ML20028F074
| ML20028F074 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 01/25/1983 |
| From: | Utley E CAROLINA POWER & LIGHT CO. |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8301310105 | |
| Download: ML20028F074 (45) | |
Text
--
CD&L Carolina Power & Light Company JAN 251983 Director of Nuclear Reactor Regulation Attention: Mr. D. B. Vassallo, Chief Operating Reactors Branch No. 2 Division of Licensing United States Nuclear Regulatory Commission Washington, DC 20555 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324 LICENSE NOS. DPR-71 AND DPR-62 ENVIRONMENTAL QUALIFICATION OF SAFETY RELATED ELECTRICAL EQUIPMENT
Dear Mr. Vassallo:
Carolina Power & Light Company (CP&L) has received your letter dated December 20, 1982, which forwarded the Safety Evaluation Report (SER) and Technical Evaluation Report (TER) concerning the environmental qualification of safety-related electrical equipment for the Brunswick Steam Electric Plant, Unit Nos. I and 2. 'Ihe purpose of this letter is to provide CP&L's 30-day response to the Staff's request for justifications for continued operation (JCO's) as outlined in the SER.
l i
As requested, the informat. ion enclosed provides a technical basis for the continued operation of the Brunswick Plant for that equipment which has been determined by the Staff to be either (1) unqualified or-(2) as not having qualification established and for which JCO's have not been previously submitted. In addition, CP&L reaffirms the previously submitted basis for continued operation for those items not addressed by this submittal.
Enclosure 1 provides a list, itemized by TER item number, of the status of JCO's for all TER items. Enclosure 2 consists of those JCO's which address the Staff's SER request discussed above. Of particular note, a JC0 for Raychem cable (TER Item 164) was provided by our submittal dated l December 31, 1982 and is not included herein. Carolina Power & Light Company
( believes that there are no known concerns relating to the environmental i qualification of safety-related electrical equipment which would interfere with safe, continued operation of the Brunswick Plant.
O 8301310105 830125 PDR ADOCK 05000324 P pop
,. . . .qetteville Street
- P. O. Box 1551
- Raleigh. N. C. 27602
Mr. D. B. Vassallo 1 If you should have any questions on this response, ~please contact our staff.
Yours very truly ,
. . . , /
E. E. Utley Executive- Vice President Power Supply and Engineering & Construction WRM/kj r (6053C12T5)
Enclosures
]
cc: Mr. D. O. Myers (NRC-BSEP)
Mr. J. P. O'Reilly (NRC-RII)
Mr. S. D. Mac Kay (NRC) j Ja
.=*-
2 i
l f
I l
l f
l' ENCLOSURE 1 STATUS OF JUSTIFICATIONS FOR CONTINUED OPERATION BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS. 1 AND 2
..s.....u... .v.. vv.. ...LJ3 v.e.uiiGN (JCO) STATUS SHEEX TER Item JC0 JC0 JC0 JC0 Not Numbe r Enclosed Submitted 3/82 Submitted 10/80 Reauired 1 X NOTE 1 2 X 3 X 4 X 5 X 6 X 7 X '
8 X 9 X -
10 X 11 X 12 X 13 X 14 X 15 X 16 X 17 X 18 X 19 X 20 X 21 X 22 X 23 X 24 X 25 X 26 X 27 X 28 X 29 X 30 X 31 X 32 X 33 X 34 X 35 X 36 X 37 X 38 X 39 X 40 X 41 X
'2
+ X 43 X 44 X 45 X 46 X 47 X 48 X 49 X 30 X Page 1
. TER Item JC0 JC0 JC0 JC0 Not Number Enclosed Submitted 3/82 Submitted 10/80 Reonired 51 X 52 X 53 X 54 X 55 X 56 X 57 X 58 X 59 X 60 X 61 X 62 X 63 X 64 X NOTE 2 65 X NOTE 2 66 X 67 X 68 X 69 X 70 X 71 X 72 X 73 X 74 X 75 X 76 X 77 X 78 X 79 X 80 X 81 X 82 X 83 X -
84 X 85 X 86 X 87 X 88 X 89 X NOTE 2 90 X 91 X 92 X NOTE 2 93 X 94 X 95 X 96 X 97 X 98 X -
99 X 100 X Page 2
TER Item JC0 JC0 JC0 'JC0 Not Numbe r Enclosed Submitted 3/82 Submitted 10/80 Required 101 X
., 102 X 103 X 104 X 105 X
, 106 X ~
107 X ~
108 X l 109 X .,
1 10 X '
111 X 112 X 113 X 1 14 .X 1 15 X 116 X 117 X 118 X 119 X NOTE.3-120 X NOTE 3 121 X NOTE 3
.) 122 X l 123 X 124 X
! 125 X 126 X
. 127 X l 128 X l 129- X 1
130 X -
131 X
~
132 'X 133 X I 134 X
( 135 X 136 X
! 137 X 138 X 139 X 140 X 141 X 142 X 143 X l'
144 X 145 X 146 X 147 X 148 X 149 X 150 X Page 3
TER Item JC0 JC0 JC0 JC0 Not Nu:nbe r Enclosed Submitted 3/82 Submitted 10/80 Required 15 1 X 152 X 153 X 154 X 155 X 156 X 157 X 150 X 159 X' '
160 X 161 X NOTE 1 162 X 163 X 164 X NOTE 4 165 X 166 X NOTE 3 167 X 168 X NOTE 1 169 X 170 X 171 X NOTE 1 172 X 173 X 174 X 175 X NOTE 3 176 X 177 X 178 X ,
179 X
~
180 X 181 X .
182 X 183 X NOTE 3 NOTES ,(
- l. Item is Category II.c.
- 2. Item has been removed from the plant.
- 3. Item is Category I.a.
- 4. Item was justified in our letter of 12/31/82 (Eury - Vassallo).
4 Page 4 4
,. - - .a
J' ENCLOSURE 2 JUSTIFICATIONS FOR CONTINUED OPERATION BRUNSWt K STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2
~.
9 i
f i
)
I
TER ITEM NUMBERS: 2,3,4,5,7,8,9, 11, 12, 13, 14, 15, 16, and 17 LIMITORQUE MOTORIZED VALVE ACTUATOR (SMB SERIES)
Plant ID Numbers: B21-F019 E21-F004A E51-F008 A-BFV-RB E11-F002A E21-F004B E51-F013 B-BFV-RB E11-F002B E21-F015A E51-F019 C-BFV-RB E11-F008 E21-F015B E51-F029 D-BFV-RB E11-F023 E21-F031A E51-F031 E-BFV-RB E11-F052B E21-F031B F-B FV-RB E11-F068A G31-F004 G-BFV-RB E11-F068B E41-F001 H-B FV-RB E11-F075 E41-F003 SW-V101 I-BFV-RB ,
E11-F103A E41-F004 SW-V102 E11-F103B E41-F006 SW-V105 N-BFV-RB E11-F104A E41-F007 SW-V106 E11-F104B E41-F008 SW-V111 CAC-V23 E11-V35 E41-F012 SW-V117 E11-V36 E41-F041 SW-V118 E11-V37 E41-F042 E11-V38 E41-F059 NOTE: ID Numbers SW-V103 and SW-V104 are not on tue safety related equipment list.
Component materials of the Limitorque Motorized Valve Actuators have been identified. These materials have been evaluated per DOR guidelines and by applying Arrhenius techniques. Results of this analysis indicate that the Class B or Class H insulation system, Durez switches, and internal wire insulation materials have greater than forty (40) years demonstrated qualified life at the maximum reactor building temperature of 104*F (
Reference:
Patel Report Number PEI-TR-83-4-3).
Therefore, continued operation is justified.
TER ITEM NUMBERS: 2, 4, 12, 16, 18. 19, AND 20 LIMITORQUE MOTORIZED VALVE ACTUATOR (SMB SERIES)
Plant ID Numbers:
B21-F016 Ell-F009 E21-F001A *E51-F007 B32-F031A Ell-F015A E21-F001B *G31-F001.
B32-F031B Ell-F015B E21-F005A c-B32-F032A Ell-F020A E21-F005B B32-F032B Ell-F020B E21-F037A
- Ell-F022 E21-F0373 Ell-F122A E41-F002 Ell-F122B
- Motors with Class H insulation.
Component materials of the Limitorque Motorized Valve Actuators have been identified. These materials have been evaluated per DOR guidelines and by applying Arrhenius techniques. Results of this analysis indicate that the Class RH insulation system, melamine switches, and internal wire insulation materials have greater than 40 years demonstrated qualified life at the maximum drywell temperature of 150'F (
Reference:
Patel Report Number PEI-TR-83-4-3).
Motors with Class H insulation systems have been identified .
above by an asterisk. These motors are superior in construction (per Limitorque Corpogation) to the Class B insulated motors successfully tested to 2 x 10 rads gamma. Additionally, these Class H motors have been successfully LOCA tested to a peak temperature of 329'F which exceeds the postulated plant accident at Brunswick.
Therefore, continued operation is justified.
TER ITEM NUMBERS: 6. 10, 12, 14, 16. and 21 LIMITORQUE MOTORIZED VALVE ACTUATOR (SMB SERIES)
Plant ID Numbers: E11-F003A Ell-F011A Ell-F028A Ell-F0033 E11-F0llB E11-F028B Ell-F004A Ell-F016A Ell-F047A E' Ell-F004B E11-F016B Ell-F047B Ell-F004C Ell-F0~.7A Ell-F048A Ell-F004D Ell-F017B Ell-F048B Ell-F006A Ell-F021A E11-F049 Ell-F006B E11-F021B E11-F052A Ell-F006C Ell-F024A SGT-V8 Ell-F006D Ell-F024B SGT-V9 Ell-F007A Ell-F027A Ell-F007B Ell-F027B NOTE: ID Numbers Ell-F007C and D are not on the list of safety related equipment.
Component materials of the Limitorque Motorized Valve Actuator have been identified. These materials have been evaluated per DOR guidelines and by applying Arrhenius techniques. Results of '
this analysis indicate that the Class H motor insulation system, -
melamine switches, and internal wire insulation materials have greater than forty (40) years demonstrated qualified . life at the maximum reactor building temperature of 104*F (
Reference:
Patel Report Number PEI-TR-82-4-2) .
Therefore, continued operation is justified..
4 TER ITEM NUMBER 18 i
LIMITORQUE MOTORIZED VALVE ACTUATOR (SMB SERIES)
Plant ID Numbers: B32-F043A B32-F043B B32-F044A B32-F944B The motorized valve actuators listed have been disabled and
- the valves locked in the required position per NRC regulations.
These valves are not required for accident mitigation and therefore, continued operation is justified.
O e
i
f TER ITEM NUMBER 46 AVC0 SOLENOID VALVE 5450-5 I Plant ID Numbers: B21-F013 A Through H B21-F013 J Through L NOTE: The TER erroneously identifies thiese valves
- as ASCO valves. --
R ese valves have been replaced with Target Rock Valves (Part Numbe r h SMS-A-01-2) . W e Target Rock Valves are qualified for use at BSEP (
Reference:
Target Rock Report Number 2199A dated December 27, 1979).
Therefore, continued operation is justified.
t i
1
-l
TER ITEM NUMBER 56 BARTON PRESSURE SWITCH MODEL 288 Plant ID Numbers:
B21-PS-N021B B21-PS-N021D ,,
E41-PDS-N004 E41-PDS-N005 E51-PDS-N017 E51-PDS-N018 These pressure switches have been replaced with Rosemount Pressure Transmitters (Part Numbers 1152GP and 1152DP Series) . The Rosemount pressure transmitters are qualified for use at BSEP (
Reference:
United Engineers Report Number UC-33229, dated August 20, 1982).
Therefore, continued operation is justified.
NOTE: New plant ID numbers were assigned to the replacement pressure transmitters as follows (respectively) :
~
B21-PT-N021B -
B21-PT-N021D E41-PDT-N004 E41-PDT-N005 E51-PDT-N017 E51-PDT-N018 1
TER ITEM NUMBERS 57 AND 59 BARTON Differential Pressure Indicating Switch Models 288A/289A Plant ID Numbers:
CAC-PDS-4222 E41-PS-N001A CAC-PDS-4223 E41-PS-N001B E41-PS-N001C E41-PS-N001D .
l Component materials of the Barton differential pressure switches have been identified. These materials have been evaluated per DOR guidelines and by applying Arrhenius techniques. Resultsofthisanagysisindicate that the nonmetallic components have greater than 1.5 x 10 years of expected life at the maximum reactor building temperature of 104*F. The pressure-switch nonmetallic materials are exposed to the plant postulated accident temperature peak of 180*F for thirty-five (35) minutes. The accident tempera-ture then decreases to 125'F within three (3) hours of event initiation. With an expected life of 96 years at 180*F, the pressure switch nonmetallic materials are insensitive to thermal degradation for the required operating period.
Additionally, similar switches have been successfully tested for six (6) hours at 10 0: RH, 7 inches W.C. with temperatures from 40*F to 212*F. Also, the radiation testing performed (3 6 x 10 s rads gamma) on similar switches exceeds the postulated TID (1 x 10 rads gamma) for these switches (Refer-ence: Pa tel Report Number PEI-TR-83-4-22) .
Therefore, continued operatica is justified. '
, TER ITEM NUMBER 58 BARTON PRESSURE SWITCH MODEL 288A Plant ID Numbers:
B21-PDS-N006A B21-PDS-N008A B21-PDS-N006B B21-PDS-N008B B21-PDS-N006C B21-PDS-N008C ..
B21-PDS-N006D B21-PDS-N008D B21-PDS-N007A B21-PDS-N009A B21-PDS-N007B B21-PDS-N009B B21-PDS-N007C B21-PDS-N009C B21-PDS-N007D B21-PDS-N009D These pressure switches have been replaced with Rosemount Pressure Transmitters (Part Number 1152DP7E22). The Rosemount pressure transmitters are qualified for use at BSEP (
Reference:
United Engineers Report Number UC-33229, dated August 20, 1982).
i iherefore, continued operation is justified.
NOTE: Pew plant ID numbers were assigned to the replacement pressure transmitters as follows (respectively):
B21-PDT-N006A
~~
B21-PDT-N006B B21-PDT-N006C B21-PDT-N006D i
B24-PDT-NC07A B21-PDT-N007B B21-PDT-N007C B21-PDT-N007D l B21-PDT-N008A B21-PDT-N008B B21-PDT-N008C-B21-PDT-N008D B21-PDT-N009A l B21-PDT-N009B i B21-PDT-N009C l B21-PDT-N009D l
i t
, TER ITEM NUMBERS 60 AND 61
- SIATIC-0-RING PRESSURE SWITCH MODEL 12N-AA4-X10TT ,
t I
1 Plant ID Numbers:
E11-PS-N010A Ell-PS-N019A Ell-PS-N010B E11-PS-N0198 ,-
i Ell-PS-N010C E11-PS-N019C Ell-PS-N010D Ell-PS-N019D Ell-PS-N0llA C72-PS-N002A Ell-PS-N011B C72-PS-N002B Ell-PS-N0 llc C72-PS-N002C Ell-PS-N011D C72-PS-N002D
, These pressure switches have been replaced with Rosemount Pressure l Transmitters (Part Number ll52GP4E22). - The Rosemount pressure transmitters
- are qualified for use at BSEP (
Reference:
United Engineers Report Number UC-33229, dated August 20, 1982).
i j Therefore, continued operation is justified.
NOTE: New plant ID numbers were assigned to the replacement pressure transmitters as follows (respectively):
~
Ell-PT-N010A E11-PT-N010B E11-PT-N0100 .
E11-PT-N010D Ell-PT-N0llA Ell-PT-N011B Ell-PT-N0 llc
, Ell-PT-N0llD
, Ell-PT-N019A Ell-PT-N019B Ell-PT-N019C E11-PT-N019D i C72-PT-N002A C72-PT-N002B C72-PT-N002C C72-PT-N002D
- TER ITEM NUMBERS 62 AND 63 STATIC-0-RING PRESSURE SWITCH MODELS SN-AA3-X9-STT AND 6N-AA21-X9-SVTT Plant ID Numbers: E21-PS-N008A E21-PS-N009A E21-PS-N008B E21-PS-N009B J' E41-PS-N010 Component materials of the Static-0-Ring Pressure Switches have been identified. These materials have been evaluated per DOR guidelines i
and by applying Arrhenius techniques. Results of this analysis indicate
, that each of the non-metallic materials (Viton A, Neoprene, General Purpose Phenolic and Ethylene Propylene rubber) has an expected life in excess of 261 years at the maximum reactor building temperature of 104*F. The switch nonmetallic materials are exposed to the plant postulated accident tempera-ture peak of 192*F for a maximum of 10 minutes. The accident temperature then decreases to 125*F within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of event initiation. With an expected life of 564 days at 192*F, the switch nonmetallic materials are insensitive to thermal degradation for the required operating period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Also, similar switches were tested at high humidity for 100 days at 150*F with no f ailures observed. Prior to the accident testing these switches had been thermally aged for 2400 hours0.0278 days <br />0.667 hours <br />0.00397 weeks <br />9.132e-4 months <br /> at 80*C.
3 Additionally, a t the highest dose rate during the initial 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ten the accident, the total accumulated dose of the non-metallics will be significantly less than the lowest radiation damage threshold.
(
Reference:
Patel Report: PEI-TR-83-4-18.)
Therefore, continued operation is justified.
TER ITEM NUMBER 70 BARKSDALE PRESSURE SWITCH MODEL B2T-M12SS Plant ID Numbers: B21-PS-N021A B21-PS-N021C E41-PS-N023A E41-PS-N023B E51-PS-N023C c-E51-PS-N023D These pressure switches have been replaced with Rosemount Pressure Transmitters (Part Numbers 1152GP Series). The Rosemount pressure transmitters are qualified for use at BSEP (
Reference:
Uniced Engineers Report Number UC-33229, dated August 20, 1982).
Therefore, continued operation is justified.
NOTE: New plant ID numbers were assigned to the replacement pressure transmitters as follows (respectively):
B21-PT-N021A B21-PT-N021C E41-PT-N023A .,
E41-PT-N023B i
E51-PT-N023C E51-PT-N023D i
i i
TER ITEM NUMBER 71, 72. 73. 74. 75, 76, 77, 78, 79, 80, 81 AND 99 Barksdale Pressure Switch Models: B2T-M12SS, D2H-M150SS, D2T-M18SS, D2T-M150SS, PlH-M340SS. TC9622-1 and T2H-M251S-12 Plant ID Numbers: Ell-PS-N016A E41-PSH-N012A RIP-PSL-1218 Ell-PS-N016B E41-PSH-N012B PIP-PSL-1219 Ell-PS-N016C E41-PSH-N012C RIP-PSL-1220 Ell-PS-N016D E41-PSH-N012D RIP-PSL-1221 Ell-PS-N020A E41-PSH-N017A RIP-PSL-1222 E11-PS-N020B E41-PSH-N017B RIP-PSLO1223 ..
Ell-PS-N020C E41-PSH-N027 RIP-PSL-1225 Ell-PS-N020D E51-PS-N019A RIP-PSL-1227 E51-PS-N019B RIP-PSL-1228 E51-PS-N019C RIP-PSL-1229 RIP-PSL-1200 E51-PS-N019D B32-PS-N018A RIP-PSL-1201 E51-PS-N020 B32-PS-N018A-1 RIP-PSL-1206 E51-PSH-N009A B32-PS-N018B RIP-PSL-1209 E51-PSH-N009B SW-TSH-1109 RIP-PSL-1210 E51-PSH-N012A SW-TSH-1110 RIP-PSL-1211 E51-PSH-N012B SW-TSH-illi RIP-PSL-1212 E51-PSH-N012C SW-TSH-1112 RIP-PSL-1217 , E51-PSH-N012D Component materials of the Barksdale switches have been identified.
These materials have been evaluated per D0R guidelines and by applying Arrhenius techniques. Results of this analysis indicate that all materials, except for Buna-N rubber, have greater than 47 years expected life at the maximum reactor building temperature of 104*F. The switch materials are exposed to the plant postulated accident temperature peak of 298'F -or only three (3) minutes.
The accident temperature then decreases to 14 F within one (1) ~
hour of event initiation. With an expected life of 146 hours0.00169 days <br />0.0406 hours <br />2.414021e-4 weeks <br />5.5553e-5 months <br /> at 298'F, the switch nonmetallic materials are insensitive to thermal degradation for the required operating period.
For Buna-N, the rubber manufacturer's literature recommends continuous use up to 200*F and extreme use to 300*F for short durations. Therefore, continued operation is justified for switches with Buna-N rubber due to the relatively mild postulated accident profile.
Also, the component nonmetallic materials have been successfully radiation aged during qualification testing (while being used in similar applications) to levels greater than 1 x 107 rads gamma, the postulated accident TID for BSEP.
In addition, the Brunswick switches are located in NEMA 3, 4,12 or 13 enclosures where the effects of direct steam impingement / humidity would be significantly reduced during the postulated accident (
Reference:
Patel Report Num be r PEI-TR-8 3-4-2 3) .
Therefore, continued operation is justified.
TER ITEM NUMBERS 84, 86, AND 87 YARWAY LEVEL SWITCH MODEL 4418C/4418EC Plant ID Numbers:
- B21-LS-N021A B21-LS-N031A
- B21-LS-N021C B21-LS-NO31B ,--
B 21-LS-N024A B21-LS-NO310 B21-LS-N024B B21-LS-NO31D B21-LS-N025A B21-LS-N042A B 21-LS-N025B B21-LS-N042B B21-LITS-NO36 B 21-LITS-NO37
- Items deleted from the list of safety-related equipment.
These level switches have been replaced with Rosemount Pressure Transmitters (Part Numbers 1152DP Series and 1152GP Series) . The Rosemount pressure transmitters are qualified for use at BSEP
(
Reference:
United Engineers Report Number UC-33229, dated August 20, 1982).
4 Therefore, continued operation is justified.
NOTE: New plant ID numbers were assigned to the replacement .,
pressure transmitters as follows (respectively):
- B21-LT-N024A B21-LT-N024B B21-LT-N025A B21-LT-N025B B 21-LT-N031A B 21-LT-NO31B B21-LT-N031C B 21-LT-N031D B 21-LT-N042A B21-LT-N042B B21-LT-N036
- B21-LT-N037 I
1 l
l l
TER ITEM NUMBER: 85 .
YARWAY LEVEL INDICATING SWITCH MODEL 4418EC Plant ID Numbers: B21-LITS-N026A
- B21-LIIS-N026B Component materials of the Yarway Level Switch, Model 4418EC."have I
been identified. These materials have been evaluated per DOR Guidelines and by applying Arrhenius techniques. Results of this analysis indicate that all nonmetallic materials, except the Buna-N O ring seal, have greater than 83 years expected life at the maximum Reactor Building maximum temperature of 104 *F. The level indicating switch nonmetallic materials are exposed to the plant postulated accident temperature peak of 225'F for only two (2) minutes. The accident temperature then decreases to 125'F within three (3) hours of event initiation. With an expected life of 757 hours0.00876 days <br />0.21 hours <br />0.00125 weeks <br />2.880385e-4 months <br /> at 225'F, the level indicating switch nonmetallic materials are insensitive to thermal degradation for the required operating period.
Also, the lowest radiation damage threshold for the nonmetallic 6
materials (1 x 10 Rads gamma) exceeds the postulated TID (1 x 10 Rads 5
gamma) for these level switches (
Reference:
Patel Report PEI-TR-83-4-21).
BSEP personnel conduct periodic (every 18 months minimum, per PT 55.3PC and PT 56.4PC) calibration / maintenance checks of these level switches to ensure proper operation. Additionally, these indicating level switches have three (3) qualified backup instruments for reliability. ,
Therefore, continued operation is justified.
I l
l
TER ITEM NUMBER 88 MAGNETROL LEVEL SWITCH MODEL 5.0-751 Plant ID Numbers:
C11-LSH-N013A C12-LSH-N013A C11-LSH-N013B C12-LSH-N013B .
C11-LSH-N013C C12-LSH-N013C C11-LSH-N013D C12-LSH-N013D Component materials of the Magnetrol Level Switch have been identified. These materials have been evaluated per DOR guidelines and by applying Arrhenius techniques. Results of this analysig indicate that the nonmetallic components have greater than 8.5 x 10 years of expected life at the maximum reactor building temperature of 104*F.
The level switch nonmetallic materials are exposed to the plant pos tulated accident temperature peak of 298'F for only two (2) minutes. The accident temperatu're then decreases to 125*F within one (1) hour of event initiation. With an expected life of seven (7) years at 298'F,-
the level switch nonmetallic materials are insensitive to thermal degradation for the required operating period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Additionally, similar switches have been thermally aged at 300*F for 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br />, then successfully tested for 480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br /> at 95-100 percent relative humidity with the temperature varying between ambient and 100*F. ,,
Also, the lowest radiation damage threshold for the nonmetallig ma terials (1 x 10' rads gamma) exceeds the postulated TID (2.4 x 10 rads gamma) for these switches. (
Reference:
Patel Report Number PEI-TR-83-4-19.)
Therefore, continued operation is justified. >
TER ITIM NUMBER 91 MAGNETROL FLOW SWITCH MODEL F-521 Plant ID Numbers: B21-FS-F015A B21-FS-F-43A E41-FS-F024A B21-FS-F015B B21-FS-F043B E41-FS-F024B B 21-FS-F015C B21-FS-F045A E41-FS-F024C B21-FS-F015D B21-FS-F045B E41-FS-F024D B 21-FS-F015E B21-ES-F047A E51-FS-F044A E 21-FS-F015F B21-FS-F047B E51-FS-F044B B Zl-FS-F015G B21-FS-F049A E51-FS-F044C 6" B21-FS-F015H B21-FS-F049B E51-FS-F044D B 21-FS-F015J B21-FS-F051A B21-FS-F015K B21-FS-F051B B21-FS-F015L B21-FS-F055 B21-FS-F015M B21-FS-1227F B 21-FS-F015N B21-FS-F015P B21-FS-F015R B21-FS-F015S Component materials of the Magnetrol Flow Switch have been identified.
These materials have been evaluated per DOR guidelines and by applying Arrhenius techniques. Results of the analysis indicate that the nonmetallic components have greater than 261 years of expected life at the maximum reactor building temperature of 104*F. The flow switch nonmetallic materials are exposed to the plant postulated accident temperature peak of 298'F for only two (2) minutes. The accident temperature then decreases to 125'F within one .
(1) hour of event initiation. With an expected life of 261 hours0.00302 days <br />0.0725 hours <br />4.315476e-4 weeks <br />9.93105e-5 months <br /> at 298'F, the flow switch nonmetallic materials are insensitive to thermal degradation for the required operating period. (
Reference:
Patel Report Number PEI-TR-83-4-20.)
Therefore, continued operation is justified.
I l
TER ITEM NUMBERS 94 AND 122 CHERRY ELECTRICAL PRODUCTS MODEL E2360H Plant ID Number: Reactor Instrument Penetration Isolation Valve Position Indicator Switch (Various)
Component materials of the Cherry switch, Model E2360H, have been-identified. These materials have been evaluated per D0R guidelines and by applying Arrhenius techniques. Results of this analysis indicate that each of the non-metallic materials (General Purpose Phenolic, acetal resin, and polybutylene terephthalate) has an expected life in excess of 660 years at the maximum reactor building temperature of 104*F. The switch nonmetallic materials are exposed to the plant postulated accident temperature peak of 298'F for only two (2) minutes. The accident temperature then decreases to 125'F within one (1) hour of event initiation. With an expected life of 732 hours0.00847 days <br />0.203 hours <br />0.00121 weeks <br />2.78526e-4 months <br /> at 298'F, the switch nonmetallic materials are insensitive to thermal degradation for the required operating period.
In addition, the Brunswick Cherry switches are located in sealed NEMA-4-like enclosures where the ef fects of direct steam impingement /
humidity would be significantly reduced during the postulated accident.
(
Reference:
Fatel Report Number PEI-TR-83-4-15.)
Therefore, continued operation is justified.
TER ITEM NUMBERS 101 THROUGH 106 NECI THERMOCOUPLE PLANT ID Numbers:
E51-TE-N021A,B E41-TE-NO30A,B E51-TE-N025D E51-TE-N022A,B G31-TE-N016A-F E51-TE-N026C,D E51-TE-N023A,B G31-TE-N022A-F E51-TE-N027C,D E51-TE-N025A G31-TE-N023A-F E51-TE-N026A E51-TE-N027A E51-TE-N025B E51-TE-N026B E51-TE-N025C E51-TE-N027B Component materials of the NECI 145C3224P1 thermocouple have been identified. These materials have been evaluated per DOR guidelines and by applying Arrhenius techniques. Results of this analysis indicate that the RTV-116 Potting Compound, SR 80 Varnish, Fiberglass Sleeving, and Nitrile-based rubber gasket (pending successful completion of the PYC0 test program currently in progress) have greater than forty (40) years demonstrated qualified life at the maximum reactor building temperature of 104*F (
Reference:
PATEL Report Number PEI-TR-83-4-6) .
In addition to the analysis performed on the NECI thermocouple using the parameters of the test program currently in progress, additional analysis were performed in the following area:
Time-Temperature Effects - Based on the expected life calculations pertormed on each material at the assumed baseline temperature of 104*F (40*Cg, all the nonmetallies have expected lives in excess
, of 2.9 x 10 years, except for the Nitrile gasket. The gasket has
~
an expected life of 11 years at the 104*F baseline temperature.
Since the cover of the thermocouple is intended only to be NEMA-1
" splash-proof,"the gasket is not used as a moisture barrier and the thermocouple will continue to function in its absence.
Further testing on similar thermocouples with those gaskets showed that, even though the thermocouples were not completely sealed against the environment, they still functioned (reference excerpts from Qualification Test Report, Nuclear Power Plant Application, PYC0 Document No. 770831, dated August 31, 1977, included as Appendix II of Patel Report No. PEI-TR-83-4-6) .
Radiation Analysis - Excluding the Nitrile rubber gasket (which is not essential for operation), the lowes t radiation threshold is 15 times greater than the worst case postulated total integrated dose in the reactor building.
Therefore, continued operation is justified.
TER ITEM NUMBER 115 NAMCO MODEL 2400XR POSITION SWITC!l Plant ID Numbers: A-BFIV-RB i
B-BFIV-RB C-BFIV-RB D-BFIV-RB ,-
Component materials of the NAMCO 2400XR Position Switch have been identified. These materials have been evaluated per DOR guidelines and by applying Arrhenius techniques. Results of this analysis indicate that all materials, except for Buna-N rubber (used as a binder in the asbestos gasket), have greater than forty (40) years demonstrated qualified life at the m'ximum a reactor building temperature of 104*F. The gasket, which is comprised of 20% Buna-N and 80% asbestos, is judged acceptable for continued operation since the Buna-N is used as a binder and once the gasket is properly installed and lef t undisturbed, no significant degradation would occur during the expected 40 year life.
The analysis performed on the D2400XR switch is based on testing conducted on NAMCO Series SL3 switches due to similarity in materials of construction. (
Reference:
Patel Report Number PEI-TR-83-4-12) .
Therefore, continued operation is justified. .,
l l
i
TER ITEM NUMBERS 116, 117 AND 118
~
BETTIS SWITCilES, Type RX-41 and RX-341 Plant ID Numbers: CAC-V49 CAC-V10 CAC-V15 CAC-V50 CAC-V9 Component materials of the Bettis Limit Switches have been identified.
These materials have been evaluated per DOR guidelines and by applying Arrhenius techniques. Results of this analysis indicate that the. switch materials have greater than forty (40) years demonstrated qualifi?ed life at the maximum reactor building temperature of 104*F (
Reference:
Patel Report Number PEI-TR-83-4-24) .
Therefore, continued operation is justified.
l d
1
L - -
.-o - J - -- 4 - - , 4d ,L TER ITEM NUMBERS 130. 131. 133, 134 AND 135 HONEYWELL MICROSW1TCH, Types: PTSEA202FB52, PTSHA201, PTKBC2221CCF9, PTKBC2221, and PTSHE202CB97 Plant ID Numbers: DL8-RS1 DM7-RS1 DLO-RSl B43-RS1 DL9-RS1 DM8-RSl DL1-RSl DH3-RSl DM2-RS1 DN6-RS1 DL2-RSl DH2-RS1 DM4-RS1 DK8-RS1 >
DL7-RS1 B50-RS1 DMS-RS1 DK9-RS1 DS4-RSl B49-RS1 B11-RS1 B41-RS1 B45-RS1 B46-RS1 B47-RS1 Bil-RS B21-CS-3412 B21-CS-3327 j B 21-CS-3329 and various valve control switches The above switches are installed on elevation 20' of the reactor building and used as isolation / selector switches in the remote shutdown system and are classified as essential passive.
According to Honeywell catalog #70 the PT series switch is a heavy duty, oiltight switch with differing configurations as below; PTS - Knob or wing lever selector PTK - Key operated selector PTH - Lighted pushbutton PTP - Unlighted pushbutton:
switch Part Numbers are developed; PTS - E -
A2 -
02 -
F -
B52 Switch Selector Cam Code Knob or Contact Block Legend Type Action Key Code Configuration Plate since the centact blocks are interchangeable among all types of switches, all PT type switches are similar. .
The switch assembly is made of high strength phenolic. Therefore, we hgve assumed the most conservative threshold value for the radiation damage of 10 rads TID (CP Phenolic P-4050, Appendix C, DOR guidelines). This value exceeds the LOCA TID of 10 rads for this area. This is upheld by the Microswitch Repo rt " Nuclear Radiation and Switch Applicatiogs" which gives an " acceptable absorbed dose" for PT series switches of 5 x 10 rads TID.
Expected life calculations for General Purpose Phenolic gives greater than 3500 years life at the LOCA temperatures of 135'F. Additionally, these switches have been thermally aged at 185*F for 767 hours0.00888 days <br />0.213 hours <br />0.00127 weeks <br />2.918435e-4 months <br />.
This aging temperature does not envelope the HELB peak temperature of 200*F, however, this peak is only above 185'F for 70 seconds. Since the switches are within enclosures, the temperature profile of the HELB would have returned below 185*F before the first signs of the peak manifests itself as a temperature rise within the switch.
Based upon the above, continued operation is justified.
.m - -, - -- --r-
TER ITEM NUMBERS 136, 137 e
RELIANCE ELECTRIC MODEL S-1000 NUCLEAR FAN MOTOR (PUMP ROOM COOLER FAN DRIVE)
"lant ID Numbers: A-FCU-RB B-FCU-RB ..
C-FCU-RB D-FCU-RB Component materials of the Reliance S-1000 Series Nuclear Fan Motors have been identified. These materials have been evaluated per DDR guidelines and by applying Arrhenius techniques.
Results of this analysis indicate that the Class RH motor insulation system has greater than forty (40) years demonstrated qualified life at the maximum reactor building temperature of 104*F (
Reference:
Patel Report Number PEI-TR-83-4-11) .
Therefore, continued operation is justified.
~
TER ITEM NUMBERS 138, 139. 140 GENERAL ELECTRIC MOTORS, Types SK821161C11, SK6346XC94A, and 5K6346XC95A Plant ID Numbers: E11-C001 A Through D E11-C002 A Through D E21-C001 A and B These< motors are installed in the plant as the prime movers for:
Residual Heat Removal (RHR) Pumps - SK6346XC95A Core Spray (CS) Pumps - SK6346XC94A J' RHR Service Water (RHRSW) Booster Pumps - SK821161C11 These are vertical or horizontal induction motors with Class B custom Polyseal insulation. Types SK6346XC94A, 95A are designed as air cooled motors to run continuously at 65'c (149'F) ambient temperature.
Type SK821161C11 is a totally enclosed air / water cooled unit designed to run continuously at 90*C (194*F) ambient.
These continuous ambient temperature ratings are both greater than the Pos t-LOCA ambient high temperature of 140*F. The lists of non-metallic materials for both types of motors have been obtained from CE and analyzed by United Engineers our A/E and have been shown jo suffer no significant pegradationat the LOCA radiation levels (1x10 Rads TID for RHRSW, 1x10 Rads TID for RHR and CS).
During a high energy line break (HELB), the peak ambient tempe ra ture .
would be 290*F for approximately one (1) minute (200*F for the RHRSW motor) with a pressure pulse to 6.9 psig at one (1) second which falls to 0.4 psig at one (1) minute.
If the motors were running at design operating temperature before a H ELB (wo rs t case), the motor insulation would never experience the high peak temperature due to the thermal lag in the motor. In the absence of the rmal lag, the insulation woul'd experience a total temperature of 188'C (65*C ambient plus 45'C design rise plus 78*C accident peak) for only one (1) minute. This is less than the 198'C at which a similiar motor was tested for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (NEDM-10672, " ENVIRONMENTAL QUALIFICATION TEST OF VERTICAL INDUCTION MOTOR FOR ECCS SERVICE IN NUCLEAR POWER I
PLANTS" by P. J. Thiemann, dated August 1972). Due to thermal lag, the temperature profile of the HELB would have returned to the motor design ambient temperature before the first signs of the peak manifests itself as a temperature rise in the insulation. .
Additionally, since the RHRSW motor is water / air cooled, its temperature is more dependent on the cooling water supply temperature than the surrounding ambient.
Based upon the above, continued operation is justified.
TER ITEM' NUM3ER 142 AND 146 CENERAL ELECTRIC RELAY MODELS CR2810 AND CR2811 Plant ID Numbers:
DOO-RS *DP5-936X DA6-3 B49-BN7-RS DA6-3-1 B50-B28-RX '
i DPS-3 Bil-B09-RS '
DPS-3-1 B43-B28-RS DBO-TS-936X B41-B28-RS B45-BN7-RS B46-B28-RS B47-B28-RS
- B50-B28-RS
- ITDi deleted from the List of safety-related equipment.
- Erroneously identified as a CR2810 relay and is, in fact, a ilFA51 relay covered under TER Item Number 145.
Component materials of the General Electric Relays, Models CR2810 and CR2811, have been identified. These materials have been evaluated per DOR guidelines and by applying Arrhenius techniques. Results of this analysis indicate that each of the nonmetallic materials (wood flour .
filled phenolic, nylon, acetate film, and polyvinyl formal magnet wire insulation) has an expected life in excess of 79 years at the maximum reactor building temperature of' 140*F (includes heat rise) . The relay nonmetallic materials are exposed to the plant postulated accident temperature peak of 261*7 (includes heat rise) for thirty-five (35) i minutes. The accident temperature then decreases to 206*F (includes heat rise) within three (3) hours of event. initiation. With an expected life of 2680 hours0.031 days <br />0.744 hours <br />0.00443 weeks <br />0.00102 months <br /> at 261*F, the relay nonmetallic materials are insensitive to thermal degradation for the operating period.
In addition, the BSEP relays are located inside cabinetry where the effects of direct steam impingement / humidity would be significantly reduced during the postulated accident. Further analysis indicates that all the nonmetallic materials have a radigtion damage threshold significantly greater than the required level of 1 x 10 rads gamma (
Reference:
Patel Report Num be r PEI-TR-83-4-17) .
Therefore, continued operation is justified.
TER ITEM NUMBER 145 GENERAL ELECTRIC RELAY MODEL HFA51A49H Plant ID Numbers:
D00-RX DLO-RX DK9-RX .,
DL1-RX DL2-RX
- B49-BN7-RS
- B50-B28-RX B50-B28-RS
- B11-B09-RS
- B43-B28-RS
- B41-B28-RS
- B45-BN7-RS
- B46-B28-RS
- B47-B28-RS
- Items erroneously identified as a General Electric Model HEA51 Series relay, and is in fact a General Electric CR2811 Series relay covered under TER Item Numbers 142 and 146.
Component materials of the General Electric HFA51A Series relay
- have been identified. These materials have been evaluated per DOR
- guidelines and by applying Arrhenius techniques. Results of this analysis indicate that the plexiglass and general purpose phenolic have greater than forty (40) years expected life at a temperature of 259.8'F (includes heat rise) in the reactor building, and are, therefore, insensitive to thermal degradation for a period of 13 years.
The requirements of 1 x 10 rads gamma was enveloped for the relay by test.
The lexan bobbin was agt included in the test but has a i radiation damage threshold of 4.3 x 10 rads gamma and, therefore, is insensitive to the radiation requirements of 1 x 105 rads gamma.
Carolina Power & Light personnel conduct a visual inspection for cracking and/or melting of the spool (bobbin) in the magnetic coil assembly of all safety-related HFA relays on a monthly basis per Inspection Procedure MI-2Z. This action was recommended by General Electric Service Information Letter (SIL) Number 44, Supplement Number 2.
If degradation of any type is detected then CP&L replaces the magnetic coil assembly or entire relay, if required, with a General Electric Century Series Unit which is fully qualified to IEEE 323-1974 and to BSEP environmental parameters (
Reference:
Patel Report Number PEI-TR-83-04-16) .
l Therefore, continued operation is justified.
i
, P TER g NUMBER M AMP MODL'1 PIDG KYNAR Plant ID Number: Wire Terminatif.ns Component materials of the AMP PIDG terminations have been identified.
These materials have been evaluated per DOR guidelines and by applying Arrhenius techniques. Results of this analysis indicate that the"KYNAR insulated terminations have greater than forty (40) years demonstrated.
qualified life at the maximum drywell temperature of 150*F (
Reference:
PATEL Report Number PEI-TR-83-4-5) .
Therefore, continued operation is justified.
4 i
i i
TER ITD4 NUMBER 155 TERRY STEAM TURBINE HPCI PUMP DRIVE MODEL CCS Plant ID Number: E41-C002 An operational analysis has been performed on the Terry Steam Turbine Model CCS HPCI Pump Drive. Results of this analysis indidate
, that the HPCI turbine and auxiliaries could be subjected to a harsh temperature environment af ter a HPCI steam supply line break. However, no credit is taken for the operation of the HPCI turbine following a rupture of its own steam supply line. Therefore, safe reactor shutdown does not depend on the operation of t' 's device for a HPCI steam supply line break.
In the event of a small high energy line break (HELB), one for which the HPCI system can maintain the reactor primary vessel water level, the core is never exposed and hence core cooling is maintained aid no significant radiation exposure is experienced by the HPCI system.
This small break will cause a short duration accident temperature peak of 180*F approximately 35 minutes af ter accident initiation, decreasing to 125*F in less than three (3) hours. The HPCI system may be called upon to operate intermittently for a maximum duration of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter accident initiation.
In the event of a large LOCA, one for which the HPCI system cannot maintain the reactor primary vessel water level, the HPCI system may ,.
be subjected to high radiation exposure. However, in this case, the HPCI system is not required since the primary vessel will be depressurized by either the break or the actuation of the Automatic Depressurization System (ADS). Adequate core cooling is then provided by the low 1
pressure Emergency Core Cooling Systems (ECCS) acting in the place of the HPCI system.
In the event of a small break LOCA for which the HPCI can maintain reactor primary vessel water level, the core never uncovers and hence core cooling is maintained and the radiation environment is not present.
In this case, the temperature environment is limited to self-generated heat.
No common mode failures to date have been reported on the Terry S team Turbine HPCI system, even after many years of operational experience throughout the nuclear power industry.
In addition, a General Electric-lead Owners Group has recently completed a successful environmental qualification program on the Terry Steam Turbine Model CCS HPCI system.
Therefore, continued operation is justified.
i j TER ITEM NUMBER 156 SBGT MOTOR CONTROL, FARR COMPANY MODEL NUMBER D51423
)
Plant ID Number: SGT-FILT-2A-RB SCT-FILT-1A-RB SGT-FILT-2B-RB SGT-FILT-1B-RB This item is located on the 50-foot elevation of the reactor building.
The post-LOCA temperature profile in this area is a gradual increase from normal (maximum 104*F) to equilibrium at 128'Figapproximately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. The total integrated radiation dose is 10 rads for the 40 year life plus the accident.
This item was constructed in the early 1970's of high quality, heavy-duty components. Typically, such components will continue to i operate in the thermal environment described above.
The DOR Guidelines, Appendix C, gives a value of 105 rads TID for the majority of electrical components.
It can be concluded, based upon the above information, that these items will continue to operate during and af ter a LOCA event.
The SBGT cannnt, due to the lack of directly applicable qual-ification data, be assumed to remain operable in the more severe post-HELB environment, but as discussed below its operation la not necessary
- for this event.
The radioactive release from a HELB in the reactor building is substantially less than that assumed for the main steam line break which is released directly to the atmosphere and results.in much less site boundary dose than that permitted by 10 CFR 100.
Since the inventory loss pr'ior to isolation for a HELB is less than the main steam line break, the of fsit'e HELB dose is also correspondingly low even if the SBGT is not immediately operable. The HELB analyses for BSEP have shown that no fuel damage is expected as a result of the event. Therefore, there will be no excessive radiation levels in the reactor coolant when long term recovery from the event _is underway. -
Thus, there is no need for the SBCT system to maintain a negative pressure in the reactor building during recovery.
Therefore, continued operation is justified.
j .
TER ITEM NUMBER 157 BOSTON INSULATED WIRE T11ERM0 COUPLE CABLE Plant ID Number: TC16, XA16 Component materials of the Boston Insulated Wire thermocouple cable have been identified. These materials have been evaluated per DOR guide-lines and by Arrhenius techniques. Results of this analysis indicate that the cross-linked polyethylene insulation / neoprene jacket system used on these cables has greater than forty (40) years demonstrated qualified life at the maximum drywell temperature of 150*F (
Reference:
Pa tel Repo rt Numbe r PEI-TR-83-4-13) .
Therefore, continued operation is justified.
O H
m - -
TER ITLM NUMBER 158, 159 BOSTON INSULATED WIRE THERMOCOUPLE AND INSTRUMENTATION CABLE Plant ID Number: MA16, MC16, XA16, XE16, YL20 Component materials of the Boston Insulated Wire thermocouple and instrumentation cable have been identified. These materials have been evaluated per DOR guidelines and by applying Arrhenius techniques.
Results of this analysis indicate that the cross-linked polyethylene insulation / neoprene jacket system used on these cables has greater than forty (40) years demonstrated qualified life at the maximum reactor building temperature of 104'F (
Reference:
Patel Report Numbar PEI-TR-4 83-4-13).
Therefore, continued operation is justified.
I f
f- +?W tD
TER ITDi NUMBER 160 CERRO WIRE AND CABLE MODEL PYRO-TROLL III AND FIREWALL-EP (POWER CABLE AND SPECIAL CABLE)
Plant ID Numbers: BD10, BD06, VD16, JG16, Panel Wire Component materials of the Cerro cable have been identified. These materials have been evaluated per DOR guideltnes and by applying Arrhenius techniques. Results of this analysis indicate that the ethylene propylene rubber insulation and cross-linked polyethylene insulation have greater than forty (40) years demonstrated qualified life at 154*F (67.7'C),
including heat rise, for locations within the reactor building . This analysis is based on the primary conductor insulation used on the Cerro cables. There are no known synergisms between the cross-linked polyethylene or EPR insulation and the neoprene jacket (
Reference:
Patel Report Number PEI-TR-83-4-9).
Therefore, continued operation is justified.
9 l
.j
TER ITEM NUMBERS 162, 163 OKONITE POWER CABLE Plant ID Numbers: AC41 BB08 LA41 BD10 I BD06 HC25 JC25 JC50 Component' materials of the Okonite Power Cable have been identified.
These materials have been evaluated per DOR guidelines and by applying Arrhenius techniques. Results of this analysis indicate that the ethylene propylene rubber insulation has greater than forty (40) years demonstrated qualified life at 170*F (76.8'C), including heat rise, for locations within the reactor building. This analysis is based on the EPR insulation used on the Okonite cable. There are no known synergisms between the Okonite insulation and the Okoprene jacket. (
Reference:
Patel Report-Num be r PEI-TR-83-4-7) .
Therefore, continued operation is justified.
1 O
I
TER ITEM NUMBER 165 RAYCHEM FLAMTROL INSTRUMENT CABLE (600V)
Plant ID Number: NA16, RC16, FA26, GA22, IA22 Component materials of the RAYCHEM FLAMTROL INSTRUMENT cable have .
been identified. These materials have been evaluated per DOR guidelines and by applying Arrhenius tachniques. Results of this analysis indicate that the insulation and jacket materials (alkane-imide and crosslink'id polyolefin) have greater than forty (40) years demonstrated qualified life at the maximum drywell temperature of 150*F (
Reference:
PATEL Report Number PEI-TR-83-4-8) .
Therefore, continued operation is justified.
O O
j l
l l
t
TER ITEM NUMBr.R 167 SAMUAL MOORE MODEL DEKORON ECI WIRE (THERMOCOUPLE CABLE)
Plant ID Number: YA16, YC16, YE16, XA16, XC16, XE16 Component materials of the Samual Moore thermocouple cable have been identified. These materials have been evaluated per DOR guidelines and by applying Arrhenius techniques. Results of this analysis indicate that the EPDM insulation, hypalon conductor jacket, and overall hypalon cable jacket materials have greater than forty (40) years demonstrated qualified life at the maximum reactor building temperature of 104*F
(
Reference:
Patel Report Number PEI-TR-83-4-10) .
Therefore, continued operation is jusrtfied.
i i '
I l
1
\
l l
TER ITEM NUMBER 172 BURNDY HYLUG WITH OKONITE OKONEX AND NUMBER 35 TAPE Plant ID Number: Electrical Termination in the Reactor Building The Burndy Hylug is an uninsulated terminal lug made of pure copper and as such is insensitive to thermal or radiation degradation.
The Okonite Okonex tape is a butyl rubber tape which has an expected life in excess of 150 years at the maximum normal reactor building temper-ature of 104*F. The butyl rubber tape is exposed to the plant postulated accident temperature peak of 298'F for only two (2) minutes. The accident temperature then decreases to 125*F within one (1) hour of event initiation.
With an expected life of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> at 298'F, the tape is insensitive to thermai degradation for the required operating period.
The Okonite Number 35 tape is a jacketing tape and was qualified as such by Okonite Test Report NQRN-3.
Therefore, continued operation is justified.
t TER ITEM NUMBERS 173 and 174 WESTINGHOUSE PENETRATIONS (CLASS B, C, E AND F)
Component materials of the Westinghouse Class B, C, E and F penetrations have been identified. These materials have been evaluated per DOR guidelines and by applying Arrhenius techniques.
Results of this analysis indicate that the materials (silicone rubber-covered fiberglass, Sylgard 185, PVC, fiberglass-filled J' polyester, KYNAR, epoxy, cross-linked oolyethylene) have greater than forty (40) years demonstrated qualified life at the maximum ,
drywell temperature of 150*F.
This analysis is based on comparison of the postulated normal and accident conditions to the testing performed on prototype models of the Class B C, E and F Westinghouse penetrations designed specifically for use in the Brunswick plants. (
Reference:
Patel Repo rt Number PEI-TR-83-4-14.)
Therefore, continued operation is justified.
r .
l l
t l
L
TER ITEM NUMBER 176, 177 AND 178 THOMAS AND BETTS TERMINALS MODEL 54108, G971 and C1010 Plant ID Number: Wire Termination Component materials of the Thomas and Betts KYNAR (Refer ,--
ence: Print Number FP-9527-3407 and 3408) insulated STA-KON terminals have been identified. These materials have been evaluated per DOR guidelines and by applying Arrhenius techniques. Results of this analysis indicate that the KYNAR insulated STA-KON terminals have greater than forty (40) years demonstrated qualified life at the maximum drywell temperature of 150*F (
Reference:
Patel Report Number PEI-TR-83-4-4).
Therefore, continued operation is justified.
l l