ML20101K499
| ML20101K499 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 08/31/1991 |
| From: | ENERGY, DEPT. OF |
| To: | NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| Shared Package | |
| ML20101K489 | List: |
| References | |
| CON-FIN-A-9135 ORNL-NOAC-266, NUDOCS 9207020317 | |
| Download: ML20101K499 (44) | |
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ORNL/NOAC-266 li REVIEW OF THE OPERATING EXPERIENCE OF FITZPATRICK i
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DIAGNOSTIC REPORT i
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DitAIT ORNL/NOAC-2(4 REVIEW OF Tile OPERATING EXPERIENCE FOR FIT 4 PATRICK FROM JANUARY 1989 - APRIL 1991 Engineering Technology Division Nuclear Operations Analysis Center August 1931 Prepared for the NUCLEAR REGULATORY COMMISSION OFFICE FOR ANALYSIS AND EVALUATION OF OPERATIONAL DATA under Interagency Agreement DOE 18868005-2F NRC FIN No. A9135 Prepared by the OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee 37831 operated by MARTIN MARIETTA ENERGY SYSTEMS, INC.
for the U.S. DEPARTMENT OF ENERGY
TAllLE OF CONTENTS 1-
1.0 INTRODUCTION
2.0 ANALYSIS OF LERS AS A FUNCTION OF REPORTA131LITY CODES 3
2.1 10 CFR 50.73(a)(2)(v) [ Event That Could llave Prevented Fulfillment of a Safety 3
Function).................................,.
3.0 ANALYSIS OF PERSONNEL ERRORS.....
4 3.1 Actual Procedural Deficiencies,,.....................,.,.
4 3.2 - Design Related Errors............................
5 33 Operating and hiaintenance Errors............................
5 3.4 hiaintenance Deficiencies..........
6 4.0 ANALYSIS OF COhtPONENT FAILURES......
7 4.1 Check Valves.....,
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4.2 hiiscellaneous Subcomponents.
8 43 hiotor Operated Valves 8
5.0 ANALYSIS OF SYSTEh! AND TRAIN OCCURRENCES -
9 5.1 High Pressure Coolant injection System..
9 5.2 Residual Heat Removal Sptem.................
9 5.3 Train Occurrences....................
.. 10 APPENDIX A:- DATA TABLES FOR OPERATMG EXPERIENCE REVIEW FOR FITZPATRICK A.1 i -
APPENDIX B: LISTING OF ABSTRACTS FOR FITZPATRICK LERs B.1 d
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1.0 INTRODUCTION
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The Nuclear Operations Analysis Center (NOAC) was regrested by NRCs Office for Analpis and Evaluation of Operational Data (AEOD) to review selected aspects of the operating experience for Fit 7 patrick for the time
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period of January 1989 through April 1991. This review will assist NRC staff in preparing for a Diagnostic Team
-inspection of Fitzpatrick.
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IL This review differs substantially from all presious operating experience reviews conducted by NOAC in that (1) only selected areas of operations were reviewed as determined by AEOD staff based on data provided by NOAC i
and (2) as a consequence of only reviewing selected areas, no overall findings regarding plant operations will be provided. Each area reviewed was analyved separately with any findings or observations provided being relevant to that area only.
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j NOAC provided AEOD with Tables 1.1 through 1.9 as presented in Appendix A.
These tables reflect information compiled for analysis when conducting a fully comprebensive review as donc previously. AEOD staff 1
l reviewed these tables and made determinations on which particular areas to be further analyze. These areas were:
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Licensee Event Report (LER) reportability criteria 50.73(a)(2)(v) - any event or condition that alone could have prevented the fulfillment of a safety function.
(Table 1.2).
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(2)
Personnel errors invohing test / calibration activities, administrative control, maintenance, and operations (Table 1.5).
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(3)
Component failures - specifically check valves, miscellaneous subcomponents, and motor-operated valves (Table 1.7).
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System level occurrences - specifically the high pressure coolant injection system (HPCI), residual heat removal system (RHR), and the essential raw cooling /senice i
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water system (ERW/SW)[ Tables 1.8 and 1.9].
l The peer group of boiling. water reactors (BWRs) in which Fit / patrick is included is one of nine peer groups developed by NOAC for NRCs performance indicator (PI) program (Table 1.1 in Appendix A). The various l
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tables in Appendix A present data for Fit 7 patrick versus other plants in this peer groups for comparative-purposes. The data presented for the peer groups does not included data for Fitzpatrick The data in the tables are based on LER information contained in the Sequence Coding and Search System (SCSS). The indicated number of personnel errors, component failures, system occurrences, etc., presented in i-the tables reflects actual numbers of errors or failures as encoded in SCSS and not just a count of LERs. Thus, the number of errors or failures for a given category may be larger than the number of LERs noted following the tables since any one LER may report multiplc errors or failures.
Appendix B includes a listing of the abstracts of events for Fitzpatrick included in this operating experience review.
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2.0 ANALYSIS OF LERS AS A FUNCTION OF REPORTAlllLITY CODES 2.1 10 CFR 50.73(a)(2)(s) (Esent That Could liase Presented Fulfillment of a Safety function],
Fitzpatrick reported 13 events pursuant to 10 CFR 50.73(a)(2)(v)[ Event That Could llave Prevented Fulfillment of a Safety Function] during the review period, which represented a higher percentage of events reported pursuant to this category as compared to other BWRs in its peer group (Table 1.2 in Appendix A). Equipment failure was determined to be the root cause for five of the 13 events; a design, construction, or manufacturing problem was cited as the root cause of a different set of five events.
Nine of these 13 events described failures of the liigh Pressure Coolant injection (IIPCI) system, All of the failures except those described in 333/89-025 and 333/90 010 were potential failures of the 1IPCI system. These two LERs describe actual failures of the llPCI system. One was an system isolation during a test (333/89 025);
the other was a partial failure of the system due to excessive flow oscillations during an actual system demand (333/90-010).
Two of the nine IIPCI LERs (333/89-005 and 90-010) involved failures of the IIPCI system turbine speed controller. One of these events (333/89-005) was attributed to moisture-induced corrosion. An LER not included in this list (333/89-019) reported a separate moisture-induced corrosion failure of the llPCI turbine speed controller -It was initially submitted with a 50.73(a)(2)(v) reportability but was resised to a voluntary reportability based on a subsequent analysis. Only one of the llPCI LERs (333/89-018) occurred while the unit was in a mode other than full power operations.
Of these nine HPCI LERs, the corrective actions as stated by the licensee involved equipment repair in six events and procedural modifications in six events as well.
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.Y 3,0 ANAINSIS OF PERSONNEL ERRORS Table 1.5 in Appendix A indicates 74 occurrences of human error reported in 40 LERs for Dvc selected cause areas: maintenance, testing / calibration, design, administrative, an-1 operations. An evaluation of these 40 reports
- determined that the events could be grouped into four general areas:
(1) actual procedure dcDciencies (2) design errors (3) operating and maintenance errors, and (4) deGeient r.cintenance program 3.1 Actual Procedural Denciencies.
Some 16 events (333/89-007,89-008, 89-013,89-016, 89-018,89-023, 89 025, M007, %010, %011, %013,90-015, %022,90-023, %028 and 91-002,' were attributable to procedures that appeared to be unclear or misleading. Three events (333/89 018,89-025 and 99-010) and another event not directly attributable to human error (333/89-002) resulted from overly conservative setpoints and/or technical speci0 cations in the llPCI system, leading to operating constraints that were difn: ult to obsene under certain operating conditions. In two instances (LERs 333/89413 and %013) problems were reported with control of jumpers. The remaining eleven
- LERs involved missed surveillances, Technical Specification violations, and engineered safety feature (ESP) -
r.ctuations or scrams resulting from actions taken in accordance with procedures, A comparison of the number of occurrences for " task description inadequacy" (procedural denciencies and communication problems) as a cause for personnel error showed that Fitzpatrick's experience (23 occurrences) is not significantly different from other plants in the same peer group (24 occurrences per plant). No clear pattern was evident as to the nature or cause of procedural deficiencies. The utility is in the process of reviewing and upgrading their procedures.This process may uncover additional reportable cases of procedure deficiencies, possibly leading to more reports in this area in the future.
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3.2 Design Related Errors.
Another 16 events involved problems arising from design errors found during plant operation or during reviews conducted by the utility operating staff or reviews by the NRC. In one case (333/90-014), no documentation could be found for a modification made to llPCI system -heck valves sometime in the past 10 years; a second ease (333/89-011) involved a design charge made in 1979 that had not been adequately documented; and in a third instance, engineering personnel found a design error existent since plant construction where loss of instrument air could cause loss of IIVAC cooling to safety-related electrical distribution equipment (333/89-004).
Corrosion of IIPCI control circuit components from harsh conditions led to two events (333/89 005 and 89-019);
the IIPCI system also suffered from lube oilleakage that was not anticipated in the original design of the llPCI turbine (333/89414). Four events (333/90-002, %011,90-016, and %020) were submitted for problems apparently arisirg from air trapped in RilR system instrument sensing lines. The latest report (333/%020) mentioned a design modification was being planned to remedy this recurring problem.
Other design related problems include drawing errors (333/90-021) and incorrect evaluations or interpretations made in developing designs or operational parameters (333/89-010, %029 and 91-005). One event (333/89-012) reported a design deficiency uncovered during an NRC Safety System Functional Inspection (SSFI) where protective switchgear functions for the emergency diesel generator could be lost under certain operating conditions.
The number of design problems reported is more than twice that for other plants in the same peer group for the same period. This appears high for a plant in commercial operation for almost 15 years. It was not apparent from the involved LERs that the licensee was actively engaged in some type of ongoing design review or I
verification program, which typically explains increased rates of reported design errors. The rate of design error reports was about 7 to 8 reports per year for 1989 and 1990.
3.3 Operating and Maintenance Errors.
- Fitzpatrick has experienced twice the number of occurrences of operating and maintenance errors as the_ peer
- group for the resiew period (20 occurrences versus 10 for the peer group).
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in two events, inattention to detail appeared to be the cause for error, in one instance (333/84-003), an operator simply manipulated the wrong switch; in the other (333/89-009), an operator opened valves that were to be-closed. Poor communication between shift crews regarding data collection was an' identified cause in LER 333/90-008
'One rather significant event (333/91-004) resulted in the release of radioactive contamination, which initiated an NRC Augmented Inspection Team investigation. In this event, operators failed to follow procedures for waste concentrator operation, causing radioactive contamination to be released to plant emirons through the auxiliary boiler atmospheric vent, Errors by maintenance technicians were indicated in three other events (333/89-017, %001 and 90424). In LER 333/89-017, operators and instrument technicians overlooked logic circuit connections for tagout; in 333/ S(X)1, a reactor scram was caused by a technician opening an instrument sensing line valve too rapidly, causing a false low reactor water level signal in LER 333/90-024, an ESF actuation occurred because of poor communication between shifts and multiple human errors in performing equipment protective tagging.
No clear trend in personnel errors is evident, but the number of occurrences in the reports for the period is higher than typical for plants in the same peer group.
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Maintenance Deficiencies.
Two events (333/89-015 and 90-012) involved sludge buildup and fouling of pipes and valves in the Reactor
' Building Cosed Loop Cooling and Normal / Emergency Senice Water systems. The fouling was discovered during performance of the AShfE Section IX Inservice festing program or through preventive maintenance.
The sludge found in these cooling weer systems would have been detected by periodic sampling of the coolant.
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- System filters (if installed) should have been checked more frequently. One other event (333/89:022) described a lack of maintenance records for snubbers. Consequently, the snubber seals were assumed to be at end of-life and subsequently changed out.
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c 4.0 ANALYSIS OF COMPONENT FAILURES Single failure of components are not generally required to be p.~,cu m LERs. Ilowever, component failures are frequently the initiators of events which themselves are reportable. It is in this cont;xt that emmponent failure data from LERs should be used. That is, LERs should not be viewed as a comprehensive source for component failure data.
Table 1.7 in Appendix A lists the number of fai'ures for each of the primary contributors to the total number l
of component failures at Fitzpatrick. The corresponding average number of failures at the other plants in the peer group is also shown. Components with fewer than five failures are not listed in the table. Failures are defined as actual or potential urJesired equipment performance such that the specific component of interest requires a repair action. Simuy resetting switches or manipulating valves does not constitute a repair action; however, repacking a vahn rebuilding a pump, or replacing a diode would be considered repair actions.
Thc analysis in the following sections addresses those components which exhibited the highest number of failures that occurred at Fitzpatrick. The top two components (check vahrs and miscellaneous subcomponents) represent approximately 19% of the total failures during the review period. These are analped in the following two sections. The third section on motor operated valves is presided because of specific interer' in events invohing these components on the part of AEOD.
.l.1 Check Valves
- There were four events (333/B012,90-014,90-019, and %025), which reported check valve failures (three invohed check valves of unknown material and one involved carbon steel eheck valves) at Fitzpatrick.
Twenty-one check valve failures were reported in two LERs (333/%012 and %025) that were discovered during surveillance testing. Five of the 16 valves reported in LER 333/%012 failed to close during a test (also noted in 333/%025). These failures were the result of sand, silt, and corrosion products found in the normal and emergency service water systems, it was determined that these five check valves would not have closed on demand. Iloth LERs were voluntary reports. The valves were from three different manufacturers.
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4.2 Miscellaneous Subcomponents Therc were six events (333/89 AM, 89-0?ti, Wh012, MLO14,90-018, and 90-023), which reported failures of miscellaneous subcomponents at Fit 7 patrick.
Four of the events concerned problems with vahc discs (333/89-026,90-012,90-018, and Wh025). Two of the valve dises were on Targe: Rock SRVs (333/39 026 and %)-018). Ilowever, the cause of failure was different for each of the two events. The other two events involved check valves either in the normal or emergency senice water systems. Again, the cause of failure was different for each, in LER 333/8946 a leaking adapter for the Standby Liquid Control system accumulator nitrogen charging connection was reported. In LER 333/MbO14,
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the licensee reported a failure to initially install or reinstall a spring in a spring auisted check valve in the llPCI system.
4.3 Motor Operated Vuhes The licensees do not necessarily report the type of operator for vahes unleu the operator itself directly conuibutes to the valve failure. Following a review of the LERs for the period of interest, only one event for Fitzpatiick was Nuad that definitely described a failure of a motor operated valve which required repair (333/89-021). This exent was caused by an error in the fabrication of the motor. Because llPCI was inoperable it was required that the RCIC system be tested for operability. During testing the motor developed a short circuit. Prior to tripping the motor breaker, the motor was burned up. The motor was replaced and tested and RCIC was declared operable. Three other events were found which involved MOVs; however, they did not involve failure of the valve itself. In one event (89418) four MOVs were declared administratively inoperable due to missed tests, la two other events (333/90-001 and 90-026) loss of power due to open breakers caused the valves to fail to operate. Other events involved valves for which the operator was not specified. Ilowever, these events did not involve failures where these components. quired repair.
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5.0 ANALYSIS OF SYSTEM AND TRAIN OCCURRENCES Table 1.8 in Appendix A is a summary of total system occurrences at Fitzpatrick for the time period of interest.
Only those systems with more than one occurrence for the renew period are shon in this table. These faults or failures are defined to be undesired performance, alignments, or configurations of systems - not just catastrophic failures or those instances where the systems would not perform if called upon. A single LER may report more than one system fault or failure. Table 1.8 shows that for the 35 listed systern occurrences during the time period of this evaluation,16 involved the llPCI system and six involved the RHR system, 5.1 Iligh Pressure Coolant injection System.
The 16 HPCI system failures are described in eleven LE.u. Three of the LERs reported multiple occurrences:
(1) LER 333/89 019 listed four occurrences; (2) LER 313/89-025 noted two; and (3) LER 333/90-010 described two. Equipment failure was determined to be the root cause for three of the eleven events; a design, construction, or manufacturing problem was cited as the root cause of a different set of four events.
All of the failures except those described in 333/89 025 and 90-010 were potential failures of the IIPCI system.
These two events involved actual failures of the HPCI system. One was aa system isolation during a test (333/89-025); the other was a partid 'ailure of the system due to excessive flow oscillations during an actual system demand (333/90-01C).
Three of the 11 Hr"I related events (333/89-005,89-019, and 90 010) involved failures of the HPCI system turbine speed controller with two of these events (333/89-005 and 89-019) attributed to moisture-induced corrosion. Only one of the HPCI events (333/89-018) occurred while the unit was in a mode other than full power operations. Of these eleven HPCI LERs, the corrective actions for eight of them involved equipment repair and six involved procedural modifications.
5.2 Residual lleat Removut System The six RHR occurrences were reported in five LERs (LER 333/90-016 contains two RHR system occurrences).
Equipment failure was determined to be the root cause for two of the five events; procedural problems were cited as the root cause for another two events. Two of the six system occurrences were potential failures that 9
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- occurred during normal' power operations (LElis 333/89-001 and 89-012). The remaining four system occurrences were actual shutdown cooling system isolations due to false ESF signals during unit refueling or hot shutdown (333/90-011,90-013, and 90-016); these events occurred over a one month time frame during March-April 1990. The corrective actions for each of the RiiR LERs included procedural modification.
5.3 Train Occurrences Table 1.9 summarizes train failures at Fitzpatrick for the we period of interest. As noted before, these faults
- or failures are defined to be undesired performance, alignments, or configurations - not just catastrophic failures.
unly those systems that had greater than one occurrence at Fiapatrick are shown in the table, The nine train failures were reported in six LERs. Only those events for the Essential Raw Cooling /Senice Water system were reviewed in detail.
The three ERC/SW train failures were all reported in LER 333/90-025. That LER described the failure of three ERC/SW check valves to close during a test. The loss of three trains of ERC/SW created the potential for a loss of cooling flow to several area unit coolers which sene safety-relatea equipment. The root cause of the event was equipment failure; correction actions include replacing the valves' internals.
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APPENDIX A: DATA TABLES FOR OPERATING EXPERIENCE REVIEW FOR FITZPATRICK
Plants in Peer Group Used for Comparison with Fitzpatrick Table 1.1 General Electric B%Tt/2, BWR/3, and older BWR/4 Peer Group Plant name Docket BWR laitial Commercial Electrical Class Criticality Operation Rating Big Rock Point 155 0
09/24/62 03/29/63 75 Browns Ferry 1 259 4
OS/17/73 0S/01/74 1065 Browns Ferry 2 260 4
07/20/74 03/01/75 1065H Browns Ferry 3 296 4
OS/08/76 03/01/77 1065 Brunswick 1 325 4
10/0S/76 03/18/77 821 Brunswick 2 324 4
03/20/75 11/03/75 S21 Ccioper Station 298 4
02/21/74 07/01/74 77S Dresden 2 237 3
01/07/70 06/09/70 794 Dresden 3 249 3
01/31/71 11/16/71 794 Duane Arnold 331 4
03/23/74 02/01/75 533 Fitzpatrick -
333 4
11/17/74 07/28/75 821 Hatch 1 321 4
09/12/74 12/31/75 776 Hatch 2 366 4
07/04/78 09/05/79 754 Millstone Point 1 245 3
10/26/70 03/01/71 660 Monticello 263 3
12/10/70 06/30/71 545 Nine Mile Point 1 220 2
09/05/69 12/01/69 620 Oyster Creek 219 2
05/03/69 12/01/69 650 Peach Bottom 2 277 4
09/16/73 07/05/74 1065 Peach Bottom 3 278 4
08/07/74 12/23/74 1065 Pilgrim 1 293 3
06/16/72 12/01/72 655 Quad Cities 1 -
254 3
10/18/71 02/18/73 789 Quad Cities 2 265 3
04/.M/72 03/10/73 789 Vermont Yankee 271 4
03/24/91 11/30/72 514 l
A-1 For 1989,1990, and 1991
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[ Table 1.2 - Comparison of Reportabilit) L.tegories for Fit 7 patrick to all BWRs-Reportability Category Perrentage of all Peer
- Percentage of Fitzpatrick Group LElts LERs 10 CFR 50.73(aX2)(tv) ESF Actustions 46 45 l 10 CFR 5033(a)(2Xi): Shutdmms of Technical Specification Violations 33 27 10 CFR 50.73(a)(2)(v) Event that could have protnted fulfillment of a 13 22 safety function OTHER: Voluntary report, special report, Part 21 report, etc.
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10 CFR 50.73(eX2)(vii): Single failurt critena 3
3 10 CFR 20J3(aX2Xii): Unanalped condaion 7
3 10 CFR 20.405(a)(1)(i): Excess exposure of an individual to radiation i
2 10 CFR 20.405(a)(1)(v): Execss lotis of radiation or tadioactivt i
2 matenalin an unrestneted area 10 CFR 50.73(a)(2)(viii)(B): Liquid effluent radioactive release -
6 2
s - less than one per cent LERs for ESF Actuations 333/89-003 333/89-005 333/89-010 333/89 013 333/89-016 333/89-017 333/89 4 18 333/89-020 333/89-023 333/89-024 333/89 025 333/%001 333/90-002 333/90-004 333/90-009 333/B013 333/90-016-333/%011 333/90-020 333/B021 333/90-022 333/ % 023 333/B024 333/B026 333/ % 027 333/D028 333/91-001 LERs for Shutdown or Technical Specification Violations 333/89-007 333/89-00S 333/89-009 333/89 011 333/89 4 26 333/B003 333/90-007 333/ % 008 333/90 014 333/90-015_
333/B017 333/90-018 333/90-029 333/91-002 333/91-003 333/91-005 LERs for Event that Could Have Prevented Fulfidiment of a Safety Function 333/89-001 333/89-002 333/89-0CM -
333/89-005 333/894)08 333/89 012 333/89-014 333/89-018 333/89-021 333/89-025 333/B004 333/90-005 333/%010
- LERs for Voluntary report 333/89 019 333/89-022 333/E012 333/90-019 333/90-025 A-2 For 1989,1990, and 1991 I
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Table 1.3 Number of LERs reporting RPS'Actuations while Critical at Fitzpatrick and Other Peer Grcup Plants Plant Docket Number of LERs Reporting RPS Actuations While Critical
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Arnold 331 12 Brunswick 2 324 8
Fitzpatrick 333 7
Hatch 1 321 7
i Pilgrim 1 293 7
Dresden 2 237 6
P '.per 298 5
Millstone 245 5
Nine Mile Point 1 220 5
Dresden 3 249 4
Quad Cities 1 254 4
Peach Bottom 2 '
277 4
Itatch '.
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4 Peac tom 3 278 4
Vermo Vankee 271 4
Brunswick 1 325 3
Quad Cities 2 265 3
Oyster Creek 219 3
Monticello 263 3
Big Rock Point 155 1
LERs reporting RPS actuations at Fitzpatrick 5'
333/89 020 333/89 023 333/90-001 333/OO(F) 333/W-023 333/B026 333/90-027
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A-3 For 1989,1990, and 1991
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Table 1.4 Number of IIRs reporting ESF Actuations at Fitzpatrick and other Peer Group Plants Plant Dociet Number of LER$ Repor ng ESF f
Actuations i
Pilgrim 1 293 34 4
Brunswick 1 325 34 I
Brunswick 2 324 33 Arnold 331 29 Monticello 263 2A Peach Bottom 2 277 26 Fitzpatrick 333 25 l Ouad Cities 1 254 22 Dresden 2 237 21 Browns Ferry 1 259 21 i
Cooper 293 20 Nine Mile Point 1 220 20 Hatch 1 321 19 Vermont Yankee 271 19 Browns Ferry 2 260 17 Hatch 2 366 17 Peach Bottom 3 278 12 Oyster Creek 219 10 Browns Ferry 3 296 to Millstone 1 245 10 l
Dresden 3 249 8
Quad Cities 2 -
265 5
LERs Reporting ESF Actuations at Fitzpatrick 333/89-003 333/89 005 333/89-013 333/89-017 333/89-018 333/89 020 333/89-023 333/89-024
-333/89 025 -
333/90-001
-333/90 002-333/90 004 333/90 009 333/90 011 333/ % 013 333/90-016 333/90-020 333/90 021 333/90-022 333/ % 023 333/90-024 333/90-026 333/90-028 333/91-001 r
333/910M n-4 For 1989,1990, and 1991 7
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Table 1.5 Personnel ActMty versus Cause for Personnel Errors at Fittpatrick 5
Tersonnel Intrinsic Task Unknown inadequate Total ActMty lluman Error Description Cause Man Machine inadequaqr Interface Maintenance 3
9 4
2 20 Testing 7
10 0
2 19 Calibration I
Design 16 0
0 0
16 Adminisitative '
11 1
1 0
13 Operations 1
3 0
2 6
installation 0
1 2
1 4
Fabrication 1
0 3
0 4
i Radiation 2
1 0
0 3
l IrottClion w_
Construction 3
0 0
0 3
Uetncmn 0
-0 2
0 2
Total 46 23 12 7
4 LERs for Maintenance / Intrinsic Iluman Error 333/89-005 333/89 016 333/90 013 333/90-024 333/91-ON LERs for Maintenance / Task Description inadequacy Errors 333/89 016 333/89-017 333/89-022 333/90-013 333/90-021 333/90 023 333/90-024 333/904)28 333/91-004 LERs for Maintenance / Unknown Cause 333/89-019 333/89-025 333/90-014 333/R)-029 LERs for Maintenance / Inadequate Man Machine Interface 333/89 013 333/99-013 LERS for Testing / intrinsic Iluman Error 333/89-003 333/89-007 333/89-009 333/89-023 333/89-024 333/904Xll 333/90-(X')S A-5 For 1989, W)0, and 1991
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LERs for Testing / Task Description inadequacy Errors 333/89-007 333/8940S 333/89-018 333/89-023 333/89 424
%. T R13 333/90407 333/90 010 333/90-015 3 W 122 LERA for Testing /i; _J.-quate Man' Machine Interface 333/894G 333/90-001 LERS for Design / intrinsic Ilunian Errors 333/89 004 333/89408 333/89 010 333/89-011 333/89-012 333/89-013 333/89-015 333/89 4 18 333/89-025 333/90-012 333/90 419 333/90420 333/90-021 333/90429 333/91-004 L
l LER$ for Administratht/ Intrinsic lit..an E.rrors l
333/89-017-333/89 018 333/89 025 333/90-003 333/90-008 333/90-029 333/91 002 333/91404 333/91-005 LERs for Administrative / Task Description inadequacy 333/91-002 LERs for Administrative / Inadequate Man Machine laterface 333/90-003 LERs for Operations / Intrinsic Iluman Error 333/91-004 LERs for Operations / Task Description inadequacy 333/90-011 333/91-002 333/91-004 LERs for Operations /leadequate Man Machine Interface 1 333/90-023 2 333/910J4 A-6 For 1989,1990, and 1991
e w.
l Table 1.6 Pc.rsonnel ActMty versus Cause for Personnel Enots at Other Peer Group' Plants (average nutuber of cr Ors per plant)
FerSonnel Intnnue Task Unknm inadequate Total ArtMey llaman reror Description Causes Man Machine innsequacy Interface Maintenance 3
1 1
10 Testing /
12 0
2 19 Calibration Design 7
0 0
7 Mministratin 11 3
0 0
14 Operations =
2 4
0 1
7 Installatkin 1
1 0
0 2
l'abrkation 1
0 0
0 1
RaJstion '-
1 0
0 0
1 Protection Construction -
0 0
0 0
0 Unkatwa 0
0 1
0 1
Total 31 25 2
4 62 l'
A-7 For 1989,1990, and 1991 I
l
Table 1.7 Dominant Component Failures at Fitzpatrick and at Other Peer Group Plants Peer Fittpatrick Group Plants (avg)
Valve, Check, Unknown Material 2
16 Mir.cellaneous Subcomponent 6
11 Seal 4
10 Connector 2
9 Valve, Isolation / Shutoff, Unknown Material 4
6 Drain 0
5 Valve, Relief v.ith Operator, Unknown Material 1
5 Total number of component failures including those not 84 145 shown in this table LERs for Valve, Check, Unknown Material 333/90 412 333/90-014 333/90 419 LERs for Miscellaneous Subcomponent 333/89-006 333/89-026 333/90-012 333/%014 333/90-018 333/90-025 LERs for Seals 333/89-001 333/89-007 333/89-014 333/89 015 333/89-022 333/B001 333/90-024 333/E027 LERs for Connectors f
333/89 4 19 LERs for Yalve, Relief with operator, Unknown Material 333/89 014 333/89 4 15 333/89-025 333/ % 001 333/90 424 333/B026 A-8 For 1939,1990, andifM1 L
Table 1.8 Summary of System Failures at Fittpatrick' Sptem Number of Sptem Failures et Number of Sptem Failures at (per plant) at Peer Group Fitzpatrick Plants Iligh Pressure Coolant injection 4.3 16 Residual }{ cat Removal 2.2 6
Fire Protection 0.1 4
Low Pressure Core Spray 1.1 3
Reactor Core isolation Cooling 1.3 2
Circulating Water 0.1 2
Condensate and Feedwater 1.6 2
- Only those sptems that had greater than one occurrence at Fittpatrick are shown in this table LERs for liigh Pressure Coolant Injection System Failures 333/89-002 333/89-005 333/89-012 333/89-014 333/89-018 333/89 019 333/89-021 333/89-025 333/90-005 333/90-010 333/90-014 1.ERs for Residual llcat Removal 333/89-001 333/89-012 333/90-013 333/90 016 333/90-011 LERs for Fire Protection 333/89 411 A-9 For 1989,19'X), and 1991 l
4 a
f Table 1.9 Summary of Train Failures at Fitzpatrick' i
Sptern Number of Train Failures Number of Train Failures at j
(per plant) at Peer Group Fitzpatrick
~
Plants Essential Raw Cooling / Service Water 0.77 3
Residual l{ eat Removal 2
2 t
Condensate and Feedwater 0.S$
2 Control Building 0.32 2
- Only those systems that had greater than one occurrence at Fitzpatrick are shown in this table LERs for Essential Raw Cooling /Senice Water 333/90 025 LERs for Residual 11 eat Removal 333/90-002 333/90-024 1.ERs for Condensate and Feedwater 333/90-026 333/90-027 LERs for Control Building 333/91 004 I
l l
A-10 For 1989,1990, and 1991 l
l l
u....
AI'I'ENDIX II: LISTING Ol' AllSTRALTS l'OR Fli"41%TP.lCK 11Rs 1
6 r w t W o -- er w- +
.-rm,,-
ve-
.,m2my.__.,_...g_,
f FOR.M
........................ 0 8. - 0 8.. 9.1
....... 1............... L. E. R.. S. C. 5 5.. D. A T A DOCKET YEAR LCR NJMBER REVISION 005 NUMBER N5IC EVENT DATE
. 333... 1969..... 001........ 0..... 89040E.0037... 213 50.1..... 02 /. 2 8. /.8 9 ABSTRACT PUWIR LEVIL - 100%. DURING ROU11NE EQUIPMENT IN5PEC110N VHILE OPERATING AT 1001 RAf t0 POWER ON 2/28/89, PERSONNEL NOTED A LOOSE MOTOR LEADS TERMINAL BOA ON RESIDUAL HEAT REMOVAL (RHR)/ LOW PRE 55URE CDOLANT INJf CTION (LPCl) (BO) PUMP A. THE PUMP VAS ADMINISTRATIVELY MADE INOPERABLE VHEN REQUIR[D BY TECH $PEC 3.5. A.I. ADDITIONAL INSPECTION REVCALED THE SAME PROBLEM ON ALL PUMP MDlDR5 IN RNR/LPCI AND LOV PRES 5URE C0tt ! PRAY (BM) SY51[MS.
THE CAUSE OF THE LOOSE TERMINAL BOX BOL11NG WAS APPARENTL1 DUE 10 AGE RILATED HARDENING OF THE GASKET OR INADEQUATE THREAD ENGAGEMENT OF THE BOLT 5. EACH OF THE 6 PUMP 5 INVOLVED WAS OUT OF 5!RVICE FOR AN AVCRAGE OF APPROK!KATELY f.5 HOURS TO ALLOV INSTALLATION OF LONGER BOLTS, LOCrVASHERS, AND 11GHTENING OF BOLTS. PER10Dit INSPECTION VILL BE IMPLEMENTED 10 PREVtNT RECURRthrt. THE EVENT HAD NO EFFECT ON ENVIRONwtNTAL QUAllFICAil0N AND THERE RAVE NOT BEEN ANY SlHILAR LICENSEE EVENT REPORTS AT THI5 FACILiff.
....... 2................ L. t. R.. S. C. 5 5.O. A. T A...................... 09. 0 8..- 9.1 DOCKET YEAR LER NUMBER REVISION DCS NUMBIR N5]C EVENT DATE
. 333.. 1989..... 002........ 0.... 89041303 5 7. 213 58.1..... 0. 3/02./.E.9 ABSTRACT POWER LEVEL - 100% AT 1330 HOURS ON 3/2/89 DURING NORWAL OPERAi!ON AT 100% POWER, THE HlGM PRES $URE CDOLANI INJEC110N (BJ) SY$1EM VAS DECLARED INOPERABLE VHILE REQUIRED BY TECHNICAL $FECIFICATION 3.5.C.I.
DURING
!N-SIRVICE TE5flNG THE TURBlNE STOP VALVE OPENING STROKE 11ME EXCEEDED THE PROCEDURE LlMll BY 5 SECON05.
~
INVESTIGA110N ALSO REVEALED THAT THE SYSTEM STARTUP T!nf 0F 25 SECONOS CONTAINED IN THE F5AR VA5 EXCEEDED FOR SOMETEST5. TROUBLESHOOTING AND RECORDING OF KEY PARAMilERS RtvEALED A WlRING ERROR THAT APPARENTLY HAS EXI51ED 5!NCE PLANT CONSTRUC110N. THE TURBINE CONTROL SYSTEM RAMP GENERATOR, VHICH CONTROL 5 ACCELERATION DURING STARTUP, VA5 V! RCD TO BE IN!TIATED VHIN THE STOP VALVE VA5 FULL OPEN RATHER THAN VHIN LEAVING THE FULL CLOSED POSITION RESULTING IN A DELAYED INit!Ai!ON OF THE RAMP CENERATOR AND THUS DELAYtD OPENING OF THE TURBINE e
THROTTLE VALVE. CORRECT 10N OF THE VIRING ERROR, ADJUSTMENT OF STOP VALVE OPENING TIME, AND ADJUSTMENT OF THE OlL SYSTEM RE570 RED THE SYSTEM TO OPERABLE STATUS APPROXIMATELY 126 HOURS LAf tR. THE DELAYED SYSTEM STARTUP VA5 Lt$$ THAN DELAYS ASSUMED IN THE ACCIDENT ANALYSIS. THERE HAVE NOT BCEN ANY SIMILAR V! RING [RROR IVEN15 ON THIS SYSTEM,
.O.RM...... 3................L. E R. S. C. 5 5.O. A.1 A..................... 08. 08.91 F
DOCKET YEAR - LER NUMBER REVISION DC5 NUMBER N5IC EVENT DATE
. 333...1989...... 003......... 0...+ 89.04.180119.. 213582..... 03/.06./.89 ABSTRACT POWER LEVEL - 100%. ON 3/6/89 AT 1415 HOURS DURING NORMAL OPERAil0N AT 100% POVER, ONE REACTOR SAFETY RElitr VALVE (AD) VA5 INADVIRTENTLY OPENED FOR APPROXIMATELY 5 5(CONDS DUE 10 A PERSONNEL ERROR. THE VALVE WAS OPENED VHIN AN OPERATOR TURNED THE VRONG SVliCH DURING PERFORMANCE OF SCHEDULED SURVEILLANCE TE511NG. PLANT RESPONSI 10 THE EVENT VA5 NORMAL AND OPINING THE VALVE FOR 5 $ECONDS 15 ESSENTIALLY THE SAME AS A Rt0VIRED TICHNICAL SPECIFICAi!ON REQUIRED SURVE!LLANCE EXCEPT THAT THE IRROR RESULTED IN AN UNDE 51 RED MANUAL ACTUAT10N OF AN l
ENGINEEREDSAFETYFEATURE(EST). CORRECTIVE ACT10N INCLUDES CRITIQUE OF THE EVENT; COUNSELLING OF OPERATOR 5, INCLUDING THE EVENT IN OPCRATCR TkAINING; AND PREVIOUSLY PLANNED MODIFICATION OF THE CONTROL PANEL 10 CORRECT HUKAN ENGINEERING DEFICIENCIES THAT MAY HAVE BEIN A CONTRIRUT]NG CAUSE LER-87-01615 A SIMILAR EVENT lh VHICH A PERSONNEL ERROR RESULIED IN AN ESF ACTUATION.
e B!
_-c_.,
r - : -.__ x _ -. _ _ - - -. --
..................... L. E R.. S. C. 5 5.. D. A T A......................... C. 8.. 0 8.. 9.1 4
.F O. R. M DOCAIT YEAR LIR NUMBER REVISIch DCS huMBER h51C (Vthi DATE
. 333... 1989...... 004.......... 1.... 9010.19.0.l?4... 2.19.CBS..... 03./.09/.89 ABSTRACT POWER Ltytt - 100%. ON 3/9/83 DURING NORKAL CPERATION AT 100% POWER (NathttRING PERSONNEL IDINT!Flt0 A OtslGN Of f!CitNCY VHICH ORIGINAf tD DURlhG PLANT CONSTRUCTION, THE D(FlCIENCY V0VLD RESULT lh Loss 0F AREA COOLING FOR PARTS OF BOTH 5ArETY Dlvl510N5 0F SAFtif-RELATED AND NON $AFITY R[LAi[D LL(CTRICAL DISTRIBUTION SY51tMS AS A t
RL5Ull 0F LOS$ OF lh51RUMENT AIR (LD) TO THE COOLING SYST[M 1[MPIRATURt CONTROL VALVl5. INVt$11 GAT 10N Riv[Alf D THAT blMILAR DEF1CithCl[5 01D NOT EX15T FOR COOLING OF OTHER SAf TTY-RILAl[D [QUlPPINT. MANUAL BYPASS VALVE 5 FCR TH[ AFFCCT[D VALVES VERE TAGGID OPIN 10 A55URt COOLING IN THE tythi 0F LOSS OF AlR. A MODIFICAi!ON OF THf itMPERATUAt CONTROL VALVf5 TO CAUSE THE VALVI TO FAIL-OPEN UPON LOS5 Or AIR VILL Bt COMPL(itD LAllR. FAILURE OF THE VALVES IN THE CLOSED P051110N V0VLD HAVE RESULi[D IN A f(MP[RATURE INCREA$t R[QUIRlhG lhVL511 Gall 0N AND MANUAL OPENING OF BYPA55 VALV[5. THERE HAV[ N01 B(th ANY $1MILAR 0(SIGN D[FICithCY EV[hi$ R[SulflNG lh THE ILCORRLCT FAILURE MODE AT THIS FACIL11Y.
F.O. R. M...... 5............... L. t. R.. S. C. 5 5. D. A T A......................... C. 8.- 0 8 - 9.1 DOCrtT YEAR LER NUMBER REVISION DC5 NUMBER h51C
[VINT DATI
. 33 3...19 89...... 00 5........ 0..... 8 9 0. 519. 03. E. 2.. 2.13 9 09...... 0 4 /.12 /.6 9 l
ABSTRACT-POWER LEVEL 100%. DURING h0RMAL OPERAfl0N AT 100% RA1ED POVCR ON 4/12/89 Af 1225 HOURS, H]GH PRI55URE COOLANT INJECTION (HPCI) (BJ) VA5 MADI INOPERABL('TO FlND AhD CORRECT A GROUNDED ILECTRICAL CIRCULI (CJ). HFC] VA5 R[0VIRED TO BE OPERABLE BY TECHNICAL SPICIFICAtl0N 3.5.C.
LISTS OF OTHCR SYSTLM5 REQUIRED BY TECHNICAL SPECIFICA110N 4.5.C.1 VHth HPCI 15 INOPERABLE WERI COMPLtilD VIIH $Al!5 FACTORY RESULTS. CAUSE OF CIRCUlf GROUND VAS A RANDOM [ VENT DUE 10 ACCUMUL ATION OF CONDUCTIVE RU51/ CORR 0510N PRODUCTS. CLCANING PROVIDED SA115FAC10RY REFAIR. DURING RIPA!R EFFOR15 A VRONG REPAIR FART VAS lhSTALLED AND AN 150LAf!CN OF THE HPCI SYSf tM DCCURRID DURING FOST VDRK TESTING. THE VRONG FARI WAS IN51ALLID AS A R[$UL10F MI5Uh0tRSTANDING A YLhDOR itCHNICAL lei 1ER VHICH EXPLAINED PART NUMBER CHANGES. UP0 ATE OF THE 1[CHNICAL MANUALS AND $ PAR [ PARIS SY5itM VlLL BE COMPLETED TO PR[VINT R[0CCURR[NC[ IN THE FUTURE. PIR10D!C INSPECTION VILL BE DONE 10 PREvtNT RECURRENC[. HPCI WAS MADE OPERABLE.ON 4/13/89. APPRO11MAi[LY 32.2 HOURS AFTER BtlNG DECLARED lh0PERABLE. NO PREV!OUS LICEkstE
[VINI R[ PORT $ HAVE BEEN ISSUED AT THl5 FACILITY FOR HPCI INOPERABILlif OUE TO CIRCUIT GROUNDING.
.......................... 08..-9.1 08-
....... 6................ L. t R S. C. 0 5. D. A T. A DOCKET YEAR LER huMBER REVISION 005 NUMBER-N5it EVENT DATE
. 333... 1989..... 006........ 0..... 89060102 4 4. 214118...... 04/.19./89 ABSTRACT PDWER LEVEL - 100%. ON 4/19/89 AT 0400 HOURS DURING NORMAL OPERATION AT 100% POWER, BOTH STANDBY L10VID CONTROL (SLC) (BR) SYSTEMS VERE DECLARED INOPCRABLE VHFk REQUIRCD 10 BE OPERABLE BY TECHNICAL SPECIFICA110N 3.4, A DUE TO THE PUkP DISCHARGC ACCUMULATOR NITROGEN PRECHARGE BCING ZERO PSIG. SLC SYSTEM B LOST THE NITROGIN CHARGE DUE TO A SLOW LEAK IN THE CHARG!hG C0hhECTOR'S SPRING LOADED VALVE. SLC SYSTIM A LOST THE N!1C& GEN CHARGE WH(N THE BLADOCR CHARGING CONNECTION VA5 BR0(IN DURING CONN [Cil0N OF A PRES $URE GAUGI, SLC SY$1tM B VAS MADE OPERABLE
]N APPROX 1MATELY 1.9 HOURS BY CHARGING THE ACCUMULATOR. SLt SY5f tM A VAS MADE OPERABLE VITHIN APPROXIMAL [LY 8.2 HOURS - BY REPLACING AND CHARGlhG THE ACCUMJLATOR BLADDER.
NO SIMILAR LICINSEE Evthi REPORTS thv0LVING INOPERABILITY OF BOTH SLC SYSTEMS AT THE SAFE TIME HAVI OCCURRCD AT TH15 FACILITY.
B-2 u
_______.m.___
.FO. RM...... 7............... L. E. R. S. C. $. $.. D. AT. A...................... 0 8.. 0 8.. 9.1 DOCKET YEAR LER NaMBER RLVl$10N DCS NUMBER NSIC EVENT DA1E
. 33 3.. 1989..... 0.07.........1.... 89.11200070.. 215512..... 0. 5/09./ 89 ABSTRACT POVER LEVEL + 000%. ON $/9/89 DURING NORMAL 0FERAi!ON AT 100% RATED POWER, IT VAS DISCOVERED THAT $URVEltt ANCE OF $0ME FIRE BARRIER FENETRA110NS REQUIRED BY TECHNICAL $PECIFICATION 4,12.F VA$ NOT DONE V!1HIN THE Alt 0VED 11ME FOR A NUMBER OF FIRE BARRIERS. THE M!$5ED SURVEILLANCE VAS COMPLETED FOR INE IDENilFIED BARR]ERS AND DEFICIENCIES HAVE BEEN CORRECTED. THE EVENT Vt.5 CAUSED BY CHANCES TO FROCEDURE5 BASED ON A REFEkthCt MANUAL VHICH DID NOT INCLUDE ALL REQUIRED FIRE EARRIERS. CORREC11VE ACTION! INCLUDE: 1) ESTABLl!HING FIR! WATCHES AkD COMPLETION OF REQUIRED SURVEILLANCE 2) CORRECTION OF DEFICIENCl[5. 3) CORREC110N OF $URVEILLANCE PRDCEDURES.
AND 4) REVll!ON OF THE REFERENCE KANUAL. LER 86-001. 86 002, 87 017, 87 022. AND 8S 006 ARE SIMILAR MISSED
$URVE!LLANCE EVENTS.
.F O. RM...... 8................ L. E. R S. C. $. $.. D. A T A.....................08..- 08..9.1 DOCKET YEAR LER NUMBER REVIS10N 005 NUMBER N5!C EVENT DATE
. 3 33... 19.89...... 008.........1.... 9010.16029 9. 2.19686...... 05/.17/.8. 9 ABSTRACT POWER LEVEL - 100%. ON $/17/89 AT 112$ HOUR $ DURING NOR".AL OPERAil0N AT 100% POWER IT VAS FOUND THAT
!URVEILLANCE TESTING OF HIGH FRES$URE COOLANT INJECTION (HPCI) (BJ) AND REACTOR CORE ISOLA 110N COOLING (RCIC)
(BN) AUTO 150LAil0N SYSTEM (JE) TIME DELAY DEV!CES HAD N01 DEEN COMPLETED AS REQUIRED BY TECHNICAL $PECIFICATION TABLE 4.2 2 SINCE INITI AL PLANT $TARTUP. LPCI AND RCIC STEAM SUPPLY !!DLATION VALVES VERE DECLARED INOPERABLE WHEN REQUIRED 10 BE OPERABLE OR CLOSED BY TECHNICAL $PECIFICATION 3.7,0.
INVE$flGATION ALSO REVEALED THAT HIGH ENERGY LINE BREAK (HEL8) ANALYSES DONE TO SUPPORT ENVIRONMENTAL QUALIFICATION HAD ASSUMED NO TIPE DELAY.
CIRCulTS VERE MODIFIED TO REMOVE TIME DELAY DEVICES AND THE ISOLATION VALVE $ ViRE MADE OPERABLE AT 0103 HOURS ON 5/18/89. CONSERVATIVE ANALYS15 TECHN!00E5 AND OTHER ISOLATION $1GNALS (VHlCH ARE NOT DELAYED) t!MITED POTENTIAL EFFECTS OF TIME DELAY DEv!CES VHICH VERE NOT ASSUMED 10 EXIST IN HEL8 ANALYSES. TECHNICAL
$PECIFICATION TABLE 4.2 2 WILL BE CORRECTED. TRAINING AND ORGAhlZATION CHANGES VILL RIDUCE THE POTENTIAL FOR RECURRENCE. LER-86-001, 86 002.87-071. 87-022,88-006. AND 89-007 ARE AL50 MI55ED SURVEILLANCE EVENTS.
........................ 06.-08.-9.1
....... 9................ L. E. R.. S. C. is. D. A T A DOCKET YEAR LER NUMBER REY!$10N DCS NUMBER NSIC EVENT DATE
. 333...1989...... 0. 09......... 0..... 89 062 7 04 59. 214 4 09......
05/.18./.89 ABSTRACT POVER LEVEL - 100%. ON 5/18/89 A10804 HOURS DURING NORHAL OPERATION AT 100% RATED POWER PRIMARY CONTAINMENT (NH) DRYVELL 70 PRES $URE SUPPRE$$10N CHAMBER (TORUS) DIFFERENTIAL PRES $URE AND TORUS VATER LEVEL DECREASED 10 LESS THAN THC LIMITS OF TECH $PECS 3.7, A.7. A AND 3.7. A.1.8 RESPECTIVELY AS A RESULT OF A PERSONNEL ERROR.
REACTOR BUILDING (NG) VENTILATION RADIATION MON 110 RING (IL) LOGIC $TSTEM FUNCTIONAL TESTING AND LOGIC SYSTEM
$!MULATED AUTOMATIC ACTUATION TESTING VAS BE!NG PERFORMED AS REQUIRLD BY RADIOLOGICAL EFFLUENT TECH $PEC TABLE 3.10.2. DURING THE TE511NG AN OPERATOR OPENED VALVES THAT THE PROCEDURE ST ATED VERE 10 bl CHECKED CLOSED, A SUBSLOVENT STEP CPENED OTHER VALVES IN PREPARATION FOR AUTO CLOSURE AND THUS PROVIDED A PATH FOR THE DRYVELL ATMOSPHERE TO FLOV TO THE TORUS RESULTING IN LOS5 OF blFFERENTIAL PRES $URE AND A DECREASE IN TORUS LEVEL. torus CHAMBER LEVEL VAS RESTORED IN 24 MINUTES AND O!FFERENTI AL PRESSURE IN 1,3 HOUR $.
CORRECTIVE ACTIONS INCLUDE:
- 1) STOPPING THE TE$T. 2) COUN$ ELL]NG PERSONNEL INVOLVED. 3) RESTORING DIFFERENTIAL PRES $URE AND LEVEL TO VITHlh IECH $PEC LIMITS. AND 4) INSTITUTING A TRAINING PROGRAM ON SELF-VERIFICATION TECHNIQUES TO UTIL12C VHEN PERFORMING A$$1GNED TASKS. LER-84-003.84-008. 86 019.87-016. AND 88 002 ARE SIMIL AR EVENTS IN VHICH PERSONNEL ERROR $ DIREC1LY CAUSED ENTRY INTO LIMITING CONDITIONS FOR OPERATION, B-3
[
FO.RM...... 10............... L. E R S. C. 5 5.. D. A. T A.......................
08.-08.-91 Dott(T YtAR LlR huMBfR REV1510h DC5 huMBCR h51C LVIh1 DAlt
. 3 3 3.. 19 89...... 010......... 0..... 8 3 00 2 7 04 0 4........ 2.14410...... 05/19./89 ABSTRAC1 POWER LEVLL 100%. CN 5/19/89 AT 1255 HOUD5 DUI thG h0RMAL OrtRATION A1100% RAllD POW [R IT VA3 DETERMikf D ThAT THE DE5lGN OF C[RTAIN AIR CON 0!T!Dhlh5 [QUIPMthi OR01[CilVE 1 RIPS WAS h0T C0h515 TINT V!1H THE D[5IGN BA515.
THE [0UlPM[N10F CONC [RN KAlhTAlh5 A MILD [hylDONM: 1 F0R SAFt1Y-R[LAl[D 000 VAC ([D) (CAD C(N1[R$ VIThlN TH[
R[ ACTOR BUILDINS (h3) 10 A55URI CONFLl ANC[ VITH 10CFR50.49 ([NV!PONuthf AL QUALIFICATION) Fl0VIRIMih15. THE IVEhi VAS CAUS[D BY A DESIGN DEFICitNCY BY 1H[ VENDOR AND lhA0[QUAll REVIEW PY THE V[hDDR AND THE AUTHOR!lY.
THE DEFICl[NCY VAS CORRICit0 BY MDDIFICAi!ON OF THE [0U1PMENT ArFRCXIKAl[LY 8.2 HOUR $ AFi[R DISCOVERY ADDITIONAL CORRICTIVE AC110N5 INCLUOE $1 rih 51Hth!NS OF CONFIGURA110N CONTROL AhD THE PFDC[IS OF YtRIFYlh5 DESIGN AC11VITIES BY VthDORS. LIR-89-008 15 A siM!LAR [V[ht.
F.O.R M..... 1 1............... L. (. R S C. S. $.. D. A T A........................
08 08.-9.1 DOCLET YEAR LtR Ne>BER REV1510N DCS hJMBER L51C EVENT DAll
)
. 333.. 1989...... 0.11........ 0..... 8907100054... 2.14 548......
05/31/89 ABSTRACT j
POWER 1[VIL - 100%. ON 5/31/89 A1 1500 HOURS DURINS hDRMAL OFERA110N AT 100% POWER 11 WAS DETERMih[D THAT MODIFICATION (PLUGGlhG) 0F N0h 5AF(TY RELAT[0 FLOOR ORAINS (WK) th THE EMERLthCY DitSIL C[hERATOR (fDG) RDOMS (hB) HAD BtIN PERFORM [D V!THOUT PRIOR [ VALUATION AS REQUIRED BY 10CFR50.59. THE FLOOR DRAIN 5 WIR[ PLUGGl0 lh 1979 !k R!$P0h5I TO hEV YORK stall DCPAR1MEh1 Of thVIRoswENTAL C0h5tRVAT10N CON:[RNS RElaT[0 TO THE ACClC;h1AL RELEASE OF TOAlt CORR 0510N INHIB 110R IN EDG EN3thE CDDL ANT 10 THE [NVIRONMEhi (BS). THE 10X[C CORR 0510%
lhHlBITOR VA5 R[PLAC[D WITH A N0h-TCXIC C0kR051DN thHlB110R IN 1986. THE EDG ROOM FLOOR DRAlh PLUG 5 V[Pt REMOVID ON 5/31/89. PLUGGID FLOOR DRAlh5 th TWO OTHER PL Ahl AREA 5 ARE BElhG RIIVALUATED UPGRADE OF M301 FICA 110N C0h1ROL PROCEDURES 51kC[ 1979 (AkD PRIOR TO DISCOVERY OF THIS [Vthi) WILL REDUC [ THE P01[hil AL FOR RfCURRENCE. [ DUCE THE POT [NTIAL FOR R[CURRENCE.
FO.RM...... 12............... L. E. R S.C. 5 5. D. AT A........................ 0 8. - 0 8..- 9 1 DOCKET YEAR LIR kUMBER REVI5!DN DCS NJMBER N51C fvENT DATE
. 333... 1989...... 0.12......... 0..... 8908100091. 214995...... 07/.14/09 ABSTRACT POVER LEVEL - 100%. TECHNICAL $PICIFICATIONS REQUIRE A MONTHLY l HR FULL LOAD TEST FOR EACH OF THE 2 SE15 Of EMERGthCY DI(5tL GEhtRATORS (EDGS)(EK). THis IS ACCOMFLISHED BY CONh!CilhG THE EDG5 IN FARALLil [L(CTRICALLY VITH THE NORMAL STATION SERVICE TRANSFORMER (h55T)(EA) 10 EUPPLY LDA01h3 TO THE GEhiRA10R. AN AhALY515 REQUEST [D BY AN hRC SAFETY SYST[M FUACil0NAL INSFECTION (55F1) FOUhD THA1 IN THIS [LECTRICAL CONFIGURATION, THE OCCURRENCE OF A HYPOTHITICAL 3-PHASE BOLT [D EUS FAULT COULD RCSULT lh CURRthT5 VHICH V0VLD EXCI(D THE HDMEhTARY DUTY RATIN35 FOR CERTAIN 52 EMERGthCY SWITCH 5LAR 4160 VOLT circuli BRIAKERS (IB)(EA) BY 23% TO 27% DEFENDlhG ON THE BUS IN WhlCH THE FAULT OCCURRED. TH15 POSTULAT[D EVENT COULD RESULT lh P01EhT! AL FAILURE OF THE PPOTECTIVE FUNCTIONS OF THE SVliCHGEAR AND LOSS OF POWER TO THE EMERGthCY BUS AND EMERGENCY CORE COOLING SYSTEMS RIOUIRED FOR SAFE SHUTDDVN OF THE PLANT. THE PROBABillTY OF OCCURRENCE OF A BOLTED BUS FAUL1 COMBihED VITH 1
SWITCHGEAR FAILURE DURlhG THE 2 HR TEST FIRIOD HAS Et[N CALCULATED 10 BE LESS ThAk 9.24 E-7 BY A SIMPLIFIED BUT CONSERVATIVE PREllH1hARY ANALYSIS. AN EVALUATION AND JUSTlFICA110N FOR C0hilhutD OPERATION UNTil THE MARCH 1990 REFUEL OUTAGE HAS BEEN PREPARED. A5 A SHORT-TERM CORR!CTIVE ACTION THE SURVEILLANCE TO REOUCE RISK Or ACCIDthTAL FAuti.
E-4
FO.RM..... 13............... L. (. R !. C. S S. D. A T A......................... C. 9.- 0 3 9. 1 DDCL[1 Y[AR LIR huMEER RIVISION DCS hJ"EIR h51C filhi DAlf
.. 3 3 3...19 6 9..... 0.13......... D...... t 9 0 912 0 2 9 5....
2.152.14....... C. f / 0 3 / !. 9 ABSTRACT POWIR L(vtl - 100%. At 1:30 F.W. Oh B/3/f 3 A itC"NICI AN WA5 R[MOVlh5 A 1[*NEARY FYrA551(51 JJ8PER FRDM A Rf L AY IN ACCORCANCC V11H AN ArrROV[D rRDCIDURI. THI R(L AY WA5 LOCATID IN A CCht! AED SPACl (JL) A Sail 1Y Sf5flM DIVISIONAL 5(FARAT104 (CNDUlf (JL) EL0ct[D DIRICT ACCESS 10 THf G[L AY.
THE CONDull C0hk(C10415 JDlhtD 10 AN ADJActhi OlV1510NAL 5[ PARA 1104 REL AY thCLCSURL BCX, THE C0hhl:104 AhD thCLD5URf BC1 Aki kDY lh5ULAT[D AhD AGE A1 [LICTRICAL GRDUND. Oh! thD Of THE JUwr[R VAS (Chh[Cl[D TD A COST ACT ON TH[ LILA1. THE OTHt R thD WAS ACC10thfALLY BRUSH [D AGA]h$1 A (ChDj[Tl%[ GRDUh2[D SQRf A((,
1H]$ (AU$[D TH[ IRDI[CTIV[ IV$[ ID (L[AR AhD O![hERGilt THE 150LA110N circuli. DIlkIRL!llh31HI CIRCull CAUSID 2 LDLINCID Cr[RAl[D GRDur !! PRIMAkV C0hT Alhw[hi PUR5[ Si5T[M 150L A110h VALVES (JM) 10 CLOLL. THI IUS[ WAS RIPL A [D AND TH[ 2 k ALVt5 RIDPthtD WlfHlh 15 Mlh5.
THE CLC5thS OF 2 VALVES, lh51EAD OF ONLY 1 n Alit, FOR A S!hGLE fu!! RlVE ALID A WlRlhi (RROR.
tHDRT-TERW CORRf CTIVE ACTION ADCED A CAullCN 51 Att>[h110 THE FROC[DJR[ AhD AFPilID ELICTRICAL lh5UL Atlh3 TAFI 10 f HC CONDUli t0khtC10RS.
LONG-1tRM CORC[011W[ AC110N W!lL lh51 ALL C0hVthit h1LY LDCAl[D RIMDi[ BirA55 TERMlhAL5 FOR THE REL AY. THE WlRlhG (RROR WILL [f COLEICT[D 10 LR0v10[ A SIk3LE FU5! FOR [ ACH 150LA110h VALVI, LIR Bfr014 AhD BL 019 SHARE CDMMON [LIMthT5 W11H TH!t tyth1 FORM
.................L.t R 5 0 $ $. O. A T A.........................C..F C6-51
........14 DDCLII YEAR LfR AUME[R G[V!SION DCS huwfLR h51C
[VINT CATL
. 333... 1959...........140.........0..... 090927C210... 215313 Of/17/09 ABSTRACT 00VER LEVEL - 100%. THE STIAM SUFPLY 15DL A110N VAlbf FCR 1st HILn FRl55URE C00LA%1 1hJtCT!0h (HPCl) (BJ) 5YSitM TURBiht KAS B[th LEAK!hG A SMALL QUAhT!TY Or 51[ AW 10 TH[ TUPBlh[, WHIL[ TH[ VALV[ !$ th TH[ CLO5fD 0051110h.
5!hCI DECIMEIR 19S8. 2ECAUst THC TURBlh[ 15 IDi[, THE 51[AW Coh0th5IS lh THE TURPlhi CA$lhS AND SDw[ Or THE ACCUMULA1[D WATER 15 PUSHlD THROUGH TH[ $4Af f $[ AL$ AND [h1[451H[ (( Akih3 LUBRlf Allh3 O!L SYST[M WHEh HPCI 15 JTARIED. DUR!hG RErIAT[D IDLL AND STARTUP CYCLIS TH[ WAt[R ACCUMJLAl[5 lh TH[ TURB1h[ O!L SUNP. BfCAU5E CI TH(
CONC [RN WITH THE LEAKAGE, SAMPL[5 VERE DRAWh rRDM THC b;TTOM OF THE SUu# DN B/15/69 AND ALAlh DN 8/17/8L BOTH SAMrtis C0hT AlhCD SLIGHTLY L[SS THAh 50% HAT (R. ALTHDUGH, A LOTTDM S A.MPL[ fRDM THE 1!$ GALLON $ UMP 15 [1FLETID T0 C041 Alh WAlf R AkD 15 hDT R[PRESENTAllyt OF THE [hilRE SUND. H'C1 WAS kt0E lhDPIRABLE 0% 8/17/E3 10 R[rLAtt THE LUBRICATING Olt AND THE SYST[M VAS R[ TURNED 10 SERV!tt Arr&;xlMAT[LY 24 HOUR 5 LAT[R. CORRCCilVE ACT!Dh DURlhG THE 5(PTEMBER 19B9 MAlhTihAhCE DUTAGE VILL lhCLUD[ D!5Asif>BLY, thCP[C110N, AND REFAIR OF THE ([ AAINS 51EAM VALVE AND R[ PLACEMENT OF TH[ TURElh[ LHAFT SEALS. LONG-1[RW CORkiC11V( ACTION WILL (XF AND THE LUBRIC A110h Oll FRDGRAM TO INCLUOE SAwPLING THE HPCI LUfRICATlhi O!L Ch A CUARTERLY BASIS 10 rRDVIDE EARLY DEfLCT!0h Or ACCUMULA110N OF WATER IN THE SUMP, FO.RM...... 15............... L. E. R 5. C 5 5..O. A T A......................... C. B - L. B - 9 1 00CKIT YEAR LER hJMBER GEV1510N DCS NUMBER h5!C
[ VENT DAlf
.. 33....1989..........15.........1.... 891120023B..... 215B73...... 09/10/83 3
0 AESTRACT POWLR LEvil - 000%. ON StPT[HEER 18, 1969 DURINS A SCHEDULED OUTACE AND FERr0;MANCE OF A SCh[DUL[D ALu[,
SECTION XI th SERVICE T[51. TVD Or hlhE, RIMDIE MANUALLY CTERAT[D DI APHRASM AIR OFERATED C0hTAlkwEh1150L ATION VALVES (15V) ON THE REACTOR BUILD!h5 CLOSED LOOP CDOLlh3 WAl[R SY5itM (CC) F AILID THE ACCEPTANCE CR11[RIA FOR VALVE CLC5!h3 TIME. ONE VALVE VOULD NDI CLOSE EXCEPT BY MAhUAL OFIRAT10%. ANDTHER VALV[ CLD51hi TlwE fxC((0tD THE CR11ERIA BY 0.4 SECONDS OR 3.0 PERC[hi. SUBSIQUEh! OUTA$t CPERAfl0NS REk[ AL[D C0uMON FROBLLws VITH THE
$[ VIN VALVES VHICH th!TI ALLY PASSID THE CLO5thi 11M[ 1[51. VAL 4E5 AFFEA0tD 10 et BlhDINS Dut rRIN;1 PALLY TO THE BUILDUP OF 1RON CAlDE SLUDG[ IN lHE VALVE OPERATlh3 lhTERNAL ALTHOUSH THERC 15 SOME thDICAfl0N THAT THE ORIGIhAL PACKlh3 MAY HAVE CONTRIBUTED TO TH[ FRDBLtM CORR [CilvE ACTION: 015A5L[MBL(D AhD CL(AN[D ALL hlhE VALVES AND CHAN;[D PACK!h3 FRDu SEVEh Rlh3 GRAFO!L
' LCADFD flVE RIN3 TYPE ON FIVE VALV[5. THR(( OTHER VALVE 5 KAD PREVIOUSLY BE[h CHANGED TO THis TYPE Di f ahL lh 19BB, LONS-TERM CORRECT 1Y[ ACT10N WILL FLU 5H SYSTEM TO REDUCE 1RON 0x!DE BUILDUP AND thvE51! GATE POSSIBLE ChAk3E5 TD lhtERNAL VALVE TRIM VHlCH 15 LESS SUSCEPTIBLE TG CORRDSION FRDDUCT ACCUMULATICN. LER$ WITH CCv40N ELEMthTS: EB-00L 68-009, AND BE-COL B-5
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...... 10................ L I R.. S C. 5 5. D. A T A..................... 0 0.. 0 B. 9.1 FO.RM l
DOCLET YEAR LtR NUMBER RIV1510h DCS huMBfR h510 Iv!hi DA.i
. 3 3 3...1989...... 0.16.......... 0.... 6 9.107 5 03 2.1. 2150 0 2...... 09. /.16. /. 0 9 i
l t
ABSTRACT POWER LIVIL ODOL Oh $tPTEMiiER 18, 1909 VITH Th[ PLAh1 SHUTDO.'h FOR A PL Akh[D MAlh1[ NANCE OUTAGt, THE P0 '[R l
$USPLY FOR THE REACTOR PRDTICT10h 5Y11tM (kPS)(VC) BUS WA$ TRAur[RRID FPDM hDRMAL P0b[R SUPPLY FRDM 1HI RP$
i MDTOR Gth[RATOR Stts (MG)(MG) TO THE ALTIRNATE PDvtR $UPPLY th A000RMht! WITH AN AprROVtD PRottDURE. THE TRAN5FER VA$ htCil$ARY TO PERMIT PLANht0 MAlhT! NANCE WDRK 04 THE RPS MG $tl$. ALTHOUGH THE CON 1 Rot RDDS WERE ALREADY FULLY INSER1tD, A SCRAM $1GNAL VA$ AUTOMATICALLY lh!11 Al!D DUR!hG TH( BUS TRAN$f[R. THE PROCEDURE PROVIDIS FOR TRAh$f(R OF TH( A RP$ POWER VHICH G[h(RATES A HALF $ CRAM DVI TO THE MDMthTARY LO$$ Or P0wtR. THis l$ FOLLOVID BY MANUAL Ri$ti Or THg A $1Dg xAlr 50R4u AND THth 1 rah $ FIR Or THC B RPS POWER F0tt0VtD BY MANUAL R[5ETTING OF THE RE$ULTAhi $ $10E HALF SCRAM. THE PRDCIDURI D1D ND1 PRDv!DE GUIDAhCf ON TH[ TIM [ lhitRVAL TO Rt ALLOW [D BITVE(N BU$ TRAh1 FIR, THE RI$ET Of THt A 51D[ b A$ COMPL(TED VITH!N FOUR 5[COND5 Or THE TRANSFER.
IN THE htxt FOUR $ECOND1 THE B $1DC WAS 1RANSFIRRtD AND A FULL $ CRAW ktLULTlD. THE Malt $11AM LINT RADIAT10h MDN! TOR POVERID FROM 1HE A RP1 BUS WA$ NOT Y!T $1AB!Lil[D FOLLOWlhG THE TRANSF[R AND INITIATID A FAISE HALF SCRAM $1GhAL LES$ THAN A T(NTH 0F A $tC0hD $tr0Rt THE B SIDE VAS TRAh$f(RRED. THE COWBlhATION OF SCRAM $1GhALS ON BDTH THE A AND B RPS CHAhN[t$ R[50 Lit 0 lh FULL SCRAM.
l
.FO. R M......17............... L [ R.. S. C. 5 5.. C. A1 A......................... C. S..- C. B.. 9.1 l
DOCLC1 YEAR LER NUMBER R[V!s!0N -DCS huMB(R hSIC (VINT DATE
. 3 33... 19 89...... 0.17........ 0..... 6 9102. $03. 3 D. 2.15 0..B 3..... 09. / 2 0 / 6 9 ABSTRACT POWER LEVEL - 00CL THI PLANT W $HUTDOWN FOR A SCHEDULID MAthTENAhCE OUTAG[.
11 WAS NECESSARV 10
[LIC1RICALLY 150 LATE A V! RING TERM!hAL STRIP TO R[ PLACE A TIRMINAL PLOCK. THC PROTECTIVI 1AGGlhG RIOU[$1 (PTR)
TO ACCOMPLISH THE ISOLATION WA$ INDfPCNDIh1LY REVlfVID BY AN EXIIRl[hCfD INSTRUM[h1 AND CONTROL f tCHNICIAN AND 1
TWO EAPIRIENC[D SENIOR LlCINSED OP[RATOR$. ON $[P1[ miter 20.1989 AT 2:05 A A A FUS[ VAS PULLID IN ACCOR0ANCE WITH THE PTR. THE [NGlht(RED $AFFTY FIATURI ACTUATION SYST[M (J[) th!T!ATED A GRDUP 11 PRIMARY C0hTAINMEhi
!$0LAtl0N $1GhAL Rf 5ULTIhG IN ISOLATION OF THE $HU1DOWN C00LlhG MDD[ OF THE Rt510UAL HEAT REMOVAL /LOV PR[$$URI COOLANT INJECTION ST5 TEM (80) AND THE REACTOR BUILDING VENTILATION $T5 TEM (VA). THE FUSE WAS REPLACED AhD THE SYSTCMS VERE RESTAR7[D VITHIN FIVE MINUT[$. TH( CAUSE WAS COGh!TIVE HJMAN [RRDR IN NOT FlhD!hG TH[ !$0LAT!0N SIGNAL CONh(CTION VHICH VAS PRIS[NT ON THE LOGIC DIAGRAMS AND VIRlu DRAVING 4 CONTRIBUTING CAUSE MAY BE THE AWr=ARD VIRING AND LOGl" CIRCULI DRAWlhG SYST[M THAT RIQUIRf$ MULTIPLE CROS$-REFERENCE $ TO MULitPLE DRAvlhG$.
THE RE$PONSIBLE OPERATORS RIvi[VED THE IVINT THE IMPORTANCE OF THDROUGH Pik RIVIEV WAS EMPHASIZED AhD INTERED IN THE NIGHT ORDERS FOR INSTRUCTION Or DTHER OPERATORS. L(R5 VITH RfLAf tp [L(MEh15:B9-013 87- 016. AND 86 019 FO.RM
........................ 0 8.- 0 6. - 9.1
...... 18...............L.IR 5C.5.$.D.ATA DOCKET YEAR LER NUMBER REVISION DCS huMBER h51C
[Vthi DATE l
. 3 33... 19. B9...... 0.18..........1.... 9. 00 2 2 2. 0 3 4 4... 2.16 8. 7 5.......10. / 09 /. 89 r
i AB$ TRACT POWER LEVEL - 014L A ROUTINE SURVilLLANCC TEST OF THE HIGH PRI55URI COOLANT INJECTION (HPCI)(BJ) SY5itM WAS IN PROGRES$ ON 10/08/89 AT 14 PERC[NT POWER DURING START-UP ATT[R A PLANh(D THREE-V[EK MAINTfhANCE OUTAGE. A1 10:26 A.M., A.HPCl HIGH $1EAM FLOW $1GNAL CLOSED THE HPC! OUTBOARD STEAM SUPPLY ISOLATION VALVES. OPERATORS VERIFIED THE AB5ENCE OF STEAM LEAKAGE. 1HE SURVEILLANCE 1ESTS REQUIRED TO BE PERFORMED VHEN HPCI 15 IhDPERABLE VERE INITIATED. INSPECTION OF THE DIFFERENT] AL PRES 5URE TRANSM]TT[R, VHICH PROVIDES THE HIGH SitAM FLOV $1GNAL, FOUND THE CAllBRATION VAS ACCURATE. DURlhG RECAtlBRATION A SMALL QUANTITY Or AIR WA$ OBSERV[D TO VENT FROM THE l
PRES $URE INSTRUMEhi $tN$lhG LIN[$. IN!TIALLY IT VA5 SCO @ ECYtY Bell [VED THAT PRESENCE OF NON CONDthSIBLE, BUT COMPRESSIBLE AIR IN THE $!NS!hG LINES. COMBINED VITH ThE FAST START TRANSl[hi. RESULTED Ik OSCILLATIONS AhD A FALSE HIGH STEAM FLOW $1GNAL, $UBS(QUthTLY (LER-89-025), IT VA$ O!5 COVERED THAT THE HIGH STIAM FLOV SIGNAL VAS VALID AND CAUSED BY USE UF A MORC CONSERVAfiv[ TEST PROCT 0VRE AND OVERLY CONS [Rv AllVE T[CHNICAL
$P[CIFICATION HIGH $1EAW FLOV AND FSAR ACTIVAT!0N TIM [ LIMITS.
UPON SATISFACTORY PERFORMANCE OF THE I
SURVE!LLANCE TEST, HPCI VAS RETURhED TO SCRVICE AT 6:30 P.M THE $ AWE CAY. LIR-89-025 AND LIR-89 002 ARE R[ LATED.
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FO. RM... 19................ L. E R S. C. S. S.. D. A T A.......................08.-08.-91 DOCKET TEAR LER NUMLER REV!510h DOS hUMBER hSIC EV!hi DATE
. 333... 1989..... 0.19. 4......
1.... 9.002220303 2.10874........10. /. 31/.89 ABSTRACT
}
POVER LEVEL - 100% Oh 10/31/89 A10625. THE HIGH PRE $$URE COOLANT thJEC110N (HPCl)(BJ) $Y$1[M WA$ DECLARED
{
INOPERABLE DUE TO AN ELECTRICAL GROUND IN ibE SPEED CONTROL circuli. BUILD UP OF C0hDUCTIVE CORROSlON PRODUCT 5 BETVEEN AN AMPHENDL TYPE C0hhECTOR IN AN Olt ENVIRONMENT AND 11$ M3UNT!hG PLATE WAS C061RIBUTED 10 BY N0!STURE IN THE Oll. A CHANGE IN THE C0hNECTOR M3UNT!N3 GROMMET C0hFIGURAT10N, AND lhADEQUATE SHRlhK TUBE th5ULA110N OVER THE CONDUCIDR CONNECTION 10 THE AMPHENOL Plh5. THE CONhECICR VAS CLEAhED AhD HPCI RESTORED TO SERVICE AT 2110.
ON 11/3/89 At 0300 HPCI VA5 DECLARED lh0PERABLE FOR THE SAME REASON. THE C0hNECTOR VAS REMOVED. CLEANED, AhD A $1LICONE RES!N COATING VA5 APPLIED 10 SEAL THE PINS. INSULAil0N. AND THE HOUNilNG PLATE. THE SHAri $EAL LEAK-OFF DRAIN PIPING VAS H0DIFIED TO IMPROVE THE M0l$fURE REM 3 VAL CAPABILITY OF THE GLAhD SEAL EXHAUSTER. TV3 UhEELATED SCRAMS DELAYED COMPLEt!ON OF $URVEllLANCE TESTING UNill 11/14/89 VHEN HPCI VA5 RETURNED TO SERVICE.
ON 12/5/89 HPCI VA$ DECLARED lh0PERABLE FOR AN ELECTRICAL GROUND. THE ACTUATOR VAS REPLACED AND HPCI RE510 RED 10 $tRVICL 12/12/89, ANALYSIS COMPLETED 2/5/90 CONCLUCED THAT HPCI VOULD HAVE PERFORMED ITS SAFETY FUNCTION AND THEREFORE 11 VAS N01. IN FACT, th0PERABLE DURING THOSE TIME thTERVAL$ VMEN THE FROUND VAS PRESENT. RELATED LERS: 89-005 AND 89-014.
)
F.O.RM..... 2 0...............L. E. R.. S. C. S. S.. D. A. T A...................... 08. 4.. 91 08-DOCKET YEAR LER NUMBER REVISION DC$ NUMBER N51C EVENT DATE
.. 33... 1989..... 020......... 1.... 9002080139.. 2.10678...... 11/05./.89 3
ABSTRACT POWER LEVEL - 100%. A REACTOR SCRAM OCCURRED FROW FULL POVER Al 3:23 P.M. ON 11/5/Bo AN UNIDENTIFIED FAILURE IN AN ELECTRONIC CONTROL CARD OF THE ELECTRO-HYDRAULIC CONTROL (EHC)(JJ) Sf51F" t0R Int "AIN TURBINE (TA) 15 BEllEVED 10 hAVE OPEhED THE BYPASS VALVES AND CLOSED THE lhTERCEPT AND C0 hip 06 VALVES. THld 'rDUCTION IN % TEAM FLOW CAUSED A PRES $URE TRAN11ENT RESULilhG IN A REACTOR HIGH FLUX SCP*a $1GNAL FROM THE AVERAu* POWER RANGE MONITOR (APRM)(IG). THE HIGH FRESSURE COOLANT INJECTION (HPCI)(BJ) !Y$ TEM WAS lh0PERABLE PRIOR 10 THE SCRAM.
4 THE AUTOMATIC FEATURES OF THE PLANT RESPONDED NORMAltY 10 THE SCRAM EXCEPT inAt 1 SAFETY RELIEF VALVE PASSED A $KALL AMOUNT OF STE AM AT A PRES $URE 5% BELDV ITS DESIGN LIF11NG PRESSURE. THE REACTOR CORE ISOLATION COOLIhG (RCIC)(BN) SYSTEM VAS USED 10 RESTORE REACTOR WATER LEVEL. 1 CONTROL ROD WAS hai FULLY lhSERTED. REQUIRING KANUAL INSERTION FROM POSITION 02. SELECTED ELECTR0h!C CONTROL CARDS VERE REPLACED IN THE EHC $1 STEM. THE PLANT VA1 RESTARTED 11/10/89 AND $ CRAMMED 11/12/89 (LER-69 023) FOR UNRELATED REA50h5. THE PLANT VA$ REETARTED 11/13/89 AND RUN AT 25% POWER TO CBSERVE THE EHC SYSTEH. 11 VAS SHUTD0VN 11/20/89 F3R FURTHER VORK ON THE EHC SYSTEM. FOLLOVING TESTING AND REPLACEMENT OF ELECTRONIC CIRCUlf BOARD $. THE PLANT WAS RESTARTED ON 11/22/89.
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....... 2.1...............L E.R S.CS $..D. AT A FORM
......................... 8-08-910.......
DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE
. 333....19 89...... 02.1........ 0..... 8912.13 03 2 0.. 2.1 E. 081..... 10./.3.1/.8 9 ABSTRACT POVER LEVEL - 100%. THE PLANT VAS OPERATING AT 100% POVER. THE HIGH PRESSURE CDOLANT INJECTION (HPCI) (BJ)
SYSTEM VAS INOPERABLE. FOLLOVlhG APPROVED PROCEDLSES.. THE REACTOR CORE ISOLA 110N COOLING (RCIC) (BN) LYSTEM VAS BEING TESTED FOR OPERABIL11Y ON 10/31/89. AT 12 N00N. DURlh3 THE PERFORMANCE OF THIS TEST. THE CIRCUlf BREAKER FOR THE RCIC INBOARD INJECTION VALVE TO THE FEE 0 VATER SYSTEM (AND ULTIMATELY THE REACIDR) TRIPPED DURlhG THE REQUIRED VALVE TESTING. THE ELECTRIC HOTOR FOR THE VALVE CPERATOR FAILED DUE TO A FAULT IN THE MOTOR VIND!hG INSULATION THE VALVE VA$ OBSERVED 10 HAVE CLOSED. THE COMBINED INOPERABILITY OF BOTH RCIC AND HPC! PLACED THE PLANT IN A 24 HOUR LIMITING CONDITION FOR OPERATION (LCO). THE MOTOR VAS REPLACED. TE511NG VAS COMPLETED, AND THE RCIC $Y$ TEM VAS DECLARED OPERABLE AT 6:29 P.M., EhD!hG THE LCO VITHIN ElGHT HOURS. TESilNG OF THE MOTOR CIRCUITS. VALVE ACTUATOR AND VALVE RULED OUT KNOVN POTEh11AL EXTERhAL CAUSES. THE MOTOR HAD BEEN IN SERVICE FOR 14 YEARS. THE CAUSE VAS FAILURE OF THE ELECTRICAL lhSULATION BELIEVED 10 HAVE BEEN CAUSED BY POOR QUALITY CONTROL-IN THE VINDING OF THE SERIES FIELD OVER THE SHUNT FIELD VIND]hG, RELATED LERS: 88-001.86-014, 86-003,85-025.
A B-7
- -.~_ - - -
- - - - - - - _ - - ~ -
_ _ _. _ _ ~
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.O.RM..... 2 2................ L E R.. S. C $ $. D. AT. A.......................08.-06..9.1 F
DOCKET YEAR LER NJMBER REvls!ON DCS huuBER N51C IVEhi CATE
. 33 3.. 1989...... 022......... 1.... 9 002 200210.. 2.1694 0......11/.0. 7/.89 AB$1RACT PDVER LEYtl - 000%.1Hl$ !$ REPORTED VOLUNTARILY AS AN [YINT VHICH MAY BE OF GEhtRIC lhitRE$1 CR CONCERN. OR CONTAIN5 A LE$$0h 10 BE LEARNED, ALTHOUGH WE D0 h0i BELitVf IT EXPLICITLY MEEi$ THE CRlitRIA C0h1 AlhtD IN 10 CFR 50.73(A). A REVIEV 0F THE $AFIT1-RELATED HYDRAULIC ShuBBER RECORDS VAS BEING PERFORMED FOR THE PURPOSE OF
!DENTIFY!hG THOSE $NUBB(R$ VHICH V0VLD B[ $CHEDULED FOR REBulLD!kG DURING THE hEXT RIFUILING OUTAGE. ON 11/7/89. 11 VAS DETERMlht0 THAT LACK OF MAlh1EhthCE R(CORDS MADE IT hECE55ARY 10 CONS (RVATIVELY ASSUME ThAT ELASTOWERIC $[ALS lh 33 $NUBBIR$ HAD hof BEIN REPLACED VlTHIN THE PLANT'S 7 YEAR SERvlCE tlFE GUIDELINE. A PLANT SPECIFlt REEVALUAi!ON OF $EAL* LIFE VA5 PERF0kMED. IT DEMONSTRATED THAT CONSERVATIVE VALUES OF l$ YEARS COULD BE APPLIED 10 * $ERV!CE LIFE FOR $NUBBER5 IN PRIKARY C0hTAlhMENT Ehv!RONMENTS AND 20 YEAR $ TO SNUBBERS lh REACTOR BUILD thYlRONMENTS. BECAU$E THE PLANT VAS 5HUTDOWN AT THE TIME OF DISCOVERY, 20 lhACCES$1BLE
$NUBbER$ VERE.4PLACED AND FUNCf10hALLY TESTED. N!htTIEN OF THE THESE PA55tD THE FUNCTIONAL TEST.
Ah EhGINEER!hG EVALUATION FOR THE Oh[ VHICH FAILID DEMONSTRATED THAT IT WAS FULLY CAPABLE OF PIRr0RMING 115 $ aft 1Y FUNCTION. THIRTEEN ACCESSIBLE $NUBBERS VERE Vl$UALLY lh5P(CitD AND SCHEDULED FOR RIBUILDlhG OR REPLACEMENT At THE REF"illhG OUTAGE SCHEDULED FOR 3/90.
FO.R.M.... 2 3................ L I. R. S. C. S. S.D. A.T A....................... 08..-08..9.1 DOCKET YEAR LER NUMBER REV1510h DC$ NUMBER N$1C
[ VENT DATE
. 333...1989...... 023......... 0.... 89.12200361. 2.10204....... 1.1/.12./. 8 9 ABSTRACT POWER LEVEL 010%. AT 5:34 P.M. ON 11/12/89. A REACTOR SCR AM OCCURRED DURING A SCHEDULED LURVEILLANCE TEST OF SAFETY RELIEF VALVES ($RV)(AD). Ih PREFARA110N FOR THE TIS 1, INDICATED REACTOR POVER VAS AT AFPROXlMATELY ID%
i AND PRESSURE VA$ BilNG C0hTROLLED AT 575 PSIG U$lhG THE MAlh BYPASS VALVES. THE FIRST SRV VAS OPEhED lh ACCORDANCE WITH PROCEDURE. UPON CLO$lhG THE VALVE, THE REACTOR SCRAMMED DUE 10 A PRES $URE TRANSIENT VHICH RESULTED IN A HIGH Flut 15% AVERAGE POVER rah 3[ M0hlTOR ( APRM)(IG) TRIP Ih THE ST ART UP/ HOT ST ANDBY MODE, THE APRM TRIP PO!NT !$ SET AT A NOMihAL 15%. ACTUAL TRIP P0thi SfiT! HGS RANGED FRDM 13.$% TO 14.5% 10 ALLDV FOR INSTRUMENT DRIFT. THl$ LEFT A MARGlh DF ONLY 3.5% BE1VEEh IkD!CATED REACTOR P0d[R OF 10% AND THE TRIP P0lki.
THE SMALL FRESSURE TRAh$1 TNT CAUSED Bf OPEN!hG AhD THEN CLOSING THE VALVE VAS SurFICIENT TO CAUSE A hEUTRON FLUX SPIKE AND SCRAM THE R[ ACTOR. TH[ $URVE!LLANCE TEST PROCEDURE DID NOT PROVIDE INSTRUCTION CONCERNING Ah APPROPRI ATE MARGIN, NOR DlD li PROVlDI A CAUTION TD THE OPERATOR CONCERNlhG 'HE P05$1BLE PRt15URE TRAN51ENT AND THE NEED FOR AN ADE0VATE MARGIN. CORRECTIVf ACTION REVlifD THE FROCEDURI TO FROVIDE FOR VALVE TE5ilhG AT 940 P51G VHEN THE REACTOR MODE SV!TLH VILL BE lh THE RUN HODE. Ih 1HE RUN MODE, THE FLOV BlASED SCRAM VILL BE lh EFFLCT. THf APRM HIGH FLUX 1 RIP P0thi 0F 12?% VILL DE lh SERVICE.
.FO.RM..... 2 4................ L..E R S. C. 5. 5. D. A. T A...................... 0 8. - 0 8.. 9.1 DOCKET YEAR LER NUMBfR REVISION DC$ NUMBER N$1C EVEhi DATE
. 333...1989...... 024........... 0.... 9 00109008.1. 2.10299.......11/.29./.89 ABSTRACT POVER LEVEL - 100%. THE REACTOR CORE ISOLATION C00LlhG (RCIC)(BN) 5Y$1[M VAS REN0ERID INDPERABLE FOR 14 MINS AT 1:06 P.M. ON 11/29/89 VITH THL REACIOR OPERAI!NG AT 1001 POWER A MONTHLY SURVEILLAhCE TEST REQUIRED BY TECH SPECS IN TABLE 4.2 2 VAS BEING PERFORMED BY AN EXPERIENCED TECHNICI AN, THE TEST FUNCTIONALLY CHECKS PRESSURE SENSORS ON THE RCIC TURLINE EXHAUS1 L!ht DOWNSTREAM FROM 1HE OVERPRI$$URE RUPTURE DIAPHRAGMS. HIGH DIAPMRAGM EXHAUST PRESSURE SthD5 A CLOLE $1CNAL TO PRIMARY CONTAlhMENT 150LAT10N VALVE 5 ON THE RCIC STEAM SUPPLY LihE.
VORKlhG FROM THE PROCEDURE PREREQU111TE sitT10N. THC circuli BREAKER FOR THE lhBOARD ISOLATION VALVI (LISTED FIRST) VAS OPEhED IN THE MORNING 50 THAT TESilNG V0VLD NOT CLD5I THE VALVE AND 150 LATE RCIC. AFTER A LUNCH BREAK THE TEST VAS CONilhUED U$1NG THE FERFORMANCE SECT 10h VHICH itSTS THE INSTRUMEhiS FOR THE OUTBOARD VALVE FIRST. THE TECHh!CIAN $1GNED OFF A VALVE CIRCUlf BRIAKER VERIFICATION STEP ASSUMING THAT IT VAS FOR THE BREAKER FOR THE VALVE HE HAL WEQUESTED TO Bt OPEN BEFORE LUNCH.
HOVfVER, THE $1EP VAS FOR THE OUTBOARD VALVE.
PERFORMING THE TEST CLOSED THE OUTBOARD VALVC BECAUSE THE t!RCUli BREAKER HAD h0i BEEN OPIhED. THE VALVE CLOSING VAS h0TED BY OPERA 10RS V40 REOPENED IT W11HlN 14 MthuTES. CORRECTIVE ACTION INCLUDED DISCUS $10N V!1H TECHNICIANS OF THE EVEh1 AND VAYS TO PREVENT REOCCURRENCE.
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FO.R.M.... 2 5............... L. E. R S. C. 5 5.. C. A T A...................... 0 8.- 0 8.- 91 i
DOCKET YtAR LIR NUMBER REV1510h DCS hpMBER N51C
[ VENT DATE l
.. 3 3....19 8 9...... 02 5........ 0..... 9. 001110.18 4... 2164 00....... ! ! /.3 0. /. 8 9 3
i ABSTRACT POVtR LEVIL - 100%. AT 4:40 A.M. ON 11/30/29 VITH THE REACTOR AT FULL PC.[R, TH[ HPCI (BJ) SY57tM 150L ATED ON A HIGH SitAM FLOV $1GNAL DURlhG A SURytlLLANCE TEST lh!TtATlhG A 7 DAY LCO. AN EXTENSIVE TESTING PROGRAM VAS l
INITIAf tD 10 IDENTIFY THE CAust. HPCI WAS TESTED AND RETURh!D 10 5tRVICE EhDlhG THE LCO 0412/3/89 AT 1:05 P.M.
HPCI VA5 REMOVED FRDM 5tRVICE FOR MAthithANCE AND FURTHER TESTING AT 6:10 A.M. ON 12/5/89, INITI ATING A SECDND l
7-DAY LCO.
17 FAlltD TO MEET OPERABILITY TESTIN3 CRITERIA DURIN3 $URVE!LLANCE TE5ilhG AT 10:00 P.M.
BY REQUlRING MORE THAN 25 SECS TO ACHlIVI FULL FLOV. ADDITIONAL C ECORDIN3 th5TRUM[his VERE C0NhfCTED, VINDOR FitLD I
IN3!hEERS VERE OBTAlhED AhD EXTth!!VE TESTIh3 VAS FERFOR'OS THROUGH 12/11/89 VH[N HPCI VAS DECLARED OrERABLE AT 1:31 P.M.
THE hRC GRANTED A TEMPORARY VAlVIR OF THE SETP0thi REQUIREMENTS FOR THE HIGH STEAM FLOV !$0LATION PRIOR TO EXP! RATION OF THE 7 DAY LCO AND APPROVED A TECH SPEC AM[hDMENI thCREA51N3 TH[ HIGH STEAM FLOV DIFFERENTI AL PRESSURE SETP0thT. PORC APPROVED INCREA51NS THE FSAR DESIGN BA515 RE0VIRED ACTUATION TIMI FOR HPCl FROM 25 TO 30 SICS. A 1981 YENDOR RECOMMINDATION VAS IMPLEMENTED INCR[A5]hG TH[ TURB!ht START *UP RAMP TIME FROM 9 TO 15 SEC5. CAUSES INCLUDED AN CVERLY CONSEEVATIVE TECH SPEC LIMIT AND FSAR BA515 FOR MAX 1 MUM ACTUATION TIME AND IMPLEMENTATION OF A HORE C045tRVATIVC PROCEDURE.
F.O. RM..... 2 6................ L. E R S. C. 5 5.. D. A T A...................... 0. B..+ 0 8.- 91 DOCKET TIAR LER hUMBER REV1510N DC5 NUMBER h51C EvfN1 DATE
. 333... 1989.... 026......... 1.... 90071202.E6.. 2.18783..... 12./.26./.89 i
ABSTRACT l
POWER LEVEL - 100%. FOLLOVING A REACTOR SCRAM ON 11/5/89. TVO SAFETY R(LitF VALVES (SRv5)(AD) VERI REMOVED FOR I
TESTING.- THE VALVE TEST FACIL]TY FROVIDED VRITTEN NOTICE. RECElVED BY THE AUTHORiiY ON 12/20/89 TkAT BOTH l
YALVES ACTUATED AT SETPOINTS VHICH DEVIATED FROM THE DE51GN PolkT BY MOR[ THAN THE +/- 11 ALLOVED BY TECH SPECS, SRV *E" LIFTED EARLY AT 1 4%.
SRV "F" LIFTED AT.4.7%.
Dl5A55(MBLY AND EXAMINATION Or THE PILOT MECHANISMS FOUhD STEAM CUTS ON THE PILOT VALVE 015C SEAT AND BEtt0V5 FOR $RV *F" DISC 10 SEAT BONDlhG 15 BElltVED TO BE THE CAUSE OF THE HIGH INITI AL LIFT PRtSSURE FOR $RY "F".
NO DETIRIORATION VAS NOTED ON $RV *E" COwPONINTS AND NO CAUSE FOR EARLY LIFilNG OF THAT VALVE VAS DETERMINED. EVALUATION OF REACTOR FRC55JRE RElltF CAPABILITY $HOV5 j
OPERATION VOULD BE ACCEPTABLE WITH 2 0F 11 SRv5 JNDPERABLE AND A SETPOINT TOLERANCC Or +/-3%.
CORRECTIVE ACTIONS INCLUDED REPLAClhG FAILED SRV5 WITH RECERTIFitD VALVES. CONTINUED PARTICIPATION th THE BVR OWNER'S GROUP TO RESOLVE SRV 155UES, AND SUBMISSION TO THE hRC 0F PRDPOSED CHAN3t$ TO TECH $PECS TO TAKE CREDIT FOR EXCESS INSTALLED SRV CAPACITY, LER-85 009.85-013, 87-004,88-004, AND B8-010 ARE SIMILAR tythis thv0LVihG SRV
$[TPolNT DRIFT.
....... 2. 7.............. L..E R S. C. 5 5.D A T A..................................1 FORM 08 08-9 DOCKET YIAR LER NUMBER REVISION DCS NUMBER h51C EVIh1 CATE
-. 3 3 3... 199 0...... 00.1........
0..... 90022 004 4 8.. 2.16910.... 0.1/.19./.9. 0 ABSTRACT POVfR LEVEL - 100%. A REACTOR SCRAM FROH FULL POWER OCCURRED AT 10:45 A.M. Oh JANUARY 19, 1990 DURlhG THE CAllERATION OF REACTOR VATER LEVEL INSTRUHEhTATION. THE lh57RUMEhi VHICH VAS BElhG CALIBRATED SKARES COMMON REFERENCE AhD VARIABLE LEGS vlTH Ih5TRUMEhis OF THE REACTOR FROTECTION SYSTEM (RPS) (JE). VHlLE ISOLATING THE INSTRUMENT UNDER TE$T, A VALVE PACKIN3 LEAK DEVELOPED. DURING THE RESPONSE TO TH15 LEAK, VALVE ACTUAT]CN CAUSED A FALCE LOV VATER t.EVEL TO BE Sth5ED BY THE RPS. THl3 FALSE LOV LEVEL TRANSIENT VAS CAUSED BY RAPID VALVE MOVEMENT BY THE FIELD TECHNICIAN PERFORMING THE VALVE MANIPULATIONS. CORRECTIVE ACTIONS INCLUDE REVIEV 0F THl$
EVENT VITH ALL 160 TECHNICI Ah5, GR00MlhG OF THE COMMUNICATION SYSTEM TO EllMihATE N015(, AND REPAIR OF EQUIPMENT THAT MALFUNCTIONED DURING THE TRAN5f fhi. RELATED LIRS: 85-012 AND 87-020.
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. 333....1990..... 002......... 0..... 9 0022 6054 5 2.109 51..... 0.1/.2. C/.90 ABSTRACT POWER LEVEL - 000%. At 6:24 A.M. ON 1/20/90. VMILE TH! RIACTOR VAS MOT $HUTDOWN FOLLOW!hG A SCRAM (LER-90-001).
THE REACTOR HIGH PRIS$URE !$0LAT10N LOGIC TRIPPED AND ISOLATED RE510UAL HEAT REMOVA! (RHR)(bO)
- B" 5HUTDOWN i
C00LlhG (SDC) SYSTEM AS THE SYSTEM VA5 [ElhG STARitD. THIS LD0!C 15 $[I TO TRIP AT LESS THAN OR EQUAL 10 75 F51G TO PROTECT AND ISOLATI THE L6V FRES$URE P!P!NG OF THE 5DC fROM THE REACTOR FRE55URE. AT THE fl1C OF THE TRIP THE REACTOR PRES $URE VA5 27,5 PS!G. VHICH 15 LtSS THAN OhE HALF OF THE TRIP SETrolhi. IhvESTIGAfiON FOUND THAT THE TRIP LOGIC PRES $URI SEh50R5 VERI PROFERLY CALIERATED AND $[T AT APFROXIMATELY 60 PSIG (OUlV ALENT PE ACTOR FRE55URE AND 1KAT THE REACTOR PRES 5URE READlh3 0F 27.5 F51G VA5 ACCURATE. h3 DEFINiilVI CAUSE VAS FOUND. A HYDR *VLIC PRESSURE TRAN51[hi 15 $U5PICTED AS THE CAUSE. THE OPERAT!h3 FROCEDURE VA5 REV15ED TD RECOMMihD START!hG THE SDC AT LES$ THAN 10 PSIG kATHER THAh LESS THAN 75 PS!G REACTOR FRI55URI.
I
...... 29................ L. E R S. C. S $.. D. AT A...................... 08.. 0 B - 9.1 FO.RM DOCKET YEAR LER NUMBER RIVISION DCS NUMBER h51C EVENT DATE
. 333....1990..... 003..........1.... 9. 01004 014 6. 219020.... 01/.29./9 0
...... +....
AB51RACT POVER LEvtl - 10D%, TVO h0N-CONSERVATIVE [RROR5 IN FEED (SJ) FLOW MEASURIMENT VERI 015 COVERED DURlh3 A Slit REVIEV BIGUN lh OCTOBER 1989. FEED FLOV TRAN5MIT1ER5 VERE REPLAEED Ch 10/3/68 VITH A MODEL REQUIRING STATIC r
PRIS$URE COMPENSAi!ON. THE COWDIh5ATION VAS OMITit0 DUE TO AN INCOMPLETE REVIEV 0F THE YENDOR MANUAL. THE CALIBRATION VAS CORRECTED AND FERFCRMED ON 11/14/89. SECONDLY, THE REVl[V ALSO RAISED QUE5110NS ABOUT A VEhDOR SUPPLIED CORRELATION FOR THE FIED FLOV VthTUR15. A C0h5tRVATIVI LICEh5EE CORRILAfl0N VA5 USED TO RICALIBRATE THE TRANSMITTERS AND THE RESULilNG lhDICATED POWER EXCIEDED LICEh5ED THERMAL P0i[R. POWER VA5 FRDuPTLY REDUCED TO THE LICEN5ED VALUE. FLOV ELEMENT VENDOR lhPUT [RRQR$ HAVE $lhCE Bt[h 10thilFIED AhD CORRECTlD. ADDITIONAL CORRECTIVE ACTIONS VILL lhCLUDE ADDlil0NAL TRAIN 1hs, MODIFICATION CONTROL FRDCEDURE CHANSES, lhCRIA$ED i
CALIBRATION FREQUEhCY, EhHANCED PERFORMANCE MON 11CRlh3 0F THE FELD ILOV, AND EhMANCED CALIBRATIDN FROCEDURES, RELATED LERS: 82-002 AND 82-034 FO.RM
........................ 08..-08-9.1
...... 3. D.......... * * *.. L E. R.S C. S. S.. D. A T A 5
DDCK!T YEAR LER NUMBER REVISION DCS hUMBER N51C EVtNT DATE 0
9003220137 217432...... 02./.07/.9.D
. 333....1990..... 004 ABSTRACT PDVER LEVIL - 100%, Oh 2/7/90 AT 012D VITH THE REACTOR OPIRAilhG AT FULL F0VER, AN ISOLATION OF THE REACT 04 CORE 150LAT10N COOLING (RCIC) $YSTEM (BN) DCCURRID. THis th!11ATED A 7-DAY LCO. THE 150LAT10h VA5 CAUSED E1 THE FAILURE OF A MASTER TRIP UNIT (MTU) VHICH GEhERATED A HlGM ARIA TEMPERATURE ISOLATION SIGNAL. THE MTU VA5 REPLACED, RCIC VA5 DECLARED TO BE OFERABLE AT 1820 thDlha THE LCO APPROX!MATILY 17 HOURS AFTER 11 STARTED.
RELATED LER$: 85-028,86-005, 86-015,87-013.
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F O. RM.. 3.1............... L. E R.. S. C. S. S.. D. AT. A......................08.-08..9.1 DOCKET YEAR LER NLMBER REvlS10h DCS huMBER hSIC EVENT DATE
. 333...1990..... 005........ 1.... 9.1030.102.55. 221130..... 02./.20/90 ADiTMACT POVER LEVEL - 100%. A REGULARLY SCHEDL' LED QUARTERLY SURVE!LLANCE TEST FOR THE HIGH PRES $URE COOLANT INJECTION SYSTEM (HPC1) (EJ) WAS PERFORMED ON 02/20/90 WITH THE REACTOR OPERATING AT FULL POVER. TEST DATA lhDICATED THAT THE 25.3 SECOND TIME INTERVAL ALLOVED BY THE lhSERVICE TESilhG (IST) PROGRAM FOR THE OPENING OF THE HPCl TURB!NE STOP VALVE (23HOV 1) VAS EXCEEDED BY 1.75 SECONDS (7%). HPCI WAS CONSERVATIVELY DECLARED TO BE lh0FERABLE AT 1:20 P.M. Ihl11ATING A 7 DAY LIM 111N3 C0h0lT10N FOR OPF. RATION (LCO). DURING THE hEXT S DAYS OF TROUBLESHOOTING AND TESilhG HPCI WAS SUBJECTED 10 6 TEST RUNS. THE $10P VALVE VHlCH FAILED THE !$1 CRITERIA l$ HYORAUllCALLY CPERATED BY THE TURBihE OIL, DURING THE SEARCH FOR THE CAUSE OF THE SLOW RESPONSE TIME. THF MYORAUllt OIL SYSTEM FILTERS AND THE KAIN $ HAFT DRIVEN Dil PUMP WERE REPLACED. THE $[ ARCH NARROWED TO A PIP 1hG DEAD LEG ASSOCIATED VITH THE CONTROLLING SERVO MECHANISM VMICH VAS FOUND TO C0hTAIN METAL FILINGS. DIRT. AND SMALL PAINT CHIP PARTICLES. AFTER CLEANING THE SERVO MECKANISM, HPCI MET THE IST CRITERIA AND WAS DECLARED OFERABLE Af 2:05 P.M.
ON 02/25/90. HPCI WAS TECHNICALLY INDPERABLE FOR S DAYS, 45 MINUTES. CORRECTIVE ACT10N INCLUDES ADDING THE REMOTE SERVO MECHANISM TO THE PM PROGRAM, TESi!NG Or THE HYDRAULIC 0!L AhD CHECKING OF Olt FILTER DIFFERENT!AL PRESSURE.
.. O. R M.... 3 2............... L. E. R. S. C. S. S. D. A. T. A.......................08..-08.9.1 F.
DDCKET YEAR LER NUMBER REVISION OCS NUMBER NS!C EVENT DATE
. 3 3 3...1990..... 0. 06..........1.... 9. 00 7 09 01 E. 2.. 2.18 713...... 0 3 /.12../ 9 0 ABSTRACT POVER LEVEL - 100%. ON 3/12/90 IT WAS DETERMlNED THAT ON 3/6/90 AT APPROXIMATELY 1600 HOURS DURlh3100% STEADY STATE POWER OPERATION A RADIATION PROTECTION TECHNICIAN (RPT) RECEIVED A SIGNIFICANT EXTREMITY DOSE FROM hA-24 CONTAMlhAil0N OF THE LEFT KAND THUMB. THE RPT WAS RAD 10 LOGICALLY MONITORihG WORKERS FERFORMING A FEEDWATER FLOW MEASUREMENT TEST. FOLLOWING THE TRANSFER Or THE 240 hcl RAD 10 TRACER SOLUT10N. THE RPT lhAPPROFRIATELY PICKED UP THE HIGHLY CONTAMIhATED VIAL CAP Vl1H THE RIGHT HAhD AND PLACED IT ON THE EMPTY VI AL, CONTAMINATION OF THE LCFT HAND THUMB OCCURRED VHILE REMOVlhG THE SURGEON GLOVE 5 UPON EX111NG THE VORK AREA. THUMB DOSE WAS CALCULATED AT A DEPTH OF 64 MG/CM TH15 AND THE PREVIOUS CURRENT QUARTER EXTREMITY DOSE WAS 17.33 REM.
THE OVARTERLY OCCUPAf tokAL EXTREMITY DOSE LIMIT IN 10 CFR 20.101 IS 18.75 REM. A 49.03 REM DOSE VOULD BE CALCULATED AT A DEP1H OF 7 MG/EM USING GENERIC NRC RECOMMEkCATIONS; H0VIVER IT VOULD BE INAPPROFRIATE IN TH15 lhSTANCE DUE TO THE PRESENCE OF A HEAVY KERATIN!2ED (1.E., CALLUS) LAYER ON THE RPT'S THUMB. DEFICIENCIES INCLUDED FAILURE TO ADEQUATELY ASSESS A POTENTIAL HAZARD AND PERFORM ONLY THE ASSIGNED TASK. COPRECilVI ACTIONS INCLUCE UPGRADlhG THE ALA8a REVIEW PROCESS, REVIEW RPT TRAIN 1hG, AND COUNSEllhG, INSPECTION REPORT $0-333/90-1215 RELATED. LER 50 333/87 002 DESCRIBES A PREVIOUS EXTREMITY EXPOSURE lhCIDENT, FO. RM.... 33...............L. E R. S. C. S S..D. A. T. A......................08..08.-9.1
+
DOCKET YEAR LER NUMBER REVISION DCS NUMBER NS!C EVENT DATE
. 333... 1990......007......... 0..... 9004110241. 2.17589........ 03/.02/.90 ABSTRACT POWER LEVEL - 10C%. CURING NORMAL FULL POWER OPERA 110N AT 1510 HOURS ON 3/2/90 li WAS DISCOVERED THAT. DUE TO
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A PRDCEDURE DEFICIENCY, A PORTION OF THE HAlh STEAM ISOLATION VALVE (MSIV)(SB) FOSITION SVITCH CONTACTS WERE l
NOT SHOWN 10 OPERATE DURING MSIV PAR 11 AL CLOSURE. THE SWITCH CONTACTS PROVICE INPUT TO THE REACTOR PROTECil0N SYSTEM (JC) 10 INITIATE A SCRAM IN THE EVENT OF MSly CLOSURE DURING OPERAT10N IN THE "RUN" MODE. THE PROCEDURE i
WAS REVISED AND COMPLETED BY 1620 HOURS ON 3/2/90. THE EVENT WAS NOT 4FEYY SIGhlFICANT BECAUSE HULTIPLE UNDETECTED FAILURES WOULD BE REQUIRED AND BECAUSE THE FlhAL SAFETY AhAl $15 REPORT SPECIFICALLY CONSIDERED FAILURE OF MSlv CLOSURE 10 CAUSE A REACTOR SCRAM. CORRECTIVE ACTION IND UDED REVISl_0N OF THE PROCEDURE TO l
CORRECT THE DEFICIENCY-AhD REVIEV 0F OTHER PROCEDURES TO VERIFY ThAT THE DEFIC!tscy WAS AN ISOLATED EVENT.
LER 89 008 15 A RELATED EVENT INVolvihG A PROCEDURE DEFICIENCY T4AT RESULTED IN A MittED SURVElLLANCE.
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FO.R.M.... 34................ L. E. R s. C. 5 5..D. A T A........................
- 08. 08.-9.1 DOCKET YEAR LER NUMBER REVISION DCS NUMBER N$1C EVENT DATE
. 333... 1990...... 008........ 0..... 9.00424 02 4 3. 2.17 808...... 03./.15./.9 0 ABSTRACT POWER LEVEL - 10?%. AT FULL POWER DN 3/15/90. THE REACTOR COOLANT LEAK RATE VAS CALCULATED FOR AN INTERVAL OF 7 HOURS INSTEAD OF 4 HOURS AS REQUIRED BY TECHNICAL $P[CIFICATION 4.6.D.
THE OPERATOR (NCO) AT THE TkDNT CONTROL PANEL ACTIVATED THE PRIMARY CONTAINwENT (HN) DRAIN COLLECTION $ UMP PUMP AT 1000. THE AU11LI ARY OPERATOR (AO) DID NOT READ THE BACK PANEL PUMP OUT FLOW INTEGRATOR $ IMMEDIATELY F0Lt0VING 1HE COMPLETION OF THE SUMP PUMPDOVN$. TH15 DNIS$10N VAS DISCOVERED AT 1900 BY ANOTHER AUXILI ARY OPERATOR AND CORRECTED. THE A0 HAD TRANSFERRED FROM ANOTHER $HIFT TEAM AND EXPECTED TO BE TOLD WHtN 10 READ THE METERS. THE NCO (VH0 HAD CHANGED DUTIES WITH THE $ENIOR NUCLEAR OPERATOR (!NO) VITHIN THE $HlFT) A$$UMED THE A0 VOULD READ THE METER 5 VITHOUT FURTHER INSTRUCT 10N AS VAS CUSTOMARY ON TH11 $HIFT TEAM. THl$ EVENT !$ NOT CONSIDERED SIGNIFICANT TO PLANT SAFETY. A REACTOR COOLANT LEAK IN EICESS OF TECHNICAL $PECIF!tATION LIMIT $ l$ DI$ PLAYED AND ALARMED ON THE FRONT PANFL$ VHEN THE $ UMP PUMP RESTARTS IN LE$$ THAN B0 MINUTES FOLLOVING THE PREVIOU$ PUMP-0UT (15 MINUTES FOR THE IDENTIFlED EQUlPMENT ORA!N $ UMP) OR VMEN EXCESSIVE TIME 15 REQUIRED 10 PUMP OUT THE SUMP. TH11 P. 3 TIME 5 FASTER THAN THE 4-HOUR MANUAL CALCULATION INTERVAL. PRDCEDURES KAVE BEEN $TANDAR0llED BETVEEN SHIFT CREV5 AND A $1NCLE PERSON HA5 BEEN A$$1GNED RESPONSIBILITY YO PERFORM ALL PARTS OF 1HE TASK. RELATt0 LER$: 88-006 AND 87-022.
FO.R.M.... 3 5...............L. E. R 5.C.$. 5.D. AT A.........................
8..-09-91 0
DOCKIT YLAR LER NUMBER REVl$10N DCS NUMBER NSit EVENT DATE
. 333... 1990...... 009......... 0..... 9004 24 026.1. 217 809...... 03./.19./.90 ABSTRACT POWER LEVEL - 1001. THE REACTOR SCRAMMED FROM FULL PDVER AT 3:54 P.M. ON 3/19/90. A FALSE LOV PEACTOR WATER LEVEL SIGNAL IN THE TEEDWATER CONTROL CIRCU!T (JB) 'NCREA$[D FIED PUMP SPEED. THl$ RESULTED IN HIGH RLACTOR WATER LEVEL FOLLDVED BY A TURBINE TRIP AND AUTOMATIC REACTOR SCRAM IN!TIATED BY TURBINE STOP VALVE CLOSURE (58).
ALL SYSTEMS PERFORMED AS DESIGNED EXCEPT FOR FLOV OSCILLATIONS IN THE HIGH PRFSSURE COOLANT INJECTION SYSTEM (HPCI) (BJ) (LER-90-010) AND ONE R00 IN$ERTING ONLY TO P051110N (C2). HPCI INJECT 10N WAS CONTINUED U$1NG MANJAL OPERATOR CONTROL. A NORMAL PLANT C00LD0VN F0tt0VED. THE CAUSE Or THE $ CRAM VAS AN OVERSIGHT BY THE VENDOR IN THE ASSEMELY OF THE AMPLIFIER COMPONENT (PURCHASED 10 YEARS AGO) 0F THE STRIP CHART RECORDER FOR NARROV RANGE WATER LEVEL. A ONE-QUARTER INCH METAL STAR LOCK VASHER BAS FOUND WEDGED BETVEEN THE PRINTED circuli BOARD AND THE MOUNTING TRACK GROOVE IN THE METAL AMPLIFIER BOX. THE VASHER WAS ALSO IN CONTACT WITH A SIGNAL VlRE ON THE BOARD. THE RESULTING ELECTRICAL GROUND CREATED A SHORT CIRCUlf TO AN INSTALLED GROUND ON THE FEEDVATER LEVEL CONTROL CIRCUIT AND A FALSE LOW REACTOR VATER LEVEL $1GNAL. THE PLANT UA5 RETURNED TO SERYlCE VITH THE CENERATOR 04 LINE AT 8:27 P.M. ON MARCH 23. 1990.
FO.R.M.... 3 6................L. E R..S.C. S. $..C. AT A.....................08..- 08. - 3.1 DOCKET YEAR LER NUMBER REVISION DCS NUMBER N51C EVENT DATE
. 333.. 195 0..... 0.10......... 0..... 90042 7 0278.. 2.18020.... < 03./.19/.9 0 1
ABSTRACT POWER LEVEL - 100%. DURING A REACTOR SCRAM (LER-90-009) FROM FULL POWER AT 1554 DN 3/19/93 HPCl (83) CORRECTLY RESPONDED TO LOV REACTOR VATER LEVEL BT AUTOMATIC START AND INJECTION INTO THE REACTOR VESSEL. UNSTABLE OPERATION A$ INDICATED BY FLOV RATE OSCILLATIONS PRDMPTED OPERATOR TO TAKE MANUAL CONTROL TO RESTORE REACTOR VATER LEVEL. 4 DAYS OF TESTING DID h0T REPRODUCE OSCILLATIONS AT 150 PSIG REACTOR PRESSURE. OSCILLATIONS WERE REPLICATED AT NORMAL OPERATING FRES$URE OF 950 PSIG. HPCI HAD BEEN LATENSIVELY TESTED AND CONTROL CIRCUIT 5 ADJUSTED IN 12/89 ELER-89-025).
DURING 2/89 hPCI HYDRAULIC $ERVO VAS CLEANED (LER-90 005) RESULTING IN IMPROVEMENT IN RESPONSE TIMES COMPARED TO 12/89. WHILE NOT NOTICED AT UtAT TIME. IMPROVED RESPONSE TIME REDUCED THE CONTF0L SYSTEM DAMP!NG AND ABILITY OF THE SYSTEM TO RESPOND 10 $1GNIFICANT FLOV $1GNAL FLUCTUATIONS.
EXISTING POST WuRK TEST PROCEDURES FAILED TO REVEAL THis CONDIT10N. THE NEED TO INCREASE DAMPlNG BY READJUSTING A HYDRAUllt SPEED CONTROL LOOP NEEDLE VALVE VAS NOT IDENTIFIED AND RESULTED IN UNSTABLE OPERATION OF HPCI IN ThE AUTOMATIC MODE. THE NEEDLE VALVE VAS PROPERLY ADJUSTED. POST-WORK TEST PROCEDURES VILL BE REVISED. A Futt FLDV INJECTION TEST TO THE VESSEL WILL BE PERFORMED DURING START-UP FROM NEXT REFUELING OUTAGE. RELATED LERS:
90-009,90-005, 89-025.89-018. AND 89 002.
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4 F. O. R.M3. 7............... L. E R.. S. C. 5 5.. C. A. T A...................... 0 8.. 08. - 9.1 DDCEET YEAR LER huMPER RIVISION DCS huMBER h51C Evthi DATE
. 333.. 1990..... 0.11........ 0..... 9.00427C.234. 2.18825....... 03 /2 0. /.9 0 ABSTRACT POWER LEvil - 000%. At 9:20 A.M.
ON 3/20/90. WHILt THE REACTOR hAS th HOT 5HU1DOWN 70LLcWlhG A SCRAM (LfR 90-009), THE REACTOR HIGH PRIS$URE 150LAi!0k LD01C TRIPPED AND 150LATID RE510UAL HEAT REMDVAL (RHR) (DO)
- B" 5HUIDOWN CDOLING ($DC) SYSTEM AS 11 WAS BElhG STARTED. THl1 LOGlc 15 SET TO TRIP AT LESS THAN OR EQUAL 10 75 PSIG TO PROTECT AND ISOLATE THE LOW PRf 55URE P!P!hG OF TMt SDC FRDM HIGH REACTOR PRESSURE. AT THE TIMI Or THE TRIP THE REACTOR PRESSURE WAS 6 PSIG, WHICH !$ LESS 1kAN ONE TINTH OF THE TRIP $E1P01hi. IhvISTIGAi!0N FOUND THAT THE TRIP LDGIC PRESSURE SENSORS WERE PROPERLY Call 2 RATED AND SET AT APPROX 1kATELY 60 PSIG EQU! VALENT I
REACTOR PRI55VRE AND THAT THE RIAC10R PRES $URI READlh3 0F 6 P51G WAS ACCURAf t.
RHR PUMP "D" DISCnARGES lh10 A CDHMON HE ADIR WITH REACTOR WATER REClRCULATION (RWR) (AD) PUMP *B" OPEhlhG THE $DC INJECTION VALYE FOLLOVING RHR PUMP START lhlilATED A HYORAULIC PRESSURE TRAN51[hi VHICH WAS TRAN5N11TLD IN A REVERSE FLOW DIRECil0N
. THROUGH RWR PUNP "B" 10 THE !$0LATION LOGIC FRE15URE SWITCHES LDCATED IN THE SUCTION Llh[ FFDM THE REAC10R 10 THE RVR PUMP. THE PRES $URI 1RAN51ENT WAS SUFFICIENT TO TRIP THE FRESSURE Swif tHis AND ISOLATE THE SYSTEM, THE OPERAi!NG PRDCEDURE WAS REVISED 10 STOP AND 150 LATE THE REACTOR WATER REClsCULATION PUMPS PRIOR 10 STARilhG THE RHR PUMP, LtR-90 002 REPOR15 AN ALM 357 IDIhilCAL EVthi.
........................ 0.B.-08-91
...... 3 8................ L. E R. S. C. S. S.. D. A T A DOCKET YtAR LER NUMBER RIVISION DCS huMBER N51C IVthi DATE
. 3 3 3... 19 9. 0..... 0.12..........1.... 9 009 0 500.12.. 219 4 53............/04/.90 04 ABSTRACT POWER LEVEL 000%. DURlhG THE 1990 REFUEL OUTAGE, 61 CHECK VALVE 5 IN THE h0RMAL (N5W) (KG) AND [MfRGENCY SERVICE VATER (ESV) (BI) VERE OPINED AND VISUALLY INSPICTED. THIRTY-5tVfN OF THE VALVES VERE FROM THI ASNE SECTION X1 IST PROGRAM AND 24 WERE FRDM THE CHICK VALVE PREVEhilVE KAlNTENANCE (FM) PROGRAM. TVINTY JST AhD 10 PM PROGRAM VALVE 5 WERE IN!f! ALLY DECLARID INOPERABit DUE 10 FAILING 1HE V!5UAL lh5PEC110N CRl1ERI A.
H0VIVER.
THE 151 CHECK VALVES VERE SHOWN 10 BE OPERABLE BY ACTUAL FLOW 1EST OR CALCULAi!ON. OTHER EFFORT $ lhCLUDED i
INTERhAL INSPECTION OF 50D' 0F SMALL BORE PlP!NG AND 10 SAFETY RELATED CDOLERS AND AIR HANDLING UNITS (AHU5).
OF THESE CDOLIRS AND AHUS, 2 WERE FOUND 10 KAVE 25% lube PLUGGING VITH Silf/ SAND, BUT 5HOWN ABLE TO REMovt DE51GN BA515 HEAT LOAD. OF 1HE 500' 0F PIP!hG, 200' VERE FOUND 101+30% RESTRILTID IN CROSS-SECTIONAL AREA But A CALCULAi!ON DEMONSTRATED THAT FLOW C0h1ROL VALVES WERE HYDRAut!CALLY LIM!ithG. IN ALL. THE ESW SYSTEM WAS CONSIDERED CAPABLE OF PERr0RuthG THE DESIGN SAFETY FUNCTION, THE AFFECTED VALVES. CDOLIRS, AND PIP!hG VERE CLEANED CR REPLACED A5 NECESSARY AhD RETURNED 10 SERVICE. THE lhTAKE BAYS VERE AL50 CLEANED. PER1001C FLU 5HlhG AND PERFORMANCE itsflhG $HOULD PREVENT RECURRENCE. LERS 88 005, 89 009 AND 89-015 ARE RELA 1ED.
0 FO.R.M3 9................L. E R..S.C. 5 5.. D. AT A -......................... 8.-08-9.1 DOCKET YEAR LER NUMBER RIV!510N 005 hUMBER N51C EVENT DATE
. 333...199.0...*. 0.13........ 0..... 500510014 6. 2.16.171..... 04 / 09/.90 ABSTRACT POWER LEVEL - 000%. THE REACTOR VAS SHUTDOVN FOR REFUEL!hG CN MARCH 31, 1990. ON april 8 A JUMPER WAS th51ALLED TO BYPASS THE HIGH PRES 5URE ISOLATION RfLAY FOR THE 5HU100WN COOLING (SDC) M3DE OF THE RESIDUAL HEAT REMOVAL (RHR)[BO)SYSTEH. THE PURPOSE OF THE JUMPIR VAS TO FACILITATE CONilNUCD OPERAi!ON VHlLE REMOVING A 4100 VOL1
[EB) AND 600 VOLT [ED) ELECTRICAL D151RIBUTION BUS FROM SERVICE. THE SDC AU10MATICALLY ISOLATED THE FOLLOVihG DAY AT 12:27 P.M. DURlhG TRANSFER OF THE BUS. THE JUMPER (WHICH VAS h0i REQUIRED TO BE T AGGED) WA5 FOUND LYlNG ON THE FLDOR DUT51DE THE' RELAY CABlhET IN VHlCH IT HAD BEth INSTALLID VliH ALL11ATOR CLIPS. CONTRACTED CRAFT (LECTRICIANS KAD BEEN VORKING Ih5!DE inE CONFlhED SFACE OF THE SAME RELAY CABihET ON A MDDIFICATION. THfY STATED THAT THEY DID NOT NOTICE THE JUMPER ON THE FLOOR. IT IS BEllEVED TkAT THE JUMPER WAS UN!hTENTIONALLY KNOCKED LOOSE BY THE VORK IN PROGRESS lh THE CABINET. THE SYSTEM VAS RESTARTED ONE HOUR LATER. THERC WAS N3 SIGH!FICANT INTERRUPfl0N IN THE REMOVAL OF DECAY HEAT FROM THE RL ACTOR. CORRECTIVE ACT10h5 INCLUD(: THE PROCEDURE VHICH CONTR3L5 JUMPERS VILL BE REVISED TO REQUIRE TAGGING OF JUMFERS INSTALLED UNDER APPROVED PRDCEDURIS. RELATED LER-89-013 REPORTS AN 150LAT10N DUE 10 JUMPER GROUND!hG IN THE CONFlhED SPACE CF A RELAY CABINET.
l B - 13 l
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FO.R.M.... 4 0............... L. (. R.. S. C. 51.. D. A.T A...................... 0 $..> 0 9.- 9.1 DOCKIT YEAR LER NUMPER Rivl5!0N DOS WUMetR h51C EV[41 Daft
. 333... 1990..... 0.14......... 0.... 9 0 0 51 E. 0.14 9. 2.16.17 ?...... 04 /.10/. 9 0 AB51RACT POVER LEVEL - 000%. G4 APRIL 10.1990. A $PRINS VAS DISCOVERED 10 EC k!55th3 FROM A 2 INCH STOP AND $PRlks Ai$15 FED Pl$10N CHtCt. VALVE (23HPI-13).
TH[ VAlv[ VAS OrthfD DURINi PERFORMANCf 0F AN ALME ;'CTION Il IN-SIRVIC[ 1($ TING PRDGRAM DURlhG A PLAhi REFUILING OUTASE WHICH Sittito MARCH 31.1990. W11H001 TH[ $FRihG THE VALVE VA$ INDP[RABLE IN TH[ HAl20h1AL VALVE $UM ORIENTATIA US[D FCR THl$ thSTAL'_AT!k h CHN 1' '.L SPECIFICATION TABLE 3.71 Di$1GhA1[5 THis AS A t0hTAlhMl%T !$0LAfl0N VALY[. TABLT 4.7-2 [xEMP*i 11 FkDM 11Pt "C' LOCAL LIAK RATE TEST!hG EECAU$[ 1115 VA!:2 $tAL[D. THE VALVC !$ THE L AST DF TVO CHECK VALV[5 IN Al DRAlk LlhE FLOW PATH 10 THE TORUS FROM TH[ EXHAUST LINE OF TH[ HIGo FR[$$URl C00LAhi INJ[Cil0N (kPCI) (BJi TURB!hE.
THE !$0LAT!0N FUNCTION V0VLD HAVE PIEN PERFORMID BY THE UPSTR[AM CHECK bALVi ALTHOUGH i? !! ND1 4 DE$10~AT[D i
CONTAINMINT 150LAT10N VALVI. THE VALVE VA$ CP[hED FOR MAIN 1tkANCE IN 1976 AND 1977 Vl1H h3 '! CORD Of SvRlhG RIMOVAL. A CURRENT EN3tht(RlhG REY!!V $UPPORis A CONCLU510N THAT TURB!ht [1 HADST PRISSURf LAY WCT 4 AVE CIIN i
SUFFICl[NT UNDER SOME OPERATING C0h01110N5 (Dut TO PRES $URI DROP THROUGH THE SitAM TRAr AhD CdECL ALVI 5) 10 OPEN THE ; HECK VALVE VITH THE SFR]NS lh5 TALL [D, THE CHECA VALVL w!LL f[ ROTAf(D TO A $fiM VERilCAL Po$li10h (VHICH DOE 5 NOT R(0VIRE A SPRING A5515110 CLD5E). THI SPRING VILL ALSO BI R[ MOVED FROM THE hEIT UPSTRIAM CHECK i
VALVE.
........................ 06.-05-9.1
........1............... L. t. R. S. C. $. $. C. A T A 4
DDCKET YEAR LER huMBER R(V1510N DCS hpMBER h51C (VINT DATE
. 3 3. 3.. 199 0..... 015..........1.... 9 0 0 718 0 2 5. 5.!..! 6 5 E..l..... 04 /.19 /.9 0 l
ABSTRACT POVIR LEVEL - 000%. REACTOR VAS SHUTDOWN FOR REFUttlNG. ON 4/20/90.11 VA5 DISCOVERED THAT A FRERE 0J151"
$URVilLLANCE TI$10F $0URCE RANGE MONITORS ($RM) (IG) HAD h0T Bf(N PERFORMED PRIOR TO 6EGlhNING RIPLACEMEf f.
THE LOCAL POVER RANGE MONifORS (LPRM) (lu) Oh 4/19. FIVE OF $1X LIRMS kAD EI[N RIPLACID PRIOR TO THIS DISCOVERY.
t CAUSE VA$ THE OMI5510N OF $1EPs FOR PERFORMIN3 THE APPR0fRIATE SRM $URV[lLLANCE TIST FROM PROCEDsRE j
PREREQUlslits. THis OMiss10N VA5 NOT FOUND DURING THE TVO MOST R[C[h1 FR(VIOUS U5ES OF THis PROCIDUAE B!CAUSE ALL FUEL HAD BIEN REMOVED FROM THE CORE AND THE SURVE!LL ANCE 1!5T VAS h01 R[ QUIRED UNCER THESE CON 01110h5.
ALTHOUGH RIPLACEMENT OF LPRMS 15 TECHNICALLY A LOR [ ALi[ RATION. AND THEREFORE SRM $URV[lLLANCE 15 REQUIRE 0. TH!$
OPERA 110N DOES NOT lh ACTUAL FACT PRODUCE $1GNIFICANT (MEASURABLE) Al1[ RAT 10ki TO THE CORE RIACTIVITY OR NEU1Roh AB50RPT10N CAPABILITY. FURTHER. TH[ $RMS VERE k[RIFl[0 FUNCT10hAL VHIN TEST [D ON 4/19/90 UPON Di$C0VERY OF THE DMI$5104 AND HAD PREVIOUSLY BEEN TESTED AND 10UND FUNCTIONAL ON 4/1B/90. L(55 THAh 23 HOURS PRIOR TO START Ot LPRM R[PLACEMthi. ACCORDINGLY. THtRE VERE NO SAFETY-R[ LATED C0h5EQUENCES FRDM THIS EVENT. THE LFRM P[PLACEM[Ni PRDCEDURI AND OTHER $1H]LAR PROCEDURES HAVE BEEN OR VILL Et Reil 5ED 10 lhCLUDC FR[REQUIS11E STEPS 10R SRM SURVEILLANCC. $1MILAR LER$ 90+007 AND 89-008.
i
....... 2...............L.E.R S.C.S$. DATA............................. 08-9.1 FO.RM
_4 08-DOCKET YEAR LER NUMBER REVl!!ON DCS NUMBER h51C EVEh1 DATE
. 333..'..199 0..... 0.16......... 1.... 9 00718. 0 3 7 8. 21 B. 8 02...... 04 /. 2 4 /. 9. 0 ABSTRACT POVER LIV (L - 000%. THE REACTOR VA5 SHUT DOWN FOR RtFUELINS ON KARCH 31. 1990 THE REACTOR VES$[L HEAD HA$ BE(N REMOVED FOR REFUELIN3. THE SHUTDOWN C00LlhG (SDC) MODE OF THE RESIDUAL HEAT REMOVAL SYSTEM (RHR) (BO) ISOLATED DURING START-UP ON APRIL 24 AND AGA!N DURlh5 PUMP TRAN5 FIR ON APRIL 25. THE SYSTEM VAS SUCCEd. 'LLY RISTAR1[0.
THERE VA$ hO $1GNIFICANT INTERRUP110N IN TH[ REMOVAL CF DECAY HEAT FROM THE R(ACTOR CORE WHICH HAD SEEN $HU100V4 FOR 24 DAYS. SUBS [OVEhTLT A $[Rl[5 0F Six TEST RUNS U5thG A00!T10NAL lhSTRUMENT A110N. CONTROLLED VARI ABLES.
- AND DUPLICATE PRES $URE SVITCHES VERE CONDUCTED FROM MAY 9 THROUGH KAY 14 A CHAf t(RING PR[S$URE LWlTCH CAUSED THE ISOLATION. TEST RESULTS SHOVED A PO$$1BLE CAUSE OF THE CkATTERIN3 MAY HAVE BEEh A $KALL QUANilif UF A!R IN
' ONE OF THE TVO PRES $URE SWITCH [$ VHlCH lh!TI AT[ THE ISOLATION $1GNALS. THE TEST $ $HOVED THAT STARTING OF THE PUMPS DOES NOT PRODUCE A PRESSURE TRAN11thi 0F SUFFICIENT MAGh!TUDE 10 INITIATE AN !$0LAT10N $1GNAL VHEN THE INSTRUMENT LlhES ARE FILLED VITH VAT [R.
THE $VITCH, VHICH VA5 FOUND 10 BE CHATTERING PRIOR TO VENTlh3 WAS REPLACfD. THE VENDOR EXAMihATION COULD NOT [xPLAIN THE CHATTERING AND D!D h3T FlkD A DEFECT lh THE SV11CH.
RELATED LERS90-002 AhD 90-011 DESCRIBE 51MILAR ISOLATIONS OF THE SDC Vd!CH MAY FL50 hAvt BEEN CAUSED BY AIR IN THE PREStVRE SV11CH lh5iRUMEhi LINE.
B - 14
.%e rw. e a-=we-.=.m
........................ 00.00.-9.1
....... 3................ L. E. R S. C. 5 5. O. AT A l
4 DOCKli YEAR LIR h;MBER RIVI$10N DC$ NUMBER h$lt (VEhi DATI
. 3 3 3.. 19 3 0..... 0.17........ 0..... 9 0. 0 0 2 7 C. 2. t. h. 2.1 B../ 4 0..... 0 5 /. 2. E. /. 9 0 AS$1kACT POWER LEVEL - 000%. THE RE ACTOR WAS $HUTD0ih FOR REFUEllh3 Ch 3/31/90. AN th$1RUMEh1 SURVE!LLANCE FAOCEDURE, "RE ACTOR LEVEL (tCC5) 1RANSM!Ti!R Call 6 RAT 10N AND FUNCT!DNAL TEST ( ATTS)" WA$ PERIORMFD 04 5/26/90. THE lt$f FOUND THAT THE HIGH REAtiOR WATER LlVIL TRIP P0thi CN A 1 rah 5Mli1ER FOR THE HIGH PRES $URE C00LA41 ]h)tL110h (HPCI) (BJ) TURBINE WA$ 0.2 thCH!$ (0.3 rERC(h10F RANSE) ABOVE THE $EIPO!hi VALUE RIOUIRED BY T ECHN IC AL
$PECIFICATIONS. THE SAFE 1Y FUNCTION PREVENT $ P01EN11AL VAf tR CARRY 0VfR DAMAGE 101 URB! hts BY SHUTDDWN Of HPCl 10 PREVENT EXCES$1VELY HIGH RIAC10R WATER LIVEL. THE TRAN5Mif f ER WAS RICAllBRATED AND SURVEllLANCE VAs TEMPORARILY INCREASE 0 FROM ONCE PER CYCLE 10 ONCE IVERY $11 H0hTHS uni]L THREE C0hsttullvt AS-FOUhD kEADlhG$
l ARE VITHlN $PECIFICAfl0N$ UNLt$$ A CON 5151[h1 h0N-C0htfRVATIVE 1RthD IN THE DIRECTION OF THE $t1 POINT ORIFT INDICATES A httD '0R EARLIER REPLACIMENT. THE UPWARD $EIP0th1 DRIFT OF 0.2 INCHES WAS h0T SIGNIFICAhi 10 $ATETY SINCE IT V0VLD N01 HAVE RESULTED IN A MEASURABLE INCREA$t IN M0!$1URE CARRYOVER LHDULD A HIGH LEVEL $110ATION tAYE OCCURRID. MDRE THAN FIVE FIET REMA!NED BETWEEN THE 1 RIP $tiPOINT AND THE STEAM tlht$.
RELAl[D LERS:85-014,85-015 AND B6-008.
FO.RM....... 4............... L. E R S. C. $. 5.. D. A T A........................
02.-08.9.1 4
DOCKET YEAR LCR huMBER REVISION DCS kbMetR ASIC IVEh! DATE
. 3. 3 3... 19 9 0...... 0.18........ 1.... 9 00911018 2.. 2.19 4 29...... 0 0. /15 /. 9 0 ABSTRACT PDVER LEVEL - 100%. OURING THE REFUILING OUTAGE EIG!hhlkG 03/31/90, THE ACTUAllhG 10PVORKS NECHANISM FOR ElGHT p
SAf t1Y RELIEF VALVES (5RVS) ( AD) WERI RIMOVED FOR itSilhG. TWO VALVES ACTUATED At FRESSURES WHICH OEVI ATED FRDM THE hAMEPLATE StiPOINT BtYOND THE +/-1.0% TOLERANCE ALLO 1D BY itCHhlCAL $PECIFICA110NS. ONE SRV LIFTED 1.1%
BELOW THE SCIPOINT, THE OTHER LIFTED 2.3% ABOVE THE SETPOINT. h3 CAUSE WAS DEltRM!htD. A PLAhi $PECIFIC AhALY$15 STATES THAT AN 5RV SETPOINT TOLERANCE OF +/-3% V!LL HAVE NO $1GNIFICANT SAFETY IMPACT ON Vl5$EL OVERPRES5URE MARGIN. THERMAL LIMITS, ECCS/LOCA PERFORMANCE. HPCl/RCIC OPERABILITY. CONTAlhWEh1 PRES $URE. OR CONTAINMENT INTEGRITY, THERE WAS AN ADCOUAlt MARG!N BETWith THE As-r0uhD SRv stif thGs AND h0RMAL REACioR PRES $URES DURlhG POWik 1rtRAT10N. CORREC11VE ACTION INCLUDED REFLAtlNG THE FAILED $RV WITH A RECER11FIED VALVI.
CONTINUED PARTICIPAT10N IN THE B CVhER$' GROUP 10 RESOLnt SRV liSUES, AND PREVIOUS SULMISSION 10 THE hRC OF PROPOSED CHANCES 10 f tCHNICAL SPECIFICATIONS TO TAKE CREDIT FOR EXCESS INSTALLED SRV CAPAtl1Y. LER-85-009, 85 013,87-004, 88 004, B0-010, AND 89-026 ARE $1MILAR EvthTS INVOLV!hG $RY Stirolhi DRIF1.
....................... 0 8..- 0 8. - 9.1
....... 5.............. ~ L. E R S. C. S. S.D. A. T A 4
DDCKET YEAR LIR NUMBER REVISION DC$ huMBER N51C EVEki DATE
. 333.. 1990..... 0.19......... 0..... 900B0B0201. 2.19.02.1..... 0$/.30/90 ABSTRACT POVER LEVEL - 000%. A ROUTINE Lf AK RATE TEST OF THE ADS PNEUMATIC $UPPLY CHECK VALVE $ WAS IN PROGRESS ON 5/30/90 DURING THE 1990 REFUELING OUTAGE. THE AU10MAftC CEFRESSURlZATION SYSTCH (ADS) SAFETY RIllEF VALVI ($RV) (AD)
ACCUMULATOR PRES $URE VAS DISCOVERED 10 b[ 40 PS! LDWER THAN THE TEST REQUIRED 110 PSIG. THE SY51tM INGINEER WAS NOTIFICD VHO CDMMENCID AN INVESilGA1!ON. THE CAUSE OF THE LOW ACCUMJLATOR PRES $URE WAS A LARGE PRES $URE DROP ACROS$ THE ACCUMULATOR CHECK VALVES. THE CHECK VALVIS VERC SFR!h3 LCADID CHECK VALVI$ WITH A RATED CRACKlh3 PRESSURE OF 25 PSI EACH. THESE CHECK VALVES VERE SPECIFIED BY THE ORIGlkAL DESIGN AND THE IMPACT OF THIS CRACKING PRES $URE WAS NOT CORRECTLY CONSIDERED. TH15 (RROR WAS CONilhutD IN LATER ADS PNEUMATIC $UPPLY ANALYSIS.
THE AUTHORITY EVALUATED THE IMPACT OF THISC CHECK VALVES ON THE ADS SYSTEM AhD DETIRMlhED THE SYSTEM 10 HAVE BEEN OPERABLE. THE CHECK VALVES VERE REPLACID W11H CHICK VALVES VITH A RAllD CRACKlh5 PRI55URI 0F 1 PSI AND I
SUCCESSf ULLY PASSED THE OPERABIL11Y it$TlhG. THis LER IS SUEulTTED IN ACCORDANCE Wliv
?GREEMENT V!1H THE hRC IN ORDER TO OtsCRIBE THE FlhDlhGS OF THIS [Vthi AND A0 VISE THE hRC OF THE CUPREN PNTUHATIC SUPPLY SYSTEM CONFIGURATION CAPABILITIES.
B - 15
4 LlR 5055 DA1A CE 0A 91
........ 0...........................................................
FORM j
i D00tti ifAR LIR haMEIR Elvl51CN 005 haMFER h51C
[\\lhi C ATE
. 3 3 3... 19 9 0... 0 2 0......... 0..... 9 0 0 E 0! 0C. ?. S. 219. :. 2 2.... C. C / 2. (. / 9 0 I
AB5 TRACT PD.ER L Evil - 000%.1FE EL ACTOR WA5 SHJiD0iN FOR MAlkilhA%:t C* JJNl 25. 1990. A1 3.20 A.M. Oh Joht 25. 1990 THE SHJTDOVN CD0llh3 (5CC) MODI Of Th[ 5[51 DUAL FI A1 RIMOVAL (E"El (LO) st5f fM 150tAffD v"[h THE RH4 PUwe WA ST AETED. THE ISOL Ai!Oh WAS FESti AhD TH{ $DC 9[5t Akl[0 21 Mlhul[$ ail [R THE lbOLATION. Il511h3 CEMDh51#.A 16AT FRE55URE TRA%51[h15 DURI AS PUMP ST ART VERL CNLY 3% (? 05!) CF THf 00 95f G rE!S5URE #f QU!Ef D FO SW11CH ACTUAT!0N AND 5'51EM 150LAT10h. A SYSTEM VAlt30sN lh510t FRlHARY C0hf AlhW!%f FOUNO A P01(N1 IN THE PRISSURE Sth$thS tlk! 10 Thl 150LA110N SiliCH. A!R IN THE Sth5!h5 tih[ 15 19{ P0510 TAT [D CAUlf CF 'Hf ISCLATION. A POSSIBLt DESIGN M00lflCAT10N 101HI 5!hilh3 Lih[ IO (LIMihATI THE AIR TEAP WILL PE IVAt9AffD.
RELAtlD LtRS: 90-002,90-011 AND 90-016.
......................... L. [. R 5 0 5 5.. C. A T.............................S.-91 A
CP C F0EM 47 DDCKET Y[AR LER NUMBER RIVISION dis h)ME!W h51C tv[hi DAl[
........ !. / 2 9. / 9 0
(
. 333... 1990..... 021........ 0..... 9008130140. 219224
/.BSTRACT P0itR LEVEL - 000%. A 1EMPERATURE SulTCH 1% TH! 51t AM Lf Ar DITECTION t00!C CF THE EL AC10R V'TER C (CE) 5YSTEM WA5 SCHIDULID FOR KIFLACIMINT DURlh5 A RE ACf CR SHUTDOWN. THE WWC ST5itM.A5 th 0F Pi[Vthi DIYl510N !! RdC PRIMARY CON 1 AlhMEhT 150LAi!CN VALYEL ht&E l>f thT10% ALLY D15A!'LtD A10518 Ch t/28/9910 AN lhA0VtRTENT ISOL ATION CUR!hG sV!1CH REFLACEMEhl, A1 ArFFCIlWAlflY 1400 A TCCHNICI AN Llfi[D THC tlADS 1Hf DRAVIN3 WAS IDINTIFlf D BY A C0hTROLLED ORAWINS AS THE POWlR SV0 PLY TO lb[ DIV1510N !! 1[MP[RA'UGt Sw!TCH IN [RROR AhD tlFilhG In( LEADS Uhihi[hT!CNALLY D15ABL(D in[ O! VISION ! LOGIC. V"[N Tel T[CHn!Clah Al 17C5 REPL ACID THE LEADS. fHf 150LAT10N % ALVI IN Dlvl510N 1 CLO5(0 THE Of A.ths (GROR Aho THt UN!hT(h710h*L D1 0F THE DlV1510N 1 LOGit VEtt DISCOWEAE0 DUR!b5 SUESIOuthi Rtyltw Dr THE DRAW!NGS Ik 1HI COURSE OF lhvE51tGAlthG 1H[ UhtXF[CTED CLOSUR* Or THE OFP0511t (DIV15tCN 1) 150LA1109 VAlv[. (4[h THOUGH hEITHIR Olvl510N WDULD AUTOMATICALLY 150LAlt0 TH( RWO, PRIMAEY C0%I Al%W{hl AhD 150(47{0% W[C[ ACI S(QU($[D 10 &[ Ortkkhi( h[CAUSE Idl REAC10R WAS SHUTCOWN AND hEhi[D. A LEAL UND[R THtst CON 0li10NS WOULO h0i NAVE [L(VATED THE Aw?t OUf flC'ENTLY TD CAUSE (ITHER AN AL AIM CR AN 150LA110%. THE 0;A4!h5 CER045 VILL PC CD&RICifD, "AS bullI" REVILIONS 10 VikDOR DRAVlh55 ARE NOW ACCC*PL15r!D EY AUTH0kitY FIRSONNf t.
......................... L. E R SC. 5 5.. D. A T A...........................
0B 91 09-FDEM AB DOCKET YEAR LER NUMBER REVISION DOS AUMEER NSIC
[VthT DATE
. 3 3 3.. 1990...... 02 2......... 0..... 9 0 l L.19 0131. 2.19. 7 4 2...... 0 9. /.10./ 9 0 ABSTRACT POWf R LIVEL - 100%. THE PLANT WAS OrEE ATlhG Al FULL P0iiQ CN 9/10/90. Af 9:36 A.M. A LOW VOL1 AGL C0A0lT1 THE OUTPUT FROM THE MOTOR O!hlRA10R (MS) F0W[k SUFFL' TO THE B RE ACTOR FR0flC110h 5f 5 TEM (RPS) (
THE F05 IN TRIFP!h3 0F THE RPS BJS ELECTRICAL PROT [CilVE A55tMELY ([T A) UNO[R YOL1 ASE RELAY AND DEEhlkGlZlh3.
RESULTED, TOGETHER V11H ISDLATION Of THC REACTOR WATIR CLCAN-UP Sf!Tf M (CE),
BUS. Ah AUTOMATIC KALF-SCRAM EEACTOR BU!LOlhG (hG) VINTILATION (VA), CONT AlhM[hi AIR DlLUTION (CAC) (LK), DWYWELL EQUIPMENT AND FLOOR ORAth PUMP 5 (WK), AND C0hlAlkWEh1 SAMPLE (LN) SYSTLM1. THE kP5 VA5 SwifCHED TO 115 ALT [RNATE POWIR SUPPLY AND 1Hf 150LAT10h5 Vf RE RESET. THE LOW VOLT AGE CON 01110N ELSUL1[D FRDM DR!ff Of THE M3 ![T VOLTAGE FIGU THE OPERATORS DOWNWARD DRif f 0F THE OUTPUT V0tTAGf TO THE ras 5E OF THE TRIP 5tiP0lhi WAS kOT RECDGN1?ID eY DUR!hG THElR SMlff SURVIILLANCC BECAUSC A CEFICIENCY lh THE DAILY SURVilLLANCE TEST PROCEDURE SP EXCESSIVELY VIDE rah 5[ FOR ACCEPT ABLE OUTPUT VOLTAGE. CORRE.CT !V[ ACTION (hCLUD[D REPL ACt#EN REGULATOR BOARDS, THE EPA ASSEMBLY CIRCulf BREAKER. ANO IMrE0VEM!hT510 inE SURvtlLLANCE rROCEDuRES WHICH D[5CRIBE TH[ HONITORlh3 Or DuiPUT VOLTACE.
E - 1L l
I l
FO.RM...... 9................ L. E. R.. S. C. $. :. C. A T A....................... C. 6 - 0 9 - 9.1 4
DSCKti YEAR LIR hpMBER Rivl510N DCS kJMBER h5]C EVthi DATE
. 3. 3 s.. 199 0..... 02 3........ 0..... 9 0.112 6 02 2 2.. 22 0.159.......10. /19. / 9 0 ABSTRACT PostR LEvil 0451. AT 0912 0% 10/19/90 THE REACTOR VAS MANUAtt? $ CRAMMED FRDM 4$ PIRCEht P0iLR A$ A CON 5tRVAi!YE MEASURE DUE TO 1hAB1tifY 10 ; LEAR THE CIRCULAllNG WATER sf 5f tM 1RAVEllkG $CREENS. THE RAf t OF
$UPPLY 0F CIRCULAT1h5 HATER HAD BEEN REDUCID Duf TO CLOGGlhG Or THE C00L1hG HATER ikTAtt SCRitNS (Kt) bY LEAVES AND LAkt VEEDS. OhE OF THRIE $0% CALACITY SCREth$ VA5 Oui 0F SERVICE 10R SCHEDutt0 PREVthilVE MAINTEhAhCI.
A $HIFT IN VIND DIRECTION CONTR!BUTED 10 AN UNU5UALLY LARGE DIER15 ACCUMJLATION ON THE REMAlb!hG TV3 $CREINS.
i THE SCRt[N DIFFER [hilAL PRE 5$URE ALARh AND SCREINVA$H SYSTEMS, VHICH VOULD HAVt FROY!DID EARLY thD! tall 0N Or FOULlh3 hA0 OtEN UNihitN110hALLY DISABLED CURlhG ThE RA!hithANCE OF THE DUT Or 5tR)lCE SCRIEN DUE 10 A PRDCEDURAL DEFICIEhtY. $ HEAR Plh5 DN THE REMAlhlhG TWO 5tkEthS FAILED. 11 VA5 htCI$SAPY TO RIM?VE TVO Or THI THREE Math CONDfh5ER CIRCULAllN3 VATER FUwPS (LE) FROM $ERV!CC 10 CLEAR AND RITAIR THE $CE! ENS.
THIS hECE5511ATED THE P0i[R REDJF110h AND SUB5t0Uthi MANu!L $ CRAM. THE PL Ahi RtTURhED 10 SERVltE AT 0418 DN 10/21/90
( A3 K'DURS. 6 MINJit$ OrF L!ht). AFFROPRI Alt tf EPS HAVE Btth AD0ID 10 PROCEDuRis. THE 01FFEREh11AL FRE11URE INSTRUMENT ATION HA$ BEEN AUDED 10 OPERATOR LOG 5Htt15. PLANT FLOW DI AGRAM$. AND thSTRUMthi CAlifRA!!DN SCHEDULES.
F.O. R. M.... 50................ L. E. R.. S. C. S. S.. D. A. T A......................08..- 09. -9.1 DOCKET YEAR LER NJMBER REV1510N DOS hUMBER h5!C Eithi DATI
.. 33.. 1990..... 024..... r.... 0.... 9 0.12.12 019 3. c.2 2 03 7 0...... u.1/ 01/. 9 0 3
1 l-ABSTRACT o
POVER ttVil
- lui%. V11tt THE RE AC10R WAS OFIRAllNS Al '.00% POVER A12332 0411/01/90 A HlGHLIVEL ALARM FOR 1HE PRIMARY CONTAlhMthi (0RYWELL) (QUIPMEhi DRAIN SUMP VAS RICE!VID. A FU$t CFthED Al ?338 RESUtilhG IN A FAR11 AL ACTUATION OF THE B $1DE OF THE GROUP !! PRIMARY CONTAlhwENT 150LAi!04 SYSTIM (PCl$)(JM). THE REACTOR bu!LDlhG VENillATIOh SYSTEM (VA) 150 LATED AND STANDBT GAS TREATMEh1 B (50GT)(EM) $TARitD As DEilGkED. THREE VALVES THAT RfCtlVE THE ISOLATION SIGNAL VIRE ALRCA0Y CLOSID. TWD OiHER$ CLOSED AS DI51GNED. THE !$0LAT10NS VERE Rt!ET WITHIN 37 MINUTI$. A REPAIR 1A5K ON A SMALL 50LEN0!D CPERAllD A!R ATM15$10N VALVE ONEAPCCTEDLY 'X1th0ED OVER THREE SHIFT; DURING VHICH ONE TRAIN OF Dit RESIDUAL HEAT REMOVAL $Y51EM (20) VA5 BElk3 RES10 RID 10 SERVICE AND RADIDACTIVE RE$1N TRAN$TER WA$ lh FROGPISS. MJLilPLE HUMAN (RKDRS AhD PDOR COMMuh!CA110N BETVEEN SHIFT 5 l
OCCURRID DURlhG IQUlPMthT PROTECilVE TAGGIN3 PROCEDLRE IMPLEWINTATION. THE 50LEN0!D VAX REMOVID FROM THE VAtyt BUT WAS LEFT [LECTRICALLY CONNECTfD DUI TO THL HUMAN (RROR5. Ah AUTOMATIC SIGNAL FROM THE 5 UMP HIGH ttVEL i
SUBSEQUth1LY [htRGlZED THE 50LtholD DAMAGlhG TH[ lhSULA110N RESULT!hG IN HIGH CURREh! AND OPENING OF THE FUtE.
RELAlt0 LERS: 89 017. 69 013.67-016. AND 86-019.
(
FOR.M...... 5.1............... L. I. R S. C. 5 5.. D. A T A..................... 0 8. - 0 2 9.1 DatitT YEAR LER NUMBER dEV1510N 00$ hJMEET hSIC EVENT DATE s
. 333... 199 0...... 02 5........ 1..... 91013..! 015 6.. 2 2 08. 7 2...... 11. /.15. /.9 0 ABSTRACT POVIR LEVEL - 100%. ON 11/15/90 AND 1L/26/90 1HE PLANT VAS OPERAtih3 AT 100% POWER. O,R'NG A SCHIDULED ASw!
SECTION XI IN-5tRVICE TE51 PROGRAM CURVIf LLANC[ 1[ST. THREE 3.!NCH SV!NG CHICK VALVES FAILED TO CLO:E ON 11/15/90. TV3 ADD 1110NAL VALYt> FAILID ON 12/26/90. F0tt0V!h3 THE lhlil AL TESTS, FOUR OF THE FIVE VALVES C.0 SED VHTN TAPPED WITH A TOOL HANDLE. THE VALVIS SUPPLY SERV!CE VATER (SVS) (KO) TO N!NE AREA VEh11LAT104 UN11 COOLERS LOCATED IN $ PACES CONTAlhlhG SAFETY-RELATED ILICTRICAL SVITCHGEAR AhD IMERGthtY CORE CDOLING Sv51EM EQUIPMENT, THE VALVL3 ARE INTENDED 10 CLOSE UPON LOSS OF SERVICE VATER PRESSURE TO PREvtNT DIVERSION OF lHE IMERGthCY SERVICE VATEi (ESVI (BI) SUPPLY AVtY FROM THE COOLERS. ON 11/16/90 AND 12/27/90 CARBON S1Ett VALVI lhTERNALS VIRE REPLACfD MlTH STAlhtESS STEIL COMPONENTS TO AVOID COEROSION PROBLEMS WHICH CONTRIBUTID TO THE a
$f!CKING CONDlil0N. THE AS-FOUND STbCK OPiN VALVE C0hDli10N VDULD h01 BE (XPECTED 10 RESULT lh CONDITIONS ADVERSE TO SAFEff }N THE EVENT OF AN FSAR Poh1ULATED ACCIOTNT. THIS l$ A VOLUNTARY REPORT. RELATED LER$:
88-01E. 88 009 AND 90 012.
B - 17
+
FO.R.M.... 52................ L. I. R 5.C. 5 5..C. A1 A........................ 0 6.- 0 8.. 9.1 DOCKET YEAR LtR NJMBER REVll10N DCS huMDIR h51C
[V!hi DATE
. 3 3 3... 199 3..... 02 6........
- 0..... 9.10110 0168 2 2. C 0 68...... 12. /.12. /. 9 0 AB5 TRACT P0i[R LEVEL 100%. A REACTOR SCRAM FRDW FULL P0itR OCCusRID A11352 Ch 12/12/90 DUR1h5 CAllBRATION OF RCACTOR WATER L[v[L IN51RUW[h1 Ail 0N. THE lh5TRUWEN1 B[lh5 CAllBRAf tD SHARID COMMON R(FIRthCC AND VARIABLE LIVIL LEG 5 WITH IN51RUMEh15 0F THE REACTOR PR0?[CT10h $YST[M (JC) TML RfACTOR SCRAMMED AS THE INSTRUMfNT HIGH PRIS5URE 150LAil0N VALVE VA5 BElhG CRACKID OPIN DURING RfiURN 10 5tRvitt. THE SCRAM RESUtitD FROM A FAL$t LOV RE ACTOR VAftR LEVEt SIGNAL. CUR!hG THE ACTUAL LEV [L TRAh51thi FOLLOW!h31HI SCRAM DIFFICULTV VAS LAf tRith?ID V!1H Rf 5f AR11NG THE REAC10R Fl[DWAf tR PUMP 5 AhD A FAILURt OF THt EtAC10R FitDWAf tR LOV FLOV C0h1ROL VALVI OCCURRt0.
A SECOND SCRAM OCCURRIO DUt 10 AN ACTUAL 10V R[ ACTOR VAltR L VIL AT 1816 DUC to FAILURE OF 1HE REACTOR FitDWAllR LOV FLOV CONTROL VALVE AIR OF(RATOR ClAPHRACM. THE FLAhi RE1URh;D 10 LIRvitt A10658 DN 12/17/90 AF1[R CIlhG OFF L!ht FOR A DAY 5, 17 HOUR $ AhD 6 MINV1ES. A R001 CAUSE INvt$11GA110N OF TH15 $ CRAM 151N FROGRt$$ AND 15 EXPECTED 10 Bf COMPLET[0 PRIOR 10 THE IND OF lH! FALL 1991 REruttlkG OUTAGE. UNill THth. IUTURE CAllbRA110NS VILL BE CONDUCT (D DURlhG SCHEDULED OUTAGil. R(LAf tD t[is: 90-001 AhD 90 C27.
F.O.R.M..... $. 3..... * *........ L. ( R. 5. C. 5 5.. D. A T. A.. * *................. 0 8. C. 6.. 9.1 DOCKET YEAR. LER NUMBER REV1510N DCS huMBlR h51C EVEhi DATE
. 33 3....199 0....... C.2 7......... 0.... 9.10 !.! 8 017 5.. 2 2 08.17.....12. /..l t. / 9 0 f
AB51RACT POWER L(VIL - 000%. A REACTOR SCRAM FROM $11 PtRCthi POWER OCCURRID DURING A START-UP GN DlCIMBER 15, 1990 At
,t 21 A0. THE CAUSE VA5 A FAILURE Or A FABRIC-V0V[N PUNA-N DI APHRAGM VHICH HA0 BIEN IN 5tRVICI FOR 15 YtAR$ IN THE AIR OPCRATOR FOR THE REACICR FEEDWAl[R LOV Flow CONTROL VALV[ AND AIR LCAKAGE FROM THE Or[RATOR LitM PACK!kG GLAND. THE FAILURf 0F THE VALVI 10 $1ROKE FULL OPEN RESUL1[D IN AN thABIL11Y 10 SUFPLY SUFFIC1 TNT F([DWATER FLOV TO THE REACTOR. THIS RE5 ULT [D th DECREA$lhG REACTOR VAf tR LEVIL AND THE htCE5511Y TO USE 1HE R[AC10R FEtO PUMP i
(RFP) DISCHARGE VALVE 10 C0hf ROL VAltR LIVtt. ArrER $[ VIRAL CohTROLLED JDG5 th 18C OPEh DIRECTION OF 1Mt RFP Dl$ CHARGE VALVE, THE INCREAst IN VATER FL0d RESULTED IN A HIGH h(UTRON FLUX LCRAM OF THE REAC10R DUE TO EXCEIDlNG THE 15 PERCEh1 POVER tlM!i VMILE THE MDDE SWITCH VA5 IN THE START-UP M00E POSITION. CORRECTIV(
AC110NS INCLUDED REPAIRlhG THE REACTOR FIEDWATER LOV FLOV CONTROL VALVI OPIRATOR, REVI$lhG THE ST ART-UP PROCEDURE TO VERIFY FULL LiROKE CAFABILITY OF THE VALVE, AND ADDlNG A CAUTlCh LIMII]hG THE ACCIPTALLE OPEN DEMAhD $1GNAL FOR THE VALVE TO 10 P(RCtNT. THE PLAhi VA5 RESTORED TO SERvitt AT 0658 ON 12/17/90. RELATED LER:
90-026.
s.
.FO.RM..... 5.A................ L. E. R S. CS S..D A T. A..........i............. 08-08-91 DOCKET TEAR 1.ER hl>BER RCV!$10N D05 hukBER N!!C
[Vthi DATE
. 333....1990..... 02 8......... 0..... 9102 0602 50. 22 09 3 3..... 12./.21/.9 0 4
AB51RACI POWER LEVEL - C00%. THE PLAhi VAS OPERAllNG AT FULL POVER ON OtttMEER 27, 1990.
THE REACTOR BulLDIh5 VENilLAi!ON EXHAUST RADIATION MON!iOR (IL) VA5 EE!hG REMOVED FROM 5tRY!CE A5 A PREREQUISITI TO PRivthilvi MAINTINANCE ON THE M0hlTOR LAMPLE PUMP. THE 1[CHNICIAN 10VCHED A VOLT METER PROBE 10 THE MONITOR HIGH VOLT AGE circuli TEST C0hhtCT10N 'O DEitRMiht THE VOLTAGE lh ACCORDANCE VITH PROCEDURE. THE MDh! TOR SFlKED HIGH kESUlilNG lh 150LA110N C' THE REACTOR BUILDING VENi1LATION SYS1[M (VA) AND PRIMARY CONTAlhwEhi ATMOSPHERE SAMPLING SYSTEM (E3) AND START OF THE STANDBY GAS TREATMENT SYSTEM (DH) AT 0830. THE SYSTEMS VERE RESET AT 0835 THE RAD! A?!ON MONITDR AND VUti MET [R WERE CHICKED AND DEitRMihED 10 Et OPERAilhG PROPERLY. VOLTAGE SPlKE5 ARC NOT A; UNUSUAL GJCURRENCE VMEN MAINTAIN 1h5 RADI A110N MONiiOR [LECTR0klCS, CORR [CilVE ACTION VILL BE TO GENIRATE A SPECIFIC PROCEDURE F08 THIS MAlhTENANCE EVOLUTION. TH15 PROCEDURE VILL D($CR18E THE POTINTI AL FOR $PUR1005 !$0LA110N AS THE RADI ATION MONITOR 15 RCMOVED AND RETURhED TO SERVf CE.
f B - 18 o
F.O.RM...... $. $............... L. E. R.. S C. $ $.. D. A T A.......................08.-08.-9.1 DOCKET YEAR LIR NUMBER REVISION DC$ NUMBtR N51C EVEkt DAlt
. 333.. 1990..... 029......... 0..... 9.10.225010.7. 2 2.104 6........!!./.0. 7/.9 0 AB$fRACT POWER LEVIL - 100%. DURING PERFORMANCE OF AN 157 ON 4/9/90 THE TEMPERATURE INPUT FROM 10F 16 DUAL (LEME11 RES!$1ANCE TEMPERATURE DETECTOR 5 (R10$) VAS READING DOWNSCALE FOR THf PRIMARY CONTAlhMENT (NH) $UPPRES$10N POOL (TORUS) VATER TEMPERATURE. ON 12/1/90 DURING PERFORMANCE OF THE VORK REQUEST TD INVE5flGATE THE 00VN5CALE READINGS. IT WAS 0l$ COVERED THAT BOTH CHANNELS OF THE TORUS BULK AVERAGE VAfiR TEMPIRATURI MDNITOR VERf I
INCLUDING A FALSE VALUE OF 30 DEGREt$ F FROM THE 00VN5CALE R10. THE $10hAL FROM A FUNCTIONING RfD VA$ THEN TEMPORARILT CONNECTED INTD THE AVERAGING CIRCUff IN PLACE OF THE DOVN5CALE RfD $lGNAL. THE EVENT VA$ P[ PORTABLE BECAUSE THi INDIVIDUAL RfD LOV VALUE VA$ INCLUDED IN THE AVERAGE RESULTING IN A LOVER THAN ACTUAL TORUS BULK TEMPERATURE INDICATION AND THUS BOTH CHANNEL $ VERE INOPERABLt. At THE f!ME 11 VA$ THOUGHT THAT ALTERhATE i
INSTRUMLNTATION Ll$1ED IN A FORMAL TECHN] CAL $PECIFICATION INTERPRITATION COULD BE USED IN PLACE OF THE INOPERABLE INSTRUMENT. THE IVENT VA1 RECLA551Fl[D AS REPORTABLE ON 1/14/91 VMEN AN [RROR IN THE FORMAL TECHNICAL
$PCClFICATION INTERPRETAil0N VA$ BROUGHT 10 THE AT1EN110N OF THE PLANI HANAGEMENT $1AFF BY THE NRC RISIDENT INSPECTOR. THE FAL$E LOV AVERAGE READINGS R[$ULIED FROM THE AB5EhCE OF A DUAL ELEMENT RfD WHICH HAD NOT BEIN INSTALLED AhD AN INGINEERING ERROR IN CHANGE TO $0FTWRC PROGRAM.
F D. R. M..... $. 6................ L. E R.. $. 0. $ $. O. A. T. A...................... 08..- 0 9 - 9.1 DOCKET YEAR LER NUMBER REV!$10N DCS NUMBER N$1C (VENT DATE
. 3 33....199.1..... 001........ 0.... 9.102.14 00 7 4..... 220944....... 0.1/. 09 /.9.1 ABSTRACT POWER LEVEL - 100%. A FAL$E PRIMARY CONTAINMENT HIGH RADI A110N ISOLAT!ON (JM) SIGNAL OCCURRED AT 1105 ON 1/9/91 VHILE THE PLANT WAS OPERATING AT FULL POWER. THE $1X FRlMARV CONTAINMENT 150LAT10N VALVES WHICH ARE ACTIVAT[0 BY in!5 $1GhAL VERE ALR[ADY lh THE CLOSED POSlfl0N. THE TRIP VAS RESET AT 1109 REDUNDANT IN5tRUMENTAT10N CONrlRMED THAT RADI ATION LEVELS VERE NORMAL. AN ELECTRICAL NOISE $1GhAL VAS GENERATED VHIN CONTACTING A VOLT METER PROBE WITH THE HlGH VOLTAGE TEST CONNECil0N DURlhG A CALIBRATION OF AN UhRELATED IN11RUMENT. THE UNRELATED INSTRUMENT VAS PHYSICALLY LOCATED ABOUT ONE FOOT ABOVE THE ELICTRONIC DRAVLR FOR THE PRIMARY C0hTAlkMENT HIGH RADIA110N MONITOR VHICH GENERATED THE FAL$t $1GNAL. CAllBRATION PROCEDURES FOR RADI A110N
.SNITORING INSTRUMENTS VILL BE REVl$tD TO NOTE THAT THE POS$1BILITY OF $PURIDU$ TRIP SIGhAL BEING 1RAN$MITTED TO ADJACENT INSTRUMENTS ExtSTS. CURING THE 1991 REFUELINC OUTAGE THE PAhEL CONTAINING THE RADIATION MONITOR 5 VILL BE TWROUGHLY IN5PECTED AND ANALYZED TO DETERMINI IF IMPROVEMENTS IN NCl$[ $UPPRE5510N ARE POSSIBLE.
RELATED LER: 90-028.
FORM 08..-08..91
........ $.7.............. L E. R $.C. S S. D. A. T. A.......................
DOCKET YEAR LER NUMBER REVl510N DC$ NUHBER hSIC EVENT DATE
. 333... 1991..... 002......... 0.... 9.1041002.10.. 2213.72..... 03./ 04 /.91 ARSTRACT POWE4 LEVEL - 100%. ON 3/4/91 AN OPENING VA$ MADE IN A VALL 10 INSTALL A 3+1hCH TELEPHONE CABLE CONDUlf SLEEVE.
THE VALL l$ A FIRE BARRIER IN APPENDIX A 10 NRC BRANCH TEC9NlCAL POSIIl0N 9.5-1.
THE PLANT VA5 OPERATING Al FULL POWER. NO FIRE VATCH VAS PROVIDED. THE OPCNING WAS LEFT UNATithDEC VHIN VORKtR$ VENT TO LUNCH AT il45.
THE OPENING WAS DISCOVERED UNATTENDED AT 1226. A FIRE VATCH VA$ E$1 ABLl5HED. THE VALL $EPARATED THE TELEPHONE ROOM FROM A $1AIRVELL AND ULTIMATELY FROM $TIIL CON 0Uli CONTAINING SAFETY-RELATED CABLE LOCATED MORE THAN 138 FEET FROM TME OPENING. A SECOND FIRE VALL 15 th PLACE BETVEEN THE OPENING AND THE $AFETY RELATED CABLE. THIS l
$ECOND VAtt 15 NOT DESIGhATED IN THE FIRE PROTEC110N MANUAL. THE WORK REQUEST $PECIFILD REQUIREMENTS TO PROYlDE FIRE BARRIER $ 14 THE CONDUIT IN ACCORDANCE WITH A SPECIFIC PROCEDURE. THE PROCEDURE RE0V! RED POSTING Or A FIRE l
VATCH FOR A LIST OF SPICIFIC AREA $. NOT INCLUDING THis PENETRATION. THE VORK PERMIT REQUEST FAILED 10 PROVIDE INSTRUCTIONS FOR A FIRI VATCH. THE JOB SUPERVISOR VAS EXPfR.ENCEl) 11 VORKING ON FIRE BARRIER PENETRATIONS, THE CAUSE 15 ATTRIBUTED TO HUMAN [RROR IN NOT RECOGNIZING Tdt NEED 10 (STABL15H A FIRE VATCH VHILE THE FIRL BARRIER PENETRATION VA$ OPEN AND UNSEALED AND A DEFICIENT PROCEDUPE, t
l B - 19
I
.F O. RM.... SC................L. E R.. S.C. S. S.. D. AT. A...................... 0. B.. 08..- 9.1 DOCKET YEAR LER huMBER REylSION DCS hUMBER NSIC EVEhi CATE
. 3.33....199.1..... 003......... 0..... 9.10419 0336. 22.14 57........ 03./.09. / 9.1 ABSTRACT POWER LEVEL - 000%. ON MARCH 10,1991 AT 0100 HOURS, VITH THE PLANT SHUTCOVN IN THE COLD CONDITION, A NON-LICENSED OPERATOR (NLO) DISCOVERED THE PRIMARY CONTAINMENT (CRYWELL) ENTRANCE GATE 272-11 UNLDCKED AND UNGUARDED WITH THE LATCH MECKANISM TAPED OVER. THE NLO IMMEDIATELY UNTAPED THE LATCH, SHUT AND LOCKED THE GATE.
A RADIOLOGICAL TECHNICI AN.i3 TAPED OVER THE LATCH MECHANISM TO F ACILITATE MOVING EQUlPMENT (HEPA HOSE) INTO THE PRIMARY CONTAINMEhTS
'4 7SP?"' TO REMOVE TME TAPE AND LOCK TP' GATE WHEN HE Flk!SHED BEC/USE HE LEFT THE g
PRIMARY CONTAINMENT THRobu9
- 1,cEENT ACCESS POINT. THE GATE VAS LEF) !N THE Uht0CKED AND UNGUARDED CONDITION FOR A PERIOD OF APPROXIMATELt t0VR MOURS AND TVENTY MINUTES. A CHECK OF THE ACCESS LOG FOR THE REACTOR BUILDING
(..
ACTIONS INCLUDE A REAFFIRMATION OF HIGH RADIATION AREA LOCKED GATE REQUIREMINTS VITH ALL PLANT RADI ATION VORKERS (PB) AhD A TOUR OF THE PRIMARY CONTAlTMENT SHOWED THAT ND UNAUTHORIZED PERSONNEL VERE IN THESE AREAS. CORRECTIVE AND DISCIPLINARY ACT10h.
w FORM
........................ 0 8..- 0 8.. 9.1
....... 59................L. E. R S. CS.S..D. AT A DOCKET YEAR LER HUMBER REVISION DCS NUMBER NSIC EVENT DATE
. 333....1991..... 0.,04....... 0..... 9.104240274...... 2219 41...... 03. /.1 B./.9.1 re ABSTRACT POVER LEVEL - 002%. AN UNUSUAL EVENT VAS DECLARED AT 1445 ON 3/18/91 AFTER THE UNMONITOP.ED RELEASE OF RADIDACTIVITY VAS DETECTEb OUTSIDE THE AUXILI ARY B0!LER BUILDING. CONTAMINATION VAS DETECTED CN THE GROUND, BUILDING VALLS, AND ROOFS. DUE TO RAIN SHOWER ACTIVITY, RAIN VATER FROM CONTAM!kATED BUILDING RDOFS DRAINED TO THE STORM DRAINAGE SYSTEM, RESULTING IN AN UNMONITOREO RELEASE INTO LAKE ChTAR10. THIS EVENT RESULTED FROM N
ACTIVITIES ASSOCIATED VITH THE LIQUID RADIDACTIVE WASTE CONCEN1RATOR (WD). THE CONCENTRATOR VASTE LINE VAS PLUGGED AND OPERATORS VERE PERF0241NG TROUBLESHOOTING ACTIVITIES TO CLEAR THE LINE. STEAM FROM THE AUXILI ARY B0llER (SA) PROVIDES PROCESS STEAM 13 THE VASTE CONCENTRATOR AND THE CONDENSATE IS RETURNED TO THE BOILER. DUE TO INAPPROPRI ATE ACTIONS BY THE OPERATORS. THE CONDENSATE RETURN BECAME CONTAMINATED VITH RADIDACTIVE CONCENTRATED WASTE. THE RAD 10ACT!vE VASTE ESCAPED WITH THE VENTING STEAM VIA THE ATMOSPHERIC VENT ON THE AUXILIARY B0!LER ROOF. AT APPROXIMATELY 1315 THE VENT VAS CLOSED, VHICH TERM! HATED THE SOURCE CF UNMONITORED RELEASE. THE STORM DRAINAGE SYSTEM VAS BLOCKED AT 1939. TERMINATING THE UNMONITORED RELEASE TO LAKE DATAR10.
ACTIVITIES DURING AND AFTER THE UNUSUAL EVENT INVOLVED PREVENTING CONTINUED RELEASE TO THE ENVIRONMENT, CLEAN-UP. DECONTAMINATION, AND RDOI CAUSE OF THE EVENT.
FO.RM 6................L.E. R S.C.S S..D AT A.................................1
...r..
0 08-0B-9
~
DOCAts YEAR LER NUMBER REvlSION DCS NUMBER NSit EVENT DATE
. 333....1991..... 005..........0.... 9105210193.. 22.1997...... 04/.12./.9.1 ABSTRACT POVER LEVEL - 000%. THE PRIMARY CONTAINMENT (NH) SUPPRESSION P00L (TORUS) VATER TEMPERATURE INSTRUMENTATION VAS M001FIED IN 1981 TO PROVIDE FOR ONE DUAL ELEMENT RESISTANCE TEMPERATURE DETECTOR (RTD) IN EACH OF THE 16 BAYS OF THE TORUS.. AN EVALUATION OF THE SYSTEM ON 4/11/91 FOUND THAT A MAXIMUM OF ONE DEGREE OF NON-CONSERVATIVE--
j ERROR COULO BE INTRODUCED DURING CONDITIONS OF NONUNIFORM ADDITION OF HEAT TO THE TORUS PROVIDING TORUS COOLING MODE OF RESIDUAL HEAT REMOVAL SYS1EM VAS IN OPERATION. THIS DATA WAS THEN INCLUDED IN AN INSTRUMENT CHAhhEL UNCERTAINTY CALCULATION PERFORMED ON 4/12/91. A C0hSERVATIVE MAXIMUM INSTRUMENT CHANNEL UNCERTAINTY OF 3.52 FAHRENHEIT DEGREES VAS CALCULATED, H0VIVER, THIS 15 VITHIN A CALCULATED 4 DEGREE MARGIN 10 THE NUREG 0783 PEAK LDCAL TEMPERATURE LIMIT. CONSERVATIVELY, THE TGRUS WATER TEMPERATURE INSTRUMENTATION IS BEING RFPORTED AS INOPERABLE. THE TORUS VATER TEMPERATURE INSTRUMENT.50FTVARE VAS REPROGRAMMED TO ADD A 4 DEGREE BI AS SO THAT THE CALCULATED BULK (AVERAGE) TEMPERATURE WOULD BE DISPLATED Al A VALUE OF 4 DEGREES ABOVE THE VATER TEMPERATURE ACTUALLY 5ENSED BY THE RTDS. ANALYSIS SHOVS THAT EVEN IF TH!$ MAXIMUM UNCERTAINTY HAD OCCURRED IN THE NON-CONSERVA?!VE (LOV) DIRECTION. THE RESULTING INDICATED TEMPERATURE V0VLD NOT HAVE HAD AN ADVERSE EFFECT ON OPERATOR ACTIONS IF EVENTS MANAGED BY THL EMERGENCY OPERATING PROCEDURES HAD OCCURRED.
B - 20
'o I)lSCLAIMER E
E This arport was prepared as an account of work sponsored by an agency of the United States Go$crnment. Neither the United States Gosernment nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or rrsponsibility for the accuracy, completeness, or usefulness of any Information, apparatus, prc. duct or process disclosed, or typresents that its use would not infringe privately owned r!ghts. Iteferences herein to any specific commerdal product, process, or se:Tice by trade name, trademark, manufactuirr, or otherwise, does not necessarily constitute or imply its end.en ment, treommen dation, or favoring by the United States Geernment or :ity agency thereof.
E The views and opinioi:s of the authors expressed herein do not necessarily state or reflect tho,e of AEOD, ar other NitC office, or any other gmernment agency.
U 5
U E
11 R
R R
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