Text

Forwards Revised Justification for Continued Operation Re Environ Qualification of safety-related Electrical Equipment,Per 840818 & 20 Commitments.No Concern Exists to Prevent Continued Safe Operation of Plant
	ML20098E305
Person / Time
Site:	Brunswick
Issue date:	09/26/1984
From:	Cutter A; CAROLINA POWER & LIGHT CO.
To:	Vassallo D; Office of Nuclear Reactor Regulation
References
	NLS-84-425, NUDOCS 8409280325
	Download: ML20098E305 (61)
	v • d • e

.

A CfEL Cdfolha Power & Light Company SERIAL: NLS-84-425 SEP 2 61984 Director of Nuclear Reactor Regulation Attention:

tk. D. B. Vessallo, Chief Operating Reactors Branch No. 2 Division of Licensing United States Nuclear Regulatory Commission Washington, DC 20555 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-324/ LICENSE NO. DPR-62 ENVIRONMENTAL QUALIF ICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT REVISED JUSTIFICATIONS FOR CONTINUED OPERATION

Dear Mr. Vassallo:

SUMMARY

On September 18, 1984 and September 20, 1984, meetings were held with your staff to discuss the review of the justifications for continued operation (JCOs) for Brunswick-2 equipment qualification.

As a result of these meetings, Carolina Power & Light Conpany (CP&L) agreed to modify our existing JCOs in order to more closely conform with review guidelines, developed subsequent to their initial preparation.

DISCU SSION Paragraph (1) of 10 CFR 50.49 contains five criteria which are the primary considerations for the acceptability of JCOs, where appropriate.

Since the original JCOs for Brunswick-2 were written concurrent with the development of 10 CFR 50.49(1), the applicable definitive criteria to a specific JC0 is not always readily apparent to the reviewer.

The attached JCOs have been revised, where appropriate, so that the JCO's are based on at least one of the five criteria of 10 CFR 50.49(1).

In addition, some JCOs have been divided into two or more JCOs and others have been deleted due to qualification of the subject equipment subsequent to initial submittal of the JCO.

Therefore, the total number of JCOs differs from the previous submittals. is a complete list of equipment installed in Brunswick-2 for which a schedular exttasion pursuant to 10 CFR 50.49(g) has been requested.

This listing does not include any new equipment, but consolidates previous listings provided in our letters dated April 25, 1984 and August 30, 1984 into a more concise format suggested by members of your review staff and references the 10 CFR 50.49(i) criteria used in the applicable JC0's.

Those items for which our extension request is based on non-availability of qualified replacements are indicated by a footnote. As stated above, some equipment has been deleted from the list.

These items are shown in Attachment 2 for completeness.

The revised JCO's are included as Attachment 3.

8409280325 840926 PDR ADOCK 05000324 P

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In our letter of April.25, 1984, CP&L referred to environmental severity reduction' modifications which would require an extension due to material delivery.. These modifications are needad to provide a permanent design change to eliminate the potential for exceeding the plant design basis as a result of an undetected small scale leak in a limited number of high eneq3y lines in the reactor building. This potential was reported to NRC Rqgion 11 pursuant to 10 CPR 21, and a schedule for installation of the necessary modifications was transmitted to Rqgion II on April 26, 1983. A schedule extension has been discussed with Region II and the company has' agreed to establish a surveillance prq; ram to ensure the timely detection of this type of leak pending completion of necessary design modifications. Consequently, no action is required on this from your staff.

CONCLUSION Carolina Power & Light Company continues to believe that there are no known concerns relating to the environmental qualification of safety-related electrical equipment which would prevent continued safe operation of the Brunswick Plant.

Should you have any questions concerning this submittal, please contact Mr. Sherwood R. Zimmerman at (919) 836-6242.

Yours very tr L,

A 9 M f resident A. B. Cutter - Vice Nuclear Engineering & Licensing WRM/ccc (642WRM)

Attachments cc:

Mr. D. O. Myers (NRC-BNP)

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

Mr. M. Grotenhuis (NRC)

Attachment I to NLS-84-425 Pego 1 of 5 L'

JUSTIFICATION FOR CONTINUED OPERATION EQUIPMENT LIST Applicable -li0 CFR NRC TER #-

Brunswick Tag No.

Description Criteria 17 E51-F019 Limitorque Motorized (2)

Valve Operator 20 B21-F016 Limitorque Motorized (2)

E11-F022 Valve Operator-E51-F007 22,23,31, B21-F004 ASCO Valves (5)

'33,43,49,-

TD-SV-3598 1,-

51,54,55 TD-SV-3601 C12-F009 A, B B21-F028 A thru D B32-FG19 SW-124 thru 126 SW-129 thru 131 24 thru 30, CAC-V4 thru 10 ASCO Solenoid Valves (2) i 35 thru 39, CAC-V15 41,42,44,-

CAC-V47 thru 50 45,47,48, CAC-V55 50,52,53 CAC-V56 CAC-SV-4222, 4223 CAC-PV-1260 thru 1262 B32-F020 C12-F110 A, B C12-F009 A, B G16-F003 G16-F004 G16-F019 G16-F020 SW-V136 SW-V139 2(A-D)-BFIV-RB CAC-PV-3439 CAC-PV-3440 34,113,114 VA-IS-936 A thru F Johnson Services (1)

VA-ZS-936 A, B Temperature Switch VA-SV-936 A, B 62

E41-PS-N010 Static-0-Ring (2)

E51-PSL-N006 Pressure Switch 67 CAC-PT-125 7-2 Bailey Transmitters (1) 68 C32-PT-N005 A, B General Electric (2)

Pressure Transmitter (648DCS/ccc)

Attachment I to NLS-84-425

.c Ptgs 2 cf 5

s, Applicable 10 CFR NRC TER f.

Brunswick Tag No.

Description Criteria 69 E11-PDT-N002 A, B General Electric (1)(5)

Pressure Transmitter 71,72,73,74

J E11-PS-N016 A thru D Barksdale. Switches (2) 76,77,78,79.

'E11-PS-N020 A thru D 80,81,99 RIP-PSL-1200 RIP-PSL-1201

' RIP-PSL-1206 RIP-PSL-1209 thru 12 RIP-PSL-1217

-'E41-PSH-N012 A thru D

.E41-PSG-N017'A, B

-* E41-PSit-N02 7

E51-PSH-N020 I

L-

^ E51-PSH-N009 A, B

E51-PSH-N012 A thru D RIP-PSL-1218 thru 1223 RIP-PSL-1225 RIP-PSL-1227 thru 1229

B32-PS-N018 A

'* B32-PS-N018 A-1

B32-PS-N018 B -

SW-TSH-1109 thru 1112

SW-PS-1175

SW-PS-1176

~IA-PSL-3594

IA-PSL-3595 82

E41-LSH-N015 A, B Robertshaw Level Switches (2) 91 821-FS-F015 A thru H Magnetrol Flow Switch (2)

B21-FS-F015 J.thru N 821-FS-F015 P B21-FS-F015 R,-S B21-FS-F043 A, B B21-FS-F045 A, B B21-FS-F047 A, B B21-FS-F049 A, B B21-FS-F051 A, 8 B21-FS-F055 B21-FS-1227 F E41-FS-F024 A thru D E41-FS-F044 A thru D 93 VA-FT-2577 Bailey Transmitter (2) 94,122 CAC-PV-1218 C Cherry Switches (2)

CAC-PV-1219 B, C CAC-PV-1220 C CAC-PV-1221 C (648DCS/ccc)

_ _. _ _ _J

Attechment 1 to NLS-84-425 Pagm 3 of-5 Applicable 10 CFR NRC TER #

' Brunswick Tag No.

Description Criteria E41-PV-1218 D E41-PV-1219 D E41-PV-1220 D E41-PV-1221 D 95 E41-FT-N008 General Electric Flow (2)

Transmitter 96,97,98

Ell-PDIS-N021 A, B Barton Differential (1)(2)(5)

E21-FS-N006 A, B Pressure Switches

E41-FSL-N006 9 7A

E51-FS-N002 Barton Differential (2)

Pressure Switch 100~-

CAC-TE-1258-1 thru 14 Pyco Temperature (2)

CAC-TE-1258-17 thru 24 Elements 107,108,110, E41-TS-3314 thru 3318 Fenwall Temperature (1)(2) 111,112 E41-TS-3354 Switches E41-TS-3488 E41-TS-3489 I

E41-TS-3319 thru 23 L

E51-TS-3355 E 51-TS-348 7 109 B21-TS-N010 A thru D Fenwall Temperature (2)(4)

Switch 115 2(A-D)-BFIV-RB NAMCO Position Switch (2) 124,125,126,

B32-F019 lioneywell Limit (2)(4)(5) 127,128,129

B32-F020 Switches

CAC-V47

CAC-V48

CAC-V55

CAC-V56 CAC-PV-1200 B CAC-PV-1205 E CAC-PV-1209 A, B CAC-PV-1211 E CAC-PV-1225 B CAC-PV-1227 A thru C CAC-PV-1227 E CAC-PV-1231 B CAC-PV-1260 thru 1262

B21-F003

B21-F004 (648DCS/ccc) l

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Attcchment 1 to NLS-84-425 Paga 4 of 5 tj_

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.o Applicable 10 CFR NRC TER #

Brunsdick Tag No.,

Description Criteria

~

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130,131,133, * - DL8-RS1 s -

Honeywell Microswitches (2) 134,135

. DL9-RS1 b

. DM4-RSI

DMS-RS1'

B11-RS1

B4 7-RS1

DM7-RSI.

DM8-RS1

DN6-RS1

DK8-RS 1

B41-RS1

B11-RS

DLO-RS1'

DL1-RS1 DL2-RSI-.

DS 7-RSI -

-* B45-RS1

B 21-CS-3412

B43-RS1

'DH3-RS1

DH2-RS1

-B50-RS1

B49-RS1

321-CS-3327

. B21-CS-3329

B21-CS-3345

'132,142,144,'*

MCC-2XA General Electric IC7700 (2) 145,146,147

MCC-2XA-2 Motor Control Center

MCC-2XB

MCC-2XB-2

MCC-2XC

MCC-2XD

MCC-2XDA

MCC-2XDB

- MCC-2XE,

MCC-2XF'

- MCC-2XH

MCC-2XJ

MCC-2XK 138

E11-C001 A thru D General Electric Motors (2) 141,155

E41-C002 Tarry Steam Turbine (1)

HPCI Pump System 143 DBO-74-17 Agastat Time Delay Relay (1) 148 D21-RE-N010 A, B General Electric (1)(2)(3)(4)

Radiation Detectors (648DCS/ccc)

em; x-to NLS-84-425 1

Paga 5 of 5

'f.

Applicable 10 CFR NRC TER #

Brunswick Tag No.

Description Criteria 151 Various Termination AMP (Nylon Insulation (5)~

Lugs Outside Sleeves)

' Containment 156

SGT-FILT-2A-RB FARR Standby Gas (1)(5).

SGT-FILT-28-RB

' Treatment System Components

' 1 64

Cable Raychem Control Cable (2) 172 SKV Terminations Burndy Electrical Lugs (2) with OKonex Butyl Rubber Tape and OKonite No. 35 Jacketing Tape

-179,181

Terminal Blocks General Electric (2)

Terminal Blocks 18 0 Terminal Blocks General Electric (2)

Terminal Blocks 182

: Terminations Outside Curtis Type "L" (2)

Containment Te rminal -Blocks NONE B21-FT-4157 thru 4167.

NDT Int'l Accelerometers (2)

NONE E51-C002-H Square D. Float Switch (2)(5)

NONE' C12-F010-L Namco Limit Switch (2)

C12-F011-L E51-C002-LS4 NONE

B32-CS-F019 Sentry Control Switches (2)

B32-CS-F020 NONE

. NP6-MOT-M1, M2 DOERR Motors and Control (2)

NP 7-MOT-M1, M2 Panels-LA-RX 18-RX

-

These items to be' deferred due to non-availability of qualified replacements.

All other items to be deferred due to installation problems.

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(648DCS/ccc)

I'"3i Attechnent 2 to NLS-84-425 Pzgs 1 of 1 i -

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~ Equipment which has been qualified and therefore removed f rom the list of equipment

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. to be' deferred..

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-Brunswick Tag No.

Description-l'

. CAC-V22 Limitorque Motorized Valve

[

Operator

(

57,59 CAC-PDS-4222 Barton Switches CAC-PDS-4223 CAC-PSE-N001 A thru D

'60,61,63' C72--PS-N002 A thru D Static-0-Ring (2)

E11-PS-N010 A thru D Pressure Switch E11-PS-N011 A'thru D E11-PS-N019 A thru D

~

E21-PS-N008 A, B E21-PS-N009 A, B.

75 -

E51-PS-N019 A thru D Barksdale Switches 85 B21-LITS-N026 A, B -

Yarway Switches

-116,117,118 CAC-V9 Bettis Limit Switches CAC-V10 CAC-VIS CAC-V49 CAC-V50'

-169-NS2 Pyle National' Connector s

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k 5

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O ATTACHMENT 3 70 NLS-84-425 JUSTIFICATION FOR CONTINUED OPERATION

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t TER NO.:

1 COMPONENT I.D. NO.:

CAC-V22 MFG /M00. N0.:

LIMITORQUE MODEL SMB-000 VALVE OPERATOR LOCATION:

REACTOR BUILDING -17' TECHNICAL DISCUSSION:

Component materials of the Limitorque Motorized Valve Operator have been identified and qualification documentation located. The qualification data has been evaluated par D0R guidelihes and by applying Arrhenius techniques.

Results of this evaluation indicate that the' Class B insulation system, melamine switches, and internal wire insulacion materials are qualified for 40 years at the reactor building maximum service temperature (104'F) and postulated accident conditions. (

Reference:

Limitorque Test Report No.

B003).

The 'above items have been removed from the list titled " Items to be deferred due to qualified replacements not available."

r Thertifore, continued operation is justified.

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TER NO.:

17 COMPONENT I.D. NO.:

E51-F019 MFG /M00. NO.:

LIMIT 0P.QUE MODEL SMB-000 VALVE OPERATOR LOCATION:

REACTOR BUILDING -17' TECHNICAL DISCUSSION:

Component materials of the Limitorque Motorized Valve Operator have been identified and qualification documentation located. The qualification data has been evaluated per D0R 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 ef fects 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 reactor building parameters (

Reference:

Limitorque Test Report No. 600376A).

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

~ 22.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months (

Reference:

Limitorque Test Report No. B0003). A comparative analysis ' f the Limitorque "Superheat" test reveals that the internal o

. temperature of the valve operator and motor will not reach 250*F during the initial 100 seconds of accident exposure. Thus, it is judged that the test temperatufe profile was actually more severe that the plant requirement.

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

Therefore, continued operation is justified.

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20 COMPONENT I.D. NO.:

B21-F016, E11-F022, E51-F007 MFG / MOD. NO.:

LIMITORQUE MODEL SMB-00 VALVE OPERATOR LOCATION :

DRYWELL ELEVATION 17', 80'

TECHNICAL' 0ISCUSSION:

Component materials of the Limitorque Motorized Valve Operators have been identified and qualification documentation located. The qualification data has _ been evaluated per D0R guidelines and by applying Arrhenius techniques.

Results or this evaluation indicate that _the Class H motor insulation system, malamine switches, and internal wire insulation materials are insensitive to

-thermal aging effects at tile 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 accident 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 motors were successfully tested to 2'X 10gditionally, the Class li insulated the postulated plant accident at BSEP.

rads gamma total integrated dose

'(

Reference:

Limitorque Report No. FC-3327) which envelops the BSEP 3

requirement.of 1.1-X 10 Rads gamma.

Thus, it is judged that the Class H insulated motors meet the criteria set forth in 10CFR50.49, paragraph (1)(2).

There fore, continued. operation is justified.

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1 TER NO.:

22, 23, 31, 33, 43, 49, 51, 54, 55 COMPONENT I.D. NO.:

821-F004 B21-F028A, B, C, D TD-SV-3598, 3601 B32-F019 C12-F009A,B SW-124, 125, 126, 129, 130, 131 MFG / MOD. N0.:

ASCO HBB302C25 HT8302 HTX8320A70 WPHT8321A1 8344A5 HT8344A5 HT832322

.HB8342A4 LOCATION:

DRYWELL, RHR ROOM CORE SPRAY ROOM, REACTOR BUILDING TECHNICAL DISCUSSION:

These valves have been replaced with fully qualified valves (ASCO NP series),

are not.in service, or no longer have a safety related function and would not cause significant degradation of any safety function under an accident environment.

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

'Therefore, continued operation is justified.

1 l i

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24, 25, 26, 27, 28, 29, 30, 35, 36, 37, 38, 39, 41, 42, 44, 45, 47, 48, 50, 52, 53 i

l COMP 0NENT I.D. NO.:

CAC-V4, 5, 6, 7, 8, 9, 10, 15, 47, 48, 49, 50, 55, 56 i

CAC-SV-4222, 4223 CAC-PV-1260, 1261, 1262 i

^

832-F020 C12-F110A, B' C12-F009A, B G16-F003, 04,19, 20 SW-V136, 1?9 2( A-D)-BFIV -RB CAC-PV-3439 CAC-PV-3440 MFG / MOD. NO.:

HT832322 JV-182-084 HT8316 HB8302C25RU HT8262C71 8302 Hi8211833 WPHT8321A1 8321A6 HT8321A6 HB8342A4 8262023 HV-180-414 HT80033 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 materials of the ASCO solenoid valves have been identified and

~ qualification documentatirn. located. The qualification data has been evaluated per D0R guidelities 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 maximum 104*F temperature.

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

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 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 UE&C Specification. The Class 'H' coils utilized in these valves have been designed for satisfactory

. operation at 165 F ambient.

i TER-24-53 Page 2 Valves 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 themal 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . 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 these valves are capable of satisfactory operation during the accident and post-accident conditions stated in the UE&C Specification".

Although ethylene propylene was the elastomer in the tested valves, the acs ual 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 Masoneilan test report (1003, dated 4-19-73) shows that WFHT8300861 valves successfully functioned during and after exposure to 310*F and 65 psig steam for 23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months .

Information or; radiation damage ' values shows that the postulated TID of 1 X 107 will not significantly degrade the function nf the nonmetallic materials except for the acetal disc holder. Testing has been performed on acetal retaining washers to 1 X 107 rads with successful results (

Reference:

MCC

. Powers Report No. 734-79.002, Rev. 3).

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

Therefore, continued operation is justified.

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-TER NO.:

34, 113, 114, 123 COMPONENT I.D. NO.:

VA-TS-936A, B, C, D, E, F VA-ZS-9368, A VA-SV-936B, A MFG / MOD. NO.:

JOHNSON SERVICES; ALLEN BRADLEY LOCATION:

RHR ROOM TECHNICAL DISCUSSION:

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 rcoms and the HPCI room simulataneously via interconnecting HVAC ductwork.

The room B fan cooling unit equipment (VA-TS-9368, C, F; VA-ZS-936B; VA-SV-9368) 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 (i)(1) and (i)(5).

Therefore, continued operation is justified.

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TER NO.:

~ 57 & 59 COMPONENT I.D. NO.:

CAC-PDS-4222 CAC-PDS-4223 CAC-PSE-N001A-D MFG / MOD.. NO. :

BARTON 288A, 289A LOCATION:

REACTOR BUILDING ELEVATION 20' and 50'

' TECHNICAL DISCUSSION:

Test data has been obtained which qualifies the subject switches to the BSEP environmental corditions, including the postulated accident conditions (Reference Barton Engineering Reports R3-288A-1 and R1-288A-11).

The above items have been removed from the list titled " Items to be deferred

'due to qualified replacements not available."

Therefore, continued operation is justified.

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60, 61, & 63 COMPONENT I.D. N0.:

C72-PS-N002A E11-PS-N0100 E11-PS-N019C C72-PS-N002B E11-PS-N011A E11-PS-N0190 C72-PS-N002C E11-PS-N0118 E21-PS-N008A C72-FS-N002D E11-f3-N011C E21-PS-N008B E11-PS-N010A E11-PS-N0110 E21-PS-N009A E11-PS-N010B E11-PS-N019A E21-PS-N009B E11-PS-N010C E11-PS-N019B

- NFG/M00. NO.:

STATIC 0-RING MODEL 12NAA4-X10TT AND SN-AA3-X9-STT PRESSUE SWITCH LOCATION:

REACTOR BUILDING TECHNICAL DISCUSSION:

4 Component materials of the Static 0-Ring pressure switch have been identified and qualification documentation located. The qualification data has been evaluated per D0R guidelines and by applying Arrhenius techniques. This evaluation qualifies the pressure switches to the postulated normal and.

accident conditions at BSEP for 40 years (

Reference:

Viking Lab. Report No.

30203-2, dated November 20,1973).

These prescure swithes complete their safety function in less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the accident initiation.

- The dbove items have been removed from the list titled " Items to be deferred due to qualified replacements not available."

Therefore, continued operat1on is justified.

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1 TER N0.:

62 COMPONENT I.D. NO.:

E41-PS-N010 E51-PSL-N006 MFG / MOD. NO.:

STATIC 0 RING PRESSURE SWITCH 6N-AA21X9SVTT AND 6N-AA21-X9-ST LOCATION:

REACTOR BUILDING EL. -17' TECHNICAL DISCUSSION:

Component materials of the Static-0-Ring (SDR) pressure switch have been identified and qualification documentation on a similar S0R pressure switch has been obtained. The qualification data has been evaluated per DDR guidelines and by applying Arrhenius 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 tem;+rature peak of 288*F for 70 seconds. The accident temperature then decreases to 205*F at 100 seconds and returns to ambient after approximately 20 minutes. This postulated peak temperature transient has been conpseed to accident test data obtained (212*F for 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ) 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 is based on the severity of the test performed in ccmparison 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 threshold greater than the maximum postulated total integrated dose of 2 X 106 rads gamma.

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

Therefore, continued operation is justified.

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.2-4 TER NO.:

67 COMPONENT I.D. NO.:

CAC-PT-1257-2

- NFG/M00. NO.:

BAILEY--KQ12C

-LOCATION:

RHR ROOM TECHNICAL DISCUSSION:

The information provided the operator by these transmitters is also provided by an independent, redJndant, and fully qualified transmitter (Rosemount). As such-the safety function of this equipment can be accomplished by alternative equipment.

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

Therefore, continued operation is justified.

)

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i TER N0.:

68 COMPONENT I.D. NO.:

C32-PT-N005A, B MFG / MOD. NO.:

GENERAL ELECTRIC MODEL 551032GKZZ2 PRESSURE TRANSMITTER LOCATION:

REACTOR BUILDING 50' TECHNICAL DISCUSSION:

Partial qualification' documentation has been obtained for a similar pressure transmitter with the same components and of similar application. The data was evaluated per the 00R guidelines.

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

Testing ha's 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. 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 200*F..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 requirenent of 1 X 10 rads gamma is less than the lowest radiation damage threshold of the transmitter components.

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

Therefore,' continued operation is justified.

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TER N0.:

69 COMPONENT I.D. NO.:

E11-PDT-N002A, & B MFG /M00. N0.:

GENERAL ELECTRIC 552032HKZZ2 PRESSURE TRANSMITTER LOCATION:

REACTOR BUILDING RHR ROOM TECMICAL DISCUSSION:

These instruments measure the o P across the 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 (i)(5).

Therefore, continued operation is justified, i

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e TER NO.:

'71, 72, 73,.74', 76, 77, 78, 79, 80, 81, A 99 COMP 0NENT I.D. N0.:

E11-PS-N016A E41-PSH-N012A RIP.-PSL-1220 E11-PS-N016B E41-PSH-N012B RIP-PSL-1221 E11-PS-N016C E41-PSH-N012C RIP-PSL-1222 E11-PS-N016D' E41-PSH-N012D RIP-PSL-1223 E11-PS-N020A E41-PSH-N017A RIP-PSL-1225 E11-PS-N0209 E41-PSH-N017B RIP-PSL-1227 1

E11-PS-N020C E41-PSH-N027 RIP-PSL-1228 E11-PS-N0200 E51-PS-N020 RIP-PSL-1229~

RIP-PSL-1200 E51-PSH-N009A B32-PS-N018A RIP-PSL-1201 E51-PSH-N0098 B32-PS-N018A-1 RIP-PSL-1206 E51-PSH-N012A B32-PS-N018B RIP-PSL-1209 E51-PSH-N012B SW-TSH-1109 RIP-PSL-1210' E51-PSH-N012C SW-TSH-1110 RIP-PSL-1211 E51-PSH-N012D SW-TSH-1111 l

RIP-PSL-1212 RIP-PSL-1218 SW-TSH-1112 RIP-PSL-1217 RIP-PSL-1219 IA-PSL-3594,3595 SW-PS-1175,1176 MFG / MOD. NO.:

BARKSDALE B2T-M12SS D2H-M150SS D2T-M18SS 02T-M150SS c

P1H-M340SS TC9622-1 T2H-M251S-12 D2T-M80SS t.

. 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

- this evaluation indicate that all materials, except for Buna-N rubter, 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 accident temperature then decreases to 145*F within one (1) hour of event

. i niti ati on. This postulated peak temperature transient has been compared to accident test. data obtained (212*F for 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , Ref. AETL TR #596-0398) for

' those switches. _ Although the testing does not envelop the postulated peak c.

l 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 l.

exposure to the accident peak temperature.

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 reduced to nil during the postulated accident.

Also, the component' nonmetallic materials have been sucessfully radiation aged levels greater _ than 1 X 10gg (while being used in similar applications) to during. qualification testi l

j rads gamma, the postulated accident TIO for BSEP.

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

.Therefore, continued operation is justified.

l

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t TER N0.:

75 COMPONENT I.D. NO.:

' E51-PS-N019A, B, C, D MFG / MOD. NO.:

BARKSDALE MODEL P1HM85SSV LOCATION:

REACTOR BUILDING ELEVATION 20' TECHNICAL DISCUSSION:

Test data has been obtained which qualifies the subject switches to the BSEP environmental conditions, including postulated accident conditions.

(

Reference:

AETL Test Report 596-0398)

The above items have been removed from tae list titled " Items to be deferred due to qualified replacements not available."

~ Therefore, continued operation is justified.

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TER NO.

82 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' TECWICAL DISCUSSION:'

Partial qualification documentation has been located for the Robertshaw level switches.. The qualification data has been evaluated per DOR guidelines and by applying Arrhenius techniques.

The' switch nomnetallic components are exposed to the the reactor building postulated accident temperature peak of_ 288'F for only 70 seconds. 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 compared to.the accident-test data obtained-(212'F,10 psig for 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months ,

Reference:

Robertshaw unnumbered test report dated March 28,-1983)

-for these switches.

Although the testing does not envelop the postulated peak accident

. temperature, it is judged that no detrimental effects to switch o~ eration will p

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.

In addition, the Brunswick switches are located in a Nema Type 7,9 explosion f proof enclosure where the effects of direct steam impingement / humidity would

'be reduced to nil during the postulated accident.

Also, the component nonmetallic materials have been successfully radiation aged during qua.lification testing. (while.being used ig similar applications)

' to levels-greater. than the BSEP requirement of 1 X 10 rads gamma.

0perationally, the level switcher located outside containment are used to signal high suppression pool level to the HPCI system.

i

'In. the event of a large break LOCA for which the HPCI system cannot maintain RP'l 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 system can maintain RPV' 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'i.

t'

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._~.

J TER-82 Page 2 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, i

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85

. COMPONENT I.D. NO.:

821-LITS-N026A B21-LITS-N026B MFG / MOD. N0.

fARWAY 4418 EC LOCATION :

REACTOR BUILDING 50' TECHNICAL DISCUSSION:

Test data has been obtained which qualifies the subject switch to the BSEP reactor building normal and the postulated accident conditions (

Reference:

Yarway Report No. 3232-3155 and 5628-3509).

The above items have.been removed from the list t1tled " Items to be Deferred Due to Installation Problems".

Therefore, continued operation is justified.

A s

TER NO.:

91 COMPONENT I.D. NO.:

B21-FS-F015A B21-FS-F015N 821-FS-F051A B21-FS-F015R B21-FS-F015P B21-FS-F051B B21-FS-F015C B21-FS-F015R B21-FS-F055 B21-FS-F0150 B21-FS-F015S B21-FS-1227F B21-FS-F015E B21-FS-F043A E41-FS-F024A B21-FS-F015F B21-FS-F043B E41-FS-F0243 B21-FS-F015G B21-FS-F045A E41-FS-F024C B21-FS-F015H B21-FS-F045B E41-FS-F0240 821-FS-F015J B21-FS-F047A E41-FS-F044A B21-FS-F015K B21-FS-F0478 E41-FS-F044B B21-FS-F015L B21-FS-F049A E41-FS-F044C B21-FS-F015M B21-FS-F049B E41-FS-F044D MFG /M00. NO.:

MAGNETROL MODEL F-521 FLOW SWITCH LOCATION:

REACTOR BUILDING (VARIOUS ELEVATIONS)

- TECmICAL DISCUSSION:

Component materials of the Magnetrol flow switch have been identified. These materials have been evaluated per D0R 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 maximan 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 material used in the flow switch. The analysis indicated that each material t

l has a radiation damage threshold level equgl to or greater than the maximum postulated total integrated dose of 1 X 10 rads gamma.

l

[

In addition, an operational analysis has been performed to determine the effects of failure '(misleading information, grounds and spurious operation) of i

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 or there are alternate systems available to achieve the same safety functions. Sufficient procedural direction and alternate information is

_available for.the operator to diagnose or respond safely to misleading l

indications.

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

Therefore, e.ontinued operation is justified.

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TER N0.:

93 COMPONENT I.D. NO.:

VA-FT-2577 MFG /M00. NO.:-

BAILEY BQ13221

. LOCATION:

REACTOR BUILDING ELEVATION 50' TECHNICAL DISCUSSION:

- Component materials of the Bailey transmitters have been identified and 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 techniqu'es. Results of this

-evaluation indicate that these transmitters consist of essentially the same

, materials and components as Rosemount 1153 transmitters. The Bailey transmitter includes Teflon and Viton 0-rings. These o-rings are used as static seals between the flange adapter and process flange (Teflon), the process flange and sensor module (Viton), and the electrical housing and cover

-(Viton). -These materials were evaluated at the normal and peak accident conditions and will not experience significant degradation of performance.

The Rosemount transmitters were tested to parameters which envelop' the BSEP reactor building conditions (

Reference:

Rosemount Reports 3788, 108025..and 08300040). Based on the similarity of the Bailey transmitters to the Rosemount transmitters, the testtag levels, and the environment at this logation (104*F normal,. _< 200*F fnr less' than 10 minutes peak accident,1 X 10 rads TID) use of the Bailey transmitters is justifed.

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

Therefore, continued operation is justified.

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TER NO.:

94, 122 COMPONENT. I.D. NO.:

VARIOUS; CAC-PV-1218C, -1219B, -1219C, -1220C, -1221C E41-PV-12180, -1219D, 12200, 12210 MFG / MOD. NO.:

CHERRY E2360H LOCATION':

REACTOR BUILDING 20' AND 50'; RHR ROOM 1 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. The qualification data has been evaluated per D0R guidelines and by applying Arrhenius techniques. Results of this evaluation indicated that the nonmetallic compc.ients have greater than 66 years expected life at the n3ximum 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ). 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.

h Additionally, -radiation testing on switches of-the samg material and

. application supports a qualification level of 3.6 X 100 rads gamma, although the' testing does not envelop.the postulated total integrated dose of 1 X 107 rads gamma, a radiation threshold analysis shows that the radiation threshold analysis for each material 'rsed in switch is greater that 1 X 107 rads gamma except for the Delrin button. For the Delrin button there is testing to support the use of this materiai in a mechanical application to a radiation level of 1 X 107 rads ganna (

Reference:

MCC Powers Report No. 734-79.002,

~Rev. 3).

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

P effects of ~ failure (misleading information, grcunds, and spurious operation) l of these items in both LOCA and HELB environments. The operational analysis i

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).

l l

Therefore, continued operation is justified.

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95 COMPONENT I.D. N0.:

E41-FT-N008 MFG /M0D. N0.:

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 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).

The accident simulation included a peak temperature of 180*F. Additionally, a separate test subjected the transmitter to a 68'F to 158'F at 100%RH test.

The tests do not enveicp the BSEP requirement of 199'F (3" RCIC line break).

However, the accident )2d temperature excursion will not cause significant degradation of equipme.t cperation during that period of exposure above the test maximum temperature t

Reference:

General Electric Report 327, File DV145C3007 and General Esectric 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.

In:the vdnt 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 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 provide.d 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, l

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

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

Therefore, continued operation is justified.

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._TER NO.:

96, 97, 98 COMPONENT I.D NO.:

E11-PDIS-N021A,B E21-FS-N006A,B-E41-FSL-N006 MFG / MOD. N0.:

BARTON 289 LOCATION:

RHR ROOM, CORE SPRAY ROOM TECHNICAL DISCUSSION:

~

'These items control the minimum flow valves for the RHR, Core Spray and HPCI pumps. A minimum flow valve is generally 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 circuit would act to open the valve. Unplanned opening of the minimum flow valve during injection

- would divert very little emergency flow from the RPV because of flow restricting orifices in each of the minimum flow lines.

-If the instrument were to fail, showing high flow, the circuit Jnould act to shut the valve. During injection the valve would aircady 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.

(

~.The plant can be safely shutdown without these instruments.

- An. additional analysis has been performed to insure that pressure switches will maintain electr' cal integrity during the postulated accident.

Component materials of the Barton differential pressure switches have been identified 'and qualification documentation located. The qualification data has-been evaluated per D0R guidelines and by applying Arrhenius techniques.

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 exposed to the plant postulated accident temperature peak of-288'F for 70 seconds. The accident temperature

'then deceases to 205'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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ) for this switch

(

Reference:

AETL Test Report No. 596-0399). 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 tenperature transient. This assessment is based on the severity of the test performed and the short time for heat transfer through the heavy L

' metal casing.

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TER-96-98 Page 2

' Additionally, radiation testing on the subject switches supports a qualification level of 3.6 X-106 rads gamma. Though the testing does not envelop the ~ postulated total integrated dose of 1 X 10Y rads gamma, a radiation threshold analysis shows that the radiation threshold for each material used in the switch is-greater than 1 X 107 rads gamma. For the Viton

'o-ring -there is testing to support the use gf this material in an 0-ring application up to radiation level of 2 X 10' rads gamma (

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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g TER NO.:

97A COMPONENT I.D. NO.:

E51-FS-N002 MFG / MOD. NO.:

BARTON 289 LOCATION:

REACTOR BUILDING RHR ROOM TECH:., CAL DISCUSSION:

Component materials of the Barton differential pressure switches have been identified and qualification documentation located. The qualification data has been evaluated per D0R guidelines and by applying Arrhenius techniques.

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

The pressure switch nonmetallic materials are exposed to the plant postulated dCCident temperature peak of 288'F for 70 seconds. The accident temperature then decreases to 205'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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ) 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 is based on the severity of the test performed and the short time for heat transfer through the heavy metal casing.

Additional!y, radiation testing on the subject switches supports a 6

envelop the postulated total integrated dose of 1 X 10ge testing does not qualification level of 3.6 X 10 rads gamma. Though t rads gamma, a material used in the switch is greater than 1 X 10}on threshold for each radiation threshold analysis shows that the radiat rads gamma except for the 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).

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

Therefore, continued operation is justified.

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TER NO.:

100 COMPONENT I.D. N0.:

CAC-TE-1258-1 T0 14 CAC-TE-1258-17 TO 24 MFG / MOD. NO.:

PYC0 100 0HM PLATINUM RTD

- LOCATION :

DRYWELL

' TECHNICAL DISCUSSION:

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 enveloping BSEP normal and accident service conditions (

Reference:

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

. The similarity of the installed equipment has been confirmed by Pyco.

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

Therefore, continued operation is justified.

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107,108,110,111, A 112 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-3323 E41-TS-3354 E51-TS-3355 E41-TS-3488~

E51-TS-3487 E41-TS-3489 MFG / MOD. NO.:

FENWAL TEMPERATURE SWITCH 17002-40 LOCATION :

REACTOR BUILDING EL. -17' AND AB0VE TECHNICAL DISCUSSION:

These instruments are temperature sensors which monitor +c~,peratui-e5 he arecs where the HPCI/RCIC steam line is located and initiate an 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 until 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 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 107 rads gamma over 30 days.

In the Fenwal temperature switches the Teflon lead wire is sandwiched between two 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 (1)(1) and (i)(2).

1-Therefore, continued operation is justified.

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TER NO.:

109 COMP 0NENT I.D. NO.:

821-TS-N010A B21-TS-N010C B21-TS-N010B B21-TS-N010D MFG / MOD. NO.:

FENWAL TEMPERATURE SWITCH 17002-40 LOCATION:

REACTOR BUILDING (TUNNEL) EL. 20' TECHNICAL DISCUSSION:

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

Reference:

Patel Engineer's Qualification Report No. PEI-TR-831200-1). The tested model was identical to the one iMPilled a; B'AP except the -lead wire insulation was different. The installed switches have teflon insulated lead wires and the tested unit had Rockbestos crosslinked polyethylene insulated lead wires.

Te{lon has a high temperature rating and the radiation threshold value is 5 X 10 rads for electrical 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 function immediately after the accident initiation. Therefore, the temperature switch lead wires will n estimated radiation dose of.1.5 X 10gt be significantly degraded by an rads before completing their safety function.

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

Therefore, continued operation is justifled.

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115 CbMPONENT I.D. NO.:

2(A-D)-BFIV-RB MFG / MOD. NO.:

NAMC0 D2400XR LOCATION:

REACTOR BUIL9ING 80'

' TECHNICAL DISCUSSION:

Component materials of the NAMC0 2400XR position switch have been identified. The materials have' been evaluated per 00R guidelines and by applying Arrhenius 1echniques. Results of this analysis indicate that all materials, except for Buna-N rubber (used as 'a binder in the asbestos gasket),

have greater th'an forty (40) years demonstrated qualified. life at the maximum 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 left undisturbed, no significaant degradation would occur.

The analysis performed on the D2400XR switch is based on testing conducted on NAMC0 series SL3 switches (generically similar in materials, construction, and operation).. These switches were exposed to a 310*F and 65 psig steam avironment (

Reference:

Masoneilon Test Report 1003, dated 4-19-73) which c4ceeds the BSEP requirement.

A radiation analysis indicates thgt the lowest damage threshold for the nonmetallic materials is 8.6 X 10 rad This damage threshold value envelops the BSEP requiremnt of 1 X 10g gamma.

rads gamma.

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

Therefore, continued operation is justified.

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-y-,

+

-c-

-=v--e--

T-e-w-*,*,-

.+ptv-,v-~r-++,-+ww+-e p--w y-v-

a-.

.. +.:.a. >..

x TER N0.:

116, 117, a 118 COMPONENT -I.D. N0.:

CAC-V9 CAC-V49 CAC-Y10 CAC-V50

.CAC-V15 MFG / MOD. NO.:

. BETTIS TYPE RX-41 AND RX-341 LOCATION:

REACTOR BUILDING 50' AND 107'

. TECHNICAL DISCUSSION:

Component materials of the Bettis Limit Switches have been identified and qualification documentation located. The qualification data has been evaluated per D0R guidelines and by applying Arrhenius techniques. Results of this evaluation indicate that the switch mechanism -(Microswitch type BZ) have greater than forty (40) years demonstrated qualified life at the maximum reactor building temperature (104*) and postulated accident conditions.

(

References:

(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," July 15, 1975.

(3)

" Evaluation of Asbestos-Free Plastics for 250*

Basic Swith," Micr6 Switch, February 21, 1979. (4)

Environmental Test,"

9993 Barksdale, August 13, 1975.

The above items have been removed from the list titled " Items to be deferred

- due to qualified replacements not available."

Therefore, continued operation is justified.

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--y

-v--gr w e-

,v-e.,y--y-.--g-eww,-

-=.

,.---.--ar-.,.---,--

,-v

TER NO.:

124, 125, 126, 127, 128, 129 COMPONENT I.D. NO :

832-F019, B32-F020 CAC-PV-1227A*

CAC-V47, CAC-V48 CAC-PV-1227B*-

CAC-VSS, CAC-V56 CAC-PV-1227C*

CAC-PV-1200B*

CAC-PV-1227E*

CAC-PV-1205E*

CAC-PV-1231B*

CAC-PV-1209A*

CAC-PV-1260 CAC-PV-1209B*

CAC-PV-1261

.CAC-PV-1211E*

CAC-PV-1262

,CAC-PV-1225B*

B21-F003 B21-F004 MFG /M00. NO.:

HONEYWELL MODEL OP-AR AND *0PD-AR LIMIT SWITCHES LOCATION :

DW 17' (B32-F019, B21-F003, 821-F004 ONLY)

RX 20' & 50' (ALL OTHERS)

TECHNICAL DISCUSSION:

Component materials of the Honeywell limit switches have been identified and partial qualification documentation located. The qualification data has been evaluated per 00R guidelines and by applying Arrhenius techniques. Results of this evaluation indicate the limit switches inside 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 Switch, July 15, 1975, (3) " Evaluation of Asbestos-Free Flastics for 250 F Basic Switch," Micro Switch, February 21,1979, (4) " Environmental Test," 9993 Barksdale, August 13,1975).

The analysis for the switches located in the reactor building meet the criteria of10CFR50.49, paragraph (i)(2).

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 envelop the BSEP postulated drywell accident conditions.

This switch provides only valve positioq indication to the control room for the inboard reactor water sample valve (832-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.

~;

TER 124-129 Page 2

~ 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 ana~ lysis meets the criteria of 10CFR50.49, paragraph (1)(2)(4)(5).

Therefore, continued operation is justified.

i

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9

TER NO.:

130,131,133,134, AND 135 CDMPONENT I.D. N0.:

DL8-RS1 DM7-RS1 DLO-RS1 B43-RS1 DL9-RS1 DM8-RS1 DL1-RS1 DH3-RS1 DM4-RSI DN6-RS1 DL2-RS1 DH2-RS1 DMS-RS1 DK8-RS1 DS7-RS1 B50-RS1 B11-RS1 B41-RS1 B45-RS1 B49-RS1 B47-RS1 B11-RS B21-CS-3412 B21-CS-3327 B21-CS-3329 B21-CS-3345 MFG / MOD. NO.:

HONEYWELL MICR0 SWITCH, TYPES: PTSEA202FB52, TPSHA201, PTKBC2221CCF9, PTKBC2221, AND PTSHE202CB97 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.

The PT series switch have been tested at 185*F for 767 hours0.00888 days 0.213 hours 0.00127 weeks 2.918435e-4 months (more than 30 days) as per Honeywell Micro Switch Qualification Report No. 24407. For radiation the switches have been analyzed as per Honeywell Engineering Report

~

0 No. LTR 15027-1 togeacceptableto5X10 rad TID. BSEP maximum anticipated radiation is 1 X 10 rads. TID.

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 tine 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.

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

Therefore, continued operation is justified.

1x

.v

.. +. -

9 3-

-TER NO.:

132,-142, 144, 145, 146, & 147 i COMPONENT I.D NO.:

MCC-2XA, MCC-2XA-2, Mcc-2XB, MCC-2XB-2, MCC-2XC, MCC-2X0, MCC-2XDA, MCC-2XDB, MCC-2XE, MCC-2XF, MCC-2XH, MCC-2XJ, MCC-2XK MFG /M00. NO.:

- GENERAL ELECTRIC IC 7700 MOTOR CONTROL CENTER

' LOCATION:

REACTOR BUILDING TECHNICAL DISCUSSION:

Test data applicable to the environmental qualification of tha General Electric Series IC 7700 motor control center has been identified and

- qualification documentation located. The-qualification data has been evaluated per 00R guidelines and by' applying Arrhenius techniques.

i

- A preliminary assessment of the test data, performed by General Electric Co.,

indicated that the test data can be used to demonstrate qualification of tF?

1 motor control centers to be BSEP nonnal and postulated accident conditions (Reference - Environmental Qualification Assessment Report - Phase I, G. E document number 710-03-0258).

Subsequent to the preliminary assessment, G. E. issued a second document, G.'E. _ report number NEDC-30322-P...This document contains detailed Engineering Change Notice (ECN) -reviews, Product Analysis Reports, and Similiarity

' Analysis Reports on specific components contained in the motor control centers.

-(THED circuit breakers, CR109 magnetic starters, and a control power-

' transformer). This report also indicates that the test data obtained demonstrated qualification of the IC 7700 motor control center to the BSEP normal and postulated accident conditions.

p I

The ' final report on the qualification status cf the IC 7700 motor control

. center is currently being prepared by-General Electric.

Based upon *.he. test data obtained and the assessments performed, this analysis meets the criteria of 10CFR50.49, paragraph (1)(2).

Therefore, continued operation is-justified.

l.

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1 TER NO.:

138 COMPONENT I.D.- NO.:

E11-C001A, B, C, D i

MFG / MOD. NO.:

GENERAL. ELECTRIC SK821161C11

' LOCATION :

REACTOR. BUILDING - 50' TECHNICAL DISCUSSION:-

1The' above motor is a horizontal induction motor with a Class B custom Polyseal insalation.

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.

- Test data has been obtained for vertical induction motors with the same insulation class (G. E. -Document NEDC-30294). The test data obtained envelops the postulted 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 conditions.

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 (1)(2).

Therefore, continued operation is justified.

f.

l l

I'

TER NO.:

+

141, 155 COMPONENT I.D. NO.:

E41-C002

- MFG /M00. 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 conjunction 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 Proceaures.

~

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

- Therefore, continued operation is justified.

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TER N0.:

143 COMP 0NENT I.D. NO.:

DB0-74-17 MFG / MOD. NO.:

AGASTAT 7022AC TIME DELAY RELAY LOCATION:

REACTOR BUILDING RHR ROOM TECHNICAL DISCUSSION:

~

BSEP has one' Agastat time delay (model 7022AC) installed in the control circuit of RHR pump room cooler fan A-FCU-RB. An autenatic start signal to 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 annun:'. tor alarm is sounded in the control room indicating that fan A-FCU-RB has f ailed to start.

It is important to note that this relay does not perform any controi function t) start or stop the fan; it only gives indicati:n.

The result of the failure of this relay would possibily be:

(1) Loss of control power to th* fan A-FCU-RB and (2) Loss of alarm ;o the control room that fan A-FCU-RB has failed to start.

If control power is not lost, the fan would start as desigt.cd. 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 scen 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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ITER'NO.:

-148

. COMPONENT I.D.- NO.:

D12-RE-N010A, B

-MFG / MOD.-NO.:

G.-E. MODEL 194 X 927G RADIATION DETECTORS LOCATION:

REACTOR BUILDING EXHAUST AIR PLENUM EL. 80' TECCICAL DISCUSSION:

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

radiation detectors. The test data was evaluated per the 00R guidelines and using Arrhenius techniques. The results of this evaluation indicate 'that the radiation detectors were tested at 212*F for 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 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 monitored by two redundant radiation detector sensors. The detector s provide

~

output signals which initiate the automatic start of the Standby Gas Treatment System and secondary Containment Isolation when the radiation levels exceed 11 MR/HR.

-Duringnormaloperation,thegotalintegratedradiationexposure-forthe detectors will be only 3 X 10 rads which is well below the damage threshold 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

'm experienced as a result of the accident.

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

Failure to automatically start the SBGT system and isolate the secondary l

containment during a HELB wili not result in an off-site radiation dose in excess of the 10CFR100 limitations. The resultant radiation release is less L

than.a main steam-line break in the turbine building..

(~

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.

I Additionally, the detectors are periodically tested once every 18 months by L

l physi ally removing them from their mounting and performing a complete l

functional test..

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

l Therefore, continued operation is justified.

l i

TER NO.:

151

~

COMP 0NENT I.D. N0.:

RING AND TONGUE TERMINATION LUGS t'

MFG / MOD. NO:

AMP (NYLON INSULATION SLEEVE)

LOCATION:

DRYWELL TECHNICAL DISCUSSION:

The nylon insulated lugs are used to terminate Class 1E cabics inside the drywell at the Penetration Termination Boxes. Field inspections were made of these terminals to verify that the lugs were properly aligned and the insulation sleeves were physically separated between adjacent terminals. This spacing is sufficient to prevent shorting of adjacent conductors at the maximum voltage levels without taking credit for the insulating sleeves.

This analysis meets the criteria of 10CFR50.49, 7.ragraph (i)(5).

Therefore, continued operation is justified.

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.TER NO.:

156

COMPONENT-I.D. NO.:

SGT-FILT-2A-RB SGT-FILT-28-RB-MFG / MOD. NO.:-

FARR MODEL NUMBER D51423 LOCATION:

REACTOR BUILDING 50' TECMICAL 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 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 exc'essive radiation levels in the reactor coolant when long term recovery from the event is underway. Thus, there is no need for the SBGT system to maintain a negative pressure in the reactor building during 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

'(naximum 104*F) to equilibrium pt 133*F in approximately 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months . The total

integrated radiation dose is 103 rads for the 40 year life plus the accident.

Qualification documentation was obtained for the SBGT system and analyzed _ per

' DOR Gu delines. The testing was performed on identical and/or similar

- componi its (

Reference:

Farr Test Report No. L-71167). For those safety-related components not tested specifically by The Farr Company, supplemental

qualification data was obtained and analyzed. These components include:

' 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 B insulation system which has been successfully tested tg a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , 212*F peak temperature, 100% relative humidity and 5.5 x 100 rads gamma. This testing envelops the BSEP postulated accident transient and through analysis, the post-accident period.

.. ~. -.. _.,

s-TER No. 156 Page.2

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.

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

Reference:

ANC0 letter for IEEE 323-1974 Qualified Components).

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

Therefore, continued operation is justified.

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,,,p,--,-

4 TER NO.:

-164 C0hPONENT I.D. NO.:

RAYCHEM CONTROL CABLE MFG / MOD. NO.:

RAYCHEM/FLAMTROL CABLE LOCATION:

DRYWELL, REACTOR BUILDING

' TECHNICAL DISCUSSION:

The cable discussed in TER 164 is Raychem/Flamtrol, unshielded multiconductor cable rated at 1000 volts having a combined insulation thickness of 120 mils or greater. This cable type was subjected to testing in a program submitted for NRC review in letters dated October 20, 1983 and May 16, 1983. The cable was tested satisfactorily during the period June to July 1984 The final reports for.that testing program have not yet been completed.

Upon receipt and successful analysis of this final report, this cable type will be considered fully qualified f or this application.

This analysis meets the criteria of 10 CFR 50.49, paragraph (1) (2).

Therefore, continued operation is justified.

r (655JSD/pgp)

i

.t TER NO.: '

169 COMPONENT I.D. NO.:

NONE MFG /M00. NO.:

PYLE NATIONAL MODEL NS2 CONNECTOR LOCATION:

RX 107' TECHNICAL DISCUSSION:

Component materials of the PYLE National connectors have been identified and qualification documentation located. The qualification data has been

. evaluated per D0R guidelines and by applying Arrhenius techniques. This evaluation qualifies the connectors to the postulated normal and accident conditions at BSEP for 40 years (

Reference:

PYLE National Report No. TRC-01637-QL).-

The above items have been removed from the list titled " Items to be deferred

'due.to qualified replacements not available."

Therefore, continued.cperation is justified.

3

1

.,V TER N0.:

172 COMPONENT I.D. NO.:

SKV TERMINATIONS MFG / MOD. N0.:

BURNDY ELECTRICAL LUGS INSULATED WITH OKONEX BUTYL RUBBER TAPE AND OK0 TITE NO. 35 JACKETING TAPE LOCATION:

REACTOR BUILDING TECHNICAL DISCUSSION:

Test data has bGen located on a similar splice system that justifies the continued use of the SKV sp. lice system at BSEP (

Reference:

Okonite Report NQRN-3).

. The Burndy electrical lug is an uninsulated, all metal terminal lug used as the SKV Class 1E cable terminations and is, therefore, insensitive to thermal and radiation degradation.

Of the insulation materials used in the SKY terminations at BSEP only the Okonex tape was not tested. However, an Arrhenius calculation performed shows an expected life of 330 years at the maaimum reactor building temperature of 104'F.

The postulated accident temperature will peak at 288'F 70 seconds after accident initiation, then decline below the U.L. temperature rating of tne 0konex at 300 seconds. Although the accident peak exceeds the rating of the material, no significant degradation will occur during the short period of

. exposure. This is based on time temperature testing of the material which shows that butyl rubber can withstand 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months at 290'F prior to significant loss in properties.

Additionally, a radiation analysis performed on the butyl rubber shows that 7

less than 25% lost of elongation occurs after exposure to 1 X 10 rads gamma

(

Reference:

REIC No. 21). This demonstrates minimum degradation at the BSEP requirement.

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

Therefore, continued operation is justified.

._....u_

J

i TER NO.

179, 181 COMPONENT I.D. N0.:

MFG /M00. NO.:

GENERAL ELECTRIC EB-25, CR-151 LOCATION:

REACTOR BUILDING - AB0VE 20', RHR ROOM 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 00R 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 bui,lding 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 thr.t portion of the test program with conditionc at BSEP. The average leakage current per terminal block was less than 1 ma at 120VAC. The results of tnis test coupled with the facts that:

1.

All terminal blocks are in en 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 (1)(2).

Therefore, continued operation is justified.

l l

h-.

I

. s1 TER NO.:

180 COMPONENT I.D NO.:

TERMINAL BLOCKS MFG /M00. NO.:

G. E. EB-5 LOCATION:

DRYWELL TECHNICAL DISCUSSION:

EB-5 terminal blocks are used inside the drywell as terminal points for 120V/250V/480V Class 1E control and power circuits only. The teiminal blocks are mounted in Nema 4 enclosures and are not subject to direct steam or water impingement.

~

Various industry reports indicate that only low voltage signal circuits might be in jeopardy during a DBA. Limitorque Report No. B0119 supports EB-5 L

terminal block qualification for the DBA at BSEP. Upon receipt and successful analysis of this report, these terminal blocks will be considered fully qualified for this application.

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

Therefore, continued operation is justified.

O 9

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TER NO.:

182 COMPONENT I.D. NO.:

TERMINAL BLOCKS MFG / MOD. NO.:

CURTIS TYPE "L"

~

LOCATION:

DRYWELL 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 Westinghouse Research Memo No. 76-1CC-QUAEQ-M24 entitled, " Radiation Hardness of Terminal Blocks", did result in the success of at least four types of 4 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 impingement of steam or water. This configuration is similar to the test configuration.

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

Therefore, continued operation is justified.

1 i

o*

0 -

TER N0.:

NONE COMPONENT I.D. NO.:

821-FT-4157 B21-FT-4163 821-FT-4158 B21-FT-4164 B21-FT-4159 B21-FT-4165 B21-FT-4160 B21-F T-4166 B21-FT-4161

'B21-FT-4167 B21-FT-4162 MFG / MOD. NO.:

NOT INTERNATIONAL 78IN/S ACCELEROMETER LOCATION:

DRYWELL EL. 38' TECHNICAL DISCUSSION:

NDT Iternational accelerometers, Model No. 781N/S, are qualified on the basis of-similiarity with the NDT International accelerometer, Model No. 838-1, (Reference Wyle. Qualification Report No. 45633-1). Model 838-1 was fully qualified to meet or exceed all BSEP service conditions inside the drywell.

Similiarity Model No. 78IN/S and 838-1 are similar. The only difference is in the interface connection of the cable with the accelerometer.

Should the interface connection 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 (1)(2).

Therefore, continued operation is justified.

i W

6 e

.w r

TER NO.:

NONE COMPONENT I.D. NO.:

E51-0002-H MFG / MOD. NO.:

SQUARE D 9038-AG1-54 FLOAT SWITCH LOCATION:

RHR ROOM TECHNICAL DISCUSSION:

This item is'part of the RCIC turbine assembly.

It must maintain its electical integrity -for 30 minutes during the BSEP postulated accident.

Testing has been successfully performed on a HPCI turbine that contained this component (

Reference:

Wyle Labs / Terry Turbine Report No. 20458, R14 80). The testing performed envelops the BSEP radiation requirement but does not envelop the accident peak temperature.

The maximum postulated accidant temperature is 198*F. Based on an analysis of the nonmetallic materials contained in the switch, electrical integrity should not be compromised since exposure to the accident peak will not result in any significant degradation.

, This analysis meets the criteria of 10CFR50.49, paragraph (i)(2) and (i)(5).

Therefore, continued operation is justified.

4

8 E

ai n

TER NO.:

NONE COMPONENT I.D. NO.:

C12-F010-L C12-F011-L E51-C002-LS4 MFG / MOD. NO.:

NAMC0 D1200G LI;ilT SWITCH LOCATION:

REACTOR BUILDING 50', RHR ROOM TECHNICAL DISCUSSION:

Component materials of the Namco D1200G limit switch have been identified and qualification documentation on similar equipment located. The qualification data has been evaluated per D0R guidelines and by applying Arrhenius techniques. Results of this evalua~1on indicates that the nonmetallic components have greater than 9 X 10$yearsatthemaximumreactorbuilding temperature 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 shows that the radiation threshold for each nonmetallic gaterial is greater than the maximum postulated total integrated dose of 1 X 10' rads gamma.

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

Therefore,. continued operation is justified.

21 1

b

.,f a

TER N0.:

NONE COMPONENT I.D. NO.:

B32-CS-F019 B32-CS-F020

. MFG / MOD. NO.:

SENTRY MODEL F3N1R1 SWITCH LOCATION:

REACTOR BUILDING EL. 20' TECHNICAL DISCUSSION:

The Sentry F3N1h1 switch utilizes a Series 2 Honeywell Microswitch as the internal 'switchi.1g 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 envelops the BSEP accident duration but does not envelop the 70 second BSEP peak temperature transient of 200'F. A.mnterial analysis indicates that the switch will not be significantly degraded by the short exposure to the postulated accident peak.

7 Rads (

Reference:

Additionally, the switch h~as been tested to 1 X 10 Hogeywell Report No. 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.

m...

~. _ _

l r

7 TER N0.:

NONE COMPONENT I.D. NO.:

NP6-M0T-M1, M2 NP7-M0T-M1, M2 1B-RX 1A-RX MFG / MOD. NO.:

DOERR MOTORS AND ITE CONTROL PANELS LOCATION:

REACTOR BUILDING EL. 20' TECHNICAL DISCUSSION:

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

1.

Main steam isolation valves 2.

Scram valves-3.

Scram volume vent and drain valves 4.

Safety relief valves 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 failure. 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 emergency air compressors 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 (1)(2).

Therefore, continued operation is justified.

ML20098E305

Text

Dear Mr. Vassallo:

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