ML20100F094

Forwards Revised Justifications for Continued Operation on safety-related Electrical Equipment.Equipment Failures Degrading Safety Functions Will Not Occur Due to Environ Resulting from DBA

ML20100F094
Person / Time
Site:	Brunswick
Issue date:	11/30/1984
From:	Cutter A CAROLINA POWER & LIGHT CO.
To:	Vassallo D Office of Nuclear Reactor Regulation
References
NLS-84-487, NUDOCS 8412060454
Download: ML20100F094 (57)

Text

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L Carolina Power & Light Company SERIAL: NLS-84-487 NOV 3 01984 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 NO. 1 DOCKET NO. 50-325/ LICENSE NO. DPR-71 ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT REVISED JUSTIFICATIONS FOR CONTINUED OPERATION Dear Mr. Vassallo By letter dated November 1,1984, Carolina Power & Light Company (the Company) was requested to review the justifications for continued operation (JCOs) for Brunswick-1 equipment qualification and confirm their acceptability on the same basis used in review of the Brunswick-2 JCOs. As a result of this review, the Brunswick-1 JCOs have been modified in order to more closely conform with review guidelines developed subsequent to their initial preparation. The revised JCOs are included as Enclosure 1.

The Company hereby confirms its belief that equipment failures which would mislead thi operator or significantly degrade safety functions will not occur due to the environment resulting from a design basis event. This belief is based in part on our review of the equipment including, where appropriate, consideration of:

1. Accomplishing the safety function by some designated alternative equipment.

2. Review of partial test data.

3. Limited use of administrative contrui .

4.- . Completion of the safety function prior to exposure of the equipment to an environment resulting from a design basis event which might degrade the equipment.

Also, the Emergency Operating Procedures provide cautions to the operator not to rely on any single indication and provides a list of instrumentation which may be used to verify the accuracy of suspect indications. The Company would like to note that, in general, JCOs have been submitted on equipment because of a lack of definitive documentation of qualification, not because the equipment is assumed to fail.

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Mr. D. B. Vassallo In addition, the Company hereby confirms its belief that:

1. In performing the review of the methodology to identify equipment within the scope of 10 CFR 50.49(b)(2), the following steps were addressed:

a. A list was generated of safety-related electric equipment as defined in paragraph (b)(1) of 10 CFR 50.49 required to remain functional during or following design-basis Loss of Coolant Accident (LOCA) or High Energy Line Break (HELB) Accidents. The LOCA/HELB accidents are the only design-basis accidents believed to result in significantly adverse environments to electrical equipment which is required for safe shutdown or accident mitigation. The list was based on reviews of the Final Safety Analysis Report (FSAR),

Technical Specifications, Emergency Operating Procedures, Piping and Instrumentation Diagrams (P& ids), and electrical distribution diagrams;

b. The elementary diagrams of the safety-related ciectrical equipment identified in Step a were reviewed to identify auxiliary devices electrically connected directly into the control or power circuitry of the safety-related equipment (e.g., automatic trips) whose failure due to postulated environmental conditions could prevent required operation of the safety-related equipment and;

c. The operation of the safety-related systems and equipment were reviewed to identify directly mechanically connected auxiliary systems with electrical components which are necessary for the required operation of the safety-related equipment-(e.g., cooling water or lubricating systems). This involved the review of P& ids, component technical manuals, and/or systems descriptions,in the FSAR.

d. Nonsafety-related electrical circuits indirectly associated with the electrical equipment identified in Step a by common power supply or physical proximity were considered by a review of the electrical design including the use of applicable industry standards (e.g.,

IEEE, NEFn, ANSI, UL, and NEC) and the use of properly coordinated protective relays, circuit breakers, and fuses for electrical fault protection.

2. Design basis events which could potentially result in a harsh environment, including flooding outside containment, were addressed in identifying safety-related electrical equipment within the scope of 10 CFR 50.49(b)(1).

3. Electrical equipment within the scope of 10 CFR 50.49(b)(3) is R.G. 1.97 Category 1 and 2 equipment or that justification has been provided for any such equipment not included in the environmental qualification program.

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.t Mr. . D.'B. Vassallo' w _

Should you have any questions regarding this issue, please contact

- Mr. Sherwood Zimmerman1(919) 836-6242. .

Yours very t 1, s: 7.,.

-.. A. B.'!: utter - Vice resident Nuclear Engineering & Licensing MAT /pgp (882 MAT)

Enclosyre s

i cc: Mr. D. O. Myers (NRC-BNP)'

.Mr. J. P. O'Reilly . NRC-RII) (

Mr. M. Grotenhuis (NRC)

A. B. Cutter, having been first duly sworn, did depose and say that the information contained herein le true and c6 erect to the best of his information, knowledge and belief; and the sources of his information are officers, employces, contractors, and agents of CarolinkT'?ouer & Light Company. ,

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Enclosure 1 to'NLS-84-487 Justifications for Continued Operation e

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' EQUIPMENT LIST)

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Brunswick --

=NRC Ter No.- Tag-No.

Description 10CFR Criteria

!JC0 No. .

.p p-ee ,

2 i7- ~E51-F019~ Limitorque-Motorized (2)

,, Valve. Operator Y

3 20.' B21-F016 . Ilmitorque bbtorized (2) -

Epi-F022 " Valve Operator

[' ;251-F007 l G31-F001.

y-($. ;24 th'ru.30 .B32-F020-S- 'ASCO Solenoid (2)

'33 thru'39 CAC-PV-1260-S s

- -.41,42,44

. 3 CAC-PV-1261-S j -

x '

45,47,48 CAC-PV-1262-S.

' ' ~ '

49,50,52 'CAC-PV-3439-IS ,

53 ~ - CAC-PV-3439-S CAC-PV-3440-IS

-CAC-PV-3440-S

+ ,

'CAC-SV-4222:

CAC-SV-4223' E

^: > - , CAC-V4-S-CAC-V5-IS

, CAC-VS-S

'CAC-V6-IS

'CAC-V6-S CAC-V7-S

~

-CAC-V8-S '

'CAC-V9-S-

'CAC-VIO-S CAC-V15-S

.CAC-V47-IS CAC-V47-S CAC-V48-IS CAO-V48-S.

CAC-V49-S

~ CAC-V50-S-CAC-V55-S s

CAC-V56-S C11-F110A C11-F110B

~

.E11-F053A-E11-F053B G16-F003-S G16-F004-S

'~

C16-F019-S G16-F020-S

$ I iLl-a

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$ Brunswick:

'10CFR Criteria

~

fJCO ' No'. NRC Te'r No. Tag No. Description 15(Cont'd)I '

RIP-SV-1200Al--

1233B2 (150. VALVES)

- 11A-BFIV-RB-S "1B-BFIV-RB-S:

IC-BF8V-RB-S ID-BFIV-RB-S 16' '

34,113,114p VA-TS-936A- Johnson Services ~(1),(5)

-123 thru.F' .

Temperature Switch

'VA-ZS-936A,B-

"9 .

VA-SV-936A,B

, .9. 62 E41-PS-N010 Static-0-Ring _ (2)

.6 ,

JE51-PSL-N006 Pressure Switch o

. .10 : CAC-PT-1257-2 .. Bailey Transmitters (1) 24 -

111 68' C32-PT-N005A,B General Electric  :(2)

~* ~

~

Pressure Transmitter

' 12- .69 .E11-PDT-N002A,BL General Electric (1),(5)

-Pressure Transmitter r71 thru 74, Barksdale Switches (2)

~

<- ~ 13' w. E11-PS-N016A!

76.thru 81,

- _ thru - D

'r 997 E11-PS-N020 A ,

thru D

' RIP-PSL-1200.

' RIP-PSL-1201 4

^ ' ' - -RIP-PSL-1203;

' RIP-PSL-1205.

' . RIP-PSL-1206.

RIP-PS2-1208' thru 12-RIP-PSL-1215 RIP-PSL-1217,

^

. thru'23

, . RIP-PSL-1225

-, thru.29-RIP-PSL-1231

thru 32-

- E41-PSH-N012A u thru D E41-PSH-N017A,B E51-PS-N020

• " 'E51-PSH-N009A,B E51-PSH-N012A-

- thru D.

B32-PS-N018A,B

'B32-PS-N018A-1 SW-TSH-1109 thra.1112 rv - _ 3

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i Brunswick JCO No.- NRC Ter No. Tag ' No. Description 10CFR Criteria

/

13 (Cont'd) IA-PSL-3594,3595' E41-PSH-N027 0 .SW-PS-1175,1176 15 82 E41-LSH-N015A,B . Robertshaw (2)

Level Switches' l17: 91- B21-FS-F015 A. Magnetrol Flow Switches (2) thru H B21-FS-F015 J thru N B21-FS-F015P B21-FS-F015R,S B21-FS-F043A,B B21-FS-F045A,B B21-FS-F047A,B B21-FS-F049A,B B21-FS-F051A,B

> B21-FS-F057 B32-FS-F009C

.j . , .B32-FS-F0408,C B32-FS-F055E,F B32-FS-F057A,8 ,

E41-FS-F024A thnt D E51-FS-F044A

-thru D 4

18 .93 VA-FT-2577 Bailey Transmitters (2),

t

19 94,122 ,

.B21-F014A thru H Cherry Switches (2).

B21-F014J thru N B21-F014P-B21-F014R

, .B21-F0145

- r .B21-F042A,B B21-F044A,B

, 'B21-F046A,B

-B21-F048A,B B21-F050A,B B21-F056

.B32-F0398,C B32-F041C B32-F056E,F-B32-F058A,B CAC-PV-1218C CAC-PV-1219B,C CAC-PV-1225C E11-F037A thru D

E41-PV-1218D

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. .:.., . . . .. , .,a.- .- . - . . . . . . , . . - , . - . . - . - . - . . - . . . _ - , . . . . - - - - . - . , , . . . -

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e Brunswick.

r, 'JC0 No.- LNRC Ter No. Tag No. Description -10CFR Criteria

"'I '19 (Cont'd) -E41-PV-1219D-1 . , .

sE41-PV-1220D E41-PV-1221D.

E41-F023A thru D

%. _ LE51-F043A thru D

=

.IA-PV-1201A CAC-PV-1209D

~

E11-F043A thru D 20- 95' E41-FT-N008 General Electric. (2)

. Flow . Transmitter y.

121' 96,97,98 )E11-PDIS-N021A',B Barton Differential .(1),(2),(5)

E21-FS-N006A,B Pressure Switches-

-E41-FSL-N006

'22 - NONE' E51'-FS-N002

• Barton .Dif ferential (2)

Pressure' Switch -

. 23 100. CAC-TE-1258-1 Pyco Temperature (2) thru 14 Elements CAC-TE-1258-17

-thru 24 '

24- 107,108~ E41-TS-3314. Fenwal Temperature (1),(2) 110 thru thru 3318 Switch 112 E41-TS-3354

- - E41-TS-3488 E41-TS-3489' E51-TS-3319 thru 3323 E51-TS-3355 E51-TS-3487 25' -109 'B21-TS-N010A Fenwal Temperature '(2),(4) ,

thru DL Switches

- 26 115' 1(A-D)-BFIV-RB-L NAMCO Position Switch. (2)

,  ; 28 124,'125,126 B32-F019-L: - Honeywell Limit (2),(4),(5)

, 127,128,129 .B32-F020-L Switches CAC-V47-L CAC-V48-L

, CAC-V55-L

/3 - CAC-V56-L-CAC-PV-1200B CAC-PV-1205E-L1

'CAC-PV-1205E-L2 CAC-PV-1209A-L1 CAC-PV-1209A-L2 n

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10CFR Criteria JCO-No. ' NRC Ter !k). . Tag No. Description ,__.

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s. 4 L28 (Cont'd CAC-PV-1209B-

- -CAC-PV-1211E CAC-PV-1211F-L1' a;- E 'CAC-PV-1211F-L1-

~CAC-PV-1215E CAC-PV-1225B CAC-PV-1227A-L1

, CAC-PV-1227A-L2-CAC-PV-1227B-L1 CAC-PV-1227B-L2 CAC-PV-1227C CAC-PV-1227E-L1 CAC-PV-1227E-L2 CAC-PV-1231B CAC-PV-1260-L CAC-PV-1261-L CAC-PV-1262-L CAC-PV-3439-L '

CAC-PV-3440-L

'l B21-F003-L

- B21-F004-L

'29- 130,131,133 B11-RS Honeywell Rdcroswitches (2) 134,135 B11-RS1 B21-CS-3327~

~

B21-CS-3329 B21-CS-3345 B21-CS-3412

- ' B41-RS B41-RSI B43-RS B43-RS1-B45-RS B45-RS1 B46-RS w B46-RS1-B47-RS'

.B47-RS1 B49-RS B49-RS1 B50-RS B50-RS1 DE3-RS-CS DE3-RS-SS' DH2-RS DH2-RS1 DH3-RS DH3-RS1

• DK8-RS DK8-RS1 s

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Brunswick!

- JC0 Nb. -NRC'Ter No.: Tag'No. -Description- -10CFR Criteria (Con t 'd) DK9-RS DK9-RS1-DLO-RS~

DLO-RSl-~

'DL1-RS

'DL1-RS1 y

DL2-RS_. -

.DL2-RS1

' ' ~

DL7-RS.

403 = t;: DL7-RS1 #

~

> DL8-RS.

'- , DL8-RS1

-DL9-RS DL9-RSI-DM1-RS o

/SI '

4 DM1-RSI ~

4 -

D MG-RS

~ DM2,-RS l'-

DM4 RS DF4-RSI.

D MS-RS 7

DF5-RS1 DFC-RS

Dh7-RS1 -

DFS-RSL DM8-RS1 DN1-RS DN1-RSI-

'DN6-RS 1

1 6- DN6-RS1 o

'DN9-RS

" DN9-RSI-DP2-RS DP2-RS1 Vf' DP3-CS' DPS-RS-A-SS

- DPS-RS-CS

,9/ DS4-RS

DS4-RS1 FN6-CS s

(30? 132,142,144 ->CC-lXA Ceneral Electric IC7700 (2)

Mbtor Control Center

'145,146,147 MCC-lXA-2 NCC-lXB

>E C-lXB-2 HCC-lXC lin MCC-1XD.

!A ' MCC-lXDA MCC-lXDB NCC-lXE

7-t Brunswick 1 . ' JC0 No. 'NRC Ter No. Tag No. , Description 10CFR Criteria

'30'(Cont'd)^~

MCC-IXF L -

MCC-1XG MCC-1XH Bil-B09-RS GM8-3-6A

. GN2-3-5A

31 138 - E11-0001A thru D General Electric (2)

Motors

32 141,155 E41-C002 Terry Steam Turbine (1)

J16-TB-B thur D HPCI Pump System

33 143 DB0-74-18 Agastat Time (1)

Delay Relay

34 148 D12-RE-N010A,B General Electric (1),(2),(3),(4)

Radiation Detectors 36 156 NG7-SGT-FIL.T- Farr Standby Gas (1),(5) 1A-RB Treatment System

. NG8-SGT-FILT- Components

-1B-RB 40- 179,181- Terminal Blocks General Electric (2)

Terminal Blocks

41_ 182 Terminal Blocks Curtis Type "L" -(2)

Terminal Blocks

'~

42 NONE C11-F010-L Namco Limit Switch (2),(4)

E51-0002-LS4

'43 NONE NP6-MOT-M1,M2 DOERR Motors and (3)

NP7-MOT-M1,M2 Control Panels 1A-RX,;1B-RX

-44 NONE E51-0002-H- Square D (2)

Float Switch

45 ; NONE B32-CS-F019 Sentry Control Switch (2)

~ B32-CS-F020

-46 NONE B21-FT-4157 NDT International (1)(2)

- thru 4167 Accelerometers 12(E-N)-SPLICE I2P-SPLICE

Trs- 1^'

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r/B .

' Tag No.' Description 10CFR Criteria.

. JC0 No. NRC Ter No.

A f481 ' NONE ^ QC4-P10-A,B Culton Connector- . (2)

QC4-P10-NR,Ps Receptacles and Plugs

.QC4-P2 A,B QC4-P2-MA, NB

- QC4-P3-A,B'

- QC4-P3-MA, NB .

. QC4-P4-A,B QC4-P4-MA,NB-QC4-PS-A,B-QC4-P5-ta,bs- ,

QC4-P9-A,B QC4-P9-MA, MB QC7-P10-A,B

- QC7-P10-NR,NB QC7-P6-A,B:

QC7-P6-MA, NB QC7-P7-A,B QC7-P7-NR,NE QC7-P8-A,B

^

QC7-P8-MA, MB QC7-P9-A,B.

QC7-P9-MA,MB-49 NONE -

1-IG7-TP5 HPCI. System (4) thru TP8 Test Points

...i 50 'NONE .XS4-DS5. Standby Cas '(4)

XS4-DS6 ' Treatment System Ps .:

XS4-MAN.0VRD.SW. Components

/ XT1-DS11l XT1-DS12 XT1-nan.0VRD;SW.

XTO-DS10

^

XIO-DS9 XTO-MAN .0VRD . SW.

X02-DS7

-X02-DS8 X02-MA7.0VRD.SU.

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UNIT 1 BSEP

^J- JC0 NO. 2 I .17 TER NO.: .

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COMPONENT :I.D. NO.': E51-F019 '.

- MFG / MOD. NO.:' LIMITORQUE MODEL SMB-000 VALVE OPERATOR

. LOCATION: REACTOR BUILDING -17 0 4 TEC}eIICAL DISCUSSION:

Component materials of the Limitorque Motorized Valve Operator'have'been identified and qualification-documentation located. The qualification data has been evaluated per DOR guidelines and by applying Arrhenius techniques. .

Results of this evaluation indicate that_ the Class B motor insulation system, ,

melamine switches, and internal wire insulation materials are insensitive to thermal aging effects at the maximum reactor building temperature of 104'F.

The valve operators and motor nonmetallic materials are exposed to the plant

postulated accident profile which shows a peak temperature of 288*F for 70 seconds, and then drops to 205*F after 100 seconds.

The_ valve operator is fully qualified for 40 years at the normal and accident t reactor building parameters (

Reference:

Limitorque Test Report No. 600376A).-

- The Class B motor insulation system has been successfully tested at 250*F- for

- 22.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (Raference: Limitorque Test Report No. B0003). - A comparative ~

analysis of.the Limitorque "Superheat" test reveals that the internal _

r . temperature of the valve operator and motor will not reach 250*F during the j,

initial 100 seconds of. accident exposure. Thus, it is judged that the test temperature profile was actually more severe that the plant requirement.

l ,

This-analysis meets the criteria of 10CFR50.49, paragraph (1)(2).

f

!- .Therefore, continued operation is justified. -

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14

UNIT 1 '

~

BSEP 0 . 4C0 NO. 3

.TER- NO. : . 20 COMPONENT I.D. NO.~: 'B21'-F016, E11-F022, E51-F007,~ G31-F001 MFG / MOD. NO.: LIMITORQUE MODEL SMB-00 VALVE OPERATOR

' LOCATION: DRYWELL ELEVATIGI 17',_80', 50' TECl#IICAL DISCUSSION:

t,

',- Component materials of the Limitorque Motorized Valve Operators havi been

identified and qualification documentation located. The qualification data has been evaluated per DOR guidelines and by applying Arrhenius techniques.

Results-of this evaluation indicate that the Class H motor insulation system,

, malamine switches, and internal wire insulation materials are insen'sitive to thermal aging effects at the maximum drywell temperature of-150*F. The valve operator-and motor nonmetallic materials are exposed to the plant postulated

.. accident -profile which shows a peak temperature, of 298'F.

The valve operators are qualified for 40 years at the normal and a_ccident

. drywell. parameters' . (

Reference:

Limitorque Test Report No. 600376A). ,

The motor, with Class H insulation, has been successfully tested to a peak

. temperature of 340*F (

Reference:

Franklin Report No FC-3441) which exceeds the postulated plant accident at BSEP. -

motors were successfully tested to 2 X 10jditionally, rads gammathe Class total H insulated integrated dose

(

Reference:

. Limitorque ' * "" '

8 requirement of 1.1 X 10 Rads gamma.

b Thus,11t _is judged that the Class H insulated motors. meet the criteria set l forth- in 10CFR50.49, paragraph (1)(2). .

Therefore, continued operation'is justified.

L i

UNIT 1 BSEP JC0 NO. 5 TER NO.: 24, 25, 2'6, 27, 28, 29, 30, 35, 36, 37, 38, 39, 41, r;, 42, 44, 45, 47, 48, 49, 50, 52, 53

. COMPONENT.I.D. NO.: B32-F020-S CAC-V47-IS CAC-PV-1260-S CAC-V47-S CAC-PV-12617S CAC-V48-IS CAC-PV-1262-S CAC-V48-S CAC-PV-3439-IS* CAC-V49-S CAC-PV-3439-S* car-V50-S CAC-PV-3440-IS* CAL //55-S

> CAC-PV-3440-S* CAC-V56-S. ,

CAC-SV-4222 C11-F110A CAC-SV-4223 C11-F1108

. CAC-V4-S' E11-F053A CAC-V5-IS E11-F0538

. CAC-VS-S G16-F003 -S

CAC-V6-IS G16-F004 -S

^

CAC-V6-S G16-F019-S CAC-V7-5 s G16-F020-5 CAC-V8-S RIP-SV-1200Al-123382 (150 VALVES)

CAC-V9-S 1A-BFIV-RB-S CAC-V10-S 18-BFIV-RB-S CAC-VIS-S IC-BFIV-RB-S 1D-8F1Y-RB-S

• NO TER MFG / MOD, NO.: H88302C25RU JV-182-084 HT8316 HT8211833 'HT8262C71 8302 HT8321A6 WPHT8321A1 8321A6 4 HV-180-414' H8342A4 8262D23 The "HT" AND "HB" ' prefixes denote high temperature

- coils with class "H" insulation and are rated for continuous use at-165*F ambient temperature.

additionally, documentation for the model 8302 indicated a class "H" was supplied LOCATION: RHR ROOM, CORE SPRAY ROOM, AND REACTOR BUILDING TECHNICAL' DISCUSSION:

Component naterials of the ASCO solenoid valves have been identified and qualification d' ocumentation located. The qualification data has been evaluated per DOR guidelines and by applying Arrhenius techniques. Results of this evaluation indicate that all the nonmetallic materials, except Buna-N,

- have greater than 660 years expected life at the maxinum 104*F temperature.

The Buna-N has an expected life of 11.86 years.

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-i j-; UNIT 1 7 BSEP

, , JCO NO. 5

- TER-24-53'. -

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Page:2 I'

a In a -letter dated 8-3-79, ASCO stated the following about model numbers HV180-414 and JV182-084:

"The materials.used in- the construction of these valves are brass' bodies, zinc plate steel bonnets, Buna-N (Nitrile) elastomers, copper shading coils, and all additional internal components are 302, 17-7PH, 305,-416, 430F stainless steel and monel. The valves have Class "H" coils and Nema b ' Type 4 solenoid enclosures.

Based on Engineering judgement, test of similar valves, experience, and rubber manufacturer s literature, the elastomeric components utilized in these' valve will function satisfactorily under the accident and post-accident conditions specified in the UE3C Specification. The Class 'H' coils. utilized in these valves have been designed for satisfactory operation at 165*F ambient.

Yalves of similar design utilizing the said Class 'H' coil system and ethylene propylene elastomers have been satisfactorily qualification tested for use inside containment in accordance with the requirements of IEEE 323-1974, 383-1972, and 344-1975. Part of this test program was a thermal aging test during which the valves were exposed to an ambient temperature voltage and de-energized for 5 minutes every 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. At the completion of this test, the valves functioned satisfactorily with no internal or external leakage. The results of this testing are recorded in ASCO test report AQS21678/TR. Ethylene propylene was chosen as the elastomer in these valves because they are for use inside containment and

!. it is expected that during an accident the temperature could rise .to a maximum of 346*F. Since the coils passed the 12 day exposure at 268'F, and rubber manufacturer's literature recommends Buna-N for use at 200*F continuous, it is our opinion that this is justification for stating that i .

these valves are capable of satisfactory operation during the accident and l\

post-accident conditions stated in the UEAC Specification".

Although ethylene propylene was the elastomer in the tested valves, the actual service condition of total time above 200*F of less that 3-minutes followed by a rapid drop off to approximately 135*F for these solenoid at Brunswick is such that Buna-N is an acceptable material.

[ There* is also a Rockwell test report (2972-03-02, Rev.1; dated 12-1-70) which shows that the.HTX8320A20 had successfully functioned during and after exposure to 345' and 110 psig steam for about 2-1/2 hours.

Additionally, a Masoneflan test report (1003, dated 4-19-73) shows that WPHT83008S1 valves successfully functioned during and after exposure to 310*F l and 65 psig steam for 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />.

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• UNIT 1 BSEP-JC0 NO. 5'

- TER-24-53a .

' Page 3 In 10{ormation on' radiationdegrade will not sipificantly damage_values shows that the function thenonmetallic of the postulated TID of 1 X materials except for the, acetal disc p1 der. Testing has been performed on acetal retaining washers to 1 X 10 rads with successful resdits (

Reference:

MCC Powers' Report No. 734.79.002,~ Rev. 3).

. This analysis meets the criteria' of 10CFR50.49, paragraph (f)(2).

~ Therefore, continued operation is justified.

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• UNIT 1

.~ ~ BSEP JC0 NO. 6 ,

-TER NO.: -

34, 113, 114,

~

123 a .

COMPONENT lI.D. NO.i VA-TS-936A, B, C, D,'E, F VA-ZS-9368, A '

VA-SV-936B,'A MFG /M00. NO.: " JOHNSON SERVICES; ALLEN BRADLEY LOCATION: .RHR ROOM

-TECHNICAL DISCUSSION:

0 The operation of the RHR Pump Room Cooling Systems .has' been reviewed. _ In the

' event of room A fan cooling unit failure the room B fan cooling unit will ,

- supply the post-LOCA cooling requirements of both RHR pump rooms and the HPCI room simulataneously via interconnecting HVAC ductwork.

Tie room B fan cooling unit equipment (VA-TS-9368, C, F; VA-ZS-936B; VA-SV-936B) is currently being replaced with fully qualified equipment. This completes the qualification of this redundant backup system.

This analysis meets the criteria of 10CFR50.49, paragraph (1)(1) and ('i)(5).

Therefore, continued operation is justified. ,

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UNIT .1 BSEP

• ~

JC0 N0. 9

TERl NO.: 62

' COMPONENT I.D. NO.:~ E41'-PS-N010'.

E51-PSL-N006 (N0 TER)

MFG / MOD. NO.: . STATIC 0 RING PRESSURE SWITCH 6N-AA21X9SVTT AND 0: 6N-AA21-X9-ST LOCATION: REACTOR BUILDING EL. -17' TECM ICAL DISCUSSION:

Component materials of the Static-0-Ring (SOR) pressure switch have been '

identified and qualification. documentation on a similar SOR pressure switch has-been:obtained.- The qualification data has been evaluated per D0R guidelines and by applying Arnhenius techniques. Results of this evaluation

' indicate that the lowest expected life of any nonmetallic material used in the pressure switch is 11.86 years.

-The pressure switch nonmetallic materials are exposed to the plant postulated accident temperature peak of-288*F fo'r 70 seconds. The accident temperature then decreases to 205*F at 100 seconds and return's to ambient after approximately.20 minutes. This postulated peak temperature transient has been compared to accident test data obtained (212*F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) for this switch.

Though the testing does not envelop the postulated peak accident temperature, it is judged that no significant detrimental effects to switch operation

~

.should occur as a result of the peak temperature transient. This assessment

~1s based on the severity of the test performed in comparison to the short duration of the temperature transient (

Reference:

Viking Lab Report No.

30203-2).

Additionally, a radiation analysis was performed to determine the threshold of each nonmetallic material used in the pressure switch. It was determined that each material has a radiation tgreshold greater than the maximum postulated total integrated dose of 2 X 10 rads gamma.

.This analysis meets the criteria of 10CFR50.49, paragraph (1)(2).

.Therefore, continued operation is justified.

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UNIT 1

. BSEP JC0 NO. 10

- TER NO.: . 67

. . - ~

COMPONENT I.D. NO.: CAC-PT-1257-2'

. MFG /M00. NO.: BAILEY KQ12C LOCATION:. RHR ROOM TEcleIICAL DISCUSSION:

The,information provided the operator by these transmitters is also provided by an independent, redundant, and fully qualified transmitter (Rosemount). As ,

such the safety. function of this equipment can be accomplished by alternative

. equipuent.

~

This analysis meets the criteria of 10CFR50.49, paragraph (1)(1).

-Therefore, continued operation is justified.

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UNIT 1 BSEP JC0 NO.11 TER NO.: ,

68 -

. COMPONENT I.D. NO.: 'C32'-PT-N005h.B

. MFG /M00.~ NO.~:- GENERAL ELECTRIC MODEL 551032GKZZ2 PRESSURE TRANSMITTER LOCATION: REACTOR BUILDING 50'

.TECMICAL DISCUSSION: -

Partial ' qualification documentation has been obtained for a 'similar pressure transmitter with the same components and of similar application. The data was

- e'sluated per the D0R guidelines.

The pressure transmitter measures the RPV pressure and gives the operator information regarding plant performance.

Testing has been successfully conducted to show that the device will not fail catastrophically under elevated temperature and humidity conditions

(

Reference:

General Electric Document NSE80036). The accident simulation included a peak temperature of 180*F during which time a 6 point calibration

. functional test was performed. This was estimated to take about 5 minutes.

-Additionally, a s'parate e test subjected the transmitter to a 68'F to.158'F at -

3" 100% RH test. The tests do not envelop the BSEP requirement of 200*F for 40-50 seconds and the subsequent ramp down to 150*F in 8 minutes. However, the accident peak temperature excursion will not cause significant degradation of equipment operation during that period of exposure above the test maximum temperature (

Reference:

General Electric Report No. 327, File DV145C3007 and General; Electric Document No. NSE80036).

Adgitionally, analysis indicates that the plant radiation requirement of 1 X rads gamma is less than the lowest radiation damage threshold of the L 10 transmitter components.

l This analysis meets the criteria of 10CFR50.49, paragraph (i)(2).

-Therefore, continued ope. ration is justified.

l

• I l

~ UNIT 1 -

, BSEP JC0 NO. 12 TER NO.: ' 69 - ',

COMPONENT I.D. NO.: E11-PDT-N002A, & B MFG /M00. NO.: GENERAL ELECTRIC 552032HKZZ2 -PRESSURE TRANSMITTER LOCATION:' REACTOR BUILDING RHR ROOM

. TECHNICAL DISCUSSION: -

These instruments measure the pressure across t'he RHR heat exchanger and provide a signal to the RHR service water outlet valve to regulate service water pressure so it is always greater than RHR system pressure. This function can be manually overridden if necessary, and the plant can be safely shutdown in the absence of these devices.

This analysis meets the criteria of 10CFR50.49, paragraph (1)(1) and (1)(5).

Theref. ore, continued operation is justified.

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+

UNIT 1 BSEP .

JC0 NO. 13 TER NO.:'

71, 72, 73, 74, 76, 77, 78, 79, 80, 81, & 99 -

~

COMPONENT I.D. NO.: " E11-PS-N016A RIP-PSL-1231* RIP-PSL-1220 E11-PS-N0168 RIP-PSL-1232* RIP-PSL-1221 E11-PS-N016C E41-PSH-N012A RIP-PSL-1222 E11-PS-N016D E41-PSH-N012B RIP-PSL-1223 E11-PS-N020A E41-PSH-N012C RIP-PSL-1225 E11-PS-N0208 E41-PSH-N0120 RIP-PSL-1227 E11-PS-N020C E41-PSH-N017A RIP-PSL-1228 r E11-PS-N0200 E41-PSH-N017B RIP-PSL-1229 RIP-PSL-1200 E51-PS-N020 B32-PS-N018A RIP-PSL-1201 E51-PSH-N009A B32-PS-N013A-1

-RIP-PSL-1203* E51-PSH-N0098 B32-PS-N018B RIP-PSL-1205*

E51-PSH-N012A SW-TSH-1109 RIP-PSL-1208* - E51-PSH-N0128 .SW-TSH-1110 RIP-PSL-1209 E51-PSH-N01EC SW-TSH-1111

RIP-PSL-1210 E51-PSH-N0120 SW-TSH-1112 RIP-PSL-1211 RIP-PSL-1218 IA-PSL-3594,3595*

RIP-PSL-1215* RIP-PSL-1219 E41-PSH-N027 RIP-PSL-1217 RIP-PSL-1226* RIP-PSL-1206 SW-PS-1175,1176* RIP-PSL-1212.

.

• NO TER

.. MFG / MOD. NO.: BARKSDALE- B2T-M12SS D2H-M150SS D2T-M18SS D2T-M150SS P1H-M340SS TC9622-1 T2H-M2515-12 D2T-M80SS LOCATION: REACTOR BUILDING, RHR ROOM, CORE SPRAY ROOM TECHNICAL DISCUSSION:

. Component materials of the Barksdale switches have been identified and qualification documentation located. The qualification data has been evaluated per DOR guidelines and by applying Arrhenius techniques. Results of i this evaluation indicate that all materials, except for Buna-N rubber, have greater than 261 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 288'F for only 70 seconds. The l accident temperature then decreases to 145*F within one (1) hour of event l initiation. This postulated peak temperature transient has been compared to accident test data obtained (212*F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,: Ref. AETL TR #596-0398) for those switches. Although the testing does not envelop the postulated peak accident temperature, it is judged that no detrimental effects to switch operation should occur as a result of the peak temperature transient. This is' based on the severity of the test performed and the short period of switch exposure to the accident peak temperature.

r

[ In addition, the Brunswick switches are located in NEMA 3, 4,12, or 13 enclosures where tne effects of direct steam impingement / humidity would be reduced to nil during the postulated accident.

l l

UNIT 1 BSEP

~

JC0 NO.13

- TER-71-99 .

' Page 2- . - - -

a

~

- Also, the component nonmetallic materials' have been sucessfully radiation aged a during-qualification testigg (while being used in similar applications) to

- levels greater than 1 X 10' rads gamma, the postulated accident TID for BSEP.

This analysis meets the criteria of 10CFR50.49, paragraph (i)(2).

Therefore, continued operation is. justified.

9 9

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1 UNIT 1 BSEP 1

. JC0 NO.15 L J TER NO.:~ 82

. . . - < s COMPONENT- I.D. NO.:. E41-LSH-N015A, B'

, MFG / MOD.-NO.: ROBERTSHAW MODEL SL-205-A2-R11-B11-1 LEVEL SWITCH LOCATION: REACTOR BUILDING -17' TECHNICAL DISCUSSION:

Partial qualification"docume,ntation has been ' located for the Roberts' haw level switches. The qualification data has been evaluated per. D0R guiaeiines and by applying Arrhenius techniques.

. The switch nonmetallic components are exposed to the the reactor building

. postulated accident temperature peak of 288'F for only 70 seconds. In addition, the Brunswick switches are located in a Nema Type 7.9 explosion

, proof ~ enclosure where the effects of direct steam impingment/ humidity would be reduced to' nil during the postulated accident. The accident temperature requirement then decreases to 145'F within one (1) hour of event initiation.

This postulated peak temperature transient has been evaluated and co, pared to the; accident test data obtained (212*F, 10 psig for 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />,

Reference:

Robertshaw unnumbered test report dated March 28,1983) for these switches.

Although the testing does not envelop the postulated peak accident ,

i temperature, it is judged that no detrimental effects to switch operation will occur as a result of the peak temperature transient. This is based on the severity of the test performed and the short exposure time oF the level

. switches to the 288'F accident peak.

~ Alsor the component nonmetallic materials have been successfully radiation aged during qualification testing (while being used 19 similar applications) to levels greater than the BSEP requirement of 1 X 10 ra'ds gamma.

' Operationally, the level switches located outside containment are used to signal high suppression pool level to the HPCI system. -

In the event of a large break LOCA for which the HPCI system cannot maintain RPV level, the switch may be subject to high radiation. However, in this case

, the HPCI system is not required since the RPV will be depressurized by the break _ and/or actuation of the ADS system. Adequate core cooling is then. ,

provided by the low pressure ECCS systems and safe . shutdown does not depend on the operation of this device.

In the event of a small break LOCA for which the HPCI syster; can maintain RPV

j. . level, the core never. uncovers and hence core coooling is maintained and the radiation environment is not present. The switch will perform its function

prior to an environmentally caused failure since the peak temperature reaches only.145*F.-

,, v-, .,..-,,,m...,,,.., .,m,,,., ,. ~ . , , .. ..,_,...,._-,,..,-,.______,-_,.--,,---,.U

tMIT 1 -

, BSEP JC0 NO. 15

.TER Page 2 , , s The 288* environment in this area.of the reactor building is due to the HELB'.

event. The function of'these switches is to transfer the HPCI suction from the condensate' storage tank to the _ suppression pool on a high suppression pool

. level condition. Since neither the HELB nor the actions required to mitigate

an HELB~will result in a high suppression pool level and HPCI system operation at the same time, this function is not needed -to mitigate an HELB.- ,,

This analysis meets the criteria of 10CFR50.49, paragraph (i)(2).

Therefore, continued operation is justified.

+,- _

e 99 k

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m.

UNIT 1 BSEP~

~

JCO. NO.17

~

ITER NO.:- ' 91 . .

COMPONENT 'I.D. NO.:~ Bk1-FS-F015A B21-FS-F015N -821-FS-F051A E41-FS-F024C B21-FS-F0158 B21-FS-F015P B21-FS-F051B E41-FS-F0240 B21-FS-F015C' B21-FS-F015R B21-FS-F057 E51-FS-F044A-B21-FS-F0150 B21-FS-F015S B32-FS-F009C 'E51-FS-F044B '

B21-FS-F015E ~821-FS-F043A B32-FS-F0408 E51-FS-F044C B21-FS-F015F B21-FS-F0438 B32-FS-F040C E51-FS-F044D

- B21-FS-F015G B21-FS-F045A .832-FS-F055E B21-FS-F015H B21-FS-F0458 B32-FS-F055F

' B21-FS-F015J B21-FS-F047A B32-FS-F057A '

B21-FS-F015K B21-FS-F047B B32-FS-F057B B21-FS-F015L. B21-FS-F049A E41-FS-F024A

. B21-FS-F015M B21-FS-F0498 E41-FS-F024B MFG / MOD. NO.: MAGNETROL MODEL F-521 FLOW SWITCH

-LOCATION: REACTOR BUILDING (VARIOUS ELEVATIONS)

TECiflICAL DISCUSSION:

ComponAnt materials of the Magnetrol flow switch have been identified. These materials have been evaluated per DOR quidelines and by applying Arrhenius -

-- techniques. Results of the analysis indicate that the nonmetallic components- *

have. greater than 47.6 years of expected life at the maximum reactor building "

temperature of 104*F.

'A flow switch of similar design and materials was tested to conditions more severe than the postulated conditions at BSEP for temperature, pressure and relative humidity (

Reference:

Barton Reports R1-288A-11 and R3-288A-1).

Additionally, a radiation analysis has been performed on each nonmetallic i

material . used in the flow switch. The analysis indicated that each material t

has a radiation damage threshold level equgl to or greater than the maximum

! postulated total integrated dose of 1 X 10 rads gamma.

In addition, an operational analysis has been performed to determine the

. effects of failure '(misleading information, grounds and spurious operation) of these items in both LOCA and HELB environments. The operational analysis indicates that while the flow switch failures could lead to a loss of some l- associated safety systems or indication, the loss would occur after they were L needed'cr there are alternate systems available to achieve the same safety ,

L functions. Sufficie'nt procedural direction and alternate information is

available for the operator to diagnose or respond safely tc misleading p . indications.

This analysis meets t'he criteria of 10CFR50.49, paragraph (1)(2).

l Therefore, continued operaticn is justified.

l 6

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. .a.,--www,,eco .-.-wwe-

UNIT 1

.BSEP JC0 NO.18 o _ ,

s..- TER NO.:~ 93' COMPONENT I.DA NO.: VA-FT-2577 ' ..  !

MFG / MOD. NO.: BAILEY BQ13221

-LOCATION: REACTOR BUILDING' ELEVATION 50' -

TECHNICAL DISCUSSICN

-

Component materials of the Bailey transmitters have been identified and i

compared to qualification documentation located for transmitters similar in design, construction, and ~ operation. The qualification data has .been evaluated per D0R guidelines and by Arrhenius techniques. Results of this evaluation _ indicate that,these transmitters consist of' essentially the same -

materials and components as Rosemount 1153 transmitters. The Bailey i transmitter includes Teflon and Viton o-rings. These o-rings are used as b  : static seals between the flange adapter and process flange (Teflon), the process flange and sensor module (Viton), and the electrical housing and cover I (Viton). These materials were evaluated at the normal and peak accident-

- conditi.ons and willL not experience' significant degradation of performance. '

i- .

The Rosemount transmitters were tested to parameters which envelop the BSEP i .- reactor building conditions (

Reference:

Rosemount Reports 3788, 109025, and ,

! D8300040). Based on the similarity of the Bailey transmitters to the Rosemount transmitters, the testing levels, and the environment at this lo 10gation rads (104*F normal, TID) use of the<Bailey 200*F transmitters for less than 10 minutes peak accident,1 X is justifed.

This analysis meets the criteria of 10CFR50.49, paragraph (i)(2).

Therefore, continued -operation is justified.

I l

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p. - _ _ _ _

L UNIT 1 L

BSEP JC0 NO. 19 TER NO: 94, 122 COMPONENT I.D. NO. : B21-F014 A-H B21-F048A & B CAC-PV-1218C E41-PV-1220D B21-F014J-N B21-F050A & B CAC-PV-1219B E41-PV-1221D B21-F014P B21-F056 CAC-PV-1219C E41-F023A-D B21-F014R, S B32-F039B & C CAC-PV-1225C* E51-F043A-D B21-F042A & B B32-F041C E11-F037A-D 1A-PV-1201A B21-F044A & B B32-F056E & F E41-PV-1218D CAC-PV-1209D*

B21-F046A & B B32-F058A & B E41-PV-1219D E11-F043A-D

• NO TER MFG / MOD. NO. : CHERRY E2360H

-LOCATION: REACTOR BUILDING 20' AND 50'; RHR ROOM TECHNICAL DISCUSSION:

Component materials of the Cherry switch have been identified and qualification documentation on a switch of similar materials and application has been located. o The qualification data has been evaluated per DOR guidelines and by applying Arrhenius techniques.. Results of this evaluation indicated that the nonmetallic components have greater than 66 years expected life at the maximum reactor building temperature of 104*F.

The Cherry switch nonmetallic materials are exposed to the plant postulated accident temperature peak of 288"F for 70 seconds. The accident temperature then decreases rapidly to 205'F at 100 seconds after accident initiation. This postulated peak temperature transient has been compared to accident test data obtained on a similar switch (212*F, 100% RH for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />). Though the testing does not envelop the postulated peak accident temperature, it is judged that no detrimental effects to switch operation should occur as a result of the peak temperature transient. This assessment is based on the severity of the test performed and the short exposure ,

time at the postulated accident peak temperature.

' Additionally,radiationtestingonswitchegofthesamematerialandapplication shprorts a qualification level of 3.6 x 10 radsgamma,alghoughthetestingdoes rads gamma, a radiation not envelop the postulated total integrated dose of 1 x 10 threshold analysis shows that theradjationthresholdanalysisforeachmaterial rads gamma except for the Delrin button. For

.used in switch is greater than 1 x 10

the Delrin button there is testing to support the us f this material in a 7

' mechanical application to a radiation level of 1 x 10 rads gamma (

Reference:

FCC Powers Report No. 734-79,002, Rev. 3).

In addition, an operational analysis has been performed to determine the effects of failure (misleading information, grounds, and spurious operation) of these items in both LOCA and HELB environments. The operational analysis indicated that there is sufficient information available for an operator to diagnose a misleading RIP valve position indication to response in a safe manner.

This analysis meets the criteria of 10CFR50.49, paragraph (i)(2).

Therefore, continued operation is justified. +

I.

Y .

A

A

~

UNIT.1 BSEP- ,

JC0 NO. 20

', TER NO.:. 95 COMPONENT 'I.D. NO.: E41-FT-N008 . -

. MFG / MOD. NO.:: GENERAL ELECTRIC 555111BDAA3PDH FLOW TRANSMITTER.

' LOCATION: RHR ROOM

' TECHNICAL DISCUSSION:

This flow transmitter provides' control of the HPCI Turbine Control Valve "

' position to maintain design rated HPCI flow. It also provides the control 4

room with an indication of HPCI pump flow.

!.' Partial qualification test data' has been obtained and evaluated for the flow transmitter. Testing has been successfully conducted to show that the device

? ;will: function under elevated temperature and humidity conditions (

Reference:

' G.E. Document No. NSE80036).

1 The accident simulation included.a peak temperature of 180*F for a time sufficiently long enough to perfonn a 6 point calibration estimated to take

about 5 minutes. ' Additionally, a separate test subjected the transmitter to a

68'F to 158'F at 100%RH test. The tests do not envelop the BSEP requirement of 199'F (3" RCIC line break) for 30 minutes and the subsequent ramp down to

.112*F in 8 minutes. However, the accident peak temperature excursion will not

-cause significant degradation of equipment operation during that period of exposure above the test maximum temperature (

Reference:

General Electric

, Report 327, File DV145C3007 and General Electric Document No. NSE80036).

~

In addition, an operational analysis was performed to address the effects of the postulated accident radiation environments on the operability requirements of the transmitter.

L In the event of a large break LOCA for which the HPCI system cannot maintain

'RPV level, the transmitter may be subject to high radiation. However in this casa, the HPCI system is not required since the RPV will be depressurized by l the i"eak and/or actuation of the ADS system. Adequate core cooling is then E provided by the low pressure ECCS systems. Therefore, operation of this

. device is not required for safe shutdown. In the event of a small break LOCA

! for which the HPCI system can maintain RPV level, the core never uncovers,

. hence cooling is maintained and the harsh radiation evironment is not present.

1-This analysis meets the criteria of 10CFR50.49,' paragraph (1)(2).

L Therefore, coitinued operation is justified.

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a' t ' ' UNIT 1 BSEP JC0 NO. 21 n TER NO. --

96, 97, 98 ,

~'

' COMPONENT- I.D NO.:- 'E11'-POIS-N021A,B' '

E21-FS-N006A,B

- E41-FSL-N006-MFG /M00. NO.:: BARTON289\' ,

LOCATION: RHR ROOM, CORE SPRA ROOM TECW ICAL DISCUSSION: ,

i IThese items control the minimun flow valves for tre[RHR, Core Spray and HPCI pumps. A minimum flow valve ,is gent. rally installed to prevent a pump from

. running 'at its shutoff head for an extended period of time.

If the instrument were to fail, showing low flow, the circJit would -act to open the valve. Unplanned opening of the minimum flow 7yalve during injection -

< would divert very little emergency flow from the RPV because of flow

'; restricting orifices in each of the minimum flow lines. 4' If the instrument were to fail, showing high flow, the circuit would act to shut the valve. During injection the valve would already be shut so there would be no effect. Undesirable, unplanned closing of the valve would only

. occur as the system was being secured by operator action. The operator can be expected to observe this and manually open the valve.

i I

~

i The plant' can be safely shutdown without 'these instruments.

l.

An additional analysis has been ' performed to insure that pressure switches

will-maintain electrical integrity during the postulated accident.

Component materials of the Barton differential prd~ssure switches have been

identified and qualification tiocumentation located. The qualification data l

has been evaluated per D0R guidelines and by applying Arrhenius techniques.

i Results of this evaluation indicate that the nonmetallic components have greater than 266 years of expected life at- the maximum reactor building temperature of 104*F.

l The pressure switch nonmetallic materials are exuosed to the plant postulated accident. temperature peak of 288*F for 70 seconds. The accident temperature i

then deceases to 205*F at 100 5econds and returns to ambient after k approximately 20 minutes. This postulated ;ed temperature transient has been compared to accident test data obtained (212*F fer 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) for this switch 1 -(

Reference:

AETL Test Report 50. 596-0399). Though the testing does not envelop the postulated peak acci, dent te'mperature, it is judged that no

'E, significant detrimental effects to switch operation should occur as a result

w of.the peak temperature transient.~ This assessment is based on the severity of the test performed and the'short time for heat transfer through the heavy metal casing.

~f- e L - - - -. - --- - - _ _ _--

UNIT'l-

~

BSEP JC0 NO. 21

- TER-96 Page~2 - . -

A Additionally, radiation testing on the subject switches supports a qualification level -of 3.6 X 106 rads gama. Though the testing does not envelop the postulated total integrated dose of 1 X 107 rads gama, a -

- radiation threshold analysis shows that the radiation threshold for each material used in the switch is greater than 1 X 107 rads gama. For the Viton

- o-ring there is testing to support the use qf this material in an 'o-ring

- application up to radiation level of 2 X 10' rads gama (

Reference:

-ASCO  :

' Report No. AQR 67368, Rev. O, paragraph 4.1.4).

This analysis meets the criteria of 10CFR50.49, paragraphs-(i)(1), (1)(2),'and (1)(5).

Therefore, continued operation is justified.

~

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UNIT 1 BSEP JC0 NO. 22

- l TER'NO.: -

- NONE 1

. . .r.

COMPONENT: I.D. NO.:~ E51-FS-N002 - .

MFG / MOD.-NO.: BARTON' 289 sLOCATION: REACTOR BUILDING RHR ROOM-

. TECHNICAL DISCUSSION:

' Component materials of- the Barton differential p'ressure switches have been identified and qualification documentation located. .The. qualification data

- has been evaluated. per~ DOR guidelines .and by applying Arrhenius techniques. '

'Results of this~ evaluation indicate that the nonmetallic components have greater.that_266 years of- expected life _dt .the maximum reactor building

temperature of 104*F.

.The pressure switch nonmetallic materials are exposed to the plant postulated accident' temperature peak of. 288*F for 70 seconds. The accident temperature then- decreases to 20S*F at.100 seconds and returns to ambient after

approximately-20 minutes. . This' postulated peak-temperature transient has been compared to accident test data-obtained (212*F for. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) for this switch.

Though the testing does not envelop the postulated peak accident temperature,

.it'isijudged -that no significant detrimental effects to switch operation Should occur as a result of the peak temperature transient. This assessment .

is-based on:the severity of the test performed and the short' time for heat

' transfer through the heavy metal casing.  ;

- Addition $11y, radiation testing on the subject switches supports a

-qualification level of 3.6 X .106 rads. gamma. Though the testing does not enveTop the postulated total integrated dose of.1 X 10 7 rads-. gamma, a

-radiationthresholdanalysisshowsthattheradiat}onthre'sholdforeach material used in the switch is greater 'than 1 X 10 rads gamma except for the. 4 Viton 0-Ring. For the Viton 0-Ring there is testing to support the usq of this material in an o-ring application up-to radiation level of 2 X 10' rads gamma-(

Reference:

ASCO Report-No. AQR 67368, Rev.0, paragraph 4.1.4).;

(

This analysis meets the-criteria of 10CFR50.49, paragraph (1)(2). #

y L - Therefore, continued operation is justified.

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• BSEP

' - JC0 NO. 23

. ,:TER NO. :'

100 .

, .. COMPONENT I.D. NO.:- CAC-TE-1258-1 TO 14 CAC-TE-1258-17 TO 24 MFG /M00.-NO.:  :

PYC0 100 OHM PLATINUM RTD

, [LOCAi10N: DRYWELL -

,1TECNtICA!..:0ISCUSSION:" ,  :

These temperature elements monitor drywell air space temperature for recording on a multipoint recorder located in the control room.

~

Pyco:has performed qualification testing on similar RTD env' eloping BSEP normal

- and accident service conditions (

Reference:

Pyco Qualification Test Report No.16436-82N, Rev. 5, dated 5/18/84).

Thel-similarity of the installed equipment has been confirmed by Pyco.

This analysis meets the criteria of 10CFR50.49, paragraph (1)(2).

Therefore, continued operation is justified.

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UNIT 1 1

' ' ' . BSEP

, JC0 NO. 24 107, 108, 110, 111', & 112 TER NO.:. -

l fi~ COMPONENT I.D. NO.: ' E41-TS-3314 E51-TS-3319 E41-TS-3315 E51-TS-3320 E41-TS-3316 E51-TS-3321 E41-TS-3317 E51-TS-3322 -

E41-TS-3318 E51-TS-3373 E41-TS-3354: E51-IS-3355 E41-TS-3488 E51-TS-3487 E41-TS-3489 MFG /M00. NO.: FENHAL TEMPERATURE SWITCH 17002-40 LOCATION: REACTOR BUILDING EL. -17' AND ABOVE TECHNICAL DISCUSSION:

These instruments are temperature sensors which monitor temperatures in areas (where the HPCI/RCIC steam line is located and initiate aa isolation signal in the event of a steam leak in the HPCI/RCIC steam line.

During a LOCA, these switches must not fail in such a way that produces a

~

spurious steam line leak indication unti1= the plant has' been brought to a-low

' pressure- condition. . If such a spurious signal did isolate the HPCI, the redundant- ADS system would remain available. No credit' is taken for RCIC

during a-LOCA.

Fenwal temperature switch, Model No. 17002-40 (modified per Patel Engineers l specification), has been qualified by testing to meet or exceed BSEP normal and accident conditions. The tested model was identical to the installed one, except the -lead wire insulation in the installed switch is teflon.

~

' Teflon hag excellent temperature tolerance and the radiation threshold value is 5 X 10' rads for electrical applications (Reference- REIC 21). The maximum accident exposure for these switches is 1 X 10 I ads r gamma over 30 days. In the Fenwal temperature switches the Teflon lead wire is sandwiched

! .between-ba) layers of nonradiation sensitive material which will maintain sufficient insulation resistance for the maximum inservice. voltage of 120 volts.

This analysis meets the criteria of 10CFR50.49, paragraph (i)(1) and (i)(2).

l.

'Therefore, continued operation is justified.

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WIT 1-

' BSEP

^ ' ' '

JC0 NO. 25 TER NO.:- 10 9

^

COMPONENT I.D. NO.: B21-TS-N010A B21-TS-N010C-821-TS-N0108 B21-TS-N0100 MFG /M00.'NO.: . FENWAL TEMPERATURE SWITCH 17002-40 LOCATION: REACTOR BUILDING (TUNNEL) EL. 20' .

TEcletICAL DISCUSSIm: -

Jenwal temperature switch, Model No. 17002-40 (modified per Patel ' Engineer's Specification) has been fully qualified by test which exceeds the BSEP normal cand accident service conditions (

Reference:

Patel Engineer's Qualification Report No. PEI-TR-831200-1). The tested model was identical to the one installed at BSEP except the' lead wire insulation was different. The

-instalTed switches have teflon insulated lead wires and the tested unit had Rockbes' t os crosslinked polyethylene insulated lead wires. p

' Teflon' has a high temperature rating and the radiation threshold value is 5 X 10 rads fore electrical L applications. (

Reference:

REIC 21). -

These temperature switches initiate main steam isolation valve closure on a

high temperature .in the steam line tunnel and will complete their safety j function immediately after the accident initiation. Therefore, the temperature _ switch lead wires will ngt be significantly degraded by an l estimated radiation dose of _1.5 'X 10* rads before completing their safety l-- function.

l l

This analysis meets'the criteria.of 10CFR50.49, paragraph (1)(2) and (i)(4).

Therefore, continued operation is justified.

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N1 - " UNIT 1 BSEP s - JCO NO. 26 ,..

~

s t 1,TER No.:- 1115-

COMPONENT l'.D.'NO.: '1(A-D)-BFIV-RB-L.

(MFG /MODRNO.:- ~NAMCO D2400XRl 1 LOCATION:. REACTOR BUILDING'80' TECHNICAL DISCUSSION: . - ,

p ,

JCAmpon~nt e i materials of the NAMCO 2400XR position switch have been identified.. The 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 asia binder in the asbestos gasket),

have greater than forty (40) years demonstrated qualified life at the maximum reactor building temperature - of 104*F. 4.The gasket, which is comprised of 20%

' Buna-N and 80% asbestos p 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. .

LThe' analysis performed on the-D2400XR switch is based on testing conducted on

~

NAMCO series;SL3 switches (generically similar in materials, construction, and

' operation). These-switches were exposed to a 310'F an'd 65 psig steam = ..

'f,

-; environment (

Reference:

Masonellon Test- Report .1003, dated 4-19-73) which exceedsLthe BSEP requirement.

Aradiationanalysisindicatesthgt.thelowest'damagethresholdforthe.

nonnetallicmaterials.is~8.6x10.radsgamma. This damage threshold'value (envelops the,BSEP requirements of:1-x 10 rads gamma..

This analysis. meets the criteria of 10CFR50.49, paragraph (i)(2).

1. Therefore, continued operation is justified.

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UNIT 1 BSEP JC0 NO. 28 TER NO.:' 124, 125, 126, 127, 128, 129 COMPONENT 1I.D. NO.: B21-F003-Lf CAC-PV-1227C*

. . . 821-F004-Li CAC-PV-1227E-L1 B32-F019-L CAC-PV-1227E-L2*

o B32-F020-L' CAC-PV-1231B*

CAC-PV-1200B* CAC-PV-1260-L CAC-PV-1205E-Li*, L2* CAC-PV-1261-L CAC-PV-1209A-Ll*, L2* CAC-PV-1262-L CAC-PV-1209B CAC-PV-3439-L*#

CAC-PV-1211E* CAC-PV-3440-L*#

i CAC-PV-1211F-Li*, L2** CAC-V47-L CAC-PV-1215E*f CAC-V48-L CAC-PV-12258*f CAC-V55-L CAC-PV-1227A-Li*, L2* CAC-V56-L CAC-PV-1227B-Ll*, L2*

1. NO TER

, MFG / MOD.'NO.: HONEYWELL MODEL OP-AR AND *0PD-AR LIMIT SWITCHES

$= DW 17' (B32-F019, 821-F003, 821-F004 ONLY)

LOCATION:

RX 20' & 50' (ALL'OTHERS) i

. TECHNICAL DIS'CUSSION:

' Component materials of the Honeywell limit switches have been identified and partial qualification documentation located. The qualification data has been evaluated per D0R guidelines and by applying Arrhenius techniques. Results of this evaluation indicate the limit switches insi.de the reactor building will perform their post-accident function prior to failure (

Reference:

(1)

" Nuclear Radiation and Switch Applications," Micro Switch, October 7,1974, (2) " Humidity Test of the 'W' Lever Type '2 Switches with General Purpose Phenolic, Mica-Filled Case and Cover, Melamine or Valox Plungers.," Micro f Switch . July 15,1975, (3) " Evaluation of Asbestos-Free Plastics for 250*F L

Basic Switch," Micro Switch, February 21, 1979, (4) " Environmental Test," 9993

Barksdale, August 13, 1975). -

l l The analysis for the switches located in the reactor building meet the criteria of 10CFR50.49, paragraph (i)(2).

l Limit; switch plant ID No. B32-F0g9 located inside the drywell has been type tested for radiation to 1.3 X 10 rads gamma, which envelops the BSEP .

requirement (

Reference:

" Nuclear Radiation and Switch Application", Micro Switch, October 7, 1974).

! However, the test parameters (

Reference:

(2), (3), and (4) above) do not i

envelop the BSEP postulated drywell accident conditions.

This switch provides only valve position indication to the control room for the inboard reactor water sample valve (B32-F019). The reactor water sample valve is normally open and may be closed by the control room operator or in response to an automatic isolation signal.

L

UNIT 1 BSEP JC0 NO. 28 TER 124-129 ~

Page 2; , ~

h ,

Failure of limit switch B21-F019 'has been anlayzed and may result in (1) loss of valve position-indication, (2) loss of control power to the valve solenoid,

. or (3). both (1) and (2). Loss of control power results in automatic closure of the valve. Since control power is fused, electrical fault of the limit switch would not adversely effect other safety related equipment.,

- : However,- the plant can be safely shutdown in the absence of limit switch B21-

.F019 since the valve fails shut.and is required to shut for an automatic isolation signal.

'_This analysis meets the criteria of 10CFR50.49, paragraph (1)(2)(4)(5).

Therefore, continued operation is justified.

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UNIT 1 BSEP e

JCO NO. 29 ITER NO.: 130, 131, 133, 134, AND 135

^

COMPONENT I.D. NO.:- Bil-RS B49-RS1 DL2-RS* DM7-RS1 811-RS1 . 850-RS DL2-RS1 DM8-RS~

B21-CS-3327 850-RS1 DL7-RS- DM8-RS1 B21-CS-3329 .DE3-RS-CS* DL7-RS1 .DN1-RS

- 821-CS-3345* DE3-RS-SS. DL8-RS DN1-RSI-821-CS-3412 DH2-R'S* DL8-RS1 DN6-RS B41-RS' DH2-RS1 -DL9-RS DN6-RS1  ;

B41-RS1 DH3-RS* DL9-RS1 DN9-RS B43-RS DH3-RS1 DM1-RS. DN9-RS1 B43-RS1 DK8-RS DM1-RSI DP2-RS

. B45-RS DK8-RS1 DM2-RS. DP2-RS1 B45-RS1 DK9-RS DM2-RS1 DP3-CS B46-RS DK9-RS1 DM4-RS DPS-RS-A-SS i

.'~

846-RS1 DLO-RS DM4-R$1 DPS-RS-CS 847-RS DLO-RS1 DMS-RS DS4-RS 847-RS1 DL1-RS DMS-RS1 DS4-RS1 849-RS DL1-RS1 DM7-RS FN6-CS

• NO TER l

PLUS 40 VARIOUS MISC. CONTROL SWITCHES MFG / MOD. NO.: .HONEYWELL MICROSWITCH, TYPES: PTSHA202FB52, PTSHA201, PTKBC2221CC, PTKBC2221, PTSHE201, PTK8C221CC, AND

. PTKBF3221.

I LOCATION: REACTOR BUILDING EL. 20' ,

(. TECHNICAL DISCUSSION:.

The above control and selector switches are in the remote shutdown system and their function is considered as essential passive.

l The PT series switch have been tested at 185'F for 767 hours0.00888 days <br />0.213 hours <br />0.00127 weeks <br />2.918435e-4 months <br /> (more than 30 L days) as per Honeywell Micro Switch Qualification Report No. 24407. For

! radiation the switches have been analyzed as per Honeywell Engineering Report

!- No. LTR 15027-1 to rad TID. BSEP maximum anticipated radiation is 1 X 10p acceptable rads. TID. to 5 X 100 Honeywell test conditions envelop the BSEP accident duration of 30 days.

However, the peak accident temperature of 200*F for 70 seconds was not

> enveloped. Since the switches are within enclosures, the switches will not see the peak temperature during the short exposure time because of thermal shielding. Moreover, the BSEP accident temperature will remain at 133*F for the remainder of the 30 day post-accident period. Since the switch was exposed' to 135'F for more than 30 days,'added confidence in the switch's ability to survive the accident and post-accident period is assured.

l This analysis meets the criteria of 10CFR50.49, paragraph (i)(2).

Therefore, continued operation is justified.

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UNIT 1

@ygy::g 9<,

' -BSEP: ,

y- ,

' ~

'JCO No. 30 x, -

132,142, J144,145 146, & 147 e b.ITERNO.':

? COMPONENT I.D. NO. : " MCC-1XA, MCC-1XAf2,r MCC-1XB, MCC-IXB-2, MCC-1XC, NCC-1XD, j$"1(

" + > ;MCC-1XDA, (MCC-1XDB,' MCC-1XE, MCC-!XF, ~ MCC-1XG, MCC-1XH .

a RELAYS:lB11-B09-RS, Gl6-3-6A, GN2-3-5A*

e'* NO TER' n ,

p

~ 4 MFG / MOD. ' NO. :1 :CENER!i ELECTRIC IC 7700 10 TOR CONTROL CENTER g

E iLOCATION: LREACTOR: BUILDING 3

', TECHNICAL; DISCUSSION:

7 ETest data applicableL to the environmental qualification of- the General

iElectric Series;IC 7700 motor control center has-beenitdentified and

- fqualification documentation located. The qualification data has been

-evaluated per DOR' guidelines'and by applying Arrhenius techniques.

'A preliminaryfassessment.of the test data',-performed by. General Electric Co.,.

' ~ ~

• indicated that:the testidata cantbe'used to demonstrate qualification of the:

? mator control centers to.be'BSEP. normal and postulated accident conditions V

(Reference ' Environmental-Qualification Assessment' Report - Phase I,.GE~

' document. number 710-03-025B).-

m .S sequent totthe preliminary assessment, GE-issued a-second document, GE

= report number NEDC-30322-P. -

DThis document contains detailed Engineering

!~ ' Change Notice (ECN) reviews,' Product' Analysis Reports,' Similarity Analysis

Reportsf on specific components contained. inlthe motor control centers-(such c'ar; THED circuit' breakers,'CR109. magnetic starters, control power.

transformers,7 relays s'nd other components).' This. report also indicates that t

the best data obtained demonstrated qualification =of the IC 7700 motor-control

~

- center to the'BSEP normal and postulated accident conditions.

g

$: :The final report on the qualification. status of the IC 7700-motor control

~

center is currently being prepared by General Electric. t

-Based upon the test data obtained:and the assessments performed, this analysis meets;the criteria of'10CFR50.49, paragraph (i)(2).

iTherefore, continued operation is justified.

4 UNIT 1 4 BSEP

, - y JC0 NO. 31 TER NO.:' 138 COMPONENT I.D. NO.:  ; E11-C001A, B, C, D MFG / MOD.- NO. :-; GENERAL' ELECTRIC SK821161C11 LOCATION: ,

REACTOR BUILDING - 50'

. TECittICAL DISCUSSIGl: ,

x

~ The above. motor is a horizontal induction motor with a Class B custom Polyseal sinsulation. . It is a totally enclosed air / water cooled unit designed to operate continuously at 194*F ambient temperature. Its function is to drive the RHR Service Water Booster Pump.

Tcst data has been obtained for vertical induction motors with the same insulation class- (G. E. Document NEDC-30294). The test data obtained envelops the posiulted accident conditions at BSEP (temperature, pressure humidity. -

radiation).

~

Arrhenius data obtained for the motor insulation has been evaluated. The evaluation shows a 40 year. life for the Class B insulation at the BSEP service condi tions.

The motor bearings and lubricating system are inspected and maintained in accordance with the BSEP periodic maintenance and surveillance program.

~ .

' . This analysis. meets the criteria of 10CFR50.49, paragraph -(i)(2).

Therefore, continued operation is justified.

l l

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, , UNIT 1 BESU JCO NO. 32 5-- . . . TER NO.: 141,155 COMPONENT I.D. NO.: E41-C002 (INCLUDING TERMINAL. BLOCKS J16-TB-8, C, D*)

4

• NO TER MFG / MOD. NO.: TERRY STEAM TURBINE MODEL CCS HPCI PUMP SYSTEM

^

LOCATION: REACTOR BUILDING EL. -17' ,

TECHNICAL DISCUSSION: -

' An operational analysis has been performed on the Terry Steam Turbine Model

.CCS HPCI Pump System. . The following postulated BSEP accidents were considered

- .in this evaluation:

1. HPCI Steamline Break

2. Large Break LOCA

3. Small Break in RCIC Steamline

4. Small Break LOCA In all cases alternate qualified ECCS systems in conjuction with the ADS system (auto or manual made) are available to maintain core cooling for a safe shutdown. Operator response is covered in the Emergency Operating Procedures.

4 This evaluation meets the criterial of 10CFR50.49, paragraph (i) (1).

- Th'erefore, continued operation is justified.

l l

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. UNIT 1 BESU JC0 NO. 33

, 'TER.NO.: - 143 - -

COMP' O NENT I.D. NO.: DBO-74-18 MFG / MOD. NO.: AGASTAT 7022AC TIME DELAY RELAY LOCATION: REACTOR BUILDING RHR ROOM ,

TECMICAL DISCUSSION:

BSEP has one Agastat time delay (model 7022AC) installed in the control

- - circuit of RHR pump room cooler fan A-FCU-RB. An automatic start signal to x RHR pump room cooler fan A-FCU-RB de-energizes the coil of the time delay relay which initiates the time delay function. If, after the timer delay

. setting has elapsed, the fan motor contactor has not closed, an annunciator Lalarm is sounded in the control room indicating that fan A-FCU-RB has failed tu start. It is important to note that this relay does not perform any control function to start or stop the fan; it only gives indication.

The result of the fail'ure of this relay would possibily be: (1) Loss of control power to the fan A-FCU-RB and (2) Loss of alarm to the control room that fan A-FCU-RB has failed to start. If contro.1 power is not lost, the fan

~

would start as designed. However, should the first fan fail to start the RHR pump rooms are provided with another 100% capacity fan B-FCU-RB. This fan will automatically start as soon as RHR pump room temperature reaches 145"F or above. 'There is no time delay relay involved in the control circuit of fan B-FCU-RB.

This analysis meets the criteria of 10CFR50.49, paragraph (i)(1)

Therefore, continued operation is justified. -

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l UNIT 1 BSEP JC0 NO. 34

~TER'NO.: 148 COMPONENT I.D.,NO.: ' '012-RE-N010A$B -

-MFG / MOD. NO.: G. E. MODEL 194 X 927G RADIATION DETECTORS LOCATION: -REACTOR BUILDING EXHAUST AIR PLENUM EL. 80' 4

TECimICAL DISCUSSIGt:

' Partial qualification documentation has been obtained for the General Electric.

radiation detectors. The test data was evaluated per the DOR guidelines and using Ar.rhenius techniques. The results of this evaluation ~ indicate that the radiation detectors were tested at 212*F for 6. hours and performed

~

satisfactorily before, during and after the test exposure. The test

. parameters envelop the BSEP requirement of 200*F accident peak temperature

(

Reference:

General Electric Report No. 248A9178).

The reactor building HVAC exhaust air plenum radiation levels are continuously l 4

~ monitored by two redundant radiation detector sensors. The detectors provide

, output ~ signals.which initiate the automatic start of the Standby Gas Treatment ,

. System and secondary Containment Isolation when the radiation levels exce'ed 11 MR/HR. .

During' normal operation, the detectors will be only 3 X 10gotal rads integrated radiation which is well below theexposure damage for the threshold r

' level of the detector nonmetallics. The detectors activate at 11 MR/HR and complete its function before damage due to higher levels of radiation is experienced as a result of the accident. ,

~

Since the detectors perform their mitigation function inanediately upon accident detection, failure would not prevent ECCS actuation or prevent the citigation of a HELB.

Failure to automatically start the SBGT system and isolate the secondary containment during a HELB will not result in an off-site radiation dose in 4

excess of the 10CFR100 limitations. The resultant radiation release is less than a main steam line break in the turbine building.

i SBGT and reactor building isolation may be manually initiated from the control

- . room and/or automatically initiated in response to other sensed parameters which occur during a.LOCA.

. Additionally, the detectors are periodically tested once every 18 months by physically removing them from their mounting and performing a complete -

3 functional test.

This analysis meets the criteria of 10CFR50.49, paragraph (1),(1)(2)(3)(4).

l ,

!' Therefore, continued operation is justified.

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UNIT 1 BSEP JC0 NO. 36 LTER NO.: 156 -

COMPONENT I.D.~ NO.:' NG7-SGT-FILT-1A-RB NG8-SGT-FILT-1B-RB

~ MFG / MOD..NO.: . 2FARR MODEL NUMBER D51423 LOCATION: REACTOR BUILDING 50' TECHNICAL DISCUSSION: -

The SBGT'is not assumed to remain operable in the most severe postulated HELB

' environment but as discussed below, its operation is not necessary for this t 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 10CFR100.

Since:the inventory loss prior to isolation for a HELB .is less than the main steam line . break, the offsite 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 ters recovery from the event is underway. Thus, there is no need for the SBGT

-system to maintain a negative pressure in the reactor butiding during i recovery.

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 133*F in appro(imately 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 fear life plus the accident.

qualification documentation was obtained for the SBGT system and analyzed per

00R Guidelines. The testing was performed on identical and/or similar

components (

Reference:

Farr Test Report No. L-71167). For those safety- -

i related components not tested specifically by The Farr Company, supplemental

. qualification data was obtained and analyzed. Thesp components include:

l

1. Blower Motor This is an enclosed General Electric blower motor with a Class F

! insulation system. This insulation system has been analyzed and found to be superior to the G.E. Class 8 insulation system which has been successfully tested ' tg a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, 212*F peak temperature,1005 relative humidity and 5.5 x 10' rads gamma. This testing envelops the BSEP

-postulated accident transient and through analysis, the post-accident period.

1

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' UNIT 1

. . - BSEP JC0 NO. 36 TER No. 156 Page 2 , . s

2. _ 'ITE Molded Case Circuit Breaker These breakers have been tested separately by ITE at a _ temperature and radiation dose more severe than the BSEP postulated accident. conditions .

(

Reference:

ITE-Gould Report No. CC 323.74-57, Rev. 2 dated October 6 1980).

3. Allen-Bradley Push Button Control and Selector. Switches These devices are manufactured basically from phenolic and metallic materials. Similar-switches have been tested by Honeywell to parameters which envelop the BSEP postulated accident conditions (

Reference:

Honeywell Test Report No. LTR-24407).

4 A11en-Bradley Series 700 Contactor These contactors have been successfully tested to 2 x 10 8 rads gamma and 248'F which envelops the BSEP requirements (

Reference:

ANCO letter for IEEE 323-1974 Qualified Components). ,

This analysis meets the criteria of- 10CFR50.49, paragraph (i)(1) and (i)(5).

Therefore, continued operation is justified.

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UNIT 1 BSEP JC0 NO. 40 7 TER NO.: 179, 181

COMPONENT I.D.~NO.

_ TERMINAL BLOCKS i MFG /M00. NO.:- GENERAL ELECTRIC EB-25, CR-151, EBS*

, LOCATION: REACTOR BUILDING 1

• NO TER TECHNICAL DISCUSSION:

Component materials of the General Electric terminal blocks have been

-identified and qualification documentation on similar terminal blocks has been .

located. The qualification data has been evaluated per' DOR. guidelines and by

applying Arrhenius techniques. Results of this evaluation indicates that the nonmetallic components have greater .than 5 X 108 years of expected life at the maximum reactor building temperature of 104*F.

The test data shows that similar terminal blocks were exposed to test conditions, including radiation, . significantly more severe than the postulated accident conditions at BSEP.

Leakage current was monitored during that portion of the test program with conditions at BSEP. The average leakage current per-terminal block was less than 1 ma at 120VAC. . The results of this test coupled with the facts that:

1. All terminal blocks are in an enclosure and therefore not subjected to direct impingement of steam or water.

2. There is a redundancy of all safety related systems as well as a

. physical separation.

3. All systems-are periodically tested which would detect any random failure.

further' substantiate the use of these terminal blocks in the Reactor Building

(

Reference:

Amerace Report F-C5143).

This analysis meets the criteria of 10CFR50.49, paragraph (t)(2).

Therefore, continued operation is justified.

e

. UNIT 1 BSEP JC0 NO. 41

'TER NO.: - 182 '

COMPONENT I.D. NO.: TERMINAL BLOCKS

' MFG / MOD. NO.: CURTIS TYPE "L" LOCATION: DRYWELL/ REACTOR BUILDING *

• NO TER TECHNICAL DISCUSSION:

Test documentation has been located and evaluated for these terminal blocks.

A Westinghouse Report PEN-TR-77-83 dated 9/13/77, " Test Report on the Effect of a LOCA-'on the Electrical Performance of Four Terminal Blocks", and a Westingnouse Research Memo' No. 76-1CC-QUAEQ-M24 entitled, " Radiation Hardness of Terminal Blocks", did result in the success of at least four types of

, similar terminal blocks; Westinghouse, Curtis, Marathon and Cinch Jones.

These blocks are-similar in material. construction, contact configuration and electrical characteristics to blocks installed at BSEP.

Additionally, Curtis type "L." terminal blocks were tested by Limitorque as part of their qualification of a motorized valve actuator (Limitoque Report No. B-0119). The environmental conditions seen by these test specimens meet the requirements at BSEP. All terminal blocks are in an enclosure and not subjected to direct steam impirgement of steam or water. This cor. figuration i is similar to the test configuration.

This, anal'ysis meets the criteria of;10CFR50.49, paragraph (1-)(2).

Therefore, continued operation is justified.

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. UNIT 1 BSEP JC0 NO. 42 TER NO.: NONE COMPONENT I.D. NO.:- ' C11-F010-L '

E51-C002-LS4 '

MFG / MOD. NO.: NAMCO D1200G LIMIT SWITCH o-LOCATION: REACTOR BUILDING 50', RHR ROOM

TECimICAL DISCUSSION:

?. .

Component D eaterials of the Namco D1200G limit switch have been' identified and

-qualification documentation on similar equipment located. The qualification 4 data has been evaluated per DDR guidelines and by applying Arrhenius techniques. Results of this evalua componentshavegreaterthan9X10gionindicatesthatthenonmetallic years at the maximum reactor building tenperature of 104*F except for Buna-N. The Buna-N components have an expected life of greater than 11.8 years. ,

The test data shows that the switch was exposed to test conditions more severe than the BSEP postulated accident conditions for temperature, pressure, and relative humidity (

Reference:

Masoneilan International Report No.1003).

' Additionally, a radiation analysis performed on the component materials sgows that the radiation threshold Buna-N which is the weakline material 1 X 10 rads. The switches scomplet their safety function is less tt'n one hour and thg maximum postulated total integrated radiation dose durin, this ti is 1 X

10 rads which is much lower that the Buna-N threshold value of 1 X 10 rads.

1

. This analysis meets the criteria of 10CFR50.49, paragraph (i)(2) and (i)(4).

. ;Therefore, continued operation is justified. ' \

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a.

UNIT 1 BSEP

- '.. , JC0 NO. 43

. TER NO.: NONE E COMPON ENT I .D. NO. : - - NP6-M07-M1,' M2 ' NP7-MOT-M1, M2 4

• IB-RX 1A-RX Li MFG / MOD. NO.: DOERR MOTORS AND ITE CONTROL PANELS

LOCATION

L REACTOR BUILDING EL. 20' TECM ICAL-DISCUSSION:

The above electrical compcnen$s are associated with the compressors to the standby air supply for the Non-Interruptible Air System. Non-interruptible instrument air is supplied to the following control systems:

1. Mais steam isolation valves

( 2. Scram valves

~ 3. Scram volume vent and drain valves

4. ~ Safety relief val'ves

. 5. Control rod drive flow regulators

6. Reactor instrument penetration system valves Each of the above valves are supplied with air accumulators of sufficient size -

to provide valve actuation air'in the event of total instrument air supply

, faildre. The Control Rod Drive System will perform its required safety .

function before the compressors will fail as a result of a HELB or LOCA.

A loss of the energency air comptassors could cause a loss of reactor level, pressure and monitoring instrumentation during a LOCA. It could cause a loss of HPCI/RCIC and' reactor instrumentation during a HELB until Unit l's air system could be cross-connected (<1 hour). Alternate systems, instrumentation, or procedural guidance is provided for directing the operator's response during these events. Other safety related components '

would either complete their safety function before air supply failure, have suitable accumulators, or fail in the safe direction. The air compressors do not directly control any indications.

,The above analysis meets the criteria of 10CFR50.49, paragraph (i)(3).

Therefore, continued operation is justified.

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1 UNIT 1 BSEP JC0 NO. 44 TER NO.: -NONE COMPONENT I.D. ' NO.: - E51-C002-H ' ,

, MFG / MOD. NO.:1 SQUARE D 9038-AG1-54 FLOAT SWITCH

- LOCATION: RHR ROOM

- TECMICAL DISCUSSION:

4

- This item is part of -the RCIC turbine assembly. It must maintain its

- electical integrity for 30 minutes during the BSEP postulated accident.

Testin'g has been successfully performed on a HPCI turbine that ocntained this component (Ref: Wyle Lab / Terry Turbine Report No. 20458, R14-21-80). The testing w to_1 X 10gsrads.

performed Duringatthe150*F HELBforaccident an undetermined condition,time and radiation testing the temperature

' gradually rises from 104*F.and peaks at 198*F in 30 seconds at which time

- steam leak isolation is completed. The accident radiation dose in the first 30 minutes of the accident will be less than 1 X~103 rads.

Since"the switch terminations are enclosed in a NEMA metal enclosure it is safe to assume that the switch will maintain its electrical integrity for the required duration.

, - This analysis meets the criteria of 10CFR50.49, paragraph (1)(2).

, Therefore, continued operation is justified.

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UNIT 1

. BSEP JC0 NO. 45

~

TER NO.: NONE ,

. , - s COMPONENT I.D. NO.:- B32-CS-F019 B32-CS-F020 NFG/M00.'NO.: SENTRY MODEL F3N1R1 SWITCH LOCATION: - REACTOR BUILDING EL. 20' TEC}ftICAL DISCUSSION:

. 'The Sentry F3N1R1 switch utilizes a Series 2 Honeywell Microswitch as the internal switching mechanism.

Honeywell Series 2 switches have been tested at 149*F for more than 30 days

(

Reference:

Honeywell Microswitch Test Response No. LTR-24407). This test

-onvelops the BSEP accident duration but does not envelop the 70 second BSEP peak temperature transient of 200*F. A material analysis indicates that the switch will not be significantly degraded by the short exposure to the postulated accident peak.

Additionally, the switch has been tested to 1 X 10 7 Rads (

Reference:

HogeywellReportNo.LTR-15027-1) which envelops the BSEP requirement of 1 x 10 Rads gamma.

This analysis meets the criteria of 10CFR50.49, paragraph (1) (2).

Therefore, continued operation is justified.

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" UNIT 1 BSEP JCO No. 46

'TER.NO.: 'NONE 4 B24-FT-4163 AND 12J-SPLICE L COMPONENT ' I.D. ' NO. : B21-FT-4167

_B21-FT-4158 B21-FT-4164 I2E-SPLICE I2K-SPLICE

. P21-FT-4159 B21-FT-4165 12F-SPLICE 12L-SPLICE-B21-FT-4160 B21-FT-4166 12G-SPLICE 12M-SPLICE B21-FT-4161' B21-FT-4167 12H-SPLICE I2N-SPLICE

. B21-FT-4162: 121-SPLICE I2P-SPLICE

MFG /)OD. NO.: 'NDT INTERNATIONAL 78IN/S ACCELEROPETER LOCATION: DRYWELL EL. 38' LTECHNICAL; DISCUSSION:

NDT International accelerometers, Nbd :1 No. 781N/S, are qualified on the basis of-similarity with the NDT International accelerometer, Mbdel No. 838-1,

,7

,(Reference'Wyle. Qualification Report No. 45638-1). Pbdel 838-1 was fully qualified to meet or exceed all BSEP service conditions'inside the drywell.

-Similarity Mbdel No. -781N/S and .838-l 'are similar. The only difference is in the accelerometer cable interface connections.

'Should the interface connections fail, there is a possibility of faulty

' indication of safety relief valve position in the control room. However, another independent indication system is provided for safety relief valve position indication. This' redundant channel signal is temperature-dependent. Therefore, safety relief valve position indication would not be

lost in the event of accelerometer failure.

This analysis' meets the criteria of 10CFR50.49, paragraph (1)(1) and (i)(2).

Therefore, continued operation is justified.

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,' BSEP JC0 NO. 48 TER'NO.: , NONE

. COMPONENT I.D. NO.: QC4-P10-A QC4-P3-M8 QC4-P9-MA QC7-P7-B QC4-P10-8 QC4-P4-A QC4-P9-MB QC7-P7-MA QC4-P10-MA QC4-P4-B QC7-P10-A QC7-P7-M8 QC4-P10-M8 QC4-P4-MA QC7-P10-8 QC7-P8-A:

QC4-P2-A. QC4-P4-MB QC7-P10-MA QC7-P8-8. ,

QC4-P2-8 QC4-PS-A QC7-P10-MB -QC7-P8-MA QC4-P2-MA QC4-P5-8 QC7-P6-A QC7-P8-M8 QC4-P2-MB QC4-PS-MA QC7-P6-8 QC7-P9-A QC4-P3-A QC7-P9-8 o QC4-P5-M8 QC7-P6-MA-QC4-P3-8 QC4-P9-A QC7-P6-M8 QC7-P9-MA QC4-P3-MA QC4-P9-B QC7-P7-A QC7-P9-MB MFG /M00 NO.: GULTON CONNECTOR MODEL RECEPTACLE 9012 4 PLUG 9013 LOCATION: DRYWELL TECHNICAL DISCUSSION:

These items-are coaxial cable connectors installed as part of the safety  :

relief valve monitoring system.

t l I The above connectors are located inside the drywell and the prototype unit of >

the same has been successfully tested in accordance v'th the requirement of Wyle Qualification Plan No.. 45098-1. The testing enveloped the required normal and accident service conditions. However the tested configuration differed from the BSEP operating configuration in that the tasted configuration encased the connector in heat shrinkable tubing. This was done after thermal aging to add mechanical rigidity to the assembly. .

\ ,

Since the BSEP configuration is not subject to movement after installation, l this ~ added rigidity is desirable but not necessary. The connector is sealed r

- against the 'environent by a compressible grommet at the cable entry and t compressible metal seal where the connector havles join se the sealing  !

capability of ,the heat shrink is not required.

This analysis meets the criterial of 10CFR50.49(1)(2).

Continued operation is justified.

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. UNIT 1 BSEP JC0 NO. 49 TER.N0.:. NONE COMPONENT I.D. NO.: - '

HPCI SYSTEM TEST POINTS 1-IG7-TPS 1-IG7-TP7 1-IG7-TP6 0 1-IG7-TP8 MFG / MOD N0.: N/A LOCATION: REACTOR BUILDING.50' TECHNICAL DISCUSSION:

These are phenolic 120V control power test jacks that were installed as extensions to the instrument rack terminal block test points.

These items.are located in a gasketed enclosure which would minimize the entrance of water or steam. They are electrical " banana" type test points which, because of their simple construction, have a low likelikhood of environmentally related failure due to a steam leak condition. The failure in question would result in small leakage current between the points or between a .

point' and ground similar to but probably less than that experienced for a 120V control power termination on a terminal block. The effect of such a leakage current would be negligible. ,

Because of the internal spacing and overall construction of thf s type of component, the likelihood of a failure during the required operating time for the HPCI system (24 hrs.) is vanishingly. small.

These will be deleted from the plant.

This anlaysis meets the crieria of 10CFR50.49(1)(4).

Continued operation is justified.

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UNIT 1 BSEP JC0 NO. 50 TER NO.: NONE

' COMPONENT I.D. NO. : STANDBY ' GAS TREAT >ENT SYSTE_M, COMPONENTS, XS4-DS5 -XTO-DS10 XS4-DS6 XTO-DS9 XS4-MAN.0VRD.SW. XTO-MAN.0VRD.SW XT1-DS11 X02-DS7 XT1-DS12 X02-DS8 XT1-MAN.0VRD.SW. X02-MAN.0VRD.SW.

)

ALLEN-BRADLEY INDICATOR LIGHTS (XS4-DS5, XS4-DS6, XTO-MFG / MOD NO.:

DS10, XTO-DS9, XT1-DS11, XT1-DS12, X02-DS7, X02-DS8)

HONEYWELL 914CEl-3 (XS4, XTO, XT1, X02 MANUAL OVERRIDE SWITCHES)

LOCATION:. XTO, XT1 REACTOR BUILDING 117' X34, X02 REACTOR BUILDING 80' TECHNICAL DISCUSSION:

These indicator lights and switches act in the control circuits of the supply and exhaust isolation dampers.

During a LOCA, these dampers will close on a safety signal very early in the accident, prior to the environment becoming harsh. Any loss of the control circuit thereafter will not affect the damper positions, it will remain in its safety position.

As discussed below, these items are not necessary to mitigate the effects of an HELB.

The radioactive release from a worst case 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 it results in much less site boundary dose than that permitted by 10CFR100.

Since the inventory loss prior to isolation for an HELB is less than the main steam line break, the offsite HELB dose is also correspondingly low even if these dampers do not operate. 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 reactor building to be isolated during recovery.

This analysis meets the requirements of 10CFR50.49, paragraph (1)(4).

Continued operation is justified.