ML20094E565

Forwards Resolutions for Environ Qualification of safety- Related Electrical Equipment & Justification for Continued Operation for Any Equipment for Which Documentation for Environ Qualification Not Yet Completed

ML20094E565
Person / Time
Site:	Pilgrim
Issue date:	08/03/1984
From:	Harrington W BOSTON EDISON CO.
To:	Vassallo D Office of Nuclear Reactor Regulation
References
84.119, NUDOCS 8408090248
Download: ML20094E565 (82)

Text

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I BOSTON EDISON COMPANY 800 BOYLBTON STREET BOSTON, MASDACHUSCTTs 02199 wtLLtAM D. HARRlNGTON August 3, 1984 c======="*="*""

BECo Ltr. No.

84.119 Mr. Domenic B. Vassallo, chief Operating Reactors Branch #2 Division of Licensing Office Of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission License No. DPR - 35 Docket No. 50-293

SUBJECT:

Environmental Qualification of Safety-Related Electrical Eauipment at Pilgrim Nuclear Power Station

REFERENCES:

1) BEco Letter No.84-099 dated 7/9/84, W. D. Harrington to D. B. Vassallo

2) Meeting between BECo and the NRC on May 22, 1984

Dear Sir:

Reference (1) provided you with Boston Edison's resolution for each of the equipment covered under the Technical Evaluation Report (TER) written by Franklin Research Center.

For the equipment items that have been added to the

" Master List of Electrical Equipment" and not factored'in the TER resolution process, Boston Edison stated that it would submit a response with resolution and applicable JCO's on August 3,1984. to this letter describes the current resolution for each of the added equipment, in a matrix format. provides you with a justification for continued operation (JCO) for any equipment for which the documentation for environmental qualification is not yet completed. Enclosure 2 also includes JCO's for some TER items which were not included in Enclosure 2 of Reference 1.

Your Staff has requested additional information regarding compliance with four areas in 10CFR 50.49. These four areas are:

(1)

JCO's; (2) 50.49 (b) (2) review Methodology (3)

DBE's including flooding outside containment; and (4)

Regulatory Guide 1.97 equipment Boston Edison has stated its approach with respect to these four areas in Reference (1) and reiterates the following to respond to the request for additional information.

8408090248 840803 i

PDR ADOCK 05000293 P

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30sTON E!b!N COMPANY TAugust 3,~1984-BECo Ltr.-No. 84.119

-(1).JCO's Based on:the JCO's submitted in Enclosure 2 of Reference 1 and in

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' Enclosure 2 of this letter Boston Edison states that no significant

degradation of required safety functions is expected to occur nor is

. operator confusion expected to inhibit the accomplishment of required

safety; functions due to failure of equipment under design basis accident Lenvironments.

T(2)l50.49 (b) (2) Review Methodoloav In performing its review of the methodology to identify equipment within

.the scope of 10CFR 50.49 (b) (2), Boston Edison has performed a series of (studies in response to I&E Information Notice 79-22, I.E.Bulletin 79-27 iand la review of associated circuits performed under the auspices of

, Appendix R.

,The findings of these evaluations conclude that there is no

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equipment identified under 50.49 (b) (2). Nevertheless, Boston Edison intends to verify this assessment using the guidelines provided by your

, staff. Any deviation identified as a result of this assessment will be

factored into the BEco Environmental Qualification Program and reported in accordance with PNPS Technical Specification reporting procedures.

(3)-DBEs includina floodina outside containment As part of the offort in identifying the Master List of equipment within

,the scope of 10CFR 50.49 (b) (1),' Boston Edison has reviewed all postulated design basis events documented in the FSAR including a loss Of Coolant Accident (LOCA) inside containment and High Energy Line Break

'(HELB) outside containment, including flooding outside containment.

~(4) Reaulatory Guide 1.97 equipment

The equipment within the scope of 10CFR 50.49 (b) (3) is all R. G. 1.97 Category 1 and 2 equipment and will be identified in the R. G.1.97

. submittal. After staff review and acceptance of R. G. 1.97 submittal, Boston Edison will add the applicable equipment to the E. Q. Master Equipment List and will implement its schedule of R. G. 1.97 activitics e

including environmental qualification.

As stated in Reference 1, it is requested that supplemental SER's be

issued incorporating your review of Reference 1 as well as this submittal. We would be pleased.to answer any questions you may have regarding this submittal

.or the information contained in Reference 1.

Very truly yours, m

W. D. Harrington : Resolution Matrix for new equipment. : Justification for Continued Operation.

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. PILGRI~' NUCLEAR POWER STATION - S.E.R. DFFICIENCY RESOLUTIONS' EQUIPMENT' TYPE TER TER #

MANUFACTURER /MODEL #.

DEFICIENCY' RESOLUTION..

1, 2, 3, 4b, 4c, 9, 14,.16,

. Motor.. Operator Aging. degradation, Inspection and replace compo--

17, 18, 19,'22a, 23, 24, 25, Limitorque/SMB qualified life nent parts:with qualified 26, 27, 28, 29, 30, 31, 32, parts '

33, 35, 36.-37, 38, 39, 40, 41

-Aging degradatton 4a. 5, 6, 10, 11, 12, 13, 15, Motor Operator Qualified life.

Replace with qualified motor 20,-21, 22b, 34 Limitorque/SMB Similarity operator'- Limitorque Radiation 7, 8, 97, 256, 258, Standby Gas Treatment System Inadequate documentation Design modification to 260, 261 Damper--Honeywell/M940A10671 establish qualification Humidity Detectors Honeywel1/R7088C Honeywel1/Q464A Temp. Switch:.

-Fenwall/40102010115 Transformer-GE/9T55Y46G7

' Contractor--Allen Bradley/

702LT0093 259, 262 Standby Gas Treatment-System Inadequate Documentation Replace with qualified cable--

Cable - Bronco 66 Vulkene Supreme or equivalent 95 Standby Gas Treatment System None Qualified Report 47066-HT-1 Heater - Chromalox/64-47499 453, 45f, 50, 53,

. Solenoid Valves Qualified life Qualified: Test Report 55, 56, 58, 59, 60, 61, 62a, ASC0/NP8320A184E AQS21678/TR Quallfled life 62b, 62c, 62d, 64, 65, 66, determined by Analysis Report 67, 70, 73, 74, 75, 77, 78a, 47066-SOV-2.

78b, 78c, 78d, 79, 82 85, 86 Solenoid Operator _. _

Similarity Negotiating to join testing AVC0/C5159 Quallfled life program already in progress Functional testing Est. completion date 1/85 49, 48 Solenoid Valves Inadequate documentation Replace with qualified ASC0/HVA-90-405-2A Aging degradation solenoid valves--ASCO NP8316

/WP-LB-831636 Qualtfled life 1

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MANUFACTURER /MODEL #

DEFICIENCY RESOLUTION 87, 91, 93, 94 Motors Inadequate documentation ~

Qualified: Test Report GE SK6339XC87A

-G-HK-0-16 Analysis Report 5K254AK299HIA 47066-MOT-3.1 Motor.

SK6337XC93A Terminations - Evaluate /:

SKl84AL217 Replace 54, 57a, 57b, 57c, 57d, 72 Solenoid Valves Qualified life Qualified: Test Reports Valcor/V526529231 QRS2600-5940-2 QR52600-515 Qualified life established in analysis report 47066-50V-8 81 Solenoid Operator Inadequate documentation Qualified:

Test Repoit 2199A; Target Rock /1/2SMSA01 Analysis Report 47066-SOV-6 l

233, 234, 235, 236, 237, 238, Cable None Qualified:

Test Report 239a/b/c, 240, 241a, 242 Kerite/FR/FR, HT/FR, HT/NS 17446-2 and Analysis Report 47066-CAB-3 268, 269, 109, Indicating Light Exempt No active safety-related 107alb/c/e, 108a/c GE/ET-16 function. Components will be Switch tested or replaced, when GE/CR-2940 qualified replacement items Relays:

are determined.

Johnson /SER KZ4000B Agastat/2412AN 117 Cable Inadequate documentation Qualified:

Test Rpts 2806, Rockbestos/Firewall III QR-1806, 110-11516, F-C-3798, F-C-5022-2 and Analysis Report 47066-CAB-5

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EQUIPMENT ITYPE TER-TER #

MAIRAFACTURER/MODEL #

DEFICIENCY' RESOLUTION I

l 243, 244,'245, 246, 247, 248 Cable None Qualifted:. Test: Report Okontte/Okolon-& Okoprene NORN-1 110, 111,.112, 118, 119 Instrument Rack Wiring from

. Inadequate documentation Replace with quallfled 120, 121, 122, 123, 124, J. B. to devices equipment. Vulkene Supreme or equivalent Replace some terminations with.

100 Ring Tongue Terminations Inadequate documentation qualified splices (Raychem Less Than 4KV in the Drywell WCSF-N). Where ring-tongues have been tested, verify installation adequacy.

252 Cable Inadequate documentation Test program to be initiated Electrical / Distribution 9/84 with completion expected Type S1 by 3/85 265b/d/e/f, 267a/c Terminal Block None Qualified:

Test Report GE/EB-25 QSR-010-A-01 & B0119 Inadequate documentation Design modification to enclose 88, 89, 90 Motor Control Centers Aging degradation MCC's eliminating humidity, Cutler Hammer /6AF685046 Quallfled life temperature and pressure Nelson Electric /1035E Similarity effects. Analysis to address l

Radiation radiation in prooress Test sequence 92 Motor Inadequate documentation Replace with qualifted motors-Louis A111s/ COG 4B Hestinghouse motors purchased from Buffalo-Forge using the l

D0-146F Qual. Report.

I 99 4KV Terminations Kerite Inadequate documentation Quallfled:

Test Reports F-C-4020-1 & F-C-4020-2.

Qualified life evaluation.

To be complete by 9/84

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PILGRIM IRACLEAR PONER STATION - S.E.RG DEFICIENCY RESOLUTIONS.2-

. EQUIPMENT. TYPE TER l~

TER #

MANUFACTURER /MODEL #.

DEFICIENCY-RESOUJTION :

101, 102 Spilces.

Aging degradation-.

Qualifiedil;TestReport Raychem/HCSF-N Qualified life-58442-1,-Qualified life.__

Analysis Report 47066-SPL-1.1!

103, 104a, 104f, 104g, 104h, Terminal Block.s Inadequate documentation,

' Design modification to delete 104I Buchanan /525 siellarity terminal blocks - Replace.with.

qualified splice (Raychem HCSF-N) 210, 211, 212, 213,-214, Level Switch Inadequate documentation Replace component parts with-l 226, 227 Yarway/4418C & 4418EC qualified component parts (Yarway Kit #959552) 98, 263 Accelerometer Inadequate documentation Qualified

Test Report TEC/ND Quallfled life 517-TR-03, Analysis Report 47066-MON-2 l

220, 221 Transmitter-Inadequate documentation Replace with qualified GE/555 transmitter - Rosemount 1153 Transmitter 232 Level Switch Inadequate documentation Required for radiation Robertshaw /SL702A1 only--pending vendors

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127, 129a, 129b, 129c, 129d, Electrical Penetrations Inadequate documentation, Qualified: GE prototype 128h,128j GE/238X60NLG similarity Test Report - Analysis Report-47066-PEN-1 132, 137 Radiation Detector Inadequate documentation Qualified: Test Report GE/237X731G009 943-81-003 and analysis report 47066-RAD-2

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5 of 17-PILGRIM NUCLEAR POWER S'TATION - S.E.R; DEFICIEEY RESOLUTIONS -

EQUIPMENT TYPE

'TER TER #

MANUFACTURER /MODEL #

DEFICIENCY.

RESOLUTION 223 Transmitter Aging degradation

. Qualified:

Test Report-Rosemount/1152 Qualified life 117415 Rev. 8, Analysis Report 47066-PT-1 establishes qualified life.

Installing Conax ECSA Conduit Seal.

Aging degradation 139, 140, 142, 144, 143, 145, Temperature Switch Pressure Qualified by existing Test 147, 159, 160, 161, 162, 163, Fenwall/17023 & 17002 Steam exposure Report BECo is negotiating to 164, 166 Profile obtain the rights for its use functional testing 171, 174, 175, 177, 178, 179, Pressure & Differential Aging degradation, Qualified: Test Reports; 206, 222 Pressure Switch qualified life, similarity, 145C3008, 145C3009, R3-288a-1.

Barton/288, 288a, 289a temperature, pressure, Analysis Report 47066-PS-2 radiation 173, 176, 180 Pressure & Diff. Pressure Aging degradation, Replace with qualified Switches qualified life, similarity, equipment. Static-0-Rings Barton 288, 288a 289a temperature, pressure, radiation Inadequate documentation Test Report:

30203-2.

189, 190, 191, 192, 193, 197, Pressure Switch Aging degradation Completion pending vendor's 198, 202, 203, 204, 205 Static-0-Ring /12N Pressure material list l

Radiation Inadequate documentation Replace with qualified 181, 182, 208, 209 Pressure Switch Aging degration equipment.

Static-0-Ring Static-0-Ring /5N Temperature Modei NO. 6N6.

Pressure 1

183, 186, 187, 188, 199, 200, Pressure Switch Inadequate documentation Qualified:

Test Reports 201 Barksdale/B2T 596-0398 & 15566-23 and Analysis Report 47066-PS-3 Inadequate documentation Replace with qualified 194, 196, 207 Pressure Switch Quallfled life Equipment:

Static-0-Ring.

Barksdale/82T, D2H, P1H Steam exposure (profile) j l

Radiation

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PILGRIM NUCLEAR POWER STATION - S.E.R. DEFICIENCY RESOLUTIONS T

W EQUIPMENT TYPE TER TER #

MANUFACTURER /MODEL #

DEFICIENCY-RESOLUTION'-

195 Pressure Switch Inadequate documentation Replace with qualified pres---

Mercold/DA23804 sure switch.

Static-0-Ring Model 4N6.

146 Temperature Element Inadequate documentation Replace with qualified equip-Thermo Electric /3544710 ment. Need RTD's Model No.SP-612D.

42, 152, 153, 154, 155, 156, HPCI Turbine Controls Inadequate documentation Radiation only.

Plant 157, 158, 185 Various Eqttipment modification to address radiation l

172 Pressure Switch Inadequate documentation Test Report R3-288a-l.

Barton/278 Replace component parts with l

qualified parts.

Barton 288A l

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249 Cable Inadequate documentation, Qualified: Test Report GE/Vulkene supreme similarity FC-4497-2 Analysis Report l

47066-CAB-1.1 l

250 Cable Qualified Test planned:

To be initiated GE/Vulkene SIS by 9/84 planned with completion by 3/85 251 Cable None Qualified Test Report BIW/Bostrad B901A

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PILGRIM NUCLEAR POWER STATION - S.E.R.~ DEFICIENCY RESOLUTIONS I EQUIPMENT ~ TYPE TER

~

.TER #

. MANUFACTURER /MODEL #

DEFICIENCY-

' RESOLUTION-254, 255 Limit Switch

. Similarity Replace with qualified Namco/EA740' equipment.' hAMCO EA740 with EC210 Connector Assembly.

Test Reports: 2392-2, 264b/d/e/f, 266a/c :

Swltch Inadequate 'docu:nentation 2392-14, 3030-1 Electro Switch /24/40 Switches-will be tested or-replaced when qualified replacements are determined 270 Cable Inadequate documentation, Qualified:

Test Repo.-ts:

GE/Vulkene SIS-similarity 43905-2 & EPAQ-047 271, 272, 273, 274, 275, 276, Terminal Bloc'k'.

None Qualified: Test Reports:

277, 278, 279, 280 GE/CR-151 GEN-8-18 & B0119 43 Solenoid Valve-Exempt Radiation only - completion Atkomatic/247214 pending vendor's material list

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. PILGRIM NUCLEAR POWER STATION - S.E.R.' DEFICIENCY RESOLUTIONS.

yq EQUIPMENT TYPE TER.

TER #

MANUFACTURER /MODEL #

.: DEFICIENCY

RESOLUTION.

44 Solenoid Valve Exempt.

Out of Scope of--10CFR50.49 Atkomatic/247214 i

45b, 45c, 45d, 45e, 459, 45h Solenoid Valve Qualified life Out of scope of 10CFR50.49 l

451, 46, 47, 62e, 78e, ASC0/NP8320A184E l

78f, 80, 83, 84, 282 51 Solenoid Valve Inadequate documentation Out of scope of 10CFR50.49 i

.ASC0/HVA90405 52, 57E, 57F, 57G, 57H Solenoid Valve Qualified life Out of scope of 10CFR50.49 Valcor/V5265683

/V526529231 63, 68, 69 71 Solenoid Valve Qualified life Out of scope of 10CFR50.49 Valvor/V526529212 l

76 Solenoid Valve Inadequate documentation Out of scope of 10CFR50.49 ASC0/HT8210C22-104b, 104c, 104d, 104e, Terminal Block Inadequate documentation Out of scope of 10CFR50.49 Buchanan /525 similarity l'25 Electrical Penetration Inadequate documentation Out of scope of 10CFR50.49

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Conax/ Modular Type similarity 224, 225 Level Switch Inadequate documentation Out of scope of 10CFR50.49 Yarway/4418EC 281 Switch Inadequate documentation Out of scope of 10CFR50.49 Electro Switch /24/40 107f/d, 108b Indicating Light Exempt Out of scope of 10CFR50.49 GE/ET-16

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PILGRIM IRJCLEAR PONER STATION u S E R; DEFICIENCY RESOLUTIONS EQUIPMENT. TYPE.

TER-TER #-

MANUFACTURER /MODEL #:

. DEFICIENCY-

. RESOLUTION 136 Transmitter Aging Degradation Out ofiscope of 10CFR50.49.

Rosemount/1152 e -

Qualifled-Life _

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. Electrical Swltch Inadequate Documentation Oui of scopefof5 0CFR50.'49

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Electro-Switch /24/40

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7 265a/c,~267b Terminal Block None

' Out of scope of 10CFR50.49 GE/EB-25 OuthfScopeof10CFR54.49

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2390 Cable

. None Kerite/HT-FR&HT-NS s'

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EQUIPMENT TYPE TER MANUFACTURER /MODEL #

DEFICIENCY l

RESOLUTION 126 Electrical Penetrstion Documentation Not Out of scope of 10CFR50.49 Physical Science, Canister Available Type 129e, 128alb/c/d/ elf /h/1 Electrical Penetration Inadequate documentation Out of scope of 10CFR50.49 GE/238X60NLG similarity 130 Pressure Switch Inadequate documentation Out of scope of 10CFR50.49 Heletron/92416SS5A 131, 133, 134, 168, 169, Trans.ni t ter Inadequate documentation Out of scope of 10CFR50.49 216, 217 GE/551 or exempt 138 Transmitter Inadequate documentation Out of scope of 10CFR50.49 Foxboro/611DM 148 Limit Switch Similarity Out of scope of 10CFR50.49 NAMC0/EA740 149 Limit Switch Inadequate documentation Out of scope of 10CFR50.49 NAMC0/01200G2 151 Fuse Panel Exempt Out of scope of 10CFR50.49 GE/238X278G1 l

l 165 Electric Heater Inadequate documentation Out of scope of 10CFR50.49 l

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PILGRIM NUCLEAR ~ POWER STATION iS.E.Rf DEFICIENCY RESOLUTIONS

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DEFICIENCY -

RESOLUTION' F

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. Level Swltch, 7 nadequate' documentation; I

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-228 Out ~of scope of -1.0CFR50.49-h McDonnel/63SY

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Exempt 0ut of scope of 10CFRSO.49 1229, 230,-231'

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'141 Thermostat f

Inadequate documentation-- I Out of scope,of 10CFR50.49-Johnson Controls;.

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' 184 Pressure Sw)tch-m Inadequate. documentation Out of scope of 10CFR50.49 Hercold/AP7021"15

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jr-135 Temperature Element Inadequate. ' documentation Out of scope of 10CFR50.49 l

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150 Hydrogen Analyzer Aging degradation Out of scope of 10CFR50.49 f

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114, 11-5, 116 Cable Inadequate documentation Out of scope of 10CFR50.49

- RockbestosIFirewall III 257 Temperature Switch Inadequate documentation Out o,f scope.of 10CFR50.49 Fen, wal1/180230 i

' 253, 113 Indicating Ltght Exempt Out-of scope of 10CFR50.49 GE/ET-16 Terminal Block GE/EB l

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. PILGRIM NUCLEAR PONER STATIONI-15.E.R.. DEFICIENCY.iRESOLUTIONS

~

- PNPS.ID #

D Equipment _ Type

' Resolution

' Manufacturer /Model SV0S-1/2/3/4 :

Limit Switch' Quallfled:. Radiation Only ' Analysis Report-

.47066-LSH-3 NAMC0/SL3-LS302 - 82A/B/C/D Level Switch Replace with qualified equipment. System com-ROBERTSHAN/SL-305-E7X prised of Rosemount 1153 Transmitter and Fluid-Component Inc. FR72-4HTRDLL Heated RTD.

Transmitter Replace with Quallfled equipment, Rosemount 1153 FT1049A/B GE/555 Transmitter.

M01301-62, M01301-49 M01001-19, M02301-6, Motor. Operator-Inspect and replace component parts with M04065, M04009A/B, Limitorque/SMB quallfled parts.

M04083, M03806, M03805 PS4008 Pressure Switch Evaluation in progress Barton/288A l

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' PILGRIM NUCLEAR POWER-STATION - 5,E.R DEFICIENCY RESOLUTIONS'

PNPS

ID:#:

Equipment l Type' Resolution LManufacturer/Model'f PS504A/B/C/D'

-Pressure Swltch-

- Qualified:

Test Reports 596-0398 & 15566-23.

PS503A/B/C/D Barksdale/B2T, D1T-Analysis Report 47066 - PS-3 Tip Shear & Ball Valves:

Valves' Analysis in progress for radiation only 1, 2, 3, 4 -

G.E.

requirement.

RE1735 A/B/C/D Radiation Monitor-Qualified-Test Report: QSR-015-A-01 GE/194X927-Analysis Report 47066-RAD-1.1 RE1734A/B/C/D Radiation Monitor Qualified:

Test Report 943-81-003 and analysis GE/237X731 Report 47066-RAD-2.

- B14 Motor Control-Center Plant modification to eliminate accident Nelson' Electric /1035E environment

PILGRIM NUCLEAR-POWER STATION - S.E.R. DEFICIENCY RESOLUTIONS PNPS ID #

Equipment Type Resolutions Manufacturer /Model #

Cable Codes:

Cable Currently being evaluated.

CX2, CX4, 520 Similarity analysis under evaluation.

548, 537, CXG, CX8, Some documentation complete. At the end of the CO2, C03, S19, S27, analysis all undocumented equipment will be Z3, Z3A tested / replaced.

SVL61, SVL82, SVL83 Solenoid Valve Replace with qualified equipment. ASCO NP8320 ASC0/HT8320A107, HT8320A22 P202D/E/F Motor Motor Qualified:

Test Report:

G-HK-0-163.

GE/5K Series Analysis Report 47066-MOT-3.1. Replace Terminations with Quallfled Kerite or Raychem Terminations - Evaluate P202D 512A, 210A, 712A, 912A Cable Quallfled:

Test Report NORN-1 Analysis Report Okonite/0kolon 47066-CAB-4.1 5 f28 Cable Qualified:

Test Report 2806 &

Rockbestos/Firewall III QR-1806 & 110-11516 & F-C-3798 & F-C-5022-2 Analysis Report 47066-CAB-5 B9 Cable Qualified:

Test Report NQRN-1, 17446-2 &

Okonite/Okoprene 47066-CAB-3 Analysis Reports: 47066-CAB-4.1 and or Kerite/FR/FR 47066-CAB-3 412A, 106, 212B, SC16 Cable Currently being evaluated.

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PILGRIM NUCLEAR POWER STATION - S.E.R. DEFICIENCY RESOLUTIONS PNPS ID #

Equipment Type Resolution Manufacturer /Model #

J720, J32, J217, J451, Terminal Block Replace with qualified equipment.

J866, C2257B, C2207A, J599, Unknown Manufacturers J600, J601, J602, J606, J444, J463, J561, J462, J603, J604, J874, J466 J264, J409, J57, J58, J59, Terminal Block Replace with quallfled splice - Raychem HCSF-N.

J300, J204, J205, J206, J207 Unknown Manufacturer J317, J318, J256, J257, J258, Terminal Block Qualified:

Test Report B0119 and Analysis J53, J54, J315, J316, J31, Buchanan /525/222 Report 47066-TB-1.

J33, 334, J51, J52, J562, J552, J456, J177 J859 Cable Splice Evaluation in progress.

Kerite N912, N923, N921, Control Switch Evaluation in progress.

CS42-1821, CS42-1822 GE/CR2940 C61 A & B Indicating Lights Evaluation in progress.

GE/CR2940UC C129A & B, C2257A, C2207B Instrument Rack Hiring and Replace with Qualified wire and terminal blocks Terminal Blocks Unknown Manufacturer J553 Terminal Block Evaluation in progress.

Buchanan /B1XX

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16 f 17 PILGRIM NUCLEAR POWER STATION - S,E.R. DEFICIENCY RESOLUTIONS PNPS ID #

Equipment Type Resolution Manufacturer /Model #

P229 Motor Evaluation in progress.

Baldor PS2390A & B Pressure Switch Evaluation in progress.

Static-0-Ring /6N C118, C119, J411, J412, H /02 Analyzer Replace with qualified equipment. Compsi Delphi 2

3660, 3661, J330, J523, J782, J681, J522, J684, J781, J787, J329 C68A & B, C63A & B Control Panels Plant modification to remove from scope of Various Equipment 10CFR50.49.

C2303, T2303 Local Centrol Panels Under evaluation plant modification to establish Various (quipment qualification under consideration.

C2204, C2222 Local Control Panel Evaluation in progress.

Various Equipment C2201, C2205A, C2260, J538 Instrument Wire Replace with qualified wire.

J539 Manufacturer Unknown J623, J624, J625, J626, C2202 Unknown Contents Walkdown and evaluation in progress

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117.of 17

?PILGRIMNUCLEAR(POWERSTATION-S.E.'R.LDEFICIENCYRESOLUTIONS[

~

):;;

PNPS ID #.

EquipmentjType

" Resolution:

Manufacturer /Model~#:'

.)

-J863

' Instrument' Hire and Terminal Block Replace'with qualified equipment.

Manufacturer Unknown C150 Control Switch & Indicating Lights Evaluation in progress.

Plant modification to Electroswitch Series 40 establish mild environment under' consideration

'GE~ET-16 Lights C2259, C2262,'C2252A Terminal Block Qualified:

Test Report B0119 and Analysis-GE/CR-151 Report 47066-TB-1.

Terminations 4KV Ring Tcng. Connectors Conduct Analysis / Undocumented equipment will be

.Various Manufacturers tested or.be. replaced

E.klC LO SURE -2.

f.ttachment 5 to NEDWI No. 277-p BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. Motor Termination Splices - VAC201 A, 8; VEX 210A, B

c tTER No. 91, 93:

Sheet 1 of I a

Preparer:

l Date:

8 h4

!8Y Independent Review:

e Date:

Approval:

MMw Date:

8 f:l/84

/

\\

'I

The HPCI Area Ceoling (VAC201A', 8).and the Standby Gas Treatment System Fan motor termination splices are tape type motor termination splices with a glyptal outer covering.

These splices are similar to the splices tested in FIRL Report #F-C3056.

~ Temperature and Pressure The test splices were subjected to a steam environment for 71/2 days.

The splices were electrically loaded throughout the exposure period. The peak temperature and pressure were 325'F and 80 PSIG for a duration of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.

The temperature and pressure were reduced to 220*F and 5 PSIG for the remainder. of the test.

The samples were subjected to a chemical spray isolution of 1900 ppm boron throughout the test.

Opera H lity Time The test time was 7 1/2 days. A degradation equivalency analysis shows that the test profile is thermally more degrading than the composite outside containment profile at PNPS for 30 days.

Aging An aging analysis shows that the materials of the tested splice have a minimum expected life at 105'F of 272 years (glass tape).

The expected life of the glyptal (assume alkyd varnish) at 105*F is greater than 1 x 105 years.

Radiation A radiation analysis shows that the materials of the tested splice have a minimum threshold value of 1.3E6 rads gamma (silicon rubber tape).

Per R[.lC Report No. 21, the dielectric properties of silicones are little affected unless absorbed doses exceed 2 x 108 rads gamma.

Although the insulation resistance decreased during the test, all of the cable splices remained functional throughout the test. Therefore, continued operation is justified.

(

_. to NEDWI No. 271 BOSTON EDISON COMPANY JUSTIFICATION FOR h-CONTINUED OPERATION

-Equipment Identification No. PS1001-90A/D TER No.189b,c ( A, C), 203a, b (8, D) Sheet 1 of 1 7f23fsi Preparer:

4 Date:

Independent' Review:

Date:

7/k tr 7

Approval:

NAbm Date:

7/%/%

( \\

v 11h'e function of,these pressure switches is to provide a high drywell pressure permissive to start the RHR and Core Spray pumps.

These switches will be

-exposed to a harsh steam and radiation environment following a PBOC-2B and 2I L(Reactor Water Cleanup System Pipe Breaks) and a harsh radiation environment following a PBIC and all other P80C's.

From FSAR Appendix G, high drywell pressure does not result from a PBOC.

Therefore, actuation of.these switches to mitigate the effects of a PBOC is

-not required.

-For P81C's, radiation levels of 1 x 106 rads are not reached until at least

100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after the pipe break occurs. As per GE letter the limiting

~ materials are neoprene and Buna-N.

Per D.O.R. Guidelines, Buna-N has a radiation threshold of 1 x 106 rads and neoprene a radiation threshold of I 7

x 10 rads.

Therefore, switch failure due to radiation would not be 1 expected to. occur until at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after the pipe break. Automatic start' of.the RHR and Core Spray pumps would not be required at this time.

lherefore, continued operation is justified.

f x

O r

to NEDW1 No. 271 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERAT10tl i

j..

- Equipment Identification No. PS-512A/D

. TER No. 190 (A, 8),-202 (C, D)

Sheet 1 of 1

~l W]

Preparer:

M Date:

7 - 2 S ' 9'f

' Independent Review: 7

$4/4f Date:

7/27/[V Approval:

TL Date:

~//17/84 N

Y

-The.functi_on of.these pressure switches is to provide a scram signal to the

. Reactor Protection System and to isolate Secondary Containment upon

indication of high drywell pressure. These pressure switches are exposed to

-a harsh steam and radiation environment following PBOC-2B and 2f (Reactor Water Cleanup System Breaks) and a harsh radiation environment following a P81C and-all other PBOC's.

'According to FSAR Appendix G, P80C's do not produce high drywell pressure.

Furthermore, subsequent failures of these pressure switches in the harsh environment caused by these.PB0C's will not reverse the previously accomplished safety functions of scram and secondary containment isolation.

Therefore, these switches do not need to be qualified for the effects of

= PBOC's.

6 rads are not reached until at least

. For-PBIC's', radiation levels of 1 x 10 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after the pipe break occurs. As per GE letter the limiting materials are neoprene and Bur.a-N.

Per D.O.R. Guidelines, Buna-N has a radiation threshold of-1 x 106 rads and neoprene a radiation threshold of I l x 107 rads. Therefore, switch failure due to radiation would not be expected to'. occur-until at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> af ter the pipe break.

It can reasonably-be assumed that the scram and secondary containment isolation li functions would have been completed prior to this time.

Therefore, continued operation is justified.

4 Ll i to NEDWI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment identification No. PS1001-89A/0

. 1ER No. 191 (A), 205 (B), 189a (C), 197 (D)

Sheet 1 of 1 s.

9!!!84 Preparer:

b

$4#S Date:

Y Independent Review:

Date:

/

Approval:

CNtC d 1 ---s Date:

E/l /M Q

The function of these pressure switches is to provide a high drywell pressure

- permissive to the ADS initiation logic.

These switches are exposed to a harsh steam and radiation environment following PB0C -2B and 2T (Reactor

^

Water. Cleanup System Pipe Breaks) and a harsh radiation environment following a PBIC and all other PB0C's.

According to FSAR Appendix G, PBOC's do not produce high drywell pressure.

Operator action is credited in the PNPS Emergency Operating Procedures to initiate ADS if required. Subsequent failure of these switches caused by a harsh environment will not prevent manual operation of ADS f rom the control room. Therefore, these switches do not need to be qualified for the effects of PB0C's.

For PBIC's, radiation levels of 1 x 106 rads are not reached until at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after the pipe break occurs.

As per GE letter the limiting materials are neoprene and Buna-N.

Per D.O.R. Guidelines, Buna-N has a radiation threshold of 1 x 106 rads and neoprene a radiation threshold of 1

- x 107 rads. Therefore, switch failure due to radiation would not be 7

expected to occur until at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> af ter the pipe break.

It can reasonably be ' assumed that an operator could actuate ADS if it would be required at-this time.

' Therefore, continued operation is justified.

m.

7-

. to NEDWI No. 211 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. PS1001-83A/D TER No. 192 (A), 193 (C), 198 (D), 204 (B)

Sheet 1 of 1 b2Y d Preparer:

\\L WW Date:

7 WM Independent Review:

hh Date:

f

/

Approval:

CSCO mt Date:

7/23 /24 i\\

v These pressure switches provide a drywell pressure permissive to the control logic of RHR valves 1001-23A/B,1001-26A/B and 1001-37A/B.

These valves must open in order to provide drywell and suppression pool spray for the purpose of. primary containment cooling. These pressure switches may be exposed to a harsh steam and radiation environment following PBOC-2B and 2T (Reactor Water l-Cleanup System Pipe Breaks) and a harsh radiation environment following a PBIC and all other PB0C's.

From FSAR Section 5, the Containment Spray Subsystem provides containment spra'f capability as an alternate method for reducing containment pressure following a LOCA. This subsystem is designed to remove energy from the drywell by condensing steam or to cool noncondensable gases which have collected in the suppression pool.

Since a PB0C does not result in these conditions, these pressure switches do not need to be able to withstand the environmental conditions associated with PB0C's.

For PBIC's,- radiation levels of 1 x 106 rads are not reached until at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after the pipe break occurs. As per GE letter the limiting materials are neoprene and Buna-N.

Per 0.0.R. Guidelines, Buna-N has a radiation threshold of 1 x 106 rads and neoprene a radiation threshold of I x 107 rads. Therefore, switch failure due to radiation would not be expected to occur until at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> af ter the pipe break.

It can reasonably be assumed that the containment spray subsystem would not be required at this time.

Therefore, continued operation is justified.

-wa

, m

f to NEDWI No. 271 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. PS1360-9 (A-D)

TER No. 194 Sheet 1 of 3 Preparer:

b Date:

7'3O M 7!3O[2'/

.Independen't Review:

2 Date:

Approval:

SCb Date:

7 / 30/ &t-O PS1360-9 (A-D).are Barksdale model PlH-M85SS-V pressure switches used to sense low pressure in the steam line supply lines to the RCIC pump turbine.

The switches are used to signal the closure of two motor operated valves in the RCIC steam supply line in order to prevent steam and radioactive gases from escaping through the turbine shaft seals into the reactor building over a'30. day mission length.

This protection is only required after reactor steam pressure has decreased to such a low value that the turbine can no longer be operated (approximately 100 psig or less).

This condition is expected-to be reached during reactor vessel cooldown and depressurization

.within a few minutes following a LOCA or approximately 6-8 hrs following a small break PBIC or-PBOC.

It is expected that the reactor vessel will reach atmospheric pressure approximately 2-4 hours later at which time, this protective function will no longer be required.

These switches are mounted in the RCIC steam leak instrument rack (C2257-B) located in the mezzanine ot the.RCIC quadrant. These switches could be exposed to a harsh superheated steam and radiation environment during a PBOC-5 (HPCI steam line break in the torus compartment) or PBOC-6 (RCIC steam line leak in the RCIC pump room) or to solely a harsh radiation environment during any other P80C or a PBIC.

Continued operation with these switches can be justified based on the

.following analyses.

Justification o

Temperature and Humidity The PBOC-5 service profile (246*F maximum and 100% RH) exceeds the test

. profile (extended exposure to saturated steam at 212*F) for approximately the

'first 3 rrinutes of the transient.

However, the thermal inertia of the switch and instrument rack in the presence of superheated steam should result in the

. temperatures actually experienced in the vital portions of the switch being enveloped by the test profile.

In the unlikely event the switches did fail,

.two scenarios could occur.

If the switch failed closed, RCIC, which is not credited for this transient, would remain isolated.

If the switch failed

.open, the control room operator could be reasonably assumed to close the valves several hours later following reactor vessel cooldown/depressurization and termination of RCIC.

. to NEDWI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION

. Equipment Identification No. PS1360-9 (A-0)

TER No. 194 Sheet 2 of 3 hM Date:

7-30-R4 Preparer:

Date:

7 D!EY Independent Review:

k Approval:

GCh w

Date:

7/30/84

- T

~

The PBOC-6 serviIe profile (short term exposure to 155*F and 90% humidity) is less severe than the test profile (extended exposure at 212*F and 100%

relative humidity) as documented in AETL Test Report 596-0398.

Therefore, the test temperature profile in the test is actually more severe than the service condition and continued operation of the plant is justified.

o

' Pressure

..The service profile reaches a peak of 14.9 psia, whereas the test pressure reaches a maximum of 7" H O (14.95 psia).

Based on this fact and due to 2

weathertight construction of this instrument, in our engineering judgment, no functional disparities will occur.

Therefore, continued plant operation is justified.

O Radiation From a Wyle Labs analysis -of the materials used for construction of this component, the most limiting material is a fiberboard type insulator which 6 rads for this application.

The has'a damage threshold of approximately 10 leak tight nature of the switch is expected to preclude beta radiation exposure to this components.

It is estimated that the gamma dose to this 6

component will meet the 10 threshold approximately 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> following a LOCA.

It'is expected that the RCIC steam line would have isolated on low pressure and sealed in prior to this time.

In the event that this exposure induccd a failure of the switch, tne steam line valves would be capable of 1 opening if the operator reset the isolation and either the operator held the control switches in the open position or a low reactor vessel level or high drywell pressure was sensed.

In the unlikely event that these conditions

~.were met opening of the valves would have a negligible effect on the ability to maintain exposures below 10CFR100 limits since RCIC operation should be completed and the reactor vessel would be expected to be nearing a cold depressurized condition at the time the damage threshold would be reached.

In addition, failure of the switch would not inhibit the ability to reclose these valves. Continued operation is justified.

- to NEDWI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. PS1360-9 (A-D)

TER No. 194 Sheet 3 of 3

- Preparer:

k Date:

7-70~8i 3O[27 Independent Review:

Date:

Approval:

O m

Date:

7/30/f4 x) o: Aging Wyle Labs has completed aging and thermal degradation analyses that confirm that the six hour qualification exposure documented in AETL Test Report 596-0398, envelopes the conditions experienced at PNPS over 40 years and a 30 day mission length accident exposure for these switches, a

1 r,m Y

U-.----

.,,-,-,,e-.

A

[ to NEDWI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. PS2360-1

'1ER No. 196 Sheet 1 of ~L

' Preparer:

Date:

2b Y

7/26hV Ind'ependent Review: k Date:

Approval:

Ib Date:

~7 /'2 ') / 7 4

<N v

1his pressure switch detects HPCI pump low suction pressure and is,

~ ~ 'y therefore, required solely to assure satisfactory HPCIS operation.

The HPCIS is relied upon to operate during and following Loss of Feedwater Flow, Total Loss of Offsite Power, Shutdown f rom Outside Control Room (Special Event),

Pipe Break Inside Primary Containment, Control Rod Drop Accident, and Pipe Break Outside Prinary Containment to assure continued core cooling, and thus. mitigate consequences which could result'in potential offsite exposures comparable to the 10CFR100 guidelines.

None of_the first three events listed above is expected to result in environmental conditions any more severe than those experienced during normal operation. The fourth event is addressed in the HPCIS Safety Evaluation, which states that, "The HPCIS is designed to provide adequate core cooling for small breaks... core never uncovers and is continuously cooled throughout

..the-transient so that no core damage of any kind occurs f or breaks that lie within the range of the HPCI."

Thus, the size of LOCA presumed to generate

--postulated core damage is beycad the capacity of HPCIS to provide core cooling.

1he Control Rod Drop Accident has been evaluated and no HPCIS equipment will be subjected to pressure, temperature, radiation or humidity conditions any more severe than those experienced during normal operation.

Those pipe breaks'outside containment which could be expected to result in harsh conditions of pressure, temperature and humidity in the vicinity of this pressure switch is the PBOC-3 and the P20C-5.

Each of these events, however, incapacitates the HPCIS. System operability is, therefore, not required for either P80C.

to NECWI No. 2/7 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. PS2360-1 TER No. 196 Sheet 2 of 2 1 Preparer:

Date:

MkI

((86MV Independent Review: h '

pr Date:

-Approval:

GCAbw Date:

'l /29 / TM

~

\\ \\

w On-the other hand, system operability is required f or the main steam line breaks, PBOC-7 and PB0C-8, either of which could result in cumulative

' radiation exposures to the pressure switch well in excess of 104 rads.

These values are based conservatively on the postulated core damage of NUREG 0737 and NUREG 0588. However, FSAR analysis of the PNPS design basis Main Steam Line Break Accident indicates that, with a maximum 10.5 second MSiv closure and continued core coverage (from normal or standby systems, including HPCIS), there would be no fuel damage. Without core damage, exposures will not exceed 104 rads.

.MSIV closure time is verified once per quarter under Technical Specification surveillance requirements. The closure time must be greater than 3 seconds and less-than 5 seconds for the valve to be considered operable.

For valve closure times shorter than 10.5 seconds, the postulated accident is

-considered less severe than that' analyzed.

Core. cooling systems are also verified operable periodically under plant surveillance requirements. Thus, if HPCIS must be declared inoperable as a

' consequence of the PB0C, then ADS, LPCI and Core Spray are all assumed to be operable to assure safe shutdown of the plant.

If all core cooling systems

, operate as-designed and tested, no fuel damage should occur.

Since the assumptions of NUREG 0737 and NUREG 0588 are considered unrealistic on this basis, failure of PS2360-1, as a consequence of excessive radiation exposure from the main steam line break accident, is considered highly improbable and continued cperation is justified.

. to NEDWI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR-CONTINUED OPERATION Equipment Identification No. PS-2389 (A-D)

LTER No. 207 Sheet 1 of 2 Preparer:

-Date:

Independent Review: Y b l/I Date:

$ 2-84

~

Approval:

T Date:

BP23 %

i T v

\\

PS-2389(A-0) are Barksdale model PlH-M85SS-V pressure switches used to sense

-low pressure in the. steam line supply lines to the HPCI pump turbine. The switches are used to signal the closure of two motor operated valves in the HPCI steam supply line in order to prevent steam and radioactive gases f rom

" escaping through:the turbine shaft seals into the reactor building.

This protection is only required after reactor steam pressure has decreased to such a. low value that the turbine can no longer be operated (approximately

-100 psig or less). This condition is expected to be reached during reactor vessel cooldown and depressurization within a few minutes following a LOCA or approximately 6-8 hrs following a small break PBIC or PBOC.

It is expected that the reactor vessel will reach atmospheric pressure approximately 2-4 hours laterLat which time, this protective function will no longer be required. These switches _are mounted in the HPCI steam leak instrument rack (C2257-A) located in the NW RHR quadrant.

These switches could be exposed to a harsh superheated steam and radiation envirorment during a PB0C-5 (HPCI steam line break in the torus compartment) or to solely a harsh radiation

e environment during any other PBOC or a FBIC.

Continued operation with these switches Ecan be justified based on the following analyses.

' Analytical Justification o-Temperature and Humidity AETL Test Report 596-0398 documents tne qualification testing of a similar component in a harsh steam environment. The test profile consisted of a rise to 212*FJin an unspecified time with a dwell at 212*F for six hours.

However, the PBOC-5 Service profile for this location consists of a rapid spike to 229'F with a return to less than 212*F in less than 3 minutes.

It is our engineering judgment that due to the thermal inertia of the components, the internal temperature experienced by these switches during the predicted service event will be significantly less than that which was experienced in the test. Therefore, the test temperature profile is essentially more severe than the service conditions and continued operation of the plant is justified.

It should be noted that HPCI operation is not required during.this transient and that the valves controlled by these switches will'be automatically closed in response to increased HPCI flow and space temperature resulting from the leak, f

_m_%

' to NEDWI No. 277 BOSTON EDISON. COMPANY JUSTIFICATION FOR CONTINUED OPERATION I

Equipment Identification No. PS-2389 (A-D)

TER No. 207-Sheet 2 of 2 kT l Preparer:

Date:

-Independent Review:

kl.

S-Date:

$ L h4

~

~

V U

Approval:

O CA Date:

9,l2 /84 iT v

o-Pressure-The service profile reaches a peak pressure of 15.3 psia and decays to 4

-atmospheric pressure within seconds.

In our engineering judgment, exposure to this pressure change.will cause no functional disparities due to weathertight. construction of these switches.

It should be noted that HPCI operation -is not required during this transient and that the valves

. controlled by these switches will be automatically closed in response to increased HPCI' flow and space temperature resulting from the leak.

=Therefore, continued plant operation is justified.

~

o Radiation.

From a Wyle Labs: analysis of the materials used for construction of this component,! the most limiting material is a fiberboard type insulator which has a dariage threstold of'approximately 106 rads for this application.

The

1.eak tight-nature of.the switch is expected to preclude beta radiation

. exposure to this components.

It is estimated that.the gamma dose to this py

component will m6et the 106 threshold approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> following a LOCA.

It.is expected that the HPCI steam line would have isolated on low

/

fpressure. prior toithis time.

In the unlikely event that this exposure 1

induced a failure of the switch, the steam line valves would be capable of m

J. /

opening -if ' the operato'r held the control switches in the open position or if

'"N J

a low; reactor vessel levellor high drywell pressure is sensed.

In the x

'unlikely event that these conditions were met, opening of the valves would-

'x have a negligible effect on the' ability to maintain exposures below 10CFR100

. limits' since HPCI' operation should be completed and the reactor vessel would

~ be expected to be nearing a cold depressurized condition at the time the

' damage threshold would be reached..In addition, failure of the switch would

not. inhibit the. operators ability to reclose these valves.

Continued 1 operation is~ justified.

lo-. Aging Wylel Labs.has completed aging and thermal degradation analyses that confirm

', that the six hour qualification. exposure documented in AETL Test Report 596-0398, envelopes the conditions experienced at PNPS over 40 years and a 30 day mission length accident exposure for these switches.

~

to NEDWI No. 277 a.

BOSTON EDISON COMPAN'l JUSTIFICATION FOR CONTINUED OPERATION

' Equipment Identification No. M01301-62 TER No. N/A Sheet 1 of 1 7!30!8Y Preparer:

A 8

Date:

LIndependent Review:

Date:

7 O

Approval:

CTd ')%

Date:

~1/30/34

>~

Q M01301-62 operates the block / control valve for supplying cooling water t low f rom the RCIC Pump Discharge to the Barometric Condenser (E-201).

The valve is normally closed but must open to facilitate RCIC operation.

The valve is located in the RCIC Pump Room. The only transients for which RCIC operation is credited are control rod drop, loss of all AC power and loss of feedwater.

However, these transients have been evaluated and no RCICs equipment will be subject to pressure, temperature, radiation or humidity conditions any more severe than those. experienced during normal operation.

Therefore operability of M01301-62 during exposure to a harsh environment need not be demonstrated for transients requiring RCIC operation.

Therefore, this valve is not within the scope of 10CFR50.49 and will be deleted from the EQML.

Attachm nt 5 to NEDWI ho. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. M01301-49

.TER No.<N/A Sheet 1 of 1 7!$0 4 Preparer:

M ett4-Date:

'7! ##/@Y Independent Review:

Date:

Approval:

FCd 'ww Date:

~7/30/84 J

M01301-49 operates the block valve ~in the discharge of the RCIC pump.

This valve'is located outside the containment in the RCIC Pump Room (zone 1.10).

M01301-49 utilizes a 125v DC reliance motor with class "HR" insulation for which complete qualification documentation is not available. M01301-49 is normally closed and is automatically signaled opened in response to a low reactor vessel level to facilitate injection of RCIC coolant into the vessel. The only transients for which RCIC operation is credited are Control Rod Drop, total loss of AC power and loss of feedwater.

However, these transients have been evaluated and no RCICs equipment will be subject to pressure, temperature, radiation or humidity conditions any more severe than those experienced during normal operation.

Therefore operability

'of M01301-49 during exposure to a harsh environment need not be demonstrated for transients requiring RCIC operation.

= M01301 -49 serves a second safety related function by providing containment isolation for transients not requiring RCIC operation.

M01301-49 would remain in a normally closed position during such transients and would not be required to actively function.

In the unlikely event that M01301-49 was open i

and failed open,. redundant isolation of this penetration would be provided by M01301-48, A01301-50 and 58A.

Based on these considerations, continued operation is justified.

to NEDWI No. 2/7 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. M01001-19 TER.No. N/A Sheet 1 of 1 Preparer:

h6 ft # $

Date:

7' 30" I4 Independent Review:

f Date:

7 O!E'[

~

g Approval:

C' COL m Date:

') l 30/R4 4

v M01001-19 ' operates the block valve in the cross tie between the "A" and "B" train RHR pump combined discharge headers.

The valve is located in the CR0 mezzanine pump room at elevation 2'9" (zone 1.8).

The valve is normally key-locked open and should remain open to ensure that all four RHR pumps deliver LPCI to the selected loop.

This valve could be exposed to a harsh steam and radiation environment during a PBOC-5 (HPCI Steam Line Break in the Torus Compartme';t)'and/or to a harsh radiation environment during a PBIC, or any PBOC other than a PBOC-5. The valve is required to remain functional for a 30 day mission-t he to facilitate operation of the RHR system in a variety

- of modes. Limitorque Report B0003 documents the qualification of a similar operator and motor for a harsh steam and radiation environment that nnvelopes the service profile for all postulated transients af fecting M01001-19 including the PB0C-5.

M01001-19 is therefore considered to be qualified pending completion of an inspection to verify that appropriate terminal blocks were utilized for power lead termination (required by IE Notice 72).

Inspection of the terminal blocks in the Limitorque operators is in progress. No deficiencies in the terminal blocks have been reported and therefore there is a high degree of assurance that the terminal blocks in this motor operator are qualified.

Continued operation is therefore justified.

[

to NEDWI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. M02301-6

. TER No. N/A Sheet 1 of 1 Preparer:

/A/) b e wr Date: ?O TUNC OV Independent Review: h db Date:

20 Juu& 84 Approval:

/

Date:

A f

7\\

v M02301-6 operates the valve in the line from the condensate storage tanks to the HPCI pump suction.

The valve is located in the RBCCW pump room-B (zone

. 1.22) and is normally open.

The valve will automatically be closed to facilitate a transfer of HPCI suction to tne torus in the event that high torus level or low condensate storage tank level is sensed.

During a PBOC-3 (HPCI Steam Line Break in the HPCI Pump Station) this valve would be exposed to a harsh environment. However, HPCI operation is not required for this transient so the exposure of this valve is inconsequential since it serves no function other than supporting HPCI operation.

During any other PBOC, a PBIC or a Control Rod Drop, this valve must remain operable over the five (5) hours mission time of the HPCI System.

However, analysis has indicated that the valve would not be exposed to a harsh environment during this time frame.

Based on these considerations, continued operation is justified.

h:

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Attachmsnt 5 to NEDWI No. 21/

BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. M04065

.TER No. N/A.

Sheet 1 of 1

. Preparer:

4h Date:

7dO-f4 7h0/EY Date:

Independent Review:

~

Approval:

CRCOm a Date:

7/ 30/ M

.\\

v M04065 operates the block valve in the RBCCW supply line to fuel pool heat exchanger E-206A.

This valve is-normally open and needs to be capable of closing to reduce non-e:sential loads on the RBCCW system during a design basis event-requiring RBCCW cooling water supply to the RHR heat exchanger (s). This valve is located outside containment in the fuel pool heat exchanger room (zone 1.13).

A valve' operator and motor similar to M04065 was qualified in a steam environment.to 250*F, 25 psig and 2 x 107 rads as documented in Limitorque Report B0003..This qualification profile envelopes the service profile for all potential harsh environment exposures to M04065 and is theretare considered qualified pending completion of an inspection to verify that appropriate terminal blocks were utilized for power lead termination (required by IE Notice 83-72).

Inspection of the terminal blocks in the

.Limitorque operators is in progress.

No deficiencies in the terminal blocks have been reported and therefore there is a high degree of assurance that the terminal blocks in this motor operator are qualified.

Based on this

- determination, continued operation is justified.

+_

to NEDW1 No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment I'dentification No. M04009A, M04009B

'TER No. N/A-Sheet-1 of 1

$ /![f4 Preparer:

Date:

"7[ &-

Date:

/;/EY Independent Review:

. Approval:

GOCbw Date:

% / 2 / 24 Q

M04009A and M04009B are required for isolation of non-essential loads in the "B" Loop of _the RBCCW System and may also be used for subsequent restoration of non-essential RBCCW loads once the heat load of the RHR heat exchangers has decreased. These valves are located outside containment in the RBCCW

-Pump Rocm-B and are normally open.

During a PBIC, these valves will be exposed to increased amounts of radiation.

However, the increase will be insufficient to cause a harsh environment exposure until after the valves mission length has elapsed. As a result, the operability of M04009A/8 post-PBIC is assured.

A valve operator and motor combination similar to M04009A and M040098 was qualified for extended exposure to steam at 250*F as documented in Limitorque Qualification ~ Test Report B0003. During a PBOC-3, M04009A and M040098 are exposed to_superheated steam resulting in a service profile which peaks at 301'F and exceeds the _qualf fication profile for approximately the first five minutes. However, the short tera thermal inertia of the valve operator in a superheated steam environment, as documented in-Limitorque Repo,t 80027, Iwould result in the vital portions of the operator and motor being exposed to temperatures _well ~ below those of the qualification profile.

In addition, the Ovalves would not be exposed to a harsh radiation environment.

Based o'n these

- considerations, continued operation is justified.

J

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to NEDW1 No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. M04083 TER No. N/A Sheet 1 of 1 Preparer:

b Date:

8. l [I4

/!8!

Independent Review:

Date:

Approval:

MD c-a Date:

% / l /84-N V

M04083 operates the RBCCW bypass valve around the "B" loop RBCCW to Salt Service Water Heat Exchanger E-2098. The valve is located in the RBCCW-B compartment and is normally maintained in a throttled position.

If the

-valve failed.to operate, the ability of the RHR system to cool the torus would be degraded.

During 'a PBIC, this valve is exposed to increased radiation.

However, a

-harsh exposure does not occur until well af ter the 30 day mission life of these' valves has passed. As a result, operability of this valve post-PBIC is

-assured, n ' valve operator and motor similar to M04083 was qualified for extended exposure to saturated steam at 250*F as documented in Limitorque Qualification Test Report B0003. During a PBOC-3, M04083 is exposed to superheated steam resulting in a service profile which peaks at 301*F and exceeds the qualification test profile for approximately the first 5

' minutes. However, the short term thermal inertia of the valve operator in a superheatad steam environment as documented in Limitorque Report 80027, would result in the vital portions of the operator and motor being exposed to temperatures well below those of the qualification test profile.

In addition,i the valve would not be exposed to a harsh radiation environment during the 30 day mission length. Based on these considerations, continued operation is justified.

.c tj G to NEDW1 No. 211 0

BOSTON EDISON COMPANY JUST1FICATION FOR CONTINUED OPERATION y

Equipme't!!dentification No. M03805, M03806 n

TER No,'N/A Sheet 1 of 1 i

s Preparer:

M/I Date:

/E7/dY E

8Y Independent' Review: 7 Date:

Approval:

CT Mw Date:

7/30/84 T

.v M03805 and M03806 operate the block / isolation valves in the salt service water discharge f rom TBCCW heat exchanger E-122B and RBCCW heat exchanger E-2098 respectively.

These valves are located in the RBCCW Pump Room-B and are normally maintained in a throttled open position.

Following a design basisaccider.t,h03805willclosetgapresetthrottlepositionandM03806 will go fu.Il open to ensure an ad,equata supply of salt service water to the RBCCWheat'exeqpngertofacilitateLPCI,CoreSprayorRHRtoruscooling operation as required. Both operators are equipped with Reliance Motors UtilizingClasP8Insulationforoutsidecontainmentapplication.

Duringi a PBIC, these valves are exposed to increased radiation.

However, a harsh radiation environmlnt does not develop until well atter the 30 day mission life of these valves has passed and as a result, M03805 and M03806

-need'not be qualified for a PBIC.

p A valveToperator and motor combination similar to M03805 and M03806 was qualifica for extended exposure to steam at 250*F as documented in Ltmitorque

. Report 80003.

During a PBOC-3, M03805 and 3806 are exposed to superh'eated steam resulting in a service profile which peaks at 301*F and exceeds the qualifiqat,iongrSllle for approximately the first five minutes.

However, the short tera thermal inertia of the valve operat6r in a superheated steam environment, a;s documented in Limitorque Report B0027, would result in the.

vital portiohs of the operator and motor being exposed to temperatures well below those o( the qualification profile.

In addition, the valves would not be exposed t Based on these considerations, continued oi,o a harsh radiation environment.

e'r'ation is' justified.

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. to NEDW1 No. 211 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION

. Equipment Identification No. CX2, CX4,.520, CX8

-TER No 'N/A Sheet 1 of 1 7/d 7!IY Preparer:

Date:

.h.

Date:

7/77//P Independent Review:

. Approval:

hTL Date:

9 i Tl f M V

This equipment consists of polyethylene insulated cable (installed outside of the drywell) provided by several manufacturers. While no qualification documentation or testing history has been found for these specific cables, similarly constructed cable has been successfully subjected to sequential testing (proprietary TR #17513-1), which documents qualification of the insulation system to 1.63 x 108 rads gamma and a LOCA condition including temperatures up to 325'F.

These conditions are more severe than the

-conditions at PNPS.

The generic materials which make up the insulation system have expected lives of greater than 1.5E4 years (PE) in an ambient temperature of 105'F.

Based on the above, it is judged that the PE cable installed is justified f or

. continued use pending f urther testing or replacement.

' Q R-

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' 'r j ;

'y to NEDW1 No. 277

, BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION 9

Equipment Identification No. 51, 537, CXG, 23, 23A, 548, CO2,'003, S19, 527 c

TER No. N/A Sheet 1 of 1 Preparer:

Date:

l Independent Review: kihe Date:

I i @4 U

U f.

Approval:

, QWC= e Date:

Ell /T4 e

q)

) jf.

11His* equipment consists of polyethylene insulated polyvinylchloride jacketed

'" cable (installed outside the drywell) provided by several manuf acturers.

While no qualification documentation or testing history has been found for

.these spehif N cabM.simitgrly Mnstructed cable has been successfully subjected to ii9questial' testing (proprietary TR #17513-1), which documents Oqualification bf. the insulation system' to 1.63 x 108 rads gaarna and a LOCA condition. including temperatures up, to 325'F.

These conditions are more L

severe than tde conditions at PNPS-g.

^

~ The generic materials which make up,the insulation system have expected lives

' ' of greater than _1.4E4 years (PVC), and greater than 1.5E4 years (PE) in an 3

X

. ambient temperature of 105'F.

r-Based on,the above, it is judged that the PVC/PE cable installed is justified

- for continued use< pending further testing or replacement.

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a to NEDWI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. SVL61 TER No. N/A Sheet 1 of 1 7/30!EY Preparer:

Date:

(IndependentReview: kbh Date:

7f30ff4 V

Q Approval:

f'Tdh Date:

7/ 30 /M 3

v This solenoid valve is an ASCO HT8320A10T.

It is unknown if this valve will survive the radiation due to a LOCA; however, it is not required to operate post accident. SVL61 controls the air supply to operate the cross-tie

' t damper between Reactor Building Clean Exhaust and Refueling Floor Exhaust Ducts..This' damper provides the isolation between safety-related and nonsafety-related exhaust ducts.

If the SVL61 disc fails, the air supply to

- A0/N-138 will be vented and A0/N-138, which is normally closed, will f ail open.

If A0/N-138 f ails open, when SGTS operates, a suction will be drawn on

-both the Reactor Building Clean Exhaust and the Refueling Floor Exhaust

. simultaneously.

If SVL61 f ails such that air is continuously supplied to A0/N-138, A0/N-138

.will remain. shut. However, the Reactor Building Clean Exhaust and Refueling Floor Exhaust can both be ventilated through their own ducts and isolation valves '( A0/N-100 and A0/N-101).

0-Therefore,- continued operation is justified.

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m to NE0WI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. SVL82, SVL83 TER No. N/A Sheet 1 of 1

- Preparer:

[b Date:

bb- [f Y

-Independent Review:

d e'v Date:

[M f

f f f Approval:

C S C N\\5 d m u Date:

~1/ 267/ R4

( T 1hese solenoid valves control air operated valves that allow the SGTS to obtain a suction on the suppression pool.

These valves are not required to

-operate post-PB0C; therefore, they only need to be qualified for radiation.

^1he 40 year TID plus LOCA dose is 2.86 x 106 rads.

The two materials of concern in this valve are the Buna-N elastomers and the acetal disc holder.

- In EPRI Report NP-2129, " Radiation Ef fects on Organic Materials in Nuclear

. Plants" Buna-N is listed as not reaching the 50% loss of elongation point until af ter 7 x 10' rads and acetal resins are listed as reaching the $0%

loss of. tensile: strength at'4 x 106 rads.

Each of these doses is greater than the combined 40 year TIO plus accident dose.

Therefore, no detrimental ef fects, due to radiation, are expected, and continued operation is justified.

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..., - - - - ~.,,,

., to NEDWI No. 27/

BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment identification No. P2020, E, F lER No. N/A Sheet 1 of 1 7!2 2/8'[

Preparer:

/

Date:

7[/Y Independent Review:

~

Date:

r

/

Approval:

CKdhe Date:

7/3o/M t

N, m

These pumps / motors provide the motive force for the RBCCW flow in RBCCW Loop

'B' P202E and F have motors that are environmentally qualified.

P202D has a motor that has been rewcund.

All three motors have motor lead connections that have not been environmentally qualifitd.

However, a justification for continued operation has been previously presented (Terminations-Ring Tongue

(<4C!)) for the connections.

106 EachRBCCWloopisde;ignedtotransfertheRHRsgstemheatload(64x BTUs/ hour) plus an additional heat load o+ 1 x 10 BTUs/ hour for other i

essential equipment with only 2 of the ? RBCCW pumps in the loop operating.

Therefore, if P202D did fail post-accident, the operators would be able to start the non-operating pump in the

'B' Loop and /or shitt to the

'A' Loop.

Based on the above information, continued operation is justified.

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. to NEDWI No. 271 BOSTON EDISON COMPANY JUSTIFICATION FOR y.g CONTINUED OPERATION I

Eq'uipment Identification No. 412A Okonite, 106 Kerite, 2128 Rockbestos, SCib Rockbestos TER No. N/A Sheet 1 of 1 7!27!IY

- Preparer:

Date:

Independent Review:

Date:

7 27/

' Approval':

c3C Date:

7 (17 lid t T lhe insulation. systems used on these cables are environmentally qualified.

iThe documentation packages for these cables are being completed and when

~ completed, environmental qualification will be proven.

Therefore, continued

. operation is' justified.

4-

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_ _ - - to NEDWI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION

- Equipment Identification No. J720

.TER No. N/A Sheet 1 of 1 Preparer:

Date:

/

Independent Review: h/,

9l!P4 Date:

_u u

Approval:

C_~[dhhru Date:

9/l /%

N v

This junction box is in the electrical circuit on M0/N-113 and SVL-70.

M0/N-113 and SVL-70 are only required post-LOCA. Therefore, this junction box need only be qualified for radiation.

The manufacturer of terminal blocks installed in this junction box has not yet been determined.

However, Sandia National Laboratories and other laboratories have compiled extensive test data on terminal blocks (both protected and unprotected) of various manufacturers, which has shown that the probability of failure of the terminal blocks is very low.

Sandia tested over 400 terminals in their own facilities.

- Sandia f ound "that for THI-2 accident conditions, the ef f ect of radiolysis is negligible on surface conductivity and therefore on breakdown" and therefore the radiation in the Reactor Building will not increase the probability of failure. Based on the above information, continued plant operation is justified.

M

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to NEDWI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. J32 TER No. N/A Sheet 1 of 1

-Preparer:

& b bo _e 0'b9h'/

Date:

Independent: Review: h Date:

f Approval:

QCb Date:

7/26/24 w

This junction box is in the electrical circuit of the following temperature

switches

..Swi tch ;

' Location Setpoint TS2370C HPCI Valve Room exh. duct 160

• F -170

• F TS2371C Torus Area exh. duct 190*F-200*F 152372C --

HPCI Valve Room exh. duct 160* F-170

• F

-TS2373C

. lorus Area exh. duct 190'F-200*F

. Exceeding the setpoint is an indication of a HPCI steam line break.

The

closing ~of the temperature switch contacts causes an auto isolation signal (which " seals in") to be sent.to M02301-4, 5, 35, and 36, which shut to isolate the HPCI steam supply line and the HPCI pump suction line from the Torus. : The terminal block inside the junction box need only operate up to the point.that the isolation signal " seals in".

Because the terminal block is,inside the. junction bux there will be a time lag between the temperature switches being subjected to the pipe break and the terminal blocks being

. subjected to the pipe break. The terminal blocks will probably never be subjected to an elevated temperature prior to performing their safety

~

function. Therefore, failure of the terminal block is unlikely and continued plant operation is justified.

b

h to NEDWI No. 211 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. J217 TER No. N/A Sheet 1 of 1 7!27!ET Date:

Preparer:

~

[!27[f Independent Review:

[

Date:

Approval:

CTdCh

_u Date:

7 !11/ M iN v

This junction box is in the electrical circuit for solenoid val.ve SV220-45, that in turn controls A0220-45 (Reactor Coolant Sample Line Outboard Isolation valve). SV220-45 is normally shut and is required to be shut to ensure primary containment isolation post-accident.

If A0220-45 is open at the beginning of an accident, the isolation signal causes SV220-45 to deenergize which will cause A0220-45 to shut. All credible failures of the terminal block within J217 will also deenergize SV220-45 and cause A0220-45 to shut.

Therefore, failure of J217 will have no adverse etfects, and continued plant operation is justified.

l II

M-Y to NEDWI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION n

q --

Equipment Identification No. J451 TER No. N/A.

Sheet 1 of 1 Date:

7 Y!@Y b

. Preparer:

e Independent Review: k Date:

I Approval:

T bo Date:

W27 / S4-wa 1 %

v This junction box _is in the electrical circuit for the following valves:

SV5033B Normal Purge Suppy to Drywe1I SV5033C_

Normal Nitrogen Makeup to Drywell

.SV5035B Purge Air to Drywell SV50368 Purge Air to Torus SV50428 Purge Exhaust from Torus SV50448 Purge Exhaust from Drywell SV5033C is normally energized and A05033C is normally open. All other valves are-normally deenergized and their respective air operated valves are normally shut. _ All of the air operators fail shut. All of the solenoid valves receive Containment Isolation signals upon a LOCA.

The isolation signal causes the solenoid valves to deenergize (if not already deenergized) and the air operated valves to shut. All credible failures of the terminal block within J451 will'also deenergize the solenoid valves and cause the air operated valve to shut.

Therefore, failure of J451 will have no adverse ef fects, and continued plant operation is justified.

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k to NEDWI No. 271 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION c

Equipment Identification No. J866 TER No. N/A Sheet 1 of 1 7/3c[F7 Preparer:

Date:

7!30!81 Independent Review:

Date:

l A'pproval: '

C Date:

7/3o/84-A T The manufacturer of the terminal block in this junction box has not yet been determined. However, Sandia National Laboratories and other laboratories have compiled extensive test data on terminal blocks (both protected and unprotected) of various manufacturers, which has shown that the probability of failure of the terminal blocks is very low. Sandia tested over 400 terminals in their own facilities.

~ The. worst case temperature peaks at 189.6*F and immediately starts to dec rease.- Applying 189.6*F to Sandia's probability curves for 480v/100%RH/5

. hours, the probability of f&ilure is less than.00075, for protected terminal blocks in 120V circuits. The installed terminal blocks may be required for longer than 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, but the Reactor Building temperature decreases to 140*F within 60 minutes of the PBOC, thereby reducing the probability of failure even further. Also, Sandia's research showed that most f ailures occurred in the initial phase of the tests.

Sandia found "that for TMI-2 accident conditions, the ef f ect of radiolysis is

. negligible on surface conductivity and therefore on breakdown" and therefore the radiation in the Reactor Building will not increase the probability of failure. Based on.the above information, continued plant operation is

-justified.

L

( to flEDWI No. 277 BOSTON EDISON COMPANY

/

JUSTIFICA". ION FOR CONTINUED DPERATION Equipment Identification No. C22578

-TER No. N/A Sheet 1 of 1 7!I7 I'/

Preparer:

Date:

~ Independent Reviewi M b

Date:

//

-Approval:

A Date:

7 / 2G /%

Q The manufacturer of terminal blocks in this local control panel has not yet been determined. However, Sandia National Laboratories and other laboratories.have compiled exter.sive test data on terminal blocks (both protected and unprotected) of various manuf acturers, which has shown that the probability of failure of the terminal blocks is very low.

Sandia tested over 400 terminals in their own f acilities.

The worst case temperature peaks at 246.5"F and immediately starts to decrease. -Applying 246.5'F to Sandia's probability curves for 480v/100%RH/5 hours,- the probability of f ailure is only.0018, for protected terminal

-blocks in 120V circuits'.

The installed terminal blocks may be required for longer than 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />,'but the Reactor Building temperature decreased to 122*F within 60 minutes of the PBOC, thereby reducing the probability of failure even further. Also, Sandia's research showed that most f ailures occurred in the initial phase of the tests.

Sandia found "that for 1MI-2 accident conditions, the effect of radiolysis is negligible on surface conductivity'and therefore on breakdown" and therefore

~the. radiation in the Reactor Buildir.g will not increase the probability of failure. Based on the above information, continued plant operation is justified.

?f -

m h,{ to NEDW1 No. 217

?

BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification Fo. C2207A TER No. N/A' Sheet 1 of 1 l9!b [

Preparer:

v1 Date:

__ k. lbrea/_

, /.f,I Independent Review:

Date:

' Approval:

(_3f b - > - -

Date:

7/ 2.C,/ 84

( T

-The manufacturer of ' terminal blocks in this local control panel has riot yet been' determined.

However, Sandia National Laboratories and other laboratories have compiled extensive. test data on terminal blocks (both protected and unprotected) of various manufacturers, which has shown that the probability of failure of the terminal blocks is very low. Sandia tested over 400 terminals in their own f acilities.

The-worst case temperature peaks at 246.3*F and immediately starts to

' decrease.. Applying 246.3*F to Sandia's probability curves for 480v/100%RH/5

' hours, the probability of failure is only.0018, for protected terminal blocks in 120V circuits. The installed terminal blocks may be required for longer than 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, but the Reactor Building temperature decreases to 123*F within 60 minutes of the P80C, thereby reducing the probability of failure even.further. Also, Sandia's research showed that most tailures occurred in the initial phase of the tests.

Sandia found "that for TMI-2 accident conditions, the ef fect of radiolysis is negligible on surface conductivity and therefore on breakdown" and therefore the radiation in the Reactor Building will not increase the probability of failure.. Based on the above information, continued plant operation is justified.

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( to NEDW1 No. 2/1 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION i

' Equipment Identification No. J599, J600, J601, J602, J606 TER'No. N/A Sheet 1 of 1 e

Date:

/7 Preparer:

7 /// M

Independent Review

[jhrr Date:

- -Approval:

O C

u Date:

H 4 f 24

,N v

. The manufacturer of terminal blocks installed in these junction boxes has not yet been determined.

However, Sandia National Laboratories and other

. laboratories have compiled extensive test data on terminal blocks (both protected and unprotected) of various manufacturers, which has shown that the probability of failure of the terminal blocks is very low. Sandia tested

.over 400 terminals in their own f acilities.

The worst case temperature peaks at 224.7'F and immediately starts to decrease. Applying 224.7'F to Sandia's probability curves for 480v/100%RH/S hours, the probability of failure is only.0018, for protected terminal

. blocks in 120V circuits. The installed terminal blocks may be required for longer than 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, but the Reactor Building temperature decreased to 132*F within 90 minutes of the PBOC,.thereby reducing the probability of failure even further. Also, Sandia's research showed that most failures occurred in the initial phase of the tests.

Sandia found "that for TMI-2 accident conditions, the ef fect of radiolysis is negligible on surface conductivity and therefore on breakdown" and therefore the radiation in the Reactor Building will not increase the probability of failure.

Based on the above information, continued plant operation is justified.

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f; to NEDWI No. 211

(.

' BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION p

Equipment Identification No. PS4008

. TER No.-N/A-Sheet 1 of 1 7!2 7fR*/

Date:

Preparer:

" Independent-Review:

Date:

7/lMM r

Approval:

C9tCEb e Date:

7 / 2.7 / S'4

\\T U

lhis pressure switch sends a pump low discharge pressure signal for pumps P2020, E, F to the RBCCW Loop 'B' Control Circuitry. When the discharge fpressure goes below 31 psig the pressure switch sends a signal to start either pump P202E or F, whichever is in " auto".

Failure of the pressure switch may prevent the pumps from starting automatically upon failure of one

. pump. However, each-RBCCW loop is designed to transfer the RHR system heat 6

' load'(64 x 106 BTUs/ hour) plus an additional heat load of I x 10 BTU / hour for other essential equipment.

Therefore, if failure of the pressure switch, along with a PBOC and failure of the operating RBCCW Loop

'B' pump does occur, the operator can shift to RBCCW Loop

'A' to remove the

-heat loads. The Loop 'A' equipment is located in an area of mild environment i

~

and therefore can be relied on to operate post-accident.

Based on the above information, continued operation is justified.

A t

- to NEDWI No. 277 91-BOSTON EDISON COMPANY

' JUSTIFICATION FOR CONTINUED OPERATION

[quipment' identification No. TIP Ball and Shear Valves

. TER No. N/A-Sheet 1 of 1

!y 2 7, /98'/

' Preparer:

Ne[

Date:

Independent-Review:

N 8 * -2 Date:

27///

f f

Approval:

- (CECADie Date:

7/3O/M Q

iThe safety function of these valves is primary containment isolation.

The only accident for which they must provide this safety function is a pipe break' inside primary containment. The components are located outside

. containment, and therefore, they must be qualified for radiation and aging

only.

(Continued operation with these components not qualified is justified because ithey provide diverse means of isolating the affected penetrations.

The ball

' valves are closed more than 99% of the time (TIP usage is approximately 3 hrs, per 2-week. period).and-they do not require power to remain closed.

In fthe unlikely event that a pipe break inside primary. containment occurs with the TIP probes inside the drywell, diversity in the system provides assurance

that the. penetrations will be isolated. The limit switches in the ball valve

! Provide a' signal to the Primary Containment Isolation display in the Main

. Control Room. JShould any of the ball valves spuriously open or be held open by a stuck probe under accident conditions, the operators can detect this and

-fire the shear l valves which are powered by 125V DC,' ensuring operability in

-case offsite power is lost.

.Therefore, continued operation is justified.

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! ;. to NEDWI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION

~ Equipment Identification No. J444, J463 TER No. N/A Sheet 1 of 1

?!/

-Preparer:

l-Date:

Independent Review: Y Date:

f5[l R4 v

v i

Approvali CdCAb%w Date:

3 /l / S4 i

\\

,w These junction boxes are in the electrical circuits for solenoid valves that are-required to deenergize to shut Reactor Building isolation dampers. The

. dampers shut immediately af ter low reactor water level, high drywell pressure or high radiation in refueling floor exhaust duct is sensed.

If the terminal

-block inside:the junction box fails af ter the dampers are shut, there will be no detrimental effects because the dampers are already shut.

In the unlikely event that.the terminal block fails prior to the solenoids deenergizing and the dampers shutting, the failure will simply speed up.the process of

' isolating the Reactor Building. Af ter the dampers have shut, there is no plausible failure of the terminal block that could reopen the valve.

-Therefore, continued operation is justified.

w

.- to NEDW1 No. 271 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. J561, J604, J874 J603, J553 TER No. N/A Sheet 1 of 1 Preparer:

Date:

/

I Independent Review: kl-Date:

8 / 84 U

U

. Approval:

(MJwm Date:

%/ I /M

( T v

The manufacturer of terminal blocks installed in these junction boxes has not yet been determined. However, Sandia National Laboratories and other labrratories have compiled extensive test data on terminal blocks (both protected and unprotected) of various manufacturers, which has shown that the

-pro ab bility of failure of the terminal blocks is very low. Sandia tested over 400 terminals in their own facilities.

The worst case temperature peaks at 238.l*F and immediately starts to

' decrease. Applying 238.1*F to Sandia's probability curves for 480v/100%RH/5

. hours, the probability of f ailure is only.0018, for protected terminal blocks in 120V circuits.

The installed terminal blocks may be required for longer than 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, but the Reactor Building temperature decreased to 132*F i

- within 90 minutes of the PBOC, thereby reducing the probability of failure even.further. Also, Sandia's research showed that most failures occurred in the initial phase of the tests.

Sandia found "that for TMI-2 accident conditions, the effect of radiolysis is negligible on surface conductivity and therefore on breakdown" and therefore the radiation in the Reactor Building will not increase the probability of

' failure. Based on the above information, continued plant operation is v

justified.

n

W' c to NEDWI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION

?

Equipment Identification No. J859 TER No. N/A-Sheet 1 of 1 E[/!2Y

. Preparer:

Date:

Ind pendent Review:

hb 8!/![M Date:

Approval:

-C

_A Date:

%)1/24 A

i V

The. insulation system used on the splices in this junction box are environmentally qualified.

The documentation packages for these splices are being completed and when completed, environmental qualification of the splices and this junction box will be proven.

Therefore, continued operation p.

is justified.

L

t, to NEDW1 No. 211 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment' Identification No. N912, N923 TER No. N/A Sheet 1 of.*L Preparer:

^-

Date:

7 7

'/

Independent. Review:

Date:

7/////f Approval:

tTED : - %

Date:

7/30/84 N

i

~

V N912 is a local start switch for P223 (Gland Seal Condenser Blower) and N923 is a local start switch for_ P220 (Gland Seal Condenser Condensate Pump).

These switches contribute to HPCIS turbine operation and are, therefore,

' required solely to assure satisfactory HPCIS operation.

The HPCIS is relied upon to operate during and tollowing n

' Loss of Feedwater Flow, L

Total Loss of Offsite Power, 1 Shutdown from Outside Control Room (Special Event),

Pipe Break Inside Primary Containment, Control Rod Drop Accident, and Pipe Break Outside Primary Containment

?

to assure continued core cooling, and thus mitigate consequences which could result in potential offsite exposures comparable to the 10CFR100 guidelines.

g e

None of the first three events listed above is expected to result in environmental conditions any more severe than those experienced during normal operation. The fourth event is addressed in the HPCIS Safety Evaluation, which states.that,."The HPCIS is designed to provide adequate core cooling for small breaks... core never uncovers and is continuously cooled throughout the transient so that no core damage of any kind occurs for breaks that lie within the range of the HPCI." Thus, the size ot LOCA presumed to generate postulated core damage is beyond the capacity of HPCIS to provide core

~. cooling.

The Control Rod Drop Accident has been evaluated and no HPCIS equipment will be subjected to pressure, temperature, radiation or humidity conditions any more severe than those experienced during normal operation.

Those pipe breaks outside containment which could be expected to result in harsh conditions of pressure, temperature and humidity in the vicinity ot

=these switches are the PBOC-3 and the PBOC-5.

Each of these events, however, incapacitates the HPCIS.

System operability is, therefore, not required for either PBOC, 1.

y to NEDWI No. 277 p

BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION l

Equipment Identification No. N912, N923

.TER.No.-N/A Sheet 2 of 2 7!A7!I'[

Preparer:

^

Date:

Independent'Rehiew:

50 Date:

7M///7 Approval:

OMd-w Date:

7/30/84

\\

s v

'On the other' hand, system operability is required for.the main steam line

. breaks, PBOC-7 and PBOC-8, either of which could result in cumulative 4

1

'_ radiation exposures to the switches well in excess of 10 ' rads.

These

. values are based ' conservatively on the postulated core damage of NUREG 0131 and NUREG 0588.

However, FSAR analysis of the PNPS design basis Main Steam Line Break Accident indicates that, with a maximum 10.5 second MSIV closure

.and continued core coverage (f rom normal or standby systems, including HPCIS), there would be no fuel damage. Without core damage, exposures will not exceed 104 rads.

' MSIV closure time is verified once per quarter under Technical Specification surveillance requirements. The closure time must be greater than 3 seconds and less'than 5 seconds for the valve to be considered operable.

For valve

-closure times shorter than 10.5 seconds, the postulated accident is considered less severe than that analyzed.

Core cooling systems are also verified operable periodically under plant surveillance requirements. Thus, if HPCIS must be declared inoperable as a consequence of the P80C, then ADS, LPCI and Core Spray are all assumed to be operable'to assure safe shutdown of the plant.

If all core cooling systems operate as' designed and tested, no fuel damage should occur.

Since the assumptions of NUREG 0737 and NUREG 0588 are considered unrealistic on this basis, failure of N912 and N923, as a consequence of excessive radiation exposure from the main steam line break accident, is considered highly improbable and continued operation is justified.

- ~ -. -. -

, to NEDWI No. 277 y

BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. N921 TER No. N/A.

Sheet 1 of a 7!O!IY Preparer:

Date:

. Independent' Review:

)b v

u

__ 7 bob Date:

Approval:

1 # d ') e w Date:

7/30/84 Q

~

N921 is-a local switch for P229 (HPCI Turbine Auxiliary Oil Pump).

P229 is only-required during turbine startup and shutdown.

The HPCIS'is relied upon to operate during and following Loss of Feedwater Flow,.

-Total. Loss of Offsite Power, Shutdown from Outside Control Room (Special Event),

Pipe Break Inside Primary Containment, Control Rod Orop Accident, and Pipe Break Outside Primary Containment to assure continued core cooling, and thus mitigate consequences which could resultiin potential of f site exposures comparable to the 10CFR100 guidelines.

None of the first three events listed above is expected to result in environmental conditions any more severe than those experienced during normal operation'. The fourth event is addressed in the HPCIS Safety Evaluation, which states that, "The HPCIS is designed to provide adequate core cooling

(

for small breaks... core never uncovers and is continuously cooled throughout the. transient so that no core damage of any kind occurs for breaks that lie

, within the range of the HPCI." Thus, the size of LOCA presumed to generate postulated core damage is beyond the capacity of HPCIS.to provide core

' cooling' The Control Rod Drop Accident has been evaluated and no HPCIS equipment will be subjected ~ to pressure, temperature, radiation or humidity conditions any

.more severe than those experienced during normal operation.

Y' Those pipe breaks outside containment which could be expected to result in harsh conditions of pressure, temperature and humidity in the vicinity of this switch is the PBOC-3 and the PBOC-5.

Each of these events, however,

^

_ incapacitates the HPCIS. System operability is, therefore, not required for

either PBOC.

g

Attachm:nt 5 to NEDWI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment identification No. N921 TEP. No. N/A Sheet 2 of ;L 7!3

[

Preparer:

Date:

7flofs4 Independent Review: kb%+

Date:

y U

i i

Approval:

QCdDw Date:

7/30/24 (T

On the other hand, system operability is required for the main steam line b/eaks, PB0C-7 and PBOC-8, either of which could result in cumulative 4 rads.

These values radiation exposures to the switch well in excess of 10 are based conservatively on the postulated core damage of NUREG 0131 and NUREG 0588. However, FSAR analysis of the PNPS design basis Main Steam Line Break Accident indicates that, with a maximum 10.5 second MSIV closure and continued core coverage (from normal or standby systems, including HPCIS),

- there would be no fuel damage. Without core damage, exposures will not

- exceed 104 rads.

MSIV closure time is verified once per quarter under Technical Specification surveillance requirements.

The closure time must be greater than 3 seconds and less than 5 seconds for the valve to be considered operable.

For valve closure times shorter than 10.5 seconds, the postulated accident is considered less severe than that analyzed.

Core cooling systems are also verified operable periodically under plant

. surveillance requ rements. Thus, if HPCIS must be declared inoperable as a i

consequence of the PB0C, then ADS, LPCI and Core Spray are all assumed to be operable to assure safe shutdown of the plant.

If all core cooling systems

,)-

operate as designed and tested, no fuel damage should occur.

Since the assumptions of NUREG 0737 and NUREG 0588 are considered unrealistic on.this basis, failure of N921, as a consequence of excessive radiation exposure from the main steam line break accident, is considered highly improbable and continued operation is justified.

o a to NEDW1 No. 217 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. CS42-1821, CS42-1822 TER No. N/A Sheet 1 of 1 7/M!2'[

Preparer:

Date:

7////W Independent Review:

d/ b/sw-Date:

f a

r Approval:

GCO mm Date:

7 l 2h/ E4

~-

( T

~

,The-functional requirement of these switches (that control VAC201) is that normally closed contacts internal to the switches remain shut.

The switches are mounted in an enclosed control panel.

The non-metallic portion of the

' switch is mide of Dupont Delrin.

The only way the contacts could open would be for catastrophic f ailure of the Delrin. The' parameters that could cause catastrophic failure, would be temperature (Delrin softening or embrittling) or radiation (Delrin disintegrating).

The radiation to which the switch might be subjected is 1.6

- x.105 rads, but it has been tested to 1 x 106 rads, therefore radiation is not a problem. The temperature due to the worst case postulated break is

.238.1*F, 24.5 seconds into the accident, and considering that Delrin has been

-tested to a much higher temperature (311*F) temperature is not a problem.

Therefore, continued operation is justified.

L

{

g:

1. to NEDW1 No. 2?/

BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment identification No. C61A, C618 Indicating Lights

-TER No. N/A Sheet 1 of 1 7!3#!IEY Preparer:

7 Date:

7kokM Independent Review: N MA+

Date:

! Approval:

L Date:

/30/S4 G

These lights are for indication at the local control panels.

They are not necessary for operation of the fans, however, failure could af fect the

~

control circuit and therefore degrade operation of their respective area cooling fans. There are three possible failure modes; open, short internal to the light, or short to the panel.

-If a light fails open there will be no effect on the circuit and the fan will continue to operate normally.

~ 1f'a light develops an internal short circuit there will be no ettect on the circuit and the f an will continue to operate normally.

If a light develops a short to the panel the control circuit will be disabled and'the fan will be deenergized. However, the only possible failure mechanism would be excessive moisture inside the panel such that the water created a path for current from the light connections to the panel.

This L

failure mechanism is extremely unlikely because the panels are gasketed,

-which will reduce the moisture inside, and because of the distance between the electrical connections and the panel.

The distance is enough so that creation of an electrical arc between the connections and the panel is extremely small. Also, if one of the fans becores disabled it is highly unlikely that this would cause a significant effect on the equipment in the af fected room, because the RHR rooms (areas 1.1 and 1.2) have redundant f ans.

Based on the above information, continued operation is justified.

n y

' to NEDWI No. 21/

BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION t

Equipment.ldentification No. C2257A

'TER No. N/A Sheet 1 of 1 17 [5 I Preparer:.

O Date:

Independent Review:

d/ 8 ^ r Date:

7h

. Approval:

C9 d

Date:

7I2fo/24

( N w

The manufacturer of one of the terminal blocks within this local control panel has not yet been determined. However, Sandia National Laboratories and other laboratories have compiled extensive test data on terminal blocks (both

' protected and unprotected) of various manuf acturers, which has shown that the p'robability of failure of the terminal blocks is very low.

Sandia tested

~ over 400 terminals-in their own facilities.

p e

.The. worst case temperature peaks at 228.7*F and immediately starts to

' decrease. Applying 228.7'F to Sandia's probability curves for 480v/100%RH/5 hours, the probability of failure is only.0018, for protected terminal

. blocks in 120V circuits.

The installed terminal blocks may be required for longer than 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, but the Reactor Building temperature decreased to 119'F within 60 minutes of the P80C, thereby reducing the probability of failure even further. Also, Sandia's research showed that most failures occurred in the initial phase of the tests.

Sandia found "that for TMI-2 accident conditions, the effect of radiolysis is

' negligible on surface conductivity and therefore on breakdown" and therefore the radiation in the Reactor Building will not increase the probability of failure.

The manufacturer of the instrument rack wire from the terminal blocks AA and D0 to DPIS2353 and DPIS2352, respectively,-is unknown.

The wire is in a conduit from'the switches to the enclosure for the terminal block.

The

. differential pressure switches are only required to provide a signal (which

~

" seals in") to the Primary Containment Isolation Control System during a Main Steam Line Break in the Steam Tunnel (PB0C-8).

The only environmental changes in the location of C2257A during the 1 minute that the pressure switches are required,'are a pressure spike to 15.2 psia and a temperature of 115.9"F.

These " mild" conditions will not add any abnormal stresses to the wire and the wire is expected to survive the accident.

Based on-the above information, continued plant operation is justified.

p

i. to NEDWI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment. Identification No. P229 TER No.'N/A Sheet 1 of 2 3

7!2 7!E'f

' Preparer:

Date:

~

7.,/7//IY Independent Review:

/M

/Jfd*4+7 Date:

Approval:

Ndlw Date:

~7/30/24 3

o P229 is' the HPCI Turbine Auxiliary Oil Pump.

It is only required during turbine startup and shutdown.

The HPCIS is relied _upon to operate during'and following

- Loss of F,eedwater Flow, Total Loss of Offsite Power,

' Shutdown f rora Outside Control Room (Special Event),

Pipe Break Inside Primary Containment, Control. Rod Drop Accident, and Pipe Break Outside Primary Containment to assure continued core cooling, and thus mitigate consequences which could result in potential offsite exposures comparable to the 10CFR100 guidelines.

None of the first three events listed above is expected to result in environmental conditions any more severe than those experienced during normal operation. The fourth event is addressed in the HPCIS Safety Evaluation, which states that, "The HPCIS is designed to provide adequate core cooling

.for small breaks... core never uncovers and is continuously cooled throughout the transient so that no core damage of any kind occurs for breaks t'1at lie within the range of the HPCI." Thus, the size of LOCA presumed to generate postulated core damage is beyond the capacity of HPCIS to provide core cooling.

The Control Rod Drop Accident has been evaluated and no HPCIS equipment will be subjected to pressure, temperature, radiation or humidity conditions any more severe than those experienced during normal operation.

Those pipe breaks outside containment which could be expected to result in harsh conditions of pressure, temperature and humidity in the vicinity of this pump is the PBOC-3 and the PBOC-5.

Each of these events, however, incapacitates the HPCIS. System operability is, therefore, not required for either PBOC.

I

hl to NE0WI No. 271 n

BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION

.g

-Equipment identification No. P229 1ER No. N/A Sheet 2 of llt 7/d 7!E'f Preparer:

Date:

Independent Review:

[1] hm 7/2WY Date:

. Approval:

CTkhw Date:

7/36/84.

~

Q On the other hand, system operability is required for the main steam line breaks, P80C-7 and P80C-8, either of which could result in cumulative

' radiation exposures to the pump well in excess of 104 rads.

These values are based conservatively on the postulated core damage of NUREG 0731 and NUREG 0588. However, FSAR analysis of the PNPS design basis Main Steam Line i

Break' Accident indicates that, with a maximum 10.5 second MSIV closure and continued core coverage (from normal or standby systems, including HPCIS),

'there would be no fuel damage. Without core damage, exposures will not exceed.104 rads, e

MSIV, closure time is verified once per quarter under Technical Specification surveillance requirementsi' The closure time must be greater than 3 seconds O

and less than 5 seconds for the valve to be considered operable.

For valve s

1 closure times shorter than 10.5 seconds, the postulated accident is

, considered'less severe than that analyf'ed.

'p /

g s

~

"C re cooling systems are also verified operable periodically under plant 9s6fveillance requirements.

Thus, if HPCIS must be declared inoperable as a

^-

consequence of the P80C, then ADS, LPCI and Core Spray are all assumed to be t

1 opbrable.to assure safe shutdown of the plant.

If all core cooling systems

-. operate as designed and tested, no fuel damage should occur.

s b F

Since the assumptions of NUREG 0737 and NUREG 0588 are considered' unrealistic on-this basis, failure of P229, as a consequence of excessive radiation

. exposure from the main steam line break accident, is considered highly improbable and continued operation is justified.

w

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^l[ to'NEDWI No. 271 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION I

6

+

Equipment' identification 'No. PS2390A, PS23908 ITER No.,N/A.

Sheet 1 of.1 7!30

'[

' preparer:

Date:

!!ndependent Review: ) L b*

Date:

7[30[fg

.(f v

,u a

oApproval:

.. CTdh Date:

'l/30/f4 f

r7 iX w

~

.These pressure switches provide a condensate storage tank level signal to

'HPCI valve control to'open M02301-35, 36..M02301-35, 36 also serve as HPCI isolation. valves.

The HPCI isolation signal overrides the Condensate Storage ciank level signal'provided by these pressure switches.

Therefore, PS2390A, B i

care required. solely to' assure satisfactory HPCIS operation.

The HPC15 is relied upon to operate during and following Loss'of.Feedwater Flow; Total Loss of Offsite Power, Shutdown from Outside Control Room (Special Event),

Pipe Break Inside' Primary Containment, n

4

g Control Rod Drop Accident, and a

6

Pipe Break-Outside Primary Containment to assure continued core cooling, and thus mitigate consequences which could g

result in potential offsite exposures comparable to the 10CFR100 guidelines.

None of'the first three events listed above is expected to result in

environmental conditions any more severe than those experienced during normal t

-operation. The fourth event is addressed in the HPCIS Safety Evaluation, which states 'that, "The HPCIS is designed to provide adequate core cooling

for small breaks... core never uncovers and is continuously cooled throughout the transient so that no core damage of any kind occurs for breaks that lie within the range of the HPCI.": Thus, the size of LOCA presumed to generate postulated core damage-is beyond the capacity of HPCIS to provide core cooling.

'The Control Rod Drop Accident has been evaluated and no HPCIS equipment will be subjected to pressure, temperature, radiation or humidity conditions any more severe than those experienced during normal operation.

Those pipe breaks outside containment which could be expected to result in harsh conditions of pressure, temperature and humidity in the vicinity of these pressure switches are the P80C-3 and the PB0C-5.

Each of these events, i

however,' incapacitates the'HPCIS. System operability is, therefore, not req 11 red for either PBOC.

4 4

. ~., - -,. --

... _, _ ~, _, _ _,, _ _.

.n. _ -.. _.-_. _. _, n - -,., _

o y-

J' n

n i

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• to NEDWI No. 277

• d DOSTON. EDISON COMPANY 3USTIFICATION FOR

...,e.

/

/

CONTINUER OPERATION

'V

? g Equipment. Identification No. V32390A, PS'23908

Tf.R No. N/A-

, Sheet 2 of 2

~

,,y 7!3o/;71 Preparer

Date:

N1deperhentReview:

)S-hw Date:

To/M J

Q U

^

E Approval':

CROD%

Date:

~7/30/24 tT 9

.J

~

On the otherrnar:Q/ system operability is rcquired for the main, steam line Ei ' breaks, P800-W)ou -PB0b8, either'of which could result to cumuiative

.radiationexposu/evtothepanelswellinexcessof10irads.~IheseValues s

are based conservatively on the postuiated ccre dacage-of NUREG 0737 and NUREG[0588.fHowever,FSARanalysisofthePNPS.designbasisMainSteamLine

~

~ Break Accident indicates that, with a majmum 10.5 second MSIV closure and continued core < coverage (f rom normal or~ standby systems, including HPCIS),

ai there would be4no fdel damage. Without core damage, exposures sill not 4

exceed 104 r$ i.,.> -

1 4

/

.MSIV closure)fub is verified once per quarter under Technical Specification surveillance requirements. The closure timg must be greater than 3 seconds and less than 5 seconds for the valve to be considered' operable.

For valve closure times shorter than 10.5 seconds, the postulated accident is "cqhsidered less severe than that analyzed.

Core cooling systems are also verified @erable periIdically under plant surseillance requirements. Thus, if HPCIS must be declared inoperable as a do~risequence of the PBOC, then ADS, LPCI and Core Spray are all assumed to be T

operable to assure safe shutdown of the plant.

If all core cooling systems operate as designed and tested, no fuel damage should occur.

-M nce the assumptions of NUREG 0737 and NUREli 0588 are considerej unrealistic on this basis, failure of PS2390A, B as s chhsequence of excessive radiation r -

exposure from the main steam lire break acddent, is considered highly improbable and continued operation is justified.

8 vf

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0 to NEDW1 No. 277 BOSTON EDISON COMPANY JUST1FICATION FOR CONTINUED OPERATION i

h Equipment' Identification No. C2303, T2303 TER No. N/A Sheet 1 of 2 Preparer:

Date:

7 V!EY Independent Review: h Date:

7[2V Approval:

(tab %

Date:

9 I Tl / E4-a fl These panels-contribute to HPCIS turbine start-up and speed control and are, therefore, required solely to assure satisfactory HPCIS operation.

The HPCIS is relied upon to operate during and following Loss of Feedwater Flow, Total = Loss of Offsite Power,

-Shutdown from Outside Control Room (Special Event),

Pipe Break Inside Primary Containment,

. Control Rod Drop Accident, and Pipe Break Outside Primary Containment

~

to. assure continued core cooling, and thus mitigate consequences which could result in potential of fsite exposures comparable to the 10CFR100 guidelines.

None ofEthe first three events listed above is expected to result in environmental conditions any more severe than those experienced during normal operation.

The fourth event is addressed in the HPCIS Safety Evaluation, which states that, "The HPCIS is designed to provide adequate core cooling for small breaks... core never uncovers and is. continuously cooled throughout the transient so that no core damage of any kind occurs for breaks that lie within the range of the HPCI." Thus, the size of LOCA presumed to generate postulated core damage is beyond the capacity of HPCIS to provide core cooling.

The-Control Rod Drop Accident has been evaluated and no HPCIS equipment will be subjected to pressure, temperature, radiation or humidity conditions any more severe than those experienced during normal operation.

Those pipe breaks outside containment which could be expected to result in harsh conditions of pressure, temperature and humidity in the vicinity of these panels are the PB0C-3 and the PBOC-5.

Each of these events, however, incapacitates the HPCIS. System operability is, therefore, not required for either P80C.

7 to NEDWI No. 2/7 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. C2303, T2303 TER No. N/A Sheet 2 of a

2. Yh Preparer:

Date:

Independent Review:

Date:

7/WMI Approval:

NO e o t Date:

7/T1/M

(

T w

On the other hand, system operability is required for the main steam line breaks, PBOC-7 and PBOC-8, either of which could result in cumulative radiation exposures to the panels well in excess of 104 rads. These values are based conservatively on the postulated core damage of NUREG 0731 and NUREG 0588. However, FSAR analysis of the PNPS design basis Main Steam Line Break Accident indicates that, with a maximum 10.5 second MSIV closure and continued core coverage (from normal or standby systems, inciuding HPCIS),

there would be nc fuel damage. Without core damage, exposures will not exceed 104 rads.

MSIV closure time is verified once per quarter under Technical Specification surveillance requirements. The closure time must be greater than 3 seconds

~ g and less than 5 seconds for the valve to be considered operable.

For valve fW closure times shorter than 10.5 seconds, the postulated accident is W7!"

considered less severe than that analyzed.

Core cooling systems are also verified operable periodically under plant surveillance requirements. Thus, if HPCIS must be declared inoperable as a

' consequence of the PBOC, then ADS, LPCI and Core Spray are all assumed to be s

operable to assure safe shutdown of the plar.t.

If all core cooling systems c,

i operate as designed and tested, no fuel damage should occur.

f..

-Since the assumptions of NUREG 0737 and NUREG 0588 are considered unrealistic "O

-on'this. basis, failure of C2303 and 12303, as a consequence of excessive radiation exposure from the main steam line break accident, is considered highly improbable and continued operation is justified.

b I

y to NE0WI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Touloment Identification No. J623. J624. J625. J626 TER No. N/A Steet 1 of 2 Preparer:

Date:

/ [U

~

Efl Independent Review: -

Date:

Approval:

m Date:

8/2 %

.s

~

These junction boxes are in the electrical circuit of t' e outboard MSIV h

control modules.

The MSIV's are relied upon to function to assure reactor vessel and primary containment isolation, and thus mitigate consequences which could result in potential ~offsite exposures comparable to the 10CFR100 guidelines, during the "following translents:

Pressure Regulator Failure, Loss of Feedwater Flow, Control Rod Drop Accident, Pipe Break Inside Primary Containment, and Pipe Break Outside Primary Containment None of the first three events listed above is expected to result in environmental conditions any more severe than those experienced during normal operation. Also, based on FSAR analyses and event profiles, no Pipe Break

.Inside Primary Containment is expected to result in conditions of pressure, temperature and humidity which are any more severe in the vicinity of these outboard MSIV's than those experienced during normal operation.

Of the latter two events listed above, the PBOC with core damage generates the most. severe conditions of radiation for the control modules. However, the MSIV's will receive the automatic isolation signal within 500 milliseconds'of the pipe break. This is more conservative than for the PBIC. -Since no electrical equipment within the valve control modules will be required to function subsequent to closure initiation, it is highly improbable that accident doses will prevent MSiv closure for required events.

' Only the PBOC-7, PB0C-8 and PBOC-9 are expected to result in harsh conditions of pressure, temperature and humidity in the vicinity of the junction boxes.

These conditions are not expected in the vicinity of the respective inboard

'MSIV control modules. These valves are tested periodically under controlled Technical Specification surveillance requirements and therefore, there is reasonable assurance that they will perform as desired.

It is therefore assumed that, should the junction boxes be made inoperable, the required

. containment isolation would be accomplished satisfactorily by A0 203-1A/D.

(

,, to NEDWI No. 277 BOSTON EDISON COMPANY

-JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. J623, J624, J625, J626 TER-No. N/A Sheet 2 of 2 d!'/E'/

Preparer:

Date:

Independent Review: k[ bMS N! / 84 Date:

LApproval:

Date:

8/2 /R4

. T v

a r

. Based on all of the above, continued operation is justified.

- 1.

N T

s n.:

.t-L

(

~ to NEDW1 No. 277

' BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION j

Equipment Identification No. J538, J539

-TER No. N/A.

Sheet 1 of 1 Nk Date:

J' W 1

Preparer:

7[f!M 8

1. A' Date:

Independent Review: -

r T d h m a Date:

7/% /E4-Approval:

(

NJ 1

Tnese junction boxes are in the electrical circuit for relative humidity

sensors (HS-1A,' 2A, 3A, 4A, IB, 28, 38, 48).

The relative humidity sensors

'are not required for Standby Gas Treatment System (SGTS) Operation The normal' function of the sensors is to detect high humidity in the SGTS inlet and energize relays, which in turn cause the heater relays and heaters to be V

energized.. The humidity controls have been bypassed, so that fulI heater operation is initiated upon operation of the SGTS exhaust f an.

Therefore, these junction boxes are not required and continued plant operation is

' justified.

J s

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,.4--

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y to NEDW1 No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR

_ CONTINUED OPERATION Equipment Identification No. C150

'TER No. N/A-Sheet 1 of 6 7!3Df84 Preparer:

hL kV.A Date:

l "7b[F &

EY

Independent Review:

Date:

Approval:

(Ttbm Date:

7 / 30/FA

( T

' Remote shutdown panel C150 is located in the "B" RBCCW room (zone 1.22) and is exposed to a harsh superheated steam environment (peaking at 300.9'F and 15.2 psia) following a PBOC-3 (HPCI steam line break in the HPCI pump room).

The panel contains series 40 Electroswitch Control Switches and GE Model ET-16 indicating lights.

Failure of these components could cause loss of control to the "D" and "E" RBCCW and salt service water pumps. A reanalysis of the PBOC-3 environment for this room is presently being performed and it Lis' expected to indicate a significantly less harsh environment than is a

presently assumed.

In the.mean time, continued operation car be justified on

the following basis.

Series 40 Electroswitch

o

-Temperature Temperature tests have been successfully conducted by Electroswitch on Series 24 (Report No. 2392-2) and Series 40 (Report No. 2392-14) switches.

The 1 tests were conducted at 176*F (80*C) for 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br />.

Proper operation of the switches was verified before and af ter the temperature exposure.

For this application the maximum accident temperature is 300.9'F which exceeds the

<l?6*F test temperature, for 45 minutes.

These switches are located inside an

" essentially leak tight NEMA-12 enclosure (unvented) which will cause the

-temperature experienced by the switches to lag the accident temperature experienced by the enclosure. Tests have been conducted by Wyle Laboratories on similar sized cabinets '(except with vents) which characterized the

' internal temperature of the cabinets as a function of time in a LOCA environment.

Results of these-tests (Wyle Report No. 44439-2) show that the internal temperature of the vented cabinets lagged the external temperature by a minimum of-50*F during the first 15 minutes of exposure to a saturated steam environment.

In that test the temperature and pressure were rapidly (within Lapproximately 10 seconds) ramped to 54 psig and 280*F (minimum) respectively. Because the pressure for this application is much less than the pressure for the test-(0.5 psig versus 54 psig) and in light of the

' essentially leak tight nature of the enclosure, the hot environment is expected to be_ essentially precluded from entering the panel.

It is therefore our engineering judgement that in a similar test of the unvented

to NEDWI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION Equipment Identification No. C150

-TER No. N/A Sheet 2 of 6 Preparer:

\\L hsc.n-Date:

7 t4 7/3#!D[

LIndependent Review:

Date:

Approval:

Tdhm Date:

~1/ 30 /84 0

cabinets to the same maximum temperature but significantly lower pressure that the internal temperature of the cabinet would lag the external temperature by substantially greater than the 50*F experienced in the test.

.Further, in the tests conducted by Wyle, varied components (examples:

-pressure transmitter and solenoid valve) were installed in the cabinet and their mass temperature was recorded in the test.

The temperature of a

- typical component-(pressure transmitter) mounted in the vented cabinet lagged the accident temperature by approximately 80*F af ter the first 15 minutes of the test.

It should be noted that saturated steam blanketing of high thermal inertia components such as the cabinet, as demonstrated in Limitorque Report B0027, would result in the accident environment being the equivalent of an exposure to 215'F to the exterior of the cabinet. The interior of the cabinet would be expected to lag 50 - 80*F minimum below this.

In the Electroswitch test, the. switches were maintained at 176*F for 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br />.

Based.on the above tests and engineering rational, it is judged that the test

' temperature of 176*F is_ comparable to the temperature which the switches would experience in the accident condition.

Therefore, the switches are

' judged suitable for use in the temperature application.

o Humidity The compartment within which these switches are mounted experiences 100% RH for a short time period.

The humidity then decays rather quickly to a long term equilibrium of approximately 60%.

Due to the essentially leak tight nature of the NEMA-12 enclosures,-the switches should not experience the

. initial humidity spike. Maximum voltage on the switches is 120 VAC. Wyle Laboratories has tested a variety of switches and terminal blocks at humidity conditions in the range of 90% to 100% including some LOCA tests.

In general, no problems have been experienced for these conditions where voltage never exceeds 120-volts unless the items experienced deformation resulting from temperature. -Operation of the switches at the temperature conditions is

' justified in the above paragraph. Also, Electroswitch has subjected the switches to 95% RH for 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />, unpowered. Operation of the switches was satisfactory in functional tests conducted prior to and following the humidity test. Therefore, the switches are judged suitable for use in the humidity environment.

[

i to NEDWI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR I

CONTINUED OPERATION

. Equipment Identification No. C150 TER No. N/A Sheet 3 of 6 Mbh Date: 7-70-74 Preparer:

2/3d!E'[

~

Independent Review:

Date:

~

Approval:

CSC(bm Date:

W30 /84

~

T

.v o

Pressure The maximum pressure which the switches would be exposed to in an accident is 15.2 psia (0.5 psig).

The configuration of the switches is such that they

'will not entrap air or otherwise cause a pressure imbalance which would result in inadvertent actuation of the switches.

Therefore the switches are judged suitable for use in this pressure environment.

o Radiation The maximum radiation which the switches will experience is approximately 1.8 x'103 rads. Electroswitch Test Report No. 3030-1 documents satisf actory operation of the switches following a radiation exposure of 1 x 101 rads.

Therefore, the switches are judged suitable for use in the radiation environment.

.o Aging Conditions of-aging were evaluated using the Arrhenius technique.

Based on

'the analysis'which considered all nonmetallic materials within the switch, an

.cstimated life in excess of 40 years was established.

Therefore, continued operation is justified.

A

.=

s to NEDWI No. 277

^

BOSTON EDISON COMPANY

..h e JUSTIFICATION FOR CONTINUED OPERATION

'Equipmentildentification No. C150

TER No..N/A Sheet 4 of 6 Preparer:

bW Date:

E D ~84 7[3o/FY

. Independent. Review:

Date:

~

Approval:

GObh w '

Date:

7/30/M

^

i T v

~

lGE' Model ET-16 Lichts

~*

o' nTemperature

. Temperature tests have been successfully conducted by Wyle on ET-16 lights.

The tests were conducted 'at 160*F.

Proper operation of the lights was e

verified before' and af ter the. temperature exposure.

For this application the maximum' accident temperature is 300.9'F which exceeds the 160*F test temperature for 75 minutes. These lights are located inside an essentially leak tight NEMA-12 enclosure (unvented) which will cause the temperature

. experienced by the. lights to lag the accident temperature experienced by the y-conclosure. Tests have been conducted by Wyle Laboratories on similar sized Lcabinets (except with vents).which characterized the internal temperature of L.the cabinets as a function of, time in a LOCA environment.

desultsofthesetests'(Wyle.-ReportNo.'44439-2) show that the internal itemperature of the--vented cabinets lagged the external temperature by a

' minimum of. 50*F during the first.15 minutes of exposure to a saturated steam

~

'^

environment.

In that test the temperature and pressure were rapidly (within

.approximately 10 seconds) ramped to 54 psig and 280*F (minimum) respectively.. Because the pressure for this application is much less than

'the pressure for the test (0.5 psig versus 54 psig) and in light of the

-essentially leak tight' nature of the enclosure, the hot environment is

- Lexpected to be essentially precluded from entering the panel.

It is

theref ore'our engineering jodgement that-in a similar test of the unvented

^

cabinets to the-same maximum temperature but significantly lower pressure

that the' internal-temperature of the cabinet would lag the external

temperature by substantially greater than the 50*F experienced in the test.

Further, in the tests conducted by Wyle, varied components (examples:

. pressure. transmitter and solenoid valve) were installed in the cabinet and

~

their' mass temperature was recorded in the test.

The temperature of a i

typical component (pressure transmitter) mounted'in the vented cabinet lagged the accident ~ temperature by approximately 80*F af ter the first 15 minutes of

-the. test.- It should be noted that saturated steam blanketing of high thermal inertia components such as the cabinet, as demonstrated in Limitorque Report L

e s

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mwe'e-re uqe e w&s-e-to NEDWI No. 217 BOSTON EDISON COMPANY JUSTIFICATION FOR I

CONTINUED OPERATION Equipment Identification No. C150

.TER No. N/A Sheet 5 of 6 Preparer:

hL bw Date: 7'70-MY 7 3 !E7' Independent Review:

k

~

Date:

, Approval:

' CXC[hw Date:

7/30/S4 j

(

B0027, would result in the accident environment being the equivalent of a

'215* saturated' steam exposure to the exterior of the cabinets.

The interior of.the' cabinet would be. expected to lag 50*F to 80*F below this.

In the Wyle

' test, ~ the lights'were maintained at 160*F.

Based on the above tests and engineering. rationale, it is judged that the test temperature of 160*F is comparable to the temperature which the lights would experience in the accident: condition.

Therefore, the lights are judged suitable for use in the

,..nperature application.

o Humidity.

~ '

-The compartment within which this cabinet is mounted experiences 100% RH tor

'2 a'short time period.

The humidity then decays rather quickly to an

equilibrium of approximately 60%.

Due to the leak tight nature of the NEMA-12 enclosures, the lights should not experience the initial humidity spike. Maximum voltage on the lights is 120 VAC. Wyle Laboratories has u

tested a. variety of lights at humidity conditions in the range of 90% to 100%.

In general, no problems have been experienced for these conditions where voltage never exceeds 120 volts unless the items experienced deformation resulting from temperature. ' Operation of the lights at the

-temperature (.onditions -is justified in the above paragraph. Therefore, the

. lights:are. Judged suitable for use in the humidity environment.

o.

Pressure

The. maximum pressure which the lights would be exposed to in an accident is 15.2 psia (0.5'psig). The configuration of the lights is such that they will not entrap air or otherwise cause a pressure imbalance which would result in a functional disparity in the lights.

Therefore the lights are judged suitable for use in this pressure environment.

L

~

h.;5:.

' to NEDWI No. 217 BOSTON EDISON COMPANY JUSTIFICATION FOR

~

CONTINUED OPERATION

squipment Identification No. C150 TER No.- N/A Sheet 6 of 6

- 1 Preparer:

[

W Date:

7- $0 'b 7!30fDf Independent' Review:

Date:

Approval:

C3Ckh Date:

7/30/24 I N v

o.

Radiation The~ maximum radiation which the lights will experience is approximately 1.8 x

-: 103 rads.

Proprietary Wyle Test Report No. 45625-1A documents satisfactory Joperation of the lights following a radiation exposure of 2.1 x 106 rads.

Therefore, the lights are < judged suitable for use in the radiation environment.

Based on the above information, continued plant operation is justified.

I' T

TW-""w "Pe e

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[ to NEDWI No. 277 BOSTON EDISON COMPANY JUSTIFICATION FOR CONTINUED OPERATION

. Equipment identification No. Ring Tongue Terminations (<4KV) Outside Containment TER No. M/A Sheet 1 of 2 Preparer:.

h Orm Date: '7b 84

- Independent Review:

w Date:

/

Y

^ Approval:-

(SCIh Date:

E/ I / %

\\

1v

.According'to.Wyle Laboratories Corrective Action Report No. 47066-TER-1, the installed ring tongue terminals include both insulted and non-insulated models f rom a variety 'of manuf acturers. The insulation materials used on insulated models has not been specifically identified.

The commonly used Einsulation materials for this application are nylon, PVC, PVF, a'nd PVDF.

Justification for continued operation is required as specific qualification (tests'do not exist.

Uninsulated ring tongue _ terminals are not susceptible to degradation or environmentally induced failure at the levels of stress produced by the

~

. environments at the Pilgrim I ' plant.

Failure of these interfaces is a

. function _'of installation _ configuration and terminal design.

- ; Insulated ring tongue terminals are supplied with an insulating material

^

covering'the barrel of the' terminals.

This insulation is provided to prevent

. bare-metal f rom' protruding beyond the terminal-block or connection to which

-it is fastened, thus reducing the hazard of. shock to personnel and a possible shorting path.between adjacent terminals and equipment. At the voltage-levels of-these terminations, the physical presence of any of the industry standard insulating materials is sufficient to perform this function.

The ei.vironments which could cause significant' insulation deterioration in-the Pilgrim plant _are temperature and radiation.

Degradation induced by

'these environments takes the form of material softening, material embrittlement..' increased compression set, loss of elongation capability, or Jcracking when subjected to bending stresses or dynamic loads. None of these

degradation mechanisms will impact the physical barrier. insulation capability J

of-the materials in their static termination application.

c

.The justification discussed above has been substantiated by the application t

of; numerous.tenninal lugs in nuclear equipment qualification tests. While these tests were not specifically designed to qualify the terminals and the

~

'models do not necessarily correlate with Pilgrim installed lugs, the tests demonstrate that in typical plant environments, neither insulated nor non-insulated terminal' lugs constitute a significant potential failure mechanism. Samples of tests which included representative terminals as part

.of the test specimen or part of the test equipment are Wyle 45603-1, Wyle 45638, Franklin C5257, Wyle 43703, Wyle 44282, Wyle 44300, Franklin C5022.

m

, Let to NE0WI No. 277 BOSTON EDISON COMPANY 1

I JUSTIFICATION FOR CONTINUED OPERATION p'

. Equipment Identification No. Ring Tongue Terminations (<4KV) Outside

. Containment' li TER No. N/A Sheet 2 of 2 Preparer:

b 2h Date:

~7!31$4

&b /8T in' dependent Review: )2k Date:

~

Approval:

N Date:

K /7 /84s

/ ~ \\

w Based on the above, continued operation with existing ring tongue terminals is justified.

tr*

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