ML20107M581
| ML20107M581 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 02/22/1985 |
| From: | Fiedler P GENERAL PUBLIC UTILITIES CORP. |
| To: | Zwolinski J Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8503010349 | |
| Download: ML20107M581 (141) | |
Text
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l g GPU Nuclear Corporation NggIgf Post Office Box 388 Route 9 South Forked River, New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number:
February 22, 1985 Mr. John A. Zwolinski, Chief Operating Reactors Branch No. 6 Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Zwolinski:
Subject:
Oyster Creek Nuclear Generating Station Docket No. 50-219 Schedular Extensions for Environmental Qualification of Certain Electrical Components In accordance with' the provisions of 10CFR50.49(g), GPUN hereby requests schedular extensions for certain electrical components required to be environmentally qualified. These components, reasons for schedular extension request, proposed resolution and schedule for qualification are indicated in tabular fonn provided in ATTACHMENT 1 to this letter. The schedule for qualification was derived by first determining if the qualification action (e.g. replacement) can be conducted while the plant is in operation. Those components which can be replaced without requiring a plant shutdown will be replaced by November 30, 1985 provided that the delivery of the components is made in a timely fashion. GPUN is currently developing a replacement schedule for each compor.ent in this category. Some components, however, must be replaced during a plant shutdown as indicated in ATTACHMENT 1. Our New Jersey Pollutant Discharge Elimination System permit (NJPDES) prohibits a planned outage during the December through March time frame because of environmental considerations. Therefore, our next scheduled refueling outage must occur sometime after April 1,1986. GPUN is planning to replace as many components as possible during unscheduled outages provided that time required for replacement of the components is less than the estimated length of the unscheduled outages. To this end we are currently identifying the required length of time for replacement of each component in this category. If, however, a component cannot be replaced during unscheduled plant shutdowns, it will be replaced during the next refueling outage. Since it is not possible to predict the length and number of future unscheduled outages at this time, identifications of components which cannot be replaced by November 30, 1985 are not now available. It is our plan to request further extension in September,1985 if any of the components which require a plant shutdown remain unqualified at that time or a late J.ti1very of a component prevents timely installation.
8503010349 850222 %
\ \
GPU Nuclear Corporation is a subsidiary of the General Pubhc Uhlities Corporation
Also included in ATTACHMENT I is a list of components which fall within the scope of 10CFR50.49 at Oyster Creek Nuclear Generating Station. Our recent re-examination of the previously submitted (November 1,1980, March 16,1983 and March 16, 1984) environmental qualification equipment master list and your letter dated May 9,1984 has resulted in an expanded equipment master list as shown in ATTACHMENT I. Recently identified components are those without TER numbers in the table provided in ATTACHMENT I. As indicated by Note 4 of Attachment I, some components will be deleted from the equipment master list if further analysis shows tnat they are not subjected to a harsh environment for the accident they are required to mitigate.
For each of the components listed requiring schedular extension we have included a Justification for Continued Operation (JCO). In some cases, the associated JCOs were already submitted by our previous letter dated December 21, 1984. However, for the purposes of expediting your review, all associated JC0s are provided in ATTACHMENT II. We conclude that the JCOs provided are valid until the proposed resolutions are instituted. Please note that out of twenty-three JCOs transmitted by the December 21, 1984 letter, some have been deleted since the associated components are not witnin the scope of 10CFR50.49 (JC0-0C-84-6 (ID45),12 (IPO SA, B, C & D) and 22). Those JC0s in ATTACHMENT II supersede the JCOs in our submittal of December 21, 1984.
ATTACHMENT III includes a list of components which were deleted from the original equipment list submitted on November 1,1980, March 16,1983 and March 16,1984. Reasons for deletions are also provided in the report documenting the methodology utilized to generate the equipment master list and
- are available in GPUN's corporate office.
We trust that the infonnation provided by this submittal is adequate for your approval of our extension requests; however, should you have any questions, please contact M. Laggart - Manager, BWR Licensing (201-299-2341).
Very truly yours, 0
P te r
! Vice President & Director Oyster Creek PBF:YN Attachments cc: Dr. Thomas E. Murley, Administrator
( Region I l U.S. Nuclear Regulatory Commission 631 Park Avenue
! Xing of Prussia, PA 19406 l NRC Resident Inspector i Oyster Creek Nuclear Generating Station Forked River, NJ 08731 l
ATTACHME2T I OYSTER CREEK ENVIRONMENTAL QUALIFICATION STATUS COMPONENT . SCHEDULAR REASON FOR COMPONENT TER QUALIFICATION JC0 EXTENSION EXTENSION _ QUALIFICATION SCHEDULE FOR TAG NO. GENERIC NAME MANUFACTURER MQ2 STATUS NQ. REQUESTED REOUESTED METHOD DUALIFICATIDN ATTACHMENT I OYSTER CREEK ENVIRONMENTAL QUALIFICATION STATUS NOTES:
Note la: This component is expected to be delivered between March 31, 1985 and November 30, 1985.
Note ib: This component is expected to be delivered after November 30, 1985.
Note 2: Replacement of this component requires a plant shutdown. Next plant shutdown (Cycle 11 refuel outage) is scheduled to commence sometime after March 1986. However.. GPUN will attengt to replace this component during the first unscheduled outage of a sufficient length following receipt of the component fran the supplier and completion of engineering. Otherwise, the replacement of this component will be accompitshed during the Cycle 11 refuel outage.
Note 3 This component has already been qualified at other nuclear plants, and there is a high degree of confidence that qualification will be documented for Oyster Creek. Documentation is being prepared to complete the EQ file.
I Note 4 This component will be' deleted if further analysis shows that it is not subjected to a harsh environment for the accident it is required to mitigate.
Note 5 Not Used Note 6 Additional time is required to identify components in order to establish qualification status or to specify replacement parts.
Note 7 Not Used Note 8 Final qualification of the cable will be confirmed by a supplementary test to be performed by the manufacturer. This test is scheduled to be completed by July 1986.
Note 9 Qualification report is being prepared by vendor for Oyster Creek based on the BWR MCC Qualification Program. Interim letter report concludes that a large majority of the devices are capable of showing qualification. The few remaining devices can be qualified by replacement, similarity analysis, or possibly minor testing. This qualification report will be available by the end of April, 1985 Note 10 Qualification documentation is being prepared by vendor. Interim letter report concludes that there is a high degree of confidence that this component is qualified.
044 c/p21
ATTACKMENT I OYSTER CREEK ENVIRONMENTAL-QUALIFICATION STATUS-COMPONENT SCHEDULAR REASON FOR COMPONENT TER QUALIFICATIDN . JC0 EXTENSIDN EXTENSION QUALIFICATION SCHEDULE FOR TAG NO. GENERIC NAME MANUFACTURER NQ. STATUS bQ. REDUESTED REQUESTED METHOD OUALIFICATIDN Standby Gas Treatment / Reactor Buildina Ventilation (Cont'd)
V-28-14 Limit Switches (2) Namco DC-85-33 Yes Note la- Replace Before 11/30/85 V-28-15 Limit Switches (2) Namco DC-85-33 Yes Note la Replace Before 11/30/85
.V-28-16 Limit Switches (2) Namco DC-85-33 Yes Note la Replace Before 11/30/85 V-28-21 Limit Switches (2) Namco DC-85-36 Yes Note la Replace Before 11/30/85 V-28-22 Limit Switches (2) Namco DC-85-36 Yes Note la Replace Before 11/30/85 V-28-23 Limit Switch Contromatics Qualified V-28-24 Limit Switch Contromatics Qualified V-23-27 Limit Switch Contromatics Qualified V-28-28 Limit Switch Contromatics Qualified V-28-48 Limit Switch Contromatics Qualified
)
(See Page 21 for Notes.)
04482/p20
i ATTACHMENT I OYSTER CREEK ENVIRONMENTAL QUALIFICATION STATUS COMPONENT SCHEDULAR REASON FOR COMPONENT TER ' QUALIFICATION JC0 EXTENSION EXTENSION QUALIFICATION SCHEDULE FOR TAG NO. GENERIC NAME MANUFACTURER & STATUS & REQUESTED REQUESTED METHOD OUALIFICATION Conunon Items (Cont'd)
Coaxial Cable Endevco DC-85-16 Yes Note 2 Replace Next refueling outage Cable Okonite (X-01ene) Qualified Standby Gas Treatment / Reactor Buildina Ventilation System TE28-6A Temperature Element Weed Qualified TE28-6B Temperature Element Weed Qualified TE28-7A Temperatere Elenent Weed Qualified TE28-78 Temperature Element Weed Qualified U-6-578 Solenoid Valve Asco Qualified V-6-580 Solenoid Valve Asco Qualified U-20-9 Limit Switches (2) Namco DC-85-33 Yes Note la Replace Before 11/30/85 U-28-10 Limit Switches (2) Namco DC-85-33 Yes Note la Replace Before 11/30/85 9-28-11 Limit Switches (2) Namco DC-85-33 Yes Note la Replace Before 11/30/85 U-28-12 Limit Switches (2) Namco DC-85-33 Yes Note la Replace Before 11/30/85 0-28-13 Limit Switches (2) Namco DC-85-33 Yes Note la Replace Before 11/30/85 (See Page 21 for Notes.)
044&a/p19
1 ATTACNMENT I OYSTER CREEK ENVIRONMENTAL QUALIFICATION STATUS COMPONENT . SCHEDULAR REASON FOR COMPONENT TER QUALIFICATION JC0 EXTENSION EXTENSIDN QUALIFICATIDN SCHEDULE FOR TAG NO. GENERIC NAME MANUFACTURER & STATUS & REQUESTED REQUESTED METHOD OUALIFICATION Comon Items Cable Anaconda Qualified Cable Loston Insulated Qualified Wire Cable General Electric 78 Qualtfled (Power)
Cable General Electric 79 Qualified (Control)
Cable Kerite FR 82 Qualified Cable Kerite HT 82 Qualified Cable Rockbestos EP 80 DC-85-20 Yes Note 8 Test Note 8 Cable Rockbestos DC-85-20 Yes Note 8 Test Note 8 Firewall III Cable Tenso11te 81 Qualtfled Electrical ITT Cannon 76 Qualified Connectors Electrical ITT Cannon 77 Qua11fted Connectors Electrical General Electric 72 OC-85-50 Yes Note 10 Analysis Before 11/30/85 Penetration Splice Raychem WCSF-N 83 Qualtfled Splice Raychem NPKV Qualified Cable Okontte (Okozell 'Qua11fted Terminal Blocks General Electric Qualtfled Terminal Blocks Weidmuller 71 Qualified Wire Terminals AMP-PIDG, Qualifted.
Plasti-Grip Wire Terminals Thomas and Betts Qualified Terminal Block Marathon Qualified Terminal Block Buchanan Qua11fted (See Page 21 for Notes.)
04483/pl8
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ATTACMMENT I OYSTER CREEK ENVIRONMENTAL QUALIFICATION STATUS-COMPONENT SCHEDULAR REASON FOR COMPONENT TER QUALIFICATION JC0 EXTENSION' EXTENSION QUALIFICATION SCHEDULE FOR TAG NO. GENERIC MAME MANUFACTURER & STATUS & REQUESTED REQUESTED METHOD 00ALIFICATION Reactor Plant Instrummotation System (Cont'd)
RE18C Level Indicating Bar on 60 Qualified Switch RE18D Level Indicating Barton 60 . Qualified Switch TE-50-1A Temperature Element Pyco Qualified TE-57-2A Temperature Element .Hy-Cal DC-85-37 .Yes Note la Replace - Before 11/30/85-TE-50-1B Temperature Element Pyco Qualified TE-59-28 Temperature Element Hy-Cal -OC-85-37 Yes Note la Replace Before 11/30/85 TE-130-450 Temperature Element Pyco Qualifted TE-130-451 Temperature Element Pyco Qualified TE-130-453 Temperature Element Pyco Qualified TE-130-454 . Temperature Element Pyco Qualified Reactor Protection System R037C Level Switch Magnetrol 59 Replaced / Qual.
R0888 Level Switch Magnetrol 59 Replaced / Qual.
RD91A Level Switch Magnetrol 59 Replaced / Qual.
R09sD Level Switch Magnetrol 59 Re> laced / Qual.
Safety & Relief valve Monitor System MS-VE-1 Accelerometers Endevco Note 3 OC-85-16 No Analysis Before 3/31/85 Thru 21 MS-VX-1 Line Driver Uholtz-Dickie OC-85-16 Yes Note 2 Relocate Next refueling outage Thru 21 TB-NR-28A Terminal Block TRW-Cinch DC-85-16 Yes Note 2 Relocate Next refueling outage Thru N. J Thru N. P Thru R.
TB-NR-108A Thru E.
(See Page 21 for Notes.)
04480/p15
r ATTACNMENT I OYSTER CREEK ENVIRONMENTAL QUALIFICATION STATUS COMPONENT- SCNEDULAR REASON FOR CDNPONENT TER QUALIFICATION JC0 EXTENSION EXTENSION QUALIFICATIDN SCHEDULE FOR TAG NO. GENERIC NAME MANUFACTURER -& STATUS E REQUESTED REQUESTED METN0D OUALIFICATION Reactor Plant Instrumentation System (Cont'd)
RE028 Level Indicating Yarway 62 OC-85-14 Yes Notes la&2 Replace Note 2 Switch RE02C Level Indicating Yarway 62 DC-85-14 Yes Notes la&2 Replace Note 2 Switch RE03D Level Indicating Yarway 62 DC-85-14 .Ves Notes la&2 Raplace Note 2 Switch RE03A Pressure Switch Barksdale 56 Qualified RE03B Pressure Switch Barksdale 56 Qualified RE03C Pressure Switch Barksdale 56 Qualified RE03D Pressure Switch -Barksdale 56 Qualified RE05A Level Indicating Yarway _63 DC-85-22 Yes Notes 1b&2 Replace Next Refueling Dutage Switch RE05/19A Level Indicating Yarway 61 DC-85-22 Yes Notes Ib&2 Replace Next Refueling Outage Switch RE05B Level Indicating Varway 63 DC-85-22 Yes Notes Ib&2 Replace Next Refueling Outage Switch RE05/19B Level Indicating Yarway 61 OC-85-22 Yes Notes 1b&2 Replace Next Refueling Outage Switch REISA Pressure Switch Barksdale 54 Qualified RE158 Pressure Switch Barksdale 54 Qualified REISC Pressure Switch Barksdale 54 Qualified RE150 Pressure Switch Barksdale 54 Qualified RE1CA Pressure Switch Barksdale Qualified RE16B Pressure Switch Barksdale Qualified RE10A Level Indicating Barton 60 Qualified Switch RE188 Level Indicating Barton 60 Qualified Switch (See Page 21 for Notes.)
04480/p14
r '
ATTACHMENT I OYSTER CREEK ENVIRONMENTAL QUALIFICATION STATUS I
COMPONENT SCHEDULAR REASON FOR COMPONEEf TER QUALIFICATION JC0 EXTENSION EXTENSION QUALIFICATION SCHEDULE FOR TAG ND. GENERIC MAME MANUFACTURER & STATUS E REQUESTED REQUESTED METHOD OUALIFICATION Condensate Transfer System V-6-457 Solenoid Valve Asco 32 Replaced / Qual.
(Fcr V-11-34) l V-6-458 Solenoid Valve Asco 32 Replaced / Qual.
(Fcr V-11-36)
I V-11-34 Limit Switch Fisher Governor Co. DC-85-31 Yes Note la Replace before 11/30/85 V-11-36 Limit Switch Fisher Governor Co. DC-85-31 Yes Note la Replace Before 11/30/85 l EggEgency Service Water Systeg v-3-87 Control Switch General Electric Note 4 DC-85-43 Yes Note la Replace Before 11/30/85 V-3-88 Control Switch General Electric Note 4 DC-85-43 Yes Note la Replace Before 11/30/85 Reactor Plant Instrumentation System IA-90A Differential Press. Rosemount DC-85-37 Yes Notes 1b&2 Replace Note 2 Transmitter IA-908 Differential Press. Rosemount DC-85-37 Yes Notes ib&2 Replace Note 2 Transmitter IA-91A Differ'ntial Press. General Electric DC-85-37 Yes Notes Ib&2 Replace Note 2 Transmitter IA-918 Differential Press. General Electric DC-85-37 Yes Notes Ib52 Replace Note 2 Transmitter IA-92A Pressure Transmitter Rosemount OC-85-37 Yes Notes Ib&2 Replace Note 2 i IA-928 Pressure Transmitter Rosemount DC-85-37 Yes Notes Ib&2 Replace Note 2 ID13A Level Transmitter General Electric 41 DC-85-8 Yes Notes Ib&2 Replace Next Refueling Outage 1013B Level Transmitter General Electric 41 OC-85-8 Yes Notes Ib12 Replace Next Refueling Outage ID4tA Pressure Indicating GE/MAC 38 OC-85-6 Yes Notes Ib&2 Replace Next Refueling Outage Transmitter 1046B Pressure Indicating GE/MAC 38 OC-85-6 Yes Notes Ib&2 Replace Next Refueling Outage Transmitter REo2A Level Indicating Yarway 62 Qualified Switch (Set Page 21 for Notes.)
04480/pl3
ATTACMMENT I OYSTER CREEK ENVIRONMENTAL QUALIFICATION STATUS COMPONENT SCHEDULAR REASON FOR COMPONENT TER QUALIFICATION. JCO EXTENSION EXTENSION QUALIFICATION SCHEDULE FOR TAG NO. GENERIC NAfE MANUFACTURER g(L STATUS & REQUESTED RE00ESTE0 METHOD OUALIFICATION Main Steam System (Cont'd)
CSO4A Limit Switches Namco 64 Replaced / Qual.
(V-1-9)
A.B.O NSO4B Limit Swttches Namco 64 Replaced / Qual.
(V-1-10)
A.B.O V-1-106 Motor Operator Limitorque 12 OC-85-44 Yes Notes la&2 Partial Replemt. Note 2 Limit Switches V-1-107 Motor Operator Limitorque 12 OC-85-44 Yes Notes la&2 Partial Rep 1 cat. Note 2 Limit Switches V-1-110 Motor Operator Limitorque OC-85-44 Yes Note 6 Partial Rep 1 cat. Note 2 Limit Switches V-1-111 Motor Operator Limitorque OC-85-44 Yes Note 6 Partial Rep 1 cat. Note 2 Limit Switches V4-2579 Solenoid Valve Asco 19 Qualified (tS-048-L1)
V-6-2680 Solenoid Valve Asco 19 Qualified (tS-04BL2)
V-6-2681 Solenoid Valve Asco 19 Qualified (CS-04BL3)
V 4-2683 Solenoid Valve Asco 20 Qualified (tS-O'A-L1)
V 4-2685 Solenoid Valve ASco 20 Qualifted (tS-04A-L2) v-6-2685 Solenoid Valve Asco 20 Qualified (CS-04A-L3)
(See Page 21 for Notes.)
04450/012
ATTACHMENT I OYSTER CREEK ENVIRONMENTAL QUALIFICATION STATUS COMPOWENT SCHEDULAR REASON FOR COMPOWENT TER QUALIFICATION JC0 EXTENSION EXTENSION QUALIFICATION SCHEDULE FOR TAG mo. GENERIC NAlf MANUFACTURER E STATUS & REQUESTED REQUESTED MTHOD OUALIFICATIml Main Steam System (Cont'd)
RE22F ' Differential Pressure Barton 57 Qualified Indicating Switch RE22G Differential Pressure Barton 57 Qualifted Indicating Switch RE22H Differential Pressure Barton 57 Qualified Indicating Switch RE23A Pressure Swttch Meletron 46 Qualified RE238 Pressure Switch Meletron % Qualified RE23C Pressure Switch Meletron 46 Qualified RE230 Pressure Switch Heletron 46 Qualified cS03A Limit Switches Namco Replaced / Qual.
(V-1-7)
A.D.D NS-03A-L1 Solenoid Valve Asco 17 Qualified (V-1-7)
ES-03A-L2 Solenoid Valve Asco 17 Qualified (V-1-7) cS-03A-L3 Solenoid Valve Asco 17 Qualif ed (V-1-7) 2503B Limit Switches Namco Replaced / Qual.
(V-1-8)
A.B.D NS-033-L1 Solenoid valve Asco 18 Qualified (V-1-8)
ES-038-L2 Solenoid Valve Asco 18 Qualified (V-1-8)
ES-038-L3 Solenoid Valve Asco 18 Qualified (V-1-8)
(See Page 21 for Notes.)
04480/011
ATTACNMENT I OYSTER CREEK ENVIRONMENTAL QUALIFICATION STATUS COMPONENT SCNEDULAR REASON FOR COMPONENT TER QUALIFICATION JCO EXTENSION EXTENSION QUALIFICATION SCHEDULE FOR TAG ND. EmERIC MAlf MAmuFACTURER & STATUS & REQUESTED REQUESTED METMDD QUALIFICATION Main steam system IB10A Temperature Switch Fenwal 65 Qualifted C108 Temperature Switch Fenwal 65 Qualif1ed IB10C Temperature Switch Fenwal 65 Qualified 18100 Temperature Switch Fenwal 65 Qualtfled IB10E Temperature Switch Fenwal 65 Qualtfled IB10F Teneerature Switch Fenwal 65 Qualified IB10G Temperature Switch Fenwal 65 Qualified IB10H Teneerature Switch Fenwal 65 Qualifled IB10J Temperature Switch Fenwal 65 Qualifted 1310E Temperature Switch Fenwal 65 Qualifted IBlot Temperature Switch Fenwal 65 Qualtfted IB10M Temperature Switch Fenwal 65 Qualifted IBION Temperature Switch Fenwal 65 Qualifted IB10P Temperature Switch Fenwal 65 Qua11fted 1B100 Temperature Switch Fenwal 65 Qualifted IBICR Temperature Switch Fer.wal 65 Qualifted RE22A Otfferential Press. Barton 57 Qualtfted Indicating Switch RZ22B Differential Pres. Barton 57 Qualtfted Indicating Switch RE22C Differential Press. Barton 57 Qualified Indicating Switch RE22D Difft.rential Press. Barton 57 Qualifted Indicating Switch RE22E Differential Press. b .-*.on 57 Qualified Indicating Switch (Se2 Page 21 for hotes.)
044so/p10 u_____ _ _ _ _ _ _ . . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____ _ _ _ _ _ _ __ _ _ _ _ _ _ _ . _ _
ATTACHMENT I OYSTER CREEK ENVIRONMENTAL QUALIFICATION STATUS
- COMPONENT SCHEOULAR REASON FOR COMPONENT TER QUALIFICATION JC0 EXTENSION EXTENSION QUALIFICATION SCHEDULE FOR TAG m0. ENERIC MAff MANUFACTURER & STATUS & REQUESTED REQUESTED METHOD OUALIFICATION Drywell & Sunnression System (Cont'd)
V-23-13 Limit Switches (2) Micro Switch OC-85-30 Yes Note la Replace Before 11/30/85 V-23-14 Limit Switches (2) Micro Switch OC-85-30 Yes Note la Replace Before 11/30/85 V-23-15 Limit Switches (2) Micro Switch OC-85-30 Yes Note la Replace Before 11/30/85 V-23-16 Limit Switches (2) Micro Switch OC-85-30 Yes Note la Replace Before 11/30/85 V-23-17 Limit Switches (2) Namco OC-85-30 Yes Note la&2 Replace Note 2 V-23-18 Limit Switches (2) Namco DC-85-30 Yes Note la&2 Replace Note 2 V-23-19 Limit Switches (2) banco Qualified V-23-20 Limit Switches (2) Namco Qualified V-23-21 Limit Switches (2) Namco DC-85-30 Yes Notes 1a12 Replace Note 2 V-23-22 Limit Switches (2) Nai.co OC-85-30 Yes Notes la&2 Replace Note 2 V-24-16 Limit Switches (2) Micro Switch Qualified V-25-18 Limit Switches (2) Micro Switch Qualified V-27-1 Limit Switches (2) Micro Switch OC-85-28 Yes Note la Replace Before 11/30/85 V 2 Limit Switches (2) Micro Switch OC-85-28 Yes Note la Replace Before 11/30/85 V-27-3 Limit Switches (2) Namco DC-85-28 Yes Note la Replace Before 11/30/85 V-27-4 Limit Switches (2) Namco DC-85-28 Yes Note la Replace Before 11/30/85 V-28-17 Limit Switches Namco DC-85-32 Yes Note la Replace Before 11/30/85 V l a Limit Switches Namco OC-85-32 Yes Note la Replace Before 11/30/85 V-28-47 Limit Switches (2) Namco DC-85-32 Yes Note la Replace Before 11/30/85 V-31-2 Limit Switen Namco OC-85-45 Yes Note la Replace Before 11/30/85 (See Page 21 for Notes.)
04480/p9
ATTACMMENT I OYSTER CREEK ENVIRONMENTAL QUALIFICATION STATUS COMPONENT SCHEDULAR REASON FOR COMPONENT TER QUALIFICATION JCD EXTENSION EXTENSION QUALIFICATION SCHEDULE FOR TAG NO. GENERIC MME MAhuFACTURER & STATUS & REQUESTED REQUESTED METHOD OUALIFICATION Drywell & Suceression System (Cont'd)
TE-109C Temperature Element Pyco Note 4 OC-85-34 Yes Note la Replace Before 11/30/85 TE-1090 Temperature Element Pyco Note 4 OC-85-34 Yes Note la Replace Before 11/30/85 Tip Ball Solenoid valves General Electric OC-85-39 Yes Analysis Before 11/30/85 v31:es V-5-147 Motor Operator Limitorque 14 OC-85-3 Yes Note 2 Partial Replcat. Note 2 V-5-166 Motor Operator Limitorque 11 Qualified U-5-157 Motor Operator Limitorque 14 OC-85-3 Yes Note 2 Partial Replcat. Note 2 V-4-395 Limit Switch Namco Qualified v 15-1 Motor Operator Limitorque 3 Replaced / Qual.
V-15-2 Motor Operator Limitorque 4 Replaced / Qual.
V-15-14 Motor Operator Limitorque 4 Replaced / Qual.
V-15-61 Motor Operator Limitorque 4 Qualifted V-17-1 Motor Operator Limitorque 5 Replaced / Qual.
V-17-2 Motor Operator Limitorque 5 Replaced / Qual.
V-17-3 Motor Operator Limitorque 5 Replaced / Qual.
V-17-19 Motor Operator Limitorque 3 Replaced / Qual.
V-17-54 Motor Operator Limitorque 1 Replaced / Qual.
v-17-55 Motor Operator Ltattorque 5 0C-85-40 No Replace Before 3/31/85 Limit Smitch V-17-56 Motor Operator Ltattorque 5 0C-85-40 No Replace Before 3/31/85 Limit Switch V-17-57 Motor Operator Limitorque 5 OC-85-40 No Replace Before 3/31/85 Limit S= itch V-22-1 Listt settches (21 Namco Qualtfied V-22-2 Limit Switches (2) Namco Qualifted V-22-28 Limit Smitches (2) Namco Qualified V-22-29 Limit Switches (2) Namco Qualified (See Page 21 for Notes.)
C48a/08
ATTAONIENT I OYSTER CEEEK ENVIEDNIENTAL QUALIFICATION ST;TUS COMPONENT SCNEDULAR REASON FOR CDMPONENT TER OuALIFICATION JC0 EXTENSION EXTENSION QUALIFICATION SCNEDULE FOR l _ TAG NO. GEhERIC NASE MAmuFACTuMR & STaTIE WL RE0uESTED REDuESTED METh0D QUALIFICATION l
l Containment Sarav Svsig (Cont'd) l V-21-17 Motor Operator Limitorque 15 Note 4 OC-85-4 No Partial Rplcat. Before 3/31/85 W-21-17 Key Lock Control General Electric Note 4 OC-85-24 Yes Note la Replace Before 11/30/85 l
Switch V-21-18 Motor Operator tiettorque Note 4 OC-85-4 No Partial Rpicat. Before 3/31/85 V 21-18 Key Lock Control General Electric Note 4 OC-85-24 Yes Note la Replace Before 11/30/85 Switch 1-1 Pump Motor General Electric 70 Qualtfted 1-2 Pump Motor General Electric 70 Qualtfled 1-3 Puus Motor General Electric 70 Qualifted 1-4 Puno Motor General Electric 70 Qualifted Q<well 8L Suoaression system DPS-66A Differential ITT Barton Note 4 OC-15-15 Yes Notes la&2 Replace Note 2 Pressure Switch CPS-668 Differential ITT sarton Note 4 OC-85-15 Yes Notes la&2 Replace Note 2 Pressure Switch LT-17 Level Transmitter Rosemount Qualified LT-13 Level Transmitter Rosemount Qualified PT-53 Pressure Transmitter Rosamount Qualtfted PT-54 Pressure Transmitter Rosemount Qualifted Rf04A Pressure Switch Static-0-Ring 50 Note 4 OC-85-5 Ves Notes la&2 Replace Note 2 RE048 Pressure Switch Static-0-Ring 50 Note 4 OC-85-5 Yes Notes la&2 Replace Note 2 RE04C Pressure Switch Static-0-Ring 50 Note 4 OC-85-5 Ves Notes la&2 Replace Note 2 RE043 Pressure Switch Static-0-Ring 50 Note 4 OC-85-5 Yes Notes la&2 Replace Note 2 TE-109A Temperature Element Pyco Note 4 OC-85-34 Yes Note la Replace Before 11/30/85 TE-1098 Temperature Element Pyco Note 4 OC-85-34 Yes Note la Replace Before 11/30/85 (See Page 21 for Notes.)
C44so/p7
ATTAQstENT I OYSTER CBEEK ENVIRONMENTAL QUALIFICATION STATUS COMPONENT SCHEOULAR REASON FOR COMPONENT TER QUALIFICATION JC0 EXTENSION EXTENSION QUALIFICATION SCHE 00LE FOR TAG be. GEhERIC BApE MANUFACTURER & STATUS E RE0uESTED REQUESTED METHOD OUALIFICATIfMI Containment Sarar System (Cont'd)
V-21-1 Key Lock Control General Electric Note 4 OC-85-24 Yes Note la Replace Before 11/30/85 Switch 0-31-3 Motor Operator Liettorque 6 OC-85-41 No Replace Before 3/31/85 Limit Switch V-al-3 Key Lock Control General Electric Note 4 OC-85-24 Yes Note la Replace Before 11/30/85 Switch V-21-5 Motor Operator Limitorque 7 mote 4 OC-45-2 No Partial Rplcat. Before 3/31/85 V-21-5 Key Lock Control General Electric hote 4 0C-85-24 Yes Note la Replace Before 11/30/85 Seitch U 21-7 Motor Operator Limitorque 6 0C-85-41 20 Replace Before 3/31/85 Liett Switch U-21-7 Key Lock Control General Electric Note 4 OC-85-24 Yes Note la Replace Before 11/30/85 Switch U-21-9 Motor Operator Limitorque 6 OC-85-41 20 Replace Before 3/31/85 Limit Switch U-21-0 Key Lock Control General Electric note 4 OC-85-24 Yes Note la Replace Before 11/30/85 Switch U-21-11 Motor Operator Limitorque 7 Note 4 OC-85-2 No Partial Rpicat. Before 3/31/85 V-21-11 Key Lock Control General Electric mote 4 OC-85-24 Yes Note la Replace Before 11/30/85 S= itch U-21-13 Motor Operator Limitorque 13 Note 4 OC-85-4 20 Partial Rplcat. Before 3/31/85 U-21-13 Key Lock Control General Electric hote 4 OC-85-24 Yes Note la Replace Before 11/30/85 Switch U-21-15 Motor Operator Limitorque note 4 OC-85-4 No Partial Rplemt. Before 3/31/85 U-21-15 Key Lock Control General Electric Note 4 OC-85-24 Yes Note la Replace Before 11/30/85 Switch (See Page 21 for Notes.)
04480/p6
ATTADeENT I OYSTER CREEK EnWIADMIGITAL QUALIFICATION STCTUS COBSONENT SCMEDULAR REASON FOR COMP 0mERT TER QUALIFICATION JC0 EXTENSION EXTENSION QUALIFICATION SCHEDULE FOR TAG to. CERERIC BAME MAbuFACTURER E STATIN & M(M SIED REQUESTED METN00 QUALIFICATIDM Containment Sarar System IPC3A Flow Transmitter General Electric 45 Note 4 OC-85-12 Yes notes Ib& 2 Replace Next Refueline Outage IP038 Flow Transmitter General Electric 44 Note 4 OC-85-12 Yes notes IbE2 Replace Next Refueline Outage IP15A Pressure Switch Barton 44 fbalif ted ID158 Pressure Suttcti Barton 48 Qualified 3P15C Pressure Switch Barton 43 Qualtfted IP150 Pressure Switch Barton 48 Qualtfted IP18A Temperature Suttch Ashcroft Note 4 OC-45-19 Yes Note la Replace Before 11/30/85 2P188 Temperature Switch Ashcroft hote 4 OC-85-19 Yes Note la Replace Before 11/30/85 0-21-1 Motor Operator Limitorgue 6 OC-85-41 20 Replace Before 3/31/85 Limit Switch (See Page 21 for hotes.)
04480/p5
c- ,
ATTADeqENT I OYSTER CREEK ENVIE0mMENTAL QUALIFICATION STATUS COMPONENT SCMEDULAR REASom FOR JC0 'I CDPPGmERT TER QUALIFICATI0m EXTENSION EXTENSION OUALIFICATION SCMEDULE FOR tag mD. GENERIC MAff MauFACTuaER E STATUS & MGLESTED RE0uESTED IETMDD QUALIFICATIfW care sara, ans automatte neareuurtrattan s ,stan (Cont'd)
RV46A Pressure Smitch Barton 48 Qualtfled ev4e8 Pressure Settch Barton 48 Qualifted RV46C Pressure Switch Barton 48 Qualifted Rv460 Pressure Suiten Barton 48 Qualtfted U-20-3 Motor Operator Liettorgue OC-85-42 no neplace Before 3/31/85 Limit Suiten V-20-4 Motor Operator Limitorque OC-85-42 No Replace Before 3/31/85 Limit Switch V-20-15 Motor Operator Ltattorque 2 Qualifled V-20-21 Motor Operator Ltattorque 2 Qualifled V-20-32 Motor Operator Ltattorque OC-85-42 No Replace Before 3/31/85 Liett Suiten V-20-33 Motor Operator Limitorque OC-85-42 No Replace Before 3/31/85 tiett Sutten V-20-40 Motor Operator Ltattorque 2 Qualtfted U-20-41 Motor Operator Liettorque 2 Qualtfted V-20-92 Limit Switch (2) hanco OC-85-25 Yes Note la Replace Before 11/30/85 V-20-93 Limit Switch (2) manco 0C-85-25 Yes Note la Replace Before 11/30/85 V 20-94 Limit Suiten (2) manco DC-85-25 Yes Note la Replace Before 11/30/85 V-20-95 Liett Seiten (2) manco OC-85-25 Yes Note la Replace Before 11/30/85 (See Page 21 for hotes.)
04480/04
ATTAOmqENT I OYSTER C KEK ENVIEDuMENTAL QUALIFICATION STATUS CDuPOWENT SCNEDULAR REASom FOR CDnPOWENT TER OUALIFICATION JCD EXTEmSION EXTERSION QUALIFICATION SCNEDULE FOR Tac me. GENERIC mAff MAmuFACTUAER & STATUS & EEGUESTED RE0uESTED stETN00 QuALIFICATIM Core Sarav and Ata- tic m ostrattan Svstan (Cont'd) a m2330 Piame sector General Electric 69 Qualifted m2330 Pump reotor General Electric 6a Qualifted RE17A Pressure Switch Barksdale 55 Qualtfted RE17B Pressure Switch Bartsdale 55 Qualtfted RE17C Pressure Swttch Bartsdale 53 Qualtfied RE17D Pressure Switch Barksdale 53 Qualtfted W26A Flow Transmitter GE/ptnC 43 OC-85-11 Yes notesib&2 Beplace next Befueling Outage W268 Flow Transmitter GE/MAC 43 OC-85-11 Yes Notes 1bE2 Seplace next Refueling Outage W29A Pressure Switch peercoid 51 OC-85-9 Yes Note la Replace Before 11/30/85 W298 Pressure Switch Mercoid 51 OC-85-9 Ves Note la Replace Before 11/30/85 W29C Pressure Switch Mercoid 51 OC-85-9 Tes note la Replace Before 11/30/85 W290 Pressure Switch Mercold 51 OC-45-9 Yes Note la Replace Before 11/30/85 W40A Diff. Press. Switch SOR 52 Seplaced/ Qual.
W4cB Diff. Press. Switch SOR 51 Seplaced/ Qual.
DJcCC Diff Press. Switch SDR 52 Seplaced/ Qual.
WtCD Otff. Press. Switch SOR 51 Seplaced/ Qual.
(See Paec 21 for Notes.)
044so/03
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ATTADeIENT I 01STE3 CBEEK ENVIRONMENTAL QUALIFICATION STZTUS COMP 0 MENT SCNEDULAR REASON FOR C[ pep 0 MENT TER QUALIFICATION JCD EXTENSION EXTENSION OUALIFICATIOu SCHEDULE FOR Tac am. GEmERIC BAME MAmuFACTUAER E STATE & MDuESTED REDUESTED stETuGD QUALIFICATION EmereencY Condenier Systag (Cont'd)
U-14-35 Motor Operator Limitorswe 9 seplaced/ Qual.
U-10-36 Motor Operator Liettorgue 10 Replaced /Oual.
U-14-37 Motor Operator Limitoreae 10 Replaced /Oual.
Cote Soray and Automatic Daareuurintian Sustan IAa3A Pressure Switch Dresser /Barksdale 49 OC-45-1 Yes Notes la&2 Replace Next Refuelins Outage and Controllers IA338 Pressure Switch Dresser /tarksdale 49 OC-85-1 Yes Notes la&2 Replace Next Befueling Outage and Controllers IAa3C Pressure Smitch Dresser /Barksdale 49 OC-85-1 Tes notes la&2 Replace next Refueline Outage and Controllers IAa3D Pressure Smitch Dresser /Sarksdale 49 OC-45-1 Yes Notes laE2 Replace Next Refueling Outage and Controllers IAa3E Pressure Smitch Dresser /Sarksdale 49 OC-35-1 Yes Notes la&2 Replace next Refueltne Outage and controllers me-108A Solenoid Valve Dresser 37 Oualtfted me-1088 Solenoid Valve Dresser 37 Qualtf1ed me-10aC Solenotd Valve Dresser 37 Qualifled mR-1080 Solenoid valve Dresser 37 Qualifted mR-10aE Solenoid valve Dresser 37 Qualtfied RZOTA P m Motor General Electric 67 Qualtfled mZ018 Pump Motor General Electric 67 Qualifted RZOIC Puus Motor General Electric 67 Qualified m2010 Pump Motor General Electric 67 Qualified m233A Punip Motor General Electric 69 Qualtfted l
EZG3B pump Motor General Electric 6a Qualtfied (See rage 21 for notes.)
044ac/02
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ATTAOeENT I OYSTER CREEK ENVIRouMENTAL QUALIFICATION ST;TUS l
l COMP 0utui SOeEDULAR REASON FOR COnpOnEmT TER QUALIFICATION JCD EXTENSION EXTENSION QUALIFICATION SCMEDULE FOR TAC 30. GEBERIC hasE tahauTACTuRER gL Status gL M REGMESTED METMOD QuALIFICATIfkl Emereeney Condenser Svstaa 1505-0.1 Differential Barton 54 Qualtfled eressure Switch 1605-a2 Differential Barton 5a Qualified Pressure Smitch 1805-B1 DtfferentIal sarton sa Qualtfied Pressure Smitch 1805-82 Differential Barton 18 Qualtfied Pressure Switch 1811-Al Differential sarton sa Qualtfied Pressure Suiten IB11-A2 Differenttal Barton 58 Qua11fied Pressure Switch 1011-01 Differential Barton 5a Qualtfled Pressure Switch 1311-82 Otfferential Barton 54 Qualtfled Pressure Switch IG06-A Level Transmitter GE/MAC 42 OC-85-7 Yes hotes Ibt2 Replace Next Refueltne outage IG06-B Level Transmitter CE/MAC 42 OC-SS-7 Yes Notes 1t42 Replace Next Refueling Outage V 14-1 Liett Switch Micro Settch OC-85-26 Yes hotes 14&2 Replace Note 2 i
V-14-5 Limit Switch Micro Switch OC-85-26 Yes Notes la&2 Replace Note 2 W-14-19 Limit Smitch Micro Smitch OC-45-26 Yes Notes la&2 Replace Note 2 i W-14-20 Ltait Smiten Micro Suiten DC-85-26 Yes hotes la&2 Replace note 2 W-14-30 Motor Operator Limitorque a Replaceo?1ual.
v-14-31 Motor Operator Limitorque 9 Replaced / Qual.
v-14-32 Motor Operator Limitorque 8 Replaced / Qual.
V-14-33 Motor Operator Liettorque 9 Replaced /Oual.
v 14-34 Motor Operator Limitormae 9 Replaced / Qual.
tSee rage 21 for notes.)
0448c/p1
ATTACHMENT II OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) i
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JCO-0C-85-1, Rev.1 i-January 31, 1985 '
Page 1 of 2 4
i OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) 1 I
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JCO-0C-85-1, gey, j U*"Unr1 31, 1985 Page 2 of 2 CONPONENTS Tag Numbers IA-83A, IA-838. IA-83C, IA-83D, IA-83E Description Pressure Switches to open EMRVs on high RPV pressure OBJECTIVE The objective of this discussion is to determine:
that the safety function can be accomplished by some other qualifled components; that the failure of the identified components as a result of a harsh environment will not degrade other safety functions or mislead the operator.
COMPONENT LOCATIO_N_
All of these switches are located in the Reactor Building on elevation 51'3".
Pressure switches IA-83A, B are located on the east drywell wall; switch IA-83C is located in the southeast quadrant; and switches IA-830 E are located in the northwest quadrant.
COMPONENT FUNCTION These components open the EMRVs on high RPV pressure to provide protection against over-pressurization. These switches would be required to work only for a small break LOCA which is not large enough to remove decay heat and when both isolation condensers are not available.
EVALUATION These switches are grouped in different areas in the Reactor Building such that the areas will not be in the same harsh environment simultaneously.
Consequently, simultaneous failure of all of the switches is not postulated.
In addition, there are 16 safety valves on the steam line which would provide over-pressure protection for the RPV in the event that the EMRVs are unavailable or insufficient to reifeve the increasing RPV pressure. Further, the operator can manually operate the remaining operable EMRVs to depressurize the RPV. Only one operable EMRV is required to provide adequate decay heat removal.
The failure of these switches will not mislead the operator, since the operator is instructed in the E0Ps to manually operate the EMRVs if they fail to initiate automatically. If a single isolation condenser is available, over-pressurization will not occur and the switches are not required.
CONCLUSION l
The failure of any one pressure switch will not prevent over-pressure protection of the RPV and operation of the other EMRVs. The operator will not be misled by this condition.
l L >
JCO-0C-85-2, Rev. 1 January 31, 1985 Page 1 of 2 OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO)
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JCO-0C-85-2, Rev. 1 January 31, 1985 Page 2 of 3
' COMPONENTS Tag Numbers V-21-5, V-21-11 Description Containment Spray Drywell Injection Valv' Motor Operators OBJECTIVE The objective of this discussion is to determine:
that there is a basis for concluding that the existing components will perform their required function; that the failure of these components will not degrade other safety ,
functions nor mislead the operator.
COMPONENT LOCATION These valves and their associated operators are located in the Reactor Building on the 23'6" elevation. V-21-5 is located in the Southeast quadrant and V-21-11 is located in the northeast quadrant.
COMPONENT FUNCTION These valves allow containment spray injection into the drywell to remove heat and reduce containment pressure for breaks inside the drywell.
EVALUATION The valves and their associated operators are located in the Reactor Building and thus are not subject to the harsh environment inside the drywell when they
're required to function. For a break inside the containment, the environment in the Reactor Building is not expected to become harsher than that for which the component can be quallfled. Hence these components are expected to function.
There is a failure mechanism of these motor operators due to radiation which can cause the valves to change state. It will take a considerable period of time for the integrated radiation dose to exceed the qualification value for the switch. A failure of the keylock switch at that time may cause some of the valves to reposition. If the Containment Spray system is lost as a result of this failure, the decay heat from the core would be sufficiently low so that the Containment Spray System would not be required to remove heat from the torus for some time. In addition, ambient losses from the torus shell may be sufficient to provide torus cooling. Further, it is extremely unitkely that all of the valves will reposition themselves in the worst alignment at the same time so that there is no injection path to the containment.
JCO-0C-85-2, Rev. 1 January 31, 1985 Page 3 of 3 EVALUATION (Continued)
Breaks outside containment are assumed to isolate. The only condition which would require Containment Spray is an event in which both Isolation Condensers were lost. Even under this remote scenario, Containment Spray would not be
= required to provide torus cooling for several hours. These valves are normally open and are not required to change position except during the monthly surveillance testing of the Containment Spray System. During this test, only one of the two Containment Spray Subsystems is tested at any one time. The operable subsystem provides ample capability te satisfy the function of the system.
The operator in the control room has position Indicating Ilghts for these valves and thus, the failure of these valves would not mislead the operator.
CONCLUSION The failure of one of these valves to operate does not degrade the effective-ness of the Containment Spray System to perform its function.
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JCO-0C-85-3, Rev. 1 January 31, 1985 ,
Page 1 of 2 i i-1 9
t OYSTER CREEK NUCLEAR GENERATING STATION
. . JUSTIFICATION FOR CONTINUED OPERATION (JCO) 4
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JCO-0C-85-3, Rev. 1 January 31, 1985 Page 2 of 2 COMPONENTS Tag Numbers V-5-167, V-5-147 Description RBCCW Containment Isolation Valve Motor Operators OBJECTIVE The objective of this discussion is to determine:
that the failure of these components (V-5-167,147) will not degrade other safety functions; that the safety function can be accomplished by some other qualified equipment.
COMPONENT LOCATION Both of these valves and their associated motor operators are located in the southeast quadrant in the Reactor Builaing on the 23'6" elevation.
COMPONENT FUNCTION These components function to isolate the drywell upon a containment isolation signal resulting from a design basis accident.
EVALUATION Both components are located outside the drywell and are not expected to be affected by inside containment breaks. In the event of a break inside the drywell, both valves will close upon a receipt of a containment isolation signal.
For breaks outside containment, Y-5-167 is in series with a redundant qualified isolation valve (V-5-166) inside containment which would function for those breaks (outside containment) which could create a harsh environment near V-5-167. The inlet RBCCW valve (V-5-147) is in series with a check valve (V-5-165) located inside containment which would function to prevent releases from containment in the event of a break (outside the drywell) which creates a harsh environment near V-5-147.
CONCLUSION The failure of these components will not degrade the isolation of containment since both are in series with components which will function to isolate the drywell in the event of a DBA.
_- .._=. - .. . __ -..- . . . .. . . .. . . . ... ..___ -.- . . . -- . .-. _.. -.-..
l JCO-0C-85-4, Rev. 1 January 31, 1985 Page 1 of 3 l
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OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO)
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JCO-0C-85-4, Rev. 1 January 31, 1985 Page 2 of 3 COMPONENTS Tag Numbers V-21-13, V-21-15, V-21-17, V-21-18 Description V-21-13, 17 -Containment Spray Dynamic Test Valve Motor Operators V-21-15, 18 -Containment Spray Torus Spray Valve Motor Operators OBJECTIVE The objective of this discussion is to determine:
that there is a basis for concluding that the existing components will perform their required function; that the failure of these components will not degrade other safety functions nor mislead the operator.
COMPONENT LOCATION These valves and their associated motor operators are located in the Reactor Building on the 23'6" elevation. V-21-13, 15 are located in the southeast quadrant, and V-21-17, 18 are located in the northeast quadrant.
COMPONENT FUNCTION These valves allow containment spray injection into the torus to remove heat and reduce pressure for breaks inside the drywell. The V-21-15, 18 valves are used in the normal mode of containment spray to provide a small spray flow to the torus. The V-21-13, 17 valves are used in the dynamic test mode to provide cooling to the torus pool.
EVALUATION These valves and their associated operators are located in the Reactor Building and are not subject to the harsh environment inside the drywell when they are required to function. For a break inside the containment, the environment in the Reactor Building is not expected to become harsher than that for which the component can be qualified. Hence these components are expected to function.
There is a failure mechanism of these motor operators due to radiation which can cause the valves to change state. It will take a considerable period of time for the integrated radiation dose to exceed the qualification value for the switch. A failure of the keylock switch at that time may cause some of the valves to reposition. If the Containment Spray system is lost as a result of this failure, the decay heat from the core would be sufficiently low so that the Containment Spray System would not be required to remove heat from the torus for some time. In addition, ambient losses from the torus shell may be sufficient to provide torus cooling. Further, it is extremely unlikely that all of the valves will reposition themselves in the worst alignment at the same time so that there is no injection path to the containment.
JCO-0C-85-4, Rev. 1 January 31, 1985 Page 3 of 3 EVALUATION (Continued)
Bre .s outside containment are assumed to isolate. The only condition which would require Containment Spray is an event in which both Isolation Condensers were lost. Even under this remote scenario, Containment Spray would not be required to provide torus cooling for several hours.
The operator in the control room has indication of valve position and thus, the failure of these valves would not mislead the operator.
CONCLUSION The failure of tnese valves to operate for outside containment breaks does not degrade the effectiveness of the Containment Spray System to perform its function. They are not expected to fail for breaks inside containment.
JCO'-0C-85-5, Rev. 1 January 31, 1985 Page 1 of 2 OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) t
JC0-0C-85-5, Rev. 1 January 31, 1985 Page 2 of 2 COMPONENTS Tag Numbers RE-04A, RE-04B, RE-04C, RE-04D
. Description High Drywell Pressure Switches OBJECTIVE The objective of this discussion is to determine that there is a basis for concluding that the existing component will perform its required function.
COMPONENT LOCATION All of these pressure switches are located in the northwest quadrant of elevation 51'3" in the Reactor Building.
COMPONENT FUNCTION These switches generate scram, drywell and reactor building isolation signals and initiate the Standby Gas Treatment System.
EVALUATION These switches are located outside the primary containment and are required to mitigate events inside the containment. For.a break inside'the containment, the environment in the Reactor Building is not expected to become harsher than that for which the component can be qualified. Hence these components are expected to function. The switches may be in a harsh environment for breaks outside containment; however, they are not required to function to mitigate these events. .Their failure for outside containment breaks will not--
mislead the operator since he is provided with a qualified drywell pressure indication from wnich he can determine that the break is outside the drywell.
The operator may initiate containment spray if he thinks that he has exceeded the drywell pressure action level in the E0Ps. This, however, should not result in unacceptable consequences.
C0llCLUSION The switches will perform their function whenever they are required to perform
- their safety function.
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OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) 4 I
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I' JCO-0C-85-6, Rev. 1 January 31, 1985 Page 2 of 2 COMPONENTS Tag Numbers ID-46A, ID-46B Description ID46A,B are RPV Hide Range Pressure Transmitters.
OBJECTIVE The objective of this discussion is to determine that the failure of these components will not degrade other safety functions or mislead the operator.
COMPONENT LOCATION These transmitters are located in the Reactor Building on elevation 51'3";
ID-46B is located in the southeast quadrant and ID-46A is located in the northwest quadrant.
COMPONENT FUNCTION These transmitters feed the wide range instruments in the control room. They do not provide any safety related function.
EVALUATION These transmitters are located outside the primary containment. For breaks inside the containment, the environment in the Reactor Building is not expected to become harsher that that for which the component can be quallfled. Hence they will function normally. For breaks outside the containment, their failure will not affect the operation of any safety system.
Also, these transmitters are located in different areas of the Reactor Building such that they will not see a harsh envi-onment simultaneously. The operator has alternate RPV pressure indications ava;1able in the Control Room so that he would not be misled.
CONCLUSION These transmitters do not provide any safety related function with respect to the automatic actuation of a safety system. A single break should not result in the failure of both of these components and alternate pressure indicators such that pressure indication would not be completely lost.
JCO-0C-85-7, Rev. 1 January 31, 1985 Page 1 of 2 OYSTER CREEK NUCLEAR GENERATING STATION s
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JC0-0C-85-7, Rav. 1 January 31, 1985 Page 2 of 2 COMPONENTS Tag Numbers IG-06A, IG-06B Description Isolation Condenser Secondary Side Level Transmitter OBJECTIVE The objective of this discussion is to determine that the failure of the identified components as a result of a harsh environment will not degrade other safety functions or mislead the operator.
COMPONENT LOCATION
-These transmitters are located in the east area of the Reactor Building on the 95' elevation.
' COMPONENT FUNCTION The components ' sense and transmit to the control room the water level on the secondary (shell) side of the isolation condensers.
EVALUATION The. isolation condensers can operate for at least 45 minutes each without makeup to-the secondary side. Makeup water is added from the Condensate Transfer System by operator action. Loss of level indication would not prevent the operator from adding makeup to the shell side of the isolation condensers.
If, after 45 minutes (or 100 minutes if 2 condensers are in service), the operator does not take action to makeup to the secondary side, heat transfer capacity will be reduced, and the reactor vessel will begin to re-pressurize.
The operator has the neans to determine pressure and can take action to depressurize the RPV by using alternate pressure control systems.
CONCLUSION, The failure of the level transmitters will not degrade the function of the isolation condensers since the system can operate for up to 100 minutes on the inventory of water contained on the shell side of both condensers. The failure of these transmitters will not mislead the operator since other indications will allow the operator to make use of alternate decay heat removal ' systems.
JCO-0C-85-8, Rev. 1 January 31, 1985 Page 1 of 2 OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO)
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i JC0-0C-85-8, Rav. 1 l January 31, 1985 Page 2 of 2 COMPONENTS-l Tag Numbers ID-13A, 1D-13B i Description GEMAC Water Level Transmitters !
OBJECTIVE i The objective of this discussion is to determine:
that the failure of these components will not degrade other safety functions nor mislead the operator; that the function may be accomplished by some other qualified components.
COMPONENT LOCATION Transmitter ID-13A is located in the northwest quadrant of the Reactor Building on the 51'3" elevation. Transmitter ID-138 is located in the southeast quadrant of the Reactor Building on the 51'3" elevation.
C0W ONENT FUNCTION These. level transmitters are only one of several means used by the operator for indications of RPY water level. These transmitters have nearly the same
} span as the YARWAY transmitters and thus provide redundancy to the YARWAYS.
One of two GEMACs is selected to provide an input to feedwater level control which is not safety related.
! EVALUATION These transmitters are located in the Reactor Building and are not subject to the harsh _ environment inside the drywell when they are regt.f red to function.
For a break inside the containment, the environment in the Reactor Building is not expected to become harsher than that for which the component can be qualified. Hence these components are expected to function. For breaks outside the containment, the loss of these transmitters does not prevent any
. plant. safety function from occurring since no safety-related trips are based on the GEMAC transmitters. The two GEMAC transmitters are located in different areas of the Reactor Building and will not see the same harsh environment
! simultaneously. The GEMAC transmitters provide diverse level indication to
,. the YARWAY level instruments which are the operator's primary level L indication. Also, the operator would use additional control room indication to make his determination. Even in the event the operator judges that he cannot determine the level, the E0Ps provide the operator with guidance to prevent uncovering the core..
CONCLUSION l.
l The . failure of these transmitters does not significantly degrade the ability
- of the' operator to monitor the RPV water level. There are no safety related trips which are based on the GEMAC transmitters.
JCO-0C-85-9, Rev. 1 January 31, 1985 Page 1 of 2 4
OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO)
JC0-0C-85-9, Rev. 1 January 31, 1985 Page 2 of 2 COMPONENTS Tag Numbers RV-29A, RV-298, RV-29C, RV-29D Description These are pressure switches which provide core spray booster pumps start permissive and indication of backup main pump start based on core spray main pumps discharge pressure.
OBJECTIVE The objective of this discussion is to show that the failure of these switches will not degrade core spray safety function and will not mislead the operator.
COMPONENT LOCATION These switches are located in the Reactor Building on the -19'6" elevation.
RV-29A,C are located in the northwest corner room; RV-298,0 are located in the southwest corner room.
COMPONENT FUNCTION These pressure switches start core spray booster pumps based on core spray main pumps discharge pressure.
EVALUATION These switches are located outside the containment and are not subject to the harsh environment inside the drywell when they are required to function. For breaks inside the containment, the environment in the Reactor Building is not expected to become harsher than that for which the component can be quallfled. Hence they will perform their safety function as required. The break which would cause a harsh environment in the area of the switches is a rupture of the Main Steam Line in the Reactor Building. For this event, both Isolation Condensers would be available for decay heat removal. Breaks outside containment are all assumed to isolate. Thus, if core spray would be necessary, it would only be required for RPV inventory makeup. Then, only one core spray subsystem would be required to function. These switches are grouped in different areas and it is not expected that all of these switches would be exposed to a harsh environment simultaneously. Thus, at least one core spray subsystem would be functional.
The operator has multiple indications in the control room of main and booster pump status so that the failure of these pressure switches will not cause him to be misled.
CONCLUSION The operator can take manual control, hence the failure of these switches will not degrade core spray safety function nor mislead the operator.
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i JCO-0C-85-10, Rev. 1 January 31, 1985
-Page 1 of 2 OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR LONTINUED OPERATION (JCO)
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' Tag Numbers RV-26A, RV-26B Description Core Spray System Flow Transmitters OBJECTIVE The objective of this discussion is to determine:
that the failure of these components will not degrade other safety functions nor mislead the operator; that the function may be accomplished by some other quallfled components.
COMPONENT LOCATION The RV-26A transmitter is located in the northwest quadrant of the Reactor Building on elevation 51'3". The RV-268 transmitter is located in the southwest quadrant of the Reactor Building on elevation 75'3".
COMPONENT FUNCTION These flow transmitters provide core spray flow indication to the operator during a break condition.
EVALUATION For a break inside the containment, the environment in the Reactor Building is not expected to become harsher than that for which the component can be qualified. Hence these components are expected to function. For breaks outside the drywell, the operator may confirm Core Spray System operation using RPV water level in the event that the flow transmitters were lost due to a harsh environment. In addition, the operator has indication that the core spray pumps are running and that the valves are open. The flow transmitters provide no input to other safety systems and thus will not impact other safety functions. Further, the transmitters are located in different areas of the Reactor Building and will not simultaneously experience the harsh environment.
CONCLUSION The failure of these transmitters does not preclude the operator from confirming Core Spray System operation and does not degrade other safety functions.
JCO-0C-85-12, Rev. 1 January 31, 1985 Page 1 of 3 OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO)
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- COMPONENTS Tag Numbers IP-03A,.IP-038 Description Containment Spray Flow Transmitters (IP-03A,B)
OBJECTIVE
-The objective of this discussion is to determine:
that the failure of these components will not degrade other safety functions nor mislead the operator; that there is a basis for concluding that the existing components will perform their required function.
COMPONENT LOCATION 1
Transmitter IP-03A is located in the northeast quadrant of the Reactor Bu11 ding on elevation 23'6". Transmitter IP-03B is located in the southeast quadrant in the Reactor Building on elevation 23'6".
COMPONENT FUNCTION-These components provide the operator indications of. flow to confirm the performance of the containment spray pumps. The Containment Spray System is required to remove heat and reduce pressure inside-the containment for breaks inside the drywell.-
EVALUATION These components are required to function for breaks inside the drywell.
However, the_ components are. located outside containment. For a break inside the containment, the environment in the Reactor Building is not expected to become harsher than that for which the component can be qualified. Hence these components are expected to function. Breaks outside the containment are assumed to isolate. The only outside containment break which would require Containment Spray is an event in which both Isolation Condensers were lost.
- i. Even under this remote scenario, Containment Spray would not be required to provide torus cooling for several hours. In addition, the failure of the instrument components will not mislead the operacor because he can determine from drywell parameters that the break is not inside containment and that for outside containment breaks, the Containment Spray System is not required until at least several hours after the event. If containment spray should start inadvertently as a result of the failure, there should be no safety consequence. Further, there is control room indication of pump actuation.
i-L CONCLUSION l The'fallure of these components does not degrade the effectiveness of the
- l. Containment Spray System to perform its function because the components will function for events for which they are required to operate.
'JCO-0C-85-13, Rev. 1 January 31, 1985 Page 1 of 2 OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) ;
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- JCO-0C-85-14, Rev.'I Januarf 31, 1985 Page 1 of 2 OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO)
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JCO-0C-85-14, Rev. 1 January 31, 1985 Page 2 of 2 COMPONENTS Tag Numbers RE-02B; RE-02C, RE-020 Description LOH-LOW Level Switches-OBJECTIVE The objective of this discussion is to determine that the failure of these components will not degrade other safety functions or mislead the operator.
COMPONENT LOCATION RE-02,C is located on rack RK01 in the northwest quadrant of the Reactor Building at elevation 51'3". RE-02B,D are located on rack RK02 in the southeast quadrant of the Reactor Building at elevation 51'3".
COMPONENT FUNCTION The Low-Low switches function is to turn on core spray pumps, containment spray pumps (high drywell pressure is also needed), primary and secondary containment isolation, trip recirculation pumps, and initiate the isolation condensers.
EVALUATION If the break is inside the containment, the switches will perform their safety function as required because, for a break inside the containment, the environment in the Reactor Buldling is not expected to become harsher than that for which the component can be qualified. For any condition which results in Low-Low level, the operator is tra.ined to initiate manually all systems which should have initiated automatically but failed to do so. It is a general operator training concept that the operator is to backup all automatic actuations. There are sufficient control room indications to do so. This is not a misleading condition. For large and intermediate breaks, it is expected that Low-Low level will be reached very quickly before the effect of the harsh environment becomes severe. Further, pairs of switches are located at different locations in the Reactor Building and would not be susceptible to the same harsh environment simultaneously. For a small break, even though Low-Low may not be reached quickly, the operator would have a longer period of time to act, and the break would not affect more than one pair of instruments so that auto actuation would occur.
CONCLUSION The operator should manually perform all actions required by Low-Low setpoint in case the switches fall. Both pairs of switches would not be subjected to the same harsh environment because of their diverse locations.
, JCO-0C-85-15',.Rev. 1
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F' JCO-0C-85-15 Rev. 1 January 31 0 1985 Page 2 of 2 COMPONENTS Tag Numbers DPS-66A, DPS-66B Description Switches to open vacuum breakers based on Reactor Building to torus dP OBJECTIVE The objective of this discussion is to determine that the failure of these components will not degrade other safety functions or mislead the operator.
COMPONENT LOCATION These components are located in the southeast quadrant of the Reactor Building at elevation 23'6".
COMPONENT FUNCTION These switches provide opening and closing signals to the Reactor Building - torus vacuum breakers. These vacuum breakers are used to purge air from the Reactor Building into the torus to prevent exceeding the negative design pressure of the containment during an event.
EVALUATION For a break inside the containment, the environment in the Reactor Building is not expected to become harsher than that for which the component can be quallfled. Hence these components are expected to function. For breaks outside the containment, a harsh environment may cause these switches to fall. However, for breaks outside containment, it is not expected that there will be a need to open these vacuum breakers since containment spray would not normally be required. If these switches fail and they are needed, the operator can manually open and close the vacuum breakers from the control room. The failure of these switches will not mislead the operator since there are indications of vacuum breaker position in the control room.
CONCLUSION The failure of these switches will not prevent the operator from manually opening and closing the torus to Reactor Building vacuum breakers from the control room. The operator has indication of vacuum breaker position available so that he will not be misled.
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JCO-0C-85-16, Rev. 1=
.. January 31, 1985 Page 1 of.2 'l i
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JCO-0C-85-16, Rev.'l January 31, 1985 Page 2 of 2 COMPONENTS Tag Number VMS (Valve Monitoring System)
Description Monitors status of EMRVs and SVs OBJECTIVE The objective of this discussion is to determine that the failure of these components will not degrade other safety functions nor mislead the operator.
COMPONENT LOCATION The acoustic monitors for the VMS are located inside the drywell.
COMPONENT FUNCTION The VMS is used to detect flow through or leaks from an EMRV or SV. The sensing components are all located inside the drywell and hence are unaffected by a break outside the containment.
EVALUATION For breaks inside containment, the operator can determine from other parameters.(tail pipe temperatures, EMRV position indication) that an EMRV or SV is open in the event that the VMS fails. Following a LOCA, it is not likely that a SV will be required to function, therefore, there is no concern for a stuck SV. For a small break LOCA, ADS will actuate and open all EMRVs.
A stuck open EMRV will not be a concern under these conditions. For a large break LOCA, the system will be completely depressurized without EMRV actuation. Thus, the loss of the VMS has no safety significance for a break inside containment. For a break outside containment, the VMS is unaffected and will function normally. The failure of the VMS does not affect the normal operation of the EMRVs or the ADS.
CONCLUSION The failure of the VMS will not degrade any other plant safety function nor mislead the operator.
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JCO-0C-85-17, Rev. 1 January 31, 1985 Page'1_of 2 0YSTER' CREEK NUCLEAR GENERATING-STATION JUSTIFICATION FOR CONTINUED' OPERATION (JCO)
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JCO-0C-85-19, Rev. 1 January 31, 1985 Page 1 of 3 t
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l COMPONENTS Tag Numbers IP-18A, IP-188 Description- Containment Spray Pump Hi Temperature Trip "0BJECTIVE
- The objective of-this discussion is to determine:
that there is a basis for concluding that the components will perform their intended function; that failure of the components will not degrade other safety systems or mislead the operator.
i COMPONENT LOCATION These switches are located on the discharge piping of the containment spray heat exchangers at elevation 23'6" in the Reactor Building. IP-18A is in the northeast corner and IP-18B is in the southeast corner.
COMPONENT FUNCTION These components-function to trip the containment spray pumps upon receipt of a high temperature indication at the outlet of the containment spray heat exchangers.
j EVALUATION-U
-These components are located outside the drywell on the containment spray heat exchangers discharge piping. For a-break inside the containment, the l-environment in the Reactor Building is not expected to become harsher than
'that for'which the component can be qualified. Hence these components are-
< expected to function. Breaks outside' containment are assumed to isolate. - The only outside containment break which would require Containment Spray is.an
' event in which both Isolation Condensers were lost. Even under this remote
. scenario, Containment Spray would not be required to provide torus cooling for several hours. By that time, any appropriate manual actions could be taken.
The failure of the temperature switches due to the resulting environment will not cause an. unsafe condition.
These switches.do not-interface with or centrol any other components and their failure would not affect any other safety or accident mitigation systems.
A The failure of these switches.will not mislead the operator since there exist separate temperature elements for each system which will give indication of heat exchanger discharge temperature in the control room.
CONCLUSION
! These components will function during an inside containment LOCA when the l . Containment Spray System is required to operate. The failure of the components due to a harsh environment created by an outside containment break
- does not degrade any other safety systems nor will it mislead the operator.
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JCO-0C-85-20, Rev. 1 January 31, '1985 4 Page 1 of 4 s
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OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) 4 4
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FF JCD-0C-85-20, Rev. 1 January-31, 1985' Page 2 of 4 COMPONENTS.
Cable: :Rockbestos Firewall and Rockbestos EP DESCRIPTION Manufacturer: -The Rockbestos Company Model: Firewall EP Firewall III Function: Power Cable Control Cable Voltage: 600 Volts 600 Volts 4
Rating: 90*C 90*C Insulation: EPR Cross Linked Polyethylene
-Jacket: Hypalon Hypalon
'0BJECTIVE The objective of this engineering justification for continued operation is to demonstrate that (1) the Rockbestos cable will perform its safety functivil in the event of a design basis-accident at OCNGS, and (2) that the plant can be
- 1. safely. operated in the interim until the completion of the environmental qualification program by the Rockbestos Company.
T EQUIPMENT FUNCTION The cable connects nuclear safety-related equipment in.the plant. The most severe function with respect to loading and environment has been evaluated. A-combination of analysis and inadequately documented' tests indicates that cable is acceptable for the intended functions.
. EQUIPMENT LOCATION
'The cable is installed in the Reactor Building, including the drywell. High radiation areas which include the cleanup demineralizer room and cleanup Yaom do'not contain any nuclear safety related equipment.
EVALUATION r e
An evaluation of current test information and analyses of the Rockbestos cable
- Indicates that full qualification is not demonstrated; however, partial test '
L
., data does provide a basis'for concluding the cable will perform its function.
These findings are based upon the'results of both the NRC audit of Rockbestos (Reference 6) and'the GPUN audit accomplished August 21-23, 1984 (Reference 1).
GPUN has concluded, as has the NRC staff (Reference 3), that "at this time no
! lmmediate safety problem exists in the use of Rockbestos cables".
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JCO-0C-85-20, Rev. 1 January 31, 1985 Page 3 of 4 It was the intent of Rockbestos to conduct tests in accordance with IEEE Standards 323-1974 and 383-1974. Inadequate traceability, inadequate documentation and the general poor auditability of the supporting documentation does not conform with NRC requirements. The responses on the part of Rockbestos provides limited support, however, for partial test data. Rockbestos tests and analyses on Firewall EP cable (Reference 5) demonstrated a thermal qualified life of 40 years at 90*C and radiation tolerance of 2 x 10' rads, franklin Research Center tests (Reference 10) on Firewall III with a Neoprene jacket instead of Hypalon, indicated acceptability of the insulation. Since Hypalon is superior to Neoprene for thermal qualified life and in light of the Rockbestos test results, it is anticipated that the Firewall III cable for nuclear scrvice is superior to that with Neoprene.
As a consequence of generic probisms with the Rockbestos test program for these Class 1E cables (Reference 3), Rockbestos has committed to conduct a new test program. This will be completed in July 1986. In the meantime, GPUN will verify the results of the Rockbestos supplemental program. Also GPUN conducts its surveillance on nuclear safety related equipment. The periodic review of cable performance will provide a measure of confidence of the performance function of the cable; any indications of degradation will be evaluated for its potential impact on the cable performance during and after the accident.
CONCLUSION GPUN concludes that the OCNGS can be safely operated pending completion of equipment qualification as required by 10CFR50.49, Section 1, and the NRC letter of May 25, 1984 " Request for Additional Information". This consideration includes, as appropriate, items 1 through 5 as follows:
Item 1 - Accomplishing the safety function by some designated alternative equipment if the principal equipment has not been demonstrated to be fully qualified.
This is not appropriate for the equipmert involved.
Item 2 - The validity of partial test data in support of the original qualification.
This is the basis for justification for continued operation as provided above.
Item 3 - Limited use of administrative controls over equipment that has not been demonstrated to be fully qualified.
This is not appropriate for the equipment involved.
Item 4 - Completion of the safety function prior to exposure to the accident environment resulting from a design basis event and ensuring that the subsequent failure of the equipment does not degrade any safety function or mislead the operator.
i JCO-0C-85-200 Rev. 1 January 31, 1985 Page 4 of 4 This is not appropriate for the equipment involved.
Item 5 - No significant degradation of any safety function or misleading information to the operator as a result of failure of equipment under the accident environment resulting from a design basis event.
This is not appropriate for the equipment involved.
Based upon the evaluation provided in Section 6, GPUN concludes that the cable is qualified to perform its safety function in the interim period before completion of the Rockbestos tests. No significant degradation of any safety function or misleading information to the operator is expected under the accident environment resulting from a design basis event.
REFERENCES
- 1. GPUN Memorandum QA-D/P-84-828 dated November 30, 1984 (Finding No. 3).
- 2. Oyster Creek Nuclear Generating Station - Environmental Qualification of Safety Related Equipment dated November 1, 1980.
- 3. IE Information Notice No. 84-44 dated June 8, 1984.
- 4. File EQ-0C-311, Revision 0, Rockbestos EP.
- 5. Rockbestos Report No. QR 1804 dated April 6, 1981.
- 6. NRC Trip Report - Audit of Rockbestos Company Qualification Documents, nated September 12, 1983.
- 7. GPUN - TOR 297, Revision 1 dated August 19, 1982.
- 8. Letter to Kearny (GPUN) from Littlehales (Rockbestos) dated June 28, 1984.
- 9. File EQ-0C-312, Revision 0, Rockbestos Firewall.
- 10. Franklin Research Center Final Report F-C3798 dated March 1974.
g-JCO-0C-85-21, Rev. 1 January 31, 1985 Page 1 of 2' i
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JC0-0C-85-21, Rev.1 January 31, 1985 Page 2 of 2 COMPONENTS Tag Numbers DC-2 Description 125V DC Power Supply OBJECTIVE The objective of this discussion is to determine that failure of this component will not degrade other safety systems nor mislead the operator.
COMPONENT' LOCATION.
This power supply is located in the northeast quadrant of the Reactor Building on elevation 75'3".
COMPONENT FUNCTION This component provides power to the vent valves and one of the condensate return valves for one of the isolation condensers.
EVALUATION Thu component sees a harsh environment for the isolation condenser break outside the drywell. It is likely that DC-2 satisfies its function before a harsh environment causes it to fail. If DC-2 supplies power to the affected isolation condenser and subsequently failed, it would not result in a degradation of the isolation function of the affected condenser because otner redundant isolation valves which are not powered from DC-2 will remain closed. If the power supply was associated with the intact condenser and subsequently failed, it would at worst cause the condenser to isolate. The operator would then make use of alternate dec4y heat removal systems.
' If DC-2 failed prior-to satisfying its function, decay heat removal would be
- through the EMRVs with RPV inventory makeup supplied by Core Spray. This mode could be sustained for a lengthy period due to the large heat capacity of the torus pool. The Containment Spray system could be subsequently used to provide torus pool heat removal. '
The operator would not be misled by the loss of DC-2 since the valve position indicating lights for the valves would inform him of the incorrect situation
- j. ;nd would allow him to take the appropriate actions.
, CONCLUSION The loss of DC-2 will not degrade other plant safety functions and will not provide misleading information to the operator.
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c-JCO-0C-85-22, Rev. 1 January 31, 1985 Page 1 of 2.
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JCO-0C-85-22 0 Rev. I January 31, 1985 Page 2 of 2 COMPONENTS Tag Numbers RE-05A, RE-058; RE-05/19A, RE-05/19B Description RPV Level Switches OBJECTIVE The objective of this discussion is to determine that the failure of these components will not degrade other safety functions or mislead the operator.
COMPONENT LOCATION RE-05A and RE-05/19A are located on rack RK01 in the northwest quadrant of the Reactor Building at elevation 51'3". RE-05B and RE-05/198 are located on rack RK02 in the southeast quadrant of the Reactor Building at elevation 51'3".
COMPONENT FUNCTION The function of the level switches is to generate a scram signal on low RPV level and a turbine trip on high RPV level. Further, RE-05/19A,8 provide YARHAY level indication in the control room.
EVALUATION For breaks inside the contali. ment these switches will perform their safety function as required because, tbr breaks inside the containment, the environment in the Reactor Bulldt.'g is not expected to become harsher than that for which the component can be qualified. For large breaks outside containment, a low level scram would be expected to occur very quickly.
Redundancy exists via a low pres ure JSIV closure scram. Also, a Low-Low level MSIV closure scram would also occur if low level scram failed. For small breaks outside containment, there exist redundant scram signals or indications which would assure that the scram function is accomplished.
Further, the switches are located at different locations in the Reactor Building such that they will not see the same harsh conditions simultaneously. The operator has diverse level instrumentation in the control room so that the loss of one YARNAY indicator or even both would still not result in a misleading condition. However, only one indicator would be expected to fall from a single break location.
CONCLUSION The failure of these switches will not degrade the scram function which would occur by diverse means, or they may perform their scram function before the harsh environments occur. In addition, the operator can scram manually.
There is sufficient diversity of level instrumentation to prevent a misleading condition. Further, only one of the two YARHAY Indicators would be expected to fall from a single break location.
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JCO-0C-85-23, Rev. O January 31, 1985 Page 1 of 2 OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO)
JC0-0C-85-23, Rev. O January 31, 1985 Page 2 of 2 I
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JCO-0C-85-24, Rev. O February 11, 1985 Page 2 of 3 COMPONENTS Tag Numbers V-21-1, V-21-3, V-21-5, V-21-7, V-21-9, V-21-11, V-21-13 V-21-15, V-21-17, V-21-18 Description Local Key Lock Control Switches for Containment Spray System Valves V-21-1, 3, 7, 9 - Pump Suction Valves V-21-5, 11 - Drywell Injection Valves V-21-13, 17 - Dynamic Test Valves V-21-15, 18 - Torus Spray Inlet Valves OBJECTIVE The objective of this discussion is to determine:
that there is a basis for concluding that the existing components will perform their required function; that the failure of these components will not degrade any safety function nor mislead the operator.
COMPONENT LOCATION Switches for V-21-1, 3 and V-21-7, 9 are located in the southeast and northeast corner rooms in the Reactor Building on the -19'6" elevation respectively. The switches for valves V-21-5, 13, 15 are located in the southeast quadrant in the Reactor Building on elevation 23'6". The switches for valves V-21-17, 18 are located in the northeast quadrant on Reactor Building elevation 23'6". The switch for valve V-21-11 1s located in the northeast quadrant on Reactor Building elevation 51'3".
COMPONENT FUNCTION These components provide the operator with the ability to locally operate the identified Containment Spray System valves for surveillance / maintenance. They are not required to operate when the containment spray system is actuated.
EVALUATION These switches are located in the Reactor Building and are not subject to the harsh environment inside the drywell when they are required to be functional.
Contact radiation doses have been calculated for general areas inside the Reactor Building for breaks within the drywell. Radiation doses to specific target components are expected to be lower. GPUN is obtaining the specific radiation doses in the Reactor Building which are expected to be less than that for which the component can be quallfled. Hence these components are expected to function. There is a failure mechanism of the key lock control switch due to radiation which can cause the valve to change state. It will
JCO-0C-85-24, Rev. O February 11, 1985 Page 3 of 3 EVALUATION (Continued) take a considerable period of time for the integrated radiation dose to exceed the qualification value for the switch. A failure of the keylock switch at that time may cause some of the valves to reposition. If the Containment Spray system is lost as a result of this failure, the decay heat from the core would be sufficiently low so that the Containment Spray System would not be required to remove heat from the torus for some time. In addition, ambient losses from the torus shell may be sufficient to provide torus cooling.
Further, it is extremely unlikely that all of the valves will reposition themselves in the worst alignment at the same time so that there is no injection path to the containment. He believe that the likelihood of this one failure would be sufficiently low when consideration is given to the time the failure may occur, and other options available to the operating staff to accomplish the safety function.
Breaks outside containment are assumed to isolate. The only outside containment break which would require Containment Spray is an event in which both Isolation Condensers were lost. Even under this remote scenarlo, Containment Spray would not be required to provide torus cooling for several hours. The operator has position indication available in the control room so that he will not be misled.
CONCLUSION It is expected that these components will be quallfled for the environment.
If the analysis shows that the radiation at the component is less than the value at which it can be quallfled, an EQ file will be prepared. If the analysis shows that the radiation at the component is greater than the value for which it is qualified, corrective action will be taken. Based on the above evaluation, it is concluded that the Plant can be operated with no adverse affect to the health and safety of the public in the interim until qualification is documented or corrective action is taken.
JCO-0C-85-25, Rev. O January 31, 1985 Page 1 of 3 OYSTER CREEK NUCLEAR GENERATING' STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO)
JC0-0C-85-25, Rev. O January 31, 1985 Page 2 of 3 COMPONENTS Tag Numbers V-20-92, V-20-93, V-20-94, V-20-95 Description Limit Switches for Valves OBJECTIVE The objective of this discussion is to determine:
that there is a basis for concluding that the existing components will perform their required function; that the failure of these components will not degrade any safety function nor mislead the operator.
Col #0NENT LOCATION The switches for valves V-20-93, 95 are located in the southwest quadrant of the Reactor Building on elevation 23'6". The switches for valves V-20-92, 94 are located in the northwest quadrant of the Reactor Building on elevation 51 ' 3" .
COMPONENT FUNCTION These components provide the operator with indication of the position of the minimum flow valves which recirculate core spray flow back to the torus during
.the period when the pumps are running, but the RPV injection valves are closed.
EVALUATION.
These switches are located in the Reactor Building. For a break inside the containment, the environment in the Reactor Building is not expected to become harsher than that for which the component can be qualified. Thus, for these breaks, the Ifmit switches are expected to perfom their indicating function.
For breaks outside containment, these switches are located in different areas of the Reactor Building and will not see the harsh environment simultaneously. These valves fail to the required position upon loss of air or loss of power to the solenoid.
There exists a failure mechanism of the limit switch due to radiation which can cause the valve to change state. The consequences of this valve failure will drain the Core Spray System piping if the core spray pumps are not running. However, it will take a considerable period of time (approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) for the integrated radiation dose to exceed the qualification value for the switch. It is expected that if the Core Spray System is required to mitigate the event that it will be required to function prior to the radiation exceeding the qualification value for the switch. Since breaks outside containment are assumed to isolate, core spray would only be required to provide the initial level recovery. Long tem inventory makeup to accomplish decay heat removal could be provided by alternate systems. Once the core
JC0-0C-85-25, R3v. O January 31, 1985 Page 3 of 3 spray pumps have started, the failure of the limit switch has no effect. In addition, prior to exceeding the qualification value of the switch, it is expected that the RPV will be depressurized to a safe shutdown condition.
The failure of these limit switches does not affect any other safety function and will not lead the operator to take any unsafe actions.
CONCLUSION Tne-failure of these switches will not significantly degrade the effectiveness of the Core Spray System and will not cause the operator to take any unsafe actions.
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JCO-0C-85-26, Rev. O '
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JCO-0C-85-26, Rev. O January 31, 1985 Page 2 of 2 COMPONENTS Tag Numbers V-14-1, V-14-5, V-14-19, V-14-20 (Switch)
Description Position Switches for Isolation Condenser Vent Valves OBJECTIVE The objective of this discussion is to determine that the failure of these components will not degrade any safety function nor mislead the operator.
COMPONENT LOCATION These position switches are located in the Reactor Building at elevation 95'3", east. ,
COMPONENT FUNCTION These components function to energize status lights located in the control room to give indication of the position of the isolation condenser vent valves.
EVALUATION The position switches give status only of the isolation condenser vent valves and are not interlocked to or depended upon for any other safety function.
Since the vent valves will go to the required (i.e. closed) position upon initiation of the isolation condensers, there are no operator actions which require the status of the valves. In addition, a failure of the solenoid operated valves or of the air supply to the vent valves will cause the vents to go to the closed position. The failure of these switches may cause the condensate return valves to spuriously change position. Under the worst case, this would result in an isolation of both condensers. The operator would then make use of al. ternate decay heat removal systems. The failure of the position '
switches will not cause the operator to take any unsafe actions.
CONCLUSION The failure of the position switches will not degrade the safety function since the Indicated valves will go to the required position upon initiation of the isolation condensers or upon failure of the air supply or electric power supply. No other safety functions are interlocked with these switches, hence i their failure will not degrade any other safety functions. There are no operator actions contingent upon these status lights, and the failure will not cause the operator to tane any unsafe actions.
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JCO-0C-85-28, Rev. O January 31, 1985 i Page 1 of 2
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l JCO-0C-85-28, Rev. O January 31, 1985 i Page 2 of 2 COMPONENTS Tag Numbers V-27-1, V-27-2, V-27-3, V-27-4 Description Limit Switches for Drywell Vent and Purge Valves OBJECTIVE The objective of this discussion is to determine that the failure of these components will not degrade any safety function nor mislead the operator.
COMPONENT LOCATION The switches for valves V-27-1, 2 are located in the Torus Room of the Reactor Building at elevation -19'6". The switches for valves V-27-3, 4 are located in the northwest quadrant of the Reactor Building on elevation 75'3".
COMPONENT FUNCTION These switches provide the operator with indication of the position of the drywell vent and purge valves, and are required to be closed for containment isolation post-accident.
EVALUATION These limit switches only provide the operator with position indication for the vent and purge valves and are not interlocked with any other safety function. These valves revert to their required position on loss of power to their associated solenoid valves. These valves are equipped with accumulators so that on loss of air to the solenold, the valves will also revert to their required (closed) position. The failure of these switches may cause a short which would result in a loss of indication to a number of containment isolation valves which are only required to close. However, the loss of position indication wou?d not prevent operation of the valves. The operator would be aware of the los; of position indication by the loss of both Indicating lights for the valves. Since the valves fall in their required position, the loss of the limit switches would not mislead the operator into taking any unsafe action.
CONCLUSION The failure of the 11mit switches for the identified valves will not degrade any safety function since these switches are for indication only. Also, tne valves fall in their required position. The operator will not be misled by the fallure of these switches.
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JCO-0C-85-29, Rev. O January 31, 1985 Page 1 of 2-i OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) 4 9
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JCO-0C-85-29, Rev. O January 31, 1985 Page 2 of 2 COMPONENTS Tag Numbers SS, 6K37X, 6K46X Description Valve Control Switch for CPM (SS)
Relays (6K37X, 6K46X)
OBJECTIVE L
The objective of this discussion is to determine:
that there is a basis for concluding that the existing components will perform their required function; that the failure of these components will not degrade any safety function nor mislead the operator.
COMPONENT LOCATION All of the components identified above are located in the northeast quadrant of the Reactor Building on elevation 23'6".
COMPONENT FUNCTION These components are utilized as part of the Containment Particulate Monitoring System to detect the concentrations of gas within the containment for breaks within the containment.
t EVALUATION These components are located in the Reactor Building and thus are not subject to the harsh environment inside the drywell when they are required to function.
For a break inside the containment, the environment in the Reactor Building is h not expected to become harsher than that for which the component can be qualified. Hence, these components are expected to function. For breaks outside containment, the Containment Particulate Monitoring System is not required to function to mitigate the event. Thus, the failure of these components would not cause the operator to take any unsafe actions.
The possibility exists that these control switches may cause containment isolation valves V-38-9, 10, 16 or 17 not to close. However, if this occurred, the containment barrier would still be maintained; a closed loop system of small diameter instrument piping would result which would exit and return to the containment through the Containment Particulate Monitor. There would be no radiation release to the environment or significant radiation exposure due to the instrument piping loop.
These components are not interlocked with any other safety function and thus, their failure would not prevent any other safety related actuation.
CONCLUSION The failure of the components will not degrade the function of the Containment Particulate Monitoring System nor be misleading to the operator.
r- JCO-0C-85-30, Rev.10 ~
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JCO-0C-85-30, Rev. O January 31, 1985 Page 2 of 2 COMPONENTS Tag Numbers V-23-13, V-23-14, V-23-15, V-23-16, V-23-17, V-23-18, V-23-21, V-23-22 Description Limit Switches for Drywell Nitrogen Purge & Relief Valves OBJECTIVE The objective of this discussion is to determine that the failure of these components will not degrade any safety function nor mislead the operator.
COMPONENT LOCATION The switches for V-23-13, 14 and V-23-17, 18 are located in the northeast and southwest quadrants of the Reactor Building on the 75'3" elevation respectively. The switches for V-23-15, 16 are located in the southwest quadrant of the Reactor Building on elevation 23'6". The switches for V-23-21, 22 are located in the Torus Room of the Reactor Building on elevation
-19'6".
COMPONENT FUNCTION These switches provide the operator with indication of the position of the drywell nitrogen purge and makeup valves (V-23-13, 14, 15, 16, 17, 18) and the drywell nitrogen relief valves (V-23-21, 22).
EVALUATION These limit switches only provide the operator with position indication of the purge and relief valves and are not interlocked with any other safety function.
.These valves revert to their required closed position as a result of a loss of power or air to their associated solenoid valves. The failure of these switches may cause a short which could result in a loss of position indication for a number of containment isolation valves which are only required to go to the closed position. However, the loss of the position indication would not prevent the operation of the valves. The operator would be aware of the loss of position indication by the loss of both indicating lights for the valves.
Since the valves fall to their required position, the loss of the limit switches would not cause the operator to take any unsafe actions.
CONCLUSION The failure of the limit switches for the identified valves will not degrade any safety function since these switches are for indication only. Also, the valves fail in their required position. The operator will not be misled by the failure.of these switches.
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'JCO-0C-85-31, Rev. 0-January.31, 1985 Page 1 of 2- 1 6
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JC0-0C-85-31, Rev. O j January 31, 1985 l Page 2 of 2 j COMPONENTS l I
Tag Numbers V-ll-34, V-ll-36 l Description Limit Switches for Valve Position Indication {
OBJECTIVE The objective of this discussion is to determine that the failure of these components will not degrade any safety function nor mislead the operator.
COMPONENT LOCATION These components are located in the northeast quadrant of the Reactor Building at elevation 95'3".
COMPONENT FUNCTION These components energize the position indicating lights for the valves supplying makeup water to the isolation condensers from the Condensate Transfer System.
EVALUATION The isolation condensers can operate for at least 45 minutes each without makeup to the secondary side. Addition of water to the condenser shells is a manual action and requires that the makeup valves (V-11-34, V-11-36) be open.
Failure of the position indicators would not prohibit makeup since the failure of the limit switch does not affect valve position. Failure of the indicators would not mislead the operator into taking an unsafe action since there are other methods to determine if there is sufficient heat removal with the isolation condensers (i.e. shell side level, reactor pressure). Failure to
' makeup to the' isolation condensers will result in a reduction of heat transfer and repressurization of the reactor. The operator is instructed by the Emergency Procedures to augment depressurization through the use of the EMRV, if the Isolation Condenser System is shown to be inadequate.
CONCLUSION The failure of th limit switches to indicate valve position will not degrade any safety functions since these switches do not affect valve position. The operator will not be misled into taking actions which are unsafe since there are methods to determine the effectiveness of the Isolation Condenser System and actions can be taken to augment reactor pressure control.
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.JCO-0C-85-320 Rev. O January 31, 1985 Page 1 of 7 L
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JCO-0C-85-32, Rev. O January 31, 1985 Page 2 of 2 COMPONENTS Tag Numbers V-28-47, V-28-17, V-28-18 Description Limit Switches for Torus Ventilation Valves OBJECTIVE The objective of this discussion is to determine that the failure of these components will not degrade any safety function nor mislead the operator.
COMPONENT LOCATION These limit switches are located in the Torus Room in the Reactor Building on elevation -19'6".
COMPONENT FUNCTION These limit switches provide the operator with indication of the position of the torus exhaust ventilation valves.
EVALUATION For a break inside the containment, the environment to which these components are subjected is not expected to become harsher than that for which the component can be qualified. Hence these components are expected to function.
Should the limit switch fall as a result of a harsh environment from a break inside the containment, it will not cause the valve to reposition itself.
Thus, the isolation function is preserved. For breaks outside the drywell, these switches may see a harsh environment. However, these valves are normally closed and are required to remain closed during and following an event. The failure of these switches may cause a short which would result in a loss of indication to a number of containment isolation valves which are only required to go closed. These valves are normally closed so that the operator will not be misled by the potential loss of indication. These limit switches are not interlocked with any other safety function and thus, the failure will not affect any safety actuation.
CONCLUSION These limit switches are only used for position indication by the operator.
Their associated valves are normally closed and are required to be closed following the event. The operator will not be misled by the failure of these switches.
JCO-0C-85-330 .Rev. O
~ January _31, 1985 Page 1 of 2 OYSTER' CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO)
JC0-0C-85-33, Rev. O January 31, 1985 Page 2 of 2 COMPONENTS Tag Numbers V-28-9, V-28-10 V-28-11, V-28-12, V-28-13, V-28-14, V-28-15, V-28-16 Description Lipit Switches for Reactor Building HVAC Isolation Valves OBJECTIVE The objective of this discussion is to determine: l that there is a Lasts for concluding that the existing components will perform their required function; that the failure of these components will not degrade any safety function nor mislead the operator.
COMPONENT LOCATION These switches for V-28-9, 10 and V-28-11, 12 are located in the southwest and northwest quadrants of the Reactor Buildin'g on elevation 75'3" respectively.
The switches for V-28-13,14 and V-28-15,16 are located in the southwest and northwest quadrants of the Reactor Building on elevation 51'3" respectively.
COMPONENT FUNCTION These switches provide the operator with the position status of the Reactor Building HVAC isolation valves.
EVALUATION These limit switches are located in the Reactor Building and thus are not subject to the harsh environment inside the drywell when they are required to function. For a break inside the containment, the environment in the Reactor Building is not expected to become harsher than that for which the component can be qualified. Hence, these components are expected to function. For breaks outside containment, these valves would close quickly on a secondary containment isolation signal before the environment became harsh. Thus, the limit switch would perform its indicating function to the operator. Once the valves are closed, there is no spurious mechanism for re-opening these valves. Then, the failure of the limit switches would not cause the operator to take any incorrect action since he had previously confirmed that the valves were closed. In addition, the valves fall to their required closed position on loss of air or loss of power to their solenoid valves.
CONCLUSION The failure of these limit switches would not degrade the ability to isolate the secondary containment nor any other safety function and would not mislead the operator into taking any unsafe actions.
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January 31, 1985 Page 2 of 2 COMPONENTS Tag Numbers TE-109A, TE-1098, TE-109C, TE-109D Description Torus Water Temperature Elements OBJECTIVE The objective of this discussion is to determine:
that there is a basis for concluding that the existing components will perform their requirea function; that the failure of these components will not degrade any safety function nor mislead the operator.
COMPONENT LOCATION These temperature elements are located in thermowells in the torus shell in the Torus Room of the Reactor Building on elevation -19'6".
COMPONENT FUNCTION These temperature elements provide the operator with indication of the temperature in the torus pool. There are no automatic actuations which occur based on these temperature elements.
EVALUATION For a break inside the containment, the environment to which these components are subjected is not expected to become harsher than that for which the component can be qualified. The integrated radiation dose for these components due to a break inside containment is expected to take a considerable period of time to reach the qualification limit for these temperature elements. It is expected that these components will be required to function for a fairly short period of time (about one week), and that by that time, the RPV will be depressurized and in a safe shutdown condition. In addition, if these temperature elements fall, the operator has alternate indications by which he can estimate torus pool temperature so that he will not be misled.
CONCLUSION There are no automatic safety actuations based on these temperature elements.
Thus, the operator will not be misled if these temperature elements fall due to a harsh environment, since there are alternate indications available.
JCO-0C-85-35, Rev. O January 31, 1985 Page 1 of 2 i
l 0YSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO)
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JCO-0C-852 36, Rev. O January 31, 1985 Page 1 of 2 OYSTER CREEK *iUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) '
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JC0-0C-85-36, Rev. O January 31, 1985 Page 2 of 2 COMPONENTS Tag Numbers V-28-21, V-28-22 Description Limit Switches for Reactor Building Ventilation Exhaust Valves OBJECTIVE The objective of this discussion is to determine:
that~there is a basis for concluding that the existing components will perform their required function; that the failure of these components will not degrade any safety function nor mislead the operator.
COMPONENT LOCATION These switches are located in the pipe tunnel which connects the Turbine Building to the base of the stack.
COMPONENT FUNCTION These switches provide the operator with the position status of the Reactor Building HVAC exhaust valves.
EVALUATION These limit switches are located in the pipe tunnel and thus are not subject to the harsh environment inside the drywell when they are required to function.
For a break inside the containment, the environment in the pipe tunnel is not expected to become harsher than that for which the component can be qualified. Hence these switches are expected to function. In addition, for breaks inside the containment these valves would close quickly on a containment isolation signal. For breaks outside containment, these valves would also close quickly on a secondary containment isolation signal before the environment becomes harsh. Thus, the limit switch would perform its indicating functicn to the operator. The failure of the limit switches would not cause the operator to take any incorrect actions since he had previously confirmed that the valves were closed. In addition, these valves fail to their required closed position on loss of air or loss of power to their solenoid valves.
CONCLUSION The failure of these limit switches would not degrade the ability to isolate the secondary containr!.ent nor any other safety function and would not mislead the operator into taking any unsafe actions.
JCO-0C-85-37, Rev. O January 31, 1985 Page 1 of 2
' OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO)
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JC0-SC-85-37, Ray. O January 31, 1985 Page 2 of 2 COMPONENTS Tag Numbers IA-90A, IA-90B; IA-91 A, IA-91B; IA-92A, IA-928; TE-57-2A, TE-59-28 Description dP Transmitters for RPV Fuel Zone Level (IA-90A, IA-908; IA-91 A, IA-91B; IA-92A, IA-928)
Temperature Elements for RPV Fuel Zone Level (TE-57-2A, TE-59-28)
OBJECTIVE <
The objective of this discussion is to determine that the failure of these components will not degrade any safety function nor mislead the operator.
_ COMPONENT LOCATION The dP transmitters IA-90A,B, IA-91 A,B and IA-92A,B and temperature elements TE-57-2A and TE-59-28 are located in the northwest quadrant of the Reactor Building on elevation 51'3".
C0WONENT FUNCTION The dP transmitters drive the fuel zone level indicators which provide the operator with RPV water level monitoring. The temperature elements provide input to the.dP transmitters to compensate the indications for changes in drywell and Reactor Building temperatures.
EVALUATION The transmitters are located in the Reactor Building and thus are not subject to a harsh environment for breaks inside the drywell. For a break inside the containment, the environment in the Reactor Building is not expected to become harsher than that for which the component can be qualified. For breaks outside the containment, the loss of these transmitters does not prevent any plant safety actuations from occurring, since no safety related trips are based on the fuel zone transmitters. If the failure of the transmitters or temperature elements resulted in conflicting RPV level indication to the operator, he would make use of other level indicators available in the control room. Even in the event that the operator judges that he cannot determine the RPV water level, the Emergency Operating Procedures provide the operator with guidance to. prevent uncovering the core.
CONCLUSION
--The failure of these transmitters and temperature elements does not significantly degrade the ability of the operator to monitor the RPV water level. There are no automatic safety related actuations which are based on the fuel zone transmitters. The E0Ps provide tne operator with guidance if he cannot determine RPV level so that he will not be misled into taking any unsafe actions.
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JCO-0C-85-39 0 Rev. 0 January 31, 1985
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4 JC0-0C-85-39 Rev. 0 January 31, 1905 Page 2 of 2 C0W ONENT Tag Numbers: TIP Ball Valves
Description:
Solenoid Operated Valves OBJECTIVE The objective of this discussion is to detennine that there is a basis for concluding that the required function will be satisfied by other~ equipment.
COMPONENT LOCATION
- The four (4) TIP ball valves are located outside the drywell in the northwest quadrant of the Reactor Building at elevation 33'5".
COMPONENT FUNCTION The ball valves close to isolate the drywell upon a reactor protection signal.
l EVALUATION l
The ball valves are the primary means of isolating the TIP system penetrations to the containment. These valves are open in order to insert the traversing incore probes for calibration of incore monitors. The ball valves are on small diameter instrument tubing. They are fail-safe, i.e. , upon loss of power they automatically close and are spring-loaded to close. They are normally closed and activated only during a calibration check with the system on-li ne. Normal operation is to run the TIP in one at a time; so that only one ball valve is open at a given time. The system is designed so that upon an isolation signal, the TIP detectors are withdrawn, and the ball valves go to the closed position. The detectors are quickly withdrawn in this situation, so that the ball valve is required to operate in a harsh environment for only a short period of time while isolating. Should the power be lost with the TIP in, the ball' valve will de-energize and try to clo:e. The incomplete closure would be ninimal (perhaps less than 10% of the 0.273" 1.D.). This would not result in a significant release of radioactivity into the Reactor Building.
Incomplete closure of the ball valve will result in a signal that the ball valve is.de-energized. In addition, there exist four (4) shear valves (one per guide tube) which would be electrically fired manually with a key switen from the control room (redundant circuits), and would cut and seal the TIP tubes for containment isolation. The shear valves are primarily mechanical components and are thus not subject to the normal failure modes of electrical equipment. Therefore, capability to isolate the drywell is assured.
CONCLUSION A The ball valves are located in small diameter instrument tubing and are of a fail-safe design. Therefore, the radioactivity released would be minimal should these valves fail to close completely. The failure of these components does not affect containment isolation since this function can be satisfied by other redundant equipment. ]
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' COMPONENTS Tag Numbers V-17-55, V-17-56, V-17-57 Descr, lotion Limit Switches for Shutdown Cooling Loop Isolation Valves OBJECTIVE The objective of this discussion is to determine that the failure of these components will not degrade any safety function nor mislead the operator.
COMPONENT LOCATION 1
These components are located in the Shutdown Cooling Room in the Reactor Building on elevation 51'3".
COMPONENT FUNCTION These limit switches provide the operator with position indication for the outlet isolation valves for each of the three shutdown cooling loops.
. EVALUATION The Shutdown Cooling System is not required to operate to bring the plant to a safe shutdown condition. The above valves are normally closed and are required
'to remain closed during and after the event. Thus, these valves do not have to change state in order to satisfy their containment isolation function. The Shutdown Cooling System is a closed loop system outilde of containment. The failure of the limit switches would not cause the valves to change state nor degrade any safety system. These valves do not open automatically. The operator must take manual action to open these valves. Thus, the failure of the switches would not mislead the operator.
CONCLUSION The failure of these limit switches will not degrade the effectiveness of any safety system nor mislead the operator.
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JC0-0C-85-41, Rsv. 0 January 31, 1985 j Page 2 of 2 l COMPONENTS I Tag Numbers V-21 -1, Y-21 -3, Y-21 -7, Y-21 -9 Description Limit Switches for Containment Spray Suction Valves OBJECTIVE The objective of this discussion is to determine that the failure of these components do not degrade any safety functions and will not mislead the operator.
COMPONENT LOCATION s
These components are located on the motor operators. V-21-1 and V-21-3 are located in the southeast corner room of the Reactor Building at elevation
-19'6"; Y-21-7 and V-21-9 are located in the northeast corner room at elevation -19'6".
COMPONENT FUNCTION These limit switches provide the operator with position indication for the suction valves for each of the containment spray pumps.
EVALUATION These valves are normally open and are required to remain open in order for the Containment Spray System to function. Failure of the limit switches would not cause the valves to change position; the valves do not function automatically but require operator action to change position. Since manual action is required to close the valves, failure of the limit switches will not mislead the operator.
CONCLUSION Failure of these switches will not degrade any safety functions nor mislead the operator since the valves cannot change position without operator action.
dCO-0C-85-42, Rev. 0
. January 31, 1985 Page 1 of 2 0YSTER CREEK NUCLEAR GENERATING STATION
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JCO-0C-85-42, Rev. O January 31, 1985 Page 2 of 2 COMPONENTS Tag Numbers V-20-3, V-20-4, V-20-32, V-20-33 Description Limit Switches for Core Spray Suction Valves OBJECTIVE The objective of this discussion is to determine that the failure of these components does not degrade any safety functions and will not mislead the operator.
COMPONENT LOCATION These components are located on the motor operators. V-20-3 and V-20-32 are located in the northwest corner room of the Reactor Building at elevation
-19'6"; V-20-4 and V-20-33 are located in the southwest corner room at elevation -19'6".
COMPONENT FUNCTION These limit switches provide the operator with position indication for the suction valves for each of the main core spray pumps.
< EVALUATION These valves are normally open and are required to remain open in order for
. the Core Spray System to function. Failure of the limit switches would not cause the valves to change position; the valves do not function automatically, but require operator action to change position. Since manual action is required to close the valves, failure of the limit switches will not mislead the operator.
CONCLUSION Failure of these switches will not degrade any safety functions nor mislead the operator since the valves cannot change position without operator action.
JCO-0C-85-43, Rev. 0 January 31 0 1985' Page 1 of 2 i
OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) il i
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JCO-0C-85-43, Rev. O February 11, 1985 Page 2 of 3 COMPONENTS Tag Numbers V-3-87, V-3-88 Description Key Lock Control Switches for ESH Valves OBJECTIVE The objective of this discussion is to determine:
that there is a basis for concluding that the existing components will perform their required function; that the failure of these components will not degrade any safety function nor mislead the operator.
COMPONENT LOCATION Switches are located in the Reactor Building at elevation 23'6"; the switch for V-3-87 is in the northeast quadrant; the switch for V-3-88 is in the southeast quadrant.
COMPONENT FUNCTION These components provide the ability to locally operate the subject valves for surveillance / maintenance. They are not required to operate when the containment spray system is actuated.
EVALUATION These' switches are located in the Reactor Building and are not subject to the harsh environment inside the drywell when they are required to be functional.
Contact radiation doses have been calculated for general areas inside the Reactor Building for breaks within the drywell. Radiation doses to specific target components are expected to be lower. GPUN is' obtaining the specific radiation doses in the Reactor. Building which are expected to be less than that for which the component can be qualified. Hence these components are expected to function. There is a failure mechanism of the key lock control switch due to radiation which can cause the valve to change state. It will take a considerable period of time for the integrated radiation dose to exceed the qualification value for the switch. A failure of the keylock switch at that time may cause some of the valves to reposition. If the Containment
. Spray system is lost as a result of this failure, the decay heat from the core would be sufficiently low so that the Containment Spray System would not be required to remove heat from the torus for some time. In addition, ambient losses from the torus shell may be sufficient to provide torus cooling.
Further, it is extremely unlikely that all of the valves will reposition
.themselves in the worst alignment at the same time so that there is no injection path to the containment.
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JC0-0C-85-43, Rev. O February 11, 1985 Page 3 of 3 EVALUATION (Continued)
He believe that the likelihood of this one failure would be sufficiently low when consideration is given to the time the failure may occur, and other options available to the operating staff to accomplish the safety function.
Breaks outside containment are assumed to isolate. The only outside containment break which would require Containment Spray is an event in which both Isolation Condensers were lost. Even under this remote scenario, Containment Spray would not be required to provide torus cooling for several hours. -The operator has position indication available in the control room so that he will not be misled.
CONCLUSION It is expected that these components will be qualified for the environment.
If the analysis shows that the radiation at the component is less than the value at which it can be qualified, an EQ file will be prepared. If the analysis shows that the radiation at the component is greater than the value for which it is quallfled, corrective action will be taken. Based on the above evaluation, it is concluded that the Plant can be operated with no adverse affect to the health and safety of the public in the interim intil qualification is documented or corrective action is taken.
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JCO-0C-85-44 Rev. O January 31, 1985 Page 1 of 2 OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO) k f
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p JCO-0C-85-44, Rev. O January 31, 1985 Page 2 of 2 COMPONENTS Tag Numbers. V-1-106, V-1-107, V-1-110, V-1-111 Description Limit Switches for Main Steam Line Drain Valves ,
OBJECTIVE The objective of this discussion is to determine that the failure of these components will not degrade any safety function nor mislead the operator.
COMPONENT LOCATION The limit switches for valves V-1-106, 107 are located inside the drywell, and the limit switches for V-1-110,111 are located in the steam line tunnel in the Reactor Building on elevation 23'6".
COMPONENT FUNCTION These limit switches provide the operator with position status for each of the four drain valves listed above.
EVALUATION The above valves are'normally closed and are required to remain closed during and after an event. Thus, these valves do not have to change state in order to satisfy their containment isolation function. These valves do not open automatically. However, a failure of the limit switch could cause the valve to reposition itself spuriously. In that event, the operator will be able to detect the blown r,ower fuse by the indicating lights in the control room.
Normally, the operator must take manual action to open these valves. Thus, the failure of the switches would not mislead the operator.
In addition, these valves are paired in series so that the outside containment valve is expected to function for a break inside containment and vice versa.
Thus, the containment isolation function is preserved regardless of break location.
CONCLUSION The failure of these limit switches will not degrade any safety system nor mislead the operator.
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A 1 JCO-0C-85-45, Rev. O January 31, 1985
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JC0-0C-85-45, Rev. O January 31, 1985 Page 2 of 2 COMPONENTS Tag Number V-31-2 Description Limit Switch for Reactor Head Cooling Isolation Valve OBJECTIVE
.he objective of this discussion is to determine that the failure of this component will not degrade any safety function nor mislead the operator.
COMPONENT LOCATION This limit switch is located in the northeast quadrant of the Reactor Building on elevation 75'3".
COMPONENT FUNCTION This limit switch provides the operator with position indication for the isolation valve on the Reactor Head Cooling System.
EVALUATION
~The Reactor Head Cooling System is not required to operate to bring the plant to a safe shutdown condition. Valve V-31-2 is normally closed and is required to remain closed during_and after the event. Thus, this valve does not have to change state in order to satisfy its containment isolation function. The failure of the limit switch would not cause the valve to change state nor degrade any safety system. This valve does not open automatically. The operator must take manual action to open this valve. Thus, the failure of the switch would not mislead the operator.
CONCLUSION The failure of this limit switch will not degrade the effectiveness of any safety safety nor mislead the operator.
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JCO-0C-85-46, Rev. O January 31, 1985 Page 1 of 2 OYSTER CREEK MUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO)
JCO-0C-85-46, Rev. O January 31, 1985 Page 2 of 2 INTENTIONALLY LEFT BLANK
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JCO-0C-85-470 Rev. O February 8,_1985 Page 1 of 3 l
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JCO-0C-85-47 Rev. O February 8, 1985 Page 2 of 3 COMPONENTS Tag Nos. 1A, IB, IC,-ID
Description:
4160 V Swltchgear Equipment OBJECTIVE The objective of this discussion is to justify interm operation of the Oyster Creek facility in accordance with the Federal Code of Regulations 10CFR50.49 section (1)(2).
COMPONENT LOCATION Turbine Building
- ENVIRONMENT Temperature 83*F for 40 years Radiation 3.5 X 10' Rad TID Accident condition Rise from 102*F to 172*F in 15 seconds Fall to 86*F 340 seconds after time zero Relative humidity 100%
EVALVATION Oyster Creek Turbine Butiding environmental parameters indicate that on the MSLB in the turbine building a peak temperature of 172*F is reached in 15 seconds and that the area temperature of 93*F is achieved after approximately I hr.
A simple lumped mass analysis of the switchgear units to determine their response to the temperature transient indicates that the unenergized units will not exceed an internal component temperature of 120*F (50*C) throughout the HSLB scenario. Energized units carrying 80% of the rated load will experience a 45*C temperature rise which will not exceed the ANSI C37.04-31 Ilmit of 105*C for continuous operation.
ANSI Standard C37.04-31 provides guidelines for the design of switchgear equipment and states that this equipment should have a maximum heat rise of 65'C above ambient and should therefore have components designed for 105*C (40*C ambient + 65'C heat rise).
Conservatism in switchgear design in conjunction with the actual loading of the switchgear in this scenario of 80% (max.) of nameplate capacity will assure that the switchgear will not be thermally overloaded during tl.e 10*C expected excursion.
This can be demonstrated by referring to the General Electric Co. "Switchgear Application Handbook" #53-02 page 1 issue 3 dated 8-15-72. This paragraph describes the technique used for de-rating switchgear equipment for use in high temperature (greater than 40*C environment).
Use of this technique gives an 8% de-rating factor for application at 50*C ambient.
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JCO-0C-85-47 Rev. O February 8, 1985 Page 3 of 3 4160 V Switchgear (continued)
ANSI Standard C37.20 requires switchgear designs to be subjected to the following prototype and production tests:
Basic Impulse Level Testing - 60,000 KV Dielectric Hfthstand Test 19 KV (production test)
It is expected that the switchgear units will pass these tests regardless of ambient relative humidity.
Plate out of condensate on current carrying conductors and insulating materials should be considered here. It is expected that the enclosed energized switchgear units will have sufficient self heating to keep local relative humidity below 100%. Unenergized units necessary for engineered safeguard loads should be kept clean to avoid short circuit conduction paths in the event some condensation occurs. A maintenance / surveillance program can assure that conductors and insulators are cleaned on a regular basis.
CONCLUSION Continued Operation is justified based on the above evaluation.
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JCO-0C-85-48, Rev. 0~
February 8,' 1985 Page 1 of'3 OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUE 0 OPERATION (JCO) i
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'JCO-0C-85-48 Rev. O s' February 8, 1985 Page 2 of 3
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COMPONENTS:
Tag Nos. IAl, 1B1 Description Unit Substations Unit substations lAl and 181 are comprised of the following i
o L4160/460 3 phase step down sub-station transformer o AKD-5.switchgear OBJECTIVE The objective of this discussion is to justify interim operation of the Oyster Creek facility in accordance with the Federal Code of Regulations 10 CFR50.49 section (1)(2).
COMPONENT LOCATION ,
'These units are located'in the turbine building basement columns B-3 and E-3 !
respectively.
ENVIRONMENT-
-Temperature _ 85'F for 40 years i Radiation. 3.5 X 108 Rad TID '
Accident condition Rise from 102*F to 180*F in 25 seconds Fall to 93*F 9000 seconds after time zero Relative humidity 100%
1 EVALUATION n
The sub-station transformers are sealed oil bath units with external heat exchangers. Based on the accident environmental data (ie. area temperature return to 93*F in approx. 2.5 hr. with peak temperature of 180*F 9 t - 25 sec) and a high transformer thermal time constant ~(large mass), it is believed that the transformer coil temperature will undergo an insignificant heat rise during a MSLB in the turbine building.
The AKD-5 switchgear units approximately 10 per sub-station have had .
environmental qualification avaluations performed. Literature search has
- determined that switchgear similar to the Oyster Creek units (AKD-5s) have '
component qualification documentation to support an extended qualified life
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for this equipment. The manufacturer recommends a maintenance replacement i program for the parts identified as age susceptible or having defined life :
endpoints (le. 10 yrs; 1000 cycles of operation, etc.). . The equipment was !
evaluated against the following environmental service conditions:
Qualification Level Ambient temp. 120*F Relative Humidity 90% R.H.
Radiation i 10* Rads 1
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r JCO-0C-85-48 Ray. 0 Unit Substation (continued)
Oyster Creek Turbine Building environmental parameters indicate that on the MSLB in the turbine. building a peak temperature of 180*F is reached in 25
-seconds and that the area temperature of 93*F is achieved in approx. 2.5 hr.
A simple lumped mass analysis of the switchgear units to determine their
' response to the temperature transient indicates that the unenergized units will not exceed an internal component temperature of 120*F (50*C) throughout the MSLB scenario. Energized units carrying 80% of the rated load will experience a 45*C temperature rise which will not exceed the ANSI C37.04-31
-limit of 105'C for continuous operation.
ANSI Standard C37.04-31 provides guidelines for the design of switchgear equipment and states that this equipment should have a maximum heat rise of 65 C above ambient and should therefore have components designed for 105*C (40*C ambient + 65*C heat rise).
Conservatism in AKD switchgear design in conjunction with the actual loading of the switchgear in this scenario, 80% (max.) of nameplate capacity, will assure that the switchgear will not be therma]1y overloaded during the 10*C expected excursion.
This can be demonstrated by referring to General Electric Co. "Switchgear Application Handbook" #53-02 page 1 issue 3 dated 8-15-72. This paragraph describes the technique used for de-rating switchgear equipment for use in high temperature (greater than 40*C environments).
Use of this technique gives an 8% de-rating factor for application at 50*C ambient.
Based on the above analysis, the only remaining concerns would be radiation and humidity exposure. Radiation dose is 4109 rads T.1.D. and therefore considered insignificant.
Humidity tests have been perfomed on GE AK-2-25 600 V, 600A circuit breakers carrying 460 VAC with exposure to 100% r.h., after which the equipment functioned properly. Similar data can be found on the other components commonly used in AKD-5s such as EB-5 and E8-25 terminal blocks, bus bar insulators, transfomers, roof entrance bushings and certain types of control wiring.
CONCLUSIONS:
Continued operation is justified for this equipment based on the following.
, 8% derating required for application at 50*C ambient vs. 20%
minimum conservatism factor of component and system design.
! Existence of thermal aging documentation.
Test data that indicates exposure to extremes of temperature and l
humidity does not effect AK-25 breakers.
Maintenance / surveillance program provides assurance that units are i clean and operate properly and within limits of calibration.
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JCO-0C-85-49, Rev. O February 8, 1985 Page 1 of 3 V
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JCO-0C-85-49 Rev. O February 8, 1985 Page 2 of 3 COMPONENTS Tag number (s): STD - Static' Time Delay. unit for: (1) USSIA1BKRO11B, (2)
USSIBIBKR021B
Description:
Static Time Delay unit model number TAKYUVT-3.
OBJECTIVE-The objective of this discussion is to justify the continued operation of the
' Oyster Creek Facility by demonstrating that the safety function can be accomplished by the Static Time Delay-unit in the specified environment.
- COMPONENT LOCATION Turbine-building at elevation 3.6'.
ENVIRONMENT Temperature. 85'F for 40 years Radiation l3.5 X 108 TID Accident condition Rise from 102*F to 180*F in 25 seconds Fall to 93*F 9000 seconds after time zero Relative humidity 100%
COMPONENT FUNCTION :
The Static Time Delay unit is used in conjunction with an Undervoltage (UV)
Trip device. The UV device protects against harmful drops in line voltage by automatically tripping the substation breaker. The' Static Time Delay unit provides a field adjustable delay between under-voltage fault and breaker trip to prevent nuisance tripping due to momentary loss of voltage.
EVALUATION The evaluation is based on the results of qualification tests performed on other static devices (example General Electric SFF relays) that contained a comparable assembly of. solid state components ~and non-metaille materials
-(1.e., stellar materials of construction). This testing included:
- 1. Exposure to 95% R.H. at 140*F for-a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with a-safety function check in the test environment at the end of the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> test period.
- 2. Exposure to 1 x 10 5rads gamma
- 3. - Thermal aging (unenergized) in a 212*F temperature environment for a period of three (3) months. Based on Arrhenius calculations, this test produced thermal aging of the critical non-metallic materials equivalent to 104 *F for a period of 40 years.
' Ability to perform the safety function was demonstrated at the beginning, during and at completion of the test program.
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JCO-0C-85-49 Rev. O February 8, 1985 Static Time Delay Unit (continued)
CONCLUSION Qualification data in GE's possession supports the conclusion that the Static Time Delay unit will perform the safety function in the specified environment. Therefore, continued operation is justified in accordance with the Federal Code of Regulations 10CFR50.49 section (1)(2).
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JCO-0C-85-50, Rev. O February 8, 1985 Page 1 of 2 i
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JCD-0C-85-50 Rev. O February 8, 1985 Page 2 of 2 COMPONENTS Tag Ncs. Various Description Containment Electrical Penetrations OBJECTIVE The oblective of this discussion is to justify interim operation of the Oyster Creek facility in accordance with the Federal Code of Regulations 10 CFR50.49 section (t)(2).
COMPONENT LOCATION These penetrations are located in the primary containment wall at various elevations.
ENVIRONMENT Operating time - 48 days Peak conditions (*F/psig/RH%/ duration) - 335/38/100/16 sec.
Radiation - 6.5E7 Rad (Gamma) TID (40 yr + 1 yr. DBE) 9.6E8 Rad (Beta) TID (1 year DBE)
Spray - Demineralized water ,
, EVALUATION The General Electric F01 series penetrations are presently being quallfted for Oyster Creek. DBE test data concerning the F01 series penetration envelops all the conditions of the nost severe Oyster Creek profile.
Traceability / similarity has been established between the tested penetrations and those installed at Oyster Creek. In addition, testing and analyses have been performed to determine a quallfled life estimate for the Penetrations in excess of 20 years (thermal and radiation). Similar F01 series penetrations are presently being successfully quallfled for Three Mlle Island Unit 1.
CONCLUSION Test data exists to qualify the penetration for the Oyster Creek DBE environment. Analyses are being concluded to complete the qualification of the Oyster Creek Electrical Penetrations with no problems encountered to date. Therefore, in accordance with 10CFR50.49 section (1)(2) Interim operation is justified.
t JCO-0C-85-51, Rev. O February 8, 1985 Page 1 of 2 L
OYSTER-CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO)
JCO-0C-85-51 Rev. 0 l February 8, 1985 Page 2 of 2 COMPONENTS Tag Nos. MCC-1All, MCC-1Al2, MCC-1A21A, MCC-1A218, MCC-1813, Oescription Motor Control Centers OBJECTIVE The objective of this discussion is to justify the interim operation of the Oyster Creek facility in accordance with 10CFR50.49 (i)(2).
COMPONENT LOCATIONS Reactor and turbine buildings COMP 0NENT FUNCTION The Motor Control Centers are used to protect and operate motorized valve actuators and pumps.
ENVIRONMENT Temperature 192*F/25 sec. return to 93'F/9000 sec. (see zone T profile), Normal 85'F Pressure 18.0 psla 0 1 sec., ramp to normal ambient 100 secs.
Relative Humidity 1007.
Radiation 3 x 10' Rads TID EVALUATION A preliminary assessment of the test data applicable to the environmental qualification of the General Electric IC7700 Motor Control Center has been made. The qualification data has been evaluated per DOR Guidelines and by applying Arrhenius techniques.
This test data can be used to demonstrate qualification of the Motor Control centers to Oyster Creek Power Plant's normal and postulated accident conditions. (Reference - Environmental Qualification Assessment Report -
GE letter G-EN-4-206)
General Electric is currently preparing a qualification report for these MCCs, including an analysis of all sub-assemblies (devices of the MCCs i.e.
circuit breakers, starters). The devices and MCC structures are being examined and evaluated in detail to ascertain qualification of the components comprising the MCC assembly to the specified Oyster Creek Power Plant environment. This effort, along with the preparation of qualification documentation for IC7700 MCC installed in other power plants, with similar environments, indicates that the test data demonstrates qualification of the IC7700 motor control center to Oyster Creek Power Plant's normal and postulated accident conditions.
CONCLUSION Based upon the test data obtained and the assessments performed to date, this analysis meets the criteria of 10CFR50.49, paragraph (t)(2).
Therefore, continued operation is justifled.
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JCO-0C-85-52,-Rev. 0 lc February 80 1985 Page 1 of 3 OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION FOR CONTINUED OPERATION (JCO)
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JCO-0C-85-52 Rev. O February 80 1985 Page 2 of 3 COMPONENTS Tag Mos. MCC-IAB2, MCC-IB21A, MCC-IB21B, MCC-DCI Description Motor Control Centers OBJECTIVE The objective of this discussion is to justify the interim operation of the Oyster Creek facility in accordance with 10CFR50.49 (t)(2).
COMPONENT LOCATIONS Reactor building COMPONENT FUNCTION The Motor Control Centers are used to protect and operate motorized valve actuators and pumps.
ENVIRONMENT Temperature peak 150*F/50 sec. return to 100*F/1600 sec., Normal 85'F Pressure 18.0 psia 9 1 sec., ramp to normal ambient 100 secs.
Relative Humidity 100%
Radiation 9.5 x 10' Rads TID EVALUATION A preliminary assessment of the test data applicable to the environmental qualification of the General Electric IC7700 Motor Control Center has been made. The qualification data has been evaluated per DOR Guidelines and by applying Arrhenius techniques.
This test data can be used to demonstrate qualification of the Motor Control centers to Oyster Creek Power Plant's normal and postulated accident conditions. (Reference - Environmental Qualification Assessment Report -
GE letter G-EN-4-206)
General Electric is currently preparing a qualification report for these MCCs, including an analysis of all sub-assemblies (devices of the MCCs i.e. circuit breakers, starters). The devices and MCC structures are being examined and evaluated in detail to ascertain qualification of the components comprising the MCC assembly to the specified Oyster Creek Power Plant environment. This effort, along with the preparation of qualification documentation for IC7700 MCC Installed in other power plants, with similar environments, Indicates that the test data demonstrates qualification of the IC7700 motor control center to Oyster Creek Power Plant's normal and postulated accident conditions.
The specified radiation environment exceeds that to which the MCC's have previously been quallfled. . General Pubile Utilities is in the process of recalculating the specified dose at the exact device location. The result is expected to fall within the quallflable limits of the established test and I analyses.
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l JCO-0C-85-52 Rev. O February 8, 1985 Page 3 of 3 k-Motor Control Centers (continued)
CONCLUSION Based upon the test data obtained and the assessments performed, this analysis meets the criteria of 10CFR50.49, paragraph (1)(2).
-Therefore, continued operation is justified.
ATTACHMENT III Components Deleted from the Original Equipment List Equip.
ID NO. Generic Name Containment Spray System IP-05-A D. P. Transmitter IP-05-B D. P. Transmitter IP-05-C D. P. Transmitter IP-05-D D. P. Transmitter V-21-1* Motor Operated Valve V-21 -3* Motor Operated Yalve V-21 -7
- Motor Operated Yalve V-21-9* Motor Operated Valve Control Rod Drive System RD-08-A Level Switch RD-08-B Level Switch RD-08-C Level Switch RD-08-D Level Switch RD-08-E Level Switch RD-08-F. L? vel Switch Drywell & Supression System Y-5-148 Motor Operated Valve V-17-55* Motor Operated Yalve V-17-56* Motor Operated Valve V-17-57* Motor Operated Valve V-22-1 Solenoid Yalve V-22-2 Solenoid Yalve V-22-28 Solenoid Valve V-22-29 Solenoid Valve V-23-13 Solenoid Yalve V-23-14 Solenoid Yalve V-23-15 Solenoid Yalve V-23-16 Solenoid Yalve V-23-17 Solenoid Valve V-23-18 Solenoid Valve V-23-19 Solenoid Yalve V-23-20 Solenoid Yalve V-23-21 Solenoid Valve V-23-22 Solenoid Valve V-26-16 Solenoid Yalve V-26-18 Solenoid Yalve V-27-1 Solenoid Valve V-27-2 Solenoid Valve V-27-3 Solenoid Valve
- Limit switch for these MOVs will be qualified for position indication purpose per Reg. Guide 1.97 c . .. .
t Equip.
ID No. Generic Name I
V-27-4 Solenoid Valve I V-28-17 Solenoid Valve V-28-18 Solenoid Valve V-28-47 Solenoid Valve V-31 -2 Solenoid Valve PT-52 Pressure Transmitter PT-IP-12 Pressure Transmitter PT-IP-07 Pressure Transmitter
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Isolation Condenser System 18-06-A Area Temp. Detector IB-06-B Area Temp. Detector IB-06-C Area Temp. Detector IB-06-D .
Area Temp. Detector Reactor Instrum?ntation I A-12 Level Transmitter
'IA-45 Pressure Transmitter Combustible Ga:; Monitoring V-38-9 Solenoid Yalve V-38-10 Solenoid Valve V-38-16 Solenoid Valve V-38-17 Solenoid Valve 1
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