ML18082B213

Forwards Addl Info Re Equipment Qualification,In Response to NRC 800908-09 Audit Request

ML18082B213
Person / Time
Site:	Salem
Issue date:	09/18/1980
From:	Mittl R Public Service Enterprise Group
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8009220280
Download: ML18082B213 (86)

Text

/

PS~G Public Service Electric and Gas Company 80 Park Plaza Newark, N.J. 07101 Phone 201/430-7000 September 18, 1980 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Attention:

Mr. Frank J. Miraglia, Chief Licensing Branch 3 Division of Licensing Gentlemen:

EQUIPMENT QUALIFICATION NO. 2 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-311 PSE&G hereby submits, in the enclosure to this letter, the additional information concerning Equipment Qualification requested by members of your staff at an audit conducted on September 8 and 9, 1980 at PSE&G headquarters in Newark, N

• J

• Should you have any questions in this.regard, do not hesi-tate to contact us.

Enclosure CC:

Mr. Leif Norrholm Salem Resident Inspector M P80 89 06 The Energy People 8 0 0 9 2 2*0.2¥-C)".

f Very truly yours, R. L. Mittl General Manager -

Licensing and Environment Engin~ering and Construction 95-0942

I PSE&G SALEM UNIT 2 SUPPLEMENTARY INFORMATION RELATED TO EQUIPMENT QUALIFICATION SEPTEMBER 19, 1980 M P80 91 14/1

CONTENTS

=

INTRODUCTION

1.

Classlflcatlon of Significance Definitions

2.

Item Classification

3.

Justifications for Full-Power. -Operation

4.

Qua I lty Assurance Aspects

5.

Compliance with NUREG-0588 M P80 91 14/2

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-~-

• ~

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-- -- ~--------:~~~~"'-*----~-~-*...._._______.. __ ~.. -----*-~* ---~-*--** --- *---* - ---- --~

INTRODUCTION PSE&G submitted Its report on Equipment Qua llf lcatloTI comp I lance with NUREG-0588 to the USNRC on August 25, Subsequently, a review was conducted by the NRC Staff September 8 and 9, 1980 In PSE&G headquarters.

an d

~

1 9 8 0.

on The purpose of the enclosed Information Is to respond to the NRC requirements specified during the review.

This Informa-tion consists of a "c lasslflcatlon of significance" of the Items previously Identified In our submittal, a tabulation of equipment by classlflcatlon group, justifications for Interim use In those cases Identified as possible problems with a schedule for correction, a discussion of the Quality Assurance activities pertinent to the review of NUREG-0588, and I dentlflcatlon of exceptions to NUREG-0588 requirements.

C larlf lcatlon Is necessary to understand why the report Is so extensive.

The review of envlronmenta I qua llf lcatlon of equipment at Salem was performed under these criteria:

Identification of Class 1E equipment which must operate to mitigate the consequences of design basis accidents and must operate In harsh environments.

This Included devices required for Initiation of automatic protective functions and those for post-accident recovery actions.

Identification of equipment which ls located In an area subject to a harsh environment and Is not required to perform a safety function but whose mlsoperatlon In such an environment may cause a degradation of other safety equipment required or may adversely affect the perform-ance of safety functions.

Identification of devices/components which provide Indication to the operator for which he Is presently called upon to utl llze In the emergency operating pro-cedures and are located In an area subject to a harsh environment.

This Includes devices needed for the per-formance of safety actions by the operator and those that provide supplementary Information.

The first Item ls of NUREG-0588 and tlon reviews.

se If-explanatory Its requirements and really Is the basis on equipment quallflca-The second Item deserves further exp lanatlon.

Many of ftems found In this category were contro I systems for the norm a M P80 91 14/3

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p I ant operation which have been previously assumed to remain as ls during safety analyses for accident scenarios.

These type devices and the potent la I effect on safety due to envlronmenta I Induced fa! lure came to light In September 1979 via NRC letters to operating p I ants, and Issuance of Information Notice 79-22.

It was addressed by Salem In C'ER 79-58 and, letter to NRC Staff dated October 4, 1979.

At that time a number of potent la I fal lures were analyzed and f o u n d t o b e a c c e p t a b I e a t S a I em,s I n c e p I a n t s a f e t y w a s n o t compromised.

In the spirit of addressing plant safety In this review and consideration of envlronmenta I consequent la effects, these devices, If found to be located In areas sub-ject to harsh environments, were listed with justifications to show that qua llf lcatlon In accordance with NUREG-0588 was not required.

Th Is was done for the steam atmospheric power operated rellef va Ives CMSl O's>

and ex-core neutron detec-tors even though previously Identified in LER's and letters and found not to be a safety problem.

Our equipment quali-fication review has reaffirmed that position.

Some other Items, siml lar in nature, have also been Identified such as loca I CMC swii*ches for main steam stop va Ives CMSl 67's>,

proportiona I adjusting transmitters CPAT>

for steam warmup valves CMSl S's) and electric/pneumatic converters CE/P) for residua I heat removal valves CRH18's and R20).

These devices are also used for norma I operation and not safety functions.

No safety concern was Identified without qua I 1-flcatlon per NUREG-0588 requirements provided design changes or power lockouts are Incorporated In the Salem design.

The justifications provided for these devices Indicates that th Is has been/w I I I be done.

The third Item Is not directly specified In NUREG-0588 but Is well enumerated In NRC IE Bulletin 79-019 which also deals with environmental qualification reviews.

The purpose of listing the emergency procedure Items that are located In areas subject to a harsh environment Is to assure that sufficient equipment has adequate qua liflcatlon for the performance of accident recovery actions and that the non-essential or supportive indications Clf non-qualified) will not be misleading to the operator and cause him to take actions contrary to safety.

In this supp lementa I report as well as the original, justifications are made for Items In the procedures which do not fu I ly comply with NUREG-0588 qua llflcatlon documentation requirements even If they are supportive/non-essential Indications.

These justifications provide the assurance that they are not required for safety actions or cause the operator to take Inappropriate actions.

In some cases the emergency operating procedures will be modified to alert the operator regarding unrellabll-

't'-

M PBO 91 14/4

lty of some Indications.

The Information that ls provided I s cons I st en t w I t h that re q u I re d by Bu I I et I n 79 - 0 1 B an d s u p -

p lament.

It Is Included to provide complete documentation of this review even though not speclflcal ly required by NUREG-0588.

M PBO 91 14/5

CLASSIFICATION OF SIGNIFICANCE

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DEFINITIONS

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M PBO 91 14/4

1.-

I' In our submittal of August i5, 1980, the equipment lacking complete environmental documentation was Identified.

This equipment has now been classified for significance to plant safety In accordance with the definitions provided below.

Classification group I contains Items considered to be re-quired for plant safety; group II Items are considered de-sirable but not required, to maintain the plant In a safe condition.

Subsequent portions of this report address each classifica-tion group and provide specific justification for the Issu-ance of a ful I-power license.

GROUP GROUP I I Items In this group Include:

.1 Those whose failure could directly affect equipment required to operate during design basis accident conditions.

This Includes both short and long-term requirements *

• 2 Equipment used by the operator to determine the type of event and the required recovery actions, or used to change the mode of equip-ment operation during the event.

Items In this group are:

.1 Those associated with aux I I lary equipment not required to perform a safety function and not requiring environmental qualification *

• 2 Those cited In existing plant procedures t~

be used to verify operation of the safety systems.

Operator action, however, Is Inde-pendent of the Indication *

• 3 Those which are affected only by non-deslgn-basls-event I lne breaks outside of the con-tainment.

In these cases, the operation of accident-mitigating equipment Is not required *

• 4 Those lacking complete aging Information and consequently they lack an established quali-fied life.

The aging Information Is being developed and Is expected to be aval I able by the end of 1980.

M P80 91 14/5

ITEM CLASSIFICATION

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M PBO 91 14/6

SAFETY FUNCTION Group I.1 A

• Main Steam Atmospheric Rellef Valve Control DEVICES E/P C21MS10)

E/P C22MS10)

E/P C23MS10)

E/P C24MS10)

TYPE FI scher-Governor 546 PA8593 PA8596 PA8594 PA8595 F I s c h e r - P o r.t e r 50EP1041 B

• M a I n Ste am Isolation 1

• Solenoids SV0269 C21MS7) ASCO HT8321A2 SV0279 C22MS7)

SV0278 C23MS7)

SV0288 C24MS7)

SV0587 C21MS18) s v 0 5 8 1 C 2 2M S 1 8 >

SV0585 C 2 3M S 1 8)

SV0583 C 2 4M S 1 8 )

SV0274 C21MS169)

SV0275 C21MS171)

SV0280 C22MS169)

SV0281 (22MS171)

SVo270

< 23MS1 6 9 >

SV0271 C 2 3M S 1 7 1 >

SV0284 C24MS169)

SV0285 C 2 4M S 1 7 1 )

2.

P.A.T.'s P.A.T.

Leeds & Northrop C21MS18) 10973 P.A.T.

C 2 2M S 1 8 >

P.A.T.

C23MS18)

P.A.T.

C24MS18)

M pao 62 14/1

SAFETY FUNCTION 3

• Control Switches

4.

Limit Switches DEVICES Select Switch C21MS167)

Select Switch C22MS167)

Select Switch C23MS167)

Select Switch C24MS167)

L.Sw. 21MS7 L.sw. 22MS7 L.Sw. 23MS7 L.sw. 24MS7 TYPE C M C.

9 1 0 P G D 5 3 3 Masone I Ian 496-2

~...............

c.

RHR 1 *

2.

Control E/P Controls Flow Switches

o.

Containment Pressure E /P E/P C2RH20)

C21RH18)

E/P C22RH18)

FA2569 FA2481 PA2344 PA2345 PA2346 PA2568 Fischer-Governor 546 Barton 289A Barton 332/351 E.

Containment Isolation

1.

MOV's and Solenoids M P80 62 14/2 2CC187 2CC190 2CV284 Llmltorque SMB SV0491 C2CV3)

SV0492 C2CV4>

SV0493

( 2 c v 5 )

SV0427 C2SJ123)

ASCO NP831654E

SAFETY FUNCTION

2.

Limit Switches (Inside Cont.)

3.

Limit Switches (Outside Cont.)

M pao 62 14/3 DEVICES SV0518 (2SS110)

SV0519 C2SS107)

SV0520 (2SS104)

SV0521 (2SS103)

SV0399 C2WL98) 5V0394 C2WL16>

SV0401 C2WL96)

TYPE ASCO 206-381-3F Panel 225 Panel 233 Panel 259 P5E&G Design Vertical Bay L.sw.2CV3 L.5w.2CV4 L.5w.2CV5 L.5w.2so123 L.5w.2WL98 NAMCO EA 180 1 130 2 L.sw.2WL17 L.5w.215594 L.5w.225594 L.5w.235594 L.5w.245594 L.5w.21GB4 L.5w.22GB4 Masone I I an 496-2

- ~ ------ -------- -

L.5w.23GB4 L.5w.24GB4 L.5w.21CA330 L.5w.22CA330 L. 5 w

• 2C C 2 1 5 L.5w.2CC113 L.5w.2CV7 L.5w.25J53 L.5w.25J60 L.sw.2NT32 L.5w.2SS64 L.5w.25S49 L.5w.25533 L.5w.2S527 NAMCO D-2400X

• SAFETY FUNCTION DEVICES L.sw.2WL108 L.sw.2WL13 L.sw.2WL97 L.sw.2WL99 TYPE F

• Boron Injection Tank Reclrc. Valves L.sw.2SJ78 L.sw.2SJ79 L.sw.2sJ1oa NAMCO EA-170-1 130 2 G.

Fan Cooler Control 1

• Solenoid Valves

2.

Flow Controllers M PSO 62 14/4 SV1120 C21SW65) s v 1 1 2 1 C22SW65) s v 1 122 C23SW65) s v 1 1 2 3 C24SW65)

SV1124 C25SW65)

SV1115 C21SW57) s v 1 1 1 6 C22SW57>

s v 1 1 1 7 C23SW57)

SV1118 C24SW57>

SV1119 C25SW57)

SV0621 C21SW223)

SV0624 C22SW223)

SV0627 C23SW223)

SV0630 C24SW223)

SV0633 C25SW223)

ASCO JX8342A22 FA3160 C-1,2,3 Fischer-Porter FA3165 C-1,2,3 53EG3000 FA3169 C-1,2,3 FA3172 C-1,2,3 FA3176 C-1,2,3

SAFETY FUNCTION H

• Excore Neutron. Detectors DEVICES Excore Neutron Detectors TYPE Westinghouse WC-23686 Group 1.2 A.

Hydrogen Analyzers XA-8650 XA-8651 Bacharach B

• Containment Temperature

c.

lncore Thermcouples D

• Containment Radiation E

• Containment Humidity Detectors T A4 31 2 TA4313 T A4 31 4 TA4315 TA4316 TA4318 T A4 3 4 8 TA4319 TA4320 T A4 3 2 1 Radiation Monitor HI gh Range TA6356-2 TA6357-2 TA6358-2 TA6359-2 TA6360-2 Tern.Tex.co.

304-250-T-G-12-SA2-1H CC-TC Vlctoreen 877 Foxboro 271 lAG Group I I.1 A.

PORV Control 2PR47 2PR4 8 Marotta M V 2 2 5C B

• Fan Motors.

Nozzle Support Vent Reactor Shield Vent Control Rod Drive West! nghouse Group 11.2 A*

Pressurizer Power Operated Relief Valve Indication L.sw.2PR1 L.sw.2PR2 NAMCO D-2400X M P80 62 14/5

SAFETY FUNCTION B.

Steam Generator Level and Aux II I ary Feed-water Flow

1.

Wide-Range Level

2.

Instrument Panels

3.

Operator Indication Aux. Feedwater Flow DEVICES TYPE LA0009 Barton 384 LA0015 LA0021 LA0027 Panel 448-2A NEMA 1 2 Panel 448-28 Instrument En-Panel 448-2C closure PSE&G Panel 448-2D Design FA1087 FA 1091 FA1095 FA1097 Fischer-Porter 1082495 FA3969 Fischer-Porter FA3970 50ES3212 FA3971 FA3972

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c.

Main Steam Valve Indication L.sw.21MS167 L.sw.22MS167 L.sw.23MS167 L.sw.24MS167 L.Sw.21MS10 L.sw.22MS10 L.Sw.23MS10 L.sw.24MS10 L

• S w

• 2 1 M S 1 8 L.sw.22MS18 L.sw.23MS18 L.sw.24MS18

• * * * * *

• m * * * *

• D

• Sampllng Isolation Valve Position L.sw.2SS110 L.sw.2SS107 L.sw.2SS104 L.Sw.2SS103 M P80 62 14/6 NAMCO D2400X-2 Masonellan 496-2.

NAMCO D-2400X

SAFETY FUNCTION E.

ECCS Equipment 1 *

2.

B

• I

• T *,

S

• I

• Pumps Press.

Cont. Spray Addi-tive Tank Level DEVICES PT0942 PA0227 PA7461 LA0217 TYPE Fischer-Porter 50EP1041 Barton 332/352

3.

Charging Flow to BIT FA7462 FI sch er-Porter 1 0B2496

s. I. Pump Discharge FA7464

4.

RHR Pump Disch.

RHR Pump Disch.

c.s. Additive Tank Cont a I nment Sump Level F.

Rod Position Indication Group 11.3 A

• Containment Isolation Valves M P80 62 14/7 F AO 2 2 6 FA1422 FAl 423 FAl 416 FA1419 FA0432 FA0218 Fischer-Porter 1 OB2495

• • • • • • • • • • • • • • • • • • • • • • • • • • • a***

LA0223 LA0224 s v 01 1 4 C2SJ53)

SV0575 C2SJ60)

SV0249 C2NT32)

SV0510 C21SS94)

SV0511 C22SS94)

SV0512 C23SS94)

SV0513 C24SS94)

GEMS LS-800 ASCO LBX831614 ASCO 8320A101 0

SV0558 C21CA330)

SV0559 C22CA330)

ASCO JX8342A22

SAFETY FUNCTION A

• Containment Isolation Valves (Cont'd)

M PBO 62 14/8 DEVICES SV0688 (2CC215)

SV0164 (2CC113)

SV0425 C2CV7>

SV0706 (21GB4)

SV0707 (22GB4)

SV0708 (23GB4)

SV0709 (24GB4>

TYPE ASCO LB831654 ASCO LBX8316 ASCO FT8321A2 SV0396 ASCO LB83146 (2WL13)

SV0398 ASCO FT8321A2 (2WL99)

SV0514 ASCO LB83146 (2SS64)

SV0515 (2SS49)

SV0516 C2SS33)

SV0517 (2SS27)

SV0400 (2WL97)

SV0395 (2WL 17)

SV0805 (2VC4)

SV0928 (2VC5)

SV0804 C2VC1)

SV1024

<2VC4)

SV1023 (2VC1)

SV1025 (2VC5)

ASCO J8320A21 ASCO HT834475 ASCO HTB34477

SAFETY FUNCTION A*

Containment* Isolation Valves (Cont'd)

DEVICES TYPE SV1078 ASCO JX8342A3 C2VC8l SV1080

( 2 v c 1 0 )

SV1082 I

( 2 v c 1 2 )

SV1085 C2VC14l SV0117 C2SJ78l SV0118 C2SJ79) s v 0 1 1 9 C2SJ108)

ASCO LBX831614 B

• Limit Switches L.Sw.2VC1 L.sw.2VC2 L.sw.2VC3 L.sw.2VC4 L.Sw.2VC5 L.sw.2VC6 L.sw.2WL16 Masone I I an 496-2 Group I I.4 A.

Pressurizer Level Steam Generator Narrow Range Level RCS Wide Range Pressure LA0086 LA0087 LA0088 LA0005 LA0006 LA0007 LA0011 LA0012 LA0013 LA0017 LA0018 LA0019 LA0023 LA0024 LA0025 PA0039 PA8088 Barton 764 LOT 1

M PBO 62 14/9

SAFETY FUNCTION DEVICES TYPE B

• Fan Cooler Motors Westinghouse

c.

ECCS/Contalnment Iso-lation Motor Operated Valves 2RH26 21RH29 22RH29 2RH1 2RH2 21RH19 22RH19 21CS36 22CS36

• 21CC16 22CC16 21SJ113 22SJ113 2SJ30 21SJ40 22SJ40 21SJ134 22SJ134 2SJ135 21RH4 22RH4 21SJ44 22SJ44 21SJ49 22SJ49 2PR6 2PR7 2SJ1 2SJ2 2 s J,4 2SJ5 2SJ12 2 s J 1 3 22SJ45 21 SJ45 2CV68 2CV69 2CV139 2CV140 2CV175 2CV40 2CV41 Llmltorque SM.B D

• Aux! I lary Fdwtr. Flow FA1087 FA1091 FA1095 FA1097 Fischer-Porter 10B2495 M P80 62 14/10

SAFETY FUNCTION DEVICES TYPE E

• Steam Generator Steam FA0101 Rosemount Flow FA0687 1153AHA FA0104 FA0690 FA0102 FA0688 FA0103 FA0689

• *** Iii **........................

F

• Steam Generator Steam PA0082 Rosemount Pressure PA0083 1153AGA PA0084 PA0087 PA0667 PA0671 PA0734 PA0670 PA0674 PA0736 PA0668 PA0672 PA0738 PA0669 PA0673 PA0740 G

• H

• Hydrogen Recomblners 0

Terminal Blocks Westinghouse Buchanan J

• Cables American Insulated Wire Samuel Moore Boston In-su I ated Wire Trlangle-PWC, I n c

• Anaconda Rockbestos Okonlte M PBO 62 14/11

JUSTIFICATIONS FOR FULL-POWER OPERATION

============================

M PBO 91 14/7

i e

e The fol lowing Information provides justlf lcatlon for the use of. Items lacking complete qua( lflcatlon data.

The Informa-tion has been grouped Into the classlf lcatlons defined earl !er.

Each justification for ful I-power referenced to the original report continuity.

M PSO 91 14/8.

operation ls cross-for clarlflcatlon and

GROUP I.1 JUSTIFICATIONS M PBO 89 04/1

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• ****-.**o*--,*...... * **-** **--- *****--*-.----..-.-----*-R---...-------,---------*

A.

MAIN STEAM POWER OPERATED ATMOSPHERIC RELIEF VALVES CONTROL (Basis #6 & #7)

The main steam power operated atmospheric relief valves C21MS1 O, 22MS1 O, 23MS1 0, 24MS1 0) are operated by a steam pressure transmitter control slgnal through an electric/

pneumatic conve~ter.

These are control grade lnstru-ment.s are were not required to be operable for accident mitigation.

However, a malfunction could result In the Inadvertent opening of the power operated rel lef valves.

Each MS10 valve has an Individual steam pressure trans-mitter and electric/pneumatic converter associated with It.

The 21MS10 and. 23MS10 associated transmitters and electric/pneumatic converters are located In the Inboard penetration area and the 22MS10 and 24MS10 associated devices are located In the outboard penetration area.

These are phys I cal ly separated areas In the plants, so that a high energy I lne pipe break CMSLB) In one area wl 11 not affect equl pment In the other area.

The steam pressure transmitters for the MS10 valves are Fisher-Porter 50EP1041.

We have examined an earl fer qua I lflcatlon report and found that It was conducted via separate effects and not sequential testing.

We have also found that other reports exist demonstrating proper operabl llty under environmental conditions appropriate for Salem.

This Information was derived from a sequen-tial test performed for other utl lltles.

A copy of the report has been requested from NUS Licensing Information Service.

We wl I I review the report per NUREG-0588 criteria to confirm Its acceptability for Salem.

This review will be completed by November, 1980.

The electric/pneumatic converter Is a Fisher-Governor 546.

Quallflcatlon data Is not In our possession but the vendor has Informed us that tests have been per-formed to demonstrate operabl I lty of the device In a harsh environment.

A copy of the report Is being for-

-*--*------w-a_r

__ d_En:l ___ t_o--p S E&-G-.--A--r-eYr Efli--o *f--T-li"-e--ve n-d-c5r--d at a w I I I b e completed by November, 1980.

In the meantime, the potential mlsoperatlon of the MS10 valves has previously been discussed In LER 79-58 sub-mitted for Salem #1.

A copy of this LER Is attached.

Although the analysts was written specifically for Unit 1, It Is applicable to Unit 2.

The analysts demon-strated that potential mlsoperatlon of MS10 valves In a M P8.0 89 04/2

e e

partlcul*ar penetration area due to postulated steam I lne or feedwater I lne breaks In that penetration area was not a safety conslderatlo~ for Salem and does not I nval-ldate any safety analyses performed for Salem.

Steam I lne and feedwater I lne breaks Inside the contal n-ment wl I I not cause an adverse environment In the In-board/outboard penetration areas.

A loss of coolant accident wl I I not result In an adverse environment In the Inboard/outboard penetration areas during the Ini-tial stages of the accident.

During long term recovery, the Inboard/outboard penetration areas could be subject to radiation exposure.

There Is no reason to expect that the devices would mlsoperate during the first few hours of recovery, however, such mlsoperatlon does not pose a safety hazard~

The MS10 can be closed by turning off the control power and/or air supply (fa! I closed valve) thereby assuring that the valve cannot mlsoperate during recovery.

The operating procedures have been revised to alert the operator t~ possible mlsoperatlon of MS10 1 s In the In-board/outboard penetration area as Indicated In the LER.

In addition, guidance regarding actions that can be taken to assure MS10 closure wl I I 8e provided by November, 1980.

B.

MAIN STEAM ISOLATION

1.

Solenoids for MS7, MS167 and MS18 Valves Solenoids (Basis #BA, #SB, IBC)

The 21-24 MS7, 21-24 MS167 and 21-24 MS18 valves are Isolated upon a main steam Isolation signal through operation of pl lot solenoid valves~

The solenoid valves are ASCO HT8321A2 and have a margin for higher ambient temperature operation.

ASCO has In-dlcated that data exists to demonstrate oper~bl I tty of the valves under adverse temperature conditions.

In addition to the ASCO I nformatlon, solenoids for the MS7 and MS167 valves are located In Instrument panels which afford thermal protection (lag time for peak temperature) for the devices.

Isolation occurs within 30 seconds <conservative) and the solenoids must only operate for approximately 5 seconds.

Therefore, the solenoids wl I I not see extreme high temperatures C320°F for MSLB In Inboard/outboard M PSO 89 04/3

e e

penetration areas> during their period of required operablllty.

ASCO Is supplying further Information which wt I I be reviewed against NUREG-0588 criteria by November, 1980 to conf lrm our conclusions.

2.

PAT for MS18 Valves (Basis 69)

The ~aln steam warmup valves (21-24 MS18> are oper-ated by a proportional adjusting transmitter via a pl lot solenold valve.

The MS18 valves close upon a main steam Isolation slgnal through deenerglzatlon of the solenold valve.

However, upon a reset of the main steam Isolation signal, the solenold valve reenerglzes and control Is returned to the position-ing device.

This particular Item was previously addressed In the response to NRC IE Bulletin 80-06, "Engineered Safety Feature (ESF) Reset Controls."

In the Bulletln response, Salem committed to modify the circuitry prior to ful I power operation to el lm-lnate the poss I bl llty of reopening; this has been completed.

Hence, the environmental qua I lflcatlon of the proportional adjusting transmitter ls not required.

This Item will be deleted from the envi-ronmental quallflcatlon review report.

3.

Main Steam Isolation Valve Control Switch (Basis #11)

Each main steam stop valve Is equipped with a local control switch and an enclosed termlnal block which are used for testing the valves during normal plant operation.

The control switches operate a hydraulic pump motor to open and close the valves.

The fat l-ure of these devices cannot prevent the safety func-tion of steamllne lsolatlon since the circuitry Is Interlocked with the plant protection *system to pre-vent operation of the hydraul le pump motor.

The hydraul le pump motor ls not needed for closing the valve under accident conditions.

A reset of the main steam Isolation signal removes the Interlock and could result In a valve opening If the switch contacts were fat led.

To preclude this poss I bl I tty, the motive power for the hydraul le pump motor wl I I be admlnlstratlvely locked out when the valve Is required to be open to pr~vent spurious opening of a steamllne Isolation valve following an accident.

This. wll I be Implemented prior to ful I-power operation.

M PBO 89 04/4

--~-~=-~::__-~::::::_: ~----*--*- --- -- - *-

~---

I

This locked-out condition Is slmlla'r to existing provisions for certain ECCS valves, and wl I I not affect safe operation of the plant.

4.

Main Steamllne Drain Valves' Limit Switches (Basis #lOA>

Each main steamllne drain valve Is equipped with a I lmlt switch used for valve control and position Indication.

The vendor, Masone! Ian, has been re-quested to provide data wh~ch would su~port opera-bl I tty of the_ switch In an adverse environment.

The llmlt switches provide seal-In contacts for the solenoid pilot valves.

Upon a reset of the main steam Isolation slgna I, If the I lmlt switch contacts fall, the solenoid pilot valves may re-energize and cause the MS7 valve to reopen.

The switches wl I I either be replaced, or the control circuit wl I I be modified, during or before the first refuel Ing out-age.

In the Interim, the plant operators wlll be In-structed to de-energize the contror circuit to MS7 solenoid valves before a main steamllne Isolation signal reset to prevent a potential valve re-ener-gization.

Th Is action wl I I only be necessary for the valves which are exposed to an adverse environ-ment.

The valves for two steam generators are located In each of the physically separated penetra-tion areas such that only two sets of valves can be affected by a single event.

The loss of position Indication Is not a safety consideration.

The emergency operating procedures require the operator to verify the closed status of the valves upon main steam Isolation.

_The operators wl I I be Informed that valves within the area of the harsh environment Cl nboard or outboard pen. area>

may provide false Indication.

Determination of whether main steam Isolation has been accomp I I shed can be made by checking main steam pressure and flow Indications <which are quallfled).

If any doubts exist, the operator can deenerglze the control power to the solenoid pl lot valves thereby assuring valve closure.

M PBO 89 04/5

e Plant operation may proceed pending replacement with qualified limit switches and/or circuit design c.h a n g e s d u e t o t h e f a c t t h a t a I t e r n a t e q u a I I f I e d Indications for determining wh~ther steamllne lsola-tlon was accompllshed are avallable and action can be taken to assure that Isolation valves remal n In the closed position.

C.

• RESIDUAL HEAT REMOVAL VALVE CONTROL

1.

Electrlc/Pneumatlc Converters for 21RH18, 22RH18, and 2RH20 (Basis #15)

Electrlc/pneumatlc converters are used In the posi-tion control for the RHR system valves to modulate flow through and around the RHR system heat ex-changers for temperature control during normal cool-down.

Modulation of these valves Is not required during accident conditions.

The RH18 valves must be open and RH20 ls to be closed by protedure.

The position of RH20 Is unimportant sine~ manual valves C21RH12, 22RH12) upstream are closed during power operation.

They are opened when the RHR system Is aligned for a normal cooldown.

The electrlc/pneu~atlc converters for these valves are Fischer-Governor Model 546 which the manufac-turer has Informed us were tested for In-containment use.

PSE&G has requested a copy of the test data.

These devices at Salem do not experience an adverse temperature/pressure environment but may be subject to high Integrated radiation exposure during reclr-culatlon.

We wl I I perform a review of this data by November, 1980 to confirm the manufacturer's state-ment.

Irrespective of the results of this review, a poten-tla I mlsoperatlon of the device causing valve move-ment during recirculation can be avoided~

The con-trol air supply to 21RH18 and 22RH18 can be Isolated during normal power operation.

This causes the RH18 valves to open and remal n open.

Any subsequent mis-operation of the electric/pneumatic converter would not cause the valve to change position.

The air supply would be restored only when the RHR system Is al lgned for normal shutdown operation.

No

~ct Ion I~

required by the operator during the I nltlal stages of an acclde~t or post-accident recovery.

The RH20 M P80 89 04/6

valve plays no part In safety at the Salem plant as presently designed and operated.

If the data review proves uniatlsfactory, procedures

)'Ill I be establlshed to Isolate the control air sup-ply to the RH18 valves during power operation.

2.

RHR Pump Reclrc Flow Control for Valves 21RH29 and 22RH29 <Basis 116)

Each RH29 valve Is control led by a Barton 289A flow switch measuring RHR pump discharge flow.

These devices are located In an area of the plant which wt I I not be subject to a harsh environment during the lnltlal stages of an accident.

The devices are In an area subject to high Integrated radiation exposures during the reclrculatlon, post-accident recovery phase.

The safety analysts for the p I ant considers the RH29 valves closed during recirculation.

In addition, the Salem emergency operating procedures require the operator to verify the closed status of the RH29 valves during reclrculatlon.

A fallure of the flow switch during reclrculatlon could cause the RH29 valves to open.

The operator has the capabl llty for manuaJ override of the automatic control and there-fore can close the valve.

Once closed by manual action, a fat lure of the switch wt 11 not cause valve operation.

The manual override portion of the cir-cuit Is not subject to a harsh environment.

The RH29 valves are motor operated and have adequate quallflcatlon documentation for reclrculatlon opera-tion.

The motor control centers are also quallfled.

Plant emergency operating procedures wt 1.1 be revised to require the operator to manually close the RH29 valves from the control console and verify closure for recirculation operation.

This action can be taken by the operator In the control room and wt I I not be detrimental to post-accident recovery ac-tions.

The procedure change wl I I be accomp II shed prior to ful I power operation.

M PBO 89 04/7

D.

CONTAINMENT PRESSURE CBasls #22)

The containment pressure measurement sensing bellows Is a B*arton Model 352 and. Is located Inside the contain-ment.

Westinghouse has performed quallflcatlon tests on these devices.

The report CWCAP 9157) demonstrated that the device performs satisfactorily.

An anomaly did occur during the test In that a pressure iplke was recorded.

This anomaly does not present a safety prob-lem.

The output of the device recovered following a short duration pressure pulse.

The transmitter Is located outside the containment and Is a Barton 332.

This device Is the same design as a Barton 384 (according to manufacturer's data>; another utl I tty has performed qua I If I cation tests on this device.

The test profile parameters are applicable for Salem.

We have contacted the NUS Licensing Information Ser v I c.e f or a copy of the q u a I I f I cat I on report.

A re-v I e w of the report for comp I lance with N~REG-0588 and confirmation of the qua I lflcatlon data I 1st Information wl I I be completed by November 1980.

E.

CONTAINMENT ISOLATION

1.

Flooded MOV's and Solenoid Valves (Basis 128)

The noted motor operated valves and solenoid valves could become flooded following a LOCA/MSLB Inside the containment:

2C C 1 8 7 CM 0 V)

SV0518 2CC1 90 CMOV)

SV0519 2C C 2 8 4 CM 0 V >

SV0520 SV0491 SV0521 SV0492 SV0399 SV0493 SV0394 SV0427 SV0401 The motor-operated valves are Llmltorque SMB's.

The solenoid valves are the ASCO NP series.

Both de-vices have been qua I I fled for post-accident opera-tion Inside the containment but the data does not support operabl I lty In a submerged state.

The devices are located at an elevatlon In the contain-ment that wl I I not become flooded prior to perform-ing their Isolation function.

The valves close within 30 seconds.

Afterwards when the devices M P80 89 04/8

e e

become flooded, valve operation Is not required.

Operabl I lty In a submerged state for one hour as required by NUREG-0588 Is not warranted under these conditions.

The solenold valves are located In Instrument panels which have been qua I I fled for post-accident opera-tion to assure that the solenoids perform the! r function but have not been proven In a submerged state.

This Is Immaterial In this case also, since the Isolation func+lon Is accomp I I shed prior to sub-mergence.

The electrical circuits are protected by Class IE breakers should shorts develop during submergence.

The effect of a potentlal loss of control power on other devices fed from the same circuit has been analyzed for safety Imp I !cations.

The results, as described In FSAR response to ques-tion 6.28 showed that plant safety was not af-fected.

This analysis has been reviewed by the NRC and found acceptable as reported In Safety Evalua-tlon Report for Salem Units 1 and 2, Supplement 1, June 1976.

In addition, there are redundant outside the containment for these tlonal qua I lflcatlon Is required.

Tsolatlon valves lines.

No addl-

2.

Inside Containment Isolation Valve Limit Switches (Basis #29)

The llmlt switches for the followlng valves could become submerged fol lowing a LOCA/MSLB Inside the containment:

2CV 3 2CV 4 2CV5 2SJ123 2WL98 These I lmlt switches are NAMCO EA-180-11302 and have been qua I I fled for post-accident operation but the qua I Tflcatlon data does not Include proper opera-bl I Tty In a submerged state.

These I lmlt switches are not utl IT zed to perform the containment Isola-tion function.

M P80 89 04/9

The I lmlt switches provide operator I ndlcatlon of valve position and are used In a seal-In control for valve opening.

The Isolation signal from the pro-tection system eliminates the seal-In feature.

Upon reset the I lmlt switch Is restored to the valve con-trol circuit; Its fallure could result In the re-energization of the solenold and a potential reopen-ing of the valve.

However, there are two factors that wl I I prevent this:

a.

These lsolatlon valves require control air to open and remain open (fa! I close on loss of air).

Control air Is Isolated from the contain-ment as part of automatic containment lsola-t I on.

Th e cont r o I a., r I so I at I on v a I v es are outside the containment.

Therefore, the noted lsolatlon valves cannot reopen without motive power (air) even If the solenold does reener-glze.

It should be noted that control air Inside the containment Is not required for post-accident operations.

b.

The solenold valves cannot be re~nerglzed <even If llmlt switches should fal I) unless the con-tainment lsolatlon slgnal Is reset.

The cir-cuitry which provides this feature Is not sub-ject to an adverse environment resultlng from the accident.

Post accident recovery does not require containment lsolatlon reset, only safety Injection reset.

There are separate pushbuttons for each of these actions.

A s d es c r I b e d a b o v e,

t h. e p o t e n t I a I f a I I u re o f t h e I lmlt switches due to floodlng does not have an adverse effect on plant safety.

The loss of Indication or Inaccurate Indication on these valves does not pose a safety problem or cause t h_~~~_!'--~.o_!:___!~ _ _!_~ k e a ct I o n c o n t r a r:_y __ t _C? _ ~ a f et y *

. One action that he could take due to false or I nac-curate Indication Is t'o attempt closure of the valve by pressing the control console pushbutton and checking that the redundant valve outside the con-tainment has closed, and If not, attempt closure.

The other action that could be taken Is to deener-glze the control circuit at the 125V

~C distribution M PBO 89 04/10

-~*-

_ _, '..... *. -* --- --~-

-..,.... }..

e e

cabinets <not subject to adverse environment>.

The operator wt I I be Informed of this matter.

In order to preclude any potentlal fallure If the operator decides that It would be advantageous to reset containment lsolatlon long Into the recovery process, the emergency operating procedures wt I require that the control circuit for the noted valves be deenerglzed prior to lsolatlon slgnal reset.

This wt I I prevent reenerglzatlon of the s o I e ri o I d. v a I v es *

3.

Limit Switches on Outside Containment lsolatlon Valves (Basis 18A)

The I lmlt switches for outside containment Isolation valves provide Indication and control functions

<seal-In circuitry for valve opening).

The llmlt switches utl I !zed are Masone I Ian 496, NAMCO D2400X and Mlcroswltch.

The manufacturers have been con-tacted *to provide supporting data for operabl I tty.

A review of this data wt I I be completed by November 1 9 8 0.

The limit switches are not required to accompllsh containment lsolatlon.

The containment tsolatlon slgnal <not subject to adverse environment> prevents valve opening.

Upon containment Isolation reset the I lmlt switch seal-In feature Is no longer prevented and a potentlal llmlt switch fallure could result In solenold energization and valve reopening.

Reset of containment lsolatlon Is not required for post acci-dent recovery.

There are separate pushbuttons for safety Injection reset and containment lsolatlon reset.

In order to preclude potential mlsoperatlon,* the emergency operating procedures wt I I be revised to alert the operator that the control circuit power should be deenerglzed for the affected valves prior to containment lsolatlon reset to assure that the valves remain closed.

This wt I I be accompllshed prior to full power.

The loss of accurate posltlonlndlcatlon to the oper-ator Is not a safety concern.

The valves Isolate upon containment lsolatlon and the only action that the operator can take Is to press the close push-M PBO 89 04/11 I

button from the control console or deenerglze con-trol power to effect closure should he suspect the valve Is not closed.

The operators wlll be alerted to th 1. s matte r

• The I lmlt switches wt I I be replaced If a study of the quallflcatlon data for the present switches reveals that currently aval fable switches are more suitable.

A vendor has I ndlcated approximately 4 m.onths del Ivery for new switches.

If the switches are replaced, It wlll be done during or before the first refuellng outage.

F.

BORON INJECTION TANK RECIRC VALVES LIMIT SWITCHES 2SJ78, 2SJ79, 2SJI08 (Basis 125)

The BIT reclrc valves close upon a safety I njectlon slg-na f.

Limit switches for these valves are utl II zed for seal-In control.

The I lmlt switches are not used to cause lsolatlon.

Upon a safety-Injection slgna I reset, the devices are restored to the circuit.

The I lmlt switches for 2SJ78, 2SJ79 and 2SJ108 are NAMCO EA-170-11302 switches.

NAMCO has Informed PSE&G that the 170-11302 series switch Is quallfled for harsh envi-ronments outside the containment from steam I lne breaks and 2 x 108 rads.

This Is adequate for the Intended service at Salem since they are locat~d outside the con-tainment and not subject to the same adverse conditions Inside the containment.

The manufacturer has been re-quested to provide supporting data.

We wt I I perform a review of this data to determine compliance with NUREG-0588 requirements.

G.

FAN COOLER UNIT CONTROL

1.

Solenoid Valves for Fan Cooler Unit Service Water Valves 21-25SW65, 21-25SW57 and 21-25SW223

<Basis 19)

The pl lot solenoid valves for the above noted valves are Identified as ASCO JX8342A22 In the orlglnal report.

These valves have been upgraded to M6del X8342B22.

This model represents modifications by ASCO of the standard solenoid valve.

The modifications made to this valve equip It with a M P80 89 04/12

-o..:..::.. -~,..

e 91 solenoid core and col I assembly derived from the NP series solenold valves In order to provide a voltage and temperature margin.

The NP series solenoid valves have been environmentally qua I I fled for adverse conditions.

By slml larlty of the electrical component to NP series and the operating margin of the "H" col Is In the solenoid valves, these devices wt I I operate In an adverse environment for the time frame needed to perform their function.

ASCO Is being contacted to provide supporting evidence.

During Inside containment accident conditions, high service water flow Is required for proper cooling capacity as stated In the safety analysis.

Postulated accidents outside the containment do not require high cool rng capacity for the containment fan coolers.

The SW65 valve solenoids are located w I th I n th e cont,a I nm en t an d a r e ca I I e d u po n to operate within one minute following a LOCA/MSLB Inside containment.

In addition to modlflcatlQn of the solenol d valves with NP series electrical components to provide operabl I tty In high temperature environment, the SW65 valves wt I I open to provide high flow upon loss of control air.

A LOCA/MSLB Inside the containment results In automatic containment Isolation which Isolates control air from the containment.

The solenol d valves located outside the containment

<SW57 and SW223) wll I not be subject to an adverse,

~nvlronment during the time frame needed for operation.

Any subsequent mlsoperatlon of the outside solenoids during long term recovery due to h lgh radiation exposure can be prevented by deenerglzatlon of control power for the solenoids once ful I flow Is established.

The Immediate safety function Is performed by the fan cooler units by operation of the solenoids without operator Intervention.

The procedures wt I be revised to alert the operator to any possible long term mlsoperatlon and the method to prevent any adverse effect on fan cooler operation.

In order to provide a design ellmlnatlng any long term operator actions, the solenoid valves will be replaced with an entire NP series assembly before or during the first refuel Ing outage.

M PBO 89 04/13

e e

2.

Flow Control of Fan Cooler Unit Service Water Valves 21-25SW223 (Basis #20)

Each of the noted service water -valves are provided with flow control equipment consisting of Fischer-Porter 53EG3000 flow controllers, Fischer-Porter 53El3000 electric/pneumatic converter, Fischer-Porter 50ES3212 square root extractor and Fischer-Porter 10B2495 differential pressure transmitter.

These devices are al I located outside the containment.

We have reviewed data on the 1082495 devices and found that It was based on a separate effects test.

We have also found that sequential tests were performed covering the parameters needed for quallflcatlon at Salem.

A copy of the report was requested from NUS Licensing Information Service and we wl I I perform a review per NUREG-0588 requirements by November 1980.

The manufacturers have been contacted regarding supporting Information for the remaining devices.

The safety functions of the fan cooler units can b~

accomplished Irrespective of flow control operabl llty based on the fol low Ing:

Accidents CLOCA/MSLBJ lns~de the containment wl I I not result In adverse environments In the area where the devices are located for the automatic actuation of the fan cooler units.

Accidents outside the containment do not require high cool Ing capacity for the fan coolers.

A lower flow rate caused by fat lure of the devices ls acceptable.

During long term recovery from accidents Inside the containment, the devices will be exposed to high Integrated radiation doses which could affect service water flow.

The operator can assure high flow by deenerglzlng control power to the flow control devices which wl 11 cause the valves to remain open.

The safety function of the containment fan coolers wl I I b*

accomplished without the need for operator action.

Long term recovery may require operator action to compensate for equipment fat lures.

The emergency operating procedures wl I I M P80 89 04/14

l..

e e

be revised prior to ful power operation to provide this Information as part of accident recovery actions.

The flow control wl I I be redesigned or replaced with suitably qua I I fled components before or during the f lrst refuel Ing outage to provide a design that does not require long term operator actions.

H.

EXCORE NEUTRON DETECTORS (Basis #13)

These are Westinghouse suppl led detectors which the manufacturer has tested to verify operabl I lty for sixteen hours at 300°F.

LER79-58 was submitted by PSE&G (copy attached) to Iden-tify that a potent la I mlsoperatlon of the detectors may cause the DNBR I lcenslng criteria to be exceeded for a very short period of time.

An evaluation was performed for this postulated event and determined not to be a problem at Salem because of the 300°F detector operabl 1-1 ty specification.

The physical location of the device precludes It from seeing extreme the period of time In question.

adverse environment for Plant safety Is not compromised since reactor trip Is accomplished lr~espectlve of detector fal lure.

A qua I lflcat.lon documentation review Is not needed sl~ce plant safety Is not affected.

M PBO 89 04/1 5

GROUP I.2 JUSTIFICATIONS M PBO 89 04/16

e

--*)

A.

HYDROGEN ANALYZERS (Basis 630)

Current operating procedures Instruct the operator to monitor hydrogen concentration as a guide for actuation of the hydrogen recomblners.

The procedures are being revised to Inform the operators of a pos~lble Incorrect reading and to Instruct them to actuate the recomblners as part of the procedure to switchover to the recirculation phase.

The actual hydrogen concentration can be determined by grab-sample analysts.

Provisions for Interim and long-term sampl Ing have been described In previous responses to NUREG-0578.

The procedural change described above Is adequate to ensure the timely lnl~latlon of the hydrogen recom-b I ners.

Additional Investigation Into the environmental qua I lflca~lon of the analyzers Is being pursued.

B.

CONTAINMENT TEMPERATURE <Basis 632)

Current procedures Instruct the operator to monitor con-tainment temperature as a guide In determining the type of accident which has occurred.

The ~hermocouples have been specified for suitable operation up to 400°F with a maximum expected containment temperature of 350°F.

It Is anticipated that the thermocouples wl I I operate prop-erly even though envlronmenta I documentation Is not cur-rently In our possession to verify this.

SI nee backup parameters are aval I able.to determine the type of accident, the operators wl I I be advised of the potential for Incorrect temperature Information and the procedures wl I I be suitably modified.

This procedural change wl 1.1 be sufficient to ensure proper diagnosis of the accident.

Additional Investigation Into the qua I I-f I cation of the thermocouples wl I I be pursued.

C.

INCORE THERMOCOUPLES <Basis 633)

Current procedures Instruct the operator to monitor the lncore thermocouples as a guide to determining the existence of adequate core cooling.

Despite a lack of aval I able documentation In our possession, Westinghouse Information and operating history of these devices Indicate that they would function.

If these devices M P80 89 04/17

---=~-~

~~..... :~,.

e 9**

were to fal I, backup parameters a-re aval I able to enable determination of adequate core cool Ing.

These backup devices are currently described In the procedure which also alerts the operator to potential errors In the I ncore temperature readings.

The existing procedure and avallablllty of alternate Indications Is sufficient to ensure proper determination of core-cool Ing adequacy.

Additional I nvestlgatlon Into the quallflcatlon of the devices wll I be pursued.

D.

CONTAINMENT RADIATION MONITOR (Basis #34)

Current procedures Instruct the operator to monitor con-tainment radiation as a guide to determining the occur-renc~ of a LOCA Inside the containment.

Th.ls determina-tion Is read I ly made by use of other direct parameters and Initiation of safety Injection systems.

The proce-dures also mention the use of these Indications.

The Information In the procedure Is pertinent to the lnltla I phase of an accident; the monitor has undergone testing to demonstrate Its capabl I lty during that time period, and Is therefore qua I I fled for the function specified.

The need for a long-term, high-level monitor has been establ I shed by the requirements of NUREG-0578.

A new monitor to meet these requirements will be Installed In the early part of 1981 In accordance with previous agreements related to NUREG-0578 Imp lementatlon sched-u I es.

E.

CONTAINMENT HUMIDITY DETECTORS <Basis 637)

These devices are not required for any safety actions.

Current procedures Instruct the operator to monitor humidity as a guide to determining the occurrence of an Inside-containment accident.

As discussed under Items 8 and D, other direct Indications are aval I able to read I ly I dentlfy an Inside-containment accident.

This negates a need for qua I I fled humidity detectors.

Plant operators wl I I be advised of the possible Inaccuracy of this device.

M P80 89 04/18

GROUP 11.1 JUSTIFICATIONS M P80 87 15/1

A.

PORV CONTROL (Basis #3)

The previous report I lsted this as a problem area because of the assumption that valves 2PR47 and 2PR48 could perform a pressure-relief function during high temperature operation.

These valves, however, are part of the low-temperature, overpressure protection system

<POPS) and are not used during design basis accident, or power operation transient conditions.

The motive power for th e so I en o I d i s I o c k.e d - out du r I n g norm a I opera t I on and cannot open spuriously.

Using these considerations, the valves can only be considered as auxiliary devices, not required for mitigation of accident conditions, and thus do not require quallflcatlon.

No changes are required for this case.

B.

MISCELLANEOUS VENTILATION SYSTEMS (Basis #36)

The nozzle support fans, reactor shlel d vent fans and the control rod drive vent fans were I lsted only because they are mentioned In the emergency procedures.

They are not required to perform a safety function, and Inadvertent operation Con or off) of these devices does not affect the proper course of actions In response to design basis accidents.

The equipment does not require environmental qualiflcatlon.

M PSO 87 15/2

I GROUP I I.2 JUSTIFICATIONS M PBO 87 15/3

I

• . :.. ~.

e A.

PORV INDICATION <Basis #4)

The original report submitted August 25, 1980, was based on the lack of qualified limit switches for the PORV's.

These switches have been replaced with qualified switches.

B.

STEAM GENERATOR LEVEL AND AUXILIARY FEEDWATER FLOW

1.

Steam Generator Wide-Range Level <Basis #5)

The wide-range level devices provide a backup method to verify that adequate auxl llary feedwater flow Is being del lvered to the steam generators.

Since these devices are backups to the narrow-range level devices and aux! I lary feedwater flow devices, qua I lflcatlon Is not currently required.

The operators are not Instructed to terminate aux! I lary feedwater flow based on wide-range level Indication.

During accidents occurring outside containment, the wide-range levels are suitable for use as either primary or backup I ndlcation.

Even though the wide-range levels need not be qualified, It Is planned to replace these transmitters on a schedule contingent upon equipment aval lab! lity and a suitable outage duration.

They wl I I be replaced no later than the Salem 2 refueling outage currently anticipated In the Spring of 1982.

2.

Instrument Panels (Basis #14>

The panels house the instruments used for wide-range level measurement described above.

To provide a qualified Installation, these panels will be equipped with louvered sections when the new wide-range transmitters are lnstal led.

_______________ 3_. ____ A_u xJ_ U a r y __ Ee e d w ate r ___ E_ Lo.w __ (Bas I s __ § l 2 A and # 1 2 B )

Auxiliary feedwater flow measurement Involves the use of transmitters and signal processing devices which are located outside containment In the penetration areas.

Indication of aux! I lary feedwater flow Is a backup method of determining the M P80 87 15/4

.:..~

-~wJ, *..

B STEAM GENERATOR LEVEL AND AUXILIARY FEEDWATER FLOW (Cont'd>

existence of an adequate heat sink during transients or accidents which require aux! llary feedwater.

The primary Indication of heat sink avat lab I I lty Is the narrow-range steam generator level measurement which Is qua I I fled for both Inside and outside containment I lne breaks.

An erroneous Indication of auxiliary feedwater flow would not lead the operator to termination of aux! I lary feedwater.

Outside containment I lne breaks cannot affect al I of the flow Indication; at least two auxlllary feedwater flow Indications would be aval I able*

Alternate methods of determining auxiliary feedwater flow are also available.

Some data exists for transmitters slml lar to the type used at Salem.

This data has been requested from the NUS Corporation and wl I I be reviewed prior to November 1980.

If the data does not confirm that the transmitters are qua I I fled, we p Ian to provide qua I I fled measuring devices and to relocate the signal processing equipment to Improve the overal I reliability of the flow measurement.

These changes are contingent upon equipment aval lab I I lty and would be made during a refueling outage.

c.

LIMIT SWITCHES FOR MAIN STEAM VALVES (Basis #108)

Position Indication of the mat n steam Isolation valves CMS167) Is used by the operator to verify the closed position of the valves.

An erroneous position I ndlcatlon would not result In operator actions

~hlch would negate post-accident system requirements.

Verification of steamllne Isolation Is aval I able by use of backup steam I lne pressure and flow measurements which are qualified.

The operators wt I I be Instructed of possible Inaccuracy In the position Indication for these valves.

A design modification wl I I be made to eliminate this condition.

The modification wlll be accomplished during or before the first refuel Ing outage.

The steam generator relief valves CMS10) position Indication situation has been described In the Salem 11 Licensee Event Report 79-58.

A spurious opening of the M P80 87 15/5

. _,; - _,,.~

~-- *-*--*'--

C.

LIMIT SWITCHES FOR MAIN STEAM VALVES (Cont'd)

MS10 valve results In plant conditions which require the operator to take corrective action.

Direct Indication of an open valve Is only one of several parameters which can be used to Identify the situation.

An erroneous closed Indication would be easl ly detected by the occurrence of fluctuations In steamllne flow and pressure.

An Improvement In system status Indication rellablllty will be achieved by replacement of the limit switch.

Th Is change requires a plant shutdown and can be deferred untl I the first refuel Ing shutdown due to the aval lab I I lty of addltlonal Indication.

D.

LIMIT SWITCHES FOR SAMPLING ISOLATION VALVES

<Basis #18B)

Current procedures Instruct the operator to verify the closed position of Isolation valves.

The I lmlt switches provide valve closure Indication.

Loss of this I ndlcatlon wl I I not cause the operator to take any action contrary to safety of the plant.

The operators wl I I be Instructed to assure that the redundant outside containment lsolatlon valves are closed and to manually actuate the close circuit of the Inside containment valves.

This wlll be accomplished prior to ful I power operation.

Based on the above and our answer to question 7.35 In the Salem FSAR, qua I If I cation of these switches Is not requ I red.

E.

ECCS EQUIPMENT

1.

Pressure Measurements (Basis #21)

Current procedures Instruct the operator to verify ECCS operation by observing boron Injection tank and safety Injection pump discharge pressures during the Initial phase of an accident.

lnoperablllty of these devices would not result In termination of safety Injection.

Alternate Indications are also avallable to verify ECCS status.

Existing qua I lflcatlon Is based on separate effects testing In I leu of sequential effects testing.

Due to the type of service required of these devices, the separate effects test method Is accep~able to quallfy the devices.

Additional qualification Is not required.

M PSO 87 15/6

2.

Spray Additive Tank Level <Basis #23)

Current procedures Instruct the operator to monitor tank level to assure that sodium hydroxide Is added during containment spray.

Operator action Is Independent of the tank level reading since spray Is continued untl I a low-low level Is reached In the refuel Ing water storage tank despite the spray additive tank level I ndlcatlon.

The system used to Inject the NaOH solution operates on a passive eductor principle.

As long as the spray pump Is pumping water, the solution from the additive tank wl I I be educted Into the spray.

Qua I If I cation of the additive tank level Instruments Is not required.

3.

ECCS Flow Instrumentation (Basis #27)

The purpose of these Instruments Is to assist the operator to assess performance of the ECCS.

The operating status of ECCS components would not be altered despite erroneous readings.

Backup system status Indication Is also aval I able.

These Instruments are used only during the lnltlal stages of an accident and can be consJdered qua I I fled for that service on the basis of the separate effects tests In I leu of sequential testing.

No further quallflcatlon ls required.

4.

Containment Sump Level (Basis #31)

Current procedures Instruct the operator to verify adequate RHR pump NPSH by observing the* sump level Indication.

The operator, however, bases his actions on the RWST level and not the sump level.

An erroneous sump level Indication wl I I cause no action to be taken which Is contrary to accident recovery operations.

As a result of NUREG-0578 requirements, these devices.are to be replaced with qua I I fled level Instruments In accordance with the schedule previously submitted <May 31, 1981).

F.

ROD POSITION INDICATION (Basis #38)

Current procedures Instruct the operator to verify a reactor trip by monitoring the rod bottom.lights.

If a successful trip Is not Indicated, a manual reactor trip Is to be Initiated.

This Is a backup type of action not required during accident conditions which generate an automatic reactor trip.

The operators wl I I be advised of the potential mis-Indication but no quallflcatlon Is required for the system.

GROUP I I.3 JUSTIFICATIONS M PSO 87 15/8

i A.

CONTAINMENT ISOLATION VALVES (Basis #17)

The outside containment Isolation valves' solenoids function pr I marl ly to provide redundant lsolatlon during accidents I nslde containment.

The outside Isolation valves' solenolds are not subjected to the containment environment and are quallfled for an accident function.

These valves can be exposed to harsh e~vlronments caused by high energy I lne breaks In the penetration area.

These breaks, however, do not require complete containment lsolatlon.

Existing Information discussed In our August 25, 1980 subml.ttal Indicates that these valves would close In the environment caused by high-energy I lne breaks, and that any subsequent fat lure would tend to keep the valve closed.

Because of the protection afforded by the solenol d valve enclosures, we conclude that the valves would remain In the safe position.

To provide an even more reliable design, these solenoid valves wl I I be replaced with qualified devices during the first refueling outage.

B.

CONTAINMENT ISOLATION VALVE LIMIT SWITCHES (Basis #18A)

The general discussion provided above for the Isolation valve solenolds ls appllcable to the valve I lmlt switches function of providing valve position Indication.

Fallure of the Indication at the time of valve closure Is Improbable; lnltlal verification of closure Is judged to be possible with a high degree of confidence.

The operators wt I I be Instructed to actuate the close circuits of these valves to assure that valves have closed during conditions requiring selective Isolation.

The valve control circuits can also be de-energized to prevent re-opening.

This topic Is also discussed In the report section deal Ing with Group I.1 I t ems.

c.

BORON INJECTION TANK SOLENOID VALVES (Basis #25)

Al I of the valves In question are located outside of the containment and not subjected to a severe environment during accidents which require the operation of the valves.

The valves are therefore quallfled for short-term operabl I tty.

During long-term conditions, the valves cannot re-position because of the envlronmental effects on the solenold and thus are qua II fled.

M PBO 87 15/9

C.

BORON INJECTION TANK SOLENOID VALVES (Basis #25)

(Continued)

Line breaks outside containment In the vicinity of these valves do not require operation of the valves to perform any safety function.

They are also protected by enclosures which I lmlt thermal damage.

No additional qualification Is required, but to Improve the overall rellablllty of the lnstallatlon, qualified valves will be provided during the first refuel Ing outage.

Additional discussion Is provided In the August 25, 1980 submittal.

M P80 87 15/10

e e

GROUP I I.4 JUSTIFICATIONS

==============

M PBO 91 16/1

e e

AGING <Basis #1 and 2)

We have contracted with Wyle Laboratories to perform an ag-ing review of equipment per the requirements of NUREG-0588.

This wll I encompass the Identification of materials which are susceptible to aging mechanisms and which could affect equipment operation.

A qualified llfe wl I I be determined and maintenance programs established for the devices.

This review Is scheduled to be essentially completed by November, 1980 with final completion before January 1, 1981.

A sup-plemental report wl I I be Issued on this topic.

This matter Is not frame In question considered a safety concern for the fol low Ing reasons:

for the time the plant wl I I be operating for only a few months prior to completion of the aging review aging degradation, If It Is a problem, Is not ex-pected to occur In this short period but on the order of many months to years equipment of the type undergoing aging review has been utl II zed In Salem #1 and we are not aware of any fal lures due to aging after a few months service some of the equipment has undergone artificial aging during the qua I If I cation tests such as the cable, NAMCO EA-180 limit switches and motors.

This data demonstrates that devices wt II work during an accident after a period of normal operation.

Wyle Labora-tories wt I I review this data as part of the aging study to determine a qualified life.

M P80 91 16/2

QUALITY ASSURANCE ASPECTS

==

M PBO 91 16/3

The review of the Salem 112 design for comp llance with NUREG-0588 was performed In accordance with the requirements of existing Engineering Department procedures which address the design Input and design verlf lcatlon aspects of the QA program.

The contents of this Information submittal and the previous submittal of August 25, 1980wt11 be Incorporated In a design analysts document prior to full-power operation.

The design analysts document wl I I serve as a record of the methods used In the review for comp I lance with NUREG-0588.

M P80 87 15/12

COMPLIANCE WITH NUREG-0588

===============

M P80 91 16/4

e e

The Salem Unit 2 plant falls under the Category II require-ments of NUREG-0588.

The fol lowing Information ts provided on the review process performed for Salem with respect to NUREG-0588 requirement Items:

I t e m 1

• 1 -

L 0 CA 0 u *a I I f I c at I o n P r of I I e NUREG-0588 states that the LOCA parameters used In the design of the containment structure and found acceptable by the NRC, are appropriate for environmental quallflcatlon purposes.

The values used in the Salem review came from the FSAR, Figures 7.5-5, 14.3-25 and page 14.3-56 and are based on the Salem LOCA analysts.

The NRC has reviewed the metho-dology and values used In this analysis and has found It acceptable CSER for Salem dated 10/11/74 pages 6-2 through 6-6 and supplemented In SER dated 4/80 page 6-3).

Item 1.2 -

MSLB Quallflcatlon Prof! le Th e doc um en t states th at th e MS LB par am et er s sh o u I d be ca 1. -

culated using a plant specific model or other model based on staff approved assumptions.

An MSLB analysis was performed speclflcally for Salem.

The results were presented In response to question 5.82 In the Salem FSAR.

This analysis has been reviewed and approved by the NRC Staff CSER for Salem 4/80 pages 6-2 and 6-3).

Item 1.3 -

Chemical Spray Salem review and requirements address chemical spray as defined on FSAR page 6.4-10.

Item 1.4 -

Radiation Doses The Salem review and cal cu lated doses conform to the re-quirements of NUREG-0588.

Item 1.5 -

Outside Containment Environmental Conditions

--Nu R E-G ---o 5 9-9--~rt-a t e_s ___ t_h_a _t -th *e-ffl:l n'I m e*t er s--f cfl* -outs Ide cont a I n -

ment should be based on the same models/calculations for Inside containment conditions.

The Salem review parameters for outside the containment were not determined by using the same computer codes used for In-containment accidents.

A summary of the method used Is as fol lows:

M P80 91 16/5 conditions are based on worst case pipe break Cf luld temperature and pressure conditions>

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d e t e r' m I n a t I o n o f s t e a m I w a t e r b I o w d o w n r a t e I n c I u d I n g al 1*owances for e.ncapsulatlon of pipes and 1 lmlted flow areas where appl !cable calculatlon of peak temperature and time frames based on steam/water flow expansion to conditions In the room space at constant enthalpy.

The temperature In the room space rises to approximately the steam temper at u re cons I st en t w Ith I sent ha I p I c exp ans I o.n

• This method,* although not In strict conformance with NUREG-0588 criteria, does provide the gross maximum tempera-ture for the area.

Our position ls*that the temperatures establ I shed bound any that would result from a more sophis-ticated analysts.

Further, the methods prescribed In Appen-dix A are geared to analysts of I n-contalnment conditions.

Item 2 -

Qua I lflcatlon Methods The review was conducted In accordance with noted require-ments as appllcable.

In the course of the review, Items which ~Id not meet some of the requirements were noted as outstanding Issues In the original qua I lflcatlon summary report.

These Included separate effects Instead of sequen-tial testing, and less than one hour proven operabl llty after comp let Ion of the safety function.

0 These were noteq In our Justifications In the original report as outstanding and under review.

In this supplemental Information we have completed our review and find that these Items do not pose a safety problem at Salem and that the testing performed and time frames Involved are adequate to determine qua I lflcatlon as noted In the "Justifications" section of the supplemental report.

Item 3 -

Margins The review considered margins In the evaluatlon process.

In the report, devices which varied from the requirements <one hour operabl llty demonstrated for short term require~ oper-ation) was noted as an outstanding Issue In the original report.

Information regarding these devices was provided In the original report and augmented by Information In this supplemental report.

Justifications for acceptablllty of lesser margins are provided.

Item 4 -

Aging The review wt I I be conducted In accordance with this sec-tion.

Further Information Ts contained In the Justifica-tions found In the original report and supplemental report.

M :P80 91 16/6

Item 5 -

Documentation Our review has been In comp I lance with this Item.

An out-standing Issue was Identified If documentation for qua I If 1-catlon was not In our possession for certain Items.

Vendors and other organizations have been contacted to provide docu-mentation to support their statements In catalogs and let-ters.

A review of any supporting documentation wl I I be com-pleted by November, 1980.

Addltlonal Information on this topic for the appropriate devices Is provided In the "Justi-fications" section of the supplemental report.

M PBO 91 16/7

PSl~G

.. '.' RECE\\"ED

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110'1 2 s \\979 f November 10, 1979 V\\CE PRESIDE.NT '..;If PRODUCTION J /.

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Mr. Boyce H. Grier Director of USNRC Off ice of Inspection and Enforcement Region 1 631 Park Avenue King of Prussia, Pennsylvania 19406

Dear Mr. Grier:

LICENSE NO. DPR-70 DOCKET NO. 50-272 REPORTABLE OCCURRENCE 79-58/0lT SUPPLEMENTAL REPORT Pursuant to the requirements of Salem Generating Station Unit No. 1 Technical Specifications, Section 6.9.1, we are submitting supple-mental Licensee Event Report 79-58/0lX-l.

Sincerely yours, l

i

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F. P. Libri'z£i General Manager -

Electric Production CC:

Director, Office of Inspection and Enforcement (30 copies)

Director, Office of Management Information and Program Control

( 3 copies) r:.*

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"'111111

Report Number:

79-58/0lX-l Report Date:

11/10/79 Occurrence Date:

9/07/79 Facility:

Salem Generating Station Public Service Electric & Gas Company Hancock's Bridge, New Jersey 08038 IDENTIFICATION OF OCCURRENCE:

Qualification of Control Systems for Adverse Environmental Conditions CONDITIONS PRIOR TO OCCURRENCE:

Operational Mode 5 DESCRIPTION OF OCCURRENCE:

We have been notified by Westinghouse that a review of the environ-mental qualifications of NSSS equipment has been conducted which showed that several systems were identified which, if subject to an adverse environment, could potentially lead to control system opera-tion which may impact protective functions.

These systems are the steam generator power operated relief valve control, pressurizer

.power operated relief valve control, main feedwater control and automatic rod control.

Each of the above systems could potentially malfunction if impacted by adverse environment due to a high energy line break inside or outside containment.

In each case, a limited set of breaks, coupled with possible consequential control malfunc-tion in an adverse direction, could result in degraded protective system function.

The consequences could yield results which are more limiting than those presented in the plant Safety Analysis Report.

DESIGNATION OF APPARENT CAUSE OF OCCURRENCE:

This occurrence is due to the numerous investigations underway which are examining the licensing bases and operating procedures of nuclear generating stations as a result of the Three Mile Island accident.

ANALYSIS OF OCCURRENCE:

Three of the four potential interactions (pressurizer power operated relief valves control, main feedwater control and rod control) do not require any action at Salem due to either equipment design, manu-fact~r~rs specifications, physical layout, or a combination of the above.

The potential adverse impact from interaction concerning the steam generator power operated relief valves is precluded through procedural changes and operator training.

Based on these considera-tions and the evaluations contained in our letter to Mr. Harold R.

Denton of the NRC dated October 4, 1979, continued operation of Salem Unit No. 1 is justified and no modifications, suspension or revocation of the operating license is warranted.

LER 79-58/0lX-l ANALYSIS OF OCCURRENCE (continued)

Concerning the Steam Generator Power Operated Relief Valve Control System, the particular sequence of events as postulated by Westing-house is not a safety consideration in that the loss of all auxiliary feedwater following a feedline rupture was factored into the Salem Safety Analysis.

Operator action will ~ssure auxiliary feedwater flow to the unaffected steam generators and can be accomplished within the alloted time frame.

The Salem design basis considered operator action in recovery from feedwater/steam line break events.

The indications and controls required for these operator actions are unaffected by the postulated break.

CORRECTIVE ACTION:

The Salem operating procedures and operator training program are being revised to address the concerns addressed in our evaluation of the Steam Generator Power Operated Relief Valve Control System.

FAILURE DATA:

Not Applicable Prepared By A. W. Kapple Manager L Salem Generating Station SORC Meeting No.

82-79

ICFORMl66 I

771 l: CENSEE EVENT REPORT

u. s. NIEAR REGULA1UMY COMMISSION CONTROL ILOCK: l 10 (PLEASE PRINT OR TYPE ALL "EOUIRED INFORMATION!

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75 RE'ORT DATE SO EVENT DESCRl~ION ANO PROBABLE CONSEQUENCES @

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I During the refueling outage, we were notified by Westinghouse that a review of the 2:lIJ I environmental qualification NSSS equipment has shown that conditions associated QJIJ I with high energy line breaks in or outside.containment and their impact on non-

§JI] I safety control systems may constitute an unreviewed safety question.

Evaluations

[1]J I indicate that the postulated sequence of events is not a safety consideration.

III) I This is the secona occurrence of this type (79-52).

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CAUSE DESCRIPTION ANO CORRECTIVE ACTIONS @

[III) I This occurrence is due to investigations conducted as a result of the Three ~ile Island accident.

The Salem operating procedures and operator training are being IIJJJ revised to address concerns addressed in the evaluation of the stea~ generator IIJ]J I power operated relief valve control system.

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OPS~G Public Service E1ectric a'ld Gas Company 80 Par". P:ace N~::;:irr<. N J 07' Cr Pr.on'= 2S...:.:,,:,. 7'f.Jr, October 4, 1979 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Denton:

EVALUATION OF POTENTIAL MALFUNCTIONS DUE TO HIGH ENERGY LINE BREAKS SALEM GENERATING STATION UNIT NO. 1 DOCKET NO. 50-272 In response to your letter of September 14, 1979, regarding high-energy line breaks and their potential impact on non-safety grade control systems reported in our letter to Mr. Boyce H. Grier*

dated September 10, 1979, we.attach the results of our evaluation.

In summary, three of the four potential interactions (pressurizer power operated relief valves control, main feedwater control and rod control) do not require any action at Salem due to either equipment design, manufacturer's specifications, physical layout, or a combination of the above.

The potential adverse impact fro~

interaction concerning the steam generator power operated relief valves is precluded through procedural changes and operator training as described in the evaluation.

As a result of the Three Mile Island accident, there are a signi-ficant number of industry, governmental and regulatory investi-gations underway which are examining the licensing bases and the operating procedures of nuclear generating facilities.

The requirements contained in NUREG 0578 for additional analyses (reference Section 3.2 Page 17 and Page A-45) encompass potential scenarios consistent with those identified by Westinghouse and made the subject of IE Information Notice 79-22.

We, therefore, believe that any additional action you may require as a result of IE, Information Notice 79-22 is consistent with the requirements of NUREG 0578.

Harold R. Denton, Director 10-4-79 Based on these considerations and our attached evaluation, con-tinued operation of Salem Unit No. 1 is justified and no modi-fication, suspension or revocation of the operating license~is warranted.

This submittal contains three (3) signed originals and forty (40) copies.

Very truly yours, l

I.

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Frank P. Librizzi General Manager -

Electric Production

U.S. NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-272

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PUBLIC SERVICE ELECTRIC AND GAS COMPANY FACILITY OPERATING LICENSE NO. DPR-70 NO. 1 UNIT SALEM GENERATING STATION Public Service Electric and Gas Company hereby submits its Evaluation of Potential Malfunctions Due To High Energy Line Breaks.

This evalua-tion has been performed in response to a letter sent from Mr. Harold R. Denton to Mr. F. P. Librizzi on September 14, 1979, and contains statements by which the NRC staff can determine that the Salem Gen-erating Station Unit No. 1 Operating License (DPR-70) should not be modified, suspended, or revoked.

Respectfully submitted, PUBLIC SERVICE ELECTRIC AND GAS C0~2A~Y By: ~tJ~

VICE PRESIDENT

STATE OF NEW JERSEY)

)

COUNTY OF ESSEX

)

SS:

FREDERICK W. SCHNEIDER, being duly sworn according to law de?oses end *says:

I am a Vice President of Public Service Electric and Gas Cowpa~y, and as such, I signed the Evaluation of Potential Malfunctio~s D~e To High Energy Line Breaks.

The matters set forth in said evaluation are true to the best of my knowledge, information, and belief.

~.;i J..k,L.;J,_

FREDERICK W.

SCli~EIDER Subscribed and sworn to before me this c_

I 1979.

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Notary ?ublic of New Jersey My Commi~s~on expires on

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EVALUATION OF POTENTIAL MA.LFUNCTIONS DUE TO HIGH ENERGY LINE BREAKS Introduction Westinghouse has notified PSE&G and its other utility customers that potential control system malfunctions resulting from high energy line pipe breaks and interacting with protective functions could result in consequences more limiting than those presented in existing safety analyses. Such control syste!i:

failures could result from the adverse envirorrnent created by such a pipe break.

The Westinghouse analysis of these events has identified four specific potential interactions which require utility review for applicability at their plants. The control systems associated with the potential interactions are the following:

Steam Generator Power Operated Relief Valves Control Syste~

Pressurizer Power Operated Relief Valves Control System Main Feedwater Control System Rod Control System lmpl icit in the four potential interaction scenarios identified by Westinghouse are worst case assumptions concerning the particular pipe break, its size and location, and the type and extent of consequential failures in control systems induced by the adverse envirorunent.

Postulated pipe break locations and control system malfunctions other than those specified would not res~lt in consequences more limiting than those already presented in the Safety Analysis Report. These assumptions are in addition to the already conservative set of assumptions ascribed to the analysis of design basis events in the Safety Analysis Report.

The particular potential interaction scenarios represent a significantly less probable subset of design basis events.

Each particular potential interaction, the applicability to Salem and any cor-rec:;ve action that may be required are described as follows:

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EVALUATION OF POTENTIAL MALFUNCTIONS DUE TO HIGH ENERGY LINE BREAKS Item No. 1 - Steam Generator Power Operated Relief Valve Control System Potential Interaction Identified by Westinghouse A feectwater line rupture occurring outside the containment in either the rr.ain or auxiliary feedwater lines between the containment penetration and check valve could have an adverse effect on the steam generator power operated relief valve control system.

The adverse environment resulting from the feedwater line rupture could impact the instrumentation and equipment which controls these valves. The potential malfunction could result in the inadvertent opening of the power operated relief valves or cause them to remain open.

This condition would cause the intact steam generators to depressurize and result in loss of the turbine driven auxiliary feedwater pump.

With an assumed single active failure of a motor driven auxiliary feedwater pump, flOn' of auxiliary feed-water can be lost to all steam generators for a period of time.

The r~ainir.g operable motor driven auxiliary feedwater pump would be injecting water into the ruptured feedwater line. The severity of this postulated event is dependent on timely operator action but could cause results more limiting than those presented in FSAR analyses.

Analysis/Safety Implications The Salem plant is provided with two penetration areas outside the con~in~er.t where this postulated feedwater line rupture could occur (refer to attached figures).

Each penetration area is utilized for routing of steam and feedwater piping for two steam generators. The penetration areas are physically separatej so that a break in one will not affect the other. Therefore this particular brr:k can only affect the steam generator power operated relief valve control systems within the particular penetration area where the break is postulated.

The turbine driven auxiliary feedwater pump is supplied steam from No. 11 and 13 steamlines which are located within the same penetration area (refer to

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Item 1 (Cont'd) attached figures). A rupture in No. 11 Steam Generator feedwater line will result in the loss of one steam feed for the turbine driven pump.

The power operated relief valve for No. 13 Steam Generator (13MS10) employs a pneumatic control system which uses steam pressure as a control parameter. A control signal from No. 13 Steam Generator pressure transmitter (not safety-related) operates the valve through an electric/pneumatic converter. The pressure instrument, electric/pneumatic converter and power operated relief valve are located in the penetration area which is subject to the adverse environment resulting from the feeO,...ater line break.

This is control grade equipment and as such was not qualified for post accident operation and therefore could be subject to misoperation.

The potential malfunction could be the opening of valve 13t1Sl0 and depressuri.zation of No. 13 Steam Generator with the eventual loss of the remaining turbine driven auxiliary feedwater pump steam supply.

The two motor driven auxiliary feedwater pumps are aligned to provide flow to all steam generators (each pump is capable of supplying adequate flow to two steam generators). With an assumed single failure of one motor driven pump, the remaining pump would be supplying the ruptured steam generator feedwater line.

This particular sequence of events res*u1 ts in the loss of all auxiliary feed-water flow to the intact steam generators.

The operator must respond by iso-lating the ruptured steam generator and providing flow to the two unaffected steam generators.

A similar analysis would result with the postulated break in No. 13 Steam Generator feedwater line.

In the analysis of feedwater line breaks presented in the Salem FSAR section 14.3, auxiliary feedwater was assumed to initiate 10 minutes following the incident with an additional 6 minutes assumed before auxiliary feedwater enters the µnaffected steam generators.

No credit has been taken in the Salem Safety Analysis for auxiliary feedwater operation in this time period (an effective loss of the system).

In addition, Westinghouse WCAP 9600, "Report on Small Accidents for Westinghouse NSSS Sys terns 11 des er ibes transient analyses for

e.

Item 1 (Cont'd) postulated loss of all main and _auxiliary feedwater {no pipe rupture).

The results indicated that the operator has at least 4000 seconds following the loss of feedwater to assure auxiliary feedwater flow to the steam gene~ators before the core begins uncovering.

The potential deficiency described above.

is similar to that presented in Section 4.2 of WCAP-9600 without the additional assumption of a feed 1 ine rupture outside the containment between the check valve and penetration. Conservatively assuming no heat removal in the steam generator liquid inventory lost through the feedwater line break, Westinghouse calculations have shown that the operator would still have at least 2808 seconds to take corrective action to assure auxiliary feedwater flow to the unaffected steam generators.

The operating procedures require that the affected steam generator be isolated to assure auxiliary feedwater flow to the remaining steam generators.

The operator i~ provided with adequate information and controls to accomplish these actions. The steam generator pressure instruments utilized for post accident monitoring (3 for each line) are located in the affected environment.

These instruments have undergone envirorvnental qualification testing for post accident operation.

The auxiliary feedwater valves which must be closed to assure flow to the unaffected steam generators are located in the auxiliary building and are not subject to the adverse environment. Other parameters required for post accident recovery such as reactor coolant system parameters and steam generator levels are not located in the potential break area.

The operators can respond in the time frame prescribed.

In this particular s~quence of events postulated, the closure of the auxiliary feedwater valves to No. 11 and 13 Steam Generators would result in the operable motor driven auxiliary feedwater pump providing flow to No. 12 and 14 Steam Generators and thereby assuring safe plant shutdown.

Westinghouse has per-fonned analyses indicating that two steam generators are sufficient for safe shutdown.

A postulated break in No. 12 or 14 Steam Generator feedwater lines in the same location will not result in the loss of the turbine driven auxiliary feedwater pump but could cause the other steam generator to depressurize because of potential misoperation of the power operated relief valve.

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Postulated main feedwater line breaks in locations other than those specific-a 1 ly addressed would not result in a sequence of even ts to cause this..

potential interaction. A break inside the containment would not create an adverse environment in the area of the s~eam generator power operated relief valves. A break in the penetration area upstream of the main feedwater check valves could cause an adverse environment for a potential misoperation of the steam generator power operated relief valves. This could cause the loss of the turbine driven pump if the break was postulated in the inboard penetration area, however, this break would not result in the operable auxiliary feedwater pump delivering flow to the pipe break.

Recovery would proceed in accordance with operating procedures.

Corrective Action This particular sequence of events as postulated by Westinghouse is not a safety consideration in that the loss of all auxiliary feedwater following a feedline rupture was factored into the Salem Safety Analysis. Operator action wi 1 l assure aux i 1 ia ry feedwa ter -fl ow to the unaffected steam genera tors and can be accomplished within the alloted time frame.

The Salem design basis considered operator action in recovery from feedwater/steam line break events.

The indications ahd controls required for these operator actions are un-affected by the postulated break.

The Salem operating procedures and operator training program are being revised to address the concerns described herein.

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EVALUATION OF POTENTIAL MALFUNCTIONS DUE TO HIGH ENERGY LINE BREAKS

• Item No. 2 - Pressurizer Power Operated Relief Valve Control System Potential Interaction Identified by Westinghouse A main feedwater line rupture occurring inside the containment could have an adverse impact on the pressurizer power operated relief valve control syste~.

The adverse envirorvnent resulting from the feedwater line rupture could im-pact the instrumentation and equipment which controls the relief valves.

The potential malfunction could result in inadvertent opening of the power operated relief valves or cause them to remain open. This would result in decreasing primary system pressure with eventual hot leg boiling dependent on the size of the feedwater line rupture.

This particular accident and sequence of events could lead to results more limiting than those presented in FSAR analyses.

Analysis/Safety Implications

. The pressurizer power operated relief valve control system at Salem is pneumatic with solenoid valves for open/close control. The solenoids are operated by electrical signals from the pressurizer control system which utilizes pressurizer pressure as the control parameter.

The pressurizer pres-sure instrumentation and the solenoid valves for the power operated relief valves were designed and qualified for post-accident operation within the con-tainment.

In addition the cabling and electrical connections for the solenoid valves and pressure transmitters have been designed and qualified for post accident operation.

The environmental qualification of instrumentation, control and electrical equipment located within the containment has been dis-cus!c1 in our response to Bulletin 79-01.

A feec:Mater line break outside of the containment cannot cause an adverse envirorvnent which could affect the pressurizer power operated relief va 1 ves.

Item No. 2 {Cont'd)

The pressurizer power operated relief valves control system has been environ-mentally qualified for post accident operation within the containment.

Potential inadvertent operation of the relief valves as postulated by Westing-house will not occur at Salem due to the design characteristics of the syste~.

Corrective Action None required.

Operator training already stresses the importance of maintaining reactor coolant system pressure above saturated pressure to prevent core boiling as a result of an open pressurizer power operated relief valve.

I

EVALUATION OF POTENTIAL MALFUNCTIONS DUE TO HIGH ENERGY LINE BREAKS Item No. 3 - Main Feedwater Control System Potential Interaction Identified by Westinghouse A small feedwater line rupture occurring outside the containment in either the main or auxiliary feedwater lines between the containment penetration and check valve could have an adverse effect on the main feedwater control syst~..

The adverse enviro11T1ent resulting from the feedwater line rupture could impact the instrumentation and equipment which controls the main feedwater syste~.

The potential malfunction could result in a condition such that all stea~

generators are at low-low water level at the time of reactor trip. This particular accident and ensuing conditions could lead to results more limiting than those presented in FSAR analyses.

Analysis/Safety Implications The postulated high energy line break for this potential interaction occurs in the mechanical penetration areas at the Salem plant (refer to attached figures).

The main feedwater control system employs a three element control scheme.

Steam generator level, steam flow and feedwater flow parameters are utilized for feedwater regulation.

The main feedwater flow instrumentation and the main feedwater control valves are located in the Turbine Building which is not affected by the adverse environment from the postulated feedwater line rupture.

The steam generator level and steam flow instrumentation is located in the containment which is also not subject to the adverse environ-ment resulting from the break.

In addition, the cabling for this equip~en~

1s not routed through the postulated break area.

A feedwater line break inside of the containment will not cause a potential misoperation of the main feedwater control system.

The steam generator level and steam flow instrumentation and associated cabling located inside the con-tainment has been qualified for post accident operation.

The remaining it~s

I tern 3 (Cont'd )

which are part of the feedwater control system are not located in the area subject to the adverse envirol1Tlent from the postulated pipe break. A main feed-water line break in the penetration area upstream of the feedwater che~k valves will not result in a plant condition where a potential feedwater control system failure could cause consequences more 1 imi ting than ex_i sting safety ana lyse.s.

The physical location and design of the main feedwater control system equip-ment and cabling at Salem precludes the possibility of occurrence of this potential interaction scenario as postulated by Westinghouse.

Corrective Action None required.

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EVALUATION OF POTENTIAL MALFUNCTIONS DUE TO HIGH ENERGY LINE BREAKS Item No. 4 - Rod Control System Potential Interaction Identified by Westinghouse An intermediate steamline rupture of approximately 0.1 to 0.25 square feet occurring inside the containment with the unit at a power level of 70-100~

could have an adverse effect on the rod control system.

The adverse en-virorvnent resulting from the steamline rupture could impact the excore detectors, connectors and associated cabling causing an erroneous signal to the rod control system.

With the rod control systen in the automatic control mode, this potential environmentally induced failure could cause the control rods t~ be~in stepping out prior to re~ctor trip. A reactor trip occurs for an intermediate steamline rupture upon overpower AT in a time frame of one to two minutes following the break (WCAP 9226). This particular sequence of events would result in a minimum DNBR below 1.30 which exceeds existing lice~sing criteria and FSAR analyses.

Analysis/Safety Implications The power range excore detectors at Salem are physically located in detector wells outside the reactor vessel on centerline with the reactor core (refer to attached figures). The location of the detectors preclude their i1Tr.1ediate exposure to an adverse environment resulting from a main steam line break.

The detector well is situated in such a fashion as to isolate it from the main containment environment.

The equipment and concrete surrounding the detector wells will act as a thennal barrier to limit immediate environmental impact.

The normal operating environment of the detectors is maintained at less than 13~~F. The manufacturer's specification for the detectors indicates a maximum temperature range of 300°F with a pressure of 180 psig and 100: relative humidity.

The cabling within the detector well is a mineral filled coaxial

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,cable hennetical ly sealed at the detector ionization chamber.

The cab1 ing from the detector well to the electrical penetrations was designed and specified by the manufacturer to survive post accident conditions. Tne electrical connections at the detector well and penetrations are via Amphenol Triaxial Connectors which are protected from the moisture effects of the envirorvnent by a Raychem heat shrink tube.

An analysis of containment environments under steamline break accident con-ditions was presented in the FSAR in response to Question 5. 82.

The series of breaks analyzed indicated that the peak ambient containment temperature is under 3S0°F and occurs in the time frame of one to two minutes.

A steam line break outside of the containment cannot cause an adverse environ-ment which could affect the excore detectors or the rod control system.

The design characteristics of the detectors and cabling and their location within the containment indicate that it is highly improbable that a main steamline break would impair their function within two minutes following the break to cause a potentia 1 mi sopera ti on of the rod contro.1 system prior to reactor trip. The potential interaction scenario as postulated by Westinghouse does not appear to be credible at Salem.

Corrective Action i~one required.

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