ML20247H189
| ML20247H189 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 09/12/1989 |
| From: | Sears P Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8909190238 | |
| Download: ML20247H189 (124) | |
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[> LICENSEEr >Yenkbe Atoinic Electric Company f" y x
' TACILITY:, Yankee Nuclear Power Station at Rowe
[SUBJECTF LNEETING TO PRESENT PLANT LIFE EXTENSION AND INDIVIDUAL PLANT D AMINATION (IPE) STUDY BY LICENSEE d ( <r '
;This meeting was held'on August 28 and 29, 1989 at the licensee's offices in Bolton,. Massachusetts. The licensee operates the lead pressurized water reactor in the industry Plant- Life Extension (PLEX) Program and will submit the.first license renewal application under the program. The NRC staff received a status briefing on Yankee's PLEX Progrum and Individual Plant , Examination Study. A list of attendees is enclosed, as well as two reports 4 containing. hand-out the material discussed at the meeting.
The licensee presented his overall program status description of the licensing documents they propose to submit and the tentative schedules. y , Yankee Atomic personnel described the screening methodology and engineering
~
evaluations used to determine which systems and systems components should be
' included in their aging analyses. The licensee has computerized these
" processes using commercially available artificial intelligence and data base programs.
4 The licensee presented their Probabilistic Risk Assessment (PRA) and IPE program. The staff was also given presentations on licensing and policy
. issues, and maintenance and engineering evaluations.
/s/
Patrick M. Sears, Project Manager Project Directorate I-3 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation i ,
Enclosures:
As stated lcc w/ enclosures: See next page
,1 [MeetingBolton,Ma8/28-29]
*SEE PREVIOUS CONCURRENCE PDI-3/DRSP
- PDI-3/DRSP/PM* NRR/PMAS
- D:PDIf DRSP I i MRushbrook PSears:rw FGillespie RWessrhan I t 9/7 /89 9/ 6 /89 9 /11/89 /g,/89 8909190238 890912 PDR ADOCK 05000029 P PDC
September 12, 1989 Docket No. 50-029 l
\
LICENSEE: Yankee Atomic Electric Company FACILITY: Yankee Nuclear Power Station at Rowe
SUBJECT:
MEETING TO PRESE.NT PLANT LIFE EXTENSION AND INDIVIDUAL PLANT EXAMINATION (IPE) STUDY BY LICENSEE This meeting was held on August 28 and 29, 1989 at the licensee's offices in Bolton, Massachusetts. The licensee operates the lead pressurized water reattor in the industry Plant Life Extension (PLEX) Program and will submit the first license renewal application under the program. The NRC staff received a status briefing on Yankee's PLEX Program and Individual Plant Examination Study. A list of attendees is enclosed as well as two reports containing hand-out the material discussed at h e me,eting. The licensee presented his overall program status description of the licensing documents they propose to submit and the tentative schedules. Yankee Atomic personnel described the screening methodology and engineering evaluations used to determine which systems and systems components should be included in their aging analyses. The licensee has computerized these processes using commercially available artificial intelligence and data base programs. The licensee presented their Probabilistic Risk Assessment (FRA) and IPE program. The staff was also given presentations on licensing and policy issues, and maintenance and engineering evaluations.
/s/
Patrick M. Sears, Project Manager Project Directorate I-3 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Enclosures:
As stated cc w/ enclosures: See next page [ Meeting Bolton, Ma 8/28-29]
*SEE PREVIOUS CONCURRENCE PDI-3/DRSP* PDI-3/DRSP/PM* NRR/PMAS* D:PDI DRSP MRushbrook PSears:rw FGillespie RWessrhan 9/7 /89 9/6 /89 9 /11/89 a /g,/89 l
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SUMMARY
VER. R0WE 8/28-29/89 i DISTRIBUTION !
' Docket F Ns. ' .$
NRC & Local PDRs P. Sears OGC E. Jordan B. Grimes ACRS (10) M. Rushbrook H,, B. Clayton J. Wiggins, R;;n. I i S. Varga T. Murley R. Wessman B. Boger F. Gillespie F. Akstulewicz M. Rubin D. Cleary, RES 1 1 e l
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[@Mo e g o UNITED STATES NUCLEAR HEGULATORY COMMISSION g .p WASHINGTON, D. C. 20666 [ September 12, 1989 Docket No. 50-029 LICENSEE: Yankee Atomic Electric Company FACILITY: Yankee Nuclear Power Station at Rowe
SUBJECT:
MEETING TO PRESENT PLANT LIFE EXTENSION AND INDIVIDUAL PLART EXAMINATION (IPE) STUDY BY LICENSEE This meeting was held on August 28 and 29, 1989 at the licensee's offices in Bolton, Massachusetts. The licensee operates the lead pressurized water reactor in the industry Plant Life Extension (PLEX) Program and will submit the first license renewal application under the program. The NRC staff received a status briefing on Yankee's PLEX Program and Individual Plant Examination Study. A list of attendees is enclosed, as well as two reports containing hand-out the material discussed at the meeting. The licensee presented his overall program status description of the licensing documents they propose to submit and the tentative schedules. Yankee Atomic personnel described the screening methodology ana engineering evaluations used to determine which systems and systems components should be included in their aging analyses. The licensee has computerized these processes using commercially available artificial intelligence and data base programs. The licensee presented their Probabilistic Risk Assessment (PRA) and IPE
' program. The staff was also given presentations on licensing and policy issues, and maintenance and engineering evaluations.
N Patrick M. Sears, Project Manager Project Directorate I-3 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Enclosures:
As stated cc w/ enclosures: See next page
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- Mr. George Papanic, Jr.' s cc w/ enclosures: j Mr. George Papanic, .Jr.
. Senior Project Engineer - Licensing Yankee Atomic Electric Company 580 Main-Street Bolton, Massachusetts 01740-1398 Dr. Andrew C. Kadak, President and Chief Operating Officer Yankee Atomic Electric Company
.580 Main Street Bolton, Massachusetts 01740-1398 Thomas Dignan, Esquire Ropes and Gray 225 Franklin Street Boston, Massachusetts 02110 Mr.'T.'K..Henderson Acting Plant Superintendent Yankee Atomic Electric Company Star. Route Rowe, Massachusetts 01367 Resident Inspector
' Yankee. Nuclear Power Station' U.S. Nuclear Regulatory Commission Post Office Box 28 Monroe Bridge, Massachusetts 01350 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road
-King of Prussia, Pennsylvania 19406 Robert M. Hallisey, Dir(ctor Radiation Control Prograr Massachusetts Department )f Public Health 150 Tremont Street, 7th Fioor Boston, Massachusetts 02111 Mr. George Sterzinger Commissioner Vermont Department of Public Service 120 State Street, 3rd Floor Montpelier, Vermont 05602 63 %
. 4 .
-__-_________.__ _____m-____________- _ _ _ _ _ _ _ ________.___._____._m________.-__.____.-m_____. _ _ . _ _ . ___________ . _ _ _ _ . ___-____-_.______.___ . _ _ _ _ - _ _ _ - _ _ _ _ -_ . _ _ _ _ . _ _ _ _ _ _ __ ___m__.3
ENCLOSURE SEVERE ACCJDENT/IPE AUGUST 28, 1989 ATTENDEES NRR YAEC-Patrick. Sears Andrew C. Kadak Frank Gillespie John De Vincentis Donalc < . .ary Donald W. Edwards Francis Akstulewicz Jane M. Grant Kamal Manoly John J. Carey-Jet Vora Joe McCumber Willism Szymczak William Hinkle John D. Heseltine STATE OF VERMONT NORTHERN STATES POWER Williem Sherman ' Jacqueline Gilchrist Terry Pickens EPRI VIRGINIA POWER Richard Burke D. R. Hostetler John J. Carey NUMARC HOPKINS:SUTTER Tricia Heroux Kathryn M. Kalbwsky SANDIA NATIONAL LABS Donnie Whitehead j. _ . _ _ _ - . ._______m.m____ __.-m-___.______________ m______-____. ______m
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ENCLOSURE
-l LICENSE RENEWAL /IPE AUGUST 29, 1989 ATTENDEES-g YAEC' Patrick Sears John Oddo Francis Akstulewicz George Harper Donald Cleary James L. Stub Frank Gillespie Alexander'R. Klein Mark Rubin David A. Rice Bob Harvey Jim Chapman Stephan P. Schultz Andrew C. Kadak Bill Hinkle John De Vincentis Ramu Sundarin John D. Haseltine Bruce W. Holmgren Jane Grant
-EPRI NORTHERN STATES POWER Richard W. Burke Jacqueline Gilchrist
- John Carey Terry Pickens-NUMARC HOPKINS:SUTTER Tricia Heroux Kathry M. Kalowsky STATE OF VERMONT William Sherman e
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g ... I. . I L [ LICENSE RENEWAL PRESENTATION I- FOR THE [ YANKEE NUCLE,AR POWER STATION [ - [ ~ [ I Yggg}EE [ : I I [ I 1 BOLTON, MASSACHUSETTS AUGUST 28-29,1989 [~ I *
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t EN. LICENSE RENEWAL' AGENDA August 28-29, 1989 I$ i 8/28 1:00-1:30 p.m. Introduction (J. Haseltine) 1:30-4:00~p.m. Review & Evaluations (J. McCumber/W. Szymezak))
'4:00-5:00 p.m. Application Overview (W.'Hinkle) 5:00-5:15 p.m. Yan". tee Schedule (J. Haseltine) g' ,5:15-6:00 p.m. Overall Yankee License Renewal Process V; (J. Grant /A. Kadak)
S/29
, 8:30-9:30 a.m. Discussion /Q&A - Previous Day's Topics I
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k I 08/25/89 SYSTEMS AND STRUCTURES PAGE NJ. 1 10:55:01 MEETING STEP 1A CRITERIA FOR FURTHER REVIEW g l========l========l==========================================: l RECORD l SYSTEM l SYSTEM / STRUCTURE :::::====ll l NUMBER l CODE l TITLE l _. l l:::::===l========l:::::::::================:r==========================l l l 1 l AR l AIR REMOVAL SYSTEM l l l 2 l AS l AUXILIARY STEAM SYSTEM l l 3 l BF l BOILER FEED SYSTEM l l 4lCH l CHARGING AND VOLUME CONTROL SYSTEM l l 5 l CS l CHEMICAL SHUTDOWN SYSTEM l l 6 l CW l CIRCULATING WATER SYSTEM l l T l CC l COMPONENT COOLING WATER SYSTEM l l 8 l CA l 00MPRESSED AIR SYSTEM l l 9 l CI l CONTAINMENT ISOLATION SYSTEM l l l 10 l Da' l DEMINERAL12ED WATER SYSTEM l 3 l 11-l EBF l EMERGENCY BOILER FEED SYSTEM l l 12 l ES l EXTRACTION GTEAM SYSTEM l l 13 l FW l FEEDWATER CONTROL SYSTEM (I&C) l l 14 l FS l FIRE DETECTION AND SUPPRESSION SYSTEM l l 15 l FO l FUEL OIL SYSTEM l l 16 l GG l GENERATOR GAS l l 17 l HV l HYDROGEN VENTING SYSTEM l l 18 l LO l LUBRICATING OIL SYSTEM l l 19 l MC l MAIN COOLANT SYSTEM s l g l 2C l MS l MAIN STEAM SYSTEM l g l 21 l NS l NITROGEN SYSTEM l l 22 l NRV l NONRETURN VALVE SYSTEM l l 23 l PW l POTABLE WATER SYSTEM l l 24 l PR l PRESSURE CONTROL AND RELIEF SYSTEM l l 25 l PU l PURIFICATION SYSTEM l l 2f l WD l RADIOACTIVE WASTE DISPOSAL SYSTEM l l N l l 27 l R: 28 l SES l l REFUELING SYSTEM SAFE SHUTDOWN SYSTEM l l l 29 l SI SAFETY INJECTION SYSTEM l l 3C l 52 l SAMPLE SYSTEM l h~ l 31 l Sa l SERVICE WATER SYSTEM l l 32 l SC l SHUTDOWN COOLING SYSTEM l
., l 33 l SF l SPENT FUEL SYSTEM l l 34 l HC1 l HEATING STM/ CONDENSATE l
{* TURBINE GENERATOR SYSTEM l l 35 l TG l l 36 l VD l VENT AND DRAIN SYSTEM l g l 37 l AD l AIR DISPOSAL (FILTERED EXHAUST) SYSTEM l 4 l 38 l HC10 l BATTERY ROOH NO3 VENTILATION SYSTEM l l 39 l H10A l BATTERY RCOMS 1+2 VENTILATION SYSTEM l l 40 j HC5 l CONTROL ROOM VENTILATION SYSTEM l 41 l HC11 EDG BUILDING VENTILATION SYSTEM l 1 f l l 42 l HCT l l NRV ENCLOSURE VENTILATION SYSTEM l l 43 l HC12 l SAFETY INJECTION BLDG VENTILATION SYS l TURBINE BLDG VENTILATION SYSTEM l l 44 l HC4 l l 45 l VCHCR l VC HEATING, COOLING, AIR RECIRCULATION l DC DISTRIBUTION SYSTEM l l 46 l DC l g 47 ' DG l DIESEL GENERATOR SYSTEM l ._ l EMERGENCY POWER SYSTEM l J l 45 l EEPS l GENERATION AND TRANSMISSION SYSTEM l l 49 l EGTS l STATION SERVICE SYSTEM l 50 l ESSS l g l ::::===l========l::::::===============================================l
08/25/89 SYSTEME AND STRUCTURES PAGE ' O. 2 10:55:01 MEETING STEP 1A CRITERIA FOR FURTHER REVIEW
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====================l
- l. l::::::==l:::::===l=================================
l 51 l FIDS l FIXED INCORE DETECTION SYSTEM l LEAK MONITORING SYSTEM l l 52 l LM l MOVABLE INCORE DETECTION SYSTEM l l 53 l MIDS l-l f l 54 l NI l NUCLEAR INSTRUMENTATION SYSTEM PROCESS RADIATION MONITORING SYSTEM l l 55 l PRM l RADIATION MONITORING SYSTEM l l 56 l RM l REACTOR CONTROL SYSTEM l l 57 l RCS l l 58 l RPS l REACTOR PROTECTION SYSTEM i l 59 l SS l SECURITY SYSTEM l 60 l VC l VAPOR CONTAINER MONITORING SYSTEM l l 61 l EN ENVIRONMENTAL SYSTEM l l l l 62 l B3B l l BATTERY ROOH NO. 3 BUILDING l l 63 l CB l COMPACTOR BUILDING l l 64 l DGE l DIESEL GENERATOR BUILDING l EQUIPHENT AND PIPE SUPPORTS l l 65 l EPS l l 66 l T ANA l FIELD FABRICATED TANKS l FIRE WATER PUMPHOUSE l l l 67 l FW: l g l ES l FT CS l FUEL TRANSFER CHUTE STRUCTURE l l l 69 l IEP l ION EXCHANGE PIT l 70 l LhE l LIFTING AND HOISTING ED'ITPMENT l METEOROLOGICAL TOWER l l 71 l FT l l 72 l N4 l NEW FUEL VAULT l PLANT MODULAR OFFICE BUILDING l l 73 l PMDE l PRIMARY AUXILIARY BUILDING l l3 l l 74-l PAB 75 l PVS l l PRIMARY VENT STACK l 76 l RSS l REACTOR SUPPORT STRUCTURE l l g l 77 ; SLEE l S.L.E. DIESEL GENERATOR BUILDING l l l l l 78- l SSSE 79 l SH l l SAFE SHUTOOWN SYSTEM BUILDING SCREENWELL HOUSE l SECONDARY PLANT VENT STACK l l 80 l SPVS l E1 l SE l SERVICE BUILDING l l f l 82 l SFF l SPENT FUEL PIT l STEAM AND FEEDWATER SUPPORT STRUCTURE l l 83 l SFSS l STORES WAREHOUSE l l l 84 l ST-W l
- 4. 85 l SEIE l SUPPORTS FOR ELECT /I+C EQUIPMENT l l
SWITCH YARD STRUCTURE l l 86 l SYS l l 87 l TAS l TRANSFORMER AREA STRUCTURE ! l 88 l TB l TURBINE BUILDING l l' VAPOR CONTAINER l l 89 l VC l l 90 l VEL l VC ELEVATOR ENCLOSURE l
; 91 l WDB l WASTE DISPOSAL BUILDING l
- 1. l 92 l YACS l YARD AREA CRANE SUPPORT STRUCTURE l l 93 l l FOUNDATIONS FOR OUTOOOR TANKS &ST ACKS l l:=======l========l=====================================================l
08/25/09 SYSTEMS / STRUCTURES PAGE NO. 1 10:5E:05 NOT MEETING STEP 1A CRITERIA FOR FURTHER REVIEW l::::::::l::::::::l::::::::::::::::===================================== l RECORO l SYS1EM l SYSTEM / STRUCTURE l NUMBER l CODE l TITLE l l l::::::==l :::::==l:::::::::=======================:.::==================l l 1 l EA l BREATHING AIR SYSTEM l CHEMICAL FEED SYSTEM l l 2 l CF l j l 3: CO l CORROSION CONTROL SYSTEM l
. l 4 l FD l FLOOR DRAINAGE SYSTEM l l 5lSO l GENERATC:1 SEAL OIL SYSTEM l l 6 l IG l INERT GAS SYSTEM l SANITARY DISPOSAL SYSTEM l I l l
T l SDS 8 l SP l l SHIELD TANK CAVITY PURIFICATION SYSTEM l l 9 l SC l SG BLDN I&C SYSTEM l l 10 l TW l WATER TREATHENT SYSTEM l l 11 l HC9 l 3!4!Ni$TRATION BLOG VENTILATION SYS l l 12 l HC6 l GJS 'STDRAGE ROOM VENTILATION SYSTEM l US NON-FILTERED VENTILATION SYSTEM l l l 1? l HC14 14 l HC13 l l 60RECNWELL PUMPHOU3E VENTILATION SYSTEM l l 3 l 15 l HCB l SERVICE BLDG VENTILATION SYSTEM l l if l SPDS l SAFETY PARAMETER DISPLAY SYSTEM l 17 l TSC TECHNICAL SUPPORT CENTER SYSTEM l
'f l l 1E l AE l
l ADMINISTRATION BUILDIN3 l
} 19 l CLP l CAMERA AND LIGHTING POLES l 7
l l l 20 l DE_R 21 l FAG l DECONTAMINATION ROOMS FENCE AND GATES l l 22 l G4 l GAT EHOUSE l l 23 l HSVS l HEATING BOILER VENT STACK l 24 l hEUE l NON ESSENTIAL UPS BUILDING l l l l 25 l PCA1 l PCA STORAGE BUILDING NO.1 l l 26 l PCA2 l PCA STORAGE BUILDING NO. 2 l l 27 l FCAA l PCA WAREHOUSE l 25 l PE l POLE BARN l l l 29 l S l SEAL PIT l g l 30 : TAE l TRAINING AREA BUILDINGS l P l 31 l FOTe l FUEL OIL TRANSFER PUMP HOUSE l l::::::::l:::::::=l========================== ::::::::::::::::==========l
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+
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+ ; Step 1b: Is degradation of the system or structure significant to plant safety?
Draft Industry i Criteria: Age-related degradation of a system or structure will be considered significant to plant safety g unless failure of the system or structure would not
] significantly contribute to increased radiological health and safety risk to the public. The degradation of a system or structure is considered significant unless:
- 1. a. The system's or structure's f ailure could not directly result in dose consequences l exceeding FSAR or other plant specific I
off-site dose limits, 1 -
- b. The system's or structure's f ailure could not result in reactor coolant pressure boundary or primary containment leakage in excess of technical specification limits,
- c. The system or structure is not otherwise required to control:
- 1) reactor criticality;
- 2) reactor coolant system integrity, inventory, or heat removal; and
- 3) containment integrity or heat removal, i
OR Although the system or structure may be required for these functions, the system's or structure's failure is detectable in a time frame which would allow shutdown i prior to requiring a manual or automatic plant trip, f
.QB i '
- 2. A plant uniaue risk assessment, if available and used, demonstrates that: '
9' *# 3 a. The system's or structure's failure does not occur in a sequence that has a core damage f requency greater than or equal to
4 :- 4 'i 1 x10-e per yeer or in a sequence that contributes 5% or more to the total core
- j. damage frequency,
.I bB
- b. When the system or structure f ailure rate is assumed to increase due to age-related degradation, the total core damage
. frequency will not increase by more than a f actor of 3 or will not exceed.1x10*d per year.
I i I 1 1 i l I l
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T L Y U O TE I S T )B RR T G E OE TN T1 T N P N NM S LI NT A E I I TLA T M C T I E R N E WNL I I FS I US T T ME O C NR C E CI N( UM A UR P RI L I TL OS GW I T ST NA OFM C I H SE I N CE E E R A / R ST T N EV E OL M U YG SM P TC ETSN C' R AED ETE S UTD I I T LSE Y S RNE EYC TAE AY RSX TSLC N R OX N DT EF VOT E A O OE O I AE I O L CE T A C RFA TE CR N A US OOEI M NGA G OLE CT A K SMA ES DAO I AFD I NY E I AS N L D R - M-I
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SYSTEME AND STRUCTUTES PAGE NO. 1 OL/25/69
. 10:59:25 MEETING STEP 1B CRITERIA FOR FURTHER REVIEW l::::::::l::=====:l::u==================================================l
' ,l l RECORD l SYSTEM l SYSTEM / STRUCTURE l l NUMBER l CODE l TITLE 1
========================l }
l========l========l::::::======================= l 1 l AS l /UXILIARY STEAM SYS TEM l BOIL ER FEED SYSTEM l l 2 l BF l 3 l CH l CHARGING AND VOLUME CONTROL SYSTEM l l l l 4 l CS l CHEMICAL SHUTDOWN SYSTEM COMPONENT COOLING WATER SYSTEM l l 5 l CC l l l '6 l CA l COMPRESSED AIR SYSTEM l 7 l CI l CONTAINMENT ISOLATION SYSTEM l 8 l DW l DEMINERALIZED WATER SYSTEM l l 9 l EBF l EMERGENCY BOILER FEED SYSTEM l l FIRE DETECTION AND SUPPRESSION SYSTEM l l 10 l FS l l 11 l FO l FUEL OIL SYSTEM l HYDROGEN VENTJNG SYSTEM l l 12 l HV l l l 13 l H: l MAIN COOLANT SYSTEM MAIN STEAM SYSTEM l l 14 l MS l NITROGEN SYSTEM l
- l. 15 l NS l 16 l h;V l NONRETURN VALVE SYSTEM l l
PRESSURE CONTROL AND RELIEF SYSTEM l g l 17 l PR , PURIFICATION SYSTEM l 3 l 15 l PU l l 19 l WD l RADIOACTIVE WASTE DISPOSAL SYSTEM l REFUELING SYSTEM l l 20 l RF l SAFE SHUTDOWN SYSTEM l l 21 l SSS l f' l 22 l SI l SAFETY INJECTION SYSTEM l SAMPLE SYSTEM l l 23 l SA l
) l 24 l S* l SERVICE WATER SYSTEM l SHUTDOWN COOLING SYSTEM l 3 l 25 l SC l l 26 l S: l SPENT FUEL SYSTEM l 27 l VD l VENT AND DRAIN SYSTEM l l
' l l 25 l AD AIR DISPOSAL (FILTERED EXHAUST) SYSTEM 29 l H;5 l CONTROL ROOM VENTILATION SYSTEM l l
l 30 l h: 11 l EDG BUILDIEG VENTILATION SYSTEM l l 31 l HC7 l NRV ENCLOSURE VENTILATION SYSTEM l l 32 l HC12 l SAFETY INJECTION BLDG VENTILATION SYS l l 33 l VCHCR l VC HEATING, COOLING, AIR RECIRCULATION l DC DISTRIBUTION SYSTEM l
. l 34 l DC l DIESEL GENERATOR SYSTEM l l l 35 l DG !
36 l EEPS l EMERGENCY POWER SYSTEM l l STATION SERVICE SYSTEM l l 37 l ESSS l NUCLEAR INSTRUMENTATION SYSTEM l l 36 l NI l PROCESS RADIATION MONITORING SYSTEM l l 39 l PRM l j RADIATION MONITORING SYSTEM l 40 l RM l REACTOR CONTROL SYSTEM l l 41 l RCS l REACTOR PROTECTION SYSTEM l l 42 l RPS l VAPOR CONTAINER MONITORING SYSTEM l l 43 l VC l BATTERY ROOM NO. 3 BUILDING l 3 l 44 l B3B l DIESEL GENERATOR BUILDING l l l l 45 l DGB 46 l EPS l l EQUIPMENT AND PIPE SUPPORTS l FIELD FABRICATED TANKS l l 47 l TANK l FIRE WATER PUMPHOUSE l l 48 l FWP l FUEL TRANSFER CHUTE STRCTURE l l 49 l FTCS l 50 l IEP l ION EXC;4ANGE PIT l . l l:::::::=l:::::===l::::=======:::::::::::::::::::::::::::::=============l {
o . 1 08/25/89- SYSTEMS AND STRUCTURES PAGE NO. 2 10:59:25 HEETING STEP 1B CRITERIA FOR FURTHER REVIEW 4
-.j l========l:::=====l:::====================================
l RECORD l SYSTEM l SYSTEH/ STRUCTURE
. l NUMBER l CODE l TITLE l J.j
! l = = 51 l = l=LHE
= = = l= l :::::::: LIFTING l :::::::::::::::= = =EQUIPMENT AND HOISTING ===============
52 l NFV l NEW FUEL val 4T l l l 53 l PAB l PRIMARY AUXILIARY BUILDING l
=l 54 l PVS PRIMARY VENT' STACK l
;. l l l
l 55 l RSS l REACTOR SUPPORT STRUCTURE 55 l SSSE l SAFE SHUTDOWN SYSTEM BUILDING l l l 57 l SH l SCREENWELL HOUSE l 58 l SPVS l SECONDARY PLANT VENT STACK l l l '59 l SFP l SPENT FUEL PIT l 60 l SFSS STEAM AND FEEDWATER SUPPORT STRUCTURE l l l l 61 l SEIE l l SUPPORTS FOR ELECT /I+C EQUIPMENT l 3 l 62 l TAS l TRANSFORMER AREA STR(ICTURE l l 63 l TB l TURBINE BUILDING l l 64 l VC l VAPOR CONTAINER l l 65 l VEL l VC ELEVATOR ENCLOSURE l l E5 l WDB l ' WASTE DISPOSAL BUILDING l l E7 l YACS l YARD AREA CRANE SUPPORT STRUCTURE l l 65 l .l FOUNDATIONS FOR OUTDOOR. TANKS & STACKS l l========l:::::= :;:::::::: ============================================l 1 1 I i <
]
I i 1 2
h '- 08/25/89 SYSTEMS AND STRUCTURES PAGE NO. 1 11:04:10 NOT MEETING STEP 1B CRITERIA FOR FURTMFP REVIEW l ::::=====l=====:e:.l :===================================================l. ' D 1 RECORD l SYSTEM j SYSTEM / STRUCTURE l-l NUMBER l CODE l TITLE 1
? :::======l::::::::::::::::::::::===============c ::::::::::=============l } l 1 l AR j AIR REMOVAL SYSTEM i l 2 l CW l CIRCULATING WATER SYSTEM l l 3 l ES l EXTRACTION STEAM SYSTEM l' l 4 l FW l FEEDWATER CONTROL SYSTEM (I&C) l 5
l 5 l GG l r"NERATOR GAS l l 6 l LO l ./BRICATING OIL SYSTEM l l T l PW l POTABLE WATER SYSTEM l l 8 l HC1 l HEATING STM/ CONDENSATE l l 9 l TG l TURBINE GENERATOR SYSTEM l l 10 l HC10 ; BATTERY ROOM NO3 VENTILATION SYSTEM l l 11 l H10A l BATTERY ROOMS 1+2 VENTILATION SYSTEM i
! 12 l HC4 l TURBINE BLDG VENTILATION SYSTEM l l 13 l EGTS 1 GENERATION AND TRANSMISSION SYSTEM l l 14 l FIDS l FIXED INCORE DETECTION SYSTEM l l 15 l LM l LEAK MONITORING SYSTEM _ l l 16 l HIDS l MDVABLE INCORE DETECTION SYSTEM l l 17 l SS l SECURITY SYSTEM l )
l 1E l EN l ENVIRONMENTAL SYSTEM l l 19 l CE l COMPAC10R BUILDING l l 20 l Mi l METEOROLOGICAL TOWER l 21 l PMDB PLANT MODULAR OFFICE BUILDINu
~1 l 22 l SLEB l
l S.L.E. DIESEL GENERATOR BUILDING l l l 23 l SB l SERVICE BUILDING l l 24 l ST-W l STORES WAREHOUSE l b l 25 l SYS l SWITCH YARD STRUCTURE l l::::::::l==:::::=l::::::::::::::::::::::::=============================l l-I I b - _ _ _ _ _ _ - _ _ _ _ _
L EA VIR m. - E E L T I - TR - NC E NW - OE PI h. M E V OR
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$ Step 2a y Step 2.Detalled Review of t Is the Component important Systems for License Renewal j ls ! E 5 to System Safety No E E Function?
E E t Yes
.I $
Step 2b y
~ "
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?
j is the Component yes "fComponent Does Not ! 3 Subject to a Require Further License E Recognized Effect.ive " $ g Renewal Review Repair, Rep!acement, or ( ; e j g g inspection Program? Step 2J r , s t t No Options Available to $ I Step 2e Address Potentially E g u f Is the Cs .:ponent Significant Age.Related ?
?
Subjee,to No Degradation j { ' 3 Age.Related - Improved Assessment ! u
? Degradt. tion? .
Preventive Actions E Mitigative Actions Yes s db t I t 1 l t t i J t t E E R
& t t
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- - -_ _ - - _- _ ___ a
( k Step 2a: Is the component important to system / structure safety function? Draft Industry Criteria: The component is important to system / structure , ~ safety function unless:
=i 1. The component is normally isolated and does not
! perform an' accident mitigating function.
2
- 2. a. Component failure would not result in either the failure of any individual train within the system or the failure of the entire system to perform its required safety function.
AND
- b. Component failure would not reduce the structural support of any other component I such that it would not perform its system safety function.
AND
- c. Component failure would not physically g
3-damage any other component such that it would not perform its system safety function. AND
- d. Component failure would not reduce the structural capacity of a building or structure beyond the point where other components necessary to system safety function could be jeopardized.
J e M J 3. For components within the scope of a plant-unique risk assessment, if available and used, the component is not credited in the risk assessment models. l I _ __
) ;l ' ; . a 1
G S Y DIT N N T )A S I EA A T E 2 T TR T:Y F P AIG OT OT E I N _ A E LT OIM NM NRR N A S T S DE D GO OG I A LT L E FPN M S Y U S U T OMI E LM L OS Y OIN T OLD T (W A O R W E WLR CE ERA OTH I S E M F ER Y V RR RO R U PNS U T PA T I S OE UM E NP L LC UC " A R AU SIFAN I I T R ST FO N I O F REIN PL E N TN TE T TH G N NN N STIS E " N OI ES N EO EEE D N O T NEO N NCCTA O P C O OI OTA OUUCR I - PD T C P E PDD AP G T A M N MDN M F MEE E C I O U ONU OE ORR M D I L P C F CAF CD C-
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4 Step 2b: Is the component subject to an established effective replacement, refurbishment, or inspection program? Draft , 7 1 Industry Criteria: A component is considered to be subject to an
- j. established effective replacement, refurbishment, j or inspection program if: l
; 1. The program is documented,- approved, and !
j routinely implemented in accordance with plant i administrative procedures, f M
- 2. The program procedures ensure that all of the component's significant safety functions are properly addressed, M
j
- 3. The program establishes specific criteria for determining the need for corrective action and requires such action to be taken if these ,
; criteria are not met.
1 lI l I l l-3 l i 1 j m. l 4
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-A A I
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= = = =
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Step 2c: 'Is the component subject to potentially significant age-related degradation?- Draft Industry Criteria: The component will be considered subject to i potentially significant age-related degradation f unless:
- 1. It is established and documented that potenti ally si gni fi cant age-rel ated degradation
<j
~
will riot occur during the license renewal period.
.2_
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- 2. A plant unique risk assessment. if available and used, demonstrates that:
i a. When the component failure rated is assumed to increase due to age-related j degradation, the total core damage i frecuency will not increase by more than a f actor of 3 or will not exceed 1x10*' per year. j 2 and
- b. When an age-related common cause failure I mechanism is postulated that may cause multiple components to fail (among those which have satisfied the above criterion),
l those components meet criterion 2a when their combined f ailures are considered as a single event. e d l J. __e.__ . . . _ . _ _ . . _ _ _ . _ _ . . . _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ . _ _ . _ _ _ . . . _ _ _ _ _ _ _ _ _ _ _
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- OBJECTIVES OF DISCUSSION ARE TO: ;
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- DESCRIBE WHAT YANKEE CURRENTLY INTENDS TO INCLUDE IN THE APPLICATION l
IDENTIFY & DISCUSS NRC CONCERNS (IF ANY) 1
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VIEWS EXPRESSED REPRESENT YANKEE'S i CURRENT THINKING i
- WE'RE WORKING WITH EPRI, DOE, NUMARC AND l NSP TO DEVELOP PROPOSED INDUSTRY GUIDELINES
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YANKEE'S THINKING AND PROPOSED INDUSTRY GUIDELINES ARE EVOLVING
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- INTRODUCTION GENERAL INFORMATION CONCERNING APPLICATION ,
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DESCRI-PTION OF: PLANT DESIGN, CURRENT LICENSING , BASIS (CLB) AND OPERATING HISTORY
- SAFETY AN ALYSIS ANALYSIS AND ACTION PLAN TO ENSURE CONTINUED SAFETY OF PLANT OPERATION THROUGH LICENSE RENEWAL PERIOD i
- ENVIRONMENTAL AN ALYSIS i
ANALYSIS TO ENSURE CONTINUED ACCEPTABILITY OF ENVIRONMENTAL EFFECTS OF PLANT OPERATION THROUGH LICENSE RENEWAL PERIOD
SUMMARY
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SUMMARY
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- APPENDICES 4-ADDITIONAL INFORMATION SUPPORTING PLANT PROFILE, ;
i SAFETY ANALYSIS AND ENVIRONMENTAL ANALYSIS l I
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- REFERENCES !
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3 i j [ LIST OF DOCUMENTS REFERENCED IN APPLICATION { -.i. +
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3 I ENVIRONVIENTAL ANALYSIS
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DESCRIPTION OF METHODOLOGY
SUMMARY
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LIST OF TIME DEPENDENCIES OF SIGNIFICANCE
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- EVALUATION OF ENVIRONMENTAL EFFECTS
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4 - DESCRIPTION OF METHODOLOGY
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- FIXED SCHEDULES.
'I- i
- LIMITS OF DISCOVERY ll
- .
- RESTRICTIONS ON ADMISSIBILITY i j I
- Y
. E J
q CURRENT LICENSING
- CONTINUATION OF CURRENT '
i l BASIS LEVEL OF SAFETY i '
- BACKFIT CONTROLS
. . l.
G R A ND FATH ERING
- INTEGRAL TO CURRENT
. LICENSING BASIS ^
t
- NOT SUBJECT TO REREVIEW ,
! l EXEMPTIONS
- TIME DEPENDENCY ONLY :
?
i SCOPE OF REVIEW
- AGING EFFECTS ONLY .
I _, f i I, ' ENVIRONMENTAL REPORTS
- EFFECTS OF CONTINUED j' OPER ATION 'ONLY
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I. YANKEE POSITION ON GENERIC ENVIRONMENT
. 1:
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i . NEPA - LICENSE RENEWAL l 1; Y f'
- GE A - REQUIRED TO DETERMINE IF RULEM AKING IMPOSES SIGNIFICANT IMPACT ON THE ENVIRONMENT
- MAINTENANCE OF CURRENT LEVEL OF SAFETY AND ENVIRONMENTAL PROTECTION
- RULEMAKING : EFFECTS OF
[ CONTINUED OPERATION
- EFFECTS ARE SMALLt
.1 ONLY - SUPPORTED BY NUMARC STUDY GEA OF GENERIC ENVIRONMENTAL REQUIRED :
- H LW - U P DAT E OF WAST E CONFIDENCE PROCEEDINGS 2
- LLW - LLWPA M ANDATED DISPOSAL
, RESPONSIBILITY TO STATES
- SEVERE ACCIDENTS - SHOULD NOT
> BE ADDRESSED IN CONTEXT OF
' LICENSING RENEWAL RULE !
o Y j NO GEIS NECESSARY!!! , 5 . 1
L l l j i I YANKEE POSITION ON SEVERE ACCIDENT RESOLUTION . 1 & 1 SEVERE ACCIDENTS LICENSE RENEWAL
'!- ' SHOULD BE i
, *NEEDS TO BE R ESOLVED l
' R ES O LVE D SEPARATELY
,
- SEVERE ACCIDENTS
[ IS A GENERIC ISSUE NOT UNLIKE OTHER GENERIC ISSUES -
;. THEREFORE,SHOULD
*NO LOGICAL REASON l BE RESOLVED IN THE TO TIE TOGETHER ;
SAME MANNER AS IN A RULE s OTHCR GENERIC I[ I ISSUES - SEPAR ATELY *NRC ALREADY H AS ! THE POWER TO I j CONDITION LICENSE RENEWAL ISSU ANCE TO ENSURE SEVERE ACCIDENT RESOLUTION -
- NO NE ED TO
- YANKEE'S SCHEDULE CONDITION THE FOR LICENSE RENEWAL LICENSE RENEWAL HAPPENS TO BE RULE COINCIDENT WITH ITS SCHEDULE FOR SEVERE ACCIDENT RESOLUTION ,
- NOT TH AT WAY F O R !
MAJORITY OF PLANTS I
- SEVERE ACCIDENT l RESOLUTION LIKELY TO OCCUR LONG BEFORE LICENSE RENEmL FOR MAJORITY OF PLANTS -
THEREFORE,NRC SHOULD WRITE A j RENEML RULE FOR ALL i 4 l PLANTS - A RULE i DEDICATED SOLELY TO , l i LICENSE RENEWAL l c i l l l .,- J
?
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- e d a l l
u s n e r R i e R a n t i e n A s d l I it e u o s R e s o e i eN p l k s k e-a n n - o n i r i a h a-P F Y T Y-t !
t - e g r a - d t d t e nr s . e ce a n ec h e f n i f e f a oc l a h d c c - slu n g i i nh mc f u ion e hc w n b s e - a bu t s. i n ies f o o r l oc - b c e t n o p e o o rS t n e d h r a : s e e mp k e n s n o a r p
- p a i t
e sy P t s b i Y a - e g n n j o t et nvl in a aiu i r o t e ct e pp e l elu i e i a s e t r ea l q f xm keR t b r k d e e a ct f e o - e c np o i d n c r a n m ua n c o n e o o - ni n u n o e s i n t n t g at o i a i t er et n e uCi n Yi r i l l e i n h ue t t t p oRd i ue o cNn a W i t M i f e o FR r f t c d s t ys 8 i ob n s f e- - s n l a
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I-! YAEC MEETING WITH NRC ON j RESOLUTION OF SEVERE ACCIDENT POLICY ISSUES l 1
. August 29,1989
- I 1
I l Presented to USNRC Presented by b JOHN M. ODDO Safety Assessment Group 1 j C _ -__ = -_-______ - ---_ _-__ - -__ _____ - - - _ _
8 4 4 3 4 YNPS SEVERE ACCIDENT l CLOSURE SUBMITTAL TO NRC i 1 1 1 1 l
1 1 Severe Accident Policy (SAP) Closure Elements
.i 1. Individual plant examination (IPE).
- 2. Individual plant examination external events (IPEEE).
- 3. Containment performance improvements (CPI).
l
- 4. Accident management (AM).
l i i
s. i
SUMMARY
OF PRESENTATION: v j o Review highlights of previous YNPS/NRC meeting. 3 YNPS plan to deliver Severe Accident closure submittal to NRC this year l YNPS bases for closure I o The Severe Accident Closure Submittal l outline 'l l h i l
1 a s
; PRIMARY OBJECTIVE q l
l Demonstrate that YNPS and NRC already have the l information needed through YNPS past PRAs and other l evaluations to allow us to move on with the associated identification of outliers and evaluation of cost justified upgrades ; I (hardware, procedures, training, etc.) i I I I I . 15' i 'l i I I Yme M
1, l 1 g METHOD l
; l t ;
?
a Provide information on what has been done, including analyses and plant changes as well as commitments in an integrated single i submittal for IPE, IPEEE, CPI, and AM.
'. IPE = Individual Plant Examination CPI = Containment Perfonnance Improvements
-l- IPEEE = Individual Plant Examination External Events AM = Accident Management I-g l
1 GL l I a
)
O t 5 lE l
f 2 L
. BASES FOR ACHIEVEMENT OF OBJECTIVE YNPS currently has sufficient infonnation as summarized in the
. subsequent overheads to meet the intent of the Severe Accident Policy elements.
.l o YNPS and NRC have done much l o YNPS has made many changes l o YNPS continuously monitors plant characteristics o YNPS is doing much now! l o YNPS and NRC have been through SEP! l YNPS can finalize improvements to Rowe's beyond design basis capabilities rapidly without waiting for further prescription and
! without excessive additional demands on NRC or YNPS resources y
and time.
' l h_ .
I I 1 (w.A i
1
-YAEC PROGRAM HISTORY l-Internal Maior Activity Period Staff Level
!l- Yankee Mini PRA 1980 1
)
1+ ! Yankee Internal Events LevelIII PRA 1981-1983 2 Systematic Evaluation Program Use 1982-Present 3 1 Seabrook Internal and External Events 1982-1984 5 LevelIII PRA j ' Yankee Tornado /High Wind Resolution 1984-1985 '5 Vermont Yankee 60-Day Containment 1986 10 Evaluation Maine, Yankee PRA/ Reliability Program 1987 13 b Started b Plant Life Extension Program Started 1987 13 Vermont Yankee Alternate Testing 1988 14 1 Maine Yankee Phase IInternal Events 1988 14
! PRA Complete a Vermont Yankee PRA Program 1989 15 i
Expected to Start b l me ___ _ .__.-a____ __,______J' - __
a , i SAMPLE OF USES AT YAEC 1 Establish risk-based design basis for external events j Containment isolation active / passive requirements Electrical penetration environmental requirements Main Coolant Pump -loss of flow performance criteria Tornado /High Wind design changes o Seismic design changes
, Battery room ventilation 1 Manual / auto control requirements for electrical buses - Reactor Vessel Water Level System requirements Safety Injection System relief valve changes Safety Injection Building ventilation requirements Turbine / generator trip control power changes
'g Emergency procedure validation Safety Parameter Display System validation g Technical Specification allowed outage time changes Technical Specification testing frequency changes g . Technical Specification alternate testing requirement changes Plant life extension - [ System reliability program development (diesels, cooling water, emergency feedwater, shutdown cooling) 'l Shutdown events evaluation Emergency planning zone evaluation ll Evaluation of containment effectiveness Severe accident policy resolution ATWS requirements exemption 7 Operator training i Safety System Functional Inspection <
; Simulator-Design -
8
- - _L
I f I [ I l
- CURRENT USES j
; OF PRA KNOWLEDGE I 1
I ; EOPS I Design Changes i Operational Reviews l Backfit Determination PRA Procedures Established 4 1 i
c , . A DECADE OF ROWE PRA STUDIES l t-t Mini PRA I 1980
.l Mini Seismic 1981 I YNPS PSS 1981-1983 SEP Use 1982 -1988 j Plant Changes 1983 -
l Tornado /High Wind 1984-1985 i Seismic 1984-1985 1 SPDS 1985 4' Containment 1987 - [. PLEX 1987 - l Simulator 1989
)
I m
.A )
- - _ __ o
,w . l
SUMMARY
~
2 i
; o YAEC has the information required to reach closure on IPE, CPI, IPEEE, and AM. (We I have been performing Safety / Risk Management proactively, even before PRA studies were initiated!)
i Continued use has not changed original g conclusions , I o We do not need to reanalyze or re-report i We are prepared to complete the information base. ik s
' 1, [ SCHEDULE: YNPS Submittal by end of 1989
'l NRC Review and response - 1990 I
i l' I I i 5 h 1 .g. sez i
-_. .___m_m.m_-_..___._mm___-___m_.-
i *t 'l l THE SEVERE ACCIDENT n CLOSURE SUBMITTAL OUTLINE c
+
1 i g i 1 1 l t v _ ,.
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41 YNPS Severe Accident Closure Snhmittak
, [: S"--rv Outline
?
I i' Section Iille
-{.
1.0 Introduction / Executive Summary 2.0- Severe. Accident Policy Elements - Summary Description
- l. 3.0 . Individual Plant Examinations (IPE) - Summary for YNPS 4.0 Individual Plant Examination External Events (IPEEE) - Summary for
-YNPS 5.0 Containment Performance-Improvements (CPI)'- Summary for YNPS 6.0 Accident Management Plan (AM) - Summary for YNPS 7.0 Summary of; Basis end Request for Closure for YNPS ~[ Addendum Title
- 1. Basis / Justification for YNPS/ SAP Submittal Approach I
I I l a t i i I 7,,3a 1
'l
k.. . :. ; i
.- 1 s
i .- .YNPS Severe Accident Closure Submittal 1: ~ .
' - ' Detailed Outline
[.-k l . .
1.0 INTRODUCTION
/ EXECUTIVE
SUMMARY
1
.1.1' . Background and Summary of Severe Accident Policy Elements l
Eo I'" 1.21 Purpose.Of Submittal 1.3 Summary of YNPS Severe Accident Policy Submittal.- 11.4 ' Request for NRC' Review and Determination of Closure of Severe Accident Policy Issues for YNPS
. 2.0 - SEVERE _ ACCIDENT POLICY ELEMENTS .
SUMMARY
DESCRIPTION
' 2.1 Individual Plant Examination (IPE) - Description
- 2. 2 .. Individual. Plant' Examination External Events (IPEEE) - Description 2.3 Containment Performance Improvements (CPI) - Description 2.4 . Accident Management (AM) -' Description r
q. i 4
.j l
7993R' p
.f'vfi:.
;p . 3.0 ~ INDIVIDUAL PLANT EXAMINATION (IPE)'-
SUMMARY
'FOR YNPS:
.U I [,, . 3.1 ' Introduction / Summary.
j 3'. 2 - YNPS PRA' Background ,l 3.2.1 YNPS PRA History _and NRC Reviews 3.. 3.2.2- - YNPS Routine FRA Uses' ( ' -1 .J
~ 3.3 ' Summary of TNPS PSS' .
' 3.3.1! ' Approach-3.3.2'. Results
- 3. 3. 2.1' . Comparison to Safety Goals 3.3.2.2 Core Damage Frequency 3.3.2.3- Bases For Results
.3.4 SEP Internal Events Probabilistic Studies 3.4.1 Containment Isolation Active / Passive Requirements 3.4.2 Electrical Penetration Environmental Requirements
; 3.4.3 Main Coolant Pump - Loss of Flow Criteria 3.4.4 Battery Room Ventilation i 3.4.5 - Manual / Auto Control Requirements for Electrical Buses 3.4.6 Safety Injection Building Ventilation Requirements 3.4.~7 NRC Contractor " Risk-Based Categorization of Yankee SEP Issues" 3.5 Current large Scope Ef forts
' 3.6 Comparisons and Findings From all Studies o Especially Low Core Damage Frequency for Internal Events o LOCAs Dominate o Containment is Effective o From Simple to Complex Studies, Conclusions are the Same 2-7993R
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, +
j f!; r 3.7 lassons/Uses From all Studies Resulting in Additions or Changes 4 3.7.1 ECCS
..l
.t.
Relief Valves Changed i lb Check Valves-Investigated 3. Ventilation Modified Alternative Recirculation Path From Containment: Implemented Li High Pressure Header' Addition i;.
'R: 3.7.2 Redundant Turbine Trip Added
- 3. 7.' 3 . TwoMotor-DrivenkuxiliaryFeedwaterPumpsAdded
,; 3.7.4- Main Steam Line Nonreturn Valves (NRVs) Check Valves Actuated Reactor Trip on NRV Position Added
' 3.'7.5 'High' Reactor Coolant System Pressure Trip Added
-3.7.6- Operator Training 3.7.7 Emergency Operating Procedures
- g
g 3.8 Conclusion That Existing Work is Sufficient for IPE
- I 1
- l {
7993R
7 . \ i: p-- 4.0 ~ INDIVIDUAL PLANT EXAMINATIONS (IPEEE) - EXTERNAL EVENTS -
SUMMARY
FOR YNPS' g3; 4.1 Introduction / Summary
.E.
>{; <
4.2 SEP 4 i-j. 4.2.1. Summaries'of SEP. Analyses r, - 4.2.2 Results/ Conclusions.
! 4.2.3 Modifications / Actions s
- List of 21 Completed Plant Modifications Correlated With l o Their External Events 1
List of 11 Plant Modifications Scheduled for.1989-1992 Correlated With Their External Events 4.3 ' External Events Applicability Assessment.
- Comprehensive' List of Natural and Man-Made Events l.-
- External Events Potentially Applicable to YNPS
-- ' Ten YNPS External Events Assessed 4.4 Tornado Wind / Straight Wind Analysis
. 4.4.1 Summary of SEP Topics III-2 and III-4.A Analyses 4.4.2 Results/ Conclusions 4.4.3 . Summary of NRC Safety Evaluation 4.4.4 Modifications / Actions 4.4.5 Summery of 1988 Review / Update
- 1. 4.4.6 Conclusions
~;
4.5 Seismic Analyses i 4.5.1 Summary of SEP Topic III-6 Seismic Design Considerations
-f 4.5.2 Yankee Scismic Analyses '! 4.5.3 Seismicity Owners Group Seismic Hazard Analyses - 1989 4.5.4 LLNL/NRC Seismic Hazard Analyses 'i 3
7993R l l
> e , # .n wi, j . .- y 4.5.5 Results/
Conclusions:
!I '4.5.6 Summary'of NRC Evaluation t o 4.5.7 Modifications / Actions; 7
- f. :
4.6- ' Fire
.}s. .
-4.6.1'.. ' Internal Fire.
2 i 4'6.1.1
. Summaries of Current / Previous Analyses-
]. 4.6.1.2 Conclusions / Findings From Current / Previous Analyses l
4.6.1.3- Modifications Implemented or Committed
- 4. 6.1. 4~ . Continued Compliance 4.6.1.5 SAP /IPEEE Topics
. .4.6.1.6 Re-Evaluation of Section 4.6.1.1'From SAP /IPEEE Intent Perspective-4,6.1.7 Additional Evaluations 4.6.1.7.1 Walkdowns (As Appropriate)
- a. . Fire / Seismic Interactions
- 1. Directed walkdown of Fire
! Protection Suppression Systems in close proximity to the seismic margin safe shutdown path components to evaluate their
" survivability".
.i 1
g i 7993R g 5 A
s
- 2. Directed walkdown to verify y '
seismic anchoring'of vessels (e.g., hydrogen. tanks or other
,4 flammable / combustible materials-susceptible to seismic events).
1 I f .'
- b. Equipment Survival b 1. Directed walkdown to evaluate.
3 potential impact of inadvertent 1 suppression actuation on safe shutdown components. 4.6.1.7.2 Miscellaneous Evaluations
~
-t
- a. Control Systems Interaction - Provide evaluation or verification that safe shutdown circuits are physically
- independent of the Control Room.
4.6.1.8 Conclusions / Findings i 4.6.1.9 Potential' Modifications
- Procedural Modifications
- Hardware Modifications
.. g g
- Adminis:.rutive Modifications (e.g., limiting combustibles, immediate removal of combustibles.
etc.) f. l 4.6.2 Forest Fire 4.6.2.1 Summary of Evaluation
. 4.6.2.2 Results/ Conclusions
- ~7993R
. -l .
1
.y : ..
f-
. pl 4.7 -Aircraft Impact:
j: 4.7.1 Summary of SEP Topic;III-4.D Analysis
} 4. 7. 2 . Results/ Conclusions 4.7.3 Summary of NRC Evaluation-4 4'. 7. 4 Summary of 1988 Review / Update 4.8 External Flooding L
1 l- 4.8.1 Summary of External Flood-Related SEP Topics 4.8.2 Precipitation-Induced Riverine Flooding 4.8.3 Local Flooding Due to Local Intense P ecipitation j 4.8.4 .. Flooding Due to Upstream Dam Failure 4.8.5 External Flooding Summary
~
4.9 Internal Flooding
- 4. 9.1 - Summary of NRC SER'
. 4.9.2 YAEC Evaluation of IN 87-49
, . 4.9.3- Other Modifications Since 1980 4.9.4 Conclusions
'4.10 Ice Cover / Snow 1- 4.10.1 Summary of SEP Topic II-2.A Review e 4.10.2 Summary of SEP Topic III-7.B Snow Loads 4.10.3 Conclusions 4.11 Lightning
- Summary of SEP Topic II-2.A Review
]
4.12 Release of Chemicals in On-Site /Off-Site Storage 4.12.1 Results of On-Site Release Evaluation Indicate Minimal
- Consequences and No Mitigative Action Needed 7993R
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... :4.12.2 .Results.of Off-Site Release Evaluation 7
t:T, .
;- 4.12.3' Conclusions-
,;LL( t-np 4.13 Turbine Generator Missiles L. pl 4.13.1 Summary of SEP Topic JII-4.B Evaluation
- 4.13.2 Operating Experience (LP Turbine Rotor Failure)
_4.13.3 -Plant Modifications Made 4.13.4. Conclusions 4.14 IPEEE Findings and Conclusions.
-l 4.15 Modifications / Actions C 4.16. Future Actions Plar.. sed /Being Considered I
l l
.1 1
l
.I
. 7993R
_ _ . _ - . - _ _ . - _ - _ . ____ __m _m_-u__. . __m. _.m _m.-_-m___._-___ __ ______ _.__._m._. _ _ _--- --mm-_-.--__ . _ . _ _ _ . _ _ . _ _ _- -_._.
W f E: p 15.0 CONTAINMENT PERFORMANCE IMPROVEMENTS (CPI) -
SUMMARY
FOR YNPS ll11 5.1' YNPS Containment SAP Topics
5.2 Background
')
5.2.1 YNPS Mini-PRA - 1980 5.2.2 YNPS PSS - 1983 5.2.3 _ Containment Analyses - 1987 to 1989-If 5.3 . CPI Findings / Conclusions i
- Depressurization of Reactor Prior to Vessel Failure Improves Containment Performance By:
I o Reduces Potential for Core Dispersal by Direct Containment ' Heating (DCH)
- o Reduces External Loads on Reactor Vessel o Improved Cavity Integrity I Increases Retention of Shield Tank Inventory o
- Make-Up Water Would Maintain Core Coolable'and Prevent Concrete
. Attack 5.4 Modifications / Actions Identified and/or Accomplished 6
5.4.1 Reactor Cavity Plugs and Stand Pipes 5.4.2 Water Cleanup System f 5.5 Actions Planned /Being Considered 5.5.1 Primary System Depressurization System 1' 5.5.2 Severe Accident Injection System Alternatives _9_
- ~7993R I
'}
.I G s.
- High Pr:s:ure Injection Ir.to MCS 3
[.
- Low Pressure Injection Into Reactor Cavity 1
r, 5.5.3 Risk / Benefit Evaluation for Final Design Selection a i i I I I I I I I i 1 4 4 a
) 7993R
- g. ( li t '
. L' . [
?
6.' 0] ' AM - ACCIDENT MAWAGEMENT PLAN -
SUMMARY
FOR YNPS
, . 6 .1 - Yankee. Approach.to Severe Accident Managerent
'l 6.1.1 Goals of the Accident Management Plan j
Evaluate Current Capabilities
- Enhance Current Capabilities
- ' Incorporate Results of Industry Research 6.1.2 Overview of Accident Management Plan
- Summary of Accident Management'Needs Role of the IFE in the Plan
- Meeting the Needs of the Management Strategy
, o Existing Emergency Plan o- Evaluating Future Enhancements .;
- 1. Multi-Discipline Evaluation Team l i
Yankee Technical Expertise and Operational
~
.- ' 2.
l Experience 6.1.3 Accident Management Strategy.During an Event l 1 - Assess Current Plant Status
- Determine Current Needs for Prevention / Mitigation
- Assess Available Equipment and Instrumentation
- Impicment Best Course of Action Predict Postulated Future Needs
- Ongoing Process
.l 1 !
! I l 7993R f
- - - - - - - _ - _ - - _ ---.-----_-_______ _ _ __ _ _ _ _ a
g b l' . 6.2 Accident Management Procedures-i , 6.2.1 ' Symptom-Based Functional Emergency Operational Procedures
- Development of Yankee E0Ps from Westinghouse Guidelines
- Conditions Covered by Existing E0Ps
- E0P Critical Safety Functions L -
6.2.2 Critical Safety Function Prioritization for Severe Accidents 7 f - Suitability of E0P CSFs to Severe Accidents
- Assessment of CSF Priority for Severe Accidents
- Potential Changes in CSF Priority as Accident Progresses
'6.2.3 Relationship Between E0Ps and Severe Accident Procedures ti . 2. 4 Severe Accident Emergency Operating Procedures I 6.2.4.1 Severe Accident Guidance Documents
- Guidance for Operators
- Guidance for ESC /TSC Staff 6.2.4.2 Severe Accident Response Systems Procedures I - Severe Accident Injection System
- Severe Accident Depressurization System 6.2.5 Procedure Transitions I
6.2.5.1 Transition Parameters I - Assessing the Parameters to be Used to Signal Transitions To and From Severe Accident Procedures i I 7993R I
- _ ______-___ ___-_-__--_____-__ - -_____ - t
6.2.5.2 Transition S:tpoints
- Assessing the Appropriate Parameter Values for Transitions To and From Severe Accident Procedures' j 6.2.5.3 Instruments
- Assessing the Instruments Which Would be Available for Determining Procedure Transitions
- Impact of Available Instruments on Parameters and Setpoints Selected to Signal Procedure Transitions 1
6.3 YNPS/ Corporate Accident Management Training I 6.3.1 YNPS 6.3.1.1 Instructor Training by YNSD 6.3.1.2 Operator Training at YNPS I - Mitigating Core Damage Training 1 - Assessment of Future Training Needs I - Incorporation of Results From Industry Research on Severe Accidents 6.3.1.3 Severe Accident Simulator (Under Evaluation)
.I 6.3.2 Corporate
! - ESC Staff Training and Guidance Documents l
I 7993R ..n,. l .-
L- /t ~ r 6.4 Computational Aids i 6.4.1 Trending and Display Upgrades to SPDS and METPAC ; 1 SPDS Upgrades
- METPAC Upgrades 6.4.2 ESC /TSC Plant-Specific Instrumentation i
- TSC SPDS Graphic Trend Display 1 - ESC /TSC SPDS Hard Coy Terminals 6.4.3 Additional Calculational and Instrumentation Tools 6-.4.3.1 Evaluation of On-Line Analysis Tools
- MAAP Code
- PC-Based Simulator Workstation 6.4.3.2 Severe Accident Simulator (Under Evaluation) 6.5 Instrumentation Status Determination 6.5.1 Environmental Qualification Efforts 6.5.2 Symptom-Based E0P Support
. 6.6 Corporate Severe Accident Communications and Decisions Responsibility Plan 6.6.1 Technical Responsibility Designations 1 Existing Emergency Plan 6.6.1.1 I 6.6.1.2 Evaluation of Future Enhancements I
U 7993R p,.
[' _ 6.6.1.3 Ongaing Assessment of Industry Severe Accident
'l Insights 6.6.2 Decisions and Communications Plan 6.6.2 1 Existing Emergency Plan p.
6.6.2.2 Evaluation of Future Enhancements 6.6.2.3 Ongoing Assessment of Industry Severe Accident Insights 6.6.3 Provisions for Multiple Shifts During Accidents 6.7 Off-Site Support Systems 6.7.1 Mutual Assistance Agreement Programs 6.7.2 . Industry Communications and Support 6.7.3 Federal and State Communications and Support E l 1 I I I I f 7993R l
- j. . .~.
-7.0
SUMMARY
OF BASIS AND, REQUEST FOR CLOSURE'FOR YNPS 7.1- IPE Results/ Conclusions Summary-Ll
. 7.2, IPEEE Results/Concittsions Summary v 7.3 CPI Resuhs/ Conclusions Summary
.II 7.4 -AM Results/ Conclusions Summary f 7.5 Commitments for Future Action (s)-(e.g., New Systems / Procedures Analyses (if any))
'f 7.6 Continuation of Monitoring 7.7 Statement Demonstrating Compliance With Severe Accident Policy Requirements Based on Section 7.1 Through Section 7.6 Inclusive 7.8 Request for NRC Determination of Closure l ADDENDUM i 1. Basis / Justification for YNPS/ SAP Submittal Approach I
I I I r 7993R __,.
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