Text

Amended SALP Repts 50-245/86-99 & 50-336/86-99 for June 1986 - Dec 1987
	ML20151H720
Person / Time
Site:	Millstone
Issue date:	02/25/1988
From:	; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20151H718	List: IR 05000245/1986099; ML20151H716;
References
	50-245-86-99, 50-336-86-99, NUDOCS 8808020015
	Download: ML20151H720 (148)
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ENCLOSURE 1 i

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! U.S. NUCLEAR REGULATORY COMMISSION

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REGION I :

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1 SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE

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! INSPECTION REPORT NUMBERS 50-245/86-99 and 50-336/86-99 (AMENDED REPOR') !

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j MILLSTONE NUCLEAR STATION, UNITS I & II [

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ASSESSMENT PERIOD: June 1. 1986 to December 31, 1987 l

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l BOARD MEETING DATE: February 25, 1988 l

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. GGOG020015 800715 PDR ADOCK 05000045 [

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, SUMMARY MILLSTONE _1 AREA HOURS P. OF TIME PLANT OPERATIONS 1019 3 RADIOLOGICAL CONTROLS 297 1 MAINTENANCE 174 SURVEILLANCE 438 1 EMERGENCY PREP 138 SEC/ SAFEGUARDS 77 OUTAGE MANAGEMENT 265 * *

TRAINING EFFECTIVENESS

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ASSURANCE OF QUALITY ENGINEERING SUPPORT 263 y TOTALS: 2671 10 .

INSPECTION HOUR SUWARY MILLSTONE 2 AREA HOURS $ OF TIME PLANT OPERATIONS 1065 3 RADIOLOGICAL CONTROLS 265 MAINTENANCE 181 $URVEILLANCE 397 1 EMERGENCY PREP 148 SEC/ SAFEGUARDS 84 OUTAGE MANAGEMENT 280 1 * *

TRAINING EFFECTIVENESS

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ASSURANCE OF QUALITY ENGINEERING SUPPORT 277 1 TOTALS: 2697 10 *The inspection hours for these composite assessments are incorporated in the 8 functional area T-1 I

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TABLE 1A SYNOPSIS OF INSPECTION _ REPORTS MILLSTONE _ UNITS 1 AND_2 REPORT NUMBERS UNIT 1 UNIT 2 TYPE TOTAL INSPECTION DATES INSPEC HOURS DESCRIPTION 86-09 86-09 RESIDENT 308 PLANT OPERATION, SURVE!LLANCE, RAINTENANCE, S/20-7/7/86 MAIN TURBINE INSPECTION, AND STATIC "0" RING O!FFERENTIAL PRESSURE SWITCHES 86-10 -

SPECIALIST 104 RESPONSE, SUBSEQUENT ANALYSIS AND MODIFI-6/23-27/86 CATIONS OF MASONRY WALLS IN RESPONSE TO IE BULLETIN 80-11, MASONRY WALL DESIGN

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86-10 SPECIALIST 0 OPERATOR LICENSING EXAMINATIONS OF 8 SRO 7/7-11/86 AND 7 R0 CANDIDATES 86-11 86-11 SPECIALIST 48 RADI0 CHEMICAL MEASUREMENTS PROGRAM USING 6/2-6/86 REGION I MOBILE RADIOLOGICAL MEASUREMENT LABORATORY 86-12 86-12 SPECIALIST 54 PERSONNEL RADIATION TRAINING AND QUALIFI-7/7-11/86 CATIONS, EXPOSURE CONTROL, SURVEYS, AUDITS, ALARA, PREVIOUSLY IDENTIFIED ITEMS 86-13 86-13 LZSIDENT 190 PLANT OPERATION, SURVE!LLANCE KAINTENANCE, 7/8-8/18/86 RADIATION PROTECTION, PHYSICAL SECURITY, FIRE PROTECTION, IE BULLETINS

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86-14 SPECIALIST 31 SURVE!LLANCE TESTING AND PROCEDURES, CALI-7/7-11/86 BRATION CONTROL, QA/QC CONTROL INTERFACES o AND PREVIOUS INSPECTION FINDINGS 86-14 86-15 SPECIALIST 36 NOTIFICATION AND COMMUNICATION EQUIPMENT, 7/7-10/86 PROCEDURES, FOLLOW-UP OF EMERGENCY PRE-PAREDNESS ITEMS FROM PREVIOUS INSPECTIONS 86-15 86-16 SPECIALIST 40 IMPLEMENTATION OF INTEGRATED SITE SECURITY 7/14-18/86 PROGRAM 86-16 86-17 SPECIALIST 70 OVALITY ASSURANCE PROGRAMS FOR RECEIPT /

7/21-8/8/86 STORAGE & HANDLING OF FUEL, PROCUREMENT CONTROL, PLANT DESIGN CHANGES, MODIFICA-TIONS T-1A-1

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Table 1A REPORT NUMBERS UNIT 1 UNIT 2 TYPE TOTAL INSPECTION DATES INSPEC HOURS DESCRIPTION

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86-18 SPECIALIST 57 PREPARATIONS FOR REFUELING INCLUDING NEW 8/11-14/86 FUEL RECEIPT AND TRAINING FOR REFUELING 86-17 86 19 RESIDENT 91 OPERATICN, SURVEILLANCE, MAINTENANCE, 8/18-9/29/86 RADIATION PROTECTION, SECURITY, FIRE PRO-TECTION, IE BULLETINS, & U-1 STANDBY GAS TREATMENT SYSTEM 86-18 -

SPECIALIST 33 MAINTENANCE PROGRAM AND PROCEDURES, ELEC-9/22-26/86 TRICAL, MECHANICAL AND INSTRUMENTATION MAINTENANCE TASKS, QA/QC CONTROL INTERFACES

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66-20 SPECIALIST 45 KANAGEMENT CONTROLS, PERSONNEL SELECTION, 10/6-10/86 QUALIFICATION & TRAINING, EXTERNAL EXPOSURE CONTROL, ALARA 86-19 86-21 RESIDENT 271 U-1 OPERATIONAL SAFETY AND MAINTENANCE:

9/30-11/3/86 U-2 REFUELING OUTAGE INCLUDING REFUELING i OPERATIONS, LOCAL LEAK RATE TE$TS, SAFETY l VALVE TESTING ,

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SPECIALIST 0 CANCELLEO 10/19-11/20/86

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86-22 SPECIALIST 0 OPERATOR LICENSING EXAMINATION OF ONE R0 12/16/85-1/30/87 AND ONE 5RO CANDIDATES 86-21 -

SPECIALIST 0 OPERATOR LICENS!NG EXAMINATIONS OF 9 R0 12/5/86-2/15/87 AND 2 SRO CANDIDATES 86-22 66-23 RESIDENT 243 PLANT OPERATION, OUTAGE ACTIVITIES, SUR-11/4/85-1/5/S7 VE!LLANCE, PERIODIC REPORTS, AND MAINTENANCE

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86-24 SPECIALIST 34 E00Y CURRENT TESTING OF STEAM GENERATOR 11/3-7/86 TUBES INCLUDING 15! PROCEDURES, EQUIPMENT, QUALITY CONTROL MEASURES, DATA COLLECTION RECORDS

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86-25 SPECIALIST 0 OPERATOR LICENSING REQUALIFICATION PROGRAM 11/12/86-1/31/87 AUDIT 86-23 S6-29 SPECIALIST 82 OBSERVATION OF LICENSEE'S ANNUAL EMERGENCY 11/18-21/86 PREPARE 0 NESS EXERCISE OF 11/19/86 AND IN-GESTICN PATHWAY EXERCISE OF 11/20/86 T-1A-2

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Table 1A REPORT NUMBERS UNIT 1 UNIT 2 TYPE TOTAL INSPECTION DATES INSPEC HOURS DESCRIPTION 86-24 86-26 SPECIALIST 26 NON-LICENSE 0 STAFF TRAINING PROGRAM 11/17-20/86

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86-27 SPECIALIST 100 LICENSEE RESPONSES, SUBSEQUENT ANALYSES 12/8-12/86 AND M00!FICATIONS OF KASONRY WALLS RELATED TO IE BULLETIN 80-11, MASONRY WALL DESIGN

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86-28 SPECIALIST 22 TEST WITNESSING AND PRELIMINARY RESULTS 12/2-5/8e EVALVATION OF LOCAL LEAK RATE TEST, PRE-VIOUS ITEMS, COMMITMENTS FOR CONTAINMENT ISOLATION VALVE PM 86-25 86-30 SPECIALIST 18 0FF-SITE REVIEW COMMITTEE (NUCLEAR REVIEW 12/1-5/86 BOARDS) ACTIVITIES

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86-31 SPECIALIST 67 CYCLE 8 STARTUP PHYSICS TESTING INCLVOING 12/8-17/86 REVIEW OF TEST PROGRAM, PRECRITICAL TESTS,

& LOW POWER PHYSIC TESTS 86-26 86-32 SPECIALIST 4 DEGRADE 0 PROTECTIVE AREA BARRIER AND 12/11-12/86 LICENSEE'S CORRECTIVE ACTICNS 87-01 87-01 RESIDENT 117 PAEVIOUS ITEMS, U-2 SHUTDOWN IE INF0P.MA-1/6-2/9/87 110N NOTICES AND BULLETINS, U-1 LERs, ELECTRICAL BUSWORK INSULATION, OPERATOR REQUALIFICATION 87 02 87-02 SPECIALIST 8 PROTECTION OF SAFEGUARDS INF0D.MATION IN-1/27-29/87 CLUDING THE USE OF REQUIRED REPOSITORIES AND HANDLING PROCEDURES 87-03 87-03 RESIDENT 201 PREVIOUSLY IDENT!FIED ITEMS U-1 STANDBY 2/10-3/9/87 GAS TREATMENT INITIATION, U-1 EMER SERVICE WATER, U-1 APRMS, PORC, U-2 FIRE PROTECTION MEETING

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87-04 SPECIALIST 29 EDDY CURRENT EXAMINATION OF STEAM GENERA-2/3-6/87 TOR TUBES, PREVIOUSLY IDENTIFIED ITEMS, INSERVICE INSPECTION DATA 87-04 87-05 kESIDENT 221 OPERATIONAL SAFETY, U-2 FUEL RECONSTITUTION, 3/10-4/13/87 U-1 ESF ACTUATION, U-1 TRIP, NEW RAD WASTE TREATMENT, EDG FUEL OIL SUPPLY, PORC, RE-FORTS T-1A-3

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REPORT NUMBERS

UNIT 1 UNIT 2 TYPE TOTAL INSPECTION _ DATES INSPEC HOURS DESCRIPTION

87-05 -

RESIDENT 164 PLANT OPERATION, SURVEILLANCE, MAINTENANCE, 4/14-5/18/87 RAD PROTECTION, SECURITY, FIRE PROTECTION, 4 NEW FUEL RECEIPT, ZINC INJECTION TRIAL i l PROGRAM l l

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87-06 RESIDENT 111 PLANT OP, RAD PROTECTION, SECURITY, FIRE

4/14-5/18/87 PROTECTION, SURVEILLANCE / MAINTENANCE,

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DIESEL GENERATOR, AUXILIARY FEEDWATER, TRIP !

REVIEWS 87-06 87-07 SPECIALIST 22 SECURITY PROGRAM RECORDS, REPORTS, PHYSICAL i 2/23-27/87 BARRIERS PROTECTIVE AREAS, POWER SUPPLIES, i ACCESS CONTROL, OETECTION AIDS, ALARM i i STATIONS !

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SPECIALIST 35 WATER CHEMISTRY CONTROL PROGRAM INCLUDING !

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2/23-27/87 MANAGEMENT CONTROL, PLANT CHEMISTRY SYSTEM, ,

4 SAMPLING / MEASUREMENT, PROGRAM IMPLEMENTA- !

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87-08 87-03 SPECIALIST 34 SOLIO RAD WASTE CLASS!FICATION, HANDLING,

! 3/9-13/87 AND TRANSPORTATION, RAD ENVIRONMENTAL !

MONITORING, RAD CHEMICAL ANALYSIS, AND j l

CHEMICAL QA CONTROL ;

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SPECI A'.IST 96 MAINTENANCE. TESTING, RECORDS, PROCEDURES, l

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VALVE DISK !

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87-09 SPECIALIST 30 MAINTENANCE CRGANIZATION, PROGRAM, ACTIVI- :

1 3/16-19/87 TIES, MEASURING AND TEST EQUIPMENT, TROUBLE l l REPORTING, INSULATION DEGRADATION, QA/QC

INTERFACES 4 87-10 87-10 SPECIALIST 16 BI0 ASSAY WHOLE BODY COUNTING PROGRAM IN- ;

S/18-20/87 CLUDING RESULT COMPARI5ON, PROCEDURE REVIEW, I DATA COMPARISON ;

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RESIDENT 136 PLANT CPERATION, SURVE!LLANCE, MAINTENANCE, I S/19-6/22/S7 RADIATION PROTECTION, PHYSICAL SECURITY,

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FIRE PROTECTION, OUTAGE PREPARATION, AL- t LEGATION i

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UNIT 1 UNIT 2 TYPE TOTAL INSPECTION DATES INSPEC HOURS DESCRIPTION

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87-11 RESIDENT 122 PLANT OPERATION, SURVEILLANCES, APPENDIX 5/19-6/29/87 R M00!FICATION, CONTROL BOARD ENHANCEMENT, k ALLEGATION RESPONSE, STEAM GENERATOR AN- !

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ALYSES 87-12 -

RESIDENT 183 PREVIOUS ITEMS, PLANT OPERATIONS, SURVEIL-6/23-8/10/87 LANCE, MAINTENANCE, RADIATION PROTECTION, PHYSICAL SECURITY, FIRE PROTECTION, ALLE- l GATION, EFS t I

87-13 87-12 SPECIALIST 21 EMERGENCY PREPARE 0 NESS PROGRAM 6/29-7/2/87

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SPECIALIST 65 SURVEILLANCE AND CALIBRATION PROGRAM IN- i 7/20-24/87 CLUDING CALIBRATION TESTING, CONTROL OF l

. MEASUREMENT AND TEST EQUIPMENT, QA/QC [

INVOLVEMENT ;

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87-13 RESIDENT il3 OPERATIONAL SAFETY, UNIT TRIP, PORC REVIEW, 6/30-8/17/87 SPENT FUEL POOL DIVING, AUXILIARY FEE 0 WATER j

SURVE!LLANCE, DIESEL SURVEILLANCES, PRE-REFUELING l I

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87-14 SPECIALIST 40 STEAM GENERATOR SURVEILLANCE, PREVENTIVE t 7/6-10/87 MAINTENANCE ACTIVITIES, ACTION 3 ON PRE- l VIOUSLY IDENTIFIED NRC ITEMS I I

87-15 87-17 :PECIALIST 117 RA0!ATION PROTECTION ACTIVITIES ASSOCIATED 7/6-10/87 WITH UNIT 1 OUTAGE, INTERNAL AND EXTERNAL EXPOSURE CONTROL, ALARA, POSTING LABELING

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SPECIALIST 36 !$! ACTIVITIES, AUGMENTED EXAMINATION PRO-7/6-10/87 GRAM FOR INTEGRATED STRESS CORROSION CRACK-ING, AND BALANCE OF PLANT EROSION / CORROSION PROGRAM 87-17 87-15 SPECIALIST 56 FOLLOW UP ON EQUIFMENT QUALIFICATION IN-7/15-20/87 INSPECTICNS 50-245/85-30 AND 50-336/85-35 INCLUDING CORPORATE FILES, CORRECTIVE AC-TIONS, AND VERIFICATION OF CONFORMANCE WITH 10 CFR 50.49

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87-16 SFECIALIST 154 TEAM INSPECTION OF THE LICENSEE'S EFFORT !

7/13-17/87 TO COMPLY WITH 10 CFR APPENDIX R, SECTIONS I

!!!.G, J. AND 0 CCNCERNING SAFE SHUT 00VN I AFTER A FIRE l

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i REPORT NUMBERS UNIT 1 UNIT 2 TYPE TOTAL INSPECTION _0ATES INSPEC HOURS DESCRIPTION l 87-18 -

SPECIAL!$T 73 CONTAINMENT INTEGRATED LEAK RATE TEST WIT- ,

7/31-8/7/87 NE551NG AND PRELIMINARY RESULTS EVALVATION !

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SPECIAL!$T 116 TEAM INSPECTION OF THE LICENSEE'S EFFORT l 8/17-21/87 TO COMPLY WITH 10 CFR APPENDIX R, SECTIONS i

,J 0 0 CONCERNING SAFE SHUTDOWN 87-20 87-18 SPECIALIST 36 RA0!0ACT!VE EFFLUENT CONTROL PROGRAM, f l 8/24-28/87 LIQUID AND GASEOUS WASTE SYSTEMS, PROCESS i l RAD MONITORING, AIR CLEANING SYSTEMS, AND 4 l AUDIT ACTIVITIES .

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RES!0ENT $9 PLANT OPERATIONS, MAINTENANCE, SURVE!LLANCE, 8/11-9/8/87 RA0!ATION PROTECTION, PHYSICAL SECURITY, FIRE PROTECTION, PERIODIC AND SPECIAL REPORTS 87-22 87-20 SPECIALIST 78 PROCEDURES, ORGAN!ZATION, PROGRAM AVOITS, 8/31-9/4/87 AND REPORTS, TESTING AND MAINTENANCE, PHYSICAL EARRIERS, LIGHTIN3, ACCESS CONTROL, SECURITY AIDS 87-23 -

SPECIALIST 0 OPERATOR LICEN5!NG EXAMINATION OF 7 SRO 9/21-10/25/87 CAN0! DATES 87-24 87-21 SPECIALIST 36 STATUS OF PREVIOUSLY IDENTIFIED !TEMS RE-9/14 24/87 LATED TO THE CAPABILITY FOR POST-ACCIDENT SAMPLING, MONITORING, AND ANALY$!$

87-25 87-19 RES!0ENT 95 OPERATIONAL SAFETY, AN ALLEGATION, U-1 CON-8/18-9/25/87 TROL R00'4 HALON TESTING, FA! LURE OF U-2 O!ESEL GENERATOR TO LOAD, U-2 CONTROL R00 ANOM.ALIES 87-26 87-22 SPECIALIST 100 ANNOUNCED EMERGENCY PREPARE 0 NESS TEAM IN-10/7-9/87 $PECTION AND OBSERVATION OF THE LICEN5EE'S ANNUAL EMER3ENCY EXERCISE PERFORMED ON 10/8/87 87-27 87-23 RESIDE 4T 114 FOLLOW UP CN PREVIOUS FINDINGS, PHYSICAL 9/26-10/26/87 SECURITY, PLANT OPERATIONS, DIESEL GENERA-TOR TRIPS, SURVE!LLANCE, MAINTENANC FEEC' DATER HYOROGEN INJECTION TESTIN3, AND IE EULLETIN 87-01 T-1A-6

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REPORT NUMBERS I UNIT 1 UNIT 2 TYPE TOTAL :

INSPECTION DATES INSPEC HOUR $ DESCRIPT!' - l l

87-28 87-24 $PECIALIST $6 NON-RA0!0 LOGICAL CHEMISTRY PROGRAM INCLUD-11/2-6/87 ING MEASUREMENT CONTROL AND ANALYTICAL

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$PECIALIST 0 CANCELLED !

11/3-20/87 j 87-30 87 25 RE5!0ENT 138 FOLLOW-UP ON PREVIOUS FINDINGS, SECURITY, !

10/27-11/30/87 OPERATIONS, SERVICE WATER OPERABILITY, DC I SWITCHGEAR VENTILATION, UNIT 2 TRIP, $UR- l VEILLANCE, COV.MITTEE ACTIVITIES, CONTROL E ROOM VENTILATION, FUEL ASSEMBLY PRES $URE !

DROP TEST, AND LERS j 87-31 87-26 SPECIALIST 16 PRIMARILY UNIT 3 OUTAGE INSPECTION, BUT 11/16-20/J7 WITH $0ME UNIT 1 AND 2 REVIEW OF TRAININ AND INTERNAL AND EXTERNAL EXPOSURE CONTROL l t

87-32 87-27 SPECIALIST 103 COvPLEX SAFETY-RELATED SYSTEM, IN-PLANT f'

11/30-12/4/87 INSTRUMENT cal.18 RATION, MEASURING AND TEST EQUIPMENT, COLD WEATHER PREPARATION, QUAL-ITY CONTROL INTERFACES l

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87-28 SPECI At.!5T 14 STEAM GENERATOR E00Y CURRENT INSPECTION, !

11/30-12/4/87 WATER CHEMISTRY CONTROLS, RADIOLOGICAL CON- !

TROLS DURING STEAM GENERATOR INSPECTION / i REPAIR {

i 87-33 87-29 RESIDENT 159 PREVIOUS INSPECTION FINDINGS, PHYSICAL i 12/1-31/87 SECURITY, PLANT OPERATIONS, IMPLEMENTATION l

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0F LICENSE AMENCMENTS, IE BULLETIN 87-02 -

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FASTENER TESTING SURVEILLANCE TESTING, f

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SCRAM O!$ CHARGE VOLUME MODIFICATIONS, COM- 5 MITTEE ACTIVITIES, AND l.!CENSEE EVENT RE- f PORTS j

87-34 87-30 SPECIALIST T9 50L10 RADWASTE AND TRANSPORTATION PROGRAM [

12/7-11/87 INCLUDIN3 KANAGEMENT CONTROL, SHIFMENTS {

OF RADI0 ACTIVE MATERIALS, TRAINING, PRO- i

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CES$1NG. PACKAGE SELECTION AND QUALITY CONTROL 87-35 -

SPECIALIST 34 LICENSEE'S RESPONSE TO GENERIC LETTER S4-11, 12/14-18/S7 INTERGRANULAR STRESS CORROS!ON CRACKIN3 OF GWR RECIRCULATION SYSTEM AND ASSOCIATED PIPING T-1A-7 l

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ENFORCEMENT SUMMARY t t

MILLSTONE 1 VIOLAT!0NS !

SEVERITY LEVEL AREA 1 2 3 4 5 DEV TOTAL ,

PLANT OPERATIONS 1 2 3 RA010 LOGICAL CONTROLS 2 2 MAINTENANCE SURVE!LLANCE 1 1 EMERGENCY PREP '

SEC/ SAFEGUARDS 1 2 3 OUTAGE RANAGEMENT TRAINING EFFECTIVENESS ASSURANCE OF QUALITY ENGINEERING SUPPORT 1 1 j i

TOTALS: 1 5 4 11 MILLSTONE 2 VIOLATIONS

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SEVERITY LEVEL [

AREA 1 2 3 4 5 DEV TOTAL i

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PLANT OPERATIONS 1 1 !

RA0!0 LOGICAL CONTROLS j MAINTENANCE I SURVE!LLANCE j i EMERGENCY PREP i l 5EC/5AFEGUARDS 1 2 3 ,

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OUTAGE MANAGEMENT 1 1 i

' TRAINING EFFECTIVENESS ASSURANCE OF QUALITY ENGINEERING SUPPORT 2 2 4

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I TABLE 2A SYNOP5!$ OF VIOLATIONS i I

MILLSTONE 1 AND 'l l REPORT NUMBERS UNIT 1 UNIT 2 REQUIREMENT SEVERIT.Y FUNCT!DNAL j INSPECTION DATES VIOLATED _ LEVEL, AQEA OESCRIPTION

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86-26 86-32 MP SECURITY 4 SEC/5AFEGR05 DEGRADATION OF THE PROTECTED !

12/11-12/86 PLAN AREA BARRIER !

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87 02 87-02 10 CFR 4 SEC/5AFEGR05 FA! LURE TO PROPERLY SECURE !

1/27-29/87 73.21(d)(2) UNATTENDED SAFEGUARD $ IN- ,

FORMATION IN A LOCKED I'

SECURITY STORAGE CONTAINER 87-05 - APPENDIX B, 5 OPERATIONS FAILURE TO UPDATE TECHNICAL 4/14-5/18/87 CRI XVI TECHNICAL SPECIFICATION TABLE 3.7.1 TO INCLUDE CON- f TAINMENT ATMO5PHERE SAMPLE i LINE ISOLATION VALVES !

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TECH SPEC 5 OPERATICNS FA' LURE TO UFCATE TECHNICAL [

4/14-5/18/87 3,6. SPECIFICATION TABLES 3.6. l l AND 3.6.1 B TO CORRECT !

l 5AFETY-RELATED 3NUEBER !

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LISTING 87-15 -

10 CFR 5 RAD CONTROL SHIPP!NG B0( CONTAINING 7/6-10/87 20.203(f) RADICACTIVE RATERIAL AND LOCATED IN THE RAILWAY

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l ACCESS AREA WAS NOT LABELED

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AS REQUIRE 0

87-15 -

TECH SFEC 5 RAD CONTROL THREE CASES OF WORKER (5)

7/6-10/87 6.11 NOT REA0!NG AND/0R FOLLOWING RADIATION WORK PERMITS

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87-15 10 CFR 50.49 4 ENG $UPPORT INADEQUATE EQU!FMENT QUAL 7/15-17/87 (f) AND (L) DOCUMENTATION OF GE $!5 WIRE i USED IN vat.VES 2-5!-654, 2 CH-501. & 2-51-644 ,

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87-15 10 CFR 50.49 4 ENG $UPPORT INA0 EQUATE EQU!FMENT QUAL 7/15-17/87 (1) 0F E!$HCP CABLE SPLICE CN MOTOR OPERATED VALVE 2-5!-654 CN RAY 31, 1937 T-2A-1

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Table 2A REPORT NUMBERS UNIT 1 UNIT 2 REQUIREMENT SEVERITY FUNCTIONAL INSPECTION DATES VIOLATE 0_ LEVEL AREA DESCRIPTION 87-17 -

10 CFR 50.49 4 ENG SUPPORT INADEQUATE EQUIPMENT QUAL 7/15-17/87 (e)(1) 0F CURTIS L-TYPF TERMINAL BLOCKS USED IN ISOLATION CONDENSER VALVE I-IC-I

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87-16 APPENDIX R , 5 ENG SUPPORT FIRE BARRIER SEPARATING THE 7/13-17/87 SEC IIIG2 WEST ELECTRICAL PENETRATION ROOM FROM THE AUXILIARY BUILDING DID NOT MEET RE-QUIREMENTS (N0 FIRE DAMPER)

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87-16 APPENDIX R, 5 ENG SUPPORT INADEQUATE DISTANCE SEPA-7/13-17/87 SEC IIIG1 RATING THE REQUNDANT AUXILI-ARY FEEDWATER HEADERS AND THEIR ISOLATION VALVES WITH INTERVENING COMBUSTIBLES 87-21 -

TECH SPEC 4 SURVEILLANCE FAILURE TO PERFORM INDEPEN-8/11-9/8/87 6.8. DENT VERIFICATION OF TEST EQUIPMENT FOR AUTO BLOWDOWN LOGIC AND FAILURE TO IM-PLEMENT MAIN STEAM LINE ISOLATION VALVE CLOSURE TEST 87-22 87-20 MP SECURITY 3 SEC/SAFEGRDS MULTIPLE EXAMPLES OF INADE- .

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8/31-9/4/87 P LAN QUATE PROTECTED AND VITAL AREA BARRIERS, TWO EXAMPLES OF VISITORS WITHOUT ESCORT, IMPROPER COMPENSATORY MEAS-URES, AND OTHER ISSUES

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87-25 10 CFR SO 4 MAINTENANCE REDUNDANT VENTILATION 10/27-11/30/87 APPENDIX B COOLERS FOR VITAL DC SWITCH-GEAR ROOMS INOPERABLE SINCE 1983 87-33 -

10 CFR 4 OPERATIONS FAILURE TO NOTIFY THE NRC 12/1-31/87 50.72(b)(2) THAT 8 0F 12 CHECK VALVES IN THE NITROGEN SUPPLY T0 i THE AUTOMATIC BLOWOOWN <

SYSTEM FAILED TO PASS THE LOCAL LEAK RATE TEST 87-29 TECH SPEC 5 0UTAGE FAILURE TO APPROVE EXCESS 12/1-31/87 6.2. MANAGEMENT OVERTIME (7 EXAMPLES) PER GUIDELINE 5 DURING AN OUTAGE l

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TABLE 3 SUMMARY OF LICENSEE EVENT REPORTS (LERs)

MILLSTONE 1 AREA CAUSE CODES CODE AREA A @ C D E TOTAL

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1 PLANT OPERATIONS 3 1 3 3 10 2 RADIOLOGICAL CONTROLS 2 1 3 3 MAINTENANCE 1 1 4 SURVEILLANCE 5 4 1 1 11 5 EMERGENCY PREP 0 6 SEC/ SAFEGUARDS 8 5 2 1 7 23 7 OUTAGE MANAGEMENT 0 8 TRAINING EFFECT 1 1 9 ASSUllANCE OF QUALITY 0 10 ENGINEERING SUPPORT 4 11 15 TOTALS: 24 21 2 5 32 64 SUMMARY OF LICENSEE EVENT REPORTS (LERs)

MILLSTONE 2 AREA CAUSE CODES CODE AREA A @ C D E TOTAL 1 PLANT OPERATIONS 3 2 6 11 2 RADIOLOGICAL CONTROLS 1 1 3 MAINTENANCE 5 5 4 SURVEILLANCE 2 1 3 6 5 EMERGENCY PREP 0 6 SEC/ SAFEGUARDS 7 3 2 6 18 7 OUTAGE MANAGEMENT 1 1 2 8 TRAINING EFFECT 0 9 ASSURANCE OF QUALITY 2 2 10 ENGINEERING SUPPORT 2 6 8 TOTALS: 20 13 2 1 17 53 CAUSE CODES A -- PERSONNEL ERROR B -- DESIGN MANUFACTURING, CONSTRUCTION / INSTALLATION C -- EXTERNAL CAUSE O -- DEFECTIVE PROCEDURE E -- EQUIPMENT FAILURE X -- OTHER T-3-1

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TABLE 3A SYNOPSIS OF LICENSEE EVENT REPORTS (LERs)

HILLSTONE 1 LER C'!ENT CAUSE AREA NUMBER DATE CODE CODE DESCRIPTION 86-17 5/21/86 E* 1 REACTOR MANUALLY TRIPPED FOLLOWING FAILURE OF MECHANICAL PRESSURE REGULATOR DURING PLANNED REACTOR SHUTOOWN TO CONDUCT TURBINE INSPECTION 86-18-01 5/24/86 B* 10 WITH UNIT SHUTP"/N, REACTOR PROTECTION ACTUATION DUE TO SOURCE RiNGE MONITOR DRIVE RELAYS CAUSING NOISE SPIKES ON INTERMEDIATE RANGE MONITORS 12 AND 16 86-19 5/31/86 A* 2 STANDBY GAS TREATMENT INITIATION CAUSED BY SPURIOUS UPSCALE TRIP 0F THE STEAM TUNNEL VEN-TILATION RADIATION MONITOR 86-25 11/14/86 B 10 NOTIFICATION THAT FEEDWATER COOLANT INITIATION RELAYS 00 NOT CONFORM TO SEISMIC QUALIFICATION 86-27 11/30/86 B 1 REACTOR TRIP ON GENERATOR TRIP CAUSED BY GENE-RATOR LOCK-00T DUE TO PHASE-TO-GROUND FAULT OF THE MAIN TRANSFORMER 86-28-01 12/3/86 B* 4 MAIN STEAM LINE LOW PRESSURE SWITCH SETPOINT ORIFT 86-29 12/6/86 E* 2 DURING SHUTDOWN, A STANDBY GAS TREATMENT ACTU-ATION CAUSED BY REACTOR BUILDING VENT PAD MONI-TOR FAILING HIGH DUE TO . AILED SENSOR / CONVERTER 86-32 12/30/86 E* 4 SURVEILLANCE OF CONDENSER LOW VACUUM SWITCHES FINDS 2-0F-4 SWITCHES WITH SETPOINT DRIFT DOWN-WARD 87-01-01 1/13/87 8 10 CRAC .ING ALONG THE HORIZONTAL NORYL INSULATORS OF 4160V OISTRIBUTION LOAD CENTER 87-04 2/1/87 D* 4 $URVEILIkN'E OF "B" STANDBY GAS TREATMENT OVFR-DUE BY f Pv0RS FOLLOWING DECLARATION THAT "A" SBGT WAS INOPERABLE 87-05 2/21/87 C* 1 STANDBY GAS TREATMENT SYSTEM INITIATION BY HIGH RADIATION IN THE STEAM TUNNEL DUE TO AIR BEING LEFT IN DEMINERALIZER "B" T-3A-1

.

Table 3A LER EVENT CAUSE AREA NUMBER DATE CODE CODE DESCRIPTION 87-07 3/22/87 A* 8 REACTOR TRIP AND ISOLATION ON LOW MAIN STEAM LINE PRESSURE DUE TO PRESSURE OSCILLATIONS CAUSED BY CONTROL PROBLEMS WITH THE MECHANICAL PRESSURE REGULATOR 87-08 3/10/87 A 3 REACTOR BUILDING VENT ISOLATION AND STANOBY GAS TREATMENT ACTUATION DURING INSTRUMENT TECHNICIAN WORK ON REACTOR BUILDING VENT RADIATION MONITOR 87-12-01 5/19/87 B* 10 EMERGENCY DIESEL GENERATOR CEILING FIRE C0ATING DISCOVERED INADEQUATE TO PROVIDE THE REQUIRED 3-HOUR FIRE RESISTANT RATING 87-13 5/27/87 0* 1 STANDBY GAS TREATMENT SYSTEM ACTUATED OUE TO HIGH RADIATION ON THE REFUELING FLOOR CAUSED BY AIR IN THE SPENT FUEL POOL COOLING SYSTEM AFTER FILLING AND VENTING 87-15-02 6/6/87 B* 4 SEVENTEEN CONTAINMENT ISOLATION VALVES, ItlCLUD-ING TWO MAIN STEAM ISOLATION VALVES, FAIL LOCAL LEAK RATE TEST 87-17 6/10/87 A* 1 REACTOR TRIP ON SCRAM VALVE AIR HEADER LOW PRESSURE OUE TO LARGE DEMAND ON STATION AIR SYSTEM AND TRIPPING OF SULLAIR AIR COMPRESSOR ON ELECTRICAL OVERLOAD 87-19 6/12/87 A 1 WHILE UNLOADING THE REACTOR CORE, FUEL ASSEMBLY LY2729 WAS FOUND MISORIENTED IN CORE LOCATION 43-18 87-20-01 6/26/87 B* 10 INTERGP. ANULAR STRESS CORROSION CRACKING INDICA-TION ON RECIRCULATION SYSTEM PIPE TO CAP WELD RMBJ-1 87-21 6/30/87 B* 10 5 0F 6 TARGET ROCK MAIN STEAM SAFETY RELIEF VALVE FOUND WITH SETPOINTS HIGHER THEN ALLOWE0 BY TECHNICAL SPECIFICATIONS 87-22 7/2/87 B* 10 BASE METAL INCLUSIONS APPR0XIMATELY 26 INCHES LONG FOUND IN THE ISOLATION CONDEf1SER RETURN LINE PIPING 87-23 7-08-87 B 10 AS-INSTALLE0 CONFIGURATION OF LOW PRESSURE COOLANT INJECTION AND CORE SPRAY SYSTEM PUMP FOUNDATIOr4 ANCHORS IN NONCONFORMANCE WITH ORIGINAL DESIGN T-3A-2

'

.

Table 3A LER EVENT CAUSE AREA NUMBER DATE CODE CODE DESCRIPTION 87-24 7/15/87 A* 2 STANDBY GAS TREATMINT ACTUATION ON REFUELING FLOOR HIGH RADIATION WHILE REPLACING LOCAL POWER RANGE MONITORS 87-26 8/3/87 B* 10 FAILUkc 0F NINE HYDRAULIC SNUBBER IN THE FIRST FEW ID% SAMPLES REQUIRED ALL HYDRAULIC SNUBBERS 10 BE TESTED IN ACCORDANCE WITH TECHNICAL SPECIFICATIONS 87-28 8/13/87 A* 4 REACTOR TRIP SIGNAL GENERATED BY INSTRUMENT TECHNICIAN WHlLE PERFORMING MAIN STEAM ISOLATION VALVE CLOSURE FUNCTIONAL TEST 87-29 7/24/87 A 10 STANDBY GAS TREATMENT SYSTEM IN0PERABLE DUE TO DEFEATED INTERLOCK ON ATMOSPHERIC CONTROL VALVE 1-AC-10 (VALVE REMOVED FOR MAINTENANCE)

87-30 7/26/87 B* 10 REACTOR TRIP SIGNAL, FROM THE INTERMEDIATE RANGE MONITORS 12 AND 16, WAS GENERATED AS SOURCE RANGE CHANNEL 23 WAS BEING DRIVEN IN 87-31 7/28/87 D* 1 REACTOR TRIP SIGNAL DUE TO INTERMEDIATE RANGE MONITOR SPIKE CAUSED BY INSTRUMENT TECHNICIAN MOVING NUCLEAR INSTRUMENT CABLES UNDER THE REACTOR VESSEL 87-32 8/11/87 B* 4 ALL FOUR TURBINE IST STAGE PRESSURE BYPASS

$ WITCHES FAIL TO MEET TECHNICAL SPECIFICATIONS SETPOINT REyv!REMENTS 87-33 8/12/87 A 4 DURING SHUTDOWN, INADVERTENT ACTU/, TION OF "A" ,

LPCI SUBSYSTEM 00E TO TEST SIGNAL INJECTION 87-34 8/14/87 A 1 REACTOR TRIP OURING STARTUP ON INTERMEDIATE RANGE HIGH FLUX DURING WITHDRAWAL OF CONTROL R0D 26-31 87-35 8/21/87 A 10 SIX FIRE DETECTION SYSTEM NOT COMPLETELY ELEC-TRICALLY SUPERVISED AND NOT DEMONSTRATED OPER-ABLE EACH 31-DAYS PER TECHNICAL SPECIFICATIONS 87-36 8/26/87 A* 4 REACTOR TRIP OURING AVERAGE POWER RANGE MONITOR SURVEILLANCE TESTING 87-37 9/8/87 A* 4 MANUAL REACTOR TRIP FUNCTION SURVEILLANCE NOT PERFORMED ON TIME T-3A-3

'

.

, -

. .

Table 3A LER EVENT CAUSE AREA NUMBER DATE CODE CODE DESCRIPTION 87-38 9/3/87 E* 1 REACTOR TRIP ON LOW SCRAM HEADER PRESSURE CAUSED BY LOW SERVICE AIR HEADER PRESSURE DUE TO SER-VICE AIR COMPRESSOR FAILURE DURING HIGH SERVICE AIR USAGE 87-39 9/21/87 A* 4 SURVEILLANCE FOUND PAST DUE ON AUTOMATIC PRES-SURE RELIEF AND LOW PRESSURE CORE COOLING PUMP INTERLOCK 87-40 9/15/87 B* 10 ALL FOUR NEW (INSTALLED DURING 1987 OUTAGE)

CONDENSER LOW VAOlVM TRIP PRESSURE SWITCHES FAILED TO MEET TS SETPOINT REQUIREMENTS 87-41 10/16/87 A* 10 FAILURE TO REQUEST TECHNICAL SPECIFICATION I

'

CHANGE FOR REMOVAL OF LOW REACTOR PRESSURE PERMISSIVE SWITCHES FROM LOW PRESSURE INJECTION AND CORE SPRAY PUMP START LOGIC 87-42 10/27/87 A 10 DURING REVIEW 0F IE INFORMATION NOTICE 86-60, IT WAS DETERMINED THAT NO SURVEILLANCE EXISTED FOR TESTING THE POST ACCIDENT SAMPLING SYSTEM PER TECHNICAL SPECIFICATION 6.13 87-43 11/16/87 E* 1 TWO HYORAULIC SNUBSERS HAD LOW RESERVOIR FLUID LEVELS: BENCH TESTING RESULTED IN DECLARING THEM INOPERABLE DUE TO SLIGHTLY HIGH LOCKUP RATES IN COMPRESSION 87-44 12/29/87 B* 4 TECHNICAL SPECIFICATION REQUIRED TESTING OF GAS TREATMENT SYSTEM NOT FULLY SATISFIED IN THAT NO FLOW OISTRIBUTION TEST WAS PERFORMED ACROSS THE CHARC0AL ABSORBERS

,

-- CAUSE CODES HAVE BEEN ASSIGNED BY OR CHANGES FROM THE LICENSEE CODES BY NRC REGION I ,

l l

!

,

l l

l >

T-3A-4 ,

,

!

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TABLE 3B SYNOPSIS OF LICENSEE EVENT REPORTS (LERs)

MILLSTONE 2 LER EVENT CAUSE AREA NUMBER DATE CODE CODE DESCRIPTION 86-03-01 S/16/86 B 1 EVALUATION IN RESPONSE TO IE INFORMATION NOTICE 83-69 IDENTIFIES 20 Ih0PERABLE FIRE DAMPERS 86-04-01 6/1/86 A* 1 REACTOR TRIP ON REACTOR COOLANT PUMP UNDERSPEED CAUSE0 BY LOSS OF POWER TO BUS 25B DUE TO IM-PROPER OPERATION OF BREAKER CONTROL SWITCH 252-2SB-2 86-05 8/12/86 8 10 REACTOR TRIP ON #1 STEAM GENERATOR LOW LEVEL AFTER LOSS OF THE "A" FEE 0 WATER PUMP 00E TO LOSS OF OIL PUMPS WHEN BUSSES 22A ANO 22B (CROSS-TIEO) LOST POWER 86-06 9/3/86 B 10 REACTOR TRIP ON LOW STEAM GENERATOR LEVEL DUE TO LOSS OF HEATER OPAINS FLOW FOLLOWING FAILURE OF AIR FITTING TO THE HEATER ORAINS CONTROL VALVE CLOSING VALVE 86-07 9/1/86 E* 4 SURVEILLANCE CHECK OF THE REMOTE SHUT 00WN PANEL FOUND TECH SPEC REQUIRED STEAM GENERATOR LEVEL TRANSMITTER LT-1113A OUT OF SERVICE

,

86-08-01 9/20/86 E' 4 SIX OF 16 MAIN STEAM SAFETY VALVES FAILE0 THE i SIMMER TEST DUE TO SETPOINT ORIFT 86-09-01 9/29/86 A* 3 TWO UNRELATED ESF ACTUATIONS ONE OUE TO PER-SONNEL ERROR AND THE OTHER DUE TO NOISE SPIKE IN RAD MONITOR RM-8262A i

86-10 10/6/86 A* 10 INCONSISTENCY BETWEEN THE NUMBER OF RCS PUMPS REQUIRED TO BE OPERATING IN MODES 3, 4 AND 5 AND THE ASSUMPTIONS USED IN THE SAFETY ANALYSIS

[

86-11 10/4/86 A* 1 TWO CASES OF IMPROPER FIRE WATCH COVERAGE RE- !

QUIRED BY TECH SPEC 3.7.10.A DURING REFUELING ,

86-12-01 10/9/86 E* 4 TYPE B AND C LOCAL LEAKAGE RATE LIMITS EXCEEDED I

86-13 10/10/86 B* 10 SAFETY INJECTION TANK "A" LEVEL TRANSMITTER FOUND OUT OF SPECIFICATION TO THE LOW SIDE

T-38-1

.

O O

.

Table 3B LER EVENT CAUSE AREA NUMBER DATE CODE CODE DESCRIPTION 86-14 10/29/86 A* 4 TWO ACTUATIONS OF THE CONTAINMENT PURGE ISOLA-TION SYSTEM CAUSED BY: 1) ELECTRONIC NOISE IN RM 8123A, AND, 2) TECHNICIAN ERROR 86-15-01 11/14/86 8 10 GENERAL ELECTRIC MODEL 12 DIESEL GENERATOR DIF-FERENTIAL RELAYS NOT SEISMICALLY QUALIFIED FOR CLASS 1E SERVICE 86-16 11/4/86 E" 7 SCHEDULED INSERVICE EXAMINATION OF STEAM GENE-RATORS IDENTIFIED SUFFICIENT NUMBER OF TUBES WITH FLAWS GREATER THAN 40*4 THROUGH-WALL 86-17 11/5/86 A* 3 DURING SHUTDOWN, LOSS OF POWER EVENT INITIATION BY TESTMAN CAUSING A PERCEIVED MAIN GENERATOR GROUND FAULT RESULTING IN OPENING OF SWITCHYARD BREAKERS 86-18 12/10/86 B* 10 PLANNED REMOVAL OF 14 HYORAULIC AND 7 HECHANICAL SNUBBERS HAVING MOVEMENTS LESS THAN 1/16 INCH:

SNUBBERS WERE REPLACED WITH RIGIO SUPPORTS 86-19 11/13/86 0* 4 OURING SHUTDOWN, OPERABILITY $URVEILLANCE OF THREE RUSKIN MODEL HVD-1-173 FIRE DAMPER HAS BEEN MISSED SINCE 1980: WERE NOT CN SP 2618G FORM 86-20 11/29/86 A 3 DURING SHUTOOWN, TWO CASES OF LOSS OF POWER ON LOAD CENTER 240 BEING SENSED BY AN IMPROPERLY INSTALLED BUS VOLTAGE POTENTIAL TRANSFORMER ORAWER ,

86-21 12/31/86 8 1 DURING SHUT 00WN, 8 VALCOR SOLEN 0!D VALVE IN THE REACTOR COOLANT VENT SYSTEM WERE LEAKING BY DUE TO SPRING FAILURES 86-22 12/23/86 A* 3 REACTOR TRIP ON LOW STEAM GENERATOR LEVEL DUE TO FEEDWATER PUMP SPEED DECREASE TO MINIMUM UPON LOSS OF POWER ON BUS 24C, CAUSED BY IMPROPERLY INSTALLED ORAWER 86-23 12/13/86 8 9 "C" CHARGING PUMP CRACKED BLOCK DUE TO HIGH INTERNAL STRESS CAUSING CRACKS TO INITIATE AT SUB SURFACE INCLUSIONS

.

87-01-01 12/22/86 E* 1 FIRE DETECTION / PROTECTION SYSTEMS FOR THE "C" REACTOR COOLING PUMP INDICATED OUT OF SERVICE DUE TO HEAT DETECTOR FAILURE l T-3B-2 i

'

l

-_ -- . . ___ - _ . . - -_ - _ _ - . - - O

, - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - --- --------- - - - . - - - - - - - - - - - - - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

.

C

.

Table 3B LER EVENT CAUSE AREA NUMBER DATE CODE CODE DESCRIPTION 87-02 1/2/87 A* 3 REACTOR TRIP ON LOW STEAM GENERATOR LEVEL FOL-LOWING LEVEL CONTROL PROBLEMS DUE TO A H0T JUMPER ARC ON THE FIRE SUPPRESSION ALARM PANEL 87-03 1/29/87 A* 7 POST OPERATIONAL REVIEW 0F E00Y CURRENT DATA IDENTIFIED TWO DEFECTIVE STEAM GENERATOR TUBES NOT REPAIRED PRIOR TO STARTUP 87-04-01 2/2/87 E* 2 DURING SHUTOOWN, TWO CASES OF ISOLATION OF CON-TAINMENT PURGE SYSTEM OCCURRED DUE TO AUTOMATIC ACTUATION OF ESAS 87-05 3/6/87 8 9 "B" CHARGING PUMP CRACKED BLOCK OUE TO HIGH IN-TERNAL STRESS CAUSING CRACKS TO INITIATE AT SUB-SURFACE INCLUSIONS 87-06 4/3/87 B* 10 FSAR TABLE ERROR RESULTED IN SERVICE WATER FLOW THRU RBCCW HEAT EXCHANGER BEING INSUFFICIENT FOR DESIGN HEAT REMOVAL 87-07 4/16/87 E* 1 REACTOR TRIP ON TURBINE TRIP CAUSED BY GENERATOR EXCITER FIELD BREAKER AND GENERATOR CREAKERS OPENING, CAUSE UNKNOWN 87-08 6/11/87 A* 4 LATE SURVEILLANCE OUE TO SCHEDULING ERROR FOR BATTERIES 201A&B (SURVEILLANCE 27368-1)

87-09 9/2/87 E* 1 REACTOR TRIP ON #1 STEAM GENERATOR LOW LEVEL DUE TO FAILURE OF FEE 0 WATER CONTROL VALVE

2-FW-51A, THE PLUG HAD SEPARATED FROM THE STEM 87-?,0 7/10/R7 A* 10 MAIN CABLE VAULT AND RACEWAY TO CHARGING PUMPS FIRE PROTECTION SUPPORTS NOT ADEQUATELY PROTECTED 87-11 7/23/87 E* 1 REACTOR TRIP ON #1 STEAM GENERATOR LOW LEVEL DURING A DOWN-POWER EVOLUTION IN RESPONSE TO DECREASING REACTOR PRESSURE CAUSED BY STUCK OPEN SPRAY VALVE 2-RC-100F 87-12 11/16/87 E* 1 REACTOR TRIP ON STEAM GENERATOR #1 LOW LEVEL FOLLOWING FAILURE OF FEE 0 WATER REGULATING VALVE; OTHER PROBLEMS WERE FAILURE OF "A" AUXILIARY FEE 0 WATER PUMP TO START AND STOPPING OF "A" AND I "C" REACTOR COOLING DUMPS OUE TO BUS TRANSFER

'

FAILURE l

l l T-38-3 l

(

i-

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Table 3B !

.

l LER EVENT CAUSE AREA l NUMBER DATE CODE CODE DESCRIPTION l 87-13-01 12/19/87 A* 1 FIRE WATCH PATROL FAILED TO CONDUCT AN HOURLY INSPECTION OF CABLE VAULT AREA THAT CONTAINS NON-QUALIFIED CABLE TRAY ENCLOSURES 87-14 12/31/87 E* 1 SIX OF 16 MAIN STEAM SAFETY VALVES FAILED ThE SIMMER TEST DUE TO SETPOINT DRIFT ,

!

l

'

-- CAUSE CODES HAVE BEEN ASSIGNED BY OR CHANGES FROM THE LICENSEE CODES BY NRC

REGION I

l l

i t

I a

f I

Y-38-4

[

'

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TABLE 3C SYN 0PSIS OF SECURITY EVENT REPORTS (SERs)

MILLSTONE SITE LER EVENT CAUSE NUMBER DATE CODE DESCRIPTION 86-2c 8/12/86 E* SECURITY RELATED EVENT FOR ALL UNITS - LOSS OF COM-PUTER PCWER 86-21 9/11/86 E" SECURITY RELATE 0 EVENT FOR UNIT 1 - LOSS OF VITAL AREA BARRIER 86-22-02 10/18/86 B SECURITY RELATED EVENT FOR ALL UNITS - LOSS OF VITAL AREA BARRIER 86-23-01 10/23/86 B SECURITY RELATED EVENT FOR UNIT 1 - LOSS OF VITAL AREA BARRIER 86-24 11/14/86 A* SECURITY RELATED EVENT FOR ALL UNITS - PERSONNEL ACCESS PROBLEM 86-26 11/24/86 A SECURITY RELATE 0 EVENT FOR ALL UNITS - LOSS OF VITAL AREA BARRIER 86-30-01 12/11/86 A* SECURITY RELATED EVENT FOR ALL UNITS - LOSS OF PRO-TECTED AREA BARRIER 86-31 12/23/86 E* SECURITY RELATED EVENT FOR ALL UNITS - COMPUTER FAILURE 87-02-01 2/6/87 B SECURITY RELATED EVENT FOR UNITS 1 AND 2 - ACCESS CONTROL PROBLEM 87-03 2/6/87 A* SECURITY RELATED EVENT FOR ALL UNITS - ACCESS CONTROL PROBLEM 87-06 3/9/87 A* SECURITY RELATE 0 EVENT FOR ALL UNITS - PROTECTED AREA ACCESS CONTROL PROBLEM 87-09 4/6/87 E* SECURITY RELATED EVENT FOR ALL UNITS - COMPUTER FAILURE 87-10-01 4/9/87 E* SECURITY RELATED EVENT FOR UNITS 1 AND 2 - COMPUTER FAILURE 87-11 S/21/87 A* SECURITY RELATED EVENT FOR UNIT 3 - ACCESS CONTROL PROBLEM T-3C-1

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Table 3C ,

LER EVENT CAUSE NUMBER DATE CODE DESCRIPTION 87-12 9/3/87 B* SECURITY RELATED EVENT FOR UNIT 1 - BREACH OF VITAL AREAS 87-13 9/5/87 C* SECURITY RELATED EVENT FOR ALL UNITS - POTENTIAL CIVIL DISTURBANCE 87-14 6/7/87 A* SECURITY RELATED EVENT FOR UNIT 1 - BREACH OF VITAL AREA 87-14 9/7/87 E* SECURITY RELATED EVENT FOR ALL UNITS - COMPUTER FAILURE 87-15 10/16/87 A* SECURITY RELATED EVENT FOR ALL UNITS - UNESCORTED ACCESS TO PROTECTED AREA ,

87-16 6/11/87 E* SECURITY RELATED EVENT FOR ALL UNITS - COMPUTER FAILURE 87-16 10/22/87 A* SECURITY RELATED EVENT FOR ALL UNITS - LOST BADGE 87-18-01 6/23/87 8 SECURITY RELATED EVENT FOR UNITS 1 AND 2 - COMPUTER FAILURE 87-18 11/12/87 A* SECURITY RELATED EVENT FOR ALL UNITS - GUARD AL-LEGEDLY NOT ALERT AT POST ;

87-19 11/19/87 A* SECURITY RELATED EVENT FOR ALL UNITS - FAILURE TO MAINTAIN PROTECTED AREA COMPENSATING MEASURES 87-20 11/24/87 A* SECURITY RELATED EVENT FOR A!.L UNITS - LOST BADGE 87-21 12/2/87 D* SECURITY RELATED EVENT FOR ALL UNITS - ALLEGED ENTRY OF DANGEROUS WEAPON 87-22 12/22/87 A* SECURITY RELATED EVENT FOR UNIT 3 - UNINTENTIONAL UNAUTHORIZED ENTRY INTO VITAL AREA 87-25 7/24/87 C* SECURITY RELATED EVENT FOR ALL UNITS - BOMB THREAT 87-27 8/14/87 D* SECURITY RELATED EVENT FOR UNIT 1 - BREACH OF VITAL AREA

-- CAUSE CODES HAVE BEEN ASSIGNED BY OR CHANGES FROM THE LICENSEE CODES BY NRC

REGION I

T-3C-2

- - ______ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ . _ _ - _ _ _ __ . _ _ _ _ _ _ _ _ _ - _

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TABLE 4

%MN ARY 9MORCEO OUTAGES. UNPLA INED TRIPS, AND POWER REDUCTIONS MILLSIONE 1 AREA A @ C Q j X TOTAL PLANT OPERATION 1 1 RADIOLOGICAL CON S 0 MAINTENANCE 1 1 SURVEILLANCE 1 1 EMERGENCY PREP O SEC/ SAFEGUARDS OUTAGE MANAGEMENT O 0 O

TRAINING INADEQUACY $ 1 1 ASSURANCE OF QUALITY 0 ENGINEERING SUPPORT b 3 3 Tk 3 4 7 SUMMARY OF FORCED OUTAGES, UN L TRIPS, AND POWER REDUCTIONS MILL AREA A B C Q j X TOTAL PLANT OPERATIONS 1 1 RADIOLOGICAL CONTROLS 0 MAINTENANCE 2 2 4 SURVEILLANCE EMERGENCY PREP

9 1

SEC/ SAFEGUARDS 0 OUTAGE MANAGEMENT 0 TRAINING INADEQUACY 0 ASSURANCE OF QUALITY 0 ENGINEERING SUPPORT 2 2

,

TOTALS: 4 2 8 CAUSE CODES A -- PERSONNEL ERROR B -- DESIGN, MANUFACTURING, CONSTRUCTION / INSTALLATION C -- EXTERNAL CAUSE

'

D -- DEFECTIVE PROCEDURE E -- EQUIPMENT FAILURE X -- OTHER T-4-1 ,

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TABLE 4 SUMMARY OF FORCEO OUTAGES, UNPLANNED TRIPS, AND POWER REDUCTIONS MILLSTONE 1 AREA A @ C Q E X TOTAL PLANT OPERATIONS 1 1 RADIOLOGICAL CONTROLS 0 MAINTENAllCE 1 1 SURVEILLANCE 1 1 EMERGENCY PREP O SEC/ SAFEGUARDS 0 OUTAGE MANAGEMENT 0 TRAINING INADEQUACY 1 1 ASSURANCE OF QUALITY 0 ENGINEERING SUPPORT 0 0 TOTALS: 3 1 4 SUMMARY OF FORCEO CUTAGES, UNPLANNED TRIPS, AND POWER REDUCTIONS MILLSTONE 2 AREA A @ C D E X TOTA PLANT OPERATIONS 1 1 RADIOLOGICAL CONTROLS 0 MAINTENANCE 2 2 4 SURVE!LLANCE 1 1 EMERGENCY PREP 0 SEC/ SAFEGUARDS 0 OUTAGE MANAGEMENT 0 TRAINING INADEQUACY 0 ASSURANCE OF QUALITY 0 ENGINEERIhG SUPPORT 1 1 TOTALS: 4 1 2 7 CAUSE CODES A -- PERSONNEL ERROR B -- DESIGN, MANUFACTURING, CONSTRUCTION / INSTALLATION C -- EXTERNAL CAUSE D -- DEFECTIVE PROCELURE E -- EQUIFMENT FAILURE X -- OTHER T-4-la

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TABLE 4A SYN 0PSIS OF FORCEO OUTAGES, UNPLANNEO TRIPS, AND POWER REDUCTIONS ;

MILLSTONE 1 POWER LER CAUSE DATE LEVEL DURATION DESCRIPTION NUMBE & AREA *

6/19/86 100% --

POWER REDUCTION TO REPAIR STEAM --

REPAIR LEAKS LEAK IN "B" SHUTDOWN COOLING HEAT (N0 AREA EXCHANGER ASSIGNED)

6/28/86 100% --

POWER REOUCTION TO REPAIR COND - --

REPA!R LEAKS SER TUBE LEAKS (ENGINEERING l I

SUPPORT)

7/16/86 100% --

POWER REDUCTION FOR C0t 00 --

ADJUSTMENT &

PATTERN ADJUSTMENT AN PAIR REPAIR LEAKS CONDENSER TUBE LEAK (ENGINEERING

/( l SUPPORT)

10/9/86 100% --

POWER REOUCTION AIR CON- --

REPAIR LEAKS DENSER TUBE L (ENGINEERING i SUPPORT) I 11/30/86 100% 15 OAYS REACTOR T GENERATOR TRIP 86-27 EQUIPMENT CAUSED A(* RATOR LOCK-0VT 00E TO FAILURE (NO PHASE- S' JND FAULT OF THE MAIN AREA ASSIGNED)

TRAN

'

3/22/87 50% 27 HR$ R ME TRIP AND ISOLATION ON LOW 87-07 TRAINING EAM LINE PRESSURE DUE TO INADEQUACY dY URE OSCILLATIONS CAUSED BY j

j

, h' ROL PROBLEMS WITH THE EPR/MPR

4/15/87 100% -

OWER REOUCTION TO REPAIR STEAM --

REPA!R LEAKS I

LEAKS IN HEATER BAY (NO AREA '

ASSIGNED) ,

8/14/87 0% --

REACTOR TRIP DURING STARTUP ON 87-34 OPERATOR INTERMEDIATE RANGE HIGH FLUX DURING ERROR WITH0RAWAL OF HIGH WORTH CONTROL (OPERATIONS) ;

R00 26-31 8/26/87 00% 21 HRS REACTOR TRIP OURING AVERAGE POWER 87-36 TESTING ERROR '

,

RANGE MONITOR SURVEILLANCE TESTING (SURVEILLANCE)

T-4A-1

_-_ _ _ _ _

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TABLE 4A SYNOPSIS OF FORCED OUTAGES, UNPLANNED TRIPS, AND POWER REDUCTIONS MILLSTONE 1 POWER LER CAUSE DATE LEVEL DURATION DESCRIPTION NUMBER & AREA *

6/19/86 100% --

POWER REDUCTION TO REPAIR STEAM --

REPAIR LEAKS LEAK IN "B" SHUTDOWN COOLING HEAT (N0 AREA EXCHANGER ASSIGNED)

6/28/86 100% --

POWER REDUCTION TO REPAIR CONDEN- --

REPAIR LEAKS SER TUBE LEAKS (N0 AREA ASSIGNED)

7/16/86 100% --

POWER REOUCTION FOR CONTROL R00 --

ADJUSTMENT &

PATTERN ADJUSTMENT AND TO REPAIR REPAIR LEAKS CONDENSER TUBE LEAKS (N0 AREA ASSIGNED)

10/9/86 100% --

COWER REDUCTION TO REPAIR CON- --

REPAIR LEAKS DENSER TUSE LEAKS (N0 AREA ASSIGNED)

11/30/56 100% 15 DAYS REACTOR TRIP ON GENERATOR TRIP 86-27 EQUIPMENT CAUSED BY GENERATOR LOCK-0UT DUE TO FAILURE (N0 PHASE-TO-GROUND FAULT OF THE MAIN AREA ASSIGNED)

TRANSFORMER 3/22/87 50% 27 HRS REACTOR TRIP AND ISOLATION ON LOW 87-07 TRAINING

,

MAIN STEAM LINE PRESSURE DUE TO INADEQUACY l PRESSURE OSCILLATIONS CAUSED BY

'

CONTROL PROBLEMS WITH THE EPR/MPR I 4/15/87 1001 --

POWER REDUCTION TO REPAIR STEAM --

REPAIR LEAKS LEAKS IN HEATER BAY (NO AREA ASSIGNED)

8/14/87 0% --

REACTOR TRIP OURING STARTUP ON 87-34 OPERATOR INTERMEDIATE RANGE HIGH FLUX DURING ERROR WITHD7AWAL OF HIGH WCRTH CONTROL (OPERATIONS)

RCD 26-31 8/26/87 100% 21 HRS REACTOR TRIP OURING AVERAGE POWER 87-36 TESTING ERROR I

'

RAN3E MONITOR SURVEILLANCE TESTING (SURVEILLANCE)

T-4A-la

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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Table 4A POWER LER CAUSE DATE LEVEL DURATION DESCRIPTION NUMBER & AREA *

9/3/87 100% 44 HRS REACTOR TRIP ON LOW SCRAM HEADER 87-38 EQUIPMENT PRESSURE CAUSED BY LOW SERVICE AIR FAILURE HEADER PRESSURE DUE TO SERVICE AIR (MAINTENANCE)

COMPRESSOR FAILURE DURING HIGH SERVICE AIR USAGE 11/14/87 100% 64 HRS REACTOR SHUTDOWN TO INVESTIGATE AND --

REPAIR LEAK REPAIR IC-1 PACKING INSIDE ORYWELL (N0 AREA ASSIGNED)

-- CAUSE AND AREA CODES HAVE BEEN ASSIGNED BY NRC REGION I

\

T-4A-2

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TABLE 4B

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.t PSI O FORCEO OUTAGES, UNPLANNED TRIPS, AND POWER REDUCTIONS MILLSTONE 2 POWER D LER CAUSE DATE LEVEL _' I R ' N DESCRIPTION NUMBER & AREA *

6/1/86 60% 13 t REA TOR TRIP DN REACTOR COOLANT 86-04-01 PERSONNEL s.P Uiu'RSPEED CAUSED BY LOSS OF ERROR BY THE

'0WER 1S 25B DUE TO IMPROPER OPERATIONS ATION OF BREAKER CONTROL STAFF SW TC([gp2-25B-2 8/12/86 984 112 HRS REACTL' ON #1 STEAM GENERATOR 86-05 PERSONNEL GENERAT L EVEL AFTER LOSS OF ERROR BY THE "A" .E ER PUMP OUE TO LOSS ENGINEERING OF OIL PUMr WIN BUSES 22A AND 228 SUPPORT (CROSS-TIEO) 'LTT POWER 9/3/86 100% 26 HRS REACTOR TRIP ON 'N' ST AM FENERA- 86-06 DESIGN DE-TOR LEVEL DUE TO L %. HEATER FICIENCY BY ORAINS FLOW FOLLOWI ' F URE OF ENGINEERING AIR FITTING TO THE hee' AINS SUPPORT CONTROL VALVE CLOSING L '_

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12/23/86 50% 20 HRS REACTOR TRIP ON LOW STEAM NERA- c'-22 PERSONNEL TOR LEVEL OUE TO FEE 0 WATER Pt ? ERROR BY SPEE0 DECREASE TO MINIMUM UPON 0 MAINTENANCE OF POWER ON BUS 24C, CAUSED BY .-

PROPERLY INSTALLE0 DRAWER 1/2/87 100% 21 HRS REACTOR TRIP ON LOW STEAM GENERA- 6-1 PERSONNEL TOR LEVEL FOLLOWING LEVEL CONTROL ERROR BY AN PROBLEMS OUE TO A HOT JUMPER ARC ON ELEC'iRICIAN THE FIRE SUPPRESSION ALARM PANEL (MAINTENANCE)

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l 1/29/87 100*. 18 OAYS CONTROLLEO SHUTOOWN FOLLOWING IN- --

. EAM

DICATIONS OF A STEAM GENERATOR TUBE GE cRATOR

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LEAK IN THE "A" GENERATOR TUBE '.EAK (SURVt LLANCE)

3/24/87 100% 0 HRS REACTOR POWER LEVEL WAS REDUCED TO --

STEAM LE *. TO REPAIR A STEAM LEAK ON THE REPAIR (N0

"B" FEE 0 WATER PUMP AREA ASSIGNE i

i T-48-1 l

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TABLE 48

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SYNOPSIS OF FORCEO OUTAGES, UNPLANNED TRIPS, AND POWER REDUCTIONS MILLSTONE 2 POWER LER CAUSE DATE LEVEL DURATION DESCRIPTION NUMBER & AREA *

6/1/86 60*4 13 HRS REACTOR TRIP ON REACTOR COOLANT 86-04-01 PERSONNEL PUMP UNDERSPEED CAUSED BY LOSS OF ERROR BY THE POWER TO BUS 25B OVE TO IMPROPER OPERATIONS OPERATION OF BREAKER CONTROL STAFF SWITCH 252-25B-2 8/12/86 93% 112 HRS REACTOR TRIP ON #1 STEAM GENERATOR 86-05 EQUIPMENT GENERATOR LOW LEVEL AFTER LOSS OF FAILURE (NO THE "A" FEE 0 WATER PUMP OUE TO LOSS AREA ASSIGNED)

0F CIL PUMPS WHEN BUSES 22A AND 228 (CROSS-TIEO) LOST POW /3/86 1004 26 HRS REACTOR TRIP ON LOW STEAM GENERA- 86-06 DESIGN DE-TOR LEVEL OVE TO LOS% OF HEATER FICIENCY BY ORAINS FLOW FOLLOWING FAILURE OF ENGINEERING AIR FITTING TO THE HEATER ORAINS SUPPORT CONTROL VALVE CLOSING VALVE 12/23/86 50% 20 HRS REACTOR TRIP ON LOW STEAM GENERA- 86-22 PERSONNEL TOR LEVEL DUE TO FEE 0 WATER PUMP . ERROR BY SPEED DECREASE TO MINIMUM UPON LOSS MAINTENANCE OF POWER ON BUS 24C, CAUSED BY IM-PROPERLY INSTALLEO ORAWER 1/2/87 100% 21 HRS REACTOR TRIP ON LOW STEAM GENERA- 87-02 PERSONNEL TOR LEVEL FOLLOWING LEVEL CONTROL ERROR BY AN PROBLEMS DUE TO A HOT JUMPER ARC ON ELECTRICIAN THE FIRE SUPPRESSION ALARM PANEL (MAINTENANCE)

1/29/87 100% 18 DAYS CONTROLLEO SHUTOOWN FOLLOWING IN- --

STEAM 0! CATIONS OF A STEAM GENERATOR TUBE GENERATOR LEAK IN THE "A" GENERATOR TUBE LEAK (SURVEILLANCE)

3/24/87 100% 0 HR$ REACTOR POWER LEVEL WAS RLOUCEO TO --

STEAM LEAK 80% TO REPAIR A STEAM LEAK ON THE REPAIR (NO

"B" FEE 0 WATER PUMP AREA AS$1GNED)

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i i Table 4B I

J POWER LER CAUSE

DATE LEVEL DURATION DESCRIPTION NUMBER & AREA *

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4/16/87 100*4 20 HRS REACTOR TRIP ON TURBINE TRIP 87-07 EQUIPMENT CAUSED BY GENERATOR EXCITER FIELD FAILURE (N0 BREAKER AND GENERATOR BREAKERS AREA ASSIGNED)

OPENING, CAUSE UNKNOWN 7/23/87 100*. 21 HRS REACTOR TRIP ON STEAM GENERATOR --

RANDCM EQUIP-

! LO'/ LEVEL DURihd DOWN-POWER IN RE- MENT FAILURE l SP0i:SE TO DECREASING PRIMARY PRES- (N0 AREAS SURE CAUSED BY A PARTIALLY (1/3 ASSIGNED)

l OPEN) STUCK OPEN SPRAY VALVE i

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9/2/87 91*; 34 HRS REACTOR TRIP ON #1 STEAM GENERATOR 87-09 EQUIPMENT LOW LEVEL DUE TO FAILURE OF FEED- FAILURE (NO WATER CONTROL VALVE #2-P4-51A, THE AREA ASSIGNED)

PLUG HAD SEPARATED FRCM THE STEM 11/16/87 100*. 26 HRS REAR"iR TRIP ON STEAM GF.NERATOR #1 87-12

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EQUIPMENT LOW LEVEL DUE TO LEVEL TRANSIENT FAILURE (N0 CAUSED BY MALFUNCTION OF THE VALVE AREA ASSIGNED)

POSITIONER FOR FEEC'JATER REGULATING l VALVE #2-Pd-SIA

-- CAUSF AND AREA CODES HAVE BEEN ASSIGNED BY NRC REGION I

T-48-2

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ENCLOSURE 2 SALP COMMENT EVALUATION AND SALP ERRATA 1. NRC Comments on Substantive Licensee Comments I.1 Millstone 2 Plant Operations Regarding Unit 2 plant operations, you stated that your performance on the DC Vital Switchgear Room coolers was more responsible than was sum-marized in the SALP. We recognize that you did engineering and 10 CFR 50.59 evaluations, provided procedural guidance for alternate cooling and demonstrated its adequacy, and planned modifications. The SALP noted (Page 4, 5th complete paragraph) that there was little safety signific-ance in this case but that operational and management review of plant conditions should have prompted earlier resolution of cooler inoper-abilit That evaluation is still considered valid; these safety-related coolers were essentially inoperable for four years. Our review concluded that the SALP paragraph involved should not be change On fire protection, you commented that fire damper concern was a docu-mentation issue only and that there was no concern for the conditions involved with the noted lack of a fireproof coating. We concur and modified the report. You also commented that, instead of some operators having difficulty with tasks involving some safe shutdown equipment a-1 breakers, one operatcr had difficulty due to use of the wrong tool but did perform the evolutio The sentence involved was delete This was a prime factor in our raising the Unit 2 Plant Operations ratirg to Category You further commented that one individual has been made responsible for overseeing the fire protection staff; we have deleted the sentence that stated otherwis On Unit 2 housekeeping, you identified improvements during the last 12 months and noted that housekeeping is now very good and improving. We have not changed our assessment or writeup on Unit 2 housekeeping and will carefully assess your improvements during the next SALP, In summary, our review of your comments on Millstone 2 plant operations resulted in raising the rating assigned to Category .2 Millstone 2 Surveillance For Unit 2 Surveillance, you disagreed with the weight given to steam generator eddy current testing problems. We agree with almost all of your comeents, including the one about steam generator leakage being from a tube that met acceptance criteri It was the 37 indicated defects which caused us concern. We credit your conservative reevaluation, though we also believe that plugging all 37 of these tubes was appro-priate. We also agree that these were a small percentage of the tubes examined and that steam generator eddy current ;esting was a small per-

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Enclosure 2 2 centage of the surveillance program. But, for these 37 tubes and three others, your program did not result in plugging tubes that should have been plugged based on available information. Assigning enough weight ,

to this matter to make the difference between a SALP Category 1 and a [

SALP Category 2 does represent a high standard. A high standard is par-ticularly appropriate for primary coolant system integrity, as is the case here. Our review therefore affirmed this Category 2 SALP evaluatio I.3 Millstone 1 Engineering Support j You commented that extensive engineering support had been applied to main condenser problems, and that main condenscr replacement had been evalu- i ated as not being an economic alternative. We recognize the extensive !

efforts applie Inasmuch as the power reductions involved have been accomplished without causing plant trips or safety system challenges, the condition has involved balancing power reduction economics against I replacement cost economics. Although the inleakage persists, the SALP did not show a safety correlation and has been changed accordingl ,

We acknowledge the additional information you provided on the LPCI/CS cnd IE Bulletin 79-14 effort. The concern remains that, in 1985, it was I found that the pump anchorings did not agree with thi design drawing Actual status of the pump anchorings was not confirmed until destructive ,

examinations were conducted in 1987.. We acknowledge your comment about i interim evaluation of ac:eptability of the installation. However, in response to our request (orally relayed by the resident inspector) for !

that evaluation, no documented review or analysis was provided. There- L fore, while we have modified the SALP to more accurately depict the r situation, our concerr. about the delay in resolution remain '

I SAlp Board Report Errata PAGE/ ;

LINE PREVIOUS WORDING REVISED WORDING !

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4/35 "On the other hand, weaknesses in en- "On the other hand, weaknesser ;

vironmental qualification, slow re- in environmental qualification l sponse to identification of short pump and slow response to identifica- I foundation bolts, and recurring main tion of short pump foundation main condenser tube leaks showed that bolts showed that significant that significant engineering support engineering support improvements improvements can be made, can be made." t Basis: Removed "recurring main condenser tube leaks" as an example because no link to safe operation was establishe l

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Enclosure 2 3 PAGE/

LINE PREVIOUS WORDING REVISED WORDING I 4/46 "The prior SALP rated five areas as "The prior SALP rated six areas

---

Catego ry 1, . . . " a s Category 1, . . . "

Basis: To reflect the prior, amended SALP Report, which modified the "Outage Man-agement" rating from "None" to Category "1."

6/47 "None#" "1" Basis: To correct the previous SALP tabulation for "Outage Management" to the ;

Category 1 rating assigned in the final SALP Repor /14 "A high level of safety performance "A high level of safety perform-was noted in Maintenance...." ance was noted in Plant Opera-tions, Maintenance...."

j Basis: To reflect amended assessment of Plant Operations.

7/51 "This SALP rated four areas as "This SALP rated five areas as i

Category 1 and seven as Category Category 1 and six as Category Basis: To reflect amended assessmen .

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l 9/36 Plant Operations (This Period)

"2" "1"

Basis: To reflect amended assessment.

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Enclosure 2 4 PAG 5/

LINE PREVIOUS WORDING REVISE 0 WORDING 15/26 "

... missing fire damper and the "missing fire damper Q documen-

"

sec . . . . tation problem only) and the seCond..."

15/28 "

...between the auxiliary feedwater "

...between the auxiliary feed-heaters and their isolation valves, water pumps and their isolation Also, fire coating material was found valves. ATso, fire coating mate-unacceptable (LER 87-10); additional rial was found unacceptable (LER compensatory measures were take The 87-10) with no significant hazard licensee has an adequate fire protec- involved; appropriate compensa-tion staff but no one person has been tory measures were taken. The been made responsible for overseeing licensee has an adequate fire fire protection..." protection staf f.",

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Basis: Correct typo and modify report to reflect accepted comments about the fire damper problem being a documentation one, about no hazard being associated with the unacceptable fire coating, and about one person being responsible for overseeing the fire protection staf /33 "

...With regard to training in Appen- (Deleted in Entirety)

dix R modifications, some operators had difficulty in performing tasks such as locating some safe shutdown equipment and removing some breakers..."

Basis: Accepted comment that only one operator had difficulty (due to wrong tool)

and that evolution was nonetheless performed acceptably 15/44 "Overall, housekeeping was evaluated "Overall, housekeeping was evalu-as f air." usted as ,sa ti sf actory."

Basis: More direct correlation to SALP category definitio /17 Category Category Basis: Amended Millstone 2 Plant Operations rating based on incorporation of accepted licensee comment '

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Enclosure 2 5 PAGE/

LINE PREVIOUS WORDING REVISED WORDING 16/25 "-- Assure proficier,:y in shutdown (Deleted in Entirety)

equipment operction."

Basis: This recommendation is not appropriate to the amended SAL /23 "SG local leak rate..." "Containment local leak rate..."

Basis: Error correcte /15 "In 1984, NNECO identified the poten- "In December 1984, NNECO identi-tial for short foundation bolts for fied the potential for short LPCI and CS pump NUCSO engineering foundation bolts for LPCI and was slow to respond to associated site CS pumps. However, NUSCO did initiatives and slow to recognize that not verify the actual bolt con-the problem existed. The presence of figuration until July 1987 when the short bolts was not confirmed and destructive examination was con-corrected until 1987." ducted, nor was documented as-sessment of acceptability pro-vided. However, NUSCO expanded the IE Bulletin 79-14 effort to include the LPCI and CS anchor-ings, and additional anchors were installed in 198?.

Basis: To more accurately depict the LPCI/CS bolt issu :

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Enclosure 2 6

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PAGE/ !

LINE PREVIOUS WORDING REVISED WORDING 46/21 "The recurrence of main condenser tube "A contributor to the August 1987 i leaks requiring frequent power reduc- reactor scram was the failure tions to identify and repair needs de- to incorporate new core design sign resolution (see Table 4A). A precautions into operating pro-contributing cause for the August 1987 cedures. This is an example of ;

reactor scram was the failure to in- the need for engineering support I corporate appropriate new core design initiatives to assure timely in- !

precautions into the operating proce- clusion of design inputs in ;

dure These examples show the need operating controls." '

for better engineering support initi-atives to resolve longstanding, recur- .

rent problems, and to assure timely com- !

pletion of design inputs into operating

) controls." [

l Basis: The assessment was changed to delete reference to the main condenser in- l leakage problem because nuclear safety significance was not show .

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! 49/53 "Two reactor trips during the assess- "One reactor trip during the t

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l ment period were caused in part by de- assessment period was caused in '

! sign deficiencies. One involved..." part by a design deficie9e '

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It involved..." -

49/56 "The second involved the improper "A second trip involved a mal- i

, overcurrent trip setpoint on plant functioning oil-filled, overcur- '

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electrical buses power the pressurizer rent trip device for the ores-heaters. Follow-up actions to iden- surizer heaters. To prevent re- i a tify and correct these deficiencies currence, all oil-filled 480 volt were proper." load center trip devices were

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replaced with solid state devices, t

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which should maintain their set- ,

points bette I i

Basts: Acceptance of licensee commen f

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Enclosure 2 7 i

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PAGE/

LINE PREVIOUS WORDING REVISED WORDING .

l T-4-1/ Unit 1 ENGINEERING SUPPtAT

{

27 E = "_3" TOTAL = "3"

_

E = "_0" TOTAL = "_0"

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29 E TOTALS = "4" TOTAL = "7" E TOTALS = "1" TOTAL = "4" l Basis: To reflect deletion of main condenser events because safety significance l was not establishe i

T-4-1/ Unit 2 ENGINEERING SUPPORT ;

47 8 = "2" TOTAL = "2" B = "1" TOTAL = "1" 49 B TOTALS = "2" TOTAL = "8" B TOTALS = "1" TOTAL = "?" I Basis: The unplanned trip due to the trip device malfunction, which was attributed [

to engineering support, has been deleted due to misclassification, f I

i T-4A-1/ f 23-33 Events datM 6/28/86, 7/15/86, and 10/9/86 i These events were attributed to These events are categorized as

"(ENGINEERING SUPPORT)." "(NO AREA ASSIGNED)." i Basis: Reevaluation of condenser events due to safety significance not being show f l

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T-48-1/ l 13 "PERSONNEL ERROR BY ENGI.EERING "EQUIPMENT FAILURE; NO AREA l

$UPPORT" ASSIGNED l

Basis: The reactor trip on 8/12/86 was caused by a malfunctioning trip device, and i not by an engineering erro I i I

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. - i' Ef4 CLOSURE 3

.r NORTHEAST UTILITIES O.,,,o On.c. . s.e.n so..i. semn. Con uncui l i UreN.ANi7[cI[c P O BOX 270

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k k J C"$CE,7, HARTFOAD. CONNECTICUT 06141-0270 (203) 665-S000 May 27, 1988 Rocket NoL _50-2.45 lh MlL31 Re: SALP U.S. Nuclear Regulatory Comission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen:

Millstone Nuclear Power Station, Unit Nos. I and 2 SX11tatic.A11edimaat_of licensAe Performar.co The Staff recently forwarded the combined SALP Board Report (I) for the 19-month period ending December 31, 1987, for Hillstone Unit Nos. 1 and Subsequent to receipt of the SALP Board Report, a meeting was held on April 28, 1988 between members of the Staff and members of Northeast Nuclear Energy Company (NNECO).

We believe that our meeting on April 28, 1988 was helpful and productive. The information exchanged during the meeting clarified several important issues and served its purpose in a very constructive fashion in our view. Consistent with our discussions during the meeting, we are responding to the findings of the SALP Board with particular emphasis on the Board recommendations for the individual evaluation categorie In two instances, namely the Plant Operations and Surveillance functional areas for Hillstone Unit No. 2, we believe that the information we discussed during the meeting and are providing herein warrants the reconsideration of the performance rating as currently assesse The responses to the Board's recommendations and other pertinent coments for both Hillstone U,11t No. I and Hillstone Unit No. 2 are contained in Attachment A to this lette The transmittal letter of the SALP Board Report indicated that our written coments should be sutaitted within 20 days af ter our trectin In a telephone conversation with Mr. E. C. McCabe on May 2, 1988, NNECO was granted a 10. day extension for the submittal of this lette . _

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(1) W. T. Russell letter to E. J. Mroczka, "Combined Systematic Assessment of Licensee Performance (SALP) Report Nos. 50 245/86 99: 50 336/86 99 (6/1/86 12/31/81)," dated March 29, 198 '

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r Regulatory Comission

- 'May 27, 1988

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NNECO takes very seriously the ratings and recommendations given by the Board as an input to the continuing process of evaluating and improving our overall performanc As reflected by our comments and observations during the April 28, 1988 meeting, wc generally concur with the Board's observations and have taken or are taking steps to address the concerns identified. It remains our objective to achieve Category I ratings in all functional areas for subsequent SALP evaluations, and the attachment to this letter describes some of the steps we will be taking in pursuit of that goa Two errors were noted in the "Summary of Results" section of the SALP Board Report concerning Millstone Unit No. l's performance in the SALP evaluation period ending May 31, 1986. On page 6, the performance rating for the Outage

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Management functional area for the last evaluation seriod is listed as "None,"

with a corresponding footnote stating "Not assessec as a separate area in the

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lest SALP." This was the SALP Board Report issued,gginal performance but was subseque rating assigned in the initial rating in the final SALP Board Resort issued.gy revised As such,tothea performance Category I rating for Outage Management shoulc be listed a. .. '"

In addition, the last paragraph on page 4 should be revised to sti'a Laat 'six" functional areas i received Category I rating in the prior SALP, 9. "five."

We trust that the actions presented '9 the attachment addressing the concerns of the Board and our general comments will be considered in the final assess-ment of our performance and in subsequent SALP evaluations. We will be updating you regarding the status of implementing the corrective actions discussed herein prior to the next SALP evaluation. Please feel free to contact us if any questions arise on these matterc, or if additional clarift-cation is neede Very truly yours, NORTHEAST l.UCLEAR ENERGY COMPANY

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.drocz.ka 1 V- W Senio#r Vice President cc: W. T. Russell, Region 1 Administrator M. L. Boyle, NRC Project Manager, Millstone Unit No. 1 D. H. Jaffe, NRC Project Manager, Millstone Unit No. 2 W. J. Raymond, Senior Resident Inspector, Millstone Unit Nos.1, 2, and 3 (2) 1. E. Hurley letter to J. F. Opeka, "SALP Report Nos. 50 245/85 o3 and 50-336/85 98," dated August 29, 198 (3) T. E. Murley letter to F. Opeka, "SALP Reports 50 245/85 98 and 50 336-85 98, and Your Reply Letter, dated November 5, 1986," dated December 4, 198 ., ,_ _ _ _____ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _

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Docket Nos. 50 245 :

50-336 A07137

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I Attachment A

Northeast Nuclear Energy Company 3 *

Millstone Unit Nos. I and 2 Response to SALP Report

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May 1988

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Functional Area: PLANT OPERATIONS Millstone Unit No. 1 Rgird Recomendation: None, hsponse:

NNECO agrees with the Staff's assessment of our performance in this functional area. We acknowledge the need to address the issues identified as having room for improvement, and will continue to strive for better performanc Millstone Unit No. 2 Rgard Recomendations:

--

Improve equipment operability overvie i

--

Assure proficiency in shutdown equipment operatio Improve housekeepin Response:

Equiement Ooerability The SALP report discusses the extended inoperability of the ventilation for the vital DC switchgear rooms. In a letter dated December 4, ,

coolgNNECO 1987 described to the NRC Staff the backgrouni details and cor actions already associated with this issue when it bec3me a topic of discus-sion with the Staf NNECO believes that this isolated situation is not represectative of our overall attention to equipment operability during the past SALP perio We believe the Staff's assessment of our overall performance in this area inappropriately places too much emphasis on this one situation. Additionally, our performance in this particular case was more responsible than sumarized by the Staff in the SALP report. Specifically,

,

o when the chilled water system was first recognized to have chronic problems, an engineering evaluation was directed to determine the t impact on the plant's design basis, o a 10CFR50.59 evaluatien was conducted to justify having the chilled !

water coolers for the vital DC switchgear room out of service for an '

extended perio !

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l (1) E. J. Mroczka letter to U.S. Nuclear Regulatory Comission, "Additional !

Information - Chilled Water System," dated December 4, 198 '

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Attachment A A07137/Page 2 o the viability of the alternate measures for cooling the vital DC switchgear room had been demonstrated, o

procedural guidance was in place to implement the alternate cooling action o steps were taken to qualify off-the shelf heat exchangers to ensure that future needs for exchangers could be me o proposed modifications were identified and implemented to facilitate maintenance on the heat exchanger o heat exchangers with a larger capacity had been ordered to allow for tube plugging if necessary and lower room temperatur NNECO's corrective actions were nearing completion when the Staff's concern was raised in the fall of 198 While restored operability of the entire system would have ideally occurred earlier in time, the safety significance of this particular situation was such that the overall result was acceptabl Again, this situation is an isolated case, and does not reflect a programmatic dnficienc Fire protectign The SALP report discusses two Appendix R violations - one for a missing fire damper and the second for insufficient separation between the auxiliary feedwater pump isolation valves. (The SALP report states that there was insufficient separation "between the auxiliary feedwater heaters' and their isolation valves. The report should be revised to say "between the auxiliary feedwater pumps.") Also, fire coating material was found unacceptable (LER

,

87-10).

The specific fire damper concern identified in the SALP report was in a barrier separating the Auxiliary Building (14' 6') General Area (stairway) and i

the West Electric Penetration Area. This condition (no fire damper) had been l reviewed in 1986 as part of a Fire Damper Upgrade project and found to be acceptable with no damper. However, no formal documentation, such as an engineering evaluation, was ever generated to document the positio Therefore, as a result of the NRC's questions. NNECO was unable to produce the necessary paperwork for the NRC's review / approval to support this positio It should be noted that NNECO's verbal position / technical justification for not having a fire damper installed was presented to the NRC Staff during the audit. As a result of this discussion, the NRC concurred that this was not a safety issue but was only a "paper' issue (no engineering evaluation on file forreview).

In response to the NRC's concern expressed during the July, 1987 audit, we properly documented the evaluation and presented it to the NRC prior to the conclusion of the audit. The NRC reviewed and approved the Generic Letter 86 10 engineering evaluation that was prepared and closed out this issue prior to the audit's exit meeting. While the absence of the engineering justifica-tion was an oversight, we do not view it to be significan __

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Attachment A

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A?7137/Fsgo 3 l The second highlight.d issue is the LER (87-10) that was issued just prior to the start of the audit and deals with the lack of fire proof coating on tray support As part of the review cycle for a project, it was determined that the coating of structural steel support members, as part of the cable fire wrapping project, was necessary to etisf approved, tested configuration as recomended b deterrained to be a concern, approyriate y the Once ve actions do this issue were taken the required, was reviewed to address and the ratte Ho.ever, thir issue was 4dentified in the audit inspection report even though no concerns were identified with the physical conditions or action take The SALP report stated that with regard to training in Appendix R modifica-tions, "some operators had difficulty in performing tasks such as locating

some safe shutcown equipment and removing some breakers." NNECO conducted a review of the NRC stated operator weaknesses in the areas of location and operaticn of safe shutdown equipment resulting from the 10CFR50, Appendix R inspection. NNECO takes exception to the findings of the SALP Board for the following reasons. The use of the term "some operators" when describing deficiencies is inappropriate, since our review shows that only one operator had difficulty performing one of the requested evolutions. The reference to locating safe shutdown equipment is questioned, since wa can identify no instance where this occurred.

I The usa of the phrase "removing some breakers" to describe the specifics of j the one deficiency appears to be inaccurate, since the operator correctly

re eved end returned the breaker to service (racked out, racked in). The

error was his inability to manually operate the breaker due to obtaining the r

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incorrect breaker operating too It is important to note that both the Operator Training and Operations

' Departments had, and continue to have, complete confidence in the evaluated operatar's ability to locally operate breaker Training had been prJvided on local breaker operation using the breaker lab as part of the operators' Appendix R training. No skills or knowledge deficien-

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cies were apparent for th< s operator at the time of the Appendix R inspection, j nor are there any curr jntly. Problems encountered during the inspection can

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mo:t likely be attributed to the stress of the "exam like" atmosphere, r

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The SALP report mistakenly states that no one person has been made responsible

! for overseeing fire protection. Millstone Unit No. 2 utilizes one of the

, Assistant Engineering Supervisors as the unit fire protection coordinator, identical to the arrangment in place at Millstone Unit No. HOMehefPh2 The SALP report discussed the need for improvement in housekeeping, and

r e fer ei.ced the failure to remove boron encrustation after leak repairs.

l ANECO's continuing review of this area indicates that housekeeping has continued to improve during the last twelve months, and is now very good and continues to improve. Considerable resources have been spent to decontaminate the Safeguards Rooms and other areas and also to paint virtually the entire Turbine Buildin Attention to active boric acid leaks and boric acid crystals has been height-ened. As part of our contamination source identification program, boric acid

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. Attachment A

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A07137/Page 4 encrusted pt >ing and valves are being identified and cleaned in a timely fashion. Adcitional information on boric acid corrosion concorpA at Millstone Unit No. 2 is containeti in NNECO's May 27, 1988 responseW to Generic Letter 88 05, which explains our preventive maintenance progra In addition, several programs to address recognized weaknesses independent of the NRC review have been initiated, o A new valve identification program was initiated in August of 1987 to improve valve tag legibility and tag durability. The program has had considerable emphasis from a human factors standpoint, o A daily status report has been in effect for almost a year. This report details various plant and equipment status, all department activities for the day and the next several days. The report provides an excellent means to access each department's activities on plant conditions, o Finally, a complete procedure rewrite program to resolve several generic procedure weaknesses has been initiate This rewrite will incorporate many of the human factor criteria identified as critical to minimize personnel error resulting from procedural flaw These actions, in addition to other more minor initiatives, will continue to previde a better overall operating and management environment. We believe our aousekeeping standards are higher now than they were during the early portion of the past SALP period and that these significant improvements were not properly reflected in the Staff's assessment of our performanc Summary In light of the above comments, we believe that reconsideration of the currently assigned Category 2 rating is appropriat .

(2) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, ' Generic Letter 88 05, Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants,' dated May 27, 198 ___ ____ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ ___ ___

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Functional Area: RADIOLOGICAL CONTROLS Millstone Unit Nos. 1 and 2 -

Rg3t{,Resommendations:

--

Improve control and supervision during outage Improve pre job planning and work efficienc Continue improving the ALARA progra Resoonse: l l

Control and Suoervision '

While not yet implemented, we initiated the following changes before the end !

of the last SALP period in recognition of the need to improve control and supervision of health physics support activities. The Health Physics Department is being reorganized creating two distinct responsibility chains.

,

There will be a Supervisor of Operations and Supervisor of Support. Each will :

l report directly to the Station Services Superintendent. With this division of '

responsibilities, it will allow increased management attention during outage Audits conducted by the NRC subsequent to the July,1987 audit of Millstone Unit No. I have not identified any problems in this are !

Work Efficiency '

As a result of the Health Physics Department reorganization, Health Physics personnel will be better able to work with job supervisors to improve both job ,

'

planning and work effi lenc ,

ALARA Prooram

+

The ALARA goal setting proc'ess is a joint effort between the corporate office and the Millt. tone unit The 1987 and 1988 ALARA goals program was more effective and corrected many of the problems noted in 1986. Additionally, in ,

!

the development of future man rem goals, more consideration will be given to

!

comparison r,f the man hour estimate used to develop the goals with actual l man hour values from past experienc t

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A major impetus in reorganizing the Health Physics Department was to improve the teplementation of the ALARA program at the station. The new organization will better utilize station ALARA coordinators in achieving this goa ,

Evidence of improvementg3jn our ALARA program can be found in the Staff's Environmental Assessment associated with our amendment request concerning i an extension of Millstone Unit No. 2's operating license, which concluded that l (3) D. H. Jaffe letter to E. J. Mroczka, "Extension of Facility Operating ;

License for Millstone Unit No. 2 (TAC #64245)," dated January 6. 198 L

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'significant progress towards establishing an effective ALARA program" is being mad Audits The SALP report stated that although procedural requirements were met, audits of the Radiation Safety Program, which were performed by the corporate staff, were compliance oriented rather than performance oriented, in that procedure adherence was audited but not procedure and program adequacy. Concerns were also identified with the incependence of auditors, specifically in the dosimetry area, in that both the auditors and the dosimetry group reported to the same superviso In the summer of 1987, NNECO recognized the need to make the following improvements:

o The responsibility for performing audits has been transferred from the Radiological Assessment Branch (RAB) to the Quality Services Department (QSD). The QSD now has responsibility for audits of the Health Physics Program, effective January 198 o The RAB has begun (as of January 1988) performance oriented apprais-als of the Radiation Protection Program to provide an overview for station and corporate managemen o The station Health Physics Department has begun a self assessment program of their activitie These actions above will serve to strengthen the entire Radiation Protection Progra (gntrol of High Radiation Areas (HRA)

The SALP HRA Our resort Letter discussed dated Aprilweaknesses 22, 1988 iglving the posting and control of discusses our intent to use the 18 inch criterion for measuring areas greater than IR/hr. This action will eliminate a number of HRAs previously required by excessively conservative interpretations. This criterion is stipulated in Millstone Unit No. 3 Technical Specification 6.12. License ame.idment requests will be processed for Millstone Unit Nos. I and 2 to formally document the 18 inch criterio We will also continue our policy of increased surveillance and double locking of doors, in selected instance (4) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, 'High Radiation Areas,* dated April 22, 198 _ . _

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Attachment A l* A07137/Page 7

l themistrv

~

I The chemistry program at Millstone has significantly improved during this SALP l

i evaluation period. The strengthened progras has been positively acknowledged by both the NRC and INPO, as evidenced by the fact that it received two INPO

' Good Practir.es' during their 1987 evaluation, and that it has not received any violations from the NRC during this period. NNEC0 realizes, however, that further improvements are possible, and will continue our comitment to support j the necessary changes for such improvement i The SA1.P report stated that ' weaknesses in laboratory calibration techniques l indicated minor inattention to detail.' In reviewing the most recent NRC Non radiological Chemistry Re (NRC Region 1 !aspection Report Nos, i 50 245/87+28 and 50 336/87-24),portit was assumed that the SALP report co

alluding to two observations made by the visiting NRC Inspecto !

o It was stated that the calibration curve for hydrazine was not j statistically fitted i.e.,

.

visually discerning th(e "bestthe fit'calibration curve line through thewas drawn by calibration points, as opposed to using the least Squares criterion to determine i

the best fit line).

o It was mentioned that the dail on a single point calibration.y ammonia calibration curve was based

~ As discussed with the NRC Inspector during the time of Inspection No /87 28 and 50 336/87 24, the above two items are not deficiencies in j calibration techniques. First, it is common practice to visually discern the

i best fit line through a set of data points. Second, the millivolt meter used in the ammonia analysis only accepts one calibration point. Further, a second

)

independently prepared standard was utilized as an unknown (near the expected j sample concentration) to verify the accuracy of the ammonia curve. The slope

,

of the millivolt meter was checked on a weekly basis. We believe these

!!

factors temper the significance of the inspector's comments. Nevertheless, f

the Chemistry Department did respond to the NRC Inspector's comments by:

o Mritjag A lf ut_1guirtLRrocedur Now, all multipoint calibration j curves are statistically fitted,

! o I

Danging_the _ method of analysis _.fDr_smmoni Since the millivolt meter would accept only one calibration standard, it was decided to

{ do ammonia by a colorimetric analysis that incorporated a multipoint

calibration curve,

\

{ Whole Bedy Countina

!

The SALP report cited a deficiency in the Whole Body Counting QC Program which

!

indicated a lack of attention to detail in this area. Personnel assigned to

perform the various functions of Health Physics support, such as Whole Body i

Counting, will be completely trained as to their responsibilities and duties, This training will be documented and confirmed by a Practical Factors Oenonstration Sheet,

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A07137/Page a

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Functional Area: MAINTENANCE Millstone Unit Nos. 1 and 2 Board Recomendation: Non Resconse:

NNECO agrees with the Staff's assessment of our performance in this functional are We acknowledge the need to address the issues identified as having room for improvement, and will continue to strive for better performanc .

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, Attachment A l A07137/Page 9

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Lingtlpaal.Ar_gn: SUR'!Elt, LANCE Millstone Unit No. 1 BP.a!L89.C.2Eithht10D: Non EttrDnJt:

MFCO ayces with the Staff's assessment of our performance in this functional are Ve acknowledge the need to address the issues identified as having room for irrprovement, and will continue to strive for better performanc Millstone Unit No. 2 EnAr1Etcoy9ndation:

- Improve the evaluation of ECT dat Improve contractor oversight and contro Etninit:

ita n_0enerator Surveillance Activities The $ ALP report stated that steam generator eddy current testing d:ficiencies existed and exhibited ineffective QA/QC review of eddy cur rent (ECT)

data reduction and evaluation. It was stated that this issue was so important that it was a major element of performanc NflECO takes exception to the Category 2 performance rating assessed for this functional area, for the following reasons:

o The 1986 ECT results were totally accurate for the tube which

-

developed a leak six weeks following return to power, o A data review was conducted as a proactive response to the tube leak. (Thisreviewwasnotrequired.)

o Thirty eight nondefects were reclassified as defects by the data

,

revie o Thirty seven of the defects resulted from one analyst who did not follow the conservative ECT data analysis guidelines which were in effect at the time, o The great majority of these reclassified defects were either within the i 10 percent ECT accuracy limit of being nondefective, or subject to legitimate interpretation differences, o NU's development of optimized inspection techniques made det : tion of the reclassified defects possible, where standard techniques would have had more difficulty in detecting the defects in the first plac _ _ _ - - - - - . - - - - . - - - - - - - - - - - - - - -

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Attachment A A07137/Page 10 i,

The inappropriate nondefect calls represented a correct call rate of 99.97 percent for the 25,000 tube ends inspecte A detailed discussion of what we judge is an exemplary steam generator inspec-tion program is provided below. We request that the SALP Board review our program discussion and consider revising the performance rating, bckgroun Northeast Utilities (NU) has gained a reputation as an industry leader in maintaining a comprehensive, state of the art steam generator ($G) inspection program. This reputation has come about as a result of NU's comitment to a quality program which consists of:

o Pre inspection qualification of data acquisition and analysis techniques, o Pre inspection planning to select test methods needed to assure SG integrity, o Assessment of inspection results, o Conservative definition of tube repair need NU has assisted the industry in maintaining a proactive approach to improving SG inspection technology through membership and active participation on both an ASME Task group, chartered to revise SG inspection requirements, and a Steam Generator Reliability Project (SGRP) Committee, chartered to revise the PWR SG Inspection Guidelines. The purpose of these committees is to provide the industry with guidance for conducting optimum SG examination Refueling Outige Inspection A well planned and executed SG inspection program was conducted during the 1986 refueling outage. Prior to this outage, much pre-inspection engineering and planning were performed. These activities included qualification of inspection techniques for eliminating interfering conditions which could have made tube flaw identification difficul Specifically, first time use of digital, three. frequency mixing to minimize both denting and tubesheet inter-ference made flaw detection at the top of the tubesheet optimum. If standard techniques had been used, many of these flaws would not have been detected, and if they were detected, many would have been classified as distorted signal The NU qualification program for the three-frequency mixing technique involved the manufacture of several prototypical flawed tube sample Knowledge gained through previous pulled tube destructive examin tions formed the basis for the design of the flawed tube samples. The final phase of the three-frequency six qualification program involved testing of the samples and manipulation of the analysis software until accurate flaw detec-tion and sizing were achieved. This was one example where NU defined a problem with standard ins)ection techniques, and took the initiative to develop and qualify an alternate technique which would provide superior capabilities for SG tube flaw detection. Development programs like this have gained NU industry recognition as having one of the very best eddy current (ECT) programs in the industry.

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L Attachment A

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AW137/Fage !!

The 1986 SG inspection program included the use of state of-the art equipmen Data acquisition equipment included digital, multi frequency instrumentation, a specialized, narrow field bobbin coil probe for copper suppression and pit detection, and rotatin or undefined signals. g pancake coil (RPC)

Data analysis probes forby was performed characterization of new contractor personnel ucing digital data analysis equipment (DDA 4). Data analysis techniques included two frequency mixing for dent and top of tubesheet signal suppres-tien. Prior to the 1986 field ECT inspection, each data analyst went through a twu day "informal" training program on data analysis criteria specific to Millstone Unit No. 2. (In this case, ' informal' refers to the fact that no written serformance without demonstrations were required of the data analysts.)

As a means of providing greater assurance that tube defects would not be missed by the data analysts NU required that each data tape be independently reviewed by two different analyst This is another example of NU exceeding requirements in order to achieve the (highest quality exam possible.) Follow-ing the reporting of the examination results by the "primary" and "secondary" analysts, the results were compared and discrepancies were identified. This comparison was performed using NU's computer based data management syste A lead analyst (typically a Level !!!) was responsible for resolving all identi-fled discrepancies and for reviewing all defect calls for final confirmatio The purpose of the defect review was to minimize the number of unnecessary tube repairs due to false calls. This served to eliminate unnecessary radia-tion exposure which would have been received during the tube repair proces The 1986 ECT program also included additional optional testing to monitor tube damage precursor Above tubesheet sludge height measurements were made in each SG plenum in an effort to bound the region most susceptible to tube corrosion and to monitor the effectiveness of the sludge lancing proces Tube denting has been routinely monitored using eight coil profilometry techniques. The information gained by the profilometry exam includes measure-ment of tube dent sizes, calculation of tube strain and determination of the effectiveness nf corrective actions in halting further denting progressio The preceding discussion clearly illustrates NU's action in maintaining a high quality, state of the art ECT inspection program which goes well beyond minimui requirceents and industry norm Also shown is NU's willingness to share its experience with the industry through active participation in devel-oping industry standards for SG surveillance. The SALP Report recognized NU's SG surveillance program as being comprehensive, demonstrating good initia-tives, being well planned, and containing detailed procedures which emphasized satisfactory performance of measurements, hnEty_1917._QWilat The 1936 ECT inspection program was cooleted in Decembe Six weeks follow-ing return to power, a primary-to seconc ary Although leakage limits had not been excee(ded, the plant was shut downP/S) in leak w order to repair the lea It was determined from hydrostatic tests that the leak was coming from the hot leg side of Tube L25/R19. A review of the ECT data from the 1986 outage revealed a small volume, 31 percent throughwall indication within 1/2 inch of the top of the tubesheet. This same indication had been present in 1985 and had not shown any signs of change between 1985

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Attachment A

A07137/Page 12 and 198 Bobbin and RPC testing of Tube L25/R19 during the 1987 Outage revealed a large volume 100 percent throughwall indication at the top of the tubeshee Ihe_t es_t_tef ult sha rl y d e mo n11r111d_that_the t h rops hw a l l d e f e c t dcyclgptd after the 19S6 ECT inspection had been ptIfAmed, and_ that the 1986

[CLrelyltLwere correct 2 _

A subsequent removal of this tube for detailed, destructive, and nondestructive examinations was performed during the 1988 outage. The small volume indication was shown to be a band of pits located about 1/2 inch above the top of the tubesheet confirming the 1986 ECT result The throughwall indication was a circumferential crack located below the band of pits at the top of the tubeshee This defect was responsible for the leak. No eddy current evidence of this defect was present in the 1986 exami-nation. The rapid flaw progression rate observed can occur with stress corrosion cracking, explaining why the flaw was not present six weeks prior to the lea At the time the leak was first identified, NU began a review of the 1986 ECT data tapes for those tubes believed to be most suspect. In the data review two tubes were identified as containing defects. These defects had not been reported during the 1936 ECT examination, it was then decided to expand the data review until all possible ' missed" defects were identifie The review of the 1986 ECT data tapes revealed a total of 38 tubes which contained defects (flaws of greater than or equal to the 40 Jercent throughwall repair limit). With the exception of one defect missed as a result of incorrectly identifying the test extent, the 37 other unreported defects were initially reported by either a primary and/or secondary analyst, but were subsequently reclassified to nondefects by one lead analyst. As mentioned above, the leaking tube, L25/R19, was not one of the unreported defects, and is entirely unrelate Cases for Missed _ Defects One defect was not reported due to an incorrect test ' extent" determination by the primary analyst. The primary analyst indicated that the test was complete to, and restricted at, the first eggerate support, with no flaws detecte This determination was incorrect since the probe was actually restricted at the top of the tubesheet and, as a result, no inspection data were collected between the top of the tubesheet and the first eggerate support. The tube was retested with a smaller diameter probe, but since it was believed that data had been previously collected and analyzed to the first support, analysis of the ECT data from the second test was performed in the area of the first support. (This analysis criterion was consciously selected so as to avoid redundant analysis.) As it turned out, the defect was located in the analyzed section of tube above the top of the tubesheet. To avoid recurrence of this situation, all inspections since the 1986 Millstone Unit No. 2 outage have required an independent verification of test extent and all data collected have been analyzed, even if the tube was tested in the same area more than onc Differences in analyst / lead analyst interpretation of the ECT data was respon-sible for the 37 other unreported defects. As stated previously, defects were initially identified in all 37 tubes by at least one analys However, the lead analyst (and in some cases the primary or secondary analyst) did not

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Attachment A A07137/Page 13

conclude that a defect was prosent and the lead analyst overruled the defect '

cal Thirty one of the thirty seven overruled defect calls were either within basic ECT fleid tolerances or not detectable with conventional methods. Many of the tube flaws which were assigned a depth estimate of greater than 40 percent throughwall during the 1987 data review, had previously been assigned depth estimates of between 30 and 39 percent throughwalls i.e., within 10 percent of being pluggabl Variables associated with estimating flaw deptis are an accepted limitation of the ECT technique and, as such, were considered when the tube repair limit was established. The three-frequency mixing method, used to analyze data at the top of the tubesheet, allowed for greater accuracy in flaw detection and sizing than had previously been possible with two-frequency mixing. Despite the obvious advantages of using the three frequency mixing technique, associated complexities with the technique lead to differ-ences in data interpretation in some cases. These differences resulted in additional defects being identified during the 1987 data revie It is important to note that had standard methods rather than the three. frequency mix been used, many of the reclassified defects would have gone undetected in the first plac Six of the thirty-seven overruled defect calls are considered to be inappro-priate final call The one lead analyst responsible for the six calls did not apply the Millstone Unit No. 2 specific analysis criteria which were in effect at the tim (qr m tive Actions .

Following the discovery of unreported SG tube defects during the 1986 Millstone Unit No. 2 outage NU initiated an aggressive prograin to identify those actions needed to prevent similar occurrences in the future, and imple-mented those corrective actions at each of its PWR plants. The following discussion will focus on the major corrective actions which NU has implement-e Detailed, plant specific guidelines were developed for data analysis, data reporting, and discrepancy identification and resolution. The guidelines for data analysis provided assurance that uniform analysis criteria were estab-lished and that, in cases where ambiguous test data were collected, conserv tive analysis practices were use Criteria for defining discrepancies between primary and secondary analysis results were also formulate These criteria included the establishment of allowable tolerances for flaw depth, flaw height and test exten The procedure for overruling defect calls was also modified so that no one analyst could eliminate a previous defect call. The revised procedure requires agreement between two lead analysts to overrule a previous defect call. Where agreement cannot be reached, the more conservative call is adopte In order to assure that data analysts are appropriately qualified and thoroughly understand the plant specific analysis guidelines, a training and formal performance testing program will be instituted prior to each ECT inspectio Prior to the January 1988 Millstone Unit No. 2 outage, for

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Attachment A A07137/Page 14 example, each of the 31 data analysts went through a five day training and qualification program. This program consisted of a review of the Millstone Unit No. 2 data Analysis Guidelines Document, review of selected field ECT

,

data tapes, and an eight hour examination. Satisfactory completion of the

,

examination was required for an analyst to be allowed to analyze data during

,' the field exam. A training program of this type assures that only qualified analysts, who are familiar with plant specific guidelines, will be allowed to perform data analysis at a particular plant, thus resulting in the most !

accurate and consistent examination possible. NU has also increased contrac-tor oversight to assure compliance with each of the above mentioned specif t-cations, procedures and guideline Implementation of the preceding corrective actions at Millstone Unit No. 2 during the January 1988 outage was given favorable reviews by the NRC, as were the ECT programs at the Haddam Neck Plant and Millstone Unit No. 3. The corrective actions taken which resulted in the favorable 1988 outage experi-

ence were taken before the start of the outage, i.e. during the SALP interval i in question here. Many of the items that are part of NU's ECT program will be ;

included in the EPRI PWR SG Inspection guidelines, now being revise Contractor Ovenight and Control During the 1988 steam generator eddy current examination program, three tubes I which should have been plugged were found unplugged on one end. Three tubes l adjacent to the unplugged tubes were inadvertently plugged on that en The i SALP report identifies this as an example of the need to better control i contractor work activitie ,

This work was performed by a contractor using their Remotely Operated Service !

>

Arm (ROSA). No templates or other tube sheet marking are required when ROSA is used for such repair operations. This results in significant reductions in [

t J personnel radiation exposure. This reduction in personnel exposure was one of (

the significant factors considered in the choice of this repair optio t

,

$1nce templates or tube sheet marking are not used to control plug installa- i

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tion location a detatied written procedure is provided to index ROSA to the i proper locatio Indexing is required after major location changes and '

! periodically during extended operations in one locatio '

In this instance, the transfer of the coordinates to the ROSA computer was not a third party verification poin Thus the operator's error was not detected l

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by the quality orgar,trations, i During the 1988 refueling outage, steam generator tube repairs were performed ;

with different equipmen This equipment required the use of templates for i position verificatio Continuous monitoring was provided by Northeast t

, Utilities quality control personne [

While the failure to plug the required tubes was a serious error, Northeast

'

Utilities considers that continuous independent inspection verification by '

company quality control personnel, combined with the routine observation of r wort control and work practices by Northeast Utilities supervision and manage- j eent personnel, is the optimum level of oversight to ensure proper contractor performance, r l

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Attachment A A07137/Page 15 Suvare and Conclusions We concluded that the sal.P rating appropriate for steam generator ECT survell-lance is Category 1. The 1986 SG ECT program was well designed and executed, featuring application of advanced equipment and techniques. These features srovided an optimum plant specific inspection that would not have been possi-ale if standard techniques were used. Although a review of the 1986 ECT data resulted in reclassification of 38 nondefects to defects, all but seven of these reclassified defects were either within the accepted 110 percent ECT accuracy limit of being nondefective, or subject to legitimate interpretation differences. Furthermore, the 1986 ECT results for the tube which developed a p/S leak in 1987 were demonstrated to be 100 percent accurate by subsequent destructive examination of the tube. Iht leakjng_1ght_witcomplettly unrelat-rd_to_the ECT _ anab11Lrrob11s Many of these reclassified defects would not have been detectable with standard ECT methods. Differences in interpre-tation, associated with the advanced ECT methods, have been effectively addressed by NU in conservative guidelines of industry model quality, that were developed subsequent to the 1987 Outage, and teplemented during the 1988 outag The inappropriate nondefect calls in the 1986 inspection represent a correct call rate of 99.97 percent when one considers the fact that over 25,000 tube ends were inspecte . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _

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Functional Area: ENERGENCY PREPAREDNES$

Millstone Unit Nos. 1 and 2 goard Rtcommendation: Non ,

Et192n11:

NNEC0 agrees with the Staff's assessment of our performance in this functional area. We acknowledge the need to address the issues identified as having room for improvement, and will continue to strive for better performance, i

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' A07137/Page 17 functiqnt]_.stu: SECURITY AND SAFEGUARDS Millstone Unit Nos. I and 2 Spard Recamendations:

Liun Re evaluate efft.ctiveness of security self assessment function, assuring thd program adequacy aspects are evaluated in addition to program complianc Reassess effectiveness of management overview of securit Reassess adequacy of the security trainin MC: Review licensee security program to the effectiveness of corrective actions on the security inadequacies which resulted in escalated enforce-r.cnt actio Reh m :

ielf hsessnent The NUSCO System Security Group and the NUSCO Quality Services Group will be conducting future audits. This coordinated approach will provide a more detailed and in depth audit function that will address both program adequacy and complianc This will enhance and strengthen the self assessment function. A revision to the Physical Security Plan will bt submitted to identify responsibilities of each group conducting security program audit MLnagement Overview Two new positions have been created within the NUSCO System Security Group, Manager Nuclear Security and Senior Agent - Nuclear. These two positions have been filled and will provide two additional people to conduct audits, support program modifications and upgrades, and provide additional assistance to the statio Corporate security management will resume active involvement in industry groups engaged in nuclear plant security matters, including the Region 1 Nuclear Security Associatio The site ranagement of the Security Department has been restructured. This change defines three specific areas of responsibility - Operations. Adailnis-trative, and Industrial Security functions and provides a supervisor or administrator for each are This provides a more reasonable span of control for the Security Supervisor and clearly separates the operational group, allowing it to be more effectiv IIAltdnifESt45 The reorganization of the Security Department will allow for better oversight in the area of training. The restructuring places training under the auspices

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A07137/Page 18 of the Assistant Security Supervisor Operations. This link will allow for the development of a more performane.ed based program in direct support of operational needs and concerns. Continuing assessment of this approach will be done by NNECO, ac well as the security force contracto Performance related Events The SALP report stated that NNECO needed to determine the root cause of performance related events and increase its oversight of the contractor to prevent recurrence. The root cause appears to be inattentiveness to routine guard force task performance by the contractor. As a result of the Security Department reorganization, oversight of the guard force will be increase This is intended to decrease the number of performance related event Sumary NNEC0 believes significant strides in improvements in this area have already been realized. This belief is supported by the recent Report No. 50 245/88 04, 50 336/88 08, and 50 423/88 06.ggC Staff Inspection

_

(5) R. P. Bell-e ietter to C. J. Mroczka, ' Combined Inspection Report No /8: M 16/R.Y J3, ann '4 423/88 06,' dated May 3, 198 _ - . - . -.

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. Attachment A

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A07137/Page 19 Eunti.lqu1 Area: OUTAGE MANAGENENT

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Millstone Unit No. 1 I

land _fltsomendation: Non Et1Mhtt:

The SALP report stated that efforts should be made to clear routine work prior to outage commencement. Valuable PORC time was spent reviewing routine procedure changes and other items that could have been accomplished prior to the outag NNECO agrees with the Staff assessment that where appropriate, routine items should be cleared prior to outage commencement. With more up front planning of outage work and plant design changes being currently trplemented, these types of items should be PORC approved prior to the outages and reduce the PORC review time during outages. Some routine items necessarily must be reviewed during outages. These items will be minimized to the extent practt-ca Millstone Unit No. 2 Board Reco-endation: None, f!nPQA11:

linLCenerator ECT Data Analysis Please refer to our response for the Surveillance functional area for our discussion of our steam generator inspection program, httjs Overtime The SALP report discusses exarrples of overtime in excess of established guidelines without ranagement approva The applicable station procedure is being revised and a station policy statement is forthcoming to ensure this situation does not occur agai Page 41 of the SALP report refers to 'SG local leak rate testing.' The report should say * Containment local leak rate testing.'

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functional Area: ASSURANCE OF QUALITY Millstone Unit Nos. 1__and 2 Igard Recomendationt Non Resronnet This functional area is essentially a synopsis of the assessments in the other ten areas. Our responses for the other functional areas address some of the specific actions which are being taken to minimin and/or eliminate problems in those area The SALP report stated that additional attention is needed to control shelf itfe for materials that age in storage. Station procedures have been deve oped and imalemented to establish a Degradable Materials Control Progra Taese procecures provide the necessary controls over all new degradable material purchased for the Millstone Site and provides the methodology to maintain stock. Degradable awareness of the shelf life status of the material in warehouse material currently in the warehou.e has been reviewed to ensure that the recommended shelf life has not been exceeded. In the cases where the expired the shelf material life discarde has exp'd. ired, replacement material is being ordered and l

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i Attachment A 707137/Page 21

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functiontLaria: ENGINEERING SUPPORT

laird Reco*Jtadation: Non Etufdlit:

NNECO understands that the NRC Staff has recently begun evaluating licensees in this functional area and appreciates constructive criticism of our perfor-ranc We acknowledge the need to address the issues identified as having room for improvement, and will continue to strive for better performance. We are confident that it will be evident that a Category I rating is appropriate for our performance in this area in future SALP evaluations. We are provid- :

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ing comments below on three issues in which we believe the SALP board may benefit from additional information, l i

Millstone Unit No. 1 K11 TLC 9adinier Performinct *

The SALP report states that the recurrence of main condenser tube leaks requiring frequent power maneuvers to identify and repair needs design resolu-tio l t

Replacement of the condenser ha[6g extensively studied, and based on the I results of the !$AP evaluations outage duration, significant costs and limited payback, a management decision was made to defer the project. The .

project continues to be actively pursued, however, to find less costly alter- f natives and to closely monitor condenser and plant parameters to ensure all ,

technical and safety issues are addressed in a timely yet cost effective l manne '

t Since 1981, NNECO has periodically been evaluating the cost benefit of replac- !

ing the Millstone Unit No. I condense All costs have been considered, including replacement power, demineralizer resin beds, ALARA, resin disposal, i and maintenanc Based on the evaluations, replacement of the condenser was !

found not to be an economical alternative to dat l l

Considerable engineering support has been directed toward improving the :

maintenance of the existing condenser, since replacement was not an economic (

alternativ !

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l J. F. Opeka letter to C. I Grim <.s !

(6) *lSAP final Report for Millstone Unit l No. 1,' dated July 31, 198 '

i (7) C, O. Thomas letter to E. J. Mroczka, ' Millstone Nuclear Power Station, Unit N Draft Integrated Safety Assessment Report (NUREG ll84),' j I dated April 2, 1987, in which the Staff ranked this topic as "medium" in ,

trplementation priority, ;

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Attachment A

- /07W ' ' t ;- 22 e .. f instituted a 100% eddy current test program during outages, i

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2,e tube plugging, and hydrolazing in an effort to minimize tube *

11 t br In 1987, a tube failure analysis was performed to better understand the specific corrosion mechanism. As a result of this is NNECO analysis, several considering addition-modifying the al activities are now being pursue criterion for preventive tube plugging from 70% wall loss to 50% wall loss in an effort to reduce power reductions for tube plugging. The inlet water boxes ,

may be coated or replaced to remove the source of iron which has been a major contributor to the tube failure mechanism. Hydrolazing techniques are being reviewed based on recent developments in the area of tube cleaning. The water box cathodic p-otection system design is being reviewed to assess the need for increased protection. Daily chlorinationIn of each circulation water bay on a addition, we are lookinn at new year-round basis is being considere These actions are methods to repair damaged tubes through the use of lining intended to reduce the corrosion rate while minimizing the effects of tube leak A working group consisting of representatives from the Plant Maintenance, Operations, and Engineering Departments with the NUSCO Project Engineer has been established to ensure that these corrective reasures are incorporated on a timely basi In summary, we believe excellent engineering support regarding resolution of condenser problems has been provide LIILanLCerLSpray itSuu o Anchouse We are providing the following description of events which were associated with the discovery and correction of the anchorage concerns discussed on page 46 of the SALP report. The discussion in the report expressed concerns thatno a problem was discovered in 1984 with the anchorage of the pumps and It is important to note that thethat resolution was reached until 198 examination of the anchorage in a potentially destructive manner could only be conducted in an outage or LCO conditio There are some statements made in the SALP report which do not correctly The SALP report indicates reflect the situation found in the 1987 outag that short bolts were found in the 1987 outage when grout was removed from the pump base. The actual field condition did not reveal short bolts but did indicate a design change had taken place on some (not all) of the six pumps involved which was not reflected in the design drawing In early 1985, review of ultrasonic inspection data taken on the pump bolting showed major inconsistencies with what would have been expected based upon the design drawings. The deviations that were identified by direct use of the UT This evidence caused data varied dramatically from the design requirement us to suspect that some design change during construction had occurred and that the UT data did not represent the anchorage which was present. Efforts during 1985 focused primarily on trying to obtain records of the suspected

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. Attachment A

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A07137/Page 23 design change, gathering and evaluating vibration data, and trying to develop a non destructive method of qualifying the actual support condition In order to address this issue fully, three major items were undertaken:

o EBASCO (the original Architect Engineer) was commissiwd to inves-tigate their records for possible construction changes and to design an alternate support system for the pump e The scope of our program addressing IE Bulletin 7914 was expanded to generically address pum tions and nozzle loadings.p and equipment anchorage as built condi-o All other safety related pumps were inspected to address potentially generic tepact issues, and the anchorage was inspected using the UT methods which had revealed the problem with the CS and LPC; pump The results of this inspection and one local grout removal showed no deficiencies in pump anchorage from the design conditio Priority was given to the redesign /reovaluation of the CS and LPCI pump supports and nozzle EBASCO completed their review and redesign following the October, 1985 to January, 1986 refueling outage. All engineering and design activities were put into place to support the 1987 refueling outag Inspections were performed on some suspect bolting on the CS pumps during the 1997 outage. This inspection required the chisping of some concrete under the pump base to expose the embedded fasteners. T11s investigation confirmed that a design modification was done during construction wherein the suspect bolting was bent over and welded to a plate. The bolts which had shown short length measurements with the UT examination were attached to this plate. MECO opted to icplement the modification prepared by EBASCO following the confirmattor, of the actual as built condition of the LPCI pumps as the same condition was assumed to exis The design modification is a significant enhancement to the design of the pump anchorage with a design capacity significantly in excess of our licensing (ISAR) requirements. The existing anchorage was not found to be non compliant with the original licensing requirements for capacity. This evaluation would have required additional destructive inspections to fully quantify the condi-tion of the as built anchorage. Design enhancemente were made in the revised pump anchorage configurations including consideration of nozzle lo: dings for safe shutdown earthquake and the Mark I containment program for piping connected to the toru C9nJ:M9ai We believe that the actions taken in response to the identified concerns tepresented reasonably timely and comprehensive actions to answer a difficult question. The operability of the involved equipment was reviewed to assure safe plant operation while the evaluation was underway. All generic irplica-tions nf the inspection data were addressed and susceptible equipment inspect-e _

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, Attachment A

. A07137/Page 24 Millstone Uni _t No. 2

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Imoroner Overcurrent Trin Setooint The subject reactor trip was not caused by an improper trip setpoint. The trip was caused by an cul filled overcurrent trip device for the pressurizer heaters, which malfunctioned and operated before its trip setpoint was reached. All oil filled 480 volt load center breaker trip devices have now been replaced with solid state trip devices, which maintain their trip setpoints better.

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Attachment A A07137/Page 25

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functional Area: TRAINING EFFf#.TIVENESS Millstone Unit Nos. 1 and 2 paard Recomendation: Non RUPanit:

NNECO agrees with the Staff's assessment of our performance in this functional are We acknowledge the need to address the issues identified as having room for improvement, and will continue to strive for better performanc Simulator txams The SALP report stated that, for Millstone Unit No. 2, the format of simulator examinations did not allow for adequate follow up questioning to distinguish individual weaknesses from group weaknesses. The conduct of simulator exami-nttions has been strengthened. The current format of simulator examinations allows for the identification of both individual and group weakness. Expected operator responses are included in the scenario evaluation forms. Additional-ly, oral examinations, related to the simulator scenarios, are administere Additional oral questions may be asked at the discretion of the examiner to follow up on identified weaknesses. Time is provided at the cortpletion of the scenarios for this purpos _ _ _ _ _ _ - _ _ _ _ _ _

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Functional Area: LICENSING ACTIVITIES

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Igard Recomendations:

Litin11t The licensee should identify any needed schedule delays to the NRC staff at quarterly meetings rather than adopt such delays unilater.

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all !HtG The NRC staff should closely monitor the licensee's progress in I

meeting their licensing obilgations and comitments.

i Resconse:

NNECO is continuing to implement maasures to maintain and improve frequent comunications with the NRC Staff concerning licensing activitie The Board's recomendation to identify schedule delays during quarterly (or more frequent) meetings will be implemented, permitting the NRC Project Managers to

improve monitoring of our comitment 1rplementation in advance. In these I meetings and in daily comunication with the Project Managers, our plans to address significant issues will be discussed, with bilateral prioritization of i activities. These measures, together with the improvements 1n our comitment tracking pro obligations, gram, will provide increased awareness of meeting regulatory i Management actions taken during the latter part of the SALP period resulted in

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measurable improvements in the timeliness of submittals to the NRC Staff. In j

addition, new corporate wide guidance concerning timeliness of regulatory correspondence has recently been issued, which provides guidelines for timely

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technical input, letter preparation, management review, and submittal to the NR !

Adherence to these guidelines will farther ensure timely responses to NRC requests for information, or early recognittan and notification to the NRC

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a Project Manager of potential schedule delays. Formalization of the Integrated

' Irplementation Schedule at Millstone Unit No.1 is also expected to irprove the visibility of the bases for ou' resource allocation decisions.

) The Millstone Unit Noa. I and 2 plant staffs and licensing personnel have been

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actively addressing timeliness of NRC required submittals. Internal comit.

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ments have been made to better the due dates of sutaittals. Also, a concise

listing of high visibility licensing issues has been developed and is being utilized in frequent meetings to track the issues and ensure timely action is being taken by the responsible individuals.

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UNITED STATES

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g y, *.[<** NUCLEAR REGULATORY COMMISSION

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'! os Atts A rcomo ENCLOSURE 4

,, / KING oF PRUS$1A.PENNSMVANI A Wo6

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Docket No. 50-245; 50-336 M 2 9 1988 Northeast Nuclear Energy Company

- ATTN: Mr. E. J. Mroczka Senior Vice President - Nuclear

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Engineering and Operations Group P. O. Box 270

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Hartford, Connecticut 06141-0270 Gentlemen:

Subject: Combined Systeratic Assessment of Licensee Performance (SALP) Report No /66-99: 50-336/86-99 (6/1/86 - 12/31/87)

in February 25, 1988, the NRC Region 1 SALP Board assessed the performan:e of Millstone 1 and 2 for the 19-month period between June 1, 1986 and December 31, 1987. Those assessments are documented in the enclosed enmbined report. We will contact you soon to schedule a n:eeting to discuss the SALP evaluation At the meeting, please be prepared to discuss the assessments and any plans you have to improve performance. You may, of course, provide any comments you have

,- regarding the SALP at the meeting. Al,o, you may provide written comments within

g 20 days after the meetin Thank you for your cooperatio

Sincerely, fd.) h William T. Russell Regional Administrator Enc 1csre: NRC Regicn 1 CemDined SALP Report 50-245/86-99; 50-336/86-99

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Northeast Nuclear Energy Company 2 M 29fg k

cc w/ encl:

W. D. Romberg, Vice President, Nuclear Operations S. E. Scace, Station Superintendent D. O. Nordquist, Manager of Quality Assurance R. M. Kacich, Manager, Generation Facilities Licensing Gerald Garfield, Esquire Chairman Zech Commissioner Roberts Commissioner Bernthal Commissioner Carr Commissioner Rogers K. Abraham, RI (13 copies)

Public Document Room (PDR)

local Public Document Room (LPDR)

Nuclear Safety Information Center (NSIC)

NRC Resident Inspector State of Connecticut (

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ENCLOSURE _5 MRIL_23u1938 SALP MANAGEMENT MEETING ATTENDEES 1. Licensee Attendees B. Fox, President, NortFeast Utilities J. Opeka, Executive Vice President, Engineering and Operations E. Mroczka, Senior Vice President, Nuclear Engineering and Operations W. Romberg, Vice President, Nuclear Operations C. Sears, Vice President, Nuclear and Environmental Engineering R. Werner, Vice President, Generation Engineering and Construction L. Johnson, Director, Generation Engineering and Design D. Nordquist, Director, Quality Services R. Harris, Director, Nuclear Engineeting R. Kacich, Manager, Generation Facilities Licensing S. Scace, Station Superintendent, Millstone H. Haynes, Station Services Superintendent, Millstone J. Stetz, Millstone 1 Superintendent J. Keenan, Millstone 2 Superintendent B. Ruth, Manag2r, Operator Training 2. HRC Attendees W. Kane, Director, Division of Reactor Projects L. Bettenhausen, Chief, Projects Branch No. 1 E. McCabe, Cnief, Reactor Projects Section IB W. Raymond, Senior Resident Inspector, Millstone L. Kolonauski, Resident Inspector, Millstone 1 P. Habighorst, Resident Inspector, Millstone 2 S. Barber, Resident Inspector, Millstone 3 J. Stolz, Project Dire-torate, NRR M. Boyle, Millstone 1 Project Manager D. Jaffe, Millstone 2 Project Manager

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IR 05000245/1986099

Text

REGION I

REGION I

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