ML20072A377

Reaffirms Justifications for Continued Operation for Equipment Identified as Categories Ib,Iia & IV for Which Qualification Programs Still in Progress.Qualification Info Not Accurate Representation.Addl Items Encl

ML20072A377
Person / Time
Site:	FitzPatrick
Issue date:	06/06/1983
From:	Bayne J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	Vassallo D Office of Nuclear Reactor Regulation
References
JPN-83-52, NUDOCS 8306100100
Download: ML20072A377 (37)

Text

r 123 Main Street White Plain' NewYork 10601 914 681.6240 3""

6' 1983

' ""'"'a a v"-

A NewYo.rkPbwer JPN-83-52 Executive VK.e President

& Authority "eciee, oeme, <e, Director of Nuclear Reactor Regulation U.S.

Nuclear Regulatory Commission Washington, DC 20555 Attention:

Mr. Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing

Subject:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Safety Evaluation-Environmental Qualification of Safety-Related Electrical Equipment

Reference:

1.

NRC letter, D.B. Vassallo to J.P.

Bayne, dated April 19, 1983.

2.

PASNY letter, J.P.

Bayne to D.B. Vassallo, dated May 20, 1983 (JPN-83-45).

Dear Mr. Vassallo:

The purpose of this letter is to reaffirm the Authority's justifications for continued operation (JCO's) for equipment identified by Reference (1) as NRC categories Ib, iia, and IV for which qualification programs are still in progress, and to submit information for equipment items for which JCO's were not previously submitted.

The Authority has reviewed the Technical Evaluation Report (TER) which was transmitted with your Safety Evaluation (Reference 1) and has concluded that the qualification information which was reviewed by the NRC and its contractor is not an accurate representation of the present level of qualification of safety-related electrical equipment installed in the FitzPatrick plant.

Since the original submittal by the Authority, an ambitious ongoing program of qualification verification and equipment replacement has resulted in a much higher level of qualification than was reviewed by the NRC.

This is further detailed in the Authority's response to 10 CFR 50.49 (Reference 2).

ft i i 8306100100 830606 PDR ADOCK 05000333 P

PDR

The Authority reaffirms previously submitted justifications for continued operation based on Enclosure 1 which summarizes the qualification status,to the best knowledge of the Authority, of the safety-related electrical equipment encompassed by NRC categories lb, iia and IV.

A number of the original JCO's no longer exist based on the following:

a.

The equipment item has been the subject of a functional evaluation and on the basis of this evaluation is judged to not require qualification in accordance with 10 CFR 50.49.

b.

The equipment item has been replaced with fully qualified equipment.

c.

A program has been completed which demonstrated environmental qualification for this equipment item.

d.

The equipment item has been relocated to a mild environment.

The Authority, in reviewing its list of equipment, has identified additional equipment items which have been added to the listing, equipment items for which the qualification status has changed, or for which justification for continued operation was not previously submitted.

A list of these additional items is provided in Enclosure (2).

Justification for continued operation for these items is included in Enclosure (3).

In Section 5.0 of the TER, concerns are expressed that some equipment listed in the NRC SER had been omitted from the Authority's response without explanation.

The Authority has examined all the items in Appendix B of Section 5.0 of the TER and confirms that the items do not require qualfication in accordance with 10 CFR 50.49 or that these items are now included in the list of equipment requiring qualification.

JCO's for these items are also provided in Enclosure (3) if they were not previously submitted and the qualification program is not complete.

Very truly yours, f

m J

. Bayne xecutive Vice President Nuclear Generation cc; Mr. J.

Linville Resident Inspector U.S.

Nuclear Regulatory Commission P.O.

Box 136

Lycoming, N.Y.

13093

ENCLOSURE (1) p 9

. JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT STATUS OF COMPONENT JCO'S JAF F.R.C.

EQ FILE JC0 (CAT.) ITEM SEC/PAGE COMPONENT ID PROVIDED COMMENTS (II.a)1 3/29, 30 10110V-25A, B Not Req.'

Note 4 (I.b)2 3/36 10MOV-31B.

'Not, Req.

Note 4,

(I.b)5 12/16 23MOV-19 Not. Req.

Note 1 (I.b)6 7/14 14MOV-llB Not Req.

Note 4 (II.a)7 7/11, 13 14MOV-llA, 12A Not Req.

Note 4 (I.b)8 8/22 20MOV-94 Yes Note 2 (II.a)l0 12/21 23110V-5.8 No't Req.

Note 4 (II.a)12 8/24 29110V-77 Not Req.

Note 4 (II.a)l3 3/52 10MOV-67 Not Req.

Note 4 (II.a)15 12/14 23MOV-17 Not Req.

Note 1 (I.b)l7 23/12 02MOV-54A, B Not Req.

Note 4 (I.b)18 8/21 20liOV-82 Yes Note 2 (II. a)19 23/11 02110V-53A, B Not Req.

Note 4 (II.a)20 3/26 10liOV-18 Not Req.

Note 4 (II.a)23 12/20 23110V-57 Not Req.

Note 4 (I.b)25 9/15

' 27PNS-101A, B Not Req.

Note 3 (I.b)27 9/25 27PNS-132A, Br Not Req.

Note 3 (I.b)28 8/29-33 29PNS-86A-D Not Req.

Note 3 (I.b)29 10/12, 13 66PNS-100A1,A2,B1 Not Req.

Note 3 j

32 1

66PNS-101A1, A2, B1 I

B2 l

(II.a)30-12/31 23PNS-18 Not Req.

Note 1 (II.a)31 12/30 23PNS-LS4 Not Req.

Note 1 (II.a)32 12/29 23PNS-LS3 Not Req.

Note 1

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ENCLOSURE (1)

Page 2

of 9 e

s JA>1ES A. FITZPATRICK h*UCLEAR POWER PLANT ENVIRON!! ENTAL QUALIFICATION OF SAFETY-RELATED EQUIPlIENT STATUS OF COMPONENT JCO'S i

JAF F.R.C.

EQ FILE JC0 (CAT.) ITEM SEC/PAGE COMPONENT ID PROVIDED COS0!ENTS l

(I.b)33 9/33 27PNS-147

'Not Req.

Note 3

..(II. a) 35 8/4, 5 2-2PN,S-40

. Not Req.

Note'1 (II.a)36 8/4, 5

~2-2PHS-39 Not Req.

Note'1 (I.b)37 9/16 27PNS-103A, B

'Not Req.

Note 3 (I.b)38 10/14 71 ACA3 Not Req.

Note 1 (I.b)39 10/16 71PT-71 ACA3 Not Req.

Note 1 (I.b)42 9/14A,-B 27P-llA Yes Note 2 (I.b)43 9/1 27DWA-PA Yes Note 2 (I.b)44 9/8A, 8B 27HAZ-101A, B Yes Note 2 (I.b)45 9/62 270 AZ-101A Yes Note 2 2

(I.b)46 10/1,. 2 17RE-430A, B Not Req.

Note 1 l

(II.a)47 15/7-11 X-100B, X-103A, B Not Req.

Note 4 X-104D (II.a)47 15/7-11 X-109, 110D Not Req.

Note 4 (I,b)48 03/65-68 10P-3A-D Not Req.

Note 4 (I.b)49 7/17 14P-1A, B Not Req.

Note 4 (IV)53 13/2 66UC-22A-K Yes Note 5 (I.b)54 3/91, 93, 95, 10PS-120B, D, F, H Not Req.

Note 4 97 (I.b)S5 3/90, 92, 94, 10PS-120A, C, E, G Not Req.

Note 4 96 (I.b)S6 4/7 02-3PT-178C, D Not Req.

Note 4 (II.a)58 23/1, 3 06PT-53A, C Not Req.

Note 1 (II.a)59 23/2 06PT-53B Not Req.

Note 1 (I.b)60 03/120 02-3PS-52A Not Req.

Note 4 (I.b)61 3/118 02PS-128A Not Req.

Note 4

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l ENCLOSURE (1)

Page 3

of 9

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT STATUS OF COMPONENT JCO'S JAF F.R.C.

EQ FILE JC0 (CAT.) ITEM SEC/PAGE COMPONENT ID PROVIDED COMMENTS (I.b)62 3/120 02-3PS-52B Not Req.

Note 4

~

(I.b)63 3/120 02-3PS-52D

! Not Req.

Note 4 e

. c (I. b)'64~

3/118 02PS-128B Not Req.

Note 4 j

(I.b)65 7/21 14PT-38A, B Yes Note 2 (I.b)68 9/37 27PT-ll4

~Yes Note 2 (I.b)71 8/6, 7 02-2SOV-39 Not Req.

Note 3 (I.b)72 8/34, 35, 36, 29s0V-80A, B, C, D Yes Note 2 37 (I.b)73 10/8, 9 66SOV-100A, B Not Req.

Note 3 (I.b)75 8/6 02-2SOV-40 Not Req.

Note 3 (I.b)76 9/43 27SOV-101A, B Not Req.

Note 3 (I.b)77 9/44,- 45, 48, 27SOV-lll, 112, 115, Not Req.

Note 3 49 116 (I.b)78 9/50, 51' 27SOV-117, 118 Not Req.

Note 3 (I.b)79 9/59, 60 27SOV-131A, B Not Req.

Note 3

}

27SOV-132A, B (I.b)80 9/58, 61 27SOV-125A, B Not Req.

Note 4 t

27SOV-135A, B (I.b)81 8/38, 39, 40, 29SOV-86A-D -

Yes Note 2 41 (I.b)82 9/56 27SOV-120A, B Not Req.

Note 4 27SOV-121A, B 127SOV-122A, B 127SOV-123A, B i

(I.b)83 3/82, 84, 88, 10PS-101B, D, C Not Req.

Note 4 86, 83, 79, 77 10PS-119C, A 10PS-100C, A (I.b)84 3/78, 80, 87, 10PS-119B, D Not Req.

Note 4 89 10PS-100B, D

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ENCLOSURE (1)

Pag, 4 of 9

JA>1ES A. FITZPATRICK hTCLEAR POWER PLANT ENVIRON!! ENTAL QUALIFICATION OF SAFETY-RELATED EQUIP >1ENT STATUS OF C0!!PONENT JCO'S JAF F.R.C.

EQ FILE JC0 (CAT.) ITD1 SEC/PAGE C051PONENT ID PROVIDED COSNENTS (I.b)85 3'/81 10PS-101A Not Req.

Note 4

. (I.b)86 5/5 02-3LIS-72B, D

'Not Req.

Note 4 (I.b)87

'5/5 02-3LIS-72A, C Not Req.

Note 4 (I.b)88 5/6 02-3LIS-83B Not Req.

Note 4 (I.b)89 5/6 02-3LIS-83A Not Req.

Note 4 (I.b)90 3/119 02-3 LITS-73 Not Req.

Note 4 (I.b)91 8/11 02-3LIS-101C, D Not Req.

Note 4 (I.b)92 8/10 02-3LIS-101A, B Not Req.

Note 4 (I b)93 12/6, 7 23LS-91A, B Not Req.

Note 1 (I,b)94 12/9 23LT-201B Not Req.

Note 1

.(I.b)95 4/2 3LS-231A, B, C, D Yes Note 2 (I.b)96 12/5 23FT-82 Not Req.

Note 1 (I,b)97 7/5 14FT-40B Yes Note 2 (I.b)98 3/9 10FT-109A Yes Note 2 (I.b)99 9/6 27FT-103A, B Yes Note 2 (I.b)l00 7/5 14FT-40A Yes Note 2 (I.b)l01 3/10 10FT-109B Yes Note 2 (I.b)l02 3/7 10FT-97B Yes Note 2 (I,b)l03 3/6 10FT-97A Yes Note 2 (II.a)104 17/2 MCC (BMCC2)

Yes Note 2 i

(II. a)105 17/20

!MCC-165 Yes Note 2 (II.a)l0E 17/14, 15, 18 MCC-153, 155, 163 Yes Note 2 (II.a)107 17/19 MCC-164 Yes Note 2 (II.a)108 21/04 L-16 Yes Note 2 e

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ENCLOSURE (1)

Page 5

of 9

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT STATUS OF COMPONENT JCO'S JAF F.R.C.

EQ FILE JC0 (CAT.) ITEM SEC/PAGE COMPONENT ID PROVIDED COSNENTS (II.a)l0S 21/03 L-15 Yes Note 2 (I.b)lil 20/2, 3, 4, 7 EB-5

'Yes Note 2 (I.15)l12 8/3 02TS-121, 124 Not Req.

Note 4 (I.b)ll3 12/4 23FS-78 Not Req.

Note 1 (I.b)ll4 7/4 14FIS-45B Not! Req.

Note 4 (I.b)ll5 7/3 14FIS-45A Not Req.

Note 4 (I.b)116 23/4, 5, 6, 7 68TE-102, 103, 105 Yes Note 2 106 (I.b)ll7 9/42 27RTD-ll2 Yes Note 2 (I.b)118 9/41 27RTD-101A-D Not Req.

Note 2

(

(I.b)120 14/46 13TE-38A Not Req.

Note 4 (I.b)121 14/50, 53, 55 13TE-100B, 106A, C Not Req.

Note 4 1

(I.b)122 8/19 13TE-ll7A, C, E Not Req.

Note 4 (I.b)123 14/49, 52, 54 13TE-100A, D, 106C Not Req.

Note 4 (I.b)124 14/56 13TE-106D Not Req.

Note 4 (I.b)125 14/51 13TE-100C Not Req.

Note 4 (I b)126 12/65, 66, 68, 23TE-92A, B Not Req.

Note 4 69 23TE-114A, B-(I.b)128 7/19 14PS-44B Not Req.

Note 4 (I.b)129 7/18, 19 14PS-41A, B, 44A Not Req.

Note 4 (I.b)l32 14/1, 2

13DPIS-83, 84 Not Req.

Note 4 (I.b)133 8/l

!02DPIS-ll6A-D Not Req.

Note 4 02DPIS-ll7A - D 02DPIS-ll8A - D 02DPIS-119A - D (I.b)134 8/i4, 15 05PS-12A, B, C, D Not Req.

Note 4 I

ENCLOSURE (1)

Page 6

of 9

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT STATUS OF COMPONENT JCO'S JAF F.R.C.

EQ FILE JC0 (CAT.) ITEM SEC/PAGE COMPONENT ID PROVIDED COMMENTS (I.b)l35 12/46, 47, 48, 23PS-86A - D Not Req.

Note 2 49

~

(I.b)l37 12/.39, 40, 41, 23PS-68A - D Not Req.'

Note 4 42 (1.b)l38 9/34 27PS-110A, B Not Req Note 4 (II.a)l40 14/57 13TS-1 Not Req.

Note 1 (II.a)l42 16/10 RSS-6-104 Not Req.

Note 4 (II.a)143 16/12.

Eaton (RB)

Not Req.

Note 4 (II.a)l44 16/13 Eaton (Drywell)

Not Req.

Note 4 (II.a)145 16/14 Anaconda FR-EP Not Req.

Note 4 I

(II.a)l47 16/18 BIW Not Req.

Note 4 (II.a)l48 16/19 Vulkene Not Req.

Note 1 (I.b)150 21/6 Circuit Breaker Not Req.

Note 6 THQB 1120 (I.b)151 27/1 23LT-203Al, B1 Not Req.

Note 4 27PT-115A1, B1, A2 B2 06PT-61A, B 23LT-203A2, B2 (I.b)l52 27/2 23LT-202A, B Not Req.

Note 4 (I.b)153 28/1 27RE-104A, B Not Req.

Note 4 23HAZ-102A, B Yes Note 2 (I.b)155 30/2 g

(II a)l56 14/7

'13MOV-15 Not Req.

Note 4 (IV)l57 36/4

!02VliE-071A-L Yes Note 5 l

I (II.a)l61 35/2 jTefrel Not Req.

Note 4 (IV)164 36/2 OTMiY-71A-L Yes Note 5 (I.b)l65 9/1 27DWA-PB Yes Note 2 (I.b)l66 9/8 27HAZ-101B Yes Note 2

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i Page 7

of __9 JASIES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRON!! ENTAL QUALIFICATION OF SAFETY-RELATED EQUIPS 1ENT STATUS OF C0!!PONENT JCO'S JAF F.R.C.

EQ FILE JC0 (CAT.) ITE}!

SEC/PAGE COSIPONENT ID PROVIDED CO3DfENTS (I.b)l67 9/62 270 AZ-101B Yes Note 2 2

(I.b)l68 9/32, 33 27PNS-147, 146 Not Req.

Note 3 (I.b)l69 9/31

!27PNS-145 Not Req.

Note 3 (I.b)170 9/31 27PNS-145 Not Req.

Note 3 (I.b)l71 9/30 27PNS-144 Not Req.

Note 3 (I.b)l72 9/30 27PNS-144 Not Req.

Note 3 (I b)l73 9/29 27PNS-143 Not Req.

Note 3 (I.b)l74 9/29 27PNS-143 Not Req.

Note 3 (I.b)l75 9/28 27PNS-142 Not Req.

Note 3 I

(I.h)l76 9/28 27PNS-142 Not Req.

Note 3 (I.b)177 9/27 27PNS-141 Not Req.

Note 3 (I.b)l78 9/27 27PNS-141 Not Req.

Note 3 (I.b)l79 3/35 10MOV-031A Not Req.

Note 4 (I.b)180 9/17, 18, 21, 27PNS-lli, 112, 11.5, Not Req.

Note 3 l

22 116 (I.b)181 9/23, 24 27PNS-ll7, 118 Not Req.

Note 3 (I.b)182 9/19, 20 27PNS-ll3, 114 Not Req.

Note 3 (I.b)183 3/12.0C 02-3PS-52C Not Req.

Note 4 (I.b)188 12/3 23DPIS-77 Not Req.

Note 4 (I.b)189 12/2

? 23DPIS-76 Not Req.

Note 4 (II.a)190 17/1, 3 i BMCC-3, 1 Yes Note 2 i

(I.b)191 8/8

'02-3LIS-57A, B Not Req.

Note 4

-(I.b)192 8/9

'02-3LIS-58A, B Not Req.

Note 4 (I.b)193 9/52, 57 27SOV-124A, B Not Req.

Note 4 27SOV-119A, B

I ENCLOSURE (1)

Page __8 of 9

JA31ES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRON!! ENTAL QUALIFICATION OF SAFETY-RELATED EQUIP}1ENT STATUS OF CO>fPONENT JCO'S JAF F.R.C.

EQ FILE JC0 (CAT.) ITEli SEC/PAGE COSIPONENT ID PROVIDED' C050!ENTS (I.b)194 9/46, 47 27SOV-113, 114 Not Req.

Note 3 (I.b)195 10/10, 11 66SOV-101A, B

'Not Req.,

Note 3 (I.b)197 10/17 71PT-71AC B3 Not Req.

Note 1

~

(I.b)198 10/15 71AC B3 Not Req.

Note 1 (I.b)199 9/14

,27P-llB Yes Note 2 l

(IV)200 36/3 Endevco Yes Note 5 l

L I,

I l

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Page 9

of 9

ENCLOSURE (1)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPliENT NOTES:

1.

This equipment item has been the subject of a functional evaluation, and on the basis of this evaluation been found not.to require qualification in accordance with 10 CFR 50.49.

2.

Justification for continued operation (JCO) for this equip-ment item hds been provided previously.

3.

This equipment item has been repisced with fully qualified equipment.

4.

A program has been completed which demonstrated environmental qualification of this equipment item.

5.

Justification for continued operation (JCO) for this equip-ment item is provided in Enclosure (3).

6.

This equipment item has been relocated to a mild environ-ment area.

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Pags 1

of 3

ENCLOSURE (2)

JAMES A. FITZPATRICK NUCLEAR. POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPl!ENT Equipment Requiring JCO's JAF E/Q FILE JC0 ENCL.(3)

COMPONENT I. D. -

SEC/PAGE PAGE NO.

t g

f 01-125FN-1A,B 6/7 1

01-125E-5A,B 6/10 2

02PS-134A-D 8/2 3

02VMY-71A-L 36/2 4

02VME-71A-L 36/3,4 4

02SOV-17 8/4,7 5

02SOV-18 8/4,8 5

02-3AU-278 4/8 6

02-3 LITS-73 3/119 7

02-3PS-102A-D 8/45,46 8

02-3PT-178A-D 4/7 9

03SOV-31A,B 23/10 10 03SOV-ll7 23/8 10 i

03SOV-ll8 23/9 10 1

03SOV-140A,B 4/11 10 10DPIS-125A,B 3/115 11 10MOV-12A,B 3/13,14 12 10MOV-20 3/27 13 r

10MOV-32 3/37 14 10MOV-89A,B 3/55,56 12 10MOV-148A 3/57 12 10MOV-149A 3/59 12

(

12MOV-15 8/16 15 i

13FS-57 14/3 16

- i 13LS-12 14/5 21 1

13MOV-16 14/8 17 i

13MOV-39 14/14 20 13MOV-41 14/15 20 13MOV-131 14/17 20 13PNS-L34 22 13PS-67A,B 14/30,31 18 13PS-72A,B 14/32 18 13PS-78A-D 14/33 19 I

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Page 2

of 3

ENCLOSURE (2)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT Equipment Requiring JCO's L

JAF E/Q FILE JC0 ENCL.(3)

COMPONENT I.D.

SEC/PAGE PAGE NO.

15MOV-103 11/3 23 16-lPNS-101A,B 31/3 24 16-lPNS-102A,B 31/4 24 16-lSOV-101A,B 31/5 25 16-lSOV-102A,B 31/6 25 16-lRTD-107 26 16-lRTD-108 26 27

{

17RE-50A,B 17RE-53A,B 26/1 27 17RT-53A,B 26/2 27 17RE-431,-432 27 17RE-434A,B 26/3 27 17RT-434A,B 26/4 27 17RE-458A,B 27 17RE-463A,B 26/5 27 17RT-463A,B 26/6 27 23SOV-83 8/42 28 20SOV-95 8/43 28 20PNS-83 8/44 29 20PNS-95 8/44 29 23FS-78 12/4 30 23MOV-14 12/11 31 23LS-99 32 23LS-100 32 23PNS-LS4 12/30 33 23PS-84A,B 12/43,44 34 23PS-97A,B 12/50,51 34 27E/P-103A,B 9/2 35 (27FCV-103A,B) 27PS-110A,B 9/34 36 27DWA-HTA,-HTB 9/67 37 27RTD-102Al,B1 9/64 37

-102A2,B2 9/65 37 27RTD-107Al,B1 9/66 37

-107A2,B2 j

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y Page 3

of 3

ENCLOSURE (2)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Y

EUVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT 1

Equipment Requiring JCO's JAF E/Q FILE JC0 ENCL.(3)

.l COMPONENT I.D.'

SEC/PAGE PAGE NO.

l 29MOV-74 8/23 38 J

29MOV-77 8/24 39 29MOV-200A,B 8/45 41

-201A,B 8/46 40

-202A,B 8/47 40 l

-203A,B 8/48 40

-204A,B 8/49 40 29PS-201A,B 8/49 41

-202A,B 8/50 41 66HV-3A,B 10/5,6 42 66UC-22A-K 13/2 43 68TE-201 thru -212 44 68TE-301 thru -310 44 71 BAT-3A,B 45 71 INV-3A,B 45 71ACA5 46.

71ACB5

~46 PT-71ACA5 46 PT-71ACBS 46 CINCH-Jones TB 20/2 47 GE EB-5,-25 Steel 20/1,3,4,5,6,7 48 JB JB-X101E 5/15 49 Misc. JCT Boxes 50 I

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Page 1 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION:

6 PAGE:

7 SY S TUM':

STANDBY GAS TREATHENT (01-125')

COMPONENT I.D.:

01-125FN-1A, 8 FAN MOTOR The Standby Gas Treatment (SBGTS) fan motor is required to function following a postulated LOCA and HELB.

The exposure to the postulated accident environment requires the motor to operate in an elevated radiation environment originating from SBGTS effluent flow following a postulated LOCA only.

No significant temperature, pressure or humidity increases are postulated for the fan motor's. location remote from the area experiencing a direct harsh environment due to these events. A preliminary.

assessment, based on the testing of motors with the same class of insulation manufactured in the same time frame, reveals that no detrimental effects to the motor's operability should occur at the postulated accident integrated radiation dose of 3.8 x 106 R

for 180 days of continuous Standby Gas Treatment System flow.

In accordance with the information supplied above concerning the SBGTS fan motor, continued operation is considered justified.

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Pega 2

.g ENCLOSURE (3) j JAMES A. FITZPATRICK NUCLEAR POWER PLANT a

ENVIRONMENTAL QUALIFICATION OF SAFETYfRELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION: 6 PAGE:T S,YSTEM:

STANDBY GAS TREATMENT (01-125)

COMPONENT I.D.:

01-125E-5A, B FILTER TRAIN ELECTRIC PREHEATER AND CONTROL UNIT

(

l-The Standby Gas Treatmerit (SBGTS) Electric Pre-Heater and Control Unit is required to function fol_ lowing a postulated LOCA and HELB.

The exposure to,the postulated L0CA environments requires the 6 eater unit to operate in an integrated radiation environdent of f.8 x 106 R for 180 days of'continuou's, standby gas treatment system flow.

No significant temperature, p,ressure or humidity increases are postulated for the heater unit's location due to the LOCA event.- The specific components involved include an electrical contactor, transformer, terminal boards, a'nd wiring.

The performanca'of similar design components in the postulated radiation exposure has been confirmed in previous type tests for i

Motor Control Centers.

The components listed above are self-contained, independent, and redundant units, each' capable of performing the intended function by itself.

Currently both components are operational.

t u

In#accordance with the information supplied above concernin'g the SBGTS heater units, continued operation is considered justified.

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Page 3 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION:

8 PAGE:

2 SYSTEM:

02PS-134 (A-D)

COMPONENT 1.D.: MAIN STEAM LINE TURBINE INLET PRESSURE SWITCH (PCIS)

BARKSOALE W -M12FJ These pressure switches are designed to provide main steam line isolation in case of a main' turbine pressure regulator failure, which could cause turbine control valves or bypass valves to open fully.

These pressure switches are located in the Turbine Building and will not experience a postulated harsh environment when performing their intended design function since they instantaneously detect pressure regulator failures.

Therefore, continued operation is considered justified.

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Paga 4 ENCLOSURE (3)

JAMES A. FITZPATRICK NOCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT.

JUSTIFICATION FOR CONTINUED OPERATION t

SECT' ION: 36

PAGE

YIT7 4 SYSTEM:

NUCLEAR BOILER (02) l COMPONENT I.D.: 02VMY-71A-L SRV ACOUSTICAL MONITORING 02VME-71A-L SYSTEM The Safety Relief Valve (SRV) Acoustical Monitoring System was installed in January 1981 in response to NUREG-0578.

Direction provided to the Authority required immediate installation of equipment to be followed by a qualification program.

A complete valve monitoring system is currently undergoing an environmental qualification test program under the sponsorship of a group of utilities.

When the final test report becomes available, the Authority will plan a modification program, if required, to bring the installed system into the tested configuration.

The part of the system which is located in a harsh environment is fully redundant.

Each SRV is equipped with two acoustical sensors and associated preamplifiers.

If one sensor channel were to fail, the other sensor channel can be connected to the system cabinet in the Relay Room.

In addition, each SRV l

discharge line is also equipped with temperature sensors which j

indicate an open SRV.

Temperature readouts and alarms are l

provided in the Control Room.

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Paga 5 ENCLOSURE (3) q JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT i

JUSTIFICATION FOR CONTINUED OPERATION q

l SECTION: 8 PAGE:47""48 SYSTEM:

NUCLEAR BOILER (02) l COMPONENT I.D.: 02SOV-17, -18 REACTOR HEAD VENT VALVE PILOT SOLENDID, ASCO NP-8316 Full qualification depends only on field inspection of the existing installation during the upcoming outage.

Identical ASCO NP Series solenoid valves have undergone extensive type testing in LOCA environments and are extensively utilized in Class 1E application in most nuclear power plants.

The required post-accident operating time for this valye (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) is much shorter than the length of the type test (30 days).

Based on these facts, continued operation is considered justified pending final qualification.

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Page 6 ENCLOSURE (3)

P JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION: 4 PAGE: 8 E

SYSTEM:

NUCLEAR BOILER VESSEL INSTRUMENT (02-3)

COMPONENT I.D.:

02-3AU-27B (A-D) - REACTOR HIGH PRESSURE ANALOG TRIP UNIT (RPS)

ROSEMOUNT 510 00 Following postulated Reactor Building HELB's, it is unlikely that reactor trip would be required based on high reactor pressure.

If required for this postulated accident, this unit would perform its intended function of providing a trip signal to the normally energized (fail-safe) protection logic.

Redundant trip l

units are provided and are located at different instrument racks.

l Based on this information, continued operation is considered justified pending relocation of these trip units to a mild environment.

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Pogs 7 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION-OF SAFETY-RELATED EQUIPMENT

. JUSTIFICATION FOR CONTINUED OPERATION SECTION:

3 PAGE: 119 SYSTEM:

NUCLEAR BOILER VESSEL INSTRUMENT (02-3)

COMPONENT I.D.: 02-3 LITS-73 REACTOR WATER LEVEL INDICATING SWITCH YARWAY 4418EC This unit provides a reactor water level. permissive signal to the RHR System and provides indication of water level to a Control Room indicator.

This unit is qualified to perform its short term function of providing reactor water level permissive signal to the RHR System.

However, test data does not cupport full qualification for providing reactor water level indication over an extended post-accident time frame after a postulated HELB in the Reactor Building.

Type test data of identical equipment at elevated temperatures and humidity in conjunction with a radiation threshold analysis of internal components supports qualification for its Reactor Building location following postulated LOCA's inside containment.

Type test data supports short term operation following postulated Reactor Building HELB's.

This is considered acceptable based on other water level indications available in the Control Room and the ability to access the Reactor Building for repairs within a short time frame following a postulated Reactor Building HELB.

Based on the above, interim operation is considered justified pending replacement.

i Page 8 ENCLOSURE (3) t JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENIAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION:

8 PAGE:45, 46 I

SYSTEM:

NUCLEAR BOILER VESSEL INSTRUMENT (02-3) l COMPONENT I.D.:

02-3PS-102 (A-D) REACTOR PRESSURE SWITCH l

(RECIRC)

These pressure switches are required to function,short term following postulated LOCA/HELB's to trip the reactor recirculation pump motors when reactor high pressure is sensed.

Post-LOCA, the environmental conditions at the switch location will not significantly change during this short time frame, since these switches are located outside the primary containment.

For HELB, redundant switches are located at separate instrument racks which experience lesser environments.

Most importantly, available type test data for switches by the same manufacturer and of a similar design indicate proper pressure switch operation at equivalent temperatures over an extended time period.

A preliminary assessment, based on type testing of these switches, reveals that no detrimental effects to the operability of the I

switches should occur during the postulated DBE.

Based on the short term operability requirement along with the type testing successfully performed on similar switches, continued operation is considered justified.

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Page 9 i

ENCLOSURE (3) l JAMES A. FIIZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT f

[

JUSTIFICATION FOR CONTINUED OPERATION' SECTION: 4 PAGE: 7 SYSTEM:

NUCLEAR BOILER VESSEL INSTRUMENT (02-3)

COMPONENT I.D.:

02-3PT-178 (A-D) REACTOR HIGH PRESSURE TRANSMITTER (RPS)

ROSEMOUNT 1151GP This transmitter provides a trip signal on high reactor pressure to the Reactor Protection System.

The Rosemount 1151GP transmitter has been fully type tested to IEEE 323-1971 for harsh environment parameters of pressure, radiation, temperature, and humidity at more severe levels than experienced in the specific JAF locations.

The only outstanding qualification issue is aging of the transmitter's electronics.

This concern can be partially resolved due to the periodic. surveillance testing which is performed to verify transmitter operation and calibration.

The function of this component is performed in the initial phases of postulated design basis accidents.

It will experience a harsh environment following postulated Reactor Building HELB's.

The type testing noted above provides a high degree of confidence l

that this unit will perform its intended function.

In addition l

to redundant sensors located at different locations, there is also diverse instrumentation which can initiate reactor trip for this postulated event.

Based on this information, continued operation is considered justified pending replacement of this item.

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I Pegs 10 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT I

ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR-CONTINUED OPERATION SECTION: 23, 4 PAGE: 8,9,11 SYSTEM:

CONTROL ROD DRIVE (CRD) (03)

COMPONENT I.D.: 03SOV-117; -118; -140A, B; -31A, 8 SCRAM AIR PILOT SOLENOID VALVES l

The solenoid valve performs its safety related function immediately upon initiation of the postulated accident.

In addition, the solenoid valve is required to de-energize in the performance of its safety related function.,

i.e.,

the valve is

" fall safe".

Based on the completion of the solenoid valve's safety related function prior to significant exposure to the accident environment, failure of this device would not cause degradation of any safety function.

An ongoing qualification program has identified type test data for identical solenoid valves for high temperature and humidity conditions.

Following postulated LOCA's inside primary containment, these valves will not experience any change in its environment in the short period required to perform its design function.

Until finalization of the qualification program for these valves, continued operation is considered justified.

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Page 11 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION:

3 PAGE:115 SYSTEM:

RESIDUAL HEAT REMOVAL (RHR)

+

COMPONENT I.D.: 10DPIS-125A, B RHR LOW FLOW DIFFERENTIAL PRESSURE SWITCH BARTON 289 t

The ongoing qualification program for this component has identified type test data directly applicable for this component.

Temperature, humidity, and radiation test data surpasses the postulated post-HELB environmental parameters which l

this component will experience in its limited operating time (6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />).

Post-HELB this component is only required to maintain i

electrical integrity and does not have to function.

Post-LOCA the only harsh environment parameter is radiation.

A radiation threshold analysis and test data indicates that no materials will experience radiation damage at the 5x105 rads t

integrated dose.

The expected life at 104*F for the limiting material relative to thermal aging is greater than the life of i

the plant.

Based on this data, continued operation is considered justified pending final qualification.

i Page 12 h

I ENCLOSURE (3) l JAMES A. FITZPATRICK NUCLEAR POWER PLANT I

ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT I

JUSTIFICATION FOR CONTINUED OPERATION I

SECTION:

3 PAGE:TT7 14, 55, 56, 57, 59 SYSTEM:

RESIDUAL HEAT REMOVAL (10)

[

COMPONENT I.D.:

10MOV-12A, B RHR HEAT EXCH. OUTLET ISOL.

l VALVE OPERATOR 10MOV-89A, B HEAT EXCH. SERV.

W. DISCH.

ISOL. VALVE OPERATOR

(

10MOV-148A SERV.

W.

CROSSTIE ISOL.

l VALVE OPERATOR 10MOV-149A SERV.

W. CROSSTIE ISOL.

{

VALVE OPERATOR t

These valves are required to operate for the residual heat removal (RHR) system. The valve operat ars are fully

+

environmentally qualified for this environment for all parameters except for a post-HELB ten (10) minute temperature transient above 250*F in which the environment exceeds type test conditions by-30*F for 10 minutes.

The probability that the valves will have to operate during this limited time period is very small and the valve stroke time is short.

1 Based on these considerations, the continued operation of the plant is considered justified pending replacement of the operator l

motors.

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Page 13 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION: 03

.PAGE: lDF-SYSTEM:

RESIDUAL HEAT REMOVAL COMPONENT-I.D.: 10MOV-20 RHR CROSSTIE VALVE OPERATOR LIMITORQUE SMB-1-40 t

This valve motor actuator is electrically disconnected with the exception of its internal position switches.

The position switches are identical to those in other fully qualified,

Limitorque valve operators.. Completion of the qualification for this item requires an administrative change to an existing qualification program to include this valve operator.

Continued operation is, therefore, considered justified pending completion of this program.

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Paga 14 ENCLOSURE (3).

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION:

3 I

PAGE: 37 l

SYSTEM:

RESIDUAL HEAT REMOVAL (10)

COMPONENT I.D.: 10MOV-32, CONTAINMENT ISOLATION VALVE The actuator performs its function during the initial' seconds of a postulated accident to provide containment isolation, simultaneously with any increase in containment temperature, pressure, or radiation level.

Qualification data for similar Limitorque actuators indicate a thermal lag before ambient temp 3ratures affect internal temperatures and acceptable performance in these environments.

The isolation function for this valve is also performed by a redundant qualified motor operated valve located outside the containment and which does not experience the accident environment during the isolation period.

For these reasons, continued operation with this valve is considered justified pending motor replacement.

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Pcgs 15 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION:

8 PAGE: 16 SYSTEM:

REACTOR WATER CLEANUP (RWCU)

COMPONENT I.D.: 12MOV RWCU INBOARD CONTAINMENT ISOLATION VALVE OPERATOR LIMITORQUE The actuator performs its function during the initial seconds of

a. postulated accident to provide containment isolation, f

siEultaneous with any increase in containment temperature, pressure, or radiation level.

Qualification data for similar Limitorque actuators indicate a thermal lag before ambient temperatures affect internal temperatures and acceptable performance in these environments.

The isolation function for-this valve is also performed by a redundant qualified motor operated valve located outside tho l

containment and which doee not experience the accident environment during the isolation period.

Post-HELB, this operator is required to isolate postulated RWCU line breaks but is located remote from the accident environment.

For these reasons, continued operation with this valve is considered justified pending motor replacement.

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4 Page 16 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT i

ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION: 14 PAGE:

3 SYSTEM:

REACTOR CORE ISOLATION COOLING (13)

COMPONENT I.D.: 13FS-57 PUMP DISCHARGE LOW FLOW SWITCH The ongoing qualification program for this component has identified type test data directly applicable for this component.

Temperature, humidity, and radiation test data surpasses the a

postulsted post-HELB environmental parameters which this component will experience in its limited operating time (6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />).

Post-HELB this component is only required to maintain electrical integrity and does not have to function.

Post-LOCA the only harsh environment parameter is radiation.

A radiation threshold analysis and test data indicates that no materials will experience radiation damage at the 5x105 rads integrated dose.

The expected life at 104*F for the limiting material relative to thermal aging is greater than the life of the plant.

Based on this data, continued operation is considered justified pending final qualification.

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Pego 17 J

I ENCLOSURE (3)

JAMES'A. FITZPATRICK NUCLEAR POWER PLANT

~ ENVIRONMENTAL QUALIFICATION OF. SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR. CONTINUED OPERATION SECTION:

14 PAGE:

8 l

SYSTEM:

REACTOR CORE ISOLATION COOLING (RCIC)

COMPONENT'I.D.: 13MOV-16 RCIC OUTBOARD STEAM ISOLATION VALVE LIMITGRQUE SMB-00-10

~

The valve actuator is identical to other Limitorque actuators qualified for postulated accident temperature transients up to e

250*F.

For a postulated line break of the HPCI or RCIC steam l

lines, a temperature transient exceeding 250*F can be postulated

[

for.up to 35 seconds.

For this break, closure of this valve will

[

isolate the source of the steam.

Isolation can also be performed j

by a redundant fully qualified valve operator located inside t

l primary containment (remote from the accident environment).

g Qualification is complete for all other environmental parameters.

Following postulated small break LOCA'.s inside primary I

containment this valve would open to admit steam to the RCIC E

turbine and later isolate following RCIC operations.

For this accident, the valve operator is fully qualified and located remote from the direct accident environment.

e Based on the above, continued operation is considered justified pending valve actuator replacement.

Pcgo 18 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION: 14 PAGE:30, 31, 32 SYSTEM:

REACTOR CORE ISOLATION COOLING (13)

COMPONENT I.D.: 13PS-67A, B RCIC PUMP LOW SUCTION PRESS. SW.

13PS-72A, 8 TURB. EXHAUST HIGH PRESS. SW.

When these pressure switches are required to perform their safety function, their environment is mild.

When their environment is harsh the only requirement on the switches is that they must not ground as they are powered from circuits that also power other safety-related equipment.

i The switches have successfully passed a type test environmental l

qualification test program.

The Authority still has to evaluate the switch specific environmental conditions to complete switch

{

qualification for this plant.

Based on the above information, continued operation with these i

switches is considered justified pending completion of the environmental qualification program.

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Pago 19 ENCLOSURE'(3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT

~

JUSTIFICATION FOR CONTINUED OPERATION SECTION: 14 PAGE:TT-SYSTEM:

REACTOR CORE ISOLA ION COOLING (13)

. COMPONENT I.D.:

13PS-78A-D TURB. EXHAUST DIAPHRAGM HIGH PRESSURE SWITCH When these position switches are required to perform their safety function, their environment is mild.

When their environment is harsh the only requirement on the switches is that they must not ground as they are powered from circuits that also power other safety-related equipment.

The switches have successfully passed a type test qualification test program.

The Authority still has to evaluate the switch specific environmental conditions to complete qualification for this plant.

Based on the above information, continued operation with these switches is considered justified pending completion of the environmental qualification program.

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Page 20 i

ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION: 14 PAGE:14, 15, 16 SYSTEM:

RCIC (13)

COMPONENT I.D.:

13MOV-39, 41, -131 - VARIOUS RCIC MOTOR OPERATED VALVES (DC)

LIMITORQUE SMB-000-5 and SMB-00-10 These valve actuators perform their intended safety function in mild environments.

However, they must maintain electrical i

~ insulation integrity during harsh environments following postulated breakc of the'RCIC steam line until the break is isolated by the RCIC isolation circuitry.

These actuators are identical to other fully qualified Limitorque actuators with the exception of the DC drive motor.

Since energization of this motor is not required to mitigate a postulated RCIC steam line break, motor failures will not impact the isolation of the line break.

This forms the basis for the ongoing qualification l

program.

Based on the above, continued operation is considered justified pending completion of the' qualification program.

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I Pega 21 ENCLOSURE (3)

JAMES A. FITZPATRiCK NUCLEAR POWER PLANT ENVIRONMENTAL QUALI.FICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION: 14 PAGE:

5 SYSTEM:

REACTOR CORE ISOLATION COOLING COMPONENT I.D.: 13LS-12 BAROMETR'IC CONDENSER TANK LEVEL SWITCH i

This switch performs its safety related design function in a mild

[

environment.

However, the existing system logic design f

i incorporates common electrical fusing for this item and other harsh environment electrical equipment requiring qualification.

The mai~n concern for this switch is that it does not lose its insulation resistance to ground when exposed to harsh environment, de-energizing other safety related equipment, f

In a postulated HELB, isolation of affected lines is detected and the isolation signal provided within a matter of seconds.

The probability that this switch could develop a significant ground within this short time is extremely small.

For these reasons, continued operation with this switch in the

(

plant is considered justified pending replacement.

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Paga 22 f

ENCLOSURE (3) l I

JAMES A. FITZPATRICK NUCLEAR POWER PLANT l

ENVIRONMENTAL QUALIFICATION'0F SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION b

SECTION:----

PAGE:----

SYSTEM:

REACTOR CORE ISOLATION COOLING (13)

COMPONENT I.D.: 13PNS-LS4 TURB. TRI'P THROTTLE

~

VALVE POSITION SWITCH When this position switch is required to perform its safety function, its environment is mild.

When its environment is harsh the only requirement on the switch is that it must not ground as it is powered from circuits that also power other safety-related equipment.

The switch has successfully passed a type test environmental qualification test program.

The Authority still has to evaluate i

the switch specific environmental conditions to complete qualification for this plant.

Based on the above information, continued operation with this switch is considered justified pending completion of the i

environmental qualification program.

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l Page 23 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION:

11 PAGE:

3 SYSTEM:

REACTOR BUILDING CLOSED LOOP COOLING (15)

COMPONENT I.D.:

15MOV-103 - DRYWELL COOLER ASSEMBLY "B"

SUPPLY VALVE OPERATOR LIMITORQUE SMB-000-2 This valve actuator may be required to operate in order to provide an alternate cooling water source to the drywell fan coolers following a postulated HELB.

Since no credit is taken for the drywell coolers in LOCA analyses these valves would not be required to operate'for this postulated accident.

This valve actuator is fully qualified for the postulated HELB environment up to 250*F.

However, the specific environment for this motor exceeds this qualified temperature by 23* for a total of 15 minutes.

If this actuator was required, due to its relatively short operating time (<1 minute), operation can be assured with high probability.

Based'on the above, continued operation is considered justified pending motor replacement.

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f Page 24 ENCLOSURE (3)

JAMES A..FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION:

31 PAGE:3, 4 SYSTEM:

CONTAINMENT SYSTEM (16)

COMPONENT I.D.: 16-1PNS-101A, B DIFFERENTIAL PRESSURE TRANSMITTER 16-1PNS-102A, 8 ISOL. VALVE POSITION SWITCH These position switches are required to function during the initial seconds of a postulated accident, to provide position information on primary containment isolation valves.

The valves isolate small instrument lines and " fail shut" in the event of a power failure.

The position switches and their valves are located outside the containment and thus are not exposed to any significant change in their environment due to the postulated LOCA.

For this reason, continued operation is considered justified pending replacement.

Page 25 u

ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION'FOR CONTINUED OPERATION i

i SECTION: 31 PAGE:D SYSTEM:

CONTAINMENT SYSTEM (16)

COMPONENT I.D.: 16-1SOV-101A, B DIFFERENTIAL PRESSURE TRANSMITTCR 16-1SOV-102A, B ISOL. VALVE PILOT SOLEN 0ID These pilot solenoid valves are required to function during the initial seconds of a postulated accident, to actuate air operated valves to isolate small instrument lines.

The valves are located outside the primary containment and thus are not exposed to any significant change due to the postulated LOCA.

For.this reason, continued operation is considered justified pending replacement.

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I Pagn 26 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFIC'ATION FOR CONTINUED OPERATION SECTION:----

PAGE:----

SYSTEM:

CONTAINMENT SYSTEM (16)

COMPONENT I.D.: 16-1RTD-107, -108 -

DRYWELL AMBIENT TEMPERATURE DETECTOR The materials in the RTO's which may be subject to deterioration due to the harsh environment are the RTD mandrel, the lead wire insulation and the terminal blocks.

However, considering the low voltage levels in these circuits (millivolt range), the RTD's can be expected to function eve'n after exposure to the harsh environment from a postulated accident.

In addition, there are a large number of thermocouples also measuring drywell ambient temperature, which provides a large measure of redundancy and diversity.

For these reasons, continued plant operation with these sensors is considered justified pending completion of the qualification program.

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Pags 27 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION: 26 PAGE: 1-6 SYSTEM:

RADIATION MOAITORING (17)

I COMPONENT I.D.:

a.

17RE-50A, B - STACK EXHAUST EFFLUENT MONITOR (LOW RANGE) 17RE-53A, B - STACK EXHAUST EFFLUENT 17RE-53A, 8 MUNITOR (HIGH RANGE) b.

17RE-431, -432 TURBINE BLDG. EXH.

EFFLUENT MONITOR (LOW RANGE) 17RE-434A, B-TURBINE BLDG. EXH.

17RT-434A, B EFFLUENT MONITOR (HIGH RANGE) c.

17RE-458A, B - RADWASTE BLDG. EXH.

EFFLUENT MONITOR (LOW RANGE) 17RE-463A, B - RADWASTE BLDG. EXH.

17RT-463A, B EFFLUENT MONITOR (HIGH RANGE)

These monitors provide for measurement of post-accident plant effluents for release assessment.

This equipment is located remote from the plant areas experiencing direct postulated accident environments.

However, sample stream radiation levela can result in higher than normal radiation levels in the area of this instrumentation.

Both the high and low range units are specifically designed to measure the required range of radioactivity.

The high range units are also shielded to j

protect its electronic components from the accident stream radiation.

Based on this data, continued operation is considered justified pending completion of re-analysis of the shielding design for these units.

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Pago 28 I

i ENCLOSU.RE (,3) s JAMES A. FITZPATRICK NUCLEAR POWERiPLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-hELATED EQUIPMENT

\\

JUSTIFICATION FOR CONTINUED OPERATION s

SECTION: 8 s

~

43 PAGE:42, f

~\\

SYSTEM:

RADWASTE (20) s COMPONENT I.D.: 20SOV-83, 20SOV-95 DRYWELL FLOOR DRAIN SUMP' OUTBOARD ISOL. VALVE PILOT SOLENOID

\\ 't The Radweste solenoid operated valves are required to func ion i

short term for primary containment isolation during a postulated LOCA.

The exposure to this accident environment requires the s'

valves to operate in an elevated radiation environment.

No significant temperature or pressure increases are postulated for the locatio6 of the solenoid valves due to the LOCA event.

In addition, the solenoid valve is required.t'o de-energize in the.

'x performance of its safety related function, i.e.,

the valve is

" fail safe".

Based on the short term operational requirement of-%

the valve, and that failure of this device would not cause.

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'i Page 29 ENCLOSURE (3)

J AMES-A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL'QUALIFICATI0h 0F, SAFETY-RELATED EQUIPMENT

'30STIFICATION FOR' CONTINUED OPERATION

. l s

SECTION:

8 PAGE: 44 SYSTEM:

RADWASTE (20) 4 i

i

} COMPONENT I.D.: 20PNS-83, 20PNS-95 DRYWELL FLOOR DRAIN SUMP OUTBOARD ISOL. VALVE POSITION SWITCH s

The Radwaste position switches are required to-function short

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• term for indicatiott of pr,1 mary containment isolation following a

-postulated LOCA.The valves on which these switchee are

'installe.d fail safd.on loss of power.

Therefore, there can be I

high confidence infthe ppsition of these1velves without relying on the functioning'of these\\ position switchep.

The exposure to this accident environment r,equires the switches to operatc in an elevated humidity,'end radiation environment.- No significant

$pmperature or! preepure'inc'resses are p'ostulated for-the location of the position swit'ches due to the LOCA event.'A preliminary assessment based on>the tssting of the similarly constructed position switches reveals that no detriment'al effects to the position swif.crt operability should occur at the postulated accident radiation leveleA For these reasons, continued operation is consibered!Ju' stifled pending replacement.

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/

ENCLOSURE (3) a

/ JAMES A. FITZPATRICK NUCLEAR-POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT a

JUSTIFICATION FOR CONTINUED. OPERATION

~

SECTION: 12 PAGE-~T-

[

]

SYSTEM:

HIGH PRESSURE COOLANT INJECTION (23)

COMPONENT I,.G.':

23FS-78 HPCI LOW-FLOW PRESSURE

, i SWI T Cit 1

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The 4cregoing qualification program for this component has identified" type test data directly applicable for this component.

Temperature, humidity, and radiation test data surpasses the postul,ated post-HELB environmental parameters which this component will experience in its limit,ed operating time (6 7

hours).

Post-HELB this component is only required to maintain

[

electrical integrity and does not have to function.

I 4

Post-LOCA the only harsh environment parameter is radiation.

A radiation threshold analysis and test data indicates that no f

materials will experience radiation damage at the 5x105 rads integrated dose.

The expected life at 104*F for the limiting f

material relative to thermal aging is greater then the life of the plant.

l Based on this data, continued operation is considered justified l

pending final qualification.

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Pcgo 31 7

i 4

ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAC QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTI[ICATION FOR CONTINUED OPERATION SECTION: 12

?

PAGE: 11 SYSTEM:

HkUH PRESSURE COOLANT INJECTION COMPONENT I.D.: 23MOV-14 HPCI TURBINE STEAM INLET ISOLATION VALVE OPERATOR I

7 This valve actuator performs its intended safety function in a mild environment.

However, it must maintain electrical insulation integrity duriny harsh environment conditions following postulated breaks of the HPCI steam line until the break is isolated by the HPCI isolation circuitry.

This actuator is identical',to other fully qualified Limitorque actuators with the exceptior/ of the DC drive motor.

Since energization of this motor is not required to mitigate a postulated HPCI steam line break, motor failure will not impact the isolation of the line break.

This forms the bas.is for the ongoing qualification program.

s Based on the above, co'ntinu'ed operation is considered justified pending completion of the qualification program.

1

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i Pcga 32 I

ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION:----

PAGE ----

SYSTEM:

HIGH PRESSURE COOLANT INJECTION ____

COMPONENT I.D.: 23LS-99 GLAND SEAL COND. HOTWELL LEVEL SW.

23LS-100 GLAND SEAL COND. HOTWELL HIGH LEVEL SW.

This switch performs its safety related design function in a mild environment.

However, the existing system logic design incorporates common electrical fusing for this item and other j

harsh environment electrical equipment requiring quGlification.

f The main concern for this switch is that it does not lose its 1

insulation resistance to ground when exposed to harsh environment, de-energizing other safety related equipment.

In a postulated HELB, isolation of affected lines is detected and the isolation signal provided within seconde.

The probability that this switch could develop a significant ground within this short time is extremely small.

For these reasons, continued operation is considered justified-pending replacement.

}

Pcgo 33 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT l

JUSTIFICATION FOR. CONTINUED OPERATION SECTION: 12 PAGE: 30 SY. STEM:

HIGH PRESSURE COOLANT INJECTION (23)

COMPONENT I.D.:

23PNS-LS4 HPCI TURB. STOP VALVE POSITION SWITCH When this position switch is required to perform its safety function, its environment is mild.

When its environment is harsh the only requirement on the switch is that it must not ground as it is powered from circuits that also power other safety-related equipment.

The switch has successfully passed a type test environmental qualification test program.

The Authority still has to evaluate the switch speciric envircnmental conditions to complete qualification for this plant.

Based on the above information, continued operation is l

considered justified pending completion of the environmental qualification program.

l

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y Pago 34 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINI!ED OPERATION

~

SECTIONi'12 PAGE:43, 44 50, 51 SYSTEM:

HIGH PRESSURE COOLANT INJECTION (23)

COMPONENT I.D.:

23PS-84A, B HPCI PUMP LOW SUCTION PRESS. SW.

23PS-97A, B TURB. EXHAUST HIGH PRESS. SW.

When these position switches are required to perform their safety '

function, their environment is mild.

When their environment is harsh the only requirement on the switches is that they must not ground as they are powered from circuits that also power other safety-related equipment.

The switches have succescfully passed a type test environmental qualification test program.

The Authority still has to evaluate the switch specific environmental conditions to complete switch qualification for this plant.

Based on the above information, continued operation is considered justified pending completion of the environmental qualification program.

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Pcgo 35 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT i

ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION: 9 PAGE: 2 SYSTEM:

PRIMARY CONTAINMENT ATMOSPHERE CONTROL / MONITORING COMPONENT I.D.: 27E/P-103A,B N2 FLOW TO CONTAINMENT ELECTRO-PNEUMATIC CONVERTER.FOR 27FCV-103A, B These inttruments are utilized to control the flow of nitrogen to the containment following a postulated LOCA inside containment.

This equipment is not exposed to the direct LOCA environment but to secondary environmental effects in the Reactor Building (radiation, elevated temperature).

Alternate methods are avellable for establishing nitrogen flow to the containment for venting purposes should this electro-pneumatic converter fail which utilizes fully qualified equipment in the Reactor Building in conjunction with manual control of the nitrogen flow from a mild environment (CAD Building).

For these reasons, continued operation is considered justified, pending completion of the qualification program.

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Pcgo 36 l

ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION: 9 PAGE:J4 SYSTEM:

PRIMARY CONTAINMENT ATMOSPHERE CONTROL / MONITORING COMPONENT I.D.: 27PS-110A, 8 DRYWELL VACUUM BREAKER ISOLATION SIGNAL PRESSURE SWITCH i

When these pressure switches are required to perform their safety function, their environment is mild.

When their environment is harsh the only requirement on the switches is that they must not ground as they are powered from circuits that also power other safety-related equipment.

The switches have successfully passed a type test environmental qualification test program.

The Authority still has to evaluate l

the switch specific environmental conditions to complete switch qualification for this plant.

Based on the above information, continued operation is considered justified pending completion of the environmental qualification program.

-~

Page 37 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION: 9 PAGE:IT-'

SYSTEM:

PRIMARY CONTAINHENT ATMOSPHERE CONTROL / MONITORING COMPONENT I.D.: 27DWA-HT'A, HTB - CONTAINMENT ATMOSPHERE MONITORING HEAT TRACING YYSTEM INCLUDING 27 RID-102A1,

-B1,

-A2,

-82 and 27 RID-107A1, B1, A2, B2 This system would be required to be functional to minimize the moisture content of containment atmosphere samples required for H2 and 02 monitoring following postulated LOCA accidents.

This equipment is located remoto from the direct accident environcent and would experience secondary environmental effects (radiation and slightly elevated temperature).

The H2 Analyzer System is presently being replaced with a fully qualified H2 analyzer and fully qualified heat trace system.

This new H2 system will be installed and operational prior to startup following the June, 1983 refueling outage.

The 02 Analyzer System will not be replaced with a fully qualified unit until the 1985 refueling outage.

Operation with the existing 02 Analyzer System and heat tracing is considered j

l jttstified because JAF operates with an inerted containment

[

(nitrogen) and following a postulated accident there is no source j

for oxygen generation.

The containment would be completely isolated from all sources of outside air.

In addition, all instrument piping uses nitrogen instead of. air for normal operations.

Therefore, hydrogen concentration would be the only l

parameter required for assessing the need for containment venting.

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l Pago 38 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMEtTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR C0h?INUED OPERATION SECTION:

8 i

PAGE:~73~

SYSTEM:

MAIN STEAM AND MAIN STEAM LEAK COLLECTION SYSTEM (29)

COMPONENT I.D.:

29MOV MAIN STEAM LINE DRAIN INBOARD ISOLATION VALVE OPERATOR i

The actuator performs its fonction during the initial seconds of a postulated accident to provide containment isolation, simultaneously with any increase in containment temperature, pressure, or radiation level.

Qualification data for similar i

limitorque ac,tuatora indicate a thermal lag before ambient temperatures affect internal temperatures and acceptable performance in these environments.

The isolation function for this valve is also performed by a redundant motor operated valve located outside the containment and which does not experience the accident environment during the isolation period.

For these reasons, continued operation is considered justified j

pending motor replacement.

Page 39 ENCLOSURE (3) m i

JAMES A. FITZPATRICK NUCLEAR POWER PLANT l

ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT I

JUSTIFICATION FOR CONTINUED OPERATION I

SECTION:----

PAGE:----

SYSTEM:

MAIN STEAM AND MAIN STEAM LEAK COLLECTION SYSTEM (29)

COMPONENT I.D.:

29MOV-77 MAIN STEAM LINE DRAIN OUTBOARD f

ISOLATION VALVE OPERATOR I

The valve actuator is identical to other Limitorque actuators qualified for postulated accident temperature transients up to 250*F.

The valve actuator is required to operate at the onset of a postulated LOCA.

The valve is located outside the containment, thus the vale environment will not significantly change during valve actuator operation.

In addition, the isolation function for this valve is also performed by a redundant motor operated valve inside the containment.

For these reasons, continued operation is considered justified pending completion of the qualification program.

Pago 40 4

I I

ENCLOSURE (3)

{

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATLD EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION:

8 PAGE:33 T9 SYSTEM:

MAIN STEAM AND MAIN STEAM LEAK COLLECTION SYSTEM COMPONENT I.D.: 20MOV-200A, B LEAK COLLECTION SYSTEM MASTER ISOL. VALVE OPERATOR 20MOV-201A, B LEAK COLLECTION SYSTEM TO STANDBY GAS ISOLATION VALVE OPERATOR 20MOV-202A, 8 LEAK COLLECTION SYSTEM TO STANDBY GAS BACKUP ISOLATION VALVE OPERATOR 20MOV-203A, B MSIV STEAM PACKING ISOL. VALVE OP.

20MOV-204A, B LEAK COLLECTION SYSTEM BACKUP DRAIN VALVE OPERATOR These valve actuators are similar to other Limitorque actuators qualified for postulated HELB accident transients in the Reactor Building.

However, these valves are only required to function after a postulated LOCA during which the local harsh environmental conditions consist of slightly elevated temperature, high radiation, and a slow humidity transient.

The only outstanding item to final qualification is a verification from the equipment vendor that a previously completed qualification program applies to these actuators (i.e., motors).

The design of these actuators and the motor insulation design (Class B) are the same as for fully qualified actuators.

Based on the above, continued operation is considered justified f

pending completion of the qualification program.

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i Page 41 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED.0PERATION.

SECTION: 8 PAGE: 49,50 SYSTEM:

MAIN STEAM & MAIN STEAM LEAK COLLECTION SYSTEM (MSLCS) (29)

COMPONENT I.D.: 29PS-201A, B LEAK COLLECTION SYSTEM 29PS-202A, B HIGH PRESSURE SWITCH These MSLCS pressure switches are required to function following a postulated LOCA.

The exposure'to this accident environment requires the switches to function in an elevated radiation environment.

No significant temperature or pressure increases are postulated for the locations of the pressure switches.

A i

preliminary aecessment, based on type testing of similar switches, reveals that no detrimental effects to the switches' operability should occur at the postulated radiation levels during and after the postulated DBE.

For these reasons, continued operation is considered justified pending final qualification.

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i Paga 42 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT 9

JUSTIFICATION FOR CONTINUED OPERATION SECTION: 10 PAGE:5, 6 SYSTEM:

REACTOR BUILDING VENTILATION COMPONENT I.D.: 66HV-3A, B CRESCENT AREA UNIT COOLER

~

LOCAL CONTROL PANEL These panels are located in the Reactor Building.

The components in the panels that may be subject to deterioration from postulated harsh accident ambient conditions are wire, terminal boards, and heavy duty GE Type SBM manual switches.

For-a postulated LOCA the panels will be required to maintain circuit continuity only.

The panel ambients may have slightly elevated temperatures, and short duration humidity transients.

These conditions are not considered to present any problem for panel functioning.

The postulated radiation levels (180 days) is 2.1x104 Rads which is less than the damage threshold for the materials in the panel (SIS wire, wood floor phenalic, terminal blocks, and phennlic switches).

For a postulated HELB the panels will also be required to maintain circuit continuity on]y.

The external panel ambients will experience te:aperature transients above 175'F for 5-10 I

3

. minutes and an integrated radiation dose of <10 R.

The panel construction will protect the terminal boards and switches from direct moisture impingement and condensation.

These conditions ere not considered to present any problems for proper functioning of panel components.

In addition, after most postulated HELB accidents the Reactor Building would be accessible for corrective maintenance within a short time after onset of accident.

j l

For these reasons, continued operation is considered justified pending modification of panels.

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a Paga 43 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION: 13 PAGE:

2 SYSTEM:

REACTOR BUILDING VENTILATION (66)

COMPONENT I.D.: 66UC22A-K FAN MOTORS The Crescent Area Unit cooler motors are Severe Duty Motors mounted within totally enclosed air-over enclosures (TEAO).

The motors are designed to operate in a continuous, ambient of 150*F with 100% relative humidity.

The maximum temperature in the Crescent Area after a postulated LOCA is 110*F and for a HELB a temperature transient above 150*F for 10 minutes occurs.

However, the motors will not experience these temperatures as they are in-duct mounted downstream of the cooling coils.

The maximum integrated radiation exposure in Crescent Area is 6.9x106 R.

Testing of similar motors with same class insulation shows no significant degradation of insulation due to th.ese j

levels of radiatior:.

J Continued operation is considered justified pending completion of the qualification program.

1 J

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Page 44 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION: ---

PAGE:

SYSTEM:

DRYWELL C00 LING SYSTEM ('68)

COMPONENT I.0.: 68TE-201 thru 212 DRYWELL 68TE-301 thru 310 THERMOCOUPLES The materials in these thermocouples consist of metal (Cu-Const),

ceramic insulators, and a pressed asbestos terminal block with material trade name "Hemit".

The ceramic insulators are aging and radiation insensitive and the "Hemit" material which is good up to 400*C is also listed as aging and radiation insensitive.

In addition, there are a total of twenty-six (26) thermocouples sensing drywell air temperature which provides a large measure of redundancy.

For these reasons, continued operation is considered justified pending replacement.

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i Pago 45 l

ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF' SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION:

22 PAGE:1, 2 SYSTEM:

ELECTRICAL SYSTEM (71)

COMPONENT I.D.:

71INV-3A, 3B and 71 BAT-3A, 3B LPCI INDEPENDENT POWER SUPPLY CHARGER / INVERTER I.

Justification for continued operation is provided based are the following:

a.

LOCA (large) - This equipment is located remote from the direct harsh environment of this accident and would perform its intended design function of providing power to the LPCI valve bus prior to the local temperature, radiation, or humidity significantly exceeding normal conditions.

The required operating time is less than 3 minutus, b.

LOCA (cmall) - This equipment is located remote from the direct accident environment.

r Although the required operating time is significantly longer for this accident (6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> maximum), no significant accident radiation exposure is expected due to its elevation in the Reactor Building and minimal fuel damage that is postulated for this accident.

The long teem temperature does not significantly increase above normal (110*F).

e

I page 45a i

i c.

HELB

- A method of plant depressurization and cooldown following a postulated HPCI or RCIC steam line break is described in NEDD-24297, Revision 1 ("High Energy Line Break Evaluation for the James A.

FitzPatrick Nuclear Power Plant" dated

.0ctober, 1980), Section 5.2.2.

This method of plant cooldown requires manual depressurizat. ion of the reactor using the Automatic.Depressurization: System (ADS) while restoring and maintaining water level using on.e of the two Core Spray Pumps. Based on this analysis, there would be some core heat-up, however, there would be considerable margin to the 10CFR50, Appendix K limit of 2200*F peak clad temperature (PCT).

Following a RWCU line break, HPCI and RCIC' Systems located in the Crescent Area experience an insignificant change in environmental conditions (5'F rise for less than 30 seconds, and a 0.5 psig pressure rise for less than 30 seconds).

Therefore, these systems will remain 1

functional to provide high pressure cooling.

Following depressurization using'RCIC or HPCI, reactor inventory can be maintained usir.g one of two Core Spray Pumps.

Refer to NEDO-24297, Revision 1 (Section 6.2.4).

II.

In addition, post-HELB temperature / pressure analyses are extremely conservative.

III. Based on the above analysis, continued operation is considered justified pending full qualification or electrical system mooifications.

1 d

Page 46 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION: 23 PAGE:TT"70 SYSTEM:

ELECTRICAL POWER (71)

COMPONENT I.D.: 71ACAS, B5 - 120 VAC DISTRIBUTION PANEL PT-71ACAS, 85 INCLUDING 600/120 VAC IRANSFORMER These panels power 2 electrical loads which may be required to j

mitigate postulated design basis accidents.

t LOAD i

a.

27NS-CA, CB - Nitrogen Instrument Supply Cabinet b.

711NV-3A, 3B - LPCI Independent Power Supply Control Power 1.

27NS-CA, CB - These panels provide 120VAC power to various instruments and control components associated with the nitrogen containment Air Dilution (CAD) System.

This equipment would only be required to perform its intended design function following a postulated LOCA inside primary containment.

These electrical panels would not be exposed to the direct accident l

environment but would be exposed to secondary l

environmental effects on elevation 300' of the i

Reactor Building.

This accident environment l

would consist of a mild increase in I

temperature (110*F maximum), a mild humidity transient, and radiation (3.0x105 Rads).

Distribution breckers of similar design have been shown by type testing to withstand I

radiation doses of 4.4x10) rads.

In l

addition, since the electrical circuits are loaded to a maximum of 80% of its trip rating by design (40-104*F), operation at a maximum temperature of 110*F will not trip the breaker.

l l

Page 46a 2.

711NV-3A, 3B - This load is the control power supply for the LPCI independent power supply charger / inverter logic.

This power source is only required for startup of the LPCI charger / inverter.

Once output voltage is established, failure of the external power source'to the control logic will not affect inverter operation.

Based on the above reasons, continued operation is considered.

justified pending relocation of these loads.to a distr.ibution panel in a mild environment.

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Paga 47 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION: 20 PAGE:. 2 SYSTEM:

MISCELLANEOUS COMPONENT I.D.:

CINCH JONES (TB) TERMINAL. BLOCK IN A GASKETED STEEL JUNCTION BOX 1.

Refer to the justification for continued operation provided for items 20 (1, 3, 4,

5, 6, 7).

2.

Five boxes have been identified which have a Cinch Jones (30A) terminal block instead or a General Electric block.

The physical design of these b1ccks and materials are

(

similar to the qualified General Electric terminal blocks.

i The Cinch Jones blocks are one piece, molded, 12 point (30A)

S manufactured of general purpose black phenolic.

They are designed for a continuous rating of 250*F and have a threshold for radiation damage greater than 1.0 x 106 R.

l 3.

Based on the above, continued operation is considered justified pending terminal block replacement.

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}

Pegs 48 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTf,L QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION 1

SECTION: 20 PAGE:TT374' 5,6,7 SYSTEM:

MISCELLANEOUS COMPONENT I.D.: GE EB-5, -25 TERMINAL BLOCK IN A GASKETED STEEL JUNCTION BOX Qualification type testing for a full post-LOCA containment environment applies to this device.

The only difference between the tested configuration and the installed configuration is the presence of a 1/4" weep hole for condensation drainage.

These junction boxes are located in the Reactor Building at JAF which i

-experiences significantly milder postulated post-accident environments.

The 1/4" weep hole has been provided in Reactor Building boxes which experience post-accident pressure transients greater than 0.8 psig.

l Post-LOCA, the Reactor Building environment consists of a maximum temperature of 110*F, no pressure increase, a slow humidity transient, and poetulated high radiation exposure.

There should be no degradation of performance for these environmental conditions based on the type test data and a radiation threshold and thermal aging analysis performed on the terminal block materials.

Post-HELB, the Reactor Building temperature transients at some locations are harsh, but not to the extremes and duration of the I

LOCA type-testing for this junction box configuration.

There is insufficient moisture present in the sealed junction boxes such that a significant condensate drainage would be expected.

l l

Based on the above, continued operation is considered justified pending drilling of weep holes in all junction boxes.

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6 4

Paga 49 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION: 15 PAGE:

5 SYSTEM:

-MISCEL'LANEOUS COMPONENT I.D.:

~JB-X-101-E ELECTRICAL PEN'ETRATION i

GENERAL ELECTRIC - CANNISTER TYPE This penetration is identical to other fully qualifie'd electrical penetrations installed at the JAF Plant.

The only safety-related circuits wired through this penetration are associated with the safety-relief valve acoustical monitoring system (NUREG-0578,

-0737) whose qualification status and JC0 is provided under System 02 (36/2, 3, 4).

Following completion of the qualification testing program, the entire acoustical monitoring system will be upgraded to the fully qualified status for all components of the system located in postulated accident harsh environments.

In the interim, a backup method of SRV position indication is provided by a temperature sensing system (thermocouple) installed on each relief valve discharge pipc.

High temperature detected on each discharge line is alarmed and indicated in the Control room.

These thermocouples are designed to measure temperatures under harsh environment conditions.

i In addition, there are other methods available to the operator to detect a stuck-open relief valve (water lever, reactor pressure) which will also automatically initiate protection and ECCS t

systems.

l Based on the above, continued operation is considered justified pending completion of the ongoing qualification program and l

planned system upgrade to a fully qualified condition.

l 4

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i

o Paga 50 ENCLOSURE (3)

JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SECTION:----

PAGE:----

SYSTEM:

MISCELLANEOUS (JUNCTION BOXES)

COMPONENT I.D.:

N/A A concern was expressed in the original NRC SER response of potential implosion of junction boxes due to postulateo HELB pressure spikes.

Due to the design and installation of the boxes and the connecting conduit raceway system which provides adequate vent paths implosion will not occur.

In addition, junction boxes with a 1/4 inch " weep hole" are used at JAF at all locations where the pressure increase will be greater than.8 psi.

It has been demonstrated by type test that boxes of similar construction have successfully withstcod accident conditions with a pressure spike of up to 40 psig without implosion while the internal terminal block maintained the required electrical continuity and insulation resistance.

Therefore, continued operation with these junction boxes is considered justified pending final qualification.

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