Steps Mere taken imedtatety to restore a DHS train to service in accordanceutth the intent of Tech Spec 3.1.1.5. This event is reportabte according to 10 CFRPart 50.73(a)(2)(IV & V). The cause of the incidents ms IlK: techn|cians troubleshooting dmorml ind|cat|on on panel ii2SFBfor OHS"8" pumpr_m (east) s_p stack Lights (18" Level indication) on panel lt2SFD. The tmmdiate cause of the spurious actuation ms an electric arc from the susp Level stack Light shan "rolling-over _ the respective bulb. The arc initiated the trip of inverter aS". As s tong tern corrective action, the DC vital peMer sq_ties uitt be modified to be e_l|pped with static transfer switches.

AM p. 6 LERDATABASE-RHR$ 08118193

PIIASE2 PLAMTNAME EVENT INITIAL PLANT RECOVERY CATEGORY DATE CON1TD 1ON T!ME ..,_...... ,...... ?.... _...,_,...... -,-.:-...... -- ......

RHR5 Robinson 06111187 zero po_er (from LERSearch)= On June 11, 1987, during zero pewer physics testing foLtouing a refueling outage, a valve packing failure resulted in primary system Leakageof approx. 23 gaLLonspar minute. An unusual event ms declared, initial indications were that the Leak was from the #1 seaL of NC" reactor coolant Ixmp. Eased on an erratic standpipe Level and high temperature in the Mat teskoff Line, the pumpwas stqw_. Later, similar indications were noted for the "8" reactor coolant pumpand it too stqppad. Subsequent invest|gaLlon of the Leakagedetermined the source of the Leak to be res|duat heat removal valve RHR-?SOinside containmant, based on high teq)erature |n its Leaker| Lines. Comuon piping is shared by this Line and the pumpseal toskoff Lines. The packing was replaced, the unusual event temineted, end plant stirrup rein|Listed. The need for additional Long-tern corrective action ts being evaluated.

RNR5 SaLem2 06/12/81 Mode4 1 hr $ rain (from NSACS2,p. A-13): During plant cooldoen, the RHRSsuction relief valve lifted. RimsFxaq_ were secured and the ItHRSsuction valves from the RCSsere closed to halt the toss of inventory. The RCS clepr/maNrizat|on transient hies terminated 1 hour 5 minutes after the relief valve actuation. The RHRS vas returned to service and RCSpressure increased to 335 psig.

Rims Satan 2 05124/83 Node5 (from seabrook Sheet S of 11): No. "21 Residual Heat Removal (RHR) pumparid No. 21 Fuel HandLingStdg exhaust fan were observed to trip. Deerwrgization of No. 21 RHRpumpresulted in no RHRLoopbeing in qperstton and Action Statment 3.4.1.4B was entered. The pumpus lamed|sLaty restarted and fLou restored. No reduction in Reactor CooLant System Boron concentration occurred with the RHRLoop inoperable.

(frm AEQD): RHRpumptrip caused by tngtc/ctrcuitry problem on the usafnguards equtpmmt controL" (SE¢) system. (Duration of event unknowl)

RHR5 San One|re 2 03/14/82 Mode6 90 a|n (Sheet 1 of 5) Shutdoun coating uas Lost due to nitrogen intrusion as a result of beckftushing a fitter in the purification system. Shutdcun cooling flow uas Lost for 90 minutes. PubLic safety ues not endangered because no irradiated fuel was in the core.

RHR5 ,_ 11106184 7 Bin (from AEG)): A procedural error in testing relays on the bus supplying the DHRpuup caused the bus to strip. The associated diesel was out for maintenance (7 rain. Loss).

RHR§ _,er'ry 1 03118189 zero Ix_er 11 hr 28 rain (from LERSearch): On 3/18/89, at 0912 hrs, it was discovered that the manualRHRand c_ponent cooling (CC) valves to the UAURHRheat exchanger (HX) were open, but the mnuaL RHRoutlet valve of the non HXuas closed. In addition, the CC isolation trip valve TC-CC-IOgAtansclosed thus prohibiting CCflea through the "An RHRllX. ConsequentLy,with 11oCC ftc_ through the nAn HX and no RHRftcu through the "B" HX, there was no RCScooling Loop in operation. This condition had existed since 21_ hrs, March 17, 1989 and is contrary to Tech Spec3.1.A.l.D.2. The CC con_irment |soLar|on trip valve for the UAURHRHX ms reopened, restoring CC flea to the HXw Laminating the toss of RHRcooling and the RCSheat up. Fotto_ing this discovery, a root cause analysis uas parfonMd by the personnel involved. The cause of the event ms operator error. The operators were under the incorrect assumption that the roB"Rim HX outlet manual isolation valve was in the openposition. In addition, the CCtrip valves were uwliputated without use of a procedure. Other contributing factors sere also identified. Persoru_t associated uith the event were disciplined. The numberof off-shift Licensed A-41 p. 7 I,|R DATA 8A$| - RHRS 08118/93

PHASE2 PLANT NAHE EVENT INITIAL PLANT RECOVERY" CATEGORY DATE _ 1T!ON T1141[

operators penattted to assume |tcensed positions on the same shift has been |talt_:l. In addition, off-shift operators mmt be retteved by an on-shift operator.

RHR5 Surry 2 02119186 Hods 5 10 sin (fr_xl CR5015): Power Level - _. On 2-19-86 with Unit 2 at cold shutdotm, operators were performing a test of the c:eNxm_t cooling (CC) check valves in the residual heat removal (RHIt) system. During this teat_ an qperator made an incorrect vatva ||noup which resulted in the isolation of CC ftou to the °AS RIIR hast exchanger lind RHIt ftow to the *80 RHRheat exchanger for approx. 10 minutes. During this period, RC8 temperature and pressure were ctosety monitored and no alx_ormt increases tmre noted. The root cause of this event was human error in that the operator felled to fottow the steps in the m'itten procedure _hich woutd have ermured the proper vatve ttnoq:. A contributing factor was poor camunicatlen between the controt rcmm operator and the operator performing the valve ttnaqp. The operators |nvotved tn this event prepared a report describing the circumstances which ted to this error and tt wiLL be placed in the operator,s required reading mnuat. This event .tit also be evaluated by the huam performm_-e evaluation coordinator.

A-42 p. 1 LERDATABASE- RHR6 08/27/93

PHASE2 PLANTNAHI EVENT |IIXIIAL PLANT REOOVIRY CATEGORY DATE CONDITION TIH! .,-,,,----,,--,---,..---.---.---...----.-.---....--...... e. Bm v.tt.y 1 "Os/al (from I$ACS2, p. A-26)s In the process of putting the HItS in operation, one of the HIS s=tion valves would not open, prectuding use of the IltS. The valve ms 'l_Wuontiy openedmmuaLty.

1116 Crystal River $ 03/0&/79 Mode4 (from IIACS2, p. A-26)8 Outing RCS¢ootdoun, both series HIS suction valves could not be openedfr_ the control ram. Both valves are Inmhle cents|rarer and _ere openedby hand to permit Hit| ties.

Ila6 Crystal Itllir 3 04/25/79 Mode4 (free ItIACSI, p. A-;_)| An W lilt|on villi failed to qxm remtety during plant eootclmm, due to lir|oll pr4blems vlth the vitvi0S motor operator. The villi ms repsirad, tested, lind restored to service tn 13-1/2 hrs.

Rill Davis _ 1 0110T/81 Node& 15 -,in (from HACS2, p. A-21)s A controL ram operator mttmpted to start a decay heat pap as a natal step tn establishing IIHI ftou during plant _otdmm. The pap did not start, end the ptmpbreaker mm reded out for impecttcm. No prebteml ire found; the breaker In ricked in, end the PUll) started aummfutty.

(fma N_): ONEpap failed to start clueto m breaker prebti. ELectricians were able to restart the IMp after • 15 minute cletw.

nm 61nno 03/o3/8& H_de & WhiLe cootlq dram the hector Cootlno lystem (RCS) to cold ahutdom condition far the annual refueling end iintonsnce _ lleri_d_test: IVt-l.4.1e,otWrofuatlngmotor operated valve =rlilitm (RIIR system- 790 valves)m in _. MOV-700_ LoopA Residual.Neat Removalsu©tlonstopvalve) filledto stroketo the I pNIIIOn dm _ from the ControlIs.

IHit6 Glnne 05111/84 Mode4 On S/14/84, uhtte cootlnlD clounthe Reactor Cootant System (RCS) to the cold shutdouncondition for sludge Lancing lind crevice ctesnlng, NOV-700 (RCSLoop A Residual Heat Removit Suction VaLve) fasted to stroke to the open pes|t|on dim actuated frm the Control Roan. FoLtouing emnuatumesting of the valve, mintermn_ persormet perfonmcl am trmpection of the valve exterior. This inqx_tion revealed that the pecking BLand fLange had shifted out of the vertical pceitlon to x point Nhera the flange ues tn c=tect uith the valve stem. This couLdhave caused • mchenicaL binding tn the stem and torqua-eut of the valve operator. The valve was then strdad mnually to verify no mchenicat binding. The verve ms then stroked toice electrically. The valve functioned satisfactorily with preper motor current readings and acceptable opening and closing tim indicating no Ichenlcat binding. A visual Inspection of the valve stem and etm threads verified adequate cleanliness and lubrication. Torque uitch settings ire verified uithin the mnufecturara design settings. On 5/22/84 Nhonthe IlCS was heatingup to hot shutdmm, the valve ms Gin stroked to verify proper operation. Again, the valve functioned prcqx_ty Hith proper motor current readings and acceptable opening and closing tim. Operation of this valve Hilt continua to he monttorad during the next cooldoun of the RCS.

ltt6 Ocenes 2 09118/81 Hods1 (from NSACS2,p. A-27): Prtmry to secondary teskage Is detected indicatlng• OTSGtube te_k. The unit ass shut dam, cooled dam lad depra, urlation initiated to reduce the Leek rate. Qhenplant I:cxdtttorm perlttted, RHRSqperatim I attempted. A RHRSsuction valve felled to openon demand. A reactor bldg entry Is trade to open the valve manually. After three ur_cce, fuL ettlpts to open the A-43 p. | LEROATAUSE ° i_ 08/27/93

PtlJUIE2 PLANTMANE Et_IIT INITIAL PLANT REt:OVERY CATEOOItY DATE CONOITION TIME

_eee e_e QIpwee e e Q_a,m o e.J I_.e, _._ fre--e _Qeoo ul e el..--..,l_, e _ ..I m _.--. _ ella.- _ Q • o o ,. e e o o e _ _ _--_e,_ ,.e o e- _ _ Q o .eeoe m .e _ valve mmuaLLy, the entire valve 0peretor ties reaved _ the valve _ uith IwluaL hoists. ItHItS operstlen ws then established, ItCIqtesctred, snd the RCSclopressurlzed to stop the OTSOtube Leak.

Itl_ Ooanso 3 10/15/85 Mode& (item I_mbrook Im4et 3 of &) An umuccasofut sttqlqDt to qxm an electric rotor operated (El0) valve (3LP-2) ms rode frm the Unit 3 Carrel Roan. Unit 3 m in hot shutdmm after cooing off-tin, for m|ntm. The veLva Is required to open in order to indicate the decay heat remover cooLtng node.

(frm C_015): Pear Level - 0S. On 10-15-85 et 0955 hrs_ en unsuccessful ettmpt to qxm en electric eater operated valve ms rode from the Unit 3 control room. Unit 3 was in hot shutdownafter caning off-Line for Im|ntenm_. The vaLve Is required to q:en in order to initiate the decW heat removal oooL|ll0 node. The rAMIe of the |r_|dmlt ties the torque I_itch settings on the valve. Rotork nuctesr Imitator sottin0s wre not sot high mo_h to operate the valve under systm premm. The U0 valve torque so|tch settings mr. not qDsolfied in the design nodiftcatien packagemad to PopLecothe valve actuator utth mmu Iotork nuclear rotor. The sorr_ttva sotlen m to open the valve frm the valve rotor .eats©to at the rotor carrot re•tore bypassing the valve ectmtor0s torque ••itch Limit control ©trcult. The enetysla supporting the Licensing basi• for Sconesdoes not require the tmedlmte qxmlng of this valve. The failure to tiloteLy qxm this valve only rmmtta in • delay in the initiation of'_¢aff hut rmmat cooling rode.

PJ_ Rablneon 07/15/82 _do 6 (Sheet 1 of &) On both occulms, motor operated volvo, RB-?59A, a residual hoot removal exchanger discharge valve, failed to qxm.

WilM Rablr_ 0T/27/82 "Node6 (Sims 1 of 4) an"both oc_icm, motoroperatedvalve, tm-?59A,mrestclu,L hut rut ex_mW dla_hea_ valve, felLed to open.

RNIM San Onofro 2 0&/_/83 Node& (Sheet 1 of 4) Prier.tiara in prngra_ for Hodo3 entry, Ib_tdmm CooLing System (SO(S) heat exchanger isolation valves 2W8150, 211V8152and 2W8153 could not be rmototy operated frm the Control Itoau UlXm initiation of shutdotm cooling to •void personnel radiation exposure frm Local operation.

RH6 Turkey Point 4 05/23/89 lkxk 4 (irma LERSesrch): On 5123189, at 1820 EST, ulth Unit 4 in Node4 and the reactor coolant tmperture betotd 3SOdegrees, m|ckmt hilt ronovaL (RHR) motor operated valve NOV-4-?51, l_ normt suction ieotutlon valve, f•lLed to open ulth the control suitch. This valve mm to be opened fr_u the aLternate ah_tdmm put to place RHRin service for its sorrel heat rosovat fur_t|o_ N part of m operability chock of the paneL. Whenthis valve f•tLed to opm, the reactor continued to be cooled by renovlng heat through the •tom generators. The valve m manuaLLyjoyed from it• •eat. The valve I_booqumtLy stroked umothLy utth the control suitch. The root cause of this event is hydraulic Locking of the valve. To prevent recurrence of thl• prabtea, the valve has been a0dlfled ulth an equalizing Line vhlch uttt assure that the pressure in the bonnet of the valve is at an equal or to_er pressure than the high pro.sure •ida of the valve. AdditionalLy, H0V-4-?50, the RIIR vaLve Immdi•teLy upstream of valve HOV-4-1_1, has been nodifled. The cart, spading Unit 3 valves uttt •Lso he nndlfled to esstre that these valves do not experience • hydraulic Looking event. The Unit 3 valves, HAY-3-750 end NOV-3-?51, ,iLL be modified prior to the Untt0s rsturn to peter.

A-44 Appendix A.2.2 Loss of 4 kV Bus (4 kV, 4 kV/P6) p. 1 LEIt DATAIMH - 4KV 08/27/93

2 PLAIITM EVENT INZTIAL PLAffT RECOVERY CAT|q_ItY DATII COHI)! T|(_ TIHE

• t m IIW!e e e dl .tme e el m e e g e_e at e e _eeeQee eee ee _meee o e m e e eae_e-see eee e • e else e e e_eee eels e mid oQ e e see e e e eee e aoe • eee e e mace ._eo /d(V Arkansas NucLear 1 1_05/_19 CIO (from IJIR Search)z On 12/5/89 •t 0645 and 12/6/89 st 2205, vhiLe the plant tms shard•to in • oslntananos outage, autmmtlc actuations of an emergencydiesel 9aneretor (E_G) occurred as • result of rosa of power to t 480 voLt (V) engineered eafetNm_ (liB) bus. Prior to both events, the 115and $6 /AOV E• bumm uere oroasosnnocted to facilitate mlntenance activltlas. The 0ec. 5 event ac_recl as I result of a personnel error vhtch occurred vhlte operators were attempting to msptltout" the 115and 06 blmses and return the III pmmr dtstrlimtlon system tlneqD to normL. The error resulted in s Loss of poller to ks 06 vhtch caused the offslta feeder breaker for 4.16 kiLovoLt bus _ to open and Initiated I start of the 'ii' EDGuhlch tied on to the J_ bus. The Dec. 6 event was aLso the result of • portwlnel error uhlch caused • lo_ of pemr to/ANN E$ Bus 115. Thl• ¢omlltton caused the offstte feeder breaker for A3 to trip end the ,A' EDQto start. The mmntary Loss of power to Jr3caused the qporatlng dicey heat remmmt (DIR) IUlP to trip. DNItfLou was Lost for q:prox. 9 minutes end resuLted tn a reactor cooLant slmtm tip. Incrs_e of 17 degrees. Nanqemnt briefings uore conducted for the •paroLing cram prtor to restart fro the outage ceverlng the Lessons teamed frea these events.

4_(V Arlumsea i_Jctear 1 12/06/89 CSO 9 sin See dmorfptton of Arkm Uuctlr 1, 1W0S/89 event.

44W Arkansas NucLosr 2 11/14/89 Shutdovn 1 ain (from I.EIt mrch)l On 11114/W, imtntenenoe personnoLinitiated s post maintenance test, using instructions in a lItnterance job older, to siluLste an undervoLtnge on a 480 Vac engineered safety featwes (ELF) motor control center (;IS) by pLac|ng • juuper across the 2115undervotta9e relay c4ntacta, ltmedlataLy foLtov|ng thl8 step, the normL offelta pmmr feeder breakor to the associated t160 Vac ESFbus (2A3) unexpectedly qxm4d resulting in the Loss of payor to ?.43. The eLectricaL bus dNnerglzed as designed. The test steps provided In the job order did not identify that 2k3 wouLd dNfterg|ze as pert of the test. _ _ usa detmergized, a Lay pressure safety in••orion (LPSI) plop, uhteh t Sul_Lytng fLeu for dmmy hit remmmLend s asrvlce uator IxmP dmmerglzed. A standby LPS! ImP _ from the redurdmt 4160 Vac ESlFeLactrlceL bus ms started tn epproxleataty one minute end fLou reeatobLIshed. Since the plant had been shutdounfor eavoraL days prior to this event, the reactor decay heat LeveL• uero Lou and the momentaryinterruption of fLotddid not resuLt in any signlf|cant reactor coolant system telR)eritura or pressure Increases. The test wss reevaluated and satisfactorily collpteted. The root cause of this event was detemined to be inadequate post mint•nonce test controls,

4XV $eaver VaLLey 1 07/03/79 Hode$ or 6 21 sin (fro NSACS2,p. A-20): I_ite stdtching a vital bus to the alternate source, contatrment isolation phase B (high-high cont•iment pressure) actuation occurred. This tripped both emergency& kV stub bus breakers resulting in Loss of the operst|ng RiOtSpuap.

/d(V Catotlba 1 01/07/89 Node § (frml LEII Search): On I/7/89 at approx. 0302 hrs, 6900V tie breaker ITC-7 tripped after reactor coolant (NC) I_P 1C had been started and power to 4160 essential bus 1ETAwas Lost. The Loss of power caused an i•otmtion of the bus utth no back-up pmmr available due to diesel generator 1Abeing out of service. The blackout resulted in a Loss of lever to residual heat removaL_ 1A, fueL pooL cooling 1A, and _t cooling Imp 1A2. SubsequentLy,due to a charging flow control valve fsitlng qxln, NCsystem pressure increased and caused a pressurizer PORVto Lift 7 tim. The 6900V sw|tchgur 1TC is ••per•ted into two sides vhlch are connectedby normlty open tie breaker ;1TC'7. The tie breaker ms closed because1T2A was out of service. The tie breaker tripped •pen on ovarcurrent because• ground overcurrent relay we• instaLLed in the time delay overcurrsnt relay Location. The ground overcurrsnt relay was not des|gnarl for the inrush current caused by starting the lIC I:uqp. The A-46 p. 2 LERDATABASE- 4KV 08/27/93

PtlASE2 PLANTNAME EVENT INITIAL PLANT RECOVERY CATEGORY DATE CONDi YION TIHIE

• .m m mm o moe, mm,_m Q o mm mm w..,m, a o o e e. a m_ m o oo*mmwoe m.m ._.oe m,. m_... *. e.D_ .mm. _ o o...... , am... o e mo ..m .. o. m...... , o.... m. unit ms tn NodeS, cold shutdovn, vhan the incident occurred and had operated in all modesof operation. This |ncldent Is attributed to an Irmpproprtate action. It el:pears the re|ays were swapped during the Initial Instal|allan during 1978. Tim delay and ground overcurrant relays were placed into the correct Locations. An inspection yes performd to verify the correct roLays Word installed in all other similar applications.

&KV Crystal River 3 08/28/89 Node5 19 gin (from LERSearch); Crystal River Unit 3 ua$ in 14ode5 (cold shutdotm) to perfom repairs to a nuclear services rim tater posp with the "B, decay heat train (OH) reaved fros service. The mA"DH train was operating and the reactor coolant system m fitted and intact with an apersbto steam generator. At 023? on Au. 28, 1989, the mA. 480 volt engineered safeguards (ES) stepdotmtransformer faulted causing ON ctoemdcycto cooling pumpmA" (OCP-IA) to de-energize. Loss of DC:P-1Aremovedthe cooling for DN train mAa. rendoring this train tnopersbte. At 0256, OHtrain aB" Innd its 8qqxwt systm were started. The UA" and "Be &80 volt ES buses were than crou-tied to provide AC power to the mArebus. The tOSSof OHcooling caused an increase In RCStosporature and an tnsdvertant entry into Node4 (hot Idmtdmm). Tho transformr has boen replaced. The transformr faulted due to insular|on degradation calmedby aging. Burtng tho event. OpMltOrs attospted to operate the roB"train OHsuction Isolation valve (DIN-&O) frm the control ram. This valve failed because the nut securing the Linkage betvmen the notor operator and the valve stem had backed off from normal operation. DHV-40 has been repaired Ind the nut (in tho Linkage his bean iltlked to _ it finn backing off. Tho redundmt valve. DlIV-39, _11Lt be staked.

Ld(V Ft. Calhoun 12/14/85 Refueling shut. 15 rain (frm CRS015): Parer Level - OZ. On 12-14-85 at 1010 hrs whlLo in a refueling shutdovn, a OCbus and 2 AC instrtamnt bum wre Lost as well as all essential &80Vbuses. The Loss of pouer initiated eafeDumrdStDnots: prmmur|zor premmre Low$19mt. asfaty tnjact|on actuation signal, c0rttaiNMnt isolation ectmtton signal, and ventilation isolation acttmtlan signal. ALso Lost due to the power failure tmre shutcimmcooling, ocqx_sead air, turb|ne plant cooling rater and sore control rein indications. The pouer fat Lure occurred due to the personnel error and an altered electrical distribution systaa Lineup due to tasting, IMIntanmlce, and ned|ficatton work. The technician tnadvartentty tr|l_ed the relay controlling the breaker that vas providing the 161 kv power to the &160v 1A4 safeguards bus and _htch in turn po_r all L_80Vbuses including the battery charers. Yith the t_s of tho bettary chargers, DCbus 112becameinaperabLe bacatme battery 42 ms d|aconnected for Imlntensnce. ALso, AC instrument buses O and D were tnopersbte N they are powered from 0C bus 42. This resulted in a partial loss of control roam indications. Corrective action Included restoring pmmr to tha _ busest,|thln 15 minutes. A meting was held with the individuals involved before all_tng thln to return to their tasting.

&KV NcGurte 1 11/29/88 zero po_r (frm LERSearch): On 11/2/)/88, operations was restoring the 1A I=usLine to service and noticed that standby breaker 1TD-6, Nhlch supplies swttchgesr 9roqp 1TDof the 6900V bus, required excessive pressure to rack in the connect position. To ensure there were no problems with the breaker, OPS de¢tded to test the breaker by cycling it. An OPSmrvtsor instructed an operator to rack the breskor to the test position, cycle It, and then restore it to the connect position. Utth the breaker in test and tho control board svttch in Imnuat, the operator closed the standby breaker at 2112 from the control loon. The normL breaker opened amdesi9ned and lover tms lost to sw|tchear group 111) resulting in a train B blackout. Residual heat resovat (lID) pump18 auto tripped because it is not a blackout Load. The diesel generator auto started and Loadedas required. Hormal power was restored and 14)pump1S was restarted to restore core coDlin. The reactor coolant tamp increased by 4F durtn9 A4T p. 3 .LERDATAeASE- 4W 08/27/93

PHASE2 PLANTNAME EVENT INITIAL PLANT RECOVERY CATEGORY DATE COliDITION TIME ,...... the time that NOpump10 was shut dour. This event is assigned a cause of managementdefi'ci_lcy the OPSsupervisor did not provide adequate written and/or verbal instructions to the operator for correctly testing standby breaker 1TO-6, and a contributory cause'of defective procedure because procedure OP/1/AJ6350/08, operation of station breakers, did not contain precautions prior to testing breakers.

4iCV Millstone 2 12109181 Mode 5 30-60 min (fram NSAC52,p. A-22): Yith the Rt:Sdepressurized and drained to the hot Leg midpoint, the running RHRSpumptripped due to 345 kV switchyard breaker testing. Core heatup from 90 F to over 208 F occurred over a period of about 30-60 minutes. Whenthe RNRSIxaapwas restarted, previously colder, stagnant IUINStemperatures |ncreasod from 90 F to 208 F in about 2 minutes. The second RHRSpumpwas started, and the hot coolant in the core area and the colder coolant in the RCSand RHRScontinued to equalize, resulting in an RNRStemperature drop to 130 F in about 7 minutes.

/dCV North Anna1 03/23/89 Mode 5 10 sac (frgm LEIt Search): At 1053 hrs on 3/23/89, uith Unit 1 in Mode5 (coLd shutcloun), the UlH" emergency bus urns|nodvertentty de-energized for approx. 10 seconds during the perforumnce of 1-PT-82.3A, 1H diesel generator test (stmJm.ated Loss of offsite I=o_mrin conjunction uith am ESF actuation signal). This is considered am inIdvertent engineered safeguards feature (ESF) actuation and is reportable pursumt to t0 CFR50.73(A)(2)(IV). A four hour report mls made in accordance with 10 CFR Srl.?2(A)(2)(II). Immediate corrective actions included automatic start of the "IP camponantcooling uater Ixsq), the manual start of the ulp residual heat rlmovat pump, and verif;.catton that de-energized equipment was e|ther returned to service or ues reset. Further corrective actions include investigating and correcting the circuit design of the trip block relay, in addition, appropriate emergencydiesel generator periodic tests uitt be revised to prevent testing _hen in an aix_rmet electrical ccmf|guretion. This event posedno significant safety implications because the alg" emergencydiesel generator (LO6) started as designed and reloaded the "IH" mergen_ bus in Less than 10 _. In addition, the UlJa ED6 WaSoperable and was capable of providing Ixmer to required systems _ the SlH" emergencybus uas inadvertently de-energized.

4ICY North Anna1 04116/89 Mode5 (fram LERSearch): On April 16, 1989 at 1115 hrs, with both units in Mode 5 (cold shutde_n) the noramt power sqpply to 1H and 2J &160V emergencybus _as Lost _hen a Lifted wire was inadvertently grounded during removal for switchyard modifications required for design changepackage (DCP) 88-05. The 1H and 2J emergencydiesel generators aut_atic_ty started on the toss of poser and re-energized their amsociated emergen_ bus. This even_, is an engineered safety feature (ESF) actuation and reportable pursumt to 10 CFR50.73(A)(2)(IV). A four hour report was madepursuant to 10 CFR50.72(B)(2)(11). In con_IJrlclcionwith the toss of pouer the Unit 1 residual heat removal (RHR) pump, 1-RH-P-1A, tripped on undervoLtage mn_ |s reportable pursuant to 10 CFR50.73(A)(2)(V)(8). Abnormal procedure 10 was initiated and 1H and 2J emergency diesel generators (LOG) were verified in operation and supplying their respective bus. Abhorrer procedure (diP) 11.2 was also initiated and residual heat removal capability subsequently restored on Unit 1 by starting .p RHRpump. No significant safety consequencesoccurred due to this event, since the redundant emergencybus remained available to supply to required plant equipment. ALso, the 1H and 2J LOGsauto started and re-energized the emergencybusses as designed.

41CV North Anna2 0L_i16/89 Mode5 See description of North Anna 1, 0/+116/89 event.

A-48 p. 4 LERDATABASE- 41_V 08127193

PHASE2 PLANTNAME EVENT INITIAL PLANT RECOVERY CATEGORY DATE COND1TI ON TIME •.: ...... n...... '."°" _°'." T'."'. "".''.''...... -...... "...... 4KV RanchoSeco 12/08/86 Mode 5" (from Seal:rook Sheet 20 of 20) The plant ms in cold shutdownremoving decay heat via the Decay Heat Relovat System (DHS) Train B on Dec. 8, 1986. Stertup transformer @1was scheduled for routine preventive maintenance. A toss of the 4A bus power attendant diesel generator start and decay heat system (DHS) isolation occurred during the transfer of the source transformer at 2:18 PMon Dec. 8, 1966.

(from Clt5015): (Rake E.IO TabLeA.4) Pouer Level - 0_. The plant uas in cold shutcloen, removing decay heat via the decay heat removal system (DHS) train "B" on Dec. 8e 1986. Startup transformer No. 1 ms scheduled for routine preventive Imtntenance. A toss of the 4A bus pouer, attendant diesel generator start, and decay heat systaR (DHS) isolation occurred during the transfer of the source transfomer at 2:18 Pit on Dec. 8, 1086. An autmmtic feature of the nuclear service bus is a five-second Limit on having taD _ feeding the bus. The control roomoperator closed startq_ transformer No. 2 sqppLybreaker 4AlO onto the 4A bus. Whenthe operator opened the supply breaker (/_01) from startl#p trlmsforlm" No. 1, breaker _10 from stertup transformer No. 2 had just completed the automtic five-second run-out and had tripped open. These events Left the 4A bus without either the nornmLor alternate supply. An attendant result was that whenpouer was restored, DHSsuction vetve IIV-20051 closed ea would be expected in this situation causing the DHS isolation. The pouer sUppLies to both IN-20001 and NV-20052sre currently racked out. The purpose for the DHS system valve interlocks is to prevent over-pressuring the DHSpiping with ItCSIX_m_ure. Since the RCSis -open to atmsphere n there is no need for the interlocks to protect the DHSpiping frm over-pr--,gmare.

_(V Satan 1 04/24/79 Mode 6 2 rain (second event on 05/08/79) (ira NSAI:52, p. A-20): Twice during performnce of tests on vital bus breakers, the vital bus st_ptytng the qpersting RHRSimp ads inadvertently de'energized causing a Loss of RHRSfloe. FLoewas restored tn 2 minutes and 4 minutes respectively.

/dCV Salem 1 05/08/79 Node 6 4 rain (1st event on O4/24/79) (from NSAC52,p. A-20): Twice during performance of tests on vital bus breakers, the vital bus supplying the operating RHRSpunlpwas inadvertently de-energized causing a Loss of RNRSfloe. FLoewas restored in 2 minutes and 4 minutes respectively.

4KV SaLem1 03116182 45 rain (from AEO0): Vital bus trilled. Componentcooling water and service water were Lost. Redundant trains were out for iRintenance (45 Ilin. toss).

410/ Satan 2 12/20/83 Mode5 22 rain (from Seal)rookSheet 8 of 20): During a umintanance shutdownm2RH1closed, resulting in a Loss of RHR fLou. The event took ptece during the transfer of 2B4k'Vvital bus from one station power transformer to the other. The backup power supply for 2B instrument inverter was cleenergized for maintenance. The transfer resulted in a luomonta_-Lossy. of the instrument bus; 2RHI closed in interlock.

(from AEOD): Loss of vita_ bus - due to personnel error resulted in closure of suction/isoLation valve (22 uin. Loss).

tdCV Surry 1 05/24/86 Mode6 (from Seabrook Sheet 10 of 11) On 5/24/86, Unit 1 was at refueling shutdom with Reactor cavity flooded and forced circulation in service; Unit 2 was at 100:; power. Due to maintenance and design

A-49 p. 5 LER DATABASI_ - l.k'V 08/27/93

PHASE2 PLANT NAHE EVENT ]NIT|AL PLANT RECOVERY CATEGORY DATE COND! T!OH T ! IdlE ...... ,...... - ...... change work in progress on Unit 1, numerous eLectricaL busses were cross tied. An_ these were 1hand 1J4161)V emergency busses and vital busses 1-iS and 1-IV. #1 emergency diesel generator was out of service. At qpproxinlatety 1520 hrs, reserve station service feeder breaker 150a opened. This resuLted in an undervoLtage transient sensed mt IJ emergency bus. _ emergency dieseL generator autostarted end esstaned Load. By design, the IJ stub bus breaker opened during the transient which resulted in the Loss of the operet|ng 111 residue! heat removal and 1a cemponont cooling IXal_. The stub bus breaker ms reset end the _ts were returned to service. Numerous spurious trip signaLs, alarms and A Hi Consequence Limiting Safeguards signal were generated during the transient.

(from CR5015): Power Level - OX. On May 24, 1986 Unit 1 was at refueling shutdown with reactor cavity flooded and forced c|rcutat|on in service; Un|t 2 was at 101_ power. Due to mintenance end design change work in progress on Unit 1, 0zmerous eLectricaL buses were cross tied. knong these uere tH and 1J 4160V onergancy buses and v|taL buses 1-I! and l-iV. #1 emergency diesel generator was out of serv|ce. At approx. 1520 hrs, I_w,erve station serv|ce feeder breaker 1501 opened. This reustted in an undervottmge transient sensed at 1J emergency bus. /t3 emergency diesel generator auto started sad essumed Load. By dea|gn, the 1J stub bus breaker opened dur|ng the transient which resulted in the Loss of the operating 1B tea|duaL heat removal and 1B coqxment cooLing IXm. The stub bus breaker tieS reset and the _ts were returned to service. Numerous spurious trip signals, alarm and hi consequence tim|ting safeguards signal were generated during the transient.

L4CV Three MILe IsLand 1 01/09/87 RefueL|ng shutdmm (from LER Search): 1341-1 was |n refueling shutcloun mode with aB" decay heat removal system in operation. APlxmd|x R modification work requ|rad that a wire be Lifted in the circuit for the 1E 4160V current transformer. Urd_t to the etectr|clan, he Lifted and opened one phase on the 1E bus current transformer. Uith th|s phase open, the tM:nLance caused an apparent neutral current and operat|on of the neutral overcurrent relay. This protective feature tr|pged and Locked out the feeder breakers to the 1E bus. The resulting undervoltage ccmcllt|on on the 1E bus caused the auto start of the 8 irgency diesel generator. The root cause of the event m |nadequate instruct|on to the electrician perfoming the turk at an energ|zed bus. The autmmt|c start of the emergency diesel generator is reportable under 10 CFR 50.73(A)(2)(IV). There was no safety significance to the auto start of the emergency dieseL generator. ALthough the operating decay heat removal system was nmentar|ty shutdetm due to the de-energized 1E 4160V bus, there was no safety significance since the backup system m avaiLabLe. Corrective action taken vas to reconnect the wire, reset the Lockout, and re-energize the bus. Uork was stopped and not restarted until the work |nstruct|ons were assessed an appropriate sequence developed.

WoLf Creek 10/16/87 17 -in (from MUREG-1410): The unit was in a refueLing outage and the refueling cavity was flooded with 60 fuel asmmbties |n the reactor. One emergency dieseL generator had been removed frem service and one safety bus had been remved frem service for cleaning. An electrician came in contact with s Lead energ|zed from the other safety bus. The energized bus desnergizad. The eaergency diesel generator started and Loaded. The emergency diesel generator output breaker was tripped to remove the etectr|cien who came in contact w|th the energized Lead. After the eLectrician was removed, the emrgency dieseL generator output breaker would not cLose. The anti-pump circuit of the output breaker prevented recLosure of the breaker once it had been opened after the diesel generator started on undervoLtage. Decay heat removal capability was Lost for 17 minutes. The industry was informed of this event in INPO Significant Event Notification (SEN) 22.

A-50 p. 1 LERDATABASE-,14_/P6 08/27/93

PHASE2 PLANTNAHE EVENT INITIAL PLANT RECOVERY. CATEGORY DATE COGDITION TINE

4¢t//P6 CaLvert CLiffs 1 11112/80 Mode5 or 6 lOmin (from NSACS2,p. A-17): A 4 kV bus normal feed breaker spuriousLy opened. The diesel generator started autmlaticaLLy, re-energizing the bus. During the voltage transients a vital bus voltage Ment Low, arid ESFASA Logic actuated, uhich caused a LOSSof shutdoun cooling fLeD.

44IV/P6 Otabto Canyon 1 01/25/85 Node$ 2 rain (fr_ Seabrook): A Loss of vital 440/ bus voltage resulted in the sat.starts of diesel generator LOG) 1-2, Contalnwnt Fan CooLerSystem1-5, and AuxiLiary saltwater pump1-2. in addition, for al:Proxiitety 2 minutes, the Decay Heat RemovalCapabiLity was Lost Mhanthe closure of the Loop4 RHR suction valve (NOV-8702) resulted tn both Residual Heat Removal(RHR) trains being isolated from the Reactor CooLant System. The RHRsuction valve was subsequently openedand RHRflow established. Within two minutes aLL other affected equipment and system uere returned to their normal standby cGrMitians.

(from DI5015): Pomr Level - OX. At 1750 PaT, 1-25-85, Math Ulltt 1 in Node5 (coLd shutdown), a Loss of vital 440/bus voltage resulted in the sat.starts of IX; 1-2, osntaiment fan cooler system 1-S, and aux saltwater pulp 1-2, and the transfer of the control roomventi Latton system to mode4. tn addition, for approx. 2 minutes, the decay heat removaLcapability was Lost tthen the closure of the Loop& RliRsuction valve (NOV-8702) resulted in both RHlttrains being isolated from the RCS. The RHR suction vaLve Mas subsequently openedarm RHRflow established within 2 minutes. ALLother affected equ|pmnt and system were returned to their normaLstandby ¢(mditicns. Investigation has shc_n that the cause of this event ues mtsadju_tmant of the mm eaitches on the bus G feeder breakers (HG 13 and l&). The arm suitches Here adjusted to a newtolerance and the breakers Mere tested Mith satisfactory results. To prevent recurrence, procedure E-51.2, "4.1610/c|rctrit breaker Pti (preventative Imintsnanos),. is being revised to identify the specific lain switch adjustment required for the bus feeder breakers. Siai Lar events 275/85-004 and 85-005.

4XV/P6 FarLey 1 01/16/81 Shutdown 1 rain (froB NSACS2,p. A-21): A Lossof power to a stsrtup transformer de-energized 4160 V poMer to the running RHRSIXmp. The emergencydieseL generator auto started and re-energized the bus, and the RHRS pumpwas restarted. FLea us Lost for 1 minute.

440/11=6 North Anna 2 0_/08183 Node$ (Sheet 4 of 11) On 4/8/83 powerwas Lost to the A Residual Heat Removal(RHR) subsystem_hen the 21( 4160 volt emergencybus was de-energized. This action resulted in Leaving only one coolant Loop (B RHR subsystem) operable. A single RHRLoopprovides sufficient heat removal capability in Nodes4 or 5.

_cv/P6 SaLem1 01104/83 Nodes& and 5

(from Seabrook Sheet 3 of 11) The Control Roomoperator observed that No. 1Bvital bus had tripped. Since it is supplied from the bus, No. 12 residual heat remvaL (RHR) punT)wasdeenergized; Loss of the pumprendered the associated RHRLoop inoperable and Action Statement 3.4.1.4A uas entered. The operator |Jmed|atety started No. 11 RHRpumpto restore core cooling fLeD. The secondpumpremained operabLe.

440//P6 SaLem2 04/13/83 Node5 < 1 rain See description of Salem2, 04118/83 event.

4KV/P6 Salem 2 04118183 NodeS < 1 rain (Sheet 4 of 11) On two separate occasions, on 4/13 and 4/18, operating Loadson the No. 2A 410/ and vital buses Mere observed to trip. in both cases, due to the de-energization of No. 21 Residual Heat Removal (RHR) pumpNo. 21 RNRLoop yes no Longer in operation. In each instance, the RNRLoopMas A-51 p. 2 LEROATASASS-4W/p6 08/2T/_

PHASE 2 PLANT NAHE EVENT IN|T;AL PLANT RECOVERY CATEGORY DATE CONDITION TIHE ...... :....., ...... -:-...-.....:- ...... ,.--.-.: ...... ilmediatety returned to operation.

&I(V/P6 Surry 1 02/04/89 Node 5 3 atn (from LER Search): On Feb. 4, 1989, at 2345 hrse _tth Units 1 and 2 in cold shutdolm, the 81 and f3 emergency diesel generators auto-started and Loaded onto the 1H and 2J emergency busses respectively. This event is reported pursuant to 10 CFIt 50.73(a)(2)(|V). The diesels started due to the de-energization of the 1H and 2J emergency busses due to the failure of a 4160 volt supply circuit breaker. The running residual heat removal (Rim) pump de-energized during the event. The redundant RHIt _ _s started approx. 3 minutes tater to supply lUUt ftou. The failed circuit breaker was cteehed _ tested satisfactory, safety-reLated 4160 volt supply circuit breaker operator mechanisms uitt be refurbished. The preventive mintonance program for 4160 volt circuit breakers is being evaluated.

&k'V/P6 Surry 1 04/06/89 Node § 1 ain (fr_ LEIt search)= On April 6, 1909 at 0&31 hrs, utth Units 1 and 2 in cold shutdovn, an electrical fault in the ptant switchyard resulted in the failure of a Lightning arrestor on the 500 kv side of the 112 auto-tie trarmformr. This reustted in a differential Lockout of the transformer shich created under voltage conditions on the eLectricaL buses, including the 1H and 2J &l(_)v emergency buses, that vere being sul_tied by the transforsmr. A number of caq:mnents were de-energized, including the operating _Ai residual heat removal (RHIt) ptalp for Unit 1 and certain Unit 1 radiation monitors. The redundant "9" ItlUt _ ms started Htthin one ainuts. The/r3 emr9ency diesel generator 8LIto started and Loaded on to the de-energized 2J emergency bus. Operators performed the appropriate procedures to restore pomtr to the affected eLectricaL buses and components. This event is being reported due to v_otattons of technical specificattorm and an unplanned engineered safety features actuation. The cause of the ttghtning arrestor failure cannot be determined. Some minor caq:onont failures were noted during the event. The cause of these fattures are being investigated and corrective actions uitt be t,_ ,s _iate.

_V/P6 Surry 2 O2/04/89 _de 5 S_e description of Surry 1, 02/0&/89 event.

4XV/P6 Surry 2 04/06/89 Hode 5 1 ain See description of Surry 1, 04/06/89 event.

A-52 Appendix A.2.3 Loss of Vital Bus (VITAL)

A-53 p. 1 LERDATABASE-V_TAL 08127/9_

PHASE2 PLANTNAME EVENT INITIAL PLANT RECOVERY CATEGORY DATE CON91TION TIME "''" .... :'" .... "':" .... "':"" ...... ""'" ...... VITAL Calvert CLiffs 2 02/04/81 Mode6 17 sin (free IISACS2,p. A-S): RWtSflow was Lost due to inadvertent cle-energ|zation of one 120 V ac vital bus. De-energizing the vital bus caused a hiDh pressure inl_Jt to an RHRSsuction valve control circuit, causing the valve to shut. The pumpwas stopped, vital bus re-energized, and floe restored in 17 mimJrt:os.

VITAL CaLvert CLiff• 2 11/22/82 Mode 6 & sin (from Sealx_x)k)- klhiLe dmmergtztng instrument power sq_pty panel 2V02, •hutckxm coating floe was Lost. A shutdolm cooling return valve, 2-SI-652, shut ellen this panel _s deenergized due to an incorrectly inst•lLed temporary japer meant to prevent 2-SI-652 closure. Shutdoencooling floe us restored four minutes tater.

(from AEGI)): Technician incorrectly de-energized • poser supply panel; caused closure of • DHRreturn valve (& rain.Los•),

VITAL CaLvert Ctiff• 2 11124182 (from ALIa)): DHRtest due to • f•i Led I:_ler supply. (Duration of event unknotm).

VITAL Calvert CLiffs 2 12/28/82 Vital inverter f•iled, caused an isolation of the DHRreturn Line. (Duration of event unkncvn.)

VITAL Calvert CLiff• 2 01/04/83 15 min Inverter tripped during stw_eittance testing - caused isolation of the DHRreturn Line. (1S sin Less)

VITAL Cataabe 1 11/26188 Node5 2 •ec (from LERSearch): On 11/26/88, at 1805:07 hrs, static inverter 1EID and 120 VACpower IXmetlxmrd 1ERP9expsrilmced an undervoLtage condition for _. 2 seconds. As • result, multiple alarm associated utth engineered eafmrds features were displayed in the control ram. The nuclear service tater (AN) pit D mergancy to Level atom occurred and cleared vhen it _s ecknoeLedged. ALL idle RII ISaqs started as expected and the suction valves and discharge valves for the systemautomticatty to the standby nuclear service water pond. There was • false reactor coolant (NC) Irystea wide range pressure signal Nhich caused residual heat removal (NO) systemsuction isolation valves 1NO1Band 111)368to close as expected, in addition, contaillent purDe (VP) system train D, Nhich vas in service, aUtOlmticaLty shutckxm. The Unit 1 control roomarea ventilation system train g filter inlet valve closed due to • false chlorine detection atom. Unit 1 us in NodeS, cold shutclmm, at the tim of this incident. Unit 2 w in Node 1, Ixx_r operation, et _ reactor polder at the time of this incident. A work request m closed out.- Following the undervottage condition, operations personnel noted • ground on 1;5 VDCdistribution center 1EIX). This condition is being investiDated. The root cause for this incident could not be detemined.

VITAL Davis Besee 1 06/28/79 Mode 5 18 :in (fro NSAC52,p. A-2): An accidental short circuit (n one channel of reactor protection was caused by a slipped eitiDator clip during surveillance testing. This btev an inverter fuse, causing the Loss of an essenti•L bus. Loss of power to the SFASchannel on that bus actuated the pressure bistable trip input on one RHRSsuction valve, causing a Loss of RHRSftov. The RHRSlamp sis secured to protect the pump. The bus was reenerDized from an oLternete source, the suction verve reopened, and floe restored.

(frcm AEO0): 18-minute toss of DHR. During surveillance testing, a stipped atiiptov clip caused a short circuit and failure of power •443ply to an SFASchannel. As • result, DHRsuction valve closed.

A-54 p. 2 _I_RDATABASI_-VITAIT 08/27/93

PHASE2 PLANTNAME EVENT INITIAL PLANT RECOVERY CATEGORY DATE CONlDTI ON TIHE

VITAL Oiabto Canyon2 01/17/86 ModeS 13 m|n (from Seabrock): At 0455 PST on i/17/86, Nhite attempting to transfer instrument AC Panel PY 2-1A from normL to backup power m.q3pty, an unLicensed operator went to the _rong paneL and inadvertently transferred instrument AC Panel PY 2-1 to its backuppower source. This momentarytoss of power caused relay •cruet•on _hich resuLted in the closure of hstduaL Heat Removal (RHR) valve 8702.

(from CILS01$)= Power Level - 08. At 0455 PST on 1-17-86, while •tteiq)ttng to transfer instrument AC panel by 2-1A from normaL to beckup IXXVersupply, an unLicensedoperator vent to the idrorigpanel arid inadvertently transferred tnstrum_t ACpaneL PY 2-1 to its backup pover source. This momentarytoss of power caused relay mctuetion Nhfch resulted in the closure of mfclumt heat removal (RHR) valve 8702. In response to the ensuing toss of fto_ sLam, RiOtpunp2-1 me secured by a Licensed operator. RHRvalve 8702 m reopened from the control room. RHRpump2-1 m interred, observed for seat delOS, and declared operable at 0508 PSTe1-17-86. No operations were in progress that invotved • reduction in reactor coolant system boron concentration. Thus, the requirements of Tech Spec 3.4.1.4.1 action li were met. To prevent recurrence, the operator involved has been counseled, operating procedures on transferring instrullierlt AC panel paver 8_q_Lies ,ttL be revised, end panet identification LabeLs in the |nstrulaeflt AC perlet• uitt be upgraded.

VITAL OiabLo Canyon2 06/29187 Node4 5 rain (fran LERSearch): At 1829 PDTon dune29, 1987, uith the Unit in Node4 (hot shutcioun), a fouLted coil on reLay 2TC_INX initiated a vettage transient on instrument pouer panel PT-24, Nhich resuLted in closure of micluat heat removal (RMt) valve B701. The momentaryvoltage transient resulted in an •citation of the RHNeutoctomJ_ interlock (ACI) function since the ACt and the foutted relay share a t instrument pouer supply. The ACI activation _ RHRvalve 8701 to cLoDs. The tic•treed operators noted that RlUtISanp2-1 discharge temperature ms decreasing rapidly and then observed RHR valve 8791 was shut. In response to the valve closure, RHItpump2-1 was secured and vaLve 8701 m _. _ pump2-1 was mterted at 1834 POT, and no a_xxrmt pumpseaL Lemkageuas•burred. The four hour non-_ report required by 10 CFR50.72(S)(2)Clll)(S) was madeat 1%5 PgT. The _atLure of the reLay was • mutt of the coiL shorting. MeLtednylon aretnd the iron core interface surfaces and some stight corrosion of the surfaces ms noted. Pgande is preparing a Letter for mittsL to the NRCrequesting and justifying removal of the RHRACI function. The RHRAC! function _i LL be removedafter NRC•taft reviev and approval of the proposedchange.

VITAL NcGuire 1 1_/24/82 6 ein (fr_a AEO0): Inverter faiLure csused closure of suction/isoLation vaLve (6 :In. Los•).

VITAL I_,ui re 1 07/13/82 Static inverter EVlA moLfunctioned causing a residual heat removal system (liD) isolation vaLve to close. Operstors restored NOfLou, but not before toss of flow effected s transition from ModeS to Node4.

VITAL Ni t Lstone 2 01106182 7 gin (from AEOD): Technician error during s preventive •mint•fiance test resulted in toss of a vital instruaent paneL, and autoctosure of the suction/isoLation valves (7 rain. toss).

VITAL North Anna 1 01/22/83 Mode6 4 rain (from Sosbrock): RHRflow m Lost for approximately & minutes.

(from AEO0)= FaiLed inverter, causedRHRsuction/isoLation valve to close. (4 min Loss)

A-55 p. 3 LERDATABASE- VITAL 08/27/93

PHASE2 PLANTIMNE EVENT INITIAL PLANT RECOVERY CATEGORY OATE CQNDI TION TI NE

,.-----,,.-. 0.-0.,... e .,_...... -...._.o,.._m.Q.--...-.,---.....,,.-.----....---0--0------0-----_------

VITAL North Anna1 04/2?./87 Mode5 (fr_ LERSearch): At 1530 hrs on April 22, 1987, w|th Unit 1 in Hocle5, the residual heat remvet (RHR) suction tim was tsar•ted uhan NOV-1701closed due to • Loss of power to the 120 VACvital bus (Va) 1-1|1. NOV-1701being closed resulted in a temporary toss of the RHItsystem. Therefore, this event I• reportabte pursmmt to 10 CFR50.73(A)(2)(Vll)(B). At 1515 hrs an Aprl L 22, 1987, during the perforce of • periodic test, the 1-111 inverter fatted, resulting in a Loss of power to VG 1-I11. Vital bus 1-111 sUppLies pete" to an uittery relay for pressure channel P-1LK)3Nhtch provides the togic to ©LoseNOV-1701on high ItCSpressure, i_an this auxl|lery retay was de-energized an auto ©t_re sign•| was sent to HOV-I?01, thereby closing NOV-1701. Power us quickly restored to t/B 1-111, NOV-1701mm rmnod and RHItfLos was restored. The cause of the |Mrtar failure tms detemtned to be e bLounfuse. To prevent recurrence of this type of event, North Anna is actively pursuing a tqNdmicat specification changeto ettntnate this ItHRinterlock.

VITAL North An 2 04/29/83 Node 6 • 1 rain (frm Seebrock Sheet 5 of 20): One of two source retie chela (N-31) and the Containamt ParticuLate end GaseousRadiation mnitors (Hlt-2S9 & 26) I_ere dmrgized. The vital bus uss pramptty reenergized and the deenergized equil:mnt promptty restored.

(fr_ AEG)); Lossof vital bus. RHRsuction/isoLation valve closed. Caused by emintanance permmL conducting • test as Loadswere being transferred (<1 win. Loss).

VITAL Pato Verde 2 01/30/86 Mode5 (fr_ CIt501S): Poser Level - 0_. At 1924 liST on 1-30-86, Palo Verde 2 ms in Node5 when an unauthorized edification on a vital power inverter caused a fafture of the train *A', class 1E, I&C peuor system, uhich resulted in a control roam essential filtration actuation signal and m temporary toss of train mA_chutdoun coating. The cause of the failure ms attrtlxlted to tmmdequatecontrol of a edification consisting of • resistor jump•red erc_ • capacitor in the circuit. The modification _.ued an excessively high current on on inverter circuit board, and resulted in 3 btoun inverter fuses. The inverter toss caused a Lossof Ix_r to a radiation monitoring unit, which in turn cued the CItEFASand the taqoorery temination of train _A_ SO(:. As corrective action, aLL inverter• were inspected for additional unauthorized edifications, the blare fuses were replaced, and inverter specs uere checked° AcldlttonatLy, work controt procedures uiLt be revised to mize the importance of reeving at[ temporary modifications prior to putting an etectricot system back in service.

VITAL RanchoSeco 06/24182 Nodes4 and 5 (from Sosbrook Sheet 1 of 11): During a preventive mintonance procedure on the B inverter, there m a momentBrytoss of power to the O bus. This in turn caused a short duration Lossof DHRS. The core teq:erature retained unaffected by this toss of ftos and the A system m avaiLabLe on standby.

(from/kEG)): Simuttane_ test _ ma|ntenar_e caused faiture of bus, closure of the suction/isoLation vatve, and Lossof DHRflow. (Duration of event t_z_m)

VITAL RanchoSeco 11/15/86 Ikxle 5 (free SeabrookSheet 19 of 20) The plant was in cord shutdotm, removingdecay heat via the Decay Neat RemoverSystem (DHS), Train A, on Nov. 15, 1986. At 1=00 PM, in preparation for • fuse reptacelnt activity in the S1A Bus inverter, S1A bus power was llmentartty ihterr14)ted, DHSoverpressure dlatabtes tripped, IN-20001 closed chtch tripped DNSA Pumpas designed. Steps were taken immediately to restore a OtiStrain to service in accordance with T.S. 3.1.1.5.

A-56 p. A LERDATABASE- VITAL 08/27/q3

PttASE2 PLANT_MME EVENT INITIAL PLANT RECOVERY CATEGORY DATE CONDITION TIME

.., ms, e o.u,e o a. o _ m,. a, e e o_. a* e.,,.,, _.. _ _ m_ _e o _ _ _. o o _ _ _....*,. *. w ,,,,,. o m ,,. ,,,,.. e ,,, ,., i.. ,.,....,. ,,,.,,,. o..,,o e o w o..,,,-g w o * ,,,, *_.. ,,,o ..--,.-.. o ,,,.. o .,,.

(from Clt5015)x Po_r Level - OX. The plant uas in cold shutdosn, removing decay heat via the decay heat rllOVet system (DHS) train UA" on 11-15-86. At 1100 in, in preparation for s fuse reptlCelMflt sottvlty in the $1A bus inverter, $1A bus power ms momentarily Interrupted, DH$over-pressure blstsbtes tripped, IN-20001 closed Mltch tril:ped DH$nAn pumpas designed. Steps were taken tmecltetoty to restore • 0H$ train to service in accorckmceutth Tech Spec 3.1.1.5. The basic cause of the inverter failure is that the original design did not mtto_ for testability of the device through the use of a Imbetltuto paver source. That design deficiency, identified as early as 1979 (NCR0s-1258, Rwisim 2). i _tzed by the _Nmt action plan for _rf_m_e tq_m_t. There Is a _wmlve mime _u_. U I_A. =ststlm Im_rter r_ino - static pr_ts im_rtara, e U_t is _ted to he mrfom_ _ W _mr on the tm_rtera, a I_NAC vital _ systm tq_wmmt m Wlrov_, but has not been leptem0nted yet.

VITAL salem 2 0S/1&/83 HodN 4 and 5 luedlltety (from saalx_k Sheet 5 of 20): On rue mrste eccast0m on Nay 14 and 15, 1983, a Residual Neat tmovaL system reaction wLve ms observed to have cLosed, thus eliminating flay in the operating RHIt Loop. In each irate, the operating pump i stq:lxld arid Action StatMient 3.4.1.&B I entered. No reduction in resctor coolant system bor_ concentration occurred uith an RNRLoopout of service. A Looptin limed|sLaty _to_ to service.

(from AI_O): RHRsuetlan valve closed during operatim. (Duration of events unknoun) The 0S/14/83 event mm trl_ _ • vital inat_mnt _ Milch m de-ene_ized for mintenance.

VITAl. Salem 2 11/28/83 Ilode 5 (frml Sasbrook Sheet 8 of 20): During a mintenance shutclmm, 2A vital tnatrumnt bus Ires transferred to its alternate pouor sq_pty to perform routine rater calibrations on 2A inverter. A vottNle trarMliont caused 2RII2 to shut resulting in a toss of RHRf|ou. The valve yes tmedlatety reqxmed and RIIRf Lea tms re-estabt i shed.

(frm AEOD): Vital bus transfer caused voLtage spike Match resulted in closure of suction/Isolation valve. (Duration of event unknolm)

VITAL South Texas 1 02/12/89 Node 5 (frml LEIt Search): On2/12/89, unit 1 mls in mode5. At 0S29 hrs a plant operator noticed evidence that inverter IV-1203, Milch supplies uninterruptabte po_r to distribution panel 9P-1203, was overheating. The operator immdtatoLy transferred the distribution panel to its alternate source and secured the Inverter. Since this yes a dead bus transfer, it resulted in a mmentary toss of paver and subsequent actuation of the control roomand fuel handling bLcl9 INACIryStelM to the emorency node, tripping of residual heat removal pump1D, and shifting of centrifugal cherging Ixmp 10 suction to the refueling water storae tank Iron the votuR control tank. These system were restored to normt operation after power was restored. The cause of this event ms a short to ground on the secondary side of the inverter ferroresonent transformer. Corrective actions include replacement and failure analysts of the failed transformer.

VXTAL Summer 11/12/83 Itxle 5 S mtn (fron Sasbrosk Sheet 7 of 20) An Engineered Safety Feature (ESF) 120 VAC vital instrulntatton paneL, APN-S901, uas transferred to alternate pouer to eccam_late nod|fications to its normal pouer source. Vith Train B Residual Heat Removal_tm tn service, its suction valve, XVG-8701Aclosed. The valve uss reopenedwithtn approximately S minutes. No adverse comequencas resulted due to plant conditions and the short duration of the event.

A-57 p. S Ln OATAIA_ - VITAL 08/27/93

PIL4U_2 UT IM/4E EVENT INITIAL UT RECOVERY CATE6ORY DATE COIIOITION T!lie

o. o. e,,m _ a, * .. e,,et,,e o o _ _ ***,* ** e*o o e, eo w* _ ode o o aJ*e I o o*l o e,o o* o o_Q a e, ,,* e,e ,a,* o moo e, .. e,.,. m ,o ,,. Ieo oe_e * ._eH. * o*,,. * * a. * a o,,...., o., .., .. e o *oel ,,.

(frm MOO): (Date is 11/12/84) Bus transfer during plant modification caused In interruption of pouer to an ElF tmtruuntatlon bus. An erroneous overpressurlzatlon signal resulted cmmtng Imcttm/Isotatlon valve closure, and interruption of DHRflow (5 gin. L_),

VITAL lumor 10/18/86 Node5 25 aln (frm ledwook Shoat 11 of 20)8 OnOct. 18, 1984, outage vlth Train A of the Residual Hut lemvet (lUiR) Jystea in service, llIR Train l out of service for r_ttne imtntononce, and the reactor coolant system (ItC$) vented It I teq:orMura of qqx_lmteLy 110 clogFahrardMIt. At 1605 hrs. a poser toso to 120 VACdistrlbutlcm panel APU-Sg01demorglzed SoLid trite Protection Syztea (raPS) ChmL! and c:mNd the irmtrumnt penal for tCS vide re pressure (PT-/_3) to initiate vetva XWi-8701A.

(frm ABm): 1 DM tsop me cut for survaiLtance tsotlnl. An Inverter failure caused closure of the operatlq toq)*8 smtlcm isolation valve (2S sin. Loss).

(frm C_015)8 Pouer rivet - _. On 10-18-84, the plant was tn Hode 5 for _ first refueling eutqe u|th train 0A0 of the ItllR system tn mice, PJIRtrain 08° mat-of-sorvlca for routine emlntm, end the _ vented at e tmperatura of qR=r_. 110 P. At 1604 hrs i pear toss to 120V ACdlstrlbutimt panel APN-5901de-energized solid 8tare protl_tlmt syataB (NPS) ch4nwL I and r,auaed the Instrumnt channel for I_ ulde range prosma'e (PT-4_3) to Initiate an auto-ctosure of the q_rabte RHRtrsln08 eu0tlm tsototlm valve XVe-8701A. Fottmdng determlnot|_l that the pear tee8 had bMn c_NMd by pertmmwi error during the perf(xmi_ of a ptant mxllfJ(:atian, mtiorm ixll'll_lnot restored p_or to MII-5901, XVa-8701Am _ and train 0A* of the ItB system returrmd to eporabLe Itetuo at 1630 hrs (total tim of INR IsoLation ms q_rox. 25 minutes). RCS tmporatura Incraemed frm 110 P to 130 F during the ewnt. The teas of tB met the conditions of an alert, and the proper notification8 we rode in eccordm=e with the mergercy plan.

VITAL Turkey Point & 03/15186 Node6 5 Idn (iron Sosbrook Eheet 18 of 20) Workus progressing to deanerglze end reptece a vital bus feeder brecker, 4P08. ktm breaker 4_8-3 us eponed, the R_ pumpsuction valve uant ctoesd. Upon receipt of the tetdmm, Isotatian alia. the RHRpu_ m stq3xld, the breaker re-energized, the valve re-qxmed, the laiR? rlmtarted, and ftce restored tn approxlmtety 5 Idnutes.

(frm CRS01S): Powr Level - C_. On 3-15-86, thtte Unit 4 was in a scheduled refueling outage Hide 6, work tms progressing to de-energize end replace a vttat bus feeder brewer 4_8. _ breaker 4PQ8-3 ms qxned, the R_ pulp suction valve ant closed. Upon receipt of the Letdoun isolation Item, the RIIRpulp m stopped, the breaker re-energized, the valve reqxmed, the lUt:) restarted and fto_ restored in approx. 5 ntnutes. Tech Specaction statemnt 3.10.7.2 us entered during the q_x_. S minutes of ftc_ toss. There tms no noticeable increase in the 93 F systn teq=erature. Whenbreaker /#08-20 us opened, 8 process tidier|on ionitor rack was Lost, CauSing the ¢ontatrtlent vantttatlan systi to isolate md the osntrot ram ventilation syston to isolate and s_ttch over to the reclrcutattcm mode, as designed, The purge valves tara secured per Tech _ action statemnt 3.10.2.A, by removing power fuses. No significant Increase in activity m recorded on the plant vent effluent monitoring system during this event; other m of monitoring ©ontatnmnt activity were avaiLabLe. No release path to outside centairmnt ms avaiLabLe. Whende-anergized, feeder breaker FP08 reptecemnt m ccxlpteted, power m restored ard the system tara then returned to their normal Line-up.

VIT_ Zion 1 03/17/82 _de 6 3 rain (frce Sesbrook Sheet 1 of 20): RHRsuction valve 1NOV-RH8702started to close due to an ir_Jvertant op_|ng of inverter 111 output breaker. The running pumpHas tripped. The inverter output breaker m A-58 p. 6 LERDATABASE- .Y|T_L 08/27/93

PNAIIE2 PIJUFrNAHE EVENT |NXT|AL PLANT RECOVERY CATEGORY OATE COla[)T|ON T|HE e ul_e_eemeeleoe eqllQieeaee e _eeql_t ewoe e u mm,e_nm,n ae _ o oaeeoee eem e ooeaupa e_a.aeee e eele_ eeee e eee_au, aeu_e eeeee e eoeeeoeoleeu recLcmcl, RHRauction vaL_ 1HOV-RH8?02ms reopemKI, and the RHRslmton was restored to operation ulthln 3 minutes.

(from AECO)| Irldvertent (©ontrsetor perllonneL) qpontng of Inverter. output breaker caused cLosura of tho I, iR pumpa,uctlon valve (3 sin. toss).

VITAL Zion 2 01/03/86 NodeS (from Beebrook Sheet 17 of 20) A momentaryfluctuation of output of inverter paver stqpptyBus 213 (Caueo Uldototm)reused tho chergtng flow control valve, 2VC-FCV121, to fall to tho 20X demnd posttim and alLaH)caused 21qov-IIH8701,tho IINII Pimp suction IsoLation vaLve to fell ©Losed. This Inormmed chlrgtq ftc_ from 39 to 190 glumand IsoLated Letclounftou resulting in Lifting of the pressurizer operated relief vaLm (POIVs). tlhlte tnvsotlimttng the cantos,Bus 213 was a_mtn deenorll(zed and tho PaWs again Lifted.

(from C115015)| Pmmr Level - 0_ At 1547 on 1-3-86 Unit 2 u_ idtut clausfor s refueling outage and tho RCStin fiLLed solid ulth no ixdd)to In the Prusurlzer. A Immntary fluctuation of output of Inverter pouer m4q)ty bus 213 (ceuso urdmmm) caused the charging fLov control valve, 2VC-FCV121, to fall to the 2001dmmd position, end aLso cauced 2NOV-RII8791the Hit pumpsuction IsoLation valve to fall L closed. This Incroesocl charging ftou from 39 to 190 p, and IsoLated totdoun fLee resuLting in Lifting of the premurlzer powr operated relief vaLves (POItV0s). WhiLe Irlveetlgatlng the caLMs, Bus 213 ms qain cloenoflllzed and the POIIV0eawln Lifted. The cause of the bus output fLuctuetlon Is c:urrontLy urdmoun. This event Is r_portad)Losines Tech Spec6.6.3.N roqutras • 30 day _rttton report on actuation OF the _ Wotoctlon systcn.

A-59 Appendix A.2.4 Loss of Component Cooling Water (CCW)

A-60 p. 1 LERDATA.BASE- C_ 08/27/93

2 PLAIITNAME EVENT XNITXALPLANT RECOVERY CATEOOaY DATE CONDITION T|ME

•u..l,-,i.b m.* eel, o el,.. e o. o o,. m s w. e t a o..I m o m.,. ii m a e_e ee ..,. I,. a._, I I _ o o.. _.. ,. I..! g o o o _ o o g, o..,,,.D m.le I.,., t, ,-,-- Q e _. Q o.. e q...I-,, o e o.., e I I,. COW Dretduood 1 01121187 zero power 14 mtn (from LERSearch): The 1D residual heat removal (RHR) heat exchanger (HX) _as out of service ulth the tube side _ratned. Preparations were modefor draining the componentcooLing water (CC) shettside of the HXto aLtou reptecemnt of e flange gasket. AT 1745 draining of the sheLL side of the HXyes started. At 1802 the 1A CC p_p tripped due to Low Level in the CC surge tank. The LowLeveL alarm on the main control board did not anma_tate although the sequence of events recorder did indicate s toy LeveL. The draining ms stopped, the surge tank refiLLed, and the isolation valves sore checked. At 1816 the 1A CCp_p ms restarted and the systemms restored to nor_L. The cause was the CC inlet isolation valve Leaking and contributing was the fat Lwe of the CC surge tank high/toy Level stem to annunciate on the main control board. AdditionaLLy, the CCmotor operated outlet valve on the 1DRHR IIX mm found 6 tul_rmoff its seat. The Leaking valve has been repaired, the Lindts for a motor operated valve mrs adjusted, the rain control tmrd eLem mm troubleshot and the symptom couLdnot be duplicated. Work Is in progress to check the cetlbratton and er,atlng on the CCsurge tank tnstrmmnt Loop.

CCW 8yren 1 0&108/87 zero Dover 17 n|n (frm LERSearch). On&IV87, st approx. 17_, a contracted maintenance crea began work on the tl_torque motor operator of the WlA" residuaL heat removal (RH) heat exchanger componentcooling water outlet isolation valve, lCC_12A. This valve ms a point of isoLation for work on the RHheat exchar_r, tihich required tt to be dro|ned of caqxxwnt cooling tlater (CC). Shift operating perscmet grlmted permission, with the under_ltancllngthat If tt becam necessary for the cr_ to stroke the vaiva, they wouLdobtain authorization. Natntm ©r_ stroked the valve in order to reLease torque m the motor 9emr set. It is unelasr dlether they ecttmtiy received authorization or not. This aLLoyedCC to Ik fLow through lCCg&12Ato the heat exchanger and out the drain. This caused the (CO) surge tank to reach the toy Level CC pep trip. The "IA" CC pumptripped at 1726 on 4/8/87. The surge tank is commonto both ira|m, consequently, both trains of componentcooling were inferable. Leak ties dleco_rad and isoLated. System ms then re-fiLLed, and the -IA" CCpumpre-sterted. TotaL tim both trains were inoperable ms 17 minutes. Causeof the event was • communicationbreskdoun between the maintenance Cm and shift qperatlng personneL. Contracted maintenance personnel nov work under the sameprocedures as station personneL. SimiLar event: 455/86-001.

CCU Mains Yankee 06102181 Mode 6 30 rain (from NSACS2,p. A-30): During refuel _i shutdmm, service water cooling to the RHRScooler was interrupted for approx. 30 minutes as • result of • breaker trip. el:hi Turkey Point 3 10107183 (fr_ AEO0): FLowrestr|ction on componentcooling _ater d|scharge valve on RHRheat exchanger. (Duration of event unknm_)

A-61 Appendix A_2.5 Inadvertent Safety Feature Actuation (ESFAS, ESFAS/SI)

A-62 p. 1 LER DATA BASE - ESFAS 08/27'/93

PHASE2 PLANT NAHE EVENT INITIAL PLANT RECOVERY CATEGORY DATE CONDI TI ON TI HE

---.-...----....._...... ¶.0...... :,.0...... ,-- ...... _....o ...... ESFAS Cook 1 09107185 Node 5 2 min (from Sesbrook): Power was tost to the Controt Roan instrument bus distribution circuits for charmer 3 and charmer 4. This resulted in various ESF reactor trip signals and Loss of the residuat heat remover pumps. Channel 3 and 4 circuits were being powered by an alternate source white the normal power source ms out of service.

(from CR5015): Power reveL - 0_. On 9/7/85 at 0720 hrs with Unit 1 in Hocle 5 power _u; tost to the control room instrument bus distribution circuits for charmer 3 and 4. This resutted in various ESF reactor trip signsts and toss of the RHR pumps. Chennet 3 and 4 circuits were being powered by an atternate source uhite the namer lover source was out of service. The circuit breaker for charmer 3 tripped as a resutt of an inadequately terminated Lead. A tic•nEed operator investigating the paver toss thouDht the charmer & circuit breaker had tripped area. The operator than attempted to reset the breakers by opening then c|ostng the breaker. This resutted in the charmer & breaker being momentarity de-energized. This caused various ESF reactor trip signets and the toss of RHR pumps (due to the refueling water storage tank Level indication re•din9 tow from power toss). This placed the unit in • LCO per Tech Spec 3.4.1.3. The RHR system ms made operabte within 2 sins after toss. To prevent recurrence the operator has been counseled not to take immediate •ctions Nhere the situation does not require it.

ESFAS Davis Bess• 1 04/19/80 Hocle 6 2 hr 30 sin (frm liSACS2, p. A-3, p. A-1S): A Loss of two vite| instrtment buses resulted in • safety features actUation (SFA) while in mode 6. Cc_ntrot paver _as also Lost to the RHRS suction valves intertock, causing them to shut. The SFA caused the suction of decay heat pump #2 to be transferred to the BUST end then emergency stamp. Water from the BUST sis gr•vtty fed to the emergency staIp during valve stroking tim (approx. 1-1/2 minutes). The RHRS pump injected about 3,500 gettons of BWSTwater into the ItCS, resulting in • smtL amount flowing out of tygon tubing used for vessel tevet indication. Approx. 1,500 gaLLons of RUST water drained to the emergency suIp. The I_HRSpu_ then tost suction and became •irbound. Extensive I=usP venting ws conducted and RHRS flow has restored 2-1/2 hrs tater. RCS temperature increased from 90 F to 170 F.

(from AEO0): 2-1/2 hr toss of DHR. Vibration from construction work actuated a ground fault relay. Due to an abnomat eLectricat Lineup associated with outage activities, toss of power resulted in SFAS actuation. Controt power to the DHR suction valves has Lost, causing the suction vatves to cross. The SFA$ actuation transferred the DHR pump suction to the BUST and then to the empty sump. The pump became s|rl=Qund. RCS teq=erature increased from 90 deg FSh to 170 deg Fah white the vesaet head _s detensioned (140 deg Fah is the maximum temp. attoued white the vesset head is detensioned).

ESFAS Davis 8ease 1 06114180 Node 6 2 min (from ItSkCS2, p. A-15): 14aintanance induced vottage transient in containment pressure section of the Safety Features Actuation System (SFAS) caused an SFAS actUation. This caused RHRS lump suction to switch from the RCS to the borated water storage tank, injecting 16,000 gattofls of BUST water into the refueting carat. Seven minutes tater, a tow BWST tevet actuation of SFAS shifted RHRSpunp suction to the empty emergency sump. The RHRSpump tost suction and v_s stopped causing a toss of decay heat fLou.

(frcm AECO): DHRpump ftou Loss for about 2 minutes, inadvertent SFAS actuation caused OHR pump reatignment to the BUST and BUST isotation. An i&C mechanic was restoring containment pressure inputs to SFAS farrowing an Integrated Leak Rate Test. Sec_se of a proceclurat irmdequacy, SFAS was actuated and the DHR pump was realigned to deriver BbfSTwater to the RCS and the refueling carat. When 8WST tevet dropped to the tow tevet timJt, SFAS tevet 5 actuation took ptace ctosing the BVST isotation verve causing 8 toss of suction to the DHR pump. A-63 p. 2 L_RO,A,TA,,eAS_- [SFAS 0a/27/93

PHASE2 PLANT IL4HE EVEHT XNITXAL PLANT RECOV1ERY CATEGORY DATE CONDXTXON TXlIE

-.------...... ,,...,...,..,...,...... ,,,......

ESFAS Davis Bease 1 07/24/80 Hede 5 2 rain (from NSA(32, p. A-4): Yhile attempting to clear Lights on the SFAS foLLot_ing maintenance, technicians adjusted a potentlometer, una_re the RHRSsuction valve trip feature had been restored. This caused an RHRS suction valve closure. The RHRSpug) was immediately stopped, and floe was restored within 2 minutes.

(from AEOO): DHR floe ns Lost for about 2 minutes due to an inadvertent closure of one of the DXR Isolation valves. SUbsequent to making a plant modification, an Z&C mechanic performed restoration work out of sequence. As a result, one of the isolation valves closed.

ESFAS Davis 8esse 1 07/24/80 Node S 50 ain (from NSAC52, p. A-&): A Loss of RIOTSfloe was caused by inadvertent automtic closure of an RHRS suction valve. The rum|ng RIIRS pump was stopped to prevent damage. A manual bypass line around the shut RHR$ suction valve was opened, system venting was performed, and the pump restarted. RHRS floe was Lost for 50 m|nutes.

(from AE(:I)): OHlt floe ms Lost for $0 minutes because of an automatic closure of an isolation valve. An etectr|c|an blew a fuse vh|te conduct|ng w|re pulling operations ossoc|ated vith a plant design change. As a result of the btom fuse, an automatic closure of one of the DHR |solar|on valves took place, and the pump became air-bound. The DHR floe path was restored by epening the manual bypass valves. During the event, the hottest in-core thermocoupte temp. rose from 104 d_g Fah to 111 (leg Fah.

ESFAS Davis 8esse 1 08/01/80 )lode 5 3 rain (frm lISAC52, p. A-5): (Date is $/3/80) During m|ntevmce, a SFAS bistable was reaoved resulting in closure of a RHRS suction valve. The operatir_ RIIRS pump vas stopped to prevent damage. The bistable was replaced am:l RHRS flow r_.-stored in approx. 3 minutes.

ESFAS Davis Bease 1 08113180 Hode 5 5 m|n (from NSAC52, P. A-S): Hafntenance performed on a SFAS cabinet resulted in RHRS suction valve closure. The operating puep was stoT_:ed. The RHRS suction valve was reopened, the pump vented, and RHRS floe reestablished in approx. S minutes.

(from AEGO): DHR floe was Lost for about $ minutes due to an inadvertent closure of one of the DHR isolation valves. Valve closure occurred during SFAS channel modification work. The l&C mechanic failed to fully defeat the automt|c isolation valve trip prior to perfoming SFAS charmel modification york.

ESFAS OiabLo Cany_ 1 09/08/86 )lode 5 2 sin (from Seal)rook): An Instrumntation and Controls IX&C) technician inadvertently grounded a poeer sn4_Ly while installing a modification in a Solid State Protection System (SSPS) cabinet. The momentary grounding of the power supply caused relay actuation _hich resulted in the closure of Residual Heat Removal (RHR) valve 8702 and an RHR toe floe alam. In response to the RHIt loe floe alarm, the operating RiOt pump was secured by a licensed operator. RHRvalve 871)2 was reopened from the Control Room. The RHR pump was restarted at 2316 POT and no seal damage was observed.

(from CRS01S): Power level - I_. AT 231/, PCT on Sept. 8, 1986, with the unit in Hode 5 (cold shutdown), an instruw_tation and controls (%&C) technician inadvertently grour_ed a power supply while A-64 p. 3 LER DATA BASE - ESFAS 08/27/93

PHASE2 PLANT NAHE EVENT INITIAL PLANT RECOVERY CATEGORY DATE CONDI TI ON TI RE ...... :...... _...... :.... _...... ---::--, installing 8 modification in a solid state protection systems (SSPS) cabinet. The momentary grcxrding of the power supply caused relay actuation wh|ch resulted in the closure of residual heat removal (RHR) valve 8702 and an RHR Low ftoH slam. In response to the RHR Low flow alarm, the operating RHR pump was secured by a Licensed operator. RHRvalve 8702 Has reopened from the control room. The RHR pump Has restarted at 2316 POT. and no seal damage was observed. A significant event report Has not filed within the &-hour time requirement of 10 CFR 50.72. The significant event report Has male at 1744 POT, Sept. 9, 1986. The event was reviet_cl at an I&C tailboard meeting emphasiz|ng attention to energized and potent|ally energized c|rcuits when working on electrical components. The circumstances and Lessons Learned fraa the event Hill be evaluated for possible inclusion in the generic net/ employee training program for I&C personnel. Additional training on 10 CFR 50.72 reporting requirements will be provided for all applicable personnel.

ESFAS North /rune 1 11106179 Node 6 5 rain (fran HSACS2, p. A-3): A j_13er Has installed in the solid state protection system for the purpose of conducting time response testing without disabling the aut_natQc closure of RHRS suction valves. A false high pressure signal caused an RHRSsuction valve to shut. FLow _es re-established in approx. 5 n_nutes.

ESFAS Salem 2 06/23183 Node 5 See description of other Salem 2, 06/23/83 event.

ESFAS Salem 2 06/23/83 Node 5 (from Seabrook Sheet 6 of 11): During routine shutdowl operation, the Control Rosa operator observed two different fnstances in Mlich a spUr ice s Safeguards . Equipment Control (SEC) System actuation caused various Loads on the No. 2A vital bus to be de-energized. In the second case, the bus irffeed breaker opened Hith no automtic transfer, rendering the bus irloperabLe. In both instances, due to the Loss of the operating Residual Heat Removal (RHR) pump, flow in the operating RHR Loop was Lost.

(from AEO0): Loss of RHR pump due to spurious actuation of the SEC (safeguards equipment control) system. (Duration of event unknoHn)

ESFAS Sequoysh I 09116182 Node 5 6 rain (fram Seabrook Sheet 3 of 20): Both trains of the Residual Beat Removal System were declared inoperable due to the inadvertent cLosing of RHR suction vaLve 1-FCV-74-2.

(from AEOD): Poser yes removed to allow nx)dification work on solid state protection system; RHR suction valve closed. (Duration of event _)

ESFAS Summer 10/02/84 Node 5 inaediateLy (from Seabrook Sheet 11 of 20) With Train B of the Residual Heat Removal (RHR) system in service, an instrument and control (I&C) technician removed two (2) fuses in Solid State Protection System (SSPS) Cabinet XPN-7020 for personnel safety during implementation of 8 modification. The fuses Here immediately replaced when the technician heard a relay activate. The deenergized circuit caused the Train A RItR suction isolation valves XVG-8702A and B (one valve in each RIIR train) to close. Operations personnel innedJateLy restored Train B RHR to service after the valve closure.

(from CR5015): Power Level - OZ. On Oct. 2, 1984, the plant Has in Node 5 with train "B" of the residual control (I&C) technician removed two (2) fuses in solid state protection of a n_dification. The fuses were inmediateLy replaced when the technician heard a relay activate. The de-energized circuit caused the train "A" RHR suction isolation valves XVG-8702 A & B (one valve in each RHR train) A-65 p. 4 l,_t OATABAS_" ESFAS O8/2?/93

PHASE 2 PLANT NAN[ EVENT INITIAL PLANT RECOVERY CATEGORY DATE CONDITION TIHE

to close. Operations personnel immediately restored train "B" RHR to service after the valve ctosure. The cause was detemined to be drawing errors. At 1700 hrs during performance of the same modification on SSPS cabinet XPN-7010, a similar RHR isolation occurred via the train mB" RHR suction isolation valves XVG-8701 A & O (one vatve in each RHR train). The l&C technician uas Lifting Leads affected by the modification to prevent a repeat of the previously mentioned isolation vhen a defective fuse holder interrupted power to the train UBm circuitry. Operations personnel immediately restored train UB" RNR to service after the valve ctosure. To prevent a potential recurrence, the Licensee initiated a draying revision and replaced the defective fuse holder on Oct. 9 and Oct. 10, 1984, respectively.

ESFA$ Surly 2 08/18/89 OZ, Unit 2 CSO Power |eve| - O'A. On 8/18/89 with Unit 2 at cold shutdmm {CSD) at 1010 hrs, 3 motor operated valves (NOVs) in the safety injection (Sl) system actusted. The vatves are designed to repesition Nhen a rectrcutation mode transfer (RMT) signal is generated t4:on a Low Level condition in the RUST. Hoverer, no toy tevet existed at the time. This spurious _ actuation constitutes an ta_ptanned engineered safety features (ESF) cceqponent actuation and was reported to the NRC per lOCFRSO.72(B)(2)(II). An electrician inadvertently energized a relay that actuated the valves white placing a jumper on an adjacent terminal in support of an engineering work request. A human performance evaluation system (HPES) investigation was conducted and 8 report prepared. Recollmendations mmde in the report HiLL be evaluated and q=p¢opriate actions taken.

A-66 p. 1 hER DATA BASE - ESFAS/S| 08/27/93

PHASE2 PLANT _ EVENT INITIAL PLANT RECOVERY CATEGORY DATE CONDI TI ON TI HE

• --..--?- ...... _ T_-.-.-....-...... ESFAS/SI Surry 1 04/26/80 CSD Maintenance Power Lave| - OX. On 6/12/85 at 1735 hrs with the unit |n refueling shutclown, a spurious safety injection occurred during PT 18.2A (SI functional test). The procedure contains 1 step that executes 2 actions, reset Ind then block SI. This Led to the spurious SI. The Si function test for units 1 and 2 uiLL be modified to provide separate steps and s|gnoffs to reset and block SI.

ESFAS/SI Surry 1 03/01/84 CSO Pouer Level - OZ. On Narch 1, safety injection signals were initiated as a result of completing 3 of 4 high contairaant pressure signals and campLeting 2 of 3 high stem ftou signals. At the time of the event, qperators were performing NOP 26.9 (removal of vital bus sole transformer l-I) when vital bus i and III Mere mistakenly cross connected out of phase. This reuatted in a voltage transient on vital bus I lind Ill, Nhich caused spurious contatraaent high pressure and high steam flow signals. The voltage transient in v|tat bus X and Ill ts beLieved to have resulted in tripping 2 of 4 containment high prmure relays. Since channel I! was in trip prior to the voltage transient, the 3 of 4 matrix for col_ta|ramnt high pressure was ¢aq)leted and safety injection was initiated. ALso, the power transient resulted in resetting the high steu flow tow tovg or Low stem pressure safety injection circuitry, since high steam flow signals were also generated Mith voltage transient and both tou header pressure Low tavg wore present, safety injection Mas actuated. The qperator ramsre-instructed in the correct manner of removing the vital bus sole transformer. LabeLs have been nmde for both unit's emnuat transfer switches. Fan 588 iris diper shaLL be run in the a_emattc mode.

ESFAS/SI Surry 1 11/16/84 RefueLing SO Pouer Level - OZ. With the unit in a refueling shutclmm condition, a spurious safety injecttan oocurre_ when returning CLS-hi (equivetant to amtainmnt delprsasurtzatien ectuetion system) to service. The contretting procedure was inadequate in that the procedure did.not require the resetting of CLS-hi. The impact on the unit ms minimal and the unit ms returned to a normal condition. Outstanding work procedures, that required the de-energizing of* CLS-hi, _ere held in abeyance until the procedures could be modified.

ESFAS/Sl Surry 1 05/11/86 Shutdown Poser Level - OZ. On Hay 11, 1986 at 0841 with Unit 1 shutdmm and on the restcluet heat removal system and Unit 2 at IOOZ power, Unit 1 experienced a teq:orary degradation of °g° (k:: bus voltage. The decrease in a8° _ _ voltage caused OB' train reactor amntmicator panels F-K to became inoperable. Safety injection actuation caused the pressurizer Level to increase and exceed the tech spec Limit of 33Z. The event and its cause were diagnosed and safety injection terminated by 0843. Pressurizer Level peaked at 42% and decreased to tess than 33:; by 0847. RCS pressure remained constant during the event. The 'B' dc _ was restored to fuLL voltage at 0900 and the annunciators returned to service at 0905. The cause of the event was an inadequate procedure in a design change to replace the °B0 batteries. Design changes for replacing the other station batteries wiLL use a revised procedure to preclude recurrence.

ESFASISl Surly 1 06/0S/89 CSD Power Level - 0%. On June 5, 1989 at 1141 hrs, with Unit 1 at cold shutcioun, an unplanned initiation of an engineered safety feature (ESF) occurred during preparations to coamence a special test. The initiating signal was a Low steam generator (S/G) Level trip that resulted in closure of the S/G btc_cio_ trip valves, opening of the auxiliary feedmater (AFW) valves, closure of the motor driven AFW Ixnp breaker and opening of the steam driven APJ pump stei supply valves. The ESF actuation occurred while simulated S/G Level signals were being transferred from one channel to another. The test directors were aware of the potential consequences of this action, I_Jt did not inform the shift supervisor. The test directors were counseled as to the necessity of keeping the shift sLq=ervisor cogF_iZant of testing activities. p. 2 _ER DATA I_ASE - ESFAS/S! 08/27/93

PHASE2 PLANT NAME EVENT ! NI T!AL PLANT RECOVERY CATEGORY DATE COND! TI ON T! HE ...... ---:.- .... -- ...... - ...... ESFAS/S! Surry 2 06112185 Refuel ! ng SU Pouer tenet - 0X. On 6/12/85 st 173S hrs w|th the unit in refueling shutdown, s spurious safety injection occurred during PT 18.2A (SI funct|onat test). The procedure contaim one step that executes 2 actions, reset and then block Sl. This Led to the spurious Sl. The SI function test for units 1 and 2 wilt be aodtfled to provide separate steps and signoffs to reset and block SI.

ESFAS/SI Surly 2 11/16/86 CSD Pover reval - OZ. On Nov. 16, 1986 with Unit 2 at cold shutdovn, during _,e performnce of periodic test 16.3 (type 'A' test), a hi consequence Limiting safeguards (CLS) signal was generated resulting in the unptannad closure of the containment radiation monitoring trip valves TV-RI4-200A, 2008, and 200C. This event was due to an inadequate procedure Nhtch did not instruct test personnat to cross the trip valves affected by the hi CLS signal. The procedure t_ttL be modified to correct this inadequacy. This report is rattled pursuant to 10CFRSO.73(A)(2)(IV).

ESFAS/SI Surly 2 09/08/89 0_; Unit 2 CSD Po_r tevet - 0_. On Sept. 8, 1989 at 2142 hrs, with Unit 2 in cotd shutdotm, during perforllance of a periodic test on the turbine I:u|tding ftood controt circuitry, two of the four condenser waterbox circulation water (CY) inlet isolation valves closed unexpectedly. The condenser inlet valves ere designed to close upon the initiation of a hi hi consequence timittng safeguards (CLS) signet in coincidence with a Loss of offsite power; however, no actual hi hi cLs signal uas present. The event |s being reported as an unptlnn_l engineered safety features (ESF) cempment actuation. A four hour non-mmrgency report ms _ to the liRC per 10CFE50.72. The event uas caused by • retay in the flood control circuit that did not drop out as required during testing ilhlch resulted in actuation of the valves. It I determined that the original relay had been replaced with a irte_ relay Idhtch required 1 Less hotd-tn current causing this event to occur. An investigation wilt be conducted to determine h_ the r_ relays were installed.

A-68 Appendix A.2.6 Loss of Instrument Air

A-69 p. 1 I.|R DATA BASE - AIR 09114193

PHASE 2 PLANT NAME EVENT INITIAL PLANT RECOVERY CATEGORY DATE CONDI T!ON T| HE

o- _------.- 0- _. _. .,---.,-----...... ,.,..,...... -....------,.------,------.... " " - " "- - -'---''" ------

AiR Beaver VaLLey 1 08129185 Hode 1 10 mtn On 8129185 at 1248 hrs, a Low •tartan air pressure alarm teas received. Operations personnel responded by starting the standby station air compressor. The reduced air pressure initiaLLy caused one main steam isolation valve to drift closed. The resultant increased steam fLou caused steam Line pressures _ to drop in the other two steam tines. The pressure drops Mere sufficient to actuate the rate-coroNa•re•ted tom •tesmtine pressure safety injection sigrmt. The safety injection signal caused a resultant reactor trip. The control room operators fottoued the q=pttcabLe emergency procedures and stabilized the plant in hot standby. An unusual event .as declared at 1300 hrs and term|nated at 1315 hrs. The tom station air pressure was the result of a failed solder fitting on the instrument air system. The solder fitting failed due to a faulty heater control on an instrument air dryer and |mproper equipment restoration. Dur|ng sube4quent plant recovery actions, Mater .as found spraying from both tom head safety injection Ixaap wedge control rod •eats. Both pumps were declared inoperable Nhtch required am entry into cold shutdoen. PostuLated failure on the control rod seats was from a mlnor ftou Induced pressure transient. FoLLowing equipment repair, a plant startup to fuLL power operations ms co|anted on 9/1105.

AIR CaLLa,my 11/05/84 Node 1 20 sin On 11/5/84 •t 1156 CaT and 11/6/84 at 0450, ,tth the plant in m 1 at 45X IsoMer and 16_ pouer, respectively, air tines sqPptytng the SG feeckmter regulating valves foiled which caused ESF actuations along wlt_ reactor trips. The f|rst event ,as due to a failure of the a|r Line connection to the feedmter control valve, FCV-531), for $6 'C'. Upon the toss of air supply, FCV-530 fai ted closed resulting in a to-to Level in Sm 'C' ,htch caused • reactor trip, turbine trip, feedHater isolation s|grmt (FWIS), ate( feed_tor actuation signal (AFAS) and SG btoudoun isolation signal (SGIBIS). The second event ,as clue to a failure of the air tins to the 'A' feedmter Ixaap recircuLatton valve FV-ZB cmJstng it to faLL open and closed then open. The resultant roadster flow osciLLations produced s high Level in SG 'C' cma_ing a turbine trip, FWlS, AFAS and SGBIS. Due to the FWIS and AFAS, SG LeveLs began to decrease end at 0454 a to-to tevat on S6 'D' caused a reactor trip. The air Line failures tare due to Imprqper amtartat aq=pLtcattons and resulting fatigue cracking caused by the vibrations imposed on the air tines during fesdwter •ysta operation, Temporary repairs using copper or stainless steel tubing were made and design changes to instaLL hangers are being made in accordance with a_kainistrstiva procedures.

AIR CaLvert CLiffs 1 01/27/87 Node 1 20 sin Power Level - 100%. At 1809 on Jan. 27, 1987 the unit ues manusLty tripped due to decreasing steam generator LeveLs, which resulted from a Loss of instrLaaent air. At 1816 the reactor coolant pumps were stopped, because coating Mater .as isolated upon the toss of air pressure. Reactor coolant system temperature Mes amintained by the use of auxiliary feed_ter and manual control of the atmospheric dump valves. At 1825 instrtment air pressure ,as restored, and t,o reactor coolant pumps Mere started. At 1840 the remaining two reactor coolant IXmPS were started and the unit ,as maintained in Hode 3 (hot standby). To prevent recurrence of this type of event on engineering analysis is being performed to |denttfy modifications that can increase the reLiabiLity of the instrtment air system, and the general supervisor-operations is revieuing this event with aLL operators.

AIR Cats,be 1 01/14/85 Node 2 15 min On 1-14-85, at 1440 hrs, the Unit 1 reactor ,as manuaLLy tripped. The reactor coolant pumps had been previously shutdoun due to the Loss of motor cooling water, and the ability to control unit reactivity was therefore tess than desirable. The Loss of motor cooling is attributed to a malfunction of service Mater valve 11tN-A83, Mhich opens to supply an alternate source of cooling Mater to the reactor coolant pumps ,hen the nomst source is not avaiLabLe. On 1-14-85, a vendor was on•its to driLL additional welts associated ,ith the cathodic protection system. WhiLe driLLing, an instrument air system Line

A-?O p. Z [.ee OATA aA_ - AIR 09/14/93

PHASE2 PLANT NAHE EVENT INITIAL PLANT RECOVERY CATEGORY DATE CONDI TION Ti ME

sam drilled through. The Vl Line t_tch Mas ruptured supplies control air to the cont. chilled Idster control panel (1YV-CP-1). The Loss of control air to 1YV-CP-1 caused the YV chiLLers and I_ to shutdown. This Iytm supplies the normal source of cooling water to the reactor coolant pump motors. The 4 service water swap-over vItves, ulhlch supply an atternIte sitar sourÙera ire supplied by a different Vl header than 1YV-CP-1. The iuIp-over valves therefore did not attempt to realign Is they did not experience a Loss of control air. ReaLizing that a sbuq)-over to service uniter had not occurred, the NCO realigned the YV control switch from auto to service sitar I:ut did not receive Indication that a swap-over had taken place. Verification revealed that 1RN-A83 was stILL closed.

AIR Cook 1 11/25185 IIode 1 20 Itn On 25 Nov, the reactor tripped white surveiLLance from the nuclear posit rlunge Instrument N-41 tll constant rate trip circuit was tn progress. The surveiLLance required the blstabtes for Ii-41 to be tripped. That resulted in a standing quadrant poser tilt (OPT) alas. Tech Specs required that the OPT be caLcuLated with Less thin 4 power range |rItItl In service. During the procesI of talcing eLectricaL current rsed|ngs froi the other power range channels, to perforI the tilt caLcuLations, a spike occurred on N-44. The spike I found to have been caused by rInoveL of the Iter probe used for taking the current readings. The Iptke tn power range trtstrIt N.t_i; resulted in the 2-out*of-4 Logic required for a rate trip.

FoiLÙsing the reactor trip w|th the turbine driven aux FU _ (TOAFP) operating It Its norIit Ipead, the turbine uouLd not reslxmd to c_trot Itlllats to reduce speed. The TOAFP ues tr|l=ped and SG Level I controLLed using iIotor driven aux FW IXm. Local control of the TOAFP could have been taken tf tt been rlqu|r_KI. Dur|ng the |nvestigatton, tt t discovered that the control air tubing to the TDAFP governor had broken off. The cause of the tub|ng fat Lure I attrtl:uted to |ti Location. It I

believed that perllonnet had _ the tul• _, .ing as I .hand.., hold whiLe checking TDAFP governor oil LeveL.

To prevent recurrence of the reactor trip, q_erattng procedures were changed to prohibit the use of a Iter probe |n an operable p(xaDr range drsuer for current readtnII _han the bistabLes were tripped in another pourer range drawer for survetttInce testing. To prevent recurrence of the control air tUbing fai Lure, a design change Has completed that tarÙuteri the tUbing to prevent Its being used as a hand hold.

AtR Davis Besse 1 12/07/87 Mode 1 31 Itn On 1217/87 the unit experienced a reactor trip at 0656 hrs from 81% reactor therIIL power. The |nit|sting event I I toss of |nstrIt Iir pressure idlich caused Ieverat secondary syatell valves to go to the|r Is|ted position. Reactor power _ncreised to the integrated control system (ICS) high deIand LtIiter, flier flow |ncreesed causing reactor coolant systI Tare to decrease. Due to a Large Ixlerator temperature coefficient, reactor power increased to the reactor protection systel high flux trip setpoint. The post-tr|p plant response was norllIt except that stem generator 1.1 pressure was stilhtty Lower and stem generator 1.2 Level was higher than the expected post trip ranges. Operator actions were required to close the misture separator reheater (MSR) second stage reheat stem source valves and to stabilize stem generator 1.1 water LeveLs. The Loss of instrument air pressure was caused by direct venting of the instrument air header to atmosphere when s solenoid vatve failed on instr/t air dr_/ers 1-1 and 1-2. This solenoid valve was repaired. The HSR second stage reheat stem source valves did not ÙLose due to a pressure suitch failure. This switch was replaced. The lOS did not respond fast enough to stem generator 1-1 decreas|ng water LeveL. A modif|cation to the ICS is scheduled.

A-71 p. 3 LERDATABASE- AIR 09/14/93

PHASE2 PLANTNAME EVENT INITIAL PLkilT RECOVERY CATEGORY ,DATE CONDITION TIME

AIR Nc_tre 1 11102185 Node 1 20 min On 11-2-85 •t 0660 • section of brstded, fLexibLe pipe on the discharge of instrument air compressorB rL.,mturedat • ueLdedseam. As a resuLt, aLL V! Loadsnot protected by check valves experienced ciacressad V! pressure. The LowV! pressure cl_usedthe fe_bmter control valves on each unit to begin to close, causing iKi LeveLsto decrease. At 06&1, Unit 1 experienced • resctor/turbtne trip on IG tA tou-tou LeveL. Pressurizer pressure dropped beLouthe safety Injection setpolnt (1845 psi|) initiating St. The pressure dropped ns partiaLLy clueto a higher than normL stem Loadswith somevalves failing open on Loss of Vl. Unit 2 tripped on SG2A Low-LowLeveL, but pressure did not decrease to the St setpolnt. Both units uera at IOOXpouer at the tim of the incident. VI cmqx.ossor B was isolated and Vl pressure returned to normt. The ruptured pipe ties replaced. Several nodtftcatlons era being revleued for possible lint.met•tiDe.

AIR N|tLetone 3 04/17./87 Node 1 20 rain On Aprl t 12, 1967 at 0618 hrs, _htte ,par•Ling •t 66g _ in Node 1, • reactor trip occurred due to Lw Lowstem generator Level in the O steam generator. The control roomq:eretors were in the process of increasing pewer during the Inttt•L stertup foLtoutng • scheduled mintanance outage. Approx. 15 minutes prtor to the trip, the qxwstor_ had tnermeed reector IXxver frem _ to Md_utthtn • 16 minute intervaL, tdhenat _ IXxier the speed controLLer to the A turbine-driven fosck_ter puap (31_IS-P2A) began to osciLLate. The respomlbLe operator osetmedmnuat controL of the pumpin an sttMIpt tO StOp the osciLLations. Ne tree unabtLeto do so before the Level in the O steam generator feLL to the Low Low Level and the plant tripped. The eperotors verified the qxming of aLL resctor trip breakers and fuLL insertion of eLL control rods. The root cause of the reactor trip is equipment faiLure. The prtsery cause tins an air Leek in an air supply Line to the D fsed_ter regulating vatve, uhlch resulted in Low tou steam generator LeveL. The Leakhad been fixed. A contributing cause wes •sol t tettone in the A turblno-drtylm f ee_ter p_. controLLer. A vendor represent•tire tm consuLted and found no probLem with the pumpcontroLLer. It respondedcorrectLy d_ring • pL'amt stortup cmApril 13, ;IM7. No other u_"k is planned.

AIR Ocomm3 08114184 Node 1 20 ale on Aug. i4, 1984 at 1126 hrs, Unit 2 tripped from 1002 fuLL pmmr (FP) _hm the instrument air Line to the POkl)EXoutlet vatvas tins accidentaLLy sheared. The Loss of air to the outlet valves caused the valves to felL shut uhlch resulted in • Loss of condensate ftou to the c(mdemate booster (CB) I:Ual_. The (:8 Ixaq_ tripped and caused the main fsedmter (MFIN) I_ tO trip on Lowsuction pressure. The toes of the NFDWpimps initiated a reactor anticipatory trip. Al=Prox. 16 minutes after the trip, the =3Au IIF1N ptlp was restarted to reestabt|sh MFDUflow. The emergency feecklater (EFDU) control valves, 3FIN-315, 316, closed on an indication of 7"=0pstg discharge pressure from the "3ASMFDUpumpas designed. The once through stem generator's (OTSG's) were isolated from aLL fesdnter flow for appro0_.9 minutes as 8 result of the eutOIMttC cLos|ng of the EFI_ control valves and Lack of MFDWftw due to insufficient discharge pressure from 3A MFIN pump. The into.diet, corrective action uas to stabilize the un|t in • hot shutdolm condition using emergencyfeed_ter (EDFU). The inroad|ate corrective action to restore fecd_ter fLos to the OTSGwes to mmusLLy open the ER)U co•trot vaLves and to increase 3A MR)Upumpspeed. The sheared etr Line uas repaired. The unit ,as restarted and reached IOOXFP at 0100 on Aug. 17, 1984.

AIR Prairie IsLand 1 05108/85 Mode 1 20 mtn on Key 8, 19(5 at about 1323, • Z-inch copper tnstrtJm_t air Line seperted at a soldered eLbowjoint (PSF). As a result, the Unit 1 side of the instrument air system (LD) deprossurtzed enoughto cause the fesd_ter regulating valve (FCV) to No. 12 stm generator to cLose. At 1326, Unit 1 tripped on Low stem 9an,tatar LeveLplus feed_ter flow/stem fLow mismatch. CabLe tray (TY) ,rapping for A-72 p. 4 J_ERDATABASE_- AIR 09114193

PHASE2 PLANTNAME EVENT INITIAL PLANT RECOVERY CATEGORY DATE CONIDTiON TINE "''''''''''''''''''''''''''''''''''''''''''''' ...... ",°,'''" ...... •.... _'.... " ...... '''_*,''' ...... "T'.'" Appendix R compliance was |n progress tn the vicinity of the break, and the air Line was movedsome_at to acciaPLtsh the wrap. This afq:arentty placed enoughstress on the e|bo_ to cause it to separate. The Unit mm mlntatned in hot shutdmm while repairs were made. A watkdovn inspection of the Inetrumnt atr system was performd tn areas of the plant _here cable tray wrapping had taken place or was in progress. No additional problem areas are identified. The Unit was returned to service at 1847 on Hay 9.

AIR 9hearon Harris 1 08/04/8? Node 1 20 rain The plant teasoperating at IOOXreactor po_r tn Mode 1, po_r operational on Aug. 1987. Instrument air dryer 1A was out of service for repairs, lnstrtamnt atr dryer usa bypeaNd at approx. 1715 hrs to replace the desiccant. Workyes completed the drye_ 1S and at approx. 2150 hrs the clearance on the dryer ties rewved. The restoration atlllment on the clearance for placing dryer 1B beck into service was incorrect. _en the valves yore rapostttoned in accordance with the clearance restoration Lineup the instrument atr c_preuors Nre isolated from the air system. The air isolation caused atr pressure to decay end caused air operated valve to 9o to ufalt safe" positions. In particular, the mltn feedkmter regulating valves to drift shut and the heater drain Level control valves to divert drain flay to the condenser. Heater drain Ixmq:s 1A and 18 tripped and a manual turbine rurtack was initiated. This ms foLLotmdby a trip of rain fesd_ter pump 1A shtch initiated an autmtic turbine rtlrdbKk. The |DiS of tnstrtslMtt sir resulted in a decrease |n stem generator water Levels. The turbine rurVbackresulted in shrink tn stem generator Levels and resulted tn a reactor trip at 2154 hrs due to steam generator feedmlter |tesmVftov mlemtch with Low stem generator rater Levels.

AIR Surly 1 01/07/80 Node 1 10 sin On 1-7-86, an 'mJ_. ventilation system safety rode initiated' slam was received in the control roomat 2059 his and the aux. ventilation eWril_-y fans, 1-VS-F-S8A and 588, auto started. Within seconds, trip vatm TV-1204 and TV-CC-107closed, whi Le the operators atteq_ted to open the trip valves, they noted Instrtanlt air header pressure decreasing _ feed flows decreasing, lmledlateLy, the operators ettampted to open the feed reg. valves (FIW). However,due to the decrease of tnstrtamlt air (|A) pressure, operators could not prevent the FRWs from fully closing and at 2104 hrs s reactor trip occurred as a result of a Lousteel generator (S.G.) Level coincident with feed flow Less than stem fLo_ signal. The dacrem of [A pressure was due to ice formation in the Unit 1 XAdryer that resulted tn blocking a|r flow. A hot gas bypass valve tn the dryer was not properly adjusted. The Unit 1 IA dryer was the source of the problem and was bypassed. IA pressure yes returned to natal and the affected systm and componant_were realigned. The bypass valve was properly adjusted by a service representative. Turbine but|dtng Logswilt be revised to incorporate IA dryer condenser temperature readings.

AIR Yankee Rove 10/04/86 Node 1 10 rain While operating at IOOZ paver in Node1 at 0917 on Oct. 4, 1986, the No. 3 (main) control air compressor air pressure switch CA-PS-453 failed. As a result of this failure and the subsequent inability of the back-up air compressors to rapidly restore the a|r pressure, the air header pressure decreased sufficiently to Lock(by design) the feedNater control valves (FCV-FW-IO00, 1100, 1200, and 1300) into the as-is position at 0918. Subsequently, the operators initiated steps by procedure to resetrestore the fesd_ter control valves to normal. As soon as the first feedwater control valve was reset (unlock) it began to rap|dry open. This preferential feeding of No. 1 steam generator caused the Levels tn the other three stem generators to rapidly decrease, resulting tn a reactor scrm on Los steam generator Level at 0938. The root cause of this event is the failure of the air pressure switch (CA-PS-4S3) coupled w|th recovery procedure gu|dance inadequacies for resetting of the fearer control valves t4)on Lock-up. Corrective action included repair of the failed equipment and cLar|ficatton of procedure |nstructions. ALL autanatic safety systems functioned as required. The A-73 p. S LER OATA BASE - AIR 09114193

PHASE2 PLAJITNAME EVENT INXTIAL PLANT RECOVERY CAT_GORY DATE CONDXT|ON Ti ME e_ae e*nweeeeeeeeeqel*_wempeeeeeee oae_ e._emm _ e,e_ e nee e eeaa Q a a eee _ee we eeeee_ _ eooe*e _eo e a e e q* eee ee*lee e ee Q _ * a *e --e* a faulty air prusuro switch for N0. $ control a4r ¢_prossor was replaced.

A-74 Appendix A.2.7 LOCAs (HCVCS, ItRIIR, JCVCS, JRItR, KRCS)

A-75 p. 1 LER DAT.A,,B.ASE- LOCAs 08/27/93

PNASE 2 PLANT NAME EVENT INITIAL PLANT RECOVERY CATEGORY DATE COND! TI ON T!NE

,_...... ,... o.. _...... _. _l,, ...,,...... ,...... NCVCS Tu_ey Point 3 01115188 being cooled down Pouer Level - 0¢. On 1/15/88, Unit 3 was being cooled doun and depressurized. Pzr spray valve PCV-3-4$SB eas identified as having erratic operation on 1/11/88, and a plant work order to correct the problem was _s.sued. With the reactor coolant system (RCS) temp. at 400 degrees F anti pressure at q3prox. _0 1:sigt it tats decided to sL¢_ the cootcloen rate from 90 deg. F per hr to Id:K_Jt 20 ¢1_ per hr, clue to the shortly q=coming shift turnover. At 0650, as the cootdoun rate uas being decreased, the pzr Level started to increase. At this peinte the reactor control operator (R(O) se_'JJred charging 3A. An adjustment Mas _ to valve PCV-3-455B 8t 0650 in order to decrease RCS pressure. 8y 0730, shortly after turnovere RCS pressece decreased to _ psig at¢l the acctmuLators started to inject. Upon noticing the RCS I:¢essure drop, an R(O Jmediatety closed PCV-3-&55B and teminated the RCS pressure ¢lecreese. Al:_)rox. 65 gallons of water t,ere injected into the RCS. The primry cause of the event uas a defective controller for valve PCV-3-455B. Contributing causes uere the It(O's failure to note the decreasing RC3 pressure in tim to take prompt corrective action, and a miscomuunication betueen the o_¢oadng and the offgoing RCO's. The controller for valve PCV-3-455B was replaced.

HRHR SecFJoyah 1 02/11/81 The residual heat rmovat (RHR) containBent WaY Iras inadvertently initiated when an assistant unit operator (AUO) opened verve 1-FCV-72-40, ehich i_tates the RHRsystem from the containment Sl:cay header. The spray continued for approx. 35 minutes releasing approx. 40,000 gallons of primry Mater and (6,000 gallons of refueling Mater storage tan_ Mater to the containment bldg. The UO received alarm indicating a rapid decrease in pressurizer Level and pressure. The UO notified the shift enginear (SE) of the condition and then tripped reactor coolant pumps 1 and 2 (puq_ 3 and 4 eere not running). The situation Mas diagnosed as 8 possible Loss of coolant accident (LOCA) and emergef_-y q:ereting |rlstruction (E|O) 0 and 1 Mere consulted. The AUO, 1411oopened the isolation va|ve, entered the control room w|th another UO discussing the vlmtve. At this tim the control rosa (mptoyee checked

the indicator Light and.....verified it Mas indeed open. The valve was shut _ |P-4 m teminated.

(frm IISAC2, p. A-13) An _RS centainment spray isolation valve eas inadvertently opened creating a Loss of coolant accident (LOCA) from the RCS via the RHRS to the container spray header. The pressurizer m_ptied _ RHItS I_P suction was essentially Lost in al_ 2 mimJtes. The LOCA Lasted for 39 minuCes, during which (minute 10) the RWSTbras valved into the RHRS in an attempt to taker4) for Lost inventory. But the RHRS suction valve from the It_ eas Left o1_, permitting the LOP,A to continue. ALso, residual RCS pressure overcame the pressure heed of the I_T, Limiting the supply of makeup wter. High pressure safety injection ms ilmLty initiated after 35 minutes. A total of approx. 40,000 gallons of reactor coolant and (_),000 gattors of RIJST water was sprayed into the contairelent.

JCVCS Robinson 2 01/29/81 A spurious safeguards actustion ar¢l reactor trip eere received resuttir_ from a high stem ftos spike caused by turbine governor vatve/E-H oi t probLeluS. Letdoen ftou was restored. RCS pressure began decreasing uith containment pressure and deu point increasing. Upon containment entry fotLouing tetdoen isolation, a tetdoen Line drain vatve (Cl/CS-2OOE) vas found partially opened and Leaking through with its pipe end-cap missing. Cause is the valve vibrating open and the pipe cap, Mhich uas serving as a pressure boundary, failing to hold at saaetime fottouing the first St. Approx. 4500-6000 9aLtons of Mater Leaked to the containment sump during the event. Letdown Mas secured. The va|ve was secured in the closed position and a nee rap was installed.

JCVCS Waterford 3 12/16/85 pLnt heatMp Power Level - 0_. On Dac. 12, 1985, Waterford Stem ELectric Station Unit 3 m condIJcting a p|ant heatup fottouing an outage. At 2340 hrs the plant was at /,44 degrees _ 1490 psi8 when a reactor coolant system (RCS) Leak was indicated inside contaimlent. The chemical and volume oontrot system (CVCS) uas isolated at 0007 hrs and the Leakage stopped. At 0040 hrs an ir_4=ection of the contairment

A-76 p. 2 LER DATA BASE- LOCAs 08127/93

PHASE2 PLANT NANE EVENT INITIAL PLANT RECOVERY CATEGORY DATE CONDI T%ON T!HE

determined the Leakage to be from relief valve RC 526 on the temporary reactor coolant pump Seal injection system. The system was isolated and the CVCS was restarted at 0046 hrs. A subsequent calculation verified RCS leakage was Less than one gallon per minute. Since there was no operable beration flow path or charging path while RC 526 was opened, the plant was in a condition prohibited by technical specifications for approx. 29 minutes. The apparent root cause of the event was the failure of utility Licensed operators to document the status of this sytem during shift turnover. The shift turnover procedure has been revised to provide for a 24 hr format turnover sheet for all licensed positions. This wilt facilitate review of turnover information from several previous shifts. Since, due to plant c(x_l|tions, the boration capability of the charging system would not have been required dnJring an accident, this event did not affect public health and safety.

JCVCS Yankee Rowe 06127/86 Node 5, Kaint Out Power Level - OX. On June 27, 1986, at 0137 hrs during a maintenance outage with the plant in Node 5, main coolant was inadvertently drained to the Low pressure surge tank (LPST). This could have resulted tn a loss of shutdown cooling. This event occurred while transferring to the alternate method of shutdovn cooling par procedure 0P-2162, Attachment C. Performance of this procedure was necessary because of the failure of the shutdoNn cooling pump's shaft seal. During the evolution, approx. 2000 gallons of rater was drained from the pzr and main coolant pressure dropped from 100 psi9 to 10 psig. The pzr did not empty. The control room operator (CRO) imediatety secured the LPST cooling pump and the primary auxiliary operator (PAD) isolated the ftou path. The CRO started all three charging pumps and restored pzr level and pressure. The root cause of this occurrence has been attributed to personnel error. While conclucting the alternate shutdmm cooling valve lineup, CH-V-654 was not fully shut, which resulted in a main coolant system to LPST flow path. The PAD thought that the valve had completed its full travel when he operated the manual valve. Th|s occurrence was reviek_d with the appropriate plant personnel and the need for strict procedural compliance was emphasized. This is the first occurrence of this nature. There Mere no adverse effects to the public health or safety as the result of this occurrence.

JRHR Braichocd 1 03/25/88 OX power level Pokier level - OZ. On Harch 25, 1988 and tlarch 27, 1988, operators noted a decreasing voltme control tank level which caused increased make-up. Reactor coolant water inventory balance surveillances confirmed that unidentified leakage was in excess of 1 gl=m. The source of the March 25, 1988 occurrence was thought to be an improperly locked closed valve which was inadvertently bumped off its closed seat. The residual heat removal (RHR) pump suction relief valves may have contributed to the generatirvj station emergency plan unusual event for both occurrences. Leakage past the seats by measuring the downstream temperature indicated the source of leakage. Subsequent investigation of one of the relief valves indicated that the disc insert pin was broken as a result of improper nozzle ring setting. The 1A RNRsuction relief valve has been repaired and reinstalled. The 1B relief valve will be tested and repaired as necessary prior to restart of the unit. There have been no previous occurrences of Crosby relief valve failures.

JRHR Braiduood 1 12101189 O_ power Level Power Level - OZ. At 0142 on 12/1/89 whi te drawing a bubble in the pzr, reactor coolant system (RCS) pressure sto_Ly increased from 375 pai9 to 404 psig. At this time the 18 residual heat removal (RHR) pump suction relief valves which had a setpoint of 450 psig, actuated and remained open. Charging flo_ Mes increased but pzr Level indicated 0"_ by 0151. Reactor operators concluded that an RHR pump suction relief valve had Lifted because hold up tank levels Mere increasing rapidly. The operating train of RNR, 1A, was isolated at 0155. At 0200 RCS pressure reached 272 psi9 and stabilized. RCS level was at _ the lamer portion of the pzr surge Line and flow into the RCS was equal to the flow exiting the RCS. _ At 0215 the Licensed supervisors decided to return the second charging pump to service per 10CFRSO.54(X). At 0235 the second charging pump was star:ed. At 0245 pzr Level was above OZ. At 0319 A-77 p. 3 LER DATA BASE - LOCAs 00/27/93

PHASE2 PLANT NNqE EVENT INITIAL PLANT RECOVERY CATEGORY DATE CONDI TI Oil TI NE

field reports identified that the 1B RHRpump suction relief had actuated' At 0350 the "B" RHR train bins isolated which terminated the event. Approx. 6L+,O00 gallons had relieved through the valve. The cause of the early Lift was dirt between the valve spindle and guide sleeve which affected valve Lift setpetnt adjustment. The cause for the valve remaining epe_ was an incorrect nozzle ring setting due to a personnel error.

JRNR NcGuire 2 08/05/84 Og p_er Level Power Level - OX. On 8/6/84, HcGuire 2 operators discovered a broken weld on the residual heat removal (RHR) system Letdown Line to the chemical and volume control system (CVCS). The RHR system was in service at the time, and water was spraying from the broken pipe and from the stem of the valve (211V-121) in the CVCS. An estimated 3000 to 7000 gallons of contaminated water was contained in the heat exchanger room, the RHRand containment spray sump, and the B floor drain sump and tank. Upon discovery the Leaking Line was isolated, a subsequent inspection revealed a runber of supports/restraints (S/RS) damged, and the broken socket weld completely separated. On 45/85, 7 socket welds w|th crack indications were discovered on the 2-inch crossover piping between valves 2N0-17 and 211D-32. No ueLd faitured occurred as in 8/84. Six of the welds with indications were in piping reptsced after the failure, but before vibration testing foLLc,ding the 8/84 failure. Causes of the August event are attributed to ccxq:xxlmt maLfunction/fai Lure, due to Loose packing in 2NV-121, and to an unusua| service conditim, because a gas void in the Line resulted in water hammer. The cause of the April event is classified as an unusual service evont, because socket welds were not designed to w{tnstand the severe Loading of the vibration testing that occurred in 8/84. Damaged piping and S/RS have been replaced. The subject valves have been modified.

JRHR Summer 1 0S/06/85 C30 (Node 5) Power Level - 0_. On Hay 5, 1985 at approx. 22 hrs, a reactor coolant system (RCS) pressure transient resulted in a challenge of a residual heat removal (RHR) suction relief valve. The plant _as in cold shutclatm (Node 5) with RHRsystem (Train "A") in operatioh. Diesel generator (D/G) surveillance testing was in progress and had resulted in a non-valid test fai Lure during an atteiapt to parallel the D/G to the ESF bus (XSW-IDB). The failure to parallel the D/G was a result of failure of the speed control smirch on the main control board. During troubleshooting activities on the D/G, a personnel error resulted in a Loss of ESF bus (XSU-1DB). Hajar equipment affected included the Loss of the "B" ccaqxwmlt cooling water (CCb/) pump, "B" service water (b'_) pump, and "B" NVAC chiller and chill water pump. The toss of CCU flow to the reactor coolant pump (RCP) required the shutdmm of the operating RCP. The breaker was rectosed to ESF bus (XSW-II)B) and the bus was reloaded. Upon restart of the RCP tdith solid plant operation, pressure spikes occurred idlich resulted in the challenge to the train "A" RHR suction relief valve. Following the relief valve actuation, an operator noted that pressurizer relief tank (PRT) Level continued to increase apparently due to a failure of the relief valve to reseat. Approx. 1600 gallons of RCb inventory were released to the PRT.

KRCS North Anna 1 06117/87 O_ power, refuel. Power Level - OX. On June 6, 1987, during the 59th day of a refueling outage with Unit 1 in Node § and Unit 2 in Node 1, a problem developed with the Unit 1 'A' reactor coolant pump (RCP) motor which required motor replacement. The RCP motor was uncoup|ed at 0420 hns on June 18, 1987, resulting in a small but expected Leak up the pump shaft of several gallons per minute. It was believed that the makeup flow rate to the reactor coolant system (RCS) compensated fan this and other inventory toss because pressurizer Level was held at approx. 20_. However, at 0130 hrs on June 21, 1987, it was discovered that pressurizer pressure was subetmospheric, and as a result, pressurizer Level was not a reliable indication of RCS inventory. RCS inventory had decreased by approx. 17,005 gaLtms. There was no |mpact on safety because the residual heat removal system remained in service throughout this event. Corrective actions have been identified to address procedural, training and RCS inventory indication inadequacies during Node 5 operation. Although there was no impact on safety, and it is not reportable under 10CFRSO.73, this report is being submitted as a voluntary LER due to the potential for A-78 t p. 4 LER DATA BASE - LOCAs 08/27/93

PHASE2 PLANT NAJI4E EVENT INITIAL PLANT RECOVERY CATEGORY DATE COND! T! ON T! HE

inadvertent rRCS ir_entor_ LOSSe

K:RCS Trojan 07103/81 Reactor coolant system Leakage of approx. 1/, gaLLons per minute was eY43erienced during the performance of s reactor coolant system integrity test. This is in excess of the technical specifications Limit of 10 9mp for identified Leakage. The source was Leaking drain vaLves on the reactor coolant Loops. The type of valve used to isolate the Loop drains mJst be torqued to ensure they seat property. PLant procedures did not specify this requirement. As a result, the valves were checked shut during preoperationaL valve tineups but were not torqued.

A-79 Appendix A.2.8 Transients

A-80 p. 1 LER DATA BASE - TRAHS 08/27/2

PHASE2 PLANT NAHE EVENT |NITZAL PLANT RECOVERY CATEGORY DATE COND1T!ON T; HE

...... ,...... ,...... 0... 0...... 0...... 0...... TRA#1S Surry 1 07/09/81 Routine startup The mximum allowable containment partial pressure of T.S. was exceeded. The unit was shut dotal and brought to cold shutdown in accordance with T.S. recFairments. The containment pressure increase was caused by a feedwater Leak through a failed flange gasket for "A" auxiliary feedwater Line fLou Limiting venturi. The high tu_ereture viton gasket used was not compatible with the slip on flanges used. The gaskets for all three AFWventuries were replaced with ftexatatic gaskets.

TRAMS Surry 1 01/05/82 Routine SO While removing the unit from service, a spurious Sl s|gnat was generated. Trip valve, TV-CC-1099, failed to close as designed. The caq=_nent cooling (CO) system is a closed system and its itnegrity was mm|ntsined (luring the event; therefore, an isolation barrier existed between the containment and the envir_t. A search for an electrical cause for the valve malfunction revealed no problems. No specif|c m_,chanicaL came has been found. Efforts have I_ initiated to further investigate the event but no definitive mrs are available at this time.

TRAILS Surry 1 01106/82 HSO A spurious safety injection caused a phase I containment isolation. Trip valve CC-1098 failed to close as required by tech spec. The cause of the event _ms determined to be a piece of teflon tape (type used to seal threaded air fittings) Idlich prevented aperation of the three way SOY that controls TV-CC-IOgB. Corrective mctions consisted of checking valve operator circuit, establishing administrltive control of the valve, removing foreign mtertaL, testing as per PT t8.68 and returning tO service.

TRAILS Surry 1 06/15/82 Routine startup The reactor coolant systm wes diluted to a boron concentration wherein the critical rod pnsitim achieved, if the control rod assemblies were withdraun in normal sequence, would have been tower than the imrtion Limit for zero I:_r. This is contrary to tech spec 3.12.A.6 and reportable per tech spec 6.6.2.B.(3). Failure to mintsin mkeup boron concentration equal to ItCS concentration resulted in an overdilution. Because of errors in the ECP calculations, the dilution was not detected until the sUlx;equant approach to criticality. The oontrotting bank was inserted and the RCS borated to the required concentration.

TRAILS Surry 1 10/17/82 Routine startup One of two recFJired bor|c acid flow paths to the core became unavailable and bit recirc was tcqllinated due to the temporary toss of boric acid transfer pump 1-CH-P-2A. This is contrary to tech epec 3.2.C.6 and 3.3.A.3, and is reportable per tech spec 6.6.2.B(2). The motor trip appears to have been a random incident. The puq) breaker was reset ard fLou verified to the VCT and blender.

TI_S Surry 1 04107/84 Routine SO I_r Level - 0:_. On April 7, 1986, with Unit 1 at 5 x 10(-11) amperes on the inteneediate rar_e and inserting control rods shutdown, a reactor trip was initiated Nhan source range 111-31 (Ells No. R1) reinstated with indication above the high flux trip setpoint. Immediately following the trip, all control and protection systems functi(xled as expected with the exception of source range HI-31, Nhich failed high. Ai_ox. 6.5 hrs foLt_ing the reactor trip, with NI-31 faiLe_ high, source range NI-32 was declared inoperable due to noise. With the unit at a hot shutdown condition, source range indication was unovai Labte for about 4 hrs. _4:pr¢_)riate abnormal procedures _ere ieptemnted to insure positive reactivity was not added to the core. The premp to El-31 was replaced and source rathe indication was established. Prior to the start-up, the source range detector for NI-32 was replaced and the channel returned to service.

A-81 p. 2 LER .DATABASE_- TRANS 08/27/93

PHASE2 PLANT NAHE EVENT INITIAL PL_T RECOVERY CATEGORY DATE Cf_D l T1ON T114E

¢ e_i, e a ie_.eplee _ _eJe e e _ ee ee ¢_ m. ee.. e _ee._.e _ o _e....e... e. amwaQ.., e..... _gme e. e e o .eee e lugs. e...., u e.ie. e .e ..e _ e e.. w .e .e _e TPJkNS Surry 1 06119/84 Routine startup Power LeveL - 010_. On June 19, 1984, with the unit just Less than 10_ power, a reactor trip resulted Nhen 2 of 6 nuclear power channels, HI 44 and H! 41 exceeded 10_ power with the turbine unlatched. A prinary plant cooLdoun of approx..8 F/sin and a primry dilution of 58 ppm contril_Jted to the power increase. FoLLowing the trip, iLL control and protection systems functioned as expected. Hain steam was isolated and the turbine stop vaLves were cLosed to Limit primry plant cooLdo_n. Precautions wiLL - be added to station proce(ka'es LOP 1.6 and PT 1S.lC) to prevent testing the stem driven auxiliary feadwater pump near the P-IO setpoint without the sin turbine being Latched.

TITANS Surry 1 01/27/85 Routine startup Power Level - 01OZ. On 1/27/85, Unit 1 ms critical with reactor power stable at 5% foLLowing a reactor trip on 1/26/85 (see LEIt 85-003-00). The stem d_ valves were isolated earlier because of knovn I:_ not specificaLLy identified or quantified Leakage. As the dumps were unisoLated, the resulting Leakge Led to a primary system temperature decrease which caused reactor power to increase. As po_ neared IOZ, it ms deci(_q:l to Latch the turbine to prev4mt a trip. Al=prox. 2 minutes after the turbine vRs Latched, the 6 turbine stop valves cLosed resulting in a reactor trip at 0768 hrs. One factor contributing to the trip vas not sufficiently considering the effect of the stem Leakage on plant piramaters. Another contributor to the event was that onLy one electro hydraulic (EH) pump was avaiLabLe end running when the turbine was Latched and it dtd not satisfy the EH demarv_s during the Latching operation. The steam (kap Leakage was identified and isolated. The htamn perform evaluation system coordinator is irtv_tigating this event arid wiLL provide feedback to the operating staff to tmprov_ huron performance in similar circumtances.

TfL_IIS Surry 1 01/28/85 Startq) Paver Level - 015X. On 1/28/85, during a Unit 1 startup, a reactor trip occurred due to a differential pressure anti-motoring tmt)ine trip. PLant parameters did not indicate that a generator mtorin9 condition existed. The trip occurred because the exhaust pressure sensing Line root valve |n the anti-motoring instrunentation was isolated. It is believed that this valve, tdhite Sll_, (_,_reLoped a smaLL Leak during a previous period of pouer operation, allowing the sing Line to bee(me pressurized. The Line remained sufficiently pressurized during the shutdovn period to cause the anti-lootoring delta p seq)oint to be exceecled as the turbine was being Loaded. Station drmdings and valve Line up checklists for the sin steam system HiLL be changed to reflect the correct position and function of the valves. (from chutowp.Ler)

TITANS Surry 1 02/07/86 Routine startup Poker Level - 015%. On 2/7/86, a Unit 1 startup was in progress. Fee(kaater control _ in manual and the trarls|tion from bypass to sin feed regulating valve (FRY) (Ells FCV) was taking place _hen the "C" steam generator feed flow suddenly had a ster_ increase above dennuv_._ _ de_ was decreased, the flow went to zero. Idhen the "C# FRVwas spared the second time, the flow again increased sign{ficantty. The operator closed the FRV; hoverer, he wa_;unable to prevent a high LeveL co, iLion in the stem generator and at 2338 hrs, a feed pump trip and turbine trip occurring initiated a reactor trip. The cause of this event was the failure to adjust the feedback cm on the "C" FRY foLLowing mintenance, _hich prevented f_le control of the valve. Due to a procedural ir_, the instrument department was not notified to check the control adjustments of the valve foLLowing mintenance just prior to startup.

TIPJUIS Surry 1 02/08/86 Routine startup Power Level - 002_. On 2/8/86, at 2:[ power, during a Unit 1 startup, the operating sin fee_ter pump tripped due to a high Level in 'C' SG. This caused the AFW pumps to auto start. The high SG LeveL occurred vhen the 'C' nmin feedwater bypass valve failed to close on desmnd. This valve was found to have a dust acctnuLation in the air pilot relay _hich blocked air to the valve operator. The bLockage A-_2 p. ] LE_DaTAeAS_- T_A.S 08/27/93

PHASE2 PLANTNALE EVEHT INITIAL PLANT RECOVERY CATEGORY DATE C_]_DITION TIIE

.....,...... -...... ,.-.,.-...---...... was removedand the valve closed. S6 Levels were returned to normal. Engineering will evaluate mthods for controlling contamination in the instrument air system.

TUNS Surry 1 09/16/86 CSO Paver Level - IOOX. OnSept. 16, 1986, with Unit No. 1 and Unit No. 2 at IOOXpower, a reactor protection system relay failed during the perfomance of Unit 1 inonthty surveillance testing. The relay failure resulted in a partial train *B' engineered safety feature (ESF) actuation. On Sept. 20, 1986 with Unt¢ 1 at hold shutdeun and Unit 2 at IOOXpower the samereactor protection relay failed. Again, a partial train 'B' ESF actuation was |nit|sled. The cause of both relay failures determined to be a failure of the coil. The coil has been replaced, the relay re-installed and satJsfactor| [y tested.

TITANS Surry 1 07/18/89 IOOX, Unit 2 - CSO Paver level - 100_. On July 18, 1989 with Unit 1 at lOOXpower and Unit 2 in cold shutdown, fottGv|ng the mntputatton of a asrvice water (SU) valve to the Unit 2 bearing cooling heat exchangers, the Unit 1 charging puap SU pumpsbecameair boundand the pumpsware declared |nopersble. This is contrary to Technical Spec|ficat|on 3.3.A.7. The cause of this event has been attributed to air entering the $W lines that supply the charging _ SWpueps. The affected pumpswere vented and returned to service. Additional high point vents _re being installed. A procedure for periodic venting has been devetqx_l and restoration procedures enhanced, in addition, a design changebeing implmented to increase the size and nmlber of the St/supply Lines to the _ has appropriate vents to facilitate removal of entrapped air. Engineering is also'continuing their investigation of the event to detemine if other actions are requ|red.

TRAILS Surry 1 07/23/89 100_, Unit 2 (:SO Power level - IOOX. OnJuly 23, 1989 at 1747 hrs, with Unit 1 at 100_ power an Unit 2 at cold shutdown, following manipulation of a service water (S",I) valve to the Unit 2 bearing cooling water heat exchangers, discharge pressures for the control room/relay roam (CR/RR) air conditioning chillers' Skl Ixaq_ and the Unit 1 and Unit 2 charging _ St/pua_ decreased due to air binding in the Ixaq0s. The puq_ were declared inoperable. ALso, the t_o operating CR/IUtchillers tripped on high condenser discharge pressure. This is contrary to Technical Specification 3.3.A.7 and 3.23.C. The cause of the event has been attributed to air entering the St/ lines that supply the _. The affected pumps_ere vented and returned to service and the CR/RRchillers _ere returned to service. Additional high point vents are being installed. A procedure for periodic venting has been developed and the abnomat procedure for restoration has been enhanced, in addition, a design change being implementedto increase the size and numberof SY supply tines to the pumpshas provisions for appropriate vents to fac|t|tate removal of entrapped air. _!ngineering is cont|nuing their investigation of the event to detem|ne if other act|ons are required.

TITANS Surry 2 12/19/82 Hot SO White conducting the RCsintegrity test (PT 11), a _eLd Leak on the 'A' steamgenerator channel head drs|n piping st 2-RC-157 bmsidentified. Inmediate action was taken to return the unit to CSD. This event is reparable pursuant to Tech Spec6.6.2.A.(3). The leakage _as within the capability of the normal rake q) system. The cause of this event is believed to be poor fusion between successive passes in a matt area of the weld. The un|t was returned to CSOand the defective weld was repaired. Liquid penetrsnt examination of the final weld revealed no defects.

TITANS Surrv 2 09/21/83 Hot SO Yith Unit 2 at hot shutdownNhite conducting the RCS integrity test (PT 11) a weld leak on the mA" steamgenerator channel head drain piping at 2-RC-1§9 _as identified. Actions were taken to return the Unit to cold shutdewn. This event is contrary to tech spec 3.1.C.4 and is reportable per tach spec 6.6.2.Bo(4). Leakage was within the capability of the normal make-up system. The exact cause has not A-83 p. 4 _ER DATABASE- TRAN$ 08127/93

PHASE2 PLANTNN4E EVENT INITIAL PLANT RECOVERY CATEGORY DATE CON1TOION T1HE

been determined. An identical event occurred on 12/19182 and the weld Leak was attributed to inadequate fusion between successive weld passes. Only the portion of the weld determined to be defective m repaired. It is suspected that undetected defects remained and propagated to the surface. The weld was c_ptetely ground out:and re_etded and tested satisfactorily.

TP,ANS Burry 2 04/15/84 2%power Power Level - O02Z. At 1604 on 4/15/84 following a maintenanceoutage, Unit 2 ties st 2_ reactor power vhen a reactor trip occurred as a result of an intermediate range (111-35) high flux trip. PLant perieters did not indicate a tmLid high flux trip. An electrician was checking for continuity across the smirch for TV-SS-201Atihen an arc occurred resulting ins spike on vital bus 1 2_hich caused the spike on 111-35. The muttimater Has selected to resistance instead of voltage. (from chutoVpotor)

Burry 2 04/20/84 Hot SO Pouer Level - OZ. OnApril 20. 1984 at 0216 hi's. vith the unit at hot shutdmm, a reactor trip occurred due to a high fLuK on source range 111-32. The reason for the trip ms personnel error by a instrtaamt technician While he Has troubleshooting 111-32vhtch had • failed preamp. The irlstrtaaMit technician tins disciplined. A replacement premap is on order and NiLL be installed when it arrives.

TRANS Burry 2 03/16/87 Hot SO Pover Level - OX. OnIlarch 16 1987 at 1704 hrs, Unit 2 ms at intermediate shutclmmwith shutdotm benks 'A' lind 'B' trithdrsim. During the perfomm_e of • periodic test of s steam flov trermittter (EIIS-FT) on Unit 2, a reactor trip'by turbine trip occurred, resulting in the insertion of the shutdotm banks (EllS-ROD). This test trips pomtssive P-7 bistabLes (EllS-33) (turbine first stage presswe) ad simulates • signal of 10_ power. Since the turbine was unlatched at the time, pemlestve P-7 (EIIS-JC) was activated, resulting in a reactor trip by turbine trip. The periodic test procedure requires the instrument technician to ensure that the turbine is Latched if plant conditions are such that the reactor trip breakers (RTB'B) are closed and reactor power is Less than 10"_. The technician believed that the RTB's were openwhen in fact, they had been closed earlier in the day. The technician continued uith the procaudre, and tripped the P-7 bistabLes, resuting in the activation of P-7 and the ensuing reactor trip by turbine trip. The technicians have been re-instructed to obtain the status of plant conditions from the shift supervisor.

TRANS Burry 2 09/10/88 Refueling outage Pover Level - 0043;. On Sept. 10, 1988 st 01S8 hrs, uith the un|t 2 reactor at 43;poller, during a shutdotmfor s refueling outa9e, a reactor trip by turbine trip occurred. The event occurred _hite oeprstors were attepttng to maintain the turbine at synchronousspeed, utth the generator output breakers open. laten the valve position Limiter was raised, per the procedure, a unexpectedly rapid opening of the turbine 9overnor valves and s rapid increase in turbine first stage pressure occurred, resulting in a turbine trip/reactor trip. Operators foLloved appropriate plant procedures and quickly stabilized the plant foLtoving the trip. The cause of the event has been attributed to a c_bination of an inadequate procedure, 8 faulty valve position Limit indication, and an unexpectedly fast valve position Limiter setting response. The controlling procedure Usedduring the event xitL be revised to ensure that the turbine control system is placed in the configuration intended. Testing will be performed on the electro hydraulic control (EHC) system, which will detemine if any additional actions wilt be required.

TRAILS Burry 2 09/16/89 SuberiticaL .+ Power Level - 0_. On Sept. 26, 1989 at 1228 hrs I_ith Unit 2 subcriticaL, during a reactor startup, a lamuat reactor trip was initiated vhen _t was determined that improper bardkoverlap existed between the 'A' and 'B' control rod banks. The reactor trip was initiated to insert all control rods and to reset " the control rod step counters to zero. A four hour non-emergencyreport tins made to the Nuclear A-$4 p. S LEROATABASE- TITANS 08/27/93

PHASE2 PLANTNANE EVENT INITIAL PLANT i_E_ CATEGORY OATE CON9!TlOti TliCE

,iQ _l,e o...., m Q.He_ _ W,. e e Q _ _._ _ _. e'! UQ o.'.,., e., .,.., o..,I ..., m.,. o,lo., em., ,.. ,.. Q o.-, .._. O O,e.,e e,V,D _. : ...... w.,..., o ,. I...,. ,.,..,....,,.. ,. Regutatcry Commissionper 10CFR50.72. TroubLeshootingdid not reveal the cause of the improper bank overlap. I)urir_ the subsequentreactor startup, no problem _ere encountered with control rod bank overlap.

TIWllS Surry 2 09/18/89 14_ Ix_r Power Level - 14X. On Sept. 18, 1989, at 1042 hrs, with Unit 2 reactor at 14:; po_r lind the turbine lit 1800 tin under no Loadconditions, • reactor trip stgnat was generated. A generator back14) differentlet tockaut relay 86 i_ tripped the turbine, and since reactor lover mm 9rester than IOX, the turbine trip initiated • reactor trip. Operators performed the appropriate plant procedures and quickly stabilized the plant loLL•ring the trip. The 86 bu generator backup Lockout retw trip ms caused by the spurious actuation of the generator backup impedlmce relay (IG)-41). The exact cause of the sptrious actuation of the relay coutd not be determined, however faults were discovered in the relay. The faulted I(D-&I retw ties replaced and appropriate testing was performed. The generator stairtqp procedure has been revised to ensure that reactor power is Less than 10; prior to closing the exciter field breaker. A four hour non-emerllonCyreport was madeto the NucLear ReguLatory Cmnlesion in _ ,ith lOCFItSO.72. APPENDIX B

Review of U.S. Nuclear Regulatory Commission (NRC)

Information Notices, Generic Letters, Bulletins and Circulars

B-1 NUREG/CR-6144 Appendix B

Review of U.S. Nuclear Regulatory Commission (NRC)

Information Notices, Generic Letters, Bulletins and Circulars

Va_a_

B.1 Summary of Findings ...... B-3

B.2 Review of NRC Information Notices ...... B-4

B.3 Review of Generic Letters, IE

Bulletins and IE Circulars ...... B-11

NUREG/CR-6144 B-2 Appendix B: Review of U.S. Nuclear Regulatory Commission (NRC) i Information Notices, Genetic Letters, Bulletins and Cireulam

B.1 - Summary of Find|n2#

Potential De mradation of Systems that Can Be Used for Accident Mitigation

1. Due to the activities in an outage, transient material and debris my be inside the containment. They may clog up the suctions from the containment sump in a LOCA. The operability of the low head injection system, inside containment redrculation system, and outside containment re,circulation system will be affected. (IN 88-22) Use the June 16, 1988 event as failure data for recirculation pumps. (IN 89-77)

2. During startup, when the power is above 10%, the intermediate power level(25%) trip will be blocked. If the reactor power is lowered below 10%, and two of the four bistable switches fails to reset, then the reactor will not trip until I09%. This applies to both power range and intermediate range channels. Tech. Spec. allows one channel t_ be taken out of service. (IN 86.105)

3. Common mode failure of four normallyclosed motor operated valves, that control the service water flow to the recirculation spray heat exchangers, to open.(IN 83-46)

Potential _tiatin_ Events

1. Inadvertent lifting of fuel assemblywhile lifting '_heupper internal,February 26, 1986 and Indian Point Oct. 90 (NSAC-129, IN88-92,IN 86-58) Damage to fuel assembly during refueling (IEC 80-13).

2. Inadvertent withdrawalof control rod. The events at Vermont Yankee on November 7, 1973 and Millstone 1 on November 12, 1976 can be used as a data.(IN 88-21)

3. Pressurizer surge Hnemovement and deformation due to thermal stratification (IEB 88-11)

B-3 NUREG/CR-6144 B.2 Review of NRC Information Notices:

90-19: Potential Loss of Effective Volume for Containment Recirculation Spray at PWR Facilities Entrapment of containment spraywater in the refueling canal may lead to insufficient water returned to the sump, and inadequate net positive suction head to the containment spray pumps and low head safety injection pumps. This may happen if the refueling canal drain valve ts closed. hmlntloa: The RWSTs of both units are cross connected, and the amount of water that can be heldup in the refueling canal is only a small fraction of the water in a RWST.

89-77: Debris in Containment Emergency Sumps and Incorrect Screen Configurations June 16,1988, Sur_. July 8, 1989, Trojan. Diablo Canyon.

88-28: Potential for Loss of Post-LOCA Recirculation Capability due to Insulation Debris Blockage March 14, 1988, Susquehanna. (It appears that any recirculation will be not likely during an outage. Due to the activities going on, it is not likely to keep the floor inside the containment dean.) Resolution: The concern regarding the ability to use recirculation while the reactor is shutdown due to the potential problem of transient mterial, debris due to the activities during shutdown need to be modelled. Surry'slow pressure injection/recirculationsystem does not have heat exchanger for heat removal Therefore, ion8 term heat removal may depend on containment spray/recirculation.

90-06: Potential for Loss of Shutdown Cooling While at Low Reactor Coolant Level July 1.8,1989, Comanche Peak. Loss of the invertor supplying power to the controller for the RHR heat exchanger flow control valve (F_ caused the FCV to open to the fully open position. The sudden increased flow caused vortexing at the RHR suction. ltmmlutiea: Use it as a loss of RHR event. It occurred prior to initial criticality. Therefore, it is not used to estimate the frequency of loss of RHR.

90-05: Inter-sys--temDischarge of Reactor Coolant December 1, 1989, Braidwood, 68000 gallon. Stuck open RHRS relief valve, and inability of operators to identify the leak(lack of EOP) ctnsed the incident to last more than 3 hours before the condition of the plant is stabilized.

Resolutioa: Use it as a loss of RHR event with RCS solid while drawing a bubble in POS 12.

89-73: Potential overpressuk-izationof Low Pressure Systems September 5, 1989, McGuire Unit 2. Containment spraysystem(design pressure 220psig) was overpressurized by 450 psig RCS pressure during functional test .)f suction valve from the sump. 2200 gallons of coolant was lost. In adequate test procedure. Resolution: Surry's RHR system does not take suction from the sump. Neither does the containment spray system. The two recirculation systems do, however, they do not take suction from RWST. The low head injection system injects into all cold and hot legs, takes suction from the RWST and the containment sump, and does not take suction from the RCS. Therefore, this scenario does not apply to Surry.The injection lines are shared between the high head injection system and low head injection system. Normally the LHSI is isolated from charging

NUREG/CR-6144 B-4 ,..... _ _._ "%_"_ and s._ _ _OOe_r_eoAImage_00,Management,_n_e,,_,te _'<" _i , _"_

___ Association for Information _@ _._,

Centimeter 1 2 3 4 5 6 7 8 9 10 1 1 12 13 14 15 mm

1 2 3 4 5 Inches _ ,-IlU_

IIIll"_ F ,,, Illllg IIIIINIllll_IIII1_

I:::IITH STI:::INDI:::IRDS _ .,,, //_/_.._

BY QPPLIED _

by one check valve and two closed MOVs. Upon spurious SI in shutdown condition, if the check valve fails open, the LHSI may be overpressurized. LHSI flow path will be open with the normally operating chargingpump injecting to it and the RCS. Similar scenario may occur at power. We need to review test procedures for high head injection system, low head injection system and RHRS to identify possible scenarios that may lead to an interfacing systems LOCA. This should be considered as a special initiator.

89-41: Operator Response to Pressurization of Low-Pressure Interfacing System March 9, 1989, Votgle. Leakage through inboard isolation valve of the RHR system cause the RHR system pressure to stay high after taken out of service. Operators opened the suction valves from the RWST to relieve the pressure. Resolution: This can be used as a data in interfacing LOCA analysis. Surrfs RI-IR system does not take suction from RWST.

89-71: Diversion of the Residual Heat Removal Pump Seal Cooling Water Flow duringRe,circulation Operation Following a Loss-of-Coolant Accident Haddam Neck PRA. During normal operation, CCW system provides cooling to the RHR heat exchangers and pump seal coolers. During a LOCA condition, CCW is isolated and service water is used to provide cooling to the RHR components. A single failure of one of the service water motor-operated valves to open following a LOCA would result in only one branch of service water being available to provide cooling to both RHR heat exchangers and the seal water coolers.

Resolution: A review of ccw and sw operations at Surryshowed that the CCW to the RHR heat exchanger will be isolated on a safety injection signal. However, RHR is not used for safety injection. The low head injection pumps at Surry are self-cooled. Loss of CCNVand SW are considered in support system failures.

89-67: Loss of Residual Heat Removal Caused by Accumulator Nitrogen Injection May 20, 1989, Salem 1, cold shutdown. Full flow test of accumulator check valve caused 1800 cubic feet of nitrogen to enter the RCS. Both RI-IRpumps were nitrogen bound. Operators had difficulty in locating the drain and vent valves. ltesdutieu: Use it as a loss of RHR data

88-23: Potential for Gas Binding of High-Pressure Safety Injection Pumps During a Loss-of-Coolant Accident February 26, 1988, Farley 1. Hydrogen from VCT may accumulate in the high point in the charging pump suction (pipe segment between the charging pump suction and the LHSI discharge). Resolution: Need to get isometri_ for LHSI, HHSI, RHR and RCS.

May 13, 1988, South Texas. October 12, 1988, Surry. Hydrogen from the volume control tank or air dissolved in RWST water may be released to ECCS piping between RWST and HPSI. ltesolution: Need to get isometrics for LHSI, HHSI, RHR and RCS.

October 30, 1989, Trojan nuclear plant. Both high head injection pump may be inoperable ff a safety injection signal occurs while the manual "bypass"valve around the motor operated VCT outlet isolation valves is opened per test procedure.

B-5 NUREG/CR-6144 Resolution: At Surry, the seal water go to the charging pump suction directly without going to the VCT. It is connected to the VCT only through a relief valve RV1362B that allows relief into the VCT.

87-57: Loss of Emergency Boration Capability due to Nitrogen Gas Intrusion May 28, 1987, Turkey Point Unit 4. Nitrogen entered the boric acid system through a failed boric acid transfer pump mechanical seal. The seal is provided with an accumulator tank partially filled with demineralized water. The accumulator is given a 40-psi nitrogen overpressure to preclude leakage of the boric acid across the seal faces. As demineralized water entered the boric acid system through the failed seal, additional nitrogen was automatically supplied to the accumulator to maintain the pressure. The falling water level then allowed the nitrogen cover gas to enter the boric acid system through the failed seal. Reselution: A failure mode of boric acid transfer pump

82-19: Loss of High Head Safety Injection Emergency Boration and Reactor Coolant Makeup Capability February 12, 1982, McGuire unit 1. Hydrogen from the positive displacement pump suction dampener entered the common suction of the charging system, causing both centrifugal charging pumps and the PDP to be inoperable. Remlntlon: A CCF of chargingpumps. Surry does not have a positive displacement chargingpump.

83-49 and 82-32: Sampling and Prevention of l_rusion of Organic Chemicals into Reactor Coolant System February 13, 1983, Hatch. April 24, 1982 Hatch Unit 1. May 5, 1983, LaSalle. (Can similar events occur at a FWR leading to degraded core heat removal or reactivity accident?) Resolution: A systematic way of identifying such scenarios is needed, otherwise, these events can be used as data for the scenarios. The Latsalleevent is considered also possl"olefor a PWR, with the condem_te storage tank replaced by primary grade water storage tank. However, the event does not seem to have any significant safety impact. It may require cleanup of the RCS, but is not an initiating event.

89-54: Potential Overpressurization of the Component Cooling Water System May 15, 1989, Sm'ry. A design deficiency in the CCW system relief valve was reported. The capacity of the relief valve is not sufficient to relieve a tube rupture of the thermal barrier heat exchanger. Additional relieve valves installed. Resolution: In des_ignchange package 89-008, the CCW inlet check valve to the thermal barrier was replaced by two check valves in series. The thermal barrier outlet is isolated by a trip valve upon high flow.

89-32: Surveillance Testing of Low-Temperature Over-Pressurization Systems Beaver Valley, Turkey Point, Shearon Harris. Stroke time requirement in inservice testing (15 sac) for PORV was not consistent with the safety analysis (2 sec). Resolution: LTOP has been identified before and will be considered.

88-92: Potential for Spent Fuel Pool Draindown October 2, 1988, Surry Unit 1. Potential drain down of spent fuel pool to 13 inch above top of fuel assembly.

NUREG/CR-6144 B-6 Resolution: Design change package ??? installed new cavity seal in the Oct. 90 refueling outage. NSAC- 129 identify events involving drop of one fuel assembly, inadvertent lifting of a fuel assembly, madmisimsitiou of 4 fuel assemblies into 4 control cells with the control rods withdrawn.

84-93: Potential for Loss of Water from the Refueling Cavity August 21, 1984, Haddam Neck. Resolution: Beyond the scope except for flooding Resolution: Consider it for internal flooding analysis

88.87: Pump Wear and Foreign Objects in Plant Piping Systems May 16, 1988 Sun7 Unit 2, auxiliaryf_ater system. June 6, 1988, Surry Units 1 and 2, recirculation spray pumps. Resolution: Get information on improvement in design or operation. Include these failure modes.

88-36: Posm_oleSudden Loss of RCS Inventory During Low Coolant Level Operation Diablo Canyon. Westinghouse cold opening scenario

87-23: Loss of Decay Heat Removal During Low Reactor Coolant level Operation Diablo Canyon.

86-101: Loss of Decay Heat Removal due to Loss of Fluid Levels in Reactor Coolant System Resolution: Already identified.

88-21: Inadvertent Criticality Events at Oskarshamn and at U. S. Nuclear Power Plants July 30, 1987, Oskarshamn. November 7, 1973, Vermont Yankee. November 12, 1976. Inadvertent withdrawalof a control rod that is adjacent to a fully withdrawn control rod leads to criticality, and the scram system shutdown the reactor. Resolution: This is a reactivity accident that should be modelled.

88-17: Summary of Responses to NRC Bulletin 87-01, "Thinningof Pipe Walls in Nuclear Power Plants" December 9, 1986, Surry Unit 2. 1987, Trojan. December 10, 1987, LaSalle Unit 1. Pipe thinning in feedwater line.

86-106: Feedwater Line Break December 9. 1986, Surry Unit 2. Resolution: Find information on improvement in surveillance of feedwater lines and determine if the scenario need to be modelled.

87-59: Potential RHR Pump Loss Potential for dead heading one of two RHR pumps in systems that have a common minifiow recirculation line serving both pumps, during a small LOCA. Capacity of miniflow recirculation line for single pump operation. Resolution: NRC or Westinghouse should have resolved this. Get resolution from NRC.

87-46: Unidentified Loss of Reactor Coolant June 21, 1987, North Anna Unit 1, cold shutdown. 17,000 gallons of coolant was lost and voids formed in the vessel head and SG tubes. Misleading pressurizer level indication.

B-7 NUREG/CR-6144 Resolution: This is a LOCA event that can be used as data. Misleading level indication need to be modelled.

84-55: Seal Table Leaks at PWRs January 20,1984, Zion Unit 1. April 19, 1984, Sequoyah. Resolution: Minor leaks during power operation. We will look into failure of temporary seals used during refueling.

87-40: Backseating Valves Routinely to Prevent Packing leakage Jue 3[7,1987, Surry Unit 1. A low flow reactor trip occurred due to the failure of the stem of a hot leg loop stop valve. Similar event occurred March 7, 1974. Resolution: Surry does not use electrical backseating of the loop stop valves any more. Use it as data for power operation.

86-105: Potential for Loss of Reactor Trip Capability at Intermediate Power Levels (If a reactivity insertion occurs above 10% power, no scram will occur until 109% power)(If a reactivity insertion occurs below 10% a single failure of a bistable switch to reset may lead to failure of scram system.) Remlution: We need to model it.

86-79: Degradation or Loss of Changing Systems at PWR Nuclear power Plants Using Swing-Pump Desigas June 26, 1985, SurtT Unit 1. Pump A was down for maintenance, and when pump B was racked out, an interlock caused the running pump C (swing pump) to trip. Resolution: This will be modeled. No additional action needed.

86-63: Loss of Safety injection Capability December 28, 1984, Indian Point 2. Two leaky valves in the discharge of the boron injection tank enabled highly concentrated boric acid to flow to the low pressure discharge line (SI pump suction) and to precipitate in the pumps, which were not heat traced. Degassing of the nitrogen cover gas dissolved in the boric acid solution is believed to be one of the likely sources of gas found in the pumps. Resolution: Surry does not have boron injection tank.

86-58: Dropped Fuel Assembly February 26, 1986, Haddam Neck. A spent fuel assembly was inadvertently lifted from the core when the upper core support structure was removed from the reactor vessel The assembly impacted the core barrel and was knocked off. Resolution: Use it as well as the recent Indian Point incident as data for failure of one fuel assembly.

85-12: Recent Fuel Handling Events Resolution: We need to model it along with events in NSAC-129.

86-01: Failure of Main Feedwater Check Valves Causes Loss of Feedwater System Integrity And Water-hammer Damage November 21, 1985, San Onofre Unit 1. Resolution: We need to model it.

80-01: Fuel Handling Events

NUREG/CR-6144 B-8 December 11, 1979, inadvertent raising of spent fuel assembly, December 17, 1979, a new fuel assembly was dropped in the fuel pool. Pilgrim. Resolution: Combine with NSAC-129.

84-70: Reliance on Water Level Instrumentation with a Common Reference Leg All the wide range level instruments have a common reference leg that developed a leak. The leak made the operators to maintain an apparent level while the actual level decreased, thereby causing the pressurizer to drain and a bubble to'enter the top of the head. Resolution: We need to fred out about the design of level instrumentation at Surry. Is this a LOCA?

84-42: Equipment Availability for Conditions During Outages Not Covered by Technical Specifications January 8, 1984, Palisades. Loss of offsite power with one diesel generator 2 down for maintenance and the service water pump that is powered from diesel generator 1 was also down for maintenance. DG I was tripped 50 minutes later upon overheating. Resolution: Already identified.

83-46: Common-Mode Valve Failures Degrade Suny's Recirculation Spray Subsystem February 9, 1983, Surry. Four normally closed MOVs that permit service water flow to the recirculation spray heat exchangers failed to open. Possible causes are corrosion, marine growth, infrequent testing, and low torque switch setting. Resolution: A common mode failure. Check if 1150 models it.

83-41: Actuation of Fire Suppression System Causing Inoperability of Safety-Related Equipment May 28, 1981, Surry unit 2. and other events. Discharge from the foam distributor system installed in the main (reserve) diesel fuel oil tank caused 4000 gallons of water be introduced and widely distributed in the diesel fuel system before a routine periodic test disclosed the presence of water. Resolutioa: Get information on fixes implemented to determine if it need to be modelled.

82-45: PWR Low Temperature OverpressurizationProtection August 1976, Turkey Point Unit 4. Po_'ble causes of inoperable PORVs: 1. Operation with both PORVs isolated (block valves closed) because of known PORV leakage.

2. Operator error during maintenance. 3. Isolation and venting of instrument air to the PORV actuators during integrated leak rate testing. 4. Low nitrogen(backup accumulator) pressure to PORV actuators. Resolution: LTOP willbe analyzed.

82-17: Overpressurization of Reactor Coolant System November 28, 29, 1981, Turkey Point 4. A pressure spike caused by starting a reactor coolant pump caused automatic isolation of the RHR suction valves. OMS failed to operate because a pressure transmitter isolation valve was found closed, the summator in the supposedly operable electrical circuitry failed, and the redundant OMS circuit was out of service. Resolution: Already identified.

B-9 NUREG/CR-6144 82-28: Hydrogen Explosion While Grinding in the Vicinity of Drained And Open Reactor Coolant System April 20, 1982, Arkansas nuclear One Unit 1. A craftsman was grinding the HPI pipe line in preparation for welding. A hydrogen explosion occurred and blew the craftsman away by about 3 feet. Resolution: Consider it in fire analysis.

81-10: Inadvertent Containment Spray due to Personnel Error February 11, 1981, Sequoyah. Resolution: Already identified.

81-09: Degradation of Residual Heat Removal March 5, 1981, Beaver Valley, midloop operation. Resolution: Already identified.

81-04: Crackingin Main Steam Lines February 23, 1981, Surry Unit 1. A lengthy crack indication in the I.D. counterbore area of a weldment on the in-line _ fitting which connects the vertical run of 30-inch piping to the safety relief valve header and 30-inch main steam line of steam generator A. Resolution: Use it in LOCA frequency assessment.

80-44: Actuation of ECCS in the Recirculation Mode While in Hot Shutdown December 5, 1980, Davis Besse. Inadvertent actuation caused a flow path to the RCS via the borated water storage tank and the DHR piping. No BWSTwater was pumped into the RCS. Rather, during the valve transition time of about 1.5 minutes, approximately 15,000 gallons of borated water was drained from the BWST to the containment emergency sump.

Resolution: RHR system at Surryis independent of safety injection system and does not take suction from RWST or containment sump.

80-20: Loss of Decay Heat Removal Capability at Davis-Besse unit While in a Refueling Mode April 19, 1980, Davis-Besse. Loss of power to non-essential 480 v bus caused recirculation mode actuation, and loss of RHR suction. (extensive maintenance act_,'ities) Resolution: Already identified in RHR data base.

80-34: Boron Dilution of Reactor Coolant During Steam Generator Decontamination May 29, 1980, Trojan. July 5, 1980, San Onofre 1, high pressure demil_eralizedflushing water leaked by a dislodged nozzle seal and the resultingreactivity addition exceeded the tech. specs. Resolution: Already included in reactivity accident.

79-04: Degradation of Engineered Safety Features Arkansas Nuclear One, units I and 2. MSIV closure at unit 1 lead to loss of ac power at unit 2 which was in hot standby. Resolution: A cause for loss of offsite power.

NUREG/CR-6144 B-10 B.3 Review of Generic Letters, IE Bulletins & IE Circulars

Condusioas GL -- generic letter IEB -- IE Bulletin IEC-- IE Circular

1) Loss of RHR. Some special concerns are: a) in midloop operation, if boiling begins in the RCS and the cold leg is open (e.g. maintenance on the pumps), the increased pressure will force water out of the opening, thus further decreas/ng time available for recovery (GL 87-12, 88-17; Diablo Canyon April 10, 1987); air ingestion will cause erroneous level indication (it should be noted that in the Diablo Canyon event, RHR pumps were able to handle a few percent air ingestion), c) maintenance activities may cause loss of redundancy during shutdown (IEB 80-12, Davis Besse, April 19, 1980). d) service water problems may cause loss of RHR 0EC 81-11, GL 89-13; Bnmswick, a BWR, Dec 8, 1980 -- SW train A in maintenance, B not aligned because a expected to be repaired before boiling reached, A train repair took too long).

2) Steam void generation during depr_tion. Hot metal out of the main flow path. Pressurizerlevel increases suddenly. Steam can interfere with heat removal. (IEC 81-10, 80-15; Crystal River 3, April 21, 1981, during cold shutdown; McGuire 1, June 1981, during heatup; also St. Lucie 1, June 11, 1980 during full power -- loss of CCW flow to RCI_ causes reactor trip -- cooldown for a few hours until rise in press, level Problems lead to eventual use of LPSI)

3) Condensate booster pumps can be used for secondary, but need to depressurize first (GL 88-20)

4) Steam binding of AFW pumps. Due to leakage past isolation valves of hot water from the MFW. IEB 85-01, GL 88-03; Crystal River 3 1982 & 1983; Robinson 2, 1981 through 1983; D. C. Cook, 1981; McGuire 2, 1984; Catawba 1, Nov. 1984)

5) Water hammer can disable important safety systems. (GL 86.07; San Onofre 1, Nov. 21, 1985, 60% power, loss of all ac power for 4 rain followed by a severe water hammer which caused a steam leak, damage to plant equipment and loss of f_ater for 3 rain; caused by check valve failure)

6) Overfill of steam generators, overcooling, overheating-- obvious concerns (GL 81-28, 81-16, 83.37)

7) Low temperature overpressurization. (GL 88-11; the critical limits and the operating window had been changed recently - to make it more difficult).

8) Boron dilution events during hot standby (only 15 rains available btw. an alarm and loss of shutdown margin)

9) Control rod withdrawalwith only one RCP operating in hot standby may violate DN-BR (for W plants -- FSAR says 2 RC'Psoperating in hot standby, but TS allows only one -- OK for most accidents except uncontrolled rod withdrawal; GL 86-13)

10) Sump screen blockage. Obvious concern, maybe more so in shutdown. Plants most vulnerable are the ones with small debris screen area (< 100 sq. fl), high ECCS recirc, pump requirements(> 8000 gpm) and small NPSH margins(< 1-2 ft of water) (GL 85-22)

B-11 NUREG/CR-6144 11) Control of heavy loads. Surry did a load drop analysis to satisfy requirements (GL 85-11).

12) Loss of shntdowa marKInduring refueling operations. Baltimore Gas & Electric presented this analysis for its Calvert Cliffs units, in March of 1987. It shows that at some intermediate positions during refueling, and assuming clustering of higher enrichment fresh fuel for the extended cycle, criticalitycan occur. (GL 89-03).

13) Pressurizer surge line movement and deformation due to thermal stratification. This occurs during heatup, cooldown and also steady state, due to stratification of hot water flowing from the pressurizer and the layer of "cold"water from the hot leg in the pressurizer surge line (IEB 88-11; Trojan observed unexpected movement of PSL at each refueling since 1982, with the "latest"one actually touching the restraints and causing plastic deformation; Beaver Valley 2 also noticed the PSL movement and the snubber movement during power ascension).

14) Pos._'ble loss of pumps (e.g. RHR pumps) in miniflow conditions. The stronger pump will cause deadheading of the weaker pump (2 pumps in parallel with a common miniflow line). The capacity of line may not be enough to support even one pump miniflow (IEB 88-04)

15) Damage to fuel assemblies during refueling. Observed extensive damage to grid straps in W fuel assemblies, caused by hitting diagonally neighboring assembly during insertion (IEC 80-13, Salem 1, LER 79-44).

NUREG/CR-6144 B-12 C_,ner_ Letters of'Interest to Sur_ Sbutd_u .Study

81-07: Control of Heavy Loads, also GL 85-11. The response should include use of electrical interlocks or mechanical stops, single failure proof cranes and load drop analysis. The last one was done at Surry, and the areas analyzed were the spent fuel pool area and the containment building. Boron dilution events duringhot standby. 15mitt availablebetween the firstalarm and loss of all shutdown margin.

81-16: Overfilling of steam generators

81-22: CVCS leak at H.B. Robinson plant -. inadvertent operation of charging pumps. 6000 gal in letdown train of CVCS. Inconsistency between tech specs and safety analysis (in 3 loop W plants). FSAR assumes 2 RCPs in operation in Mode 3 (hot standby). Tech specs say one is OK. With only one pump, an uncontrolled withdrawal of the control bank from the subcritical condition may cause violation of DNBR criteria.

86-07: Water hammer and loss of power at San Onofre. This occurred at 60% power. All 5 check valves were disabled.

85-22". Post LOCA sump screen blockage. The plants that are most vulnerable are the ones with: - small debris screen area (<100 ft**2) - high ECCS recirculation pumping requirements (>8,000 gpm)- small NPSH margins (< 1-2 ft of water).

85-16: High boron concentration. At Indian Point 2, all 3 SI pumps were frozen with crystallized H3BO3. Boron injection tank(BIT) has a high concentration of boric acid (in case of a steam line break). Surry 1 & 2 have requested to remove the BIT or reduce the concentration to 2'000 ppm. The request was granted.

85-13: Davis Besse loss of main and auxiliaryfeedwater

85_ Generic W modifications for reactor trip system. Require at power testing of undervoltage and shunt trip attachments.

85-07: Inadvertent boron dilution events. 110%pressure in the RHR. Instrumentation is not required.

85-02: SGTR prevention. TV camera for loose parts and foreign objects -- after any secondary side modifications and eddy current testing.

89-21: USIs and GSIs: water hammer SG tube integrity A-31 RHR shutdown requirements A-49 pressurized thermal shock A-47 control systems: overfill protection, overcooling, overheating transients. Surry has 2 out of 3 logic, with one channel used both for control and protection. MFW isolated by closing the MFW isolation valves and tripping the MFW pumps. System interactions -- flood, seismic, electrical reliability and operator actions

89-13: Service water system problems

B-13 NUREG/CR-6144 88-17: Loss of decay heat removaL NUREG-1269 (report on Diablo Canyonevent) identified many generic weaknesses.

NUREG/CR-6144 B-14 APPENDIX C SYSTEM FAULT TREES

C-I NUREG/CR-6144 APPENDIX C SYSTEM FAULT TREES

PAGE

C.1 Accumulator System ...... C-3

C.2 Auxiliary Feedwater System ...... C-8

C.3 Charging Pump Cooling System ...... C-72

C.4 Component Cooling Water System ...... C-82

C.5 Coc_pressed Air System ...... C-104

C.6 Component Spray System ...... C-113

C.7 Emergency Power System ...... C-119

C.8 Emergency Switchgear Ventilation System ...... C-143

C.9 High Pressure Injection & Recirculation System ...... C-159

C.10 Low Pressure Injection & Recirculation System ...... C-193

C.11 Primary Pressure Relief Sy_*,¢m...... C-235

C.12 Recirculation Spray Systems ...... C-243

C.13 Residual Heat Removal System ...... C-265

C.14 Service Water System ...... C-280 (The Fault Trees for the Service Water Components are included hi the Fault Trees of those systems that depend o_LService Water) (See App. C.3, C.4, C.8 & C.12)

C.15 Steam Generator Recirculation and Transfer System ...... C-281

C.16 Steam Generator Secondary Relief System ...... C-293

C.17 Consequence Limiting Control System ...... C-305

C.18 Reactor Protection System ...... C-308

C.I 9 Recirculation Mode Transfer System ...... C-310

C.20 Safety Injection Actuation System ...... C-313

C-2 Appendix C.I Accumulator System

C-3 t ACCUMULATORS (ACC) (DS) I A_UM TO L_

I -T I,ro_"IrLvw

I SEGTI-IMENTS_ PIPEPSI20 SE_R PS120 SEANDC_ENrSPS1PS22_I I _,_ A_m , ? , , , , HI_ PPE _EGIJENT [TIdRUPf_ SEGMENT _ PIE S[(;t_NT _ PPE SEGMENT ITI-RUPI_ SEGMENT ITHRUPPE.!E_ PS12.0 PS122 PS12.1 | Ps_

ACC4 ACC5 ACC4 ACC6 ACC5 ACC6

PAGE1, 2 &: 3 r_

cj cj

C-5 ACCUMULATORS(ACC) (ACCS)

PAC_£ I, 2 & 5 qu_ W n_

C_ C_O

Co7 Appendix C.2 Auxiliary Feedwater System

c-8 5 _ i 3_Vd

' I i I I

I , I I ' tJ ' _ '

I (7) (L-M-IV) OS _" JO L 0_1_ N___l_VM(]33-t ,l,_tVl-ll×flV w---"

Lu_

C-IO

AUXILIARY FEEDWATERTO 1-OF 3 SO (AFW-1) (AFW14) I

I_ , III-.411_.-QO-III_ _-II_U II_I,.,IW'-I'P...-IICOlUlllW-,Oll-#l,-at_l

p+I_E 7 AUXILIARY FEEDWATER TO I OF 3 SG (AFW-I) (AFW15) I

_. _ _,'_,

_ 12B

! I I.

_I I_ I

I' t

• , ,

o'/

I x _ C)"I

I 0

C-14 AUXLIAR¥ FEEDWATERTO 1 OF 3 SG (AFW-1) (AFW17) I

FLOW THRO_H PIPEI SEGMENT PS81 I

FNLUREDC BU,OFS t120a. V FALUREBUSOF,1H4KV AC SEGMENi"_,PS81 PTRN AV1 E_ 4KV1H

0 /_ @ N_N-TNK-VF-CST AFW-ACT-FA-PMP3A AFW-CCF_-LK-STMBD

IDUE TO ,_KF'I.OW I' °l '' b._ 19'2 I l ,J

' "MA-_3A

AFIY-CX'Y-CC-4_577 I I

THROUGH _ FW,._B THROUGH TDP FW2

AFW-MOP-FS--FW3B AFW--CW-OO-CV172 AFW-TDP-FS-FW2 AF'W--Q(V--OO-CV142 AUXILIARYFEEDWATERTO 1 OF 3 SG (AFW-1) (AFWiS) I

I F^U.'_aN_ /

/ °""" " / I,.'__.DLE TO 8t,C:i_,'LOW,,, r_°_" ""_" _'"

J lt_CKI_..OW O_OW I ll,ltU _ S[O i_. _ _ I"IROI_H PlOPIrW.._B __-____:,_ i_ _. _, ..

,AFW-k4DP-F'S-_ AFW-CKV-OO-CV157 A_m-UDP-FS--_/30 ,_L'W--O(V-OO-CVI72 TEST AND MAINTENANCEON AFW MDP 3A AUXILIARY FEEDWATERTO 1OF 3 SC (AFW-1) (AFW21)

,- . ..

PAGE 6 AUXILIARY FEEDWATERTO l OF 5 SO (AFW.----1) (AFW22)

I m

THRUPIPI[SEG PS95. PS06. PSi7 INSI,F51"MFLO# ]

,o I I PIPEEG THN PIE EG TItRU_ EG I"eJFS_PS95"FLOWI I NS_PSS_O_ * I I'_L_SP__"7FLO* I 0>

c_ u.. I

Z 0

Z ._ • °Z ii_.¢i ILl Z - N_-,;

Z _

D_I

C-20 C-21 AUXILIARY FEEDWATERTO 2 OF 3 SG (AFWl) (AFWl) I

.__ _ _,

! , I

P_t AUXILIARYFEEDWATER TO 2 OF ,3SC (AFW-2) (AFW2) I,, ,,

! I

I "m'nw i ,/kX ,,.-._,-_-_._/'k,x ,_.-_-_-e_ AqlellJl

PA61:9 AUXILIARY FEEDWATERTO 2 OF 3 SO (AFW-2) (AFWS) I " t

! t AUXLIARY FEEDWATERTO 2 OF 3 SG (AFW-2) (AFW15A)

PA61:2 AUXILIARY FEEDWATERTO 2" OF 3 SG (AFW-2) (AFW14A)

p_IE I1 AUXLIARY F'EEDWATERTO 2 OF 3 $C (At-'-W-2)

I i

t_-I t_--.i i__" ! "

_=_-_._A AI_-CKV-OO-CVI_7 AFIF--MI__ AFYI_--QO-C'VI_2 c_ CL

Z 0 W Z Z W z

L.U F-

C-28 AUXLtARY FEEl)WATERTO 2 OF 3 SG (AFW-2) (AFW',6A) I

I ,,, I

AFWl_A

41KVU E1B

. . _ '_ /qqI-ACT-FA-,q_38 AIrW-C_-I.K-Si14]D AI_-INK-VF-CST

__, , __ l ,,

MA_

I , I AUXLIARYFEEDWATERTO 2 OF 3 SO (AFW-2) (AFW17A) I ,

SI[GML,_N_I1 IX;_ _ 1

I I c_l_ I I

MA--IPW_Uk

I ,., I

_tX'W-_-F$-F"_V2 AF'W-.(:t(V_CV14.2 /d:'W-_-_ AF"W___ AUXILIARY FEEDWATER TO 2 OF 3 SG (AFW-2) (AFW21A) t_

PAGE 7 AUXILIARY FEEDWATER1-0 2 OF 5 SO (AFW-2) (AFW22A)

_]_ _

FLOWTHRUPIPE FLOWTHRUPiPE t_ _G PSg_ SEGPS98 PS97

_--Fr--_2 J_'W-_--_-_7 _-I_-FT--_I_ _eW-X_X_lk_ N_W.-t_-Fr-C'V178 AJ_-_1-_-_158 AUXILIARYFEEDWATERTO 1 OF 2 SG (AFW-5) (L3) I !,__.1

° ! ? ! Ire.,,! Ir_i

I_-'_I I_'."'_" i

,IIIW-¢IT-I_-II_IN _ illl_l-Cl_,-.l'l',-lg'_41 _ _ N_..O_,.41'lt_ ,LitI1.¢_,11_,..1_,.i,1t111 I !

I _,.E,,,,,,_I ,c I s,,_o I D

PAGE ! AUXILIARY FEEDWATERTO 1 OF 2 SO (AFW-3) (AFW13B)

HW..lzQr-4rlD-ZlL_E_ _qI'-.CL"II'-._0°_ _'W.,,4_mr-.I-,MOrtZ.,le H W.,,,O_-I'lt',,,.O_131

PACE! AUXILIARYFEEDWATER TO I OF 2 SG (AFW-3) (AFW14B) i

llll-.aV-.ll-lllll ,IIW..aV-I_-¢I_IIIMW-III'..II_.-II_Nl'l.,llll411r'._lll _iIii

PAGE 7 AUXILIARY FEEDWATERTO 1 OF 2 SO (AFW-3) (ArWlSB) I I _.r_, '1

N'W-klOP-F'S--FW_ krW-(](V--(X)--_157 kcW_1i:p-.FS-FW2 N'W-.C_-OO-CV142 AUXLIARYFEEDWATERTO 1 OF 2 SG (AFW-3) (AFW17B) I

_:¢.,1t1_3IxJ81_ I

FALUREOF 120 V FALLII_ OF"_ AC EGIENT PS81 FTRNI I [X_ BUS 'iA ! I BUS 1H i I_°_'''" •

E_ 4KV_I

_ @B AFW-TNK-VF--CSi" AFW-ACT-FA--PMP3A AEW-CCF._I.K-STI413D

FLC_VTHRUPUI4P N _ DUE TO BA()Clq._ e

krk-, 1_3A

I I

AF'W-M3P-FS-FW,:.'_AFW-C:KV-OO-CV172AFW-I"DP-FS-FW2 AFW-CKV-OO-CV142 AUXEIARYFEEDWATER TO I OF 2 SG;(AF3N-3) (AFW18A) I

,n o.,_.,.,.i _,_I___" __ _'_'_°I ,__.o,.,.

I I i .[ .._ 6_'I 39¥d

_tA4T tT_LLT |tAdT I_TIUidT IPtLIdlT PIIIJLdIT StJLdlT I_llAdlT

VI __I 7I _l VI, VI VI _IVI

I I I I

I__I I_-_I I__I

0 I _ I

' (_'L-_SOd- sn) (L-SMJV)OS_ JOL O± _3±vMa??_,k_WqlXnV AUXLIARY FEEDWATERTO 1 OF 3 SG (AFWS-1) (AFWS5) I I_'_1

P_8 AUXILIARY FEEDWATERTO 1 OF 3 SO (AFWS-1) (AFWS13) ° I-,_ AUXILIARY FEEDWATERTO l OF 5 SG (AFWS-1) (AFWS14) I o I-:_H

p_ 7 AUXLIARY FEEDWATERTO I OF 3 SO (AFWS-I) (AFWSm) I I--I ,

J_fSt_

_ /_ .41_-'II_-VIF_ AF'N--CCIF-4J(-_IIISD

TO W_IO'I.0W _ MBP II

! I

AFW-kI_P-rS-FW_ M'W-CIQ/-(X)-CV157 _ --_L'T-OO--._rLU AUXILIARY FEEDWATERTO l OF 5 SO (AFWS-1) (AFWS17) I

FLOW THROUGH F'PEI I S£GMENT,,_'T,"cPS81,a,a"I

FALURE OF 120 V F'ALURE OF 4KV AC SEGMENT PS81 PTRN DCBUS,1A BUS,1H o /_E1A, ;4KV1H_ _ _-_-_ ____

MA-FW3A

_6 AFW_--CC-41_577 /_ I i

THROUGH MOP FW,_ TH_U_ TOP FW2

.AFW-M[_-FS---FW3B N_"W--CKV-O0-C'V 172 AFW--CHV-O0 -CV14.2 AUXILIARYFEEDWATER- POS 3-13 (AFW41)

FAIL_H.E OF

PW151A

I ,,, i l i , I

PLUOOED PAIL8 TO OPEN NOV 151A,B Mf_f_-IHI-2 MfJV IN MAINTENANCE

APW-MOV--P0--iSIA APW-MOV--FT--lfllA AF'W--MOV-C_C-151 AUXILIARY FEEDWATER - POS 3-13 (AFW42)

0 0 0 _ All_-IPql°mll AIP_-ll_l'_mm AIP_-.IWC-.Infill0- AUXILIARY FEEDWATER - POS 3-13 (AFW43)

I I I i

I_llNmO ItAl,lJ I 1 I --I I __'-I I"IW_SI-''_ ! 0 0 0 _qf-i$OV--INi-mlO .lUPg-IiR_-IPT- Nm /.I'V -I$OY-.Ce-SM AUXILIARY FEEDWATER - POS 3-13 (AFW44)

I I I.. i

AJ4f0-li¢_r -IND-mlIi 4i/W-0lldV-IMi'-li_ .... $JW-©IIIOV_I_I AUXILIARY FEEDWATER - P0S 3-13 (AFW45)

0 0 0 AUXILIARY FEEDWATER - POS 3-13 o (AFW46)

! "-_"1 !"__l I__l I"A'"I 0 0 0 _ i.._I i,_.__°--,,l I_-. I I I I _J

I I

_O.f-31.IX-//UY L_I

(]NY _lnN i_r_ Ol JLVWILl _ g/I. I aNY 31L_mJNII oL-_vJI _ol_,_ ol _1 _'_ I

IY mU _ g/_ I ' (_sn) (_'-saJv)os _;JoL o_l_;_-__l_vaa]-_Jx_vl--llxnv AUXILIARY FEEDWATERTO 1OF 2 SO (AFWS-3) (AFWS6) I

,5,

I i

i ! i ...... zN z_ z_ Zx

P_8 AUXILIARY FEEDWATERTO 1 OF 2 SO (AFWS-3) (AFWS13B) | AUXILIARYFEEDWATER TO I OF 2 SG (AFWS-3) (AFWS14B) I

P_7 AUXLIARY FEEDWATERTO I OF 2 SG (AI:'WS-3) (AFWSmB) I

i iI

_U _8

I I

A_IOP_ AFW-Q(V.-O0--_157 AF_/-O_-OO.-_tt2 AUXILIARYFEEDWATERTO I OF 2 SO (AFWS--5) (AFWS17B) I

I""-'__'_,_" 1 I"_"_W,_ _ I !'_"_"'-'__',i"_

_ .

I I

A._ -14_-FS--I_3B Af'W-(:XV--O0--_4 f/2 AI_-(_(V-OO-CV142 AUXILIARYFEEDWATERTO 3 OF 3 SG WINDOW I - (LSWl) mmm

i i AUXILIARY FEEDWATER TO 2 OF 5 SG WINDOW 2&5 - (LSW25)

,.1[._ 1 N_ IpUIBi 2L_3

I-_''"i I @ I,,-..-I @ !-"..., ." @ '

I I ...L,. I I r,_,_,.! _ I-_! i-__ ._o ._. .-__ ..,_ ..,_. 1 .. I i I i I , t I I T*LiO_Ore I I ,,% /% /% zS, _ /% /% /% KI_WSI8 ATW4_ Kl'W814 &Y'W4,0 &tWillS &rw44 &FWSI4 Al'W48

PAGE l& 9 AUXILIARY FEEDWATER TO I OF .5 SG WINDOW 4 - (LSW4)

I ©1 AFW I_IIPJ I I I . I

THRU PiPESEG THIIU _ lEG IHIIU FIPE SEG PSILI TO SG A PSI2 TO SC II Pllln TO _ C

I . I I

_u,W_.Oly__..lSrWl ,6UpW_._W,li' Jt_#_-FT_7 _ AFWflO AFW_-IFT.-O_B /td_-OI_...4FT-CVIMI ,A.rw--GKY-CO-L_'088 I I ,o._. J i ,,,I J

"_."_I I _'._ _I I "_'_ I I'_. _ _"I I t I I I I I I I ' I I I I _ "_ I "'_" I _'_" '_" I '_'_ I sm,,E:m4 I

ArwsI3 AFW46 AFWBI4 AI'W4O ArWml_l ATW4I AFWIII4 #.]'W4I

PAGE 1 & 9 SB0. AT UNIT 1 - AFW 1 OF 3 SG (SBOU1-L) - POS 1/2, 14/15

/.TIIW _ |-II.

JIIW-ZlB--I'(_-_IWIIO SBO AT UNIT l&2 -- AFW 1 OF 3 SG (SBOUIU2-L) - POS 1/2, 14/15 SBO AT UNIT 1 -AFW 1 OF 3 SG (SBOU1-LS) - POS 3-13 5 • | 3"3¥d

IN_l.r"..OO-AII_j_- lr

I . I I

I I

I (NG.LT) (17G17)OS _" _-IOL01 M_-IVL IINFi IV OEIS 8 ]OVd

imm&1 ¢i_1&1

¢lIU-lkfACn-Air _7 _ _III-I_'A_'D.IIV _ _

(_nG'i x) (nai]) os _ jot oi Mjvt lINn IV Oa5 SBO AT UNIT 1AFW TO l OF 3 SG (LTDU) (LTDU13) I

_,.,l_'v.4rlt_ ._ II_W,,.,,I_I,,.,N_ HW41_',-_ .,IUIV...¢I_-.t16-,II4m Ltln/iS SBO AT UNIT .1AFW TO 1-OF 3 SG (LTDU) (LTDUI4) I -

IIpI-4_IV-4L'-4IIIO_rW-CI_I_.II-CIAI,,,,IIq_'IIIF-I_-ZCONI4p_w-aOlTI,.._'vgl I,mZIII

PP_E 7 SBO AT UNIT 1AFW TO 1OF ,3 SG (LTDU)

(LTDU15)I

tlllgl

PAC[ 2 .*- 3 SBO AT UNIT 1AFW TO l OF 5. SO (LTDU) (LTDU18) _ , ,

I .u.lsN_ ' i__Row

_ PACE 2, 4 & 5 AFW-CKV--OO-CV157 AIk'W-O_I72 U3

LL O o_

z k-

O U3

c-69 SBO AT UNIT 1AFW TO 1OF 3 SG (LTDU) (LTDU22) I FAILURE OF UNIT 2 TURBII',E DRIVEN AFW PUMP

I ,i,,Ivp I ,', p_ , ' '_tl il_ , _ , ? , V o "r@"_i F0'al "r©'_"l i(D' J _-30 e0al,i I_ _i,,0 j

...... i I 0 0 0 0 0 0 0 0 0 Ap_ndix C.3 Charging Pump Cooling System

C-72 CHARGINGPUMPA COOLING(CPCA) (CPCA)

PAOE 1 & 3 L 3"3vd

(LO_) (VC)clO) 0N11003 V cl_flcl ONIOSVHD CHARGING PUMP A COOLING (CPCA) (CPC2) I

I FLOWFROMPI_ AC MCC 1,11-1 SEGMENI"FISI01 • ISE(_I i_ OR _ SE(_ I_'101

I_-SWIOB GCW1815 EJI

"I'_ PSIO0

CI_-O_-I.I_-Sl'RAB i

CPC-ClO/_I_ CPC-MDP--4"R-S_I_ TEST AND MAINTENANCEON MDP SWlOB

! I i

,_ I ' i i I I 1

m ira,

__ _.... '[j !,..,4.1 ..' i_._1 I ...'- _*"° I ,..- ...., 0 0 0 () 0 0 "-4.. LD Q_ LP

C-77 TEST AND MAINTENANCEON MDP CC2B

I I ....I

,_ ! I I I I ! Oo t ._ 'I I ' .-.,|- I'"'I'-'I- CHARGINGPUMP A .COOLING(CPCA) (CPC4) ! CHARGINGPUMP B COOLING (CPCB) (CPCB)

, ..... _"""'1 I''_'-'''_I

@ i,,.4.,-i@1_4_,,,I

PAGE I & 2 CHARGING PUMP C COO.LING (CPCC) (CPCC) i m,_l'l_,w,

_ , , _ ,

¢ , C_1 PN_I Appendix C.4 Component Cooling Water System

C-82 COMPONENTCOOLINGWATERSYSTEMFAULT TREES INSUFFICIENTCC WATER TO R_ l-IXIA,CCWl

TO IIIIIIIX I_I,

CCrl

U _ 'll

0 COMPONENT COOLINO WA.TI_R _S_I'I_ FA_IILT TR]_.S INSUFFICIENT CC WATRR FROM UNIT t CC _H]_AJDER,CCWl3 '

I_. _I =_ i-_" I ? - ? TEST AND MAINTENANCE HT EXCHGR EIB

l-.m-_,,I o I .....I I i I '"I

I ' 'I I,, I 0 0 0 0 0 0 . ++++++++,+._'-- AIIM + " _+%

_ _+_ _ 'Silver100 WayneSpring,Avenue,MarylandSuite209101100 _ _ ".+t_," '_"¢_+/_ _'_+_.D +, ' _) Anociationlorlnformalionandlmmgemanagement _/ .++_ + _,_ /_" __ +++p

Centimeter 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 mm

1 2 3 4 5 Inches LLLL,po°+,+i,.o °°+,_oIIItl_

IIII1_IIII1_IIII1+

+ + •+ ..+ ..+-++/_,..

MANUFACTURED TO AIIM STANDARDS _ ,,,xx_ ._<,+ BY APPLIED IMAGE, INC. _

COMPONENT COOLING WATER SYSTEM FAULT TREES TRIP VALVE CC-TV-109 FAILURES, CCWt4

COW-AOV-MA--L0gA _J J _ _-AOV--iPT-|_A CC_--IBI--I(_Pi4A L_W--AOY --P0-|00A _--AOY-OC--i09& IA-LF-OUTIA li| COMPONENT COOLING WATER SYSTEM FAULT TREES FAILURE TO START CC-P-IB BY AUTOSTART OR OPERATOR, CCWI9

(_JS_III _ AUTOSTA_' io'oez_'ro" t Q

: T_?g_oggI c

A_ 0P A_ OF CCW-LF -PS [01A CCW-X me- !0P14.| CC'W'-ZU-FO- SPIB CC'W-XHB-10P 14.1 COMPONENT COOLING WATER SYSTEM FAULT TREES INSUFFICIENT CC WATER TO RHR HX IB, CCW2

a w8

_Wii CClit

O0 COMPONENT COOLING WATER SYSTEM FAULT TREES INSUFFICIENT CC WATER SUPPLY FROM HEADER B, CCW23

/ L_ 0 OO t,l,lp_lt T.vl_.

HFJLDII_R I (DU .TION)

cow,, _zzt ccw-x'v-P°-'C_°

SP-gl =-CLSI-IIR COMPONENT COOLING WATER FAULT TREES, TRIP VALVE CC-TV-109B FAILURES, CCW24

I CO_@AmqME_NY OUtI.LV=NF._ Q

- I LOSSor _ I I--i_s. _iR mrs qC-TV____S Ini ,SS,.S_0IcuC0LTZ1,_neC.wtOrrSq_ln TO r._-W-IO_8¥_II.U_i L088 OF I'UWCTIOIr / _._.v l OF _ , AC VITAl.! I . /k, ) ,,

CCW-AOV-PG-I_B C_W-AOY--OC-I098 EZll IA-UV-O_rIA

I ...... t_a

SP-51-O.Sl-I_

I_'_.-At_--_- l_g _-XI-_--_P_.I COMPONENT COOLING WATER SY;_TEM FAULT TREES INSUFF. CC WATER TO RHR PUMP COOLER RH-E-2A, CCW3 COMPONENT COOLING WATER SYSTEM FAULT TREES INSUFF. CC WATER TO RHR PUMP COOLER RH-E-2B, CCW4

ccv-tm.-ax ue c¢¥.-go-Kf uu

I IPTOM "hLN" _ O_'JLI_ YAI,IrIIII I

_-X]ln-l_--(_l LI //_ _ COff-XM'¢-FO-(X_Ildl t't'W'D OClt'21 COMPONENT COOLING WATER SYSTEM FAULT TREES INSUFFICIENT CC WATER TO ANY OF THE 3 RCP THERMAL BARR[ERS, CCW5

°I°

N RBI_L_f _ T_ t'_ 5L CCW-80-1¥LI(IA. CCV-IO¥ -_1_-t,KA CCW-SO-RYI LOB CC'W-AOT - IC-J_0_ OCW-BO.-RY| LOC CC'W-AOY-flC- f20C

CC'W-80,-.i_ LDO /,_ 6_ CCWL$ COMPONENT COOLING WATER SYSTEM FAULT TREES INSUFFICIENT CC WATER TO RHR HX IA, CCWIS

, i tsl_D. CC W£TJI|

TO RIIR fix IA 1 _5 _'111 r,.o_ol_.-.,I T CCW-N-ilTII_

c_Ir -ZM'T-PO..-¢¢I04 C_tr-Xl_ -..PO-¢C181 C¢1W148 COMPONENT COOLING WATER SYSTEM FAULT TREES TRIP VALVE CC-TV-IOgA FAILURES, CCWI4S

se. SS/COnT_OSU_Z SIGNAL /

_,6,ry., v.Es.[s..P.Si-.,.,/cc_,_ql z seu nlOJ_...Sl/ c_s-,

SIGNAL /

Z SP-81-CLSI-HK

CCW-AOV -FT -- I09A CCW--XH]B--10P 14_.t COMPONENT COOLING WATER SYSTEM FAULT TREES- INSUFFICIENT CC WATER TO RHR HX 1B, CCW2S

_W -_kfV-4P_.-_NL_01 OL_q-llO-IVLtOI O COMPONENT COOLING WATER SYSTEM FAULT TREES INSUFFICIENT CC WATER SUPPLY FROM HEADBR B, CCW23S COMPONENT COOLIHC WATBR SYSTEM FAULT TREES INSUFFIC[EHT CC WATER TO RHR HX IB, CCW2S

_ , ,I

_NIII

CCgftOJ COI_£PONENT COOLING WATER FAULT TREES TRIP VALVE CC-TV-109B FAILURES, CCW24S

lrltlP Y.dLI.YI_ ]

CC1_IB C_ff-XMV-PO- CC|00

0 _D

o_rsl_ CONT/IXMENt LOS_ Or CONTROLPOwJ_R - C¢-TV-_q.UOS l lSS_ AIR SYS. / TO C_-TV-LOeB21JLURE IOr 7 S_n 1.088 o, PUNCTION/ 1_ AC "V,,AL BUS it,, DURIN ON

C..¢W-AOV-0C-100B _/ CCW-A0_-PG -10_S IA - LIe-'OUTIA l_lll

:SP-SI-CLSI-I_

CCW-AOV-FT-IOSD CCW-XHII-IOIP:II.L COMPONENT COOLING WATER SYSTEM FAULT TREES INSUFF. CC WATER TO RHR PUMP COOLER RH-E--;_A, CCW3S

,,--.-I __

O_"W-X]IdV-PO-CGIH _ _ OI+"W-XMV-P• -I_'_C]II CCW+e:318 ¢'¢12 J8 COMPONENT COOLING WATER SYSTEM FAULT TREES ]NSUFF. CC wATER TO RHR PUMP COOLER RH-E-2B, CCW4S n I

DO0 <( l OOdEL

EL.r_ oLd(_9 -____ SY__ O_ 0 O- on--k--

L_j_ C [2::1 I,I

L _

C-I02 TEST AND MAINTENANCEON RCP lC

I I,,, i LA I I i ,_ I I i.... I I, I o-,. ._ .& . I I I 0 0 0 0 0 0 Appendix C.$ Compressed Air System

C-I04 LOSS OF CONTAINMENTIA TOP EVENT =CIA t,o-_'10 ©

!-,,,,-/ I IF-tO n-a'x3 ""-_-_"_ ""_'_"_ _"-_"'-_ ,_ U 0 0 v" 0 0 v" _ 0 I I I I I I ! .L J 0

I I I I I

I I I

NOIIONS _40 40 SSOfl COMPONENT COOLING WATER (CCWl) (w) I

i I _ TI_ TO

I ! J !

(_ _ PAGE 1 & 2 Loss of Turbine Building Instrumento Air Loss of Bo_h SA Compressors

I I 1

.. _ | © ._ i 1 I ' 1

r_"_'-/ ,,, _"'! i _-_'' ©' © @' ©' ©' _-i' TURBINE BUILDINO INSTRUMENT AIR-CSD Loss of Both SA Compressors - CSD

I I 1 ,_ i I I' 1

i | I I , _1o

8_$|-,_SP 1110U_6,-_! taJ_--0_PS-IrR-2S_l 0Li_l) fb(.TO,_.4ql_SA¢ e. 0LOSP S/_"0ccsr-FR-KI S_05-CPS-m-lS4¢I This page intentionally left blank.

C-112 Appendix C.6 Containment Spray System

C-113 CONTAINMENT SPRAY SYSTEM (CS-S)

ilUUJ _Jl_lW

[nuF lq,owI ms_ _ow I INSUI'FICI_.NT _LOW 3 css ,_rN *1 css 'r_, s I ,_ROM CS_

/EOU_-POMi3 ¢SSlMU6 ¢8S811118 HOUR--POSSL8

_A ' /_ CSS-TRA-MA COOLM_LS CONTAINMENTSPRAY SYSTEM (C2;S)

(CSS1Nl15) Pos_,2, _+,ila-.+l

c_ I _ F_'oI+(PuP I iF_uI.U_'OF120V ' I ! ' I .o,_,._J ._. .+-,

EI_ ¢__3 E_OH . g

I '+"'+' '+++

I I c_esm_.l ^c _ m_2 _:_-t_-_u'M_ I 1 ,,,cuccx.-2 _-_-_---,_g pem_oR.u'roI //_ I FAILUREOF4_V J _:3 /_IocT_Lcsmsm._l. I

OLS--ACI'-FA'-O_S_ CILS-ACT-QP--4_5_ + "+ Ct.S-ACT-FA-Ct$2CI.S-ACl"-0P-Q.S29 e CONTAINMENTSPRAY SYSTEM (CSS)

(cssl_51_5pos 5) !-_' !

i_,_ _ cc_ro) _G_ ,,_ rAL_or a,, I .. I I I .. _ .___ E'i_ ¢lleLira E4801H -

I '1

AC MOC 1141-2 I iFALtJREOF ,180V I IU| 8=S'D-WI-J,?e-S1O _'I_)"d0-J;)V-S'D _,S13--_'J-JDV-S'I0 _'D--d0-J,3V-S'I0

I_'_, I i-.- ! _ !_'_, I L__

I I I t--

I-_:_°I I I ''_ I ,'°°"'-_I I I

I,_.I ] t (C]l,I,NSSSO) (SSO) IAIS/SAS AV_lctS ZN31AINIVZNOO TEST AND MAINTENANCEON CSS MDP 1B

I I -i

! I ....I I I I i

I '_ i I I 0' o :"_ Appendix C.7 Emergency Power System

C-119 aSO'l-:_noH

L 30Vd _B.is-o..i-3-1w-,_)v

...... -if:)3t:lmlv..-I

I (HLmS3) (HL8±S3)HL$088NIS-U3A_Od9VDIUiO393 ELECTRICALPOWER - MCC 1HI-1 (4KV1H) (4-KVlH) TEST AND MAINTENANCEON DIESELGENERATOR1 ELECTRICALPOWER - 4801H BUS (E4-801H) (E4801H)

i FALURE OF 480V AC BItS 1H

'-'_ E411H I L tbll FAILUREOF 4VKV ACP-CRB-CO-14H1 AC BUS 1H ACP-TFM-NO-1H ACP-CRB-CO- 15H7 A,CP-BAC-ST-4801H

///_ _,CP-BAC-MA-4801H 4KV1H

PAGE ELECTRICALPOWER - MCC 1H1-1 (EH1) (EH1) I FALURE OF 80V AC MCC-1HI-1 &

UNAV._LABLE "_)UE

ACP-CRB-CO-14H14 ACP-BAC-ST-1HI-1 ACP-BAC-MA-[Ht-I 4_ C'Z

1 I tbll FAILUREOF 410/ _,CP-CRB-CU--14HIb AC BUS 11-I ACP-TFM-NO-11-11 ACP-CRB-CO-15H7 ACP-BAC-ST- 1H1

ACP-BAC-MA-1H1 PAGE 1 4KV1H ELECTRICALPOWER - MCC 1H1-2 (EH2) (EH2)

FAILURE OF I

480V AC MCC-1HI-f

E t2

480V AC BUS 1HI { BREAKER_ "" L 14H1_ I I _ .8o.._ ,.,__ ACP-CRB-CO-14H13 ACP-BAC-ST-1H1-2 ACP-BAC-MA-IHI-2

I I tbB | F.tu_or4_ AN_-2_,,.T_%21_tt AC BUS 1H ACP'-CRB-CO-15H71

ACP-BAC-_-IIHI ACP-BAC-ST-1H1/ 4KV1H I PAGE 1 ELECTRICALPOWER- STUB BUS 1J (ESTB1J) (ESTB1J)

I FALl,HI OF 4KV AC:STUB BUS 1J

o,, FALUI_ OF 4VI

/_ ACP-BAC-ST-STBU U ACP-CRB-CO-15J9 &CP-]L,_C.-MA-8TBIJ. 4KVIJ

R UB PAGE 1

ACP-XHE-F'O-STBSS

_)t._.-kOSP I I f___ 4F U I i I

ELECTRICALPOWER(4KV1J)-MCC 1J1-1 (EJ1) oT_

MOU_IlI--Ila 0._

HOUSM-MI ! I I I

• _i_ t__o . ,

HOVS-LO_JP HOU_-LO'JP ACP-lU_C-MA- 4k'V_ACP-6AC-ST-4k'V_ ' .TAn-VRL51 J TEST AND MAINTENANCEON DIESEL GENERATOR.3

I ! J

,_ [ I I I I

, i I"''-1 I'_'_----'1I"',---'_ I'_'! 0 0 0 0 ELECTRICALPOWER - MCC 1H1-2 (EH2) (EH2)

FAILURE OF" / 4BOV AC MCC-1HI-_ , r 0 FALURE OF fJNAVA],_&BLE "DUE TO 4-80V AC BUS 1H1 M E

ACP-CRB-CO- 14H13 ACP-BAC-ST - 1H1-2 ACP -BAC-MX - ! H | -2

I I tbl3 | FALURE OF 4KV AIJ-J--CI@_--U(J--]g.H]_ AC BUS 1H ACP-TFM-NO-1HI/ ACP-CRB-CO-15H7 I ACP-BAC-MA- 1H1

//_ ACP-BAC- ST- 1H1/ 4k'VIH / PAGE 1 ELECTRICALPOWER - 4801J BUS (E4801J) (E4801J)

I •

FALURE OF 480V AC BUS IJ

E41?'J t i tbll FALURE OF 4VKV AL;l-'-(.;I_I:I-C(.)-14Ji AC BUS 1J ACP-TFM-NO- 1J AC:P-CRB-CO- 15,17 ACP-BAC-ST-CSO1J

Z/_ ACP-BAC-IvIA-480"LJ 4.KVLI PAGE ELECTRICALPOWER- MCC 1J1-1 (EJ1) . (EJI) i FAILUREOF

480V AC MCC- U1-1 &

o. 480VFAL_AC BUSoF1,11 u,,.,,,_.,_

ACP--CRB-CO- 14J16 ACP-BAC-ST-UI-1 ACP-BAC-MA- tJ t-.

[ I tbll AcBu_s ^_-TF_--.O-_J1

ACP-BAC-MA- U1 ,,_ PAGE 1 4K'V1J ELECTRICALPOWER- MCC 1J1-2 (EJ2) (EJ2) I FALURE OF 480V AC MCC-1J1-2

2 FALURE OF 4BoyAC BUS IJI (BREAKER_ _ e " _, 14J14 / 4BOY A_I-Z ACP-CRB-CO- 14J14 ACP-BAC-ST- U1-2 ACP-BAC-MA-1JI-2 zlz I L tN3 FALURE OF 4KV A(._l-'-UI'(I::I-L;U-14dll AC BUS 1J ACP-TFM-NO-1J1 ACP-CRB-CO- 15J7 ACP-BAC-ST-'LI1

z//_ ACP-BA¢-IWA-IJI PAGE 1 4KV1J ELECTRICAL POWER - DC BUS IA (EIA)

I25V DCBUS 'i_ I FALUREOF" I

0 I I 125VOC I:OVER I Ik FII_)MUPS 1o,2 1_FRCU UPS I_1 I:ROMB**_J_

lk2 lkl FALU_ OF I FAL_ OF'

_ I I I I F,_LUREOF 480,/ FNLU_ OF 4a(N AC _ 141--2 AC MCC 1.11-I

(.2 _1 PAGE1 ELEC'TR_ I:_ - DC BUS IB (EIB) (EB) I

l0.,,,a I _-ST--BUSIB I ,lab iPOtt_ TO OC BUS 12_ OC PO_t _ 1B FIK_ LIPS1BI 1B2 fROU B_TTIDW18 ' ' !

I I I OCP_-2I I _ IX:PI-CRB-CO-23 " ICP-EAT-LP-..BATIB IX:P-(X_--LP-l_r I_B

AC U_ 1J1--1 _ AC Id_ 1J1..-2

/_ - /_ /_ .,-.,,, PAGE 1 u*-mBL F.J1 £J2 _A_ TEST AND MAINTENANCE ON BATTERY CHARGER B1

I I I I I ,I

@ O @ O TEST AND MAINTENANCEON BATTERY 1B

I i I

I I I I I !

I, I r" l'--',a_ I,'__ 1"--'-_."_1'I"-"_-':"T" ' I--'-I"'" 0 O 0 O 0 0 TEST AND MAINTENANCEON BATTERYCHARGERB2

, I I I

,_ I I I I I ! .- -°r- @- 0, 0 0,

ELECTRICAL POWER - VITAL BUS 111(E2111) (E2111) I

I I I .-I

N_L..I_g-_0--11 kDIi..1_1_..¢__,lllgII I !

I I"__'t'_ i I I_'-''''_T I'_"_ I_"1 I'_"'1 I__1 r""l I'__"1

_ _ AOII'0iV"I_-UP_Sl AOI-0tlW_ ACP0--TFM--NO-II I-2 ±o ACIII-0CRII.-(_-flCI_ I I

,, I I I_ I I__'_1 I_! I'_'1 I_'-_'_'1I'_'_i [- '

__l 0 .II_..0._ItB-.I_I_IE A_I_0TFld.-N_IBI-t _ OCP-Cge-4:(_)D-94 _ OCP-0a_.,-._-.gL,SlB _. I I

I I I I i I I-'"-"1' I_'_1 I_*"_1 I_-,,_ i__1 I'__'1 'I."__'_''_ " I

+ OCP-.CgB0-CO-_3 _C--lIO.d0PS82 _ 0 N:P-lt'M-N0-10 _-I 0CP-80_I'-4.P-8_ III 0¢P-CCf0-LP-Bl"lqB TEST AND MAINTENANCEON VITAL BUS SUPPLY

,,,.iI

._ I I I I I.... I

i' "' I ! I i .9.I !"'_"1 I".,_.:-_-I'-'-.__r ''='1r"_.,_,-__.-'--"_ -.--. 0 O O 0 0 O ELECTRICALPOWER- VITAL BUS lgl (E31gl) (E3111) i

I I

'¢"="-' I (_ •1_ aa,.-lw-l_-I)lPml ._¢p..le/-dle)-UiNJ_ I I _--_IM_4gO-_2-Z _ ._CP-aIB-¢O-_I_

T 041

! ! I r_ar I

, , PAGE I _T-LP-IMT_ 8C_L_F-LP_IT_R ELECTRICALPOWER(-E4VITAL1_) BUS 1V (E411V) I

_lf/L U IV/

I_''

f i I_'--_'1 l_-_

I' I F_LUE (_" 1 I I'11_1moll iJ_:l I _'__,_ _|

, _ ll:cIk11-12

_..Ii.'rlI IIICID"aO"(I°-23 _ OOI_IOC'II"8_S Z_ I _ I I

_ I_,,.IIII-II Appendix C.8 Emergency Switchgear Ventilation System

C-143 VENTILATION SYSTEM, VS (EMERGENCY SWITCHGF_,AR ROOM COOLING) (vsl)

I I . luliun,_ wA'i_tt mm,IlmquryLv_nnnmu.upwwr_AnWp, ll_JmrlP ¢IllLLtO Wllllt .u_-,noum.,.p _ u,,. aj nj_,[_.,n,.o.,._.wi_.m_ocj

GYg|OIS 0•9K715 GVSIOI3 • o'

C-145 VENTILATION SYSTEM, VS (EMERGENCY SWITCHGEAR ROOM COOLING) (GVS1313) I ]_su_XIRcooL,N, F.p_.UNN,..N,C T,

IAHU I-VS-AC-7

o_ CVS L313

[ I LOSS OF FUNCTION t-_S-AC-7 FAN MOT_)R 1-_S-FM07 POWER L IN AIR HANDLING VAILS TO RUN FOI_ FAILURE OF 480V j UNIT I-VS--AC--7 _ISSION, I-VS--FMO_7 / MCC-IHI-I, (____) (____) V S -A H U -LF-V SAC7 V 8-FM O-FR -FM07 VENTILATION SYSTEM, VS (EMERGENCY SWITCHGEAR ROOM COOLING) (GVS1613) I mSlL_ CH-.tr- WA_ sv_, F,0M Ru,ms9

O GVS 613

I , I

I _,_,_I I,.sor,,cEzrrs. I _cTT.,cRAIN ..._A LTI s

TS-MOV- SC-PI07A _ TS-CHL-FR-VSE4A /_ GVSlgI3 GVS _825

I I c,-.i-,.o-, ,o.. ,._s) (:.,_ _ ,.o_ o. :o._.o,.

F_/ SILURM_R -IOFJI-,480V A I POWERPOWE]__ TFOAILUV,TREAL OFBUSIZUVJIt,. AC

EJ| E31I]| VENTILATION SYSTEM, VS (EMERGENCY SWITCHGEAR ROOM COOLING) (cvslTls) I

_oM lrr.A_ n w._r,,, j

GVS 713

<_+ "+"+'_ I '-v.-.-`"I "+'+,,+,-I. s-_,---,,.-iswmL _ K_ .>x --MI_--rFR--VSI_ _--F'_--"V'mmlB --PG--Y_91 --BI_eI.OWDJV'ER OV-FT--PGI07B I I _u_ rr_Fr plUt.a'ro ntrx p'Oml cp-+.,,,,:-vs-.,-,,,l i e` Hill_ON I

Vg-tN_-I_I-_8,1B VS-tlltL-t_-Y_B TEST AND MAINTENANCEON CHILLER 1-VS-E-4B

i I I

,,-, I ! I ! I

...... I i i 1 I I 1_ !;_-t"O I'_O=t P_-t'0 I'_'t'O !'==O'i VENTILATION SYSTEM, VS (EMERGENCY SWITCHGEAR ROOM COOLING) (cvslsla) [ m_mn, cHn.can warn SUPPLY rnOU aOW_qO lCHILLUt 13An; C ]

CV8 +813

t-,O o L 1 |VS-BCYLOwl)rv_R _ THItOUOHCHIIJJ_ I wrrmN CHLLUt S-MDP-FR-VSP2C / ,-v.-.-,c/ ! T._,.

S--CHI.-FR--VSE4C CVSlUI3 CVS 8Z5

(z-v_-_e-_.'.-"-,_-.ss) ,c__---_J__£__ e- F,_tms oF 4eov ,_: coNvl_deo,,wm_Y),ILV_ OP [ .cc_z.,_, ] ,2povlo.ze:.o.c,vsz-v.,qr.] L| I !

2EEII - l 2E3111.1 V_TZ_ON SVSTEUVS, (EMERGENCY. SwI'rCHGEAR ROOM COOLING) (GY81913)

I !

I Ir ' _

I I VENTILATION SYSTEM, VS (EMERGENCY SWITCHGEAR ROOM COOLING) (Gvsllot3) [

i i

-t

_-IT-4LII01 _-I0T_ IBII@--EI"II-4lPOar-sel

I '_"'__"'I_ I_" _' "_'?'_

SW©I-KMI -APII.O0 ¥ II©-_ -PT-PA K vm+T,.+,.Tm, re'STEM,vs (EMBRCBNCY'b"WITCHGEARROOMCOOLING) (GVStili3)

I ,;,,

I I , I m_ w nn_ _ ....cud== " Ira' toLD _ _ 10 ll[W_ TV]Ilql,

IWI llll

! I I+E+¢_ I_""='-'=++"'=I.oJ

IWl-0Jl,(_l'-IlO-I_ I_l'_©lLY-lt_-IVlO0 I

9 i _ I I " I="._"-_r=--,_ H.-"_-_,_ __.

IICI'IILI _ Ul I11-IIl'll-IIl..-_lllmO I I

4_-_'_=-_1=....==l--. l o01,.,. Pol

IW©i..IIPI-JUIIII I I_l-¢ct_-IINt-11._)1111_ I + i , _=,I n-v=,,o,-Ip,-se,+/ ,,,,.,I_-Vs-P-uk, o,== _• Im !.=._=.=_.._1. I "" " "" / I_'=" _ 1I_,, c _.,,.

TI-MOT*TII-TIFI¢

Q I I n..-..=....., im-_;=_p._

IIP-@ZlI-J_II+O0 V_-IW-0PkllO VENTILATION SYSTEM, VS (EMERGENCY SWITCHGEAR ROOM COOLING) (GVS3113)

,NSU*V,OCHI_.,L_TL_AEDT_R F_OM1 OF 2 SACK_P

0 GVe 3113 t._

! _ i I_SUFFICIENT CHILLI_D I_SUFFICIENT CHILL_] WATER FROM BACK[.]P _ATERF_OM BACK_I _HILLER/<,I-VS-E-3_ [CHILLER/<,I-VS-E-3_ GVS3213 GVS3313 VENTILATION SYSTEM, VS (EMERGENCY SWITCHGEAR ROOM COOLING) (GVS3213)

I ,,

W4T_R FnOM BACKUI_ CpILLER I--Y S-E-3A ]

_ I ! , , I i ! _'_ I SYSTEM I C_IILLF..R t--YS-E-3A'I_ C}4ILL,R ,--'VS--E--3A i C_,LLER I-'4S-E-SAI _LI:_.R I-VS--E--3A / Ico_,,FAUL,,cUa,T T,O_j,Pu_PIT_,-vRUs-e-N Z¢_.HR,nq| s ...... /II-V¢.zS-Z78c,¢vPAzv_LUCS I | fAULTS I ""_""_"G'soVS'_C'C""Cc"OWE'_SJ_-P

GVS _1 GVS: 234 Ell VS--MDP--FR--VSP3A V B-CK'V--PG-V'_7a

I i I I I OATvA,, z.v, I I e...A_VA,.Vl I IP_-t.OO_,. I clm.uz_x-vs-z-.,I _v_,o,_Tr,,coz ,-I

V'S-XVM-FT--VSE47 VS--XVM-FT -V-qP.01 VS-XI_-F0-FCA 10 VS-C HL--IeR--VSE3A V S-EVC-F R-VSE 1A VENTILATION SYSTEM, VS (EMERGENCY SWITCHGEAR ROOM COOLING) (GVS3313)

[ 1 I i i cAT,v.,.v, I I cA,,v_ i / vcA-,.ooexvRi c/h,,.,Jr_LS . ,-TOv,m-,-3,1m I _Z_wNSvZRoeXT,-vs-_--sn_co_-',

Tbb--%OPENrS--247 ORFAILPLUGSS _| Tll_-VSOP-_ENt ORFAILPLUGS S_| AISHU'_TDOWNrACRMENTAR"1oEA YEN1'Yl]_ / 24_ HR _ J ||FA|LS _LTO W--nUN ....Il-

VS--XV'M-FT-YSE47 VS-XV'M-PT-VSESI VS-XHZ-FO-FCA 10 VS-CHL-FR-VSE3B V S-EV'C-FR-VSEtB VENTILATION SYSTEM. VS (SEnVICE WATER EIY'STEM INTERFACe) (GCW1813)

I , i m*,momu j I%. _5

I I

I [ t "__ I I'-" J

[ . I I

-,,,.1 r I I m,.. -- I

4_-I_II-r6-IIWVII [_q2-1'b_'-IPII-I_q_ IVI-©]I'_I,i _- IW-I©PII.-MI,'IPI_ ..___ © © I I , laiT , i _

nr..A0V..Oe.l_WIM IA-I,P-MI_IA _I_N)-I'I_

I ,-,w-,m rlm.*I

lw-Aov-l.ir-iwIIII C-158 Appendix C.9 High Pressure Injection and Recirculation System

C-159 C._ Z C/3 _cN _2

.i.J i "_'J .... I

'? I _ !_.._ Ao

L_LJ k-- IlCf) O>-- U') L_ C_ _2 <_ I,

C-160 FAILURE OF FEED AND SPILL USING HPI SYSTEM - WINDOW 3&4 (FSW34H)

l AJPID gPllU, UIIINO liP[ I FJU]I,U11J O7 I _5

O'l

]liP[ Blr8'TB)4

FAn.U_ OF ]

I i , FAI],Uml OF liP[ TO CC_D LIlG TO HOT LEO /-,. /_ I)1 _tllEl.a ELI.13-Wl I11[] rl,'_lt' _ 0.3 IZlcl"i

' g, Sn8 O0 I

_I_ ---'-I_ E _ I r

_-¥_"_ I 1_I I _'U| ,,r_ , /

x/ ' x¢_ x/ ' x/ ...... D__-j.II_I3)_I...... Si I I l I ,I _5

I (Ta) (TCI) DI,I,VIAIOJ_flV- l_IOIJ,D_[l'l_II _[HI1SS.qHcIHDIH C-163 TEST AND MAINq-ENANCEON CHARGINGPUMP 1A

I i._,._i

1 I I I I ' I

#. ! /'" #. ! #-.. I"-"1 l,_ I'_="=I i-!H. 1=-!'.... I.==1' 0 0 0 0 0 0 HIGH PRESSURE INJECTION- AUTOMATIC (DI) (HPI6) I

,, ,

P_5 TEST AND MAINTENANCEON HPI MDP CH1C

I I I [% I I ! I ! I

i' I I I I , r" I--=,,_'_ I_-_-"_• P,--..-_i O O O O O 0 HIGH PRESSURE INJECT'ON- AUTOMATIC (DI) (HPI7) I - ? <_ I I ':

PAGE 6 HIGH PRESSURE INJECTION - AUTOMATIC (D1) (mPl8) I

N.,,_CI'-Ir,4,,-_

PA_ 7 TEST AND MAINTENANCEON HPI MDP CH1B

I I I

I I I I I

i i I I !

0 ll,_0..llp,..llp,,-,c_I 0_ _..II0_,,,,Q, NI 1,40=_I,OP,,,.,I_CIIN 0II01_I_.,I) HIGH PRESSURE INJECTION- AUTOMATIC (DI) (HPI9) I I,__!

I I I l_,_,_ I,_<.,_,1__.!_ " i ? ' I, I ? I

PAISE 2 & HIGH PRESSUREINJECTION- AUTOMATIC(D1) (mPl21) I

.,.,I 81"1;_ i_3 II,t_iY _ 101] I,ilidkl __1 _""_°' o_-,iI__'_ I I I 1 I I 6_H

I I I /

V "_ V V \/

I 1 l J

OI,iX a'lO,,_ll"4:- ,_i

e ,_- / I-'°-__I_'_1 '

(lfl) 3I.I.VI_IO&FIV - NOI.I.D_II'NI 3_$S3_d tIDIH HPI TO HOT LEGS - MANUAL (D1HLG) (DIHLG) I AIL HIGH lqumm FLOW HOT Ll_8

DLIl._

FIIOMO44A0_I0 LIE08 DUB TO OHl_g" 8, |v_vm rAn.ummmJ Ft,t,P_

.'qlPW'HLG

/,,__ ,,_,,,, ,,,k,, ,,_

I' I I _

Hq9

i I I I I I

I JLOOVAc _wr_ I A_m,°s I IF_L_EOF_ I_ (_ _ovrm, Duot-. I f_o__-_.- o_o_,i__o _ HPQI IE1B _ 41WlJ ,...imlll-o.,._._,--C:HIB HIGH PRESSUREINJECTION- AUTOMATIC(D1) (mPl7OO) I

Mll _ L'_,_':i_''"'_' .;7-"

PAGE 6 HIGH PRESSURE INJECTION- AUTOMATIC (DI) (HPIS[II::))

PACE 7 mo_ p_xssu_ mJECTI0_ - KAHUAL(D2) (D2) l

_" z_ z2a

I i I

_ Elm _ HIGH PRRSSU[ _ INJECTION - MANUAL (D2-23) (2/3 pump,) i I

I '1 I I

! 1 I I , I 'i

, ! ,

i i , I __J

Z__. I LJ E_ (./-) L Illcn E_ • N

m -T-

C-178 HIGH PRESSURE INJECTION - MANUAL (D2) (HPISA) I

I I I I I I i--I_.- i _°'_. [ Bus_ At f...... _ Bus_ if _-_., __-_ ° /XHPIgA _ /%4KVIH /%E1A /%CPC¢ _-J-- I I I FROMCHARGING , _ __.. _ _.___, FALLRE OF" AND MAmll_ICE DUE TO ' V _ AC POWER IIPI MDP C]-1 BACKFLOW TO virAL BUS 1-t

[--HPI21I iI___'_'' El !1 _ IdA--CIIIC T HIGH PRESSURE INJECTION- MANUAL (D2) (HPI9A) I

! ,, I ,| I

, , , , , '

PJ_GE3 & 4 ""°°"°nWayne'°Avenue,:_oSuite'°r1100 °'"°n _ _"_" #._

Silver Spring. Maryland 20910 _.._,__'_1.' _._._ '._-_ '_2

• _c_\ _ __

0__ BY .PPLIED

HIGH PRESSURE INJBCTION TO HOT LEGS - MANUAL (DZHI..(]-23. 2/3 PUMPS)

I ='' '" I

ICe _ 4i0_1 _ [HI r.i1

14=D! [211 IAk-CHII FICHPRESSUnNJ(£ CnON- Re=SE_.S(03) (03)

I(Plal IHl(i_ li(d 5

I I I__1 I"

I_1 14PIt _ (! 14_1

HIGH PRESSUREINJECTION- EMERGENCYBORATION(HPI-EB) (D4)

, I I ,I I ,_ _-'t I_ -_ _,,'_-,_1[_ _ I!, _.__' 1i-_ _a'._ _1, ee , t o._ _l I _*_'_"'_I ¢ I ""_t' _

_-_-_°-_ _ _'_-_°-'_ _ I I I ,, I

EH2 E_ HIOH PRESSURE IRECIIRCUL_ATION(HPlR-H2) (u:2) I

ll-tE REI_ M_OE

I I

[u_,_rrL_,s"l_ow/TooI I,I ,__n._ ow__'_ I!

FROM t_ MW FEOM t_ MOV lq._ FROM ROW. F_OM I_ I_ CI-IM_GINGPUMP _ PUL_ UDP-CHIB MOP-CHIC

LPR-HHB LPR-1-14A HPR8

I mr a_ am_ I r m_'ra_xT ,_u_E,_- /_- -- /\ ,_ ---_.-.- (3 /_ M&-cn_uk 4KVIH _ !F4_11 1.41_13

MA--CHe HI>R_ I-I>R_ HIGHPRESSURE RECIRCULATION(HPR-H2) (H2)

1 i I_'t._l !"¢',,'_1 Y ' ' _4e..i "

I"1 _,,,,.,I.. H.m

Ifll'll ItPIIII

PIIG, ( 1 HIGH PRESSURERECIRCULATION(HPR-H2) (HPR14)

NIl-UO_m._gtl NR-IW_

PAI_ 2 HIGH PRESSURE RECIRCULATION(HPR-H2) (HPR15) 2 HIGH PRESSURE RECIRCULATION(HPR-H2) (HPR13) P4_

Eli _

PA(_ 4. HIGH PRESSURE RECIRCULATION(HPR-H2) (HPR8) F__

l,i,l F__

PAI_ 5 HIGHPRESSURERECIRCULATION(HPR-H2) (HPR12) I I_-_ I

I i

,II1_11 Ill

PA(_:6 Appendix C.IO Low Pressure Injection .and Recirculation System

C-193 FAILURE OF FEED AND SPILL USING LPI SYSTEM - WINDOW 1&::2 I _ P_iliI.go

2 LOWTOPR[SHOTL[OSSi I_SLI_CO.B TOOP[H FAILURE OF FEED AND SPILL USING LPI SYSTEM - WINDOW 3&4 (FSW34L)

, i FAILURE, OF FEED AI_D tOPILL USING LPI dh F_v 34L

LPI SYSTEM FAILURE OF i

I .E I , FAILURE OF LP! [FAILURE OF LPI

TO COLD LEG [ TO HOT LEG

I_-C LPIItLG LOW PRESSUREINJECTION(LPI) (D6)

! F'=r,.%..'.'.'_I

IA-Cllf-41,1r-of_l _--C_II I.Pl-O_-ier-_ll _-FI-OAII

f i , ' i -" -i

' 91-AOT-,I_I,-IJIJ_ I/,,I i_-._--_ I t-,,AGT4A_ I Z /,,, lilt,- lll_k 1"4a1014 _ I_,laOU Irll

I I I i l TEST AND MAINTENANCEON LPI MDPSllA

I L_ I

,_ I ' I I I ! I

. I .--..,! I I ='1 r"..-_.., I .'_rrr--,,I :_'ti---,--I" | . 0 0 0 0 0 0 TEST AND MAINTENANCEON LPI MDPSI1B

I 1 ! 1 I

1_. I=..-.t-1or__iJ F=_It .. .,----/ °'-_^_"t ...... I , ,I.._oo,,_ Io=_,o,,} [o,n_ ^_u",._,_'e:,_,_'_'I - ..!I

0PO_,...I_ LR-k(0}P-MA-SltB 0POS.-_ I.PI-0MOP--I_-SltB 0POS-.-R'IO LPI-0-klOP_IB TEST AND MAINTENANCEON LPI 1862A o i' i nS_ i' _

i i l i 0 0 0 0 O 0 I

i I I I i I"'_" _1 ,.,.. ,,. I ,,_. i_.,_,.,_1 "_" '_ I "_="__

I I I ,, t

I_-II& - '7,,.' ' /_ Loss of LPI Flow Path (De-CO) p

I '

I I ! I I

!._.--_1 I ---=-I I_-=-I !___i I-_=-I

_! i _, LOW PRESSURE INJECTIONPATH TO HOT LEGS (D61-1LG) I Prom/

CKV-(X_-gr-HII.GI RII'T--TNK--Llt'--RWS_ "CKV_-FT_ PJ:M3LO

_ M,t_' TlomuIm I

lJ'LOWTO THROUGH PIPIt THItOIICH lP'IPSI MOT t8800 8NMNNT INI4Y MOVlNOJt mmMltirr

D4J30HLO DSdOlILG D831HLO De,_IHI,Q LOW PRESSURE INJECTIONTO HOT LEGS

(D650HLO) I _LOW "_A

I I &

_G1 I, '_:t_-l_._ ION,-,_', _|

MA-51_ LOW PRESSURE INJECTIONTO HOT LEOS

(D631HLG) LOW PRESSURE INJECTIONTO HOT LEG

(D640HLG) I MOV181FROM_B LOW J HEAD_ SiFLOW_ TOB JI

I I

...... } _S&UFFC_ro, ij u_5

|°N_,_'_

I_-Sl_ LOW PRESSURE INJECTION --I-0 HOT LEGS (D64J_HLG) I _ol,,uvl,,,o._.F_.o.,o,_..o/,o, HEAD SI TRAIN B[

! I INSUF FLOW TO [ tN5 MOY 1890A [-I_Pr'MoV-I_G-=I_I_O2" TP._IN B

D64" -ILG1

I I & LOW PRESSURE INJECTION TO HOT LEGS (LPIHLG) LPIHLG

u LL

_2 I"'_I_-_-..'* i_'_"--*'- ,"-*"',

LPt_IO0 LM40Q LPISOt LP'ftOI LOW PRESSURE INJECTIONTO HOT LEGS (LPIHLG) LPR6 I

P_5 0 I 0_ _J V

W _J k-- 0 I 0 rY Z 0- o J 0 W Z

W n_ to w n_ 0_

0 _J LOW PRESSURE INJECTION TO HOT LEGS (LPIHL-O) LPISO0 ! f_on.ax TO I

£48014 E_

e 8B-SJ-cD_dl O_-O0-A)D-_ 0 0 8s_m_ I _w xoe_oI I I i O" v 0 0 0 0 k)

I I LI_ I l \/ I-: I

(97Hic:17) $9:17 /OH Ol NOI/O31_NI 3_INSS3_Id AAO7 _J t

0 I

0 z ,q- o __ _J 0

U3 ka_l c_

0 _.J

C-212 LOW PRESSURE INJECTIONTO HOT LEGS (LPlt.-ILO)

LPI401

i mM D 81 TI_IN !_

uov 18o_ F.61.U_ _ 48_ F_LU_ OF _0_ _., e_o_ _ AC 8U5 U 0¢ BUS B

I I I

e LOW PRESSUREREClRCULATION- TO RCS (LPR-LH) (HAS) LOW PRESSURE REClRCULATION- TO RCS (LPR-LH) (LPR4) I

!,,,',._I = I .r_i

i I , _ _.._ !,_",,.,,i !__1 ------¢ I '_", .t,i .. -

P_5 LOW PRESSURERECIRCULATION- TO RCS (LPR-LH) (ERR9) I

P_2 LOW PRESSURERECIRCULATION- TO RCS (LPR-LH) (LPRIO) I 25

,..., J I T . / , i i

E4OO'd £S I_-SrlB LOW PRESSURERECIRCULATION- TO RCS (LPR-LH) (LPR11) i I_..mi

I._-=-i

P_ 4 LOW PRESSURERECIRCULATION-TO RCS (LPR-LH) (HAS)

I T I , I_'_ I_'_! r__i"_°-

I I I , , I PS31OR PS3a _ _:GI4[M" _ _ psJo oR ps37 ['--[F__TS !_ F_JLTS I [_ r^u._ I [-F__UL_-

LJmll i._10

i I I i I -..... I /,,, ¢,. /,,, ILt-Sl_ e e (4aOtH £1A LPI TO HOT LEGS-RECIRCULATION MODE (HASHLG) i

< o

I |

° 0

C-221 LPI TO HOT LEGS-RECIRCULATION MODE (HASHLG) LPRIO1

• ..... I , ,

TO BACI(IEN(N_ 120V tbll Ib_ I'rHeu M:_ _ II_'II I ' ' ' t-1.9PD SIS_'

E4BO'U E1B - -I_-SI_

LPi-_P-FS-NI_ LPI__ LPI TO HOT LEGS-RECIRCULATION MODE (HASHLG) LPR800 I

I I _'_'m,_o,,_-,,..,,,.! "",_%'i,-,, I_'- ,_ _ I_ " "__" 1

E480l'1 E'_ I_-_11_'. LPI TO HOT LEGS-RECIRCULATION MODE (HASHLG) LPRS01 LOW PRESSURERECIRCULATION.- TO HPR (LPR-HH) (H1)

I I

PAGIE1 LOW PRESSURERECIRCULATION- TO HPR (LPR-HH) (LPR-I--IHA) I 9 , I__i I._',mI 1- -I

I.iiil,.llvalll,. gWll i,ipl-iil_..llo-III1_ ,_

PAGE 1 a_-_4-'d_-W'} Og/_-OO-/¢O-d31

_ I_lJ -_rt_ 1 I I I /

L _dl

AtOll .4_/ I I • (I.HH-_d-!.) (HH-_d-1) _dH Ol- NOl_LV-Ir19_IIO3_I-q_rlSS3_ldMO-I _d.16..._._l-,jl_p-_I_ltCl3"-M1

.,,¢.._,,I _ ",,,/ "7

I (_HH-_d-1) (HH-_Id-1) _dH Ol- NOIIV-lrlO_lOq_l Et_rlsSq_d aO-I LOW PRESSURE RECIRCULATION- TO HPR (LPR HH)

i FLOWFI_M LF_ I--IPMJIRN A n I ,.. I I i DCBUS_ 14 Cum_ FALS"lOaM ehempl_m_l FOR24 HOLm

[I_ [40011 l_lpo _-FR---A24HR LOW PRESSURE RECIRCULATION- TO HPR (LPR-HH) (LPR-HH5) LOW PRESSURE RECIRCULATION- TO HPR (LPR-HH) (LPR-HH2) I

P_4 LOW PRESSURE RECIRCULATION- TO HPR (LPR-HH) (LPR-HHB) °

_ '

P_6 LOW PRESSURE RECIRCULATION- TO HPR (LPR-HH) (LPR-HHE) I

P,l_! LOW PRESSURERECIRCULATION-TO HPR (LPR-HH) Appendix C.11 Primary Pressure Relief System

C-235 1+_ I_"w^o_'l_ _uo,JI I'_--'1 I I I.-.=-I I I i I

! I

I [ I i I='"=I - L j L l '_D

I--lI

i010 GI.1'04 I--!I

(LA_IOd) A_lOd H108 3S079 01 3;9N71V-I PORV FAILS TO OPEN ON DEMAND (PORV2) PRIMARYPRESSURERELIEF - l OF 2 (P1) (P1)

I I I

I I I # .a ' ' 0 ¢I

I i I I I I J ""

£1B Idk--t4_ EI_ MA-NS,_C

I I I I

AC MCCttt-Z PRIMARY PRESSURE RELIEF - 2 OF 2 (P)

(P)I

I',_,,I -t I'__u '! _I__ ,1__, i

Em £'R Im(mm-_ I_HmClk"l_k°_ imM_lp____ti,q_

! _ 1 I I

PAO(_ 1 PRIMARY PRESSURE RELIEF - ATWS (P2) (P2)

i.,7_=._i l''IW_" I I"'_" I I'__'" I r..- .... , _.__:.__e..e.,

a TEST AND MAINTENANCEON PORV 1455C

i I ! o I I I i I

!"I _'1I I'--'_-4_I I__I m-_"rI !"-_"4_1: I I'_' I _'1 0 0 0 0 0 0 TEST AND MAINTENANCEON PORV 1456

I I i

| _a, I

C-243 _ V

I I

aJtl-'Dd'H _ _IO.-,.gd--AcIq-"I_IS

_uL Va_ _ I vm_=mk =r.n 1 ! I

C) "_-" ,

L_ L J. J V

! !

t_ (L3) (asi) NOlivqnoaio3aAvaas :l(:]iSNi INSIDESPRAY RECIRCULATION(ISR) (ISR2)

" _ II:l --

I_ MIo I,m

J •

7,, I I l_s i,_,i',Isl I _os3-13 I

FRON [SR P03 3-13

P_fP8 AlVD/On corm

II_1N313

KOU_IM--SDOl HOUSM-gSOR HOIISR--LOffP

I I I , ! I i AC BUS IH OF' PIR[ SECME3fl" Containment FNLLX_ Of 12(N PS60 FlOW Bump Plu_d OC I_

E46011"1 S31KIJ_l(l 3UMPPLUG EIA

! 1

PS67. _ SEO

FLOW THROUGHPIPE ITHROUGHPPE SEG

ITH_t_H _ S_

S',_3-MOV-F:'G-106A V_I_V SWS-Ivl0V--I_- 1068 J I I

_12 5_D INSIDESPRAY RECIRCULATION (ISR) (ISR2) I

I I I I I

" I f III_IFLImI _ I

aPA'5-IIOV_-'iDGA _. _IIq_-,Id_l_lO01I i" !

PA_..6 I I FAILIJ_ OF OUTSIDI! SPRAY I_CIRCUI_TION

t I

1_8',_,",'_[ I "OS3-'3 !

]_ue _m pwn LOSS] TO FROM FLOWOSR TRAIN A TRAIN B l cmc PUMPB_vaI'vls INSUF POS 3-13

HOUU-NOT-POB313 OSRINll5 OSR,YNII5, KOUSE-POB313 I"°''_''_'_°_"-_; (:_T:"_,) I I

•f eU OSRIN313

HOUSE---qBOI HOUSE-SBOR HOUSS-LOSP I _OW/(X_O_ FROM POB 1,E,14.15

! , I I

/ A OF PS71

_ " {;O_ll_ Olr PIPE , / PiE 5EG PS72 ! OUTSIDE SPRAY RECIRCULATiON (OSR) (OSRS) I t_ PH_ $EGPS71 I I .su_m,,mII

i I F T ]

gump Pl_ged

OSR-MDP_-2AB CLS-ACT-OP-CL$2CLS--ACT-F'A--CLS2A /_ HOUSZ-NOT-_IS SUMPPLUG _E--/:_ OUTSIDE SPRAY RECIRCULATION (OSR) (OSRd) I

(_ R8

... t__ I I I I

i+0 1 +ilE'IB iI E4BO_J ii+'

Bump Plul_ed

O._-ACI'-_-O._

(sws6)I

| SEGMENTPS67 ct,s-Ac,,'r-eA-Ct,,._meswa-iloV--oF-xm"e

I]0UlgI-]P'09:313 _ 9#$9 _ _10 _ 101Ume-llO'i'-P019313

9A_-IdOV.,-t_-IO_ _ _#S-),4OV-PG-I¢4_9 I I _ /a,, /_ 0

Z 0

C3 Z

b-- W

C-258 OUTSIDE SPRAY RE:CIRCULATION(OSR) (sws9} I _ow__ _:_'1A011"IX387 1

I I , . ! i _

OF RS SW _ AC idCC 11-11-I OPEN RS SW IdOVS I NO _ OP_:_NG FALU_ OF 4_V I ANO (:eS OON"T

SWS-IvlOV-OP-I_SS SWS-MOV-.OP-RSFP (_LS-ACT-FA-CLS2A

I-IOI,,I_--POS3'I3 HOUSZ--NOT-POS313 TEST AND MAINTENANCEON SWS MOV 103A

L__ F-- ! o ! I I i I i

>._. o

C-261 INSIDESPRAY RECiRCULATION(ISR) (SWS12) i

P_4 C-263 CotGinmentf@ilureSumpof reclrculGtlonis PiuggedtcGuses I

ontainmlquentued Sum t.

2 _ 1

I ,, I I _ I Time Window PLUGGINGOF Time W'_dow PLUGGINGOF- 1-House Event THE CONTAINMENT 2-House Event THESUCONTAINMP in TimeMENT SUMPin "Time Window2 Wh'x:iow1

WINDOW1 LPR-CCF-PG-SUMP1 WINDOW2 LPR-CCF-PG-SUMP2 Appendix C.13 Residual Heat Removal Systems

C-265 0-266 TEST AND MAINTENANCEON RHR MOV 1720A INSUFF FLOW FROM SEG PS9, DISCH TO LOOP 5 (FT WS) I

ACC-CKY-ItT-CYI4? J:_J.IR-M(-_FI"- 17208 RHR-MOV-F_ -1720B RI-R--CCF-FT -_ _HHH RHR8

i I i

I "I

/ ---! ,-j I I I ! I I I I 1

cclr4 IqlI4PI_-)AVII frsreu _ Pe,mN. Maint:enence of RHR Pump B RHRMOD I

•hr-xmv- - RHR PUMP TBL1 -- I MAINTENANCE

,hr-xmv -ft-xv24 z_ MA- 17a0A MA-RHRIB L_ HOU SB-RHR6A RHRII RHR14 WS-S

i Loss of POSs 3-13

w,_s I l J InsuffFlow ! fI_grom Dis__Schegtoment

RHI2S RHR3-S I I l I TI_oSNTAND MAmqTENANCl CHECK VALVE Loss of Power InsufFfIQ_ RHR MOV 1720 CV TsOi 130OPENFALLS at IvlC011-11--2 PSThru7, HS_B._cnht / Hdr

MA-1720A EH2 RFR4-S ACC-CKV-FT-CV'130 Centimeter 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 mm

1 2 3 4 5 Inches I.O ,5_o Lt[,,_,,LI',+_°"+_umIIlll_'''+ °°"+_IIIII_ ,,-: Ulllg IIIIINIIII1_Illll_

RHR3-S

I I C_ZCK VALVE CV SI 147. FAI_S TO OPEN.

ACC-CKV-FT-CV t47 RHR4-S Insufficient Flow from

RHR ___header

n i I I I, J

,, IHI-)Md-IoI0-W_ 1_I.4Md.-f0.-tNS 1 MR.-HIX-.f°_-(14 IIlIHlX-4J(-.[lk ,a,iZ,Mi_-)I41-PO-'_iV'Z4 , oX--I_[ll IIIIR-HIX-IX-(ll I 1 l'_'o," / i ,...._. /

II IiIt-l_0lg'.-I_r -Iv N _-I 0_-It-l_,_ RHRS-S _ a _ _ L,_] o o o o o -- _r L,__I

g-TTHH_I Appendix C.14 Service Water System (The Fault Trees for the Service Water Components are included in the Fault Trees of those systems that depend on Service Water) (See App. C.3, C.4, C.8 & C.12)

C-280 Appendix C.15 Steam Generator Recirculation and Transfer System

C-281 SG RECIRCULATION 2 OUT OF 3 (SGRT1) ALL SG RECIRCULATION FAILED (SGRT2) SG RECIRCULATION AND TRANSFER - SG A

(SGRTA) SG RECIRCULATION AND TRANSFER - SG B (SGRTB)

! I _ z,, 0 SG RECIRCULATION AND TRANSFER -- SG C

(SGRTC) II ,

Co287 C-2._ SG RECIRCULATIONSYSTEM- SG C (SaRT2C)

0 I

I I I I I

_ O1P P15UOOBD _ I_-KLC _ RT-IGIC l'T--r,l_ i _ IKT--IItC LBAdaNO DV NAINIrSNANat

_ _ BORi_Q-HT'£--PG--II|C OOitT--llIQ_--LK-mC OOItT--Oll_--]Lq -_1C

'1'00MN ltq'-l_ PAI_ TO BUN I£_NTBN._CW ' U _ I ' I FA1Lq TO '_AIT I I ' I ffGI_ --L0_CV-I_ - lt'iL'41t -_.-..IEIGA'I'-IP3A0--Pit_ --ltq'f_ 8OI_I'-PMP-IeS--0 m?IC OORT--PMP0-)/A-it?IC C-290 SG RECIRCULATION - SG B (SGRT4B)

| II"i--II I

o_ I ?. [ I

I-' ' I'-'" i -' ' ',,

I1{1! r,41.1l) IOiIIr-,IIdT-.PO,-IDDL L IKllT-lff"tl(l-11811 IIOIT -1111P-i_1-1014 i_lilril_l I / _ , , , , _L , ....1 I I I I I .__ t '-"" II-_lAI O011T/A]Lqf- SOil III Hlff-J0iO1r--l_-IOeO IGIrI'-AOt'-LK-INO0 IQtl'-AO?--O00-1I)OO ± , , , , _ ,.,,

m-U_8 0 0 0 CO_Wr/k.rW -gGIf z_I_tll _llirl__A(W,_lr O_ IOIID gorjr-.IkOV-ii [-io{m IIOIIT-II_V -.OC-- IO00 C-292 Appendix C.16 Steam Generator Secondary Relief System

C-293 SECONDARYSIDE HEAT REMOVALFOR REFLUX COOLINGWINDOW1 (SSHWl)

I I ' I I ' I ! ' I

I ..... I I I I I,

I ,I I I I I I I I I-

I ri'_uTOmu_e_r.rseAm'_n_ I faI--k"b" I--" I-- I rleTMu_K._P°rs°,m, I I ! I fa__'" I rlos_u__°urs°J_

S_"iRIB S_"IWIk Sg,'INILk _D'IRm SSI,,_I_

_lt.-ll_lr-xlgl-la i

__-ll:llm-lllgO SECONDARY SIDE HEAT REMOVAL FOR REFLUX COOLING WINDOW 2&3 (SSHW23)

2_3 I l 1 i I 1 I, I I

'_ l'__J I"-I _ I J i'-s--i

I ! I I"_" ] I I,. !_" ] I'_1 I'_" ]

I_1 I *_ ! I ! I I o--i i__i, t 'w"l ±IIINRll ±m © ±m ,_m © ±_ © SECONDARY SIDE HEAT REMOVAL FOR REFLUX COOLING WINDOW 4 (SSHW4)

I i I t } t t t t

' 1_'_rI I'__I 1"°_"1 i----'''''-II-"_'-I o I i.., It, i , . I' i 1'4"'1 I ' J I"_'l l.__.i I'_"! i , ± ± © ± ± c._) c._) _ ± Illl_ll_'Im IIUI_3oI_-'IN JlII_-_-'IB C-297 SO A STEAM PATH FAILURE SO B STEAM PATH FAILURE

SG A PORV "[_ A TT TT .'_ _'_-

I Iw'"1_,

M_ SG B PORV FAILURE

PJul.I 5'o _ /I,W4HIIm H • _ _wl_r_ Am Imu-'_m_ mml_ '_0 i ....N •'_ i I H •"1_ I_ rAp'.rao.m I Nm_'O_ I I mS'.O. I I rJm.oaa'_'a, Ii 0 0 0 0 _ -_-

FJ.-O-.V-IO_ IfSg._OY-lCii.- lO.IJ o I o I"l-- 0 SG C PORV FAILURE

N_ LOSS OF INSTRUMENTAIR

I T''"I

l i I ! T""'I e i i 1 1 r 1 1 I" I I ' i , , , , , , ,

11,41.11-11111,-Iio _o 111_12o__1111I,..11 o_ n.o4,1.-._l_-II oIIILR_tJI _o ol'llql-Illlll o 0 0 Appendix C.17 Consequence Limiting Control System

c-305 E

C 0

0306 CLCS-B I

1 I NO SIGNALFROM CLCS TRAIN B

EB CLS-ACT-FA-Ct.S2B Appendix C.18 Reactor Protection System

C-308 FAILURE OF RPS TO SCRAM THE REACTOB

, I o I I

K Appendix C.19 Rectrculation Mode Transfer System

C-310 Recirculation Mode Transfer System Train A

I , o I !

! ...... I 11_'STRAINA FALUR(RMr DI.£ "

Rm4T-ACT0-FA-RI4TSA RI_r-C0OF-FA_ o w

| I I

C-312 Appendix C.20 Safety Injection Actuation System

U') c- O Q M 4 0 <(

0 • m 0 (1) C

C_ (f)

C-314 Safety Injection Actua-tion System Train B d_, I

E,I.W E1B SIS-ACT-FA-SIS8 APPENDIX D

Statistical Analysis of Time to Mid-Loop and Duration of Mid-Loop

D-1 APPENDIX D

STATISTICAL ANALYSIS OF TIME TO MID-LOOP AND DURATION OF MID.LOOP

PAGE

Appendix D.I Bayes/an Analysisof Tune to Mid-Loop ...... D-3

AppendixD.2 BayesianAnalysisof Durationof Mid-Loop ...... D-12

AppendixD.3 Recovery Curvesfor Accident InitiatingEvents ...... D-16

AppendixD.4 Assessmentof UncertaintyParametersof Basic Events ...... D-23

D-2 APPENDIX D.1

Bayesian Analysis of Time to Mid-Loop

D-3 D.I Bayesian Analysis of Thne to Mid-loop

This Appendix summarlxas st Bayesian analysis d three quantities:

• the time-to-mid.loop (POS 6) • the time.to-soared mid-loop (POS 10) • the duration of mid.loop (POS 6 and POS 10)

Note that in all three cases, the random variable is assumed to be st shifted Iognormal, i_e, the time to mid.loop T = t + t', where t' is the lower bound for T and t is lognormai distributed with density function given by (using the Tso., formulation)

1 In f(tJTs°'u") _2"_utexp

The following lists the time shift for each of the POSs:

Jl llll Illi I II PGS t' I ii

Poe 6 - hireling o_ -_, i 2 daysii i ii POS 6 - Drained Maintmumee 13 hours il Hill l_u 10. bplaNhll _talle 2068 Iin_xm i I I , ,,,,, i

These times were eslAmated using the opendlq procedures and past plant experience. The rest of this appendix deals with the distibution of t and shows the results for 3". Reference [1] is st ILaPerthstt documents the method used in the analysis in more detail.

TIMK TO MID-IA)OP DISTRIBE/IK)N

Data for the time to mid-loop consist of population data and Surry-spedfic data. _ the former, it Is assumed that the time-to-mid-loop for each plant is Iognormally distributed with parameters p and u, and that thase parametersvaryfrom plantto plant. The population data (generally) eonslst of experts' "best estimates" and "lower bouds" for each phutt.specific distribution. (In some cases, only an"best estimate" is providsd.) The "bestestimates" are interpreted as medians; the "lower bounds" are interpreted Sth per_mtiles.

1) lqk'nt gmlp_ PqJdatlm V_ _ fbr Tm ud u

F__s:

Assume Tso and u are independent, iopormally distributed parameters. Then (for the ith plant):

D-4 L(T_),, 2_°s1oT_ _ _ ,j

L(,,,)- 2¢._O_o.O_ Oo

Prior Dirlribu_e, (IS Stage):

O_oO,

Data: Tabh D.I.I

i i i

Me_ 4.1438 0.49'/1 -0.5776 0.7729 ii i i i H ii i

s,(o) = ]'ft(,,l_,.)-,(_.)d_,.

(Noee tlmt tlke f(.) are _ m p and G, ratl_r elum Tm and u.) m(Ts,,e-)mt'rm)g,(o)

The analysis distbtguhdtes between refueling outages and drained maintenance outages.

L_l_V_lPed_ For the Ith data point from Surry,

D-5 1 1 l"(t/Tz) L(:r,l:r_,o),.¢2...;._._ t_. o 2t

Prior _" n (2_I Stq,e):

f.(T...,.) = s.(l"..p)

h_,_,g Omw

H k pm _nm(k) I R 02/83 184 i n o6/83 293 R 02/84 88.1 R 09/84 161 J i| 11 (unit2) _/20/8S 239 112(Unit I) 0S/I0/86 91.9 R3 (Unit 2) 10/04/86 100 i R4 CtJnlt1) 04/09R8 174 RS (Unit 1) 09/14/88 31S n6 (Unit2) o911o_ 203 ,,

_. Ox._m_ft_ (m _ 1).1.1): Meulgl $th78.? S0thlSS

D-6 Ma_Zmam_ Oaz_:

i _t Dm _ (b) I I D 10/82 121 ill i i i i ii i i D 12/82 44.0 D O6/83 24,O i D 09/83 106 i ii D 12/83 42.1 i i i D 12_ 27.9 i ii llllll i i |ll| ii i D 03184 37.6 i D 10/84 82,5

DI (Umdt1) 04/29/8S 94,8 i ii i fill llJl i i i lliH Hll i ill llll i D2 (Unit 1) oe/(O/Ss lOS 1)3 (Unit 2) 10/29/tJS 61._ i 1)4 (Unlit1) 01/24_ 74,8 i DS (Unit 2) 06/17/86 S3.3 i i D6 (Unit 1) 12/11/86 276 1)7 (Unit 2) 12/09/86 1282 i D8 (Unit 1) 0s/16/87 74.9 D9 (Unit I) 06/23/87 S3.7 ii ii D10 (Unit 2) 12/09/87 184 DI1 (Unit 2) 05/16/88 291 ii i i DI2 (Unit 2) 10/12/89 403

_m _.stim (B¢ FigureD.I.1): MmmlgO $th27 SOthlOS 9_h61S

TO SlgCOIqDMID-LOOP (POS 10) msl'glBtrglON

D-7 Only Surry-_ data O pohats) an svdable. (Nots that tkls POS is relevant only for ref_lhq outsps.) A sUsfsbttorwsrd _ spproscb is used.

IJInmmed_

l'dor INskn,da

I_,sdml DMri, dha

r(t) = J'f_(tJjJ_)vt(p,u)d,_

,_--T/,

I I I I nl (Usdt2) _ 1909 R3 CUnlt2) 10/04/96 9_ (_slt 2) 0_/10/SS 290S

_ B_dlmlm O._ for t i i l iHi u ml p w i 7.7276 _73' i i

ke.m.d uksa,nm_ ckm.amm4_ _ Y Mlmm_19 Slii1942

9Slk_71

D-8 930902 bestf(T)s ch

PDFs forTime to MidloopDistributionsGiven Populationand Surry Data

0.01 given populationdata

givenpop. and Surrydata (20 drainedmaint,outagesonly) 0.008 given pop. and Sur_ data (all 30 outages 1982-89)

. _ givenpop. and Surrydata (10 refuelingoutagesonly)

0.006 't t

i 0.004

0 002

I • I ,i, q,

I ! 0 _ _ .. 0 100 200, . 300 400 500 600 700 800 _;900 Time to Midloop[hrs] D.I-I _ d _ to Mkl.ioop T_bts D.I-1 Pop_tloa Data os Time to Mid.loop Bsssd on _ms to GQ,rtc L,tt,r 87-12 illli ill ill i ilnniliii i iii I ll]nllnlinl l i , , ,,,,, ,, , , , , , , I IIIlill I1, MaSt # M_m(br) Stb(hr) 9$tb(hr) I I I I I I [ I ] 1 84.6 74.8 ii iH i i i i i i i i i isl 2 66 Ill i irnn inli i I III RIN I I I 3 ] I I II III II II ilnlllUi

i i i i i i i Irll_i 5 I _ nl iii ii i ill ii i ii i 6 56 II I I I IIII II

f l nnl I i ii ii lUlflUillli I i 8 77 II I Ipl I IIII IIIII II P III

I _ II I I I [11 II Inil I 10 33 i i i i i i llll 11 30 i i i |ill i i i llluH llJIH 12 44.5 I II I ii 13 48 i ill i 14 121 i iiii ii iii i 15 72 i i 16 72 I I III I I I 17 52 I i i 18 96 II i ii i ii iiii i i 19 48 i 20 99 528 j i 21 65 II iiii i i 22 147 76 ,,==, 23 50 i i 24 72 I I iiiiii 25 120 II i illl 26 23 II 27 14 i ii

D-10 Table D.I-I (_tlnmNl)

i ii illl Illlll I I I I , ,I ,,11 IIIIII I I IIIIII • Plant # MedianCar) Slk(br) 9Stb(br) I i i il IIIll Ill i II i i i i i i I I " i 28 48 24 i i i ill H ill ii 29 53 I IIIIIIII II I III III 30 42 II ]1 Ill II [11 [1111 I[1111I I II III 31 34 -- I i1_ [I III II / IIII 32 120 48 ]_ I II III Jill III] i ii 33 106 I . IIII II II I i I I|111 I 34 144 35 50 144 IL I III I IIIII I I [ 36 51

I I I I III I I [ II I iiii 37 48 IIII I IIIIII IIIIIIIII I ] II 38 116 III1[ iii i i i i iiiiiiui i ii II 39 56 16 I I II 40 168 72 |llll i i i i i ii 41 44 i i i iH i i i i i i 42 30.5 II IIIIIIII I I I 43 54 37 i i i 44 108 i i i i i 45 69 il i i lili Bill I i ii i iililiii 46 161 i

The populationdata for the time to mid-loopweremllected from responsesto NRC InformationNotice 87-12. The plantswere asked to provide informationon their time to midloop experiences. The plants provided "typk_" minimumand maximumvalues of the time to mid-loop. In this analysis,these have been interpretedu median, 5th percentile,and 95th percentilevalues.

D-11 Appendix D.2

Bayesian Analysis of Duration of Mid-Loop

D-12 D.2 _lan Analysis of Mid-loop Duration

MID-_P DURATION DIgFRIBUTION

OulySurty-speciflcdataareavailable.Analysesaredoneforrefueling outagesanddrainedmaintenance In the cam of the former,POS 6 and POS 10 mid-loopsare distinguished. A straishfforward Bayesian apprm_ is used.

IJlulfheed hndlm

Pdsr INs_lhuf_ , 1 5ro(jS,ujoo-- 0

Kspsdsdllklbu_t

r(t) - (tJ p_ryu%(p_v)dpdev _ omp. poe 6 Data: t i till lhuat The(h) II III I I IIIIII I II rill(Unit9,9 _, L_ lllllll ii iiiiii iiii i il2 (Unit 1) 0S/10/86 61.7 ii i i R3 (Unit 2) 10/(0/86 41.6 i i ,ll R4 (Unit 1) 04/09/n _._ i]1 RS (Unit 1) 09/14/88 100 i i u R6 (Unit 2) 09/10/88 681 illl l It

l Nil I NlUl p • liil il Mean 4.714 1.146 i i i Sth 3.923 0.652

D-13 i i i S0th 4.710 1.046 ill| 9Stla $.SD 1.999 I I1'11 _mlation 0.028 i_ r'i. Zspedmtnb_tbeam Ckmdm4Jam (Sam)

Mesa2_ .qlk14.2 S0tbll2. gStb876

_ OUTAC_POS 10 Note tli the expectod distribution ires a reverse J-shaped pdf. Auziliav3 MathCAD computations confirm that this b approprlete, even thouKh the Joint pdf is p and o is unlmodsl and well-behaved.

i i _emt nmo _me 0,) I I I IIIII al aJ_ 2) 1 __ ,,, 14! g3 (Unit 2) 10/(W86 S2_

It6 (Ullt 2) m/lo_ 4S2

ii p • IIII Mean $.021 1.S93 , llll ill• 3.413 0.619 i H S0tk &011 1.2'/3 H 95th 6.6S6 3.798 ! I I Comdation 0.039 i i i i i i

Mmm4_ _h6.3 SOthIS1 9Sthg.q_

D-14 D NKD MAINT/_qANCK OUTAGL POS 6

i i swat nine (b) II II II I I D1 (Unit 1) 04/29/8S lS3 D2 (Omit 1) 08/06/8S 42J i I)3 (Unit 2) 10/29/8S 79.4 i IM (Unit 1) 01/24/86 182

D$ (Unit 2) 06/17/86 191 I)6 (Unit 1) 12/11/86 1072 i i i i i)7 (Unit 2) 12/09/86 59.9

D8 (Unit 1) 05/16/87 144 1)9 (Unit 1) 06/23/87 10.9 i llll i D10 (Unit 2) 12/09/87 28_3 DU (Unit 2) 0S/16/88 206 i i t i D12 (Unit 2) 10/12/89 266 i

II I t I p w I II Mean 4.692 1.282

Stk 4.065 0.892 i 50th 4.691 1.232

9Stb $.318 1.849

Correlation 0.001

Expeebd Dblribmflom _ Me=n_$ $4b11.9 .q0thl09 9Sth9S8

D-15 Appendix D.3 Recovery Curves for Accident Initiating Events

D-16 ! Appendix D.3 Recovery Curves for Accident Initiating Events

The data that was used in estimating the frequency of initiating events show that the initiating event can be terminated by eliminating their causes. In the event trees, recovery of the following initiating events was modeled as top events in the event trees:

Loss of Offsite Power Loss of 4 KV Bus Loss of Inadvertent Safety Injection Loss of Vital Bus Loss of Instrument Air

Recovery of offsite power is discussed in section 4.3.1. The recovery of the rest of the above initiating events was analyzed using a single stage bayesian analysis. The computer code BAYF__D written by lq. Siu was used. The recovery time is assumed to be lognormaUy distn'buted. The prior distn'bution of the parameters It and u is assumed to be inversely proportional to u. The likelihood function is the density function of the Iognonnal distn'bution. Tables D3-1 to D.3-5 list the input data and results.

D-17 Table D.3-1

Statistical Analysis of Recovery Time of 4KV Bus

I i Recovery Times of Loss Percentiles of Time of 4 KV Bus(minutes) to Recovery

1 9.00 Probability Percentile (%) (minutes) 2 1.00 5 033 3 21.00 10 0.62

, 4 1.53 15 0.96 5 19.00 20 1.34 ii i 6 1.50 25 1.80 iii 7 2.00 30 234 i 8 15.00 35 2.99 9 60.00 40 3.77 ii 10 0.17 45 4.71

11 2.00 50 5.87 12 4.00 55 7.32

13 45.00 60 9.15 14 22.00 65 11.54

15 17.00 70 14.72 75 19.13 iiiiiii ii 80 25.68 lit 85 36.08 90 55.52 95 104.87

Mean=29.09minutes

D-18 Table D_-2

Statistical Aulysls of E,_eoveryTime d CCW

i Recovery Tunes of Loss Percentiles of Time of CCW (minutes) to Recovery Probability Percentile (%) (minutes) 1 14.00 5 7.17 2 17.00 10 8.97 3 30.00 15 10.44 20 11.76 , , ,,, 13.04 30 14.30

35 15.58

40 16.90 45 18.29 50 19.76

55 21.35 60 23.10 65 25.06

70 27...30 75 29.94

80 33.20

85 37.42 90 43.54 95 54.46 Mean =26.16 I

D-19 Table D.3-3

StatisticAanla135iof, RecoveryTime of Im_wtut Safety

i i i Recovery Times of Inadvertent Percentiles of Time

Safety Injiieiiction(minutes) toi Recovery i Probability Percentile

(%)ii (minutes)i 1 5.00 5 0.86 i i ii i i 2 15.00 10 1.39 i i|l ii 3 23.00 15 1.92 ,,|,,, , , ,,, ,,, 4 9.00 20 2.48 LII[ I ii 5 2.1111 25 3.10 i i i ii 6 150.00 30 3.78 7 2.00 35 4.54

8 50.00 40 5.40 9 2.1111 45 6.39 i i i i i i 10 3.00 50 7.55 ii i i i 11 5.t111 55 8.91 ii 12 2.00 60 10.55 13 7.111 65 12.56 Jl i i, 14 5.00 70 15.10

15 19.11tl 75 18.39 -- i 16 6.00 80 22.97 ,| 85 29.67 iiiii 90 41.07 ill i i i i 95 66.35 ii i i i Me._= 18.70 i|l

D-20 Table D.3..4

Statistical Analysis of'Recovery Time of Loss of Vital Bus

RecoveryTimesofLoss PercentilesofTune of Vital Bus(minutes) to Recovery Probabil/ty Percentile (minutes) . 1 0.03 5 0.33

2 1.00 10 0.60 3 3.00 15 0.91 4 4.00 20 1.25 its 5 4.00 25 1.65 6 5.00 30 2.12 7 5.00 35 2.67

8 5.00 40 333 9 6.00 45 4.12 10 %00 50 5.08

11 13.00 55 6.26 iii 12 15.00 60 7.74 13 17.00 65 9.64 14 18.00 70 12.15

15 25.00 75 15.58

16 15.00 80 20.60 85 28.45 i i 90 42.84 95 7836 Mean=21.34

D-21 Table D.3-$

Statistical Analysis of Recovery Time of Loss of Instrument Air

ii I I i i Recovery Times of Loss PercentilesofTime of Vital Bus(mlnutes) to Recovery I II I I I III Probability Percentile (minutes) I II I i 1 10.00 5 932 I 2 20.00 10 10.68 a I II 1ll81 I I I 3 20.00 15 11.70 i I I 4 15.00 20 12.58 II I I 5 20.00 25 13.40 i i iiii I i 6 31.00 30 14.17 I I I I I I, 7 20.00 35 14.93 I III ii II 8 20.00 40 15.68 i I I I 9 20.00 45 16.45 ii I I 10 20.00 50 17.24 i 11 20.00 55 18.07 12 10.00 60 18.95 I i ii 13 10.00 65 19.91 70 20.98 , II 75 22.18 80 23.62

25.40 9O 27.84 I i 95 31.89 Mean-- 18.55 ,, ,, I i ii

D-22 APPENDIX D.4

Assessment of Uncertainty Parameters of Basic Events

D.23 A48pmdlx D.4 Assessment of Unity Parame/e of Basic Events

Frmuencv of Outases

The uncertainty of the frequency,of outages was estimated judgmental_. Table D.4-1 lists the characteristic parameters of the outage frequenc/es. In chapter 9, the point estimates of the frequencies were calculated at using the Suny plant experience. They are taken to be the mean frequencies of the outages. It was the Surry experience that the time between 2 refueling outages sometimes is as longer as 2 years. Therefore, a frequency of 0.5 per year is quite poss/ble to occur. It was taken to be the 25th percentile of the distribution of the frequency of refueling. That is, the probability that the time between 2 refueling outages is longer than 2 years is 0.25. A lognormal distribution is assumed with these parameters. In the case of drained maintenance, it is the plant experience that a unit may go to drained maintenance twice a year. Therefore, 3 times a year is taken to be an upper bound (95th percentile) of the frequency of drained maintenance.

Another paran_ter used in the qmmtiflcation is the haction(FRAC-POS10) of time that the plant goes to mid-loop operation after refueling is completed in a refueling outage. The review of plant experience found that in 3 out of 6 refueling outages the plant went to the second mid_atlon. Therefore, the mean of this fraction is taken to be 03. It is assumed to be uniformly distributed between 0 and I.

Cond/tiO0al pmbab_ty of a T'nn.• Window Given That the Initiatinl Event Occurred in a POS

In order to correctly determine the success criteria for the mitigating functions, the time window approach was developed. Given that the accident initiating event had occurred in a POS, the probability that it occurred in a given time window was determined in section 9.3 using data coHoected on time to mid-loop and duration of mid-loop. The uncertainty in the time when the accident occurs is reflected in the probabilities of the time windows. Therefore, the cond/tlonal probabilities of the time windows take on single values with no uncertainty parameter. They are listed in Table D.4-2 for the accident initiating events. For all htitiating events except over draining to mid-loop(RHR2A), the distribution of the time when the accident occurs is cakulated as the time to mid-loop plus the duration of mid-loop times an uniform distribution. For RHR2A, the time when the initiating event occurs is the same as the time to mid-kp.

Maintenance UnavailabiHW

Two types of maintenance events are used, the ones adopted from NUREG-1150 study, and those estimated using plant outage data. For those maintenance evetns of NUREG-1150, the uncertainty data of NUREG-1150 was used. Chapter 9 documents the maintenance data collected from the plant_ The uncertainty of these maintenance nnavailaldlities was derived by judgement using the following rules:

I.H the mean issmall,a lognormaldistributionwithsome EF isassumed.

2. H the mean is dose to 1, a uniform distn'vution is assumed. If the mean is larger than 0.5, the upper bound is set to 1. If the mean is lower than 0.5, the upper bound is set to twice the mean.

D-24 TableD.4.1

Chsrs_rlstic Valuesot Freq_scy ot Ostslps(per year)

, L , I II Mean Error 5% 50% 95% p o Factor L ,,,,, ,,,,, Refueling 0.6 1.46 0.40 0.584 0.854 -0.5375 0.2309 i i i ii i i i iiii i , H, .,, , , , , ii ii i i i H Drained 1.2 3.22 0.29 0.932 3.00 -0.07 0.7104 Maintenance i i ii i i i i

D-25 Tsble D_-2 Cosdltio_ Prob,billty tkst am IE Occurs Innthe Tim, Wisdo_ Given It Omsrr_ Is tim POS

uiiinllii ill in i i IIINI i iii nlnlllnllnnnii i i llnlllnl iiiiiiiii i n WINDOW I WINDOW 2 WINDOW 3 WINDOW 4 I iinl I'll Illll II I I Ill bl II I I I I Ill Deflnfllon < = 75 hours > 75 hours and >,, 240 hours and > 32days <= 240 hours <= 32 days I I i iiii ii I I i I I lllIll I II i All I_ exceptRHR2A ...... II II l I I I l 11HI l I nll l I IllI HI D6 1.17E-01 0.436 0375 7.20E-02

i ii] I I I I I II I I[ I] I IIll II I iiii nln fill R6 1.7E4)2 0.543 0.41 3.4E-02

I I I II IIIIIIIl I II lllll l II I I I I R10 0.0 0.0 0.016 9.84E-01

• I ii I llll|lllll I I I rill

, ,,,, ,, I IHlll Iml I ill D6 031 0.454 0.21 2.6E-02 I I I iii]11111iiii II1[111IIIIIIII I II IIIII I I is ill sl R6 5.8_MI2 0.7 0.24 1.48F_3 iH iH i i H i i i i i i i i , ,,,,,,,,ml RI0 0.0 0.0 2.2E-02 0.98 I II I I I i I I I , i llllll i i ii [

D-26 Kiyoharu Abe Ephraim Asculai Dept. of Reactor Safety Research Division of Nuclear Safety Nuclear Safety Research Center Wagramestrasse, 5 Tokai Research Establishment P.O. _ox 100 JAERI A- 1400 Wien Tokai -mura, Naga- gun AUSTRIA Ibaraki -ken, JAPAN Vladimar Asmolov Head, Nuclear Safety Department Sarbes Acharya I.V. Kurchatov Institute Department o_" Energy of Atomic Enegry NS- I/FORS Moscow, 123181 Washington, DC 20585 RUSSIA

Dr. Ulvi Adalioglu J. de Assuncao Cekmece Nukleer Arastraima ve Cabinete de Proteccao Egitim Merekezi Seguranca Nuclear P.K. 1 Ministerio da Indusstria Havaalanl/ISTANBUL Ave. de Republica 45-6 TURKEY 1000 Lisbon PORTUGAL

Dr. Eng. Kiyoto A/zawa Senior Engineer H.P. Balfanz, Head Reactor Eng. Dev. Department Institute of Probabilistic PNC Safety Analysis 9-13, Chome, Akasaka TUV Nord MAnato-K, Tokyo Grosse Bahnstrasse 31 JAPAN D-22525 Hamburg 54 GERMANY

Harry Alter Manager Applied Tech Pat Baranowsky Nuclear Systems Tech USNRC-AEOD/TPAB NE-46 MS : T-4A9 US DOE Washington, DC 20585 Robert A. Bari, Deputy Chairman Dept of Nuclear Energy R.M. Andrews Bldg 197C Nuclear Installations Insp. Brookhaven National Laboratory St. Peters House Upton, NY 11973 Balliol Raod, Bootle Merseyside L20 31Z UNITED KINGDOM Librarian Technical Information Section Battelle Pacific Northwest Lab George Apostolakis P.O. Box 999 UCLA Richland, WA 99352 Boelter Hall, Room 5532 Los Angeles, CA 90024-1597 Dr. John Baum Dept of Nuclear Energy Director of Reactor Engineering Radiological Sciences Div Argonne National Laboratory Bldg 703 M 9700 S Case Ave Brookhaven National Laboratory Bldg 208 Upton, NY 11973 Argonne, IL 60439

Dist- 1 Eric Beckjord Dennis Bley USNRC- RES/DO Buttonwood Consulting MS: T-10FI2 17291 Buttonwood St. Fountain Valley, CA 92708

Robert Bernero USNRC- NMSS/DO Roger Blond MS : T-8A23 Booz-Allen & Hamilton 4330 East West Highway Bethesda, MD 20814 Andrea Besi Institute for Systems Engineering and Informatics M.P. Bohn CEC Joint Research Centre Division 6449 CP N 1 Sandia National Laboratories 1-21020 Ispra (Varese) Albuquerque, NM 87185 ITALY

Dr. Mario Bonaca John Bicke] Manager, Reactor Engineering Idaho National Engineering Lab. Northeast Utilities EG&G MS: 3850 P.O. Box 270 P.O. Box 1625 Hartford, Conn. 06141 Idaho Falls, ID 83415

Robert B. Borsum Vicki Bier Nuclear Power Division Dept. of Industrial Engineering B & W Nuclear Tech University of Wisconsin-Madison 1700 Rockville Pike 1513 University Avenue, Room 389 Suite 525 Wisconsin, WI 53706 Rockville, MD 20852

Scott Bigelow Stephen Boult S- CUBED Electrowatt Engineering Services 2501 Yale SE, Suite 300 (UK) Ltd. Albuquerque, •NM 87106 Grandford House 16 Carfax, Horsham West. Sussex RHI2 IUP Prof. Dr. Dr.-Ing. E. H. Adolf ENGLAND Birkhofer Gesellschaft f_r Anlagen und Reaktorsicherheit (GRS) mbH Gary Boyd ForschungsgelAnde Safety & Reliability Optimization D- 8046 Garching Services Federal Republic of Germany 9724 Kingston Pike, Suite 102 Knoxville, TN 37922

David Black American Electric Power Brookhaven National LaboraUory (2) 1 Riverside Plaza Attn: Lev Neymotin Columbus, OH 43215 Arthur Tingle Building 130 Upton, NY 11973 Harold Blackman Idaho Nanional Engineering Lab. EG&G MS: 3850 David M. Brown P.O. Box 1625 Paul C. Rizzo Associates, Inc. Idaho Falls, ID 83415-3850 300 Oxford Drive Monroeville, PA 15146-2347

Dist-2 Tom D. Brown A.L. Camp Sandia National Laboratories Division 6412 Dept. 6413 MS: 0748 P.O. Box 5800 Sandia National Laboratories Albuquerque, NM 87185 Albuquerque, NM 87185-0748

Robert J. Budnitz John Forbes Campbel Future Resources Associates, Inc. HM Superintending Inspector 2039 Shattuck Avenue, Suite 402 Health & Safety Executive Berkeley, CA 94704 St. Peter's House Balliol Road Bootle L20 31Z Gary Burdick UNITED KINGDOM USNRC- RES/SAIB MS : T- 10FI3 Leone1 Canelas • New University of Lisbon Arthur Buslik Quinta de Torre USNRC-RES/PRAB 2825 Monte de Caparica MS : T- 9F31 PORTUGAL

Edward Butcher Harold Careway USNRC-NRR/SPSB General Electric Co., M/C 754 MS: O-10E4 175 Curtner Ave. San Jose, CA 95129

Technical Library B&W Nuclear Service Co D.D. Carlson P. O. Box 10935 Division 6411 Lynchburg, VA 24506 Sandia National Laboratories Albuquerque, NM 87185

Stefaan Caeymaex Safety & Systems Section Jose E. De Carlos Nuclear Generation Dept. CSN International Coordinator TRACTEBEL Consejo de Seguridad Nuclear Avenue Ariane 7 Calle Justo Dorado 11 B-1200 Bruxelles 28040 Madrid BELGIUM SPAIN

Leonard Callan, Administrator Annick Carnino U.S. Nuclear Regulatory Commission International Atomic Energy Agency Harris Tower and Pavilion Wagramerstrasse 5, P.O. Box i00 611 Ryan Plaza Drive, Suite 400 A-1400 Vienna Arlington, TX 76011-8064 AUSTRIA

J. Calvo S. Chakraborty Division of PSA & Human Factors Swiss Federal Nuclear Safety Consejo de Seguridad Nuclear Inspectorate Calle Justo Dorado, 11 Hauptabteilung f_r die Sicherheit 28040 Madrid der Kernanlagen SPAIN CH-5232 Villigen-HSK SWITZERLAND

Erulappa Chelliah US_RC-RES/PRAB MS: T-gF31

Dist-3 Mike Cheok S. Daggupaty NUS Environment Canada 910 Clopper Road 4905 Dufferin Street Gaithersburg, MD 20878 Downsview Ontario, M3H ST4 CANADA Nilesh Chokshi USNRC- RES/SSEB MS: T-10L1 Louise Dahlerup Inspectorate of Nuclear Inst. Danish Civil Defense & T. L. Chu Emergency Planning Agency Brookhaven National Laboratory 16, Datavej Department of Nuclear Energy DK-3460 Birkerod Bldg. 130 DENMARK Upton, NY 11973

John Darby Peter Cooper SEA, Inc. SRD/AEA Technology 6100 Uptown Blvd. NE Wigshaw Lane Albuquerque, NM 87110 Culcheth Cheshire WA3 4NE England Gerald Davidson Fauske and Associates, Inc. 16 W 070 West 83rd Street Susan E. Cooper Burr Ridge, IL 60521 Science Applications Int' 1. Corp. 11251 Roger Bacon Drive Reston, VA 22090 Peter Davis PRD Consulting P.O. Box 2046 Michael Corradini Sheridan, WY 82801 University of Wisconsin 1500 Johnson Drive Madison, WI 53706 P. De Gelder Secretary, BELGIAN NUCLEAR SOCIETY (BNS) E.R. Corran A V Nuclear ANSTO Reasearch Establishment Avenue du Roi 157 Lucas Heights Reserch Labs. B-1060 Brussels Private Mail Bag 1 BELGIUM Manai, NSW 2234 AUSTRALIA Lennart Devell Studsvik Nuclear Massimo Coszone Studsvik Energiteknik AB A.N.P.A. S-611 82 Nykoping Via V. Brancati, 48 SWEDEN 1-00144 Rome ITALY J. Devooght Service de la Metrologie Nucl George Crane University Libre de Bruxelles 1570 E. Hobble Creek Dr. Faculte des Sciees Appliqu. Springville, Utah 84663 50 Avenue F-D Roosevelt Bruxelles 5 BELGIUM Mark Cunningham USNRC -RE S /PRAB MS: T-gF31

Dist-4 G. Diederick John Flack CoaBonwealth Edison Co. USNRC-RES/SAIB LaSalle County Station MS: T-10F13 RR1, Box 220 2601 North 21st Rd. Marsielles, IL 61341 Karl Fleming Pickard, Lows & Garrick 2260 University Drive Chuck Dobbs Newport Beach, CA 92660 Idaho National Engineering Lab. EG&G MS : 3840 P.O. Box 1625 Terry Foppe Idaho Falls, ID 83415 Safety Analysis Engineering Rocky Flats Plant Energy Systems Group Mary Drouin Rockwell International Corp USNRC-RES\SAIB P.O. Box 464 MS: T-10F13 Golden, CO 80401

Duke Power Co. (2) R H. Gauger Attn: Duncan Brewer Manager-Reliability Engr Steve Deskevich A/E Div 422 South Church Street Holmes & Narver Inc. Charlotte, NC 28242 R Roanne Circle Irvine, CA 92714

Bill Eakin Northeast Utilities Robert Gobel Box 270 Clark University Hartford, CT 06141 Center for Technology, Envirca_ment and Development 950 main St. Stewart D. Ebneter Worcester, MA 01610-1477 USNRC I01 Marietta St., Suite 2900 Atlanta, GA 30323- 0199 Paul Govaerts Studiecentrum voor Kernenergie (SCK/C_) Adel A. Ei-Bassioni Boeretang, 200 USI_RC- NRR/PRAB B-2400 1601 NS : O- 10E4 BELGIOM

ENEA/DISP (2) Hr. Gubler Attn: Alvaro Valeri International Atomic Energy Agency Alfredo Bottino RENS/SAD B0842 Via Vitaliano Brancati, 48 Wagramerstrasse 5, P.O. Box i00 00144 Roma EUR A-1140 Vienna ITALY AUSTRIA

Walter P. Engel Paul M. Haas, President PRAG _GR Analysis & Reg Matter Concord Associates, Inc. NE-60 725 Pellissippi Parkway CRYCITY Suite I01, Box 6 US DOE Knoxville, TN 37933 Washington, DC 20585

Dist-5 F. T. Harper Steven Hedge Divisio_ 6413 Oak R/dee National Laboratories MS: 0748 P.O. Box Y Sandia National Laboratories Oak Ridge, TN 37831 Albuquer_ue, NM 87185-0748

Gary Holahan Dr. U. HauprJnannB USNRC-AEOD/OSP Gesellschaft FQr Anlagen und MS : T-4A9 Reaktoreicherheit (GRS) mgH Schwertnergasse I D-5000 KSln I N.J. Holloway GERMANY A72.1 Attic Weapons Establishment Ademaston Sharif Heger Reading RG7 4PR Chemical and Nuclear UNITED KXIK]DGU Engineering Department Farris Engineering, Roan 209 Albuquerque, NM 87131 Grill Holmes Westinghouse Electric Co. Energy Center East Jo_ C. Helton Bldg. 371 Dept. of Nathe_atics P.O. Box 355 Arisona State University Pittmbur_, PA 15230 Tempa, AZ 85287

William Hopkins Dr. P. M. Harttrlch Bechtel Power Corporation Gesellschaft f_r Anlagen und 15740 Shady Grove Road Rea_torsic4m_it (GR8) ndi Gai_/_ersburg, _)30877 Schwertnergasse I 5000 r,51n I GEI_W_NY Dean Houston USRRC -ACRe MS: P-315 Dr. D.J. Higso_ Radiolc_ical Safety Bureau Australian Nuclear Science & Der-Yu Hsia Technology Organisation Institute of Nuclear EnersY Research P.O. Box 153 lamg-Tan 325 Roeeberry, NSW 2018 TAIMAN AUSTRALIA

Alej andre Huerta-Bahena Dr. Mitsumasa Hirano National Ccew/ssion on Nuclear Deputy General Manager Safety and Safeguards (CNSNS) Institute of Nuclear Safety Insurgentes Sur N. 1776 NUPEC C.P. 04230 Mexico, D. F. 3 -6 -2, Torano_z_n, M_L_tc -ku MEXICO Tokyo 1o8 JAPAW Peter Humphreys US Atcmic Energy Authority Dr. S. Hirechberg Wigehaw Lane, Culcheth paul Scherrer Institute warrins_c_, Cheshire Vurenlingen and Villigen UNITED KIRGDON, WA3 4NE CH-5232 Villigen PSI SWITZE_

i DiSt-6 i W. Huntington Dr. H. Kalfsbeek Ccenonwealth Edison Co. DG/XII/D/1 LaSalle County Station Comnission of the European RRI, Box 220 Coununities 2601 North 21st Rd. Rue de la Loi, 200 Marsielles, IL 61341 B-1049 Brussels BELGIUM

J.S. Hyslop USNRCo RES/PRAB Yoshio Kano MS: T-gF31 General Mngr. & St. Engineer Systems Analysis Section O-arai Engineer. Centr, PNC Idaho National Engineering Lab. (2) Higashi-Ibaraki-gun Attn: Doug Brownson Ibaraki-Ken, 133-13 Darrel Knudson JAPAN EG&G MS: 3840 P.O. Box 1625 Idaho Falls, ID 83415 William Kastenberg UCLA Boelter Hall, Room 5532 Idaho National Engineering Lab. (2) Los Angeles, CA 90024 Attn: Art Rood Mike Abbott EG&G MS: 2110 Barry Kaufer P.O. Box 1625 OECD/NEA Idaho Falls, ID 83415 "Le Seine St. Germain. 12 Boulevard des Iles 92130 Issy-les-Moulineaux Hanspeter Isaak FRANCE Abteilung Strahlenschutz Hauptabteilung fQr die Sicherheit der Kernanlagen (HSK) Paul Kayser CH-5303 Wurenlingen Division de la Radioprotection SWITZERLAND Avenue des Archiducs, 1 L- 1135 Luxembourg-Belair LUXEMBOURG Brian Ires UNC Nuclear Industries P. O. Box 490 Ken Keith Richland, WA 99352 TVA W 20 D 201 400 West Surmnit Hill Kamiar Jamili Knoxville, TN 37092 DP- 62/I_i'N Department of Energy Washington, D.C. 20585 G. Neale Kelly Comnission of the European Comnunities Robert Jones Joint Research Centre USNRC-NRR/DSSA Rue de la Loi 200 MS : 0-8E1 B-1049 Brussels BELGIUM

Edward Jordan USNRC-AEOD/DO John Kelly MS : T-4D18 Sandia National Laboratories P. O. Box 5800 MS 0742 Albuquerque, NM 87185

Dist-7 Knolls At_nic Power Laboratory (2) Dr. J.M. Lanore Attn: Ken McDonough CEA/IPSN/DAS D_ninic Sciaudone Centre d'Etudes Nucl6aires de Box 1072 Fontenay- aux- Roses Schenectady, NY 12301 B.P. n ° 6 92265 Fontenay-aux-Roses CEDEX FRANCE Dr. K. Koberlein Gesellschaft for Reaktorsicherheit mbH Jose A. Lantaron Forschungsgel&nde Consejo de Seguridad Nuclear D-8046 Garching Sub. Analisis y Evaluaciones GERMANY Calle Justo Dorado, 11 28040 Madrid SPAIN Alan Kolaczkowski Science Applications International Corporation Josette Larchler-Boulanger 2109 Air Park Rd. S.E. Electricte de France Albuquerque, NM 87106 Direction des Etudes Et Recherches 30, Rue de Conde 75006 Paris Jim Kolanowski FRANCE Camnonwealth Edison Co. 35 1st National West Chicago, IL 60690 H. Larsen Head of Department Riso National Laboratory John G. Kollas P.O. Box 49 Institute of Nuclear Technology and DK-4000 Roskilde Radiation Protection DENMARK N.R.C.P.S. "Demokritos" P.O. Box 60228 GR- 153 10 Aghia Paraskevi Lawrence Livermore Nat' 1 Lab. (4) Attiki Attn: George Greenly GREECE Marvin Dickerson Rolf Lange Sandra Brereton S. Kondo Livermore, CA 94550 Department of Nuclear Engineering Facility of Engineering University of Tokyo Shengdar Lee 3 -1, Hongo 7, Bunkyo-ku Yankee Atanic Electric Company Tokyo 580 Main St. JAPAN Boston, MA 17407

D. Lamy B.T.F. Liwaang CEN/SCK Dept. of Plant Safety Assessment Dept. Scientific Irradiation Swedish Nuclear Power Inspec. Experiment & Study BR2 P.O. Box 27106 Boeretang, 200 S-i0252 Stockholm B-2400 14ol SWEDEN BELGIUM

Peter Lohnberg Expresswork International, Inc. 1740 Technology Drive San Jose, CA 95110

Dist-8 Steven M. Long Hideo Matsuzuru USNRC-NRR/SPSB Tokai Research Establishment MS : O- 10E4 Tokai -mur Maka- _jn Ibaraki -ken, 319-11 D. Eugenio Gi.1 Lopez JAPAN Consejo de Seguridad Nuclear Calle Justo Dorado, 11 28040 Madrid Jim Mayberry SPAIN Ebasco Services 60 Chubb Ave. Lyndhurst, NJ 07071 Los Alamos National Laboratory (2) Attn: Kent Sasser N-6, K-557 Andrew S. McClymont Los Alamos, NM 87545 IT-Delian Corporation 1340 Saratoga-Sunnyvale Rd. Suite 206 Christiana H. Lui San Jose, CA 95129 USNRC- RES/PRAB MS: T-gF31 Michael McKay Los Alamos National Laboratory John Luke A- 1, MS F600 Services Florida Power & Light P.O. Box 1663 P.O. Box 14000 Los Alamos, NM87545 Juno Beach, FL 33408

Zen Mendoza Daniel Manesse SAIC ISPN 5150 E1 Camino Real Boite Postale n ° 6 Suite C3 1 92265 Fontenay-aux-Roses CEDEX Los Altos, CA 94022 FRANCE

Dr. J. Mertens Fred Mann Division of Risk Analysis & Westinghouse Hanford Co. Reactor Technology WIA-53 Institute of Safety Research P.O. Box 1970 Research Centre Julich (KFA) Richland, HA 99352 D-52425 Julich GERMANY

Nadia Soido Falcao Martins Comissao Nacional de Energia Nuclear Jim Meyer R General Severianao 90 S/408-1 Scientech Rio de Janeiro 11821 Parklawn Dr. BRAZIL Suite 100 Rockville, MD 20852

Harry F. Martz Analysis and Assessment Division Joe Minarick Los Alamos National Laboratory Science Applications Int' 1 Corp. Los Alamos, NM 87545 301 Laboratory Road P.O. Box 2501 Oak Ridge, TN 37830 Herbert Massin Commonwealth Edison Co. 3_ Ist National West Chicago, IL 60690

Disto9 Jose I. Calvo Molins, Head Ken O'Brien Division of P.S.A. and Human Factors University of Wisconsin Consejo de Seguridad Nuclear Nuclear Engineering Dept. Calle Justo Dorado, 11 153 Engineering Research Blvd. 28040 Madrid Madison, WI 53706 SPAIN \ Theresa Oh Ken Muramatsu INEL Tech Library Risk Analysis Laboratory EG&G MS: 2300 Japan Atomic Energy Research P.O. Box 1625 Institute Idaho Falls, ID 83415-2300 Tokai -mura, Naka- gun Ibaraki-ken, 319-11, Tokyo JAPAN N.R. Ortiz, Director Nuclear Energy Technology Division 6400 Joseph A. Murphy Sandia National Laboratories Division of Safety Issue Resolution Albuquerque, NM 87185 U.S. Nuclear Regulatory Coanission MS: T-10ES0 Washington, DC 20555 Robert Ostmeyer U.S. Department of Energy Rocky Flats Area Office Kenneth G. Murphy, Jr. • P.O. Box 928 US Department of Energy Golden, CO 80402 19901 Germantown Rd. Germantown, MD 20545 Robert Palla USNRC -NRR/SPSB Shankaran Nair MS : O- 10E4 Central Electricity Generating Board Berkeley Nuclear Laboratories Gareth Parry Berkeley NUS Corporation Gloucestrshire CL13 9PB 910 Clopper Rd. UNITED KINGDOM Gaithersburg, MD 20878

Ray Ng Vern Peterson NEI Building T886B 1776 Eye St. N EG&G Rocky Flats Suite 300 P.O. Box 464 Washington, DC 20006-2496 Golden, CO 80402

G. Niederauer G. Petrangeli Los Alamos National Laboratory ENEA Nuclear Energy ALT Disp P. O. Box 1663 Via V. Brancati, 48 MSK 575 00144 Rome Los Alamos, NM 87545 ITALY

Oak Ridge National Laboratory (2) Ing. Jose Antonio Becerra Perez Attn: Steve Fisher C_ision Nacional De Seguridad Sherrel Greene Nuclear Y Salvaguardias MS-8057 Insurgentes Sur 1806 P.O. Box 2009 01030 Mexico_ D. F. Oak Ridge, TN 37831 MEXICO

Dist- I0 William T. Pratt M. Roch Brookhaven National Laboratory Manager of Design, Nuclear Building 130 Department Upton, NY 11973 TRACTEBEL Avenue Ariane 7 B-1200 Bruxelles Urho Pulkkinen BELGIUM Technical Research Centre of Finland A.E. Rogers Laboratory of Electrical & General Electric Co Automation Engineeering 175 Curtner Ave Otakaari 7B, 02150 Espoo 15 MC-489 FINLAND San Jose, CA 95125

Blake Putney Judy Rollstin Science Applications GRAM Inc International Corporation 8500 Menual Blvd. NE 5150 E1 Camino Real, Suite C31 Albuquerque, NM 87112 Los Altos, Ca 94022

Marc Rothschild Dr. V. M. Raina Halliburton _JS Project Manager-Risk Assessment 1303 S. Central Ave. Ontario Hydro Hll G1 Suite 202 700 University Ave. Kent, WA 98032 Toronto, Ontario MSG IX6 CANADA Christopher Ryder US_C-RES/PRAS William Raisin MS: T-gF31 NEI 1726 M. St. NW Suite 904 Takashi Sato, Deputy Manager Washington, DC 20036 Nuclear Safety Engineering Section Reactor Design Engineering Dept. Nuclear Energy Group, Toshiba Corp. Ann Ramey-Smith Isogo Engineering Center USNRC-RES/PRAB 8, Shinsugita-cho, Isogo-ku, MS: T-gF31 Yokohama 235, JAPAN

Dale Rasmuson Martin Sattison USNRC-AEOD/TPAB Idaho National Engineering Lab. MS: T-4A9 P.O. Box 1625 Idaho Falls, ID 83415

John Ridgely USNRC-RES/SAIB Dr. U. Schmocker MS: T-10F13 Hauptabteilung fQr die Sicherheit der Kernanlagen CH-5232 Villigen HSK Richard Robinson (2) SWITZERLAND USNRC-RES/PRAB MS: T-gF31 A.J. Seebregts ECNNuclear Energy Westerduinweg, 3 Postbus 1 NL-1755 Petten ZG THE NETHERLANDS

DiSt-ll Dr. S. Serra Stone & Webster Engineering Corp Ente Naxionale per I'Energia Technical Information Center Electtrica (ENEL) A. Hosford via G.B. Martini 3 245 Summer Street 1-00198 Rome 245/01 ITALY Boston, MA 02210

H. Shapiro Dennis Strenge Licensing & Risk Branch Pacific Northwest Laboratory Atomic Energy of Canada Ltd. RTO/ 125 Sheridan Park Research Comm. P.O. Box 999 Mississauga, Ontario LSK IB2 Richland, WA 99352 CANADA

Technadyne Engineer. Consultants (3) Nathan O. Siu Attn: David Chanin Center for Reliability and Risk Jeffery Poster Assessment Walt Muff in Idaho National Engineering Lab. Suite A225 EG&G MS: 3850 8500 Menus1 Blvd. N P.O. Box 1625 Albu_uerque, NM 87112 Idaho Falls, ID 83415-3855

Ashok Thadani 2. Soederman USNRC- NRR/ADT ES-Konsult AB MS: O-12GIB Energy and Safety P.O. Box 3096 S-16103 Bromma _T. G. Theofanous SWEDEN University of California, S. B. Department of Chemical and Nuclear Engineering Desmond Stack Santa Barbara, CA 93106 Los Alamos National Laboratory Group Q-6, Mall Stop K556 Los Alamos, NM 87545 Catherine Thompson USBRC-RES/SAIB MS : T- 10F13 Jao Van de Steen KEMA Laboratories Utrechtseweg, 310 Soren Thykier-Nielsen Postbus 9035 Riso National Laboratory NL 800 ET Arnhem Postbox 49 THE NETHE_S DK4000 Roskile DERMARK

Eli Stern Israel AEC Licensing Div. R. Toossi P.O. Box 7061 Physical Research, Inc. Tel-Aviv 61070 25500 Hawthorn Blvd. ISRAEL Torrance, CA 90505

Dr. Egil Stokke Ennio Traine Advisory Group ENEL OECD Halden Reactor Project Via Vialiano, 48 P.O. Box 173 00144 Rome N- 1751 Halden ITALY NORWAY

Dist- 12 Ulf Tveten Envlronmental Physics Section Seppo Vuori Institutt for Energitekmikk Technical Research Centre of Finland Postboks 40 Nuclear Engineering Laboratory N- 2007 Kjeller Lonnrotinkatu 37 NORWAY P.O. Box 169 Sf-00181 Helsinki 18 FINLAND US Department of Energy Energy Library Room G 034/GTN Dr. Ian B. Wall AD- 622.1 81 Irving Avenue Washington, DC 20585 Atherton, CA 94027

US Department of Energy Edward Warman NS-50 (GTN) Stone & Webster Engineering Corp. NS-10.1 P.O. Box 2325 S-161 Boston, MA 02107 Washington, DC 20585

J.E. Werner U.S. Environmental Reactor Research & Techn Division Protection Agency (2) US DOE Idaho Operations Attn: Allen Richardson MS: 1219 Joe Logsdon 850 Energy Drive Office of Radiation Programs Idaho Falls, ID 83401-1563 Environmental Analysis Division Washington, DC 20460 Dr. Wolfgang Werner Safety Assessment Consulting GmbH Harold VanderMolen Veilchenweg 8 USNRC-RES/PRAB D 83254 Breltbrunn MS : T- 9F31 GERMANY

Dr. A. Valeri Westinghouse Electric Corp A. N.P.A. Technical Library Via Vitaliano Brancati, 48 P.O. Box 355 1-00144 Rome East 209 ITALY Pittsburgh, PA 15230

Magiel F. Versteeg Westinghouse Electric Corp Ministry of Social Affairs NTD and Employment Central File Nuclear Safety P.O. Box 90804 P.O. Box 355 2509 LV Den Haag 408 1-A THE NETHERIJ_S Pittsburgh, PA 15230

Martin Virgilio Westinghouse Electric Company (3) USNRC-NRR/DSSA Attn: John Lacovln MS : 0-8E2 Burr Morris Griff Holmes Energy Center East, Bldg. 371 R. Virolainen, (Chairman PWG5) P.O. Box 355 Systems Znteg. Off. (STUK) Pittsburgh, PA 15230 P.O. Box 268 Kumpulanite 7 SF- 60101 Helsinki FINLAND

Dist -13 t Westinghouse Savannah River Co. (2) Attn: Kevin O'Kula Jackie East Safety Technology Section 1991 S. Centennial Ave., Bldg. 1 Aiken, SC 29803

Donnie Whitehead Department 6412, MS: 0747 Sandia National Laboratories P.O. Box 5800 Albuquerque, NM 87185-0747

Keith Woodard PLG, Inc. 7315 Wisconsin Ave. Suite 620 East Bethesda, MD 20814-3209

John Wreathall john Wreathall & Co. 4157 MacDuff Way Dubin, OH 43017

M. K. Yeung University of HongKong Mechanical Engineering Dept. Polfulam HONG KONG

Bob Youngblood Brookhaven National Laboratory Department of Nuclear Energy Bldg. 130 Upton, NY 11973

Carlo Zaffiro A.N.P.A. Directorate for Nuclear Via Vitaliano Brancate, 48 1-00144 R_ne ITALY

Dr. X. Zikidis Greek Atomic Energy Coma. N.R.C.P.S. "Demokritos" GR-153 10 Agia Paraskevi Attiki GREECE

Dist - 14 q_