ML19322E652
| ML19322E652 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/11/1980 |
| From: | Herbein J METROPOLITAN EDISON CO. |
| To: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| IEB-79-01B, IEB-79-1B, NUDOCS 8004020056 | |
| Download: ML19322E652 (33) | |
Text
_. _ _ -
Wc Metropolitan Edison Company j{Q
('
Middletown, Pennsylvan,a 17057 f
Post Office Box 480 i
717 944-4041 Writer's Direct Dial Number March 11, 1980 TLL 111 Office of Inspection and Enforcement Attn: Boyce H. Grier, Director Region I U. S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pa.
15 ^6
Dear Sir:
Three Mile Island Nuclear Station, Unit I (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Response to Bulletin 79-01B Enclosed please find the initial response to IE Bulletin 79-OlB, " Environmental Qualification of Class lE Equipment." The schedule for subsequent submittals is given in enclosure 1.
Sincerely,
~
/
7 J. G. Herbein Vice President D
Nuclear Operations JGH:CFM: hah Enclosures i
cc:
J. T. Collins 8004020 0 56 Q Metrcpohtan Edson Cor oany is a Memoer of rne General Putlic Ut6t:es System
METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY AND PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION, UNIT I Operating License No. DPR-50 Docket No. 50-289 This letter is submitted in support of the Nuclear Regulatory Commission request concerning IE Bulletin 79-OlB, " Environmental Qualification of Class lE Equipment," dated January 14, 1980 for Three Mile Island Nuclear Station, Unit I.
As part of this response a preliminary report on the Th.ca. Mile Island Unit I IE Bulletin 79-01B program is attached.
- Further, all statements contained in this report have been reviewed and all such stateme its made and matters set forth therein are true and correct to the best of my knowledge, information and belief.
METROPOLITAN EDISON COMPANY s
By/
O Vice President Sworn and subscribed to me this lith day of Januarv
, 1980.
1
/h 1
By W h'!
'I /' <<<
No,tary Publicr CATHY L B;;EY. N:ta y PuMic LeNen< ferry ina.. Caus.rt C n:y Pa.
6 Cc.m. sacn Ec res Oct. 24, in3 1
TLL 111 ITEM 1.
Provide a " master list" of all Engineered Safety Feature Systems (Plant Protection Systems) required to function under postulated accident conditions. Accident conditions are defined as the LOCA/HELB inside con-tainment, and HELB outside containment. For each system within (including cables, EPA's terminal blocks, etc.) the master list identify each Class lE electrical equipment item that is required to function under accident conditions. Pages 1 and 2 of Attachment 2 are standard formats to be used for the " master list" with typical information included.
Electrical equipment items, which are components of systems listed in Appendix A of Attachment 4, which are assumed to operate in the FSAR safety analysis and are relied on to mitigate design basis events are considered within the scope of this Bulletin, regardless whether or not they are classified as part of the engineered safety features when the plant was originally licensed to operate.
The necessity for further up grading of nonsafety-related plant systems will be dependent on the outcome of the licensees and the NRC reviews subsequent to TMI/2.
RESPONSE
1.
A preliminary master list of Engineered Safety Feature Systems is given in Enclosure 2.
The major components of these systems are identified in the list.
The detailed listing requested by the Bulletin will be provided in final master lists. These lists will be submitted by July 15, 1980.
ITEM 2.
For each class lE electrical equipment item identified in Item 1, provide written evidence of its environmental qualification to support the capa-bility of the item to function under postulated accident conditions.
For those class 1E electrical equipment items not having adequate qualification data available, identify your plans for determining qualifications of these items and your schedule for completing this action.
Provide this in the format of Attachment 3.
RESPONSE
2.
The information showing evidence of qualification will be submitted within 45 days of the submittal of the master list of components for each system.
(Note:
Open items indicating continuing investigation of environmental qualification may still exist on these.submittals.)
ITEM 3.
For equipment identified in Items 1 and 2 provide service condition profiles (i.e., temperature, pressure, etc., as a function of time).
These data
- Enclosure 1 TLL 111 ITEM 3 Con't.
should be provided for design basis accident conditions and qualification tests performed.
This data may be provided in profile or tabular form.
RESPONSE
3.
The service conditions for the items identified above are given on pages 19 thru 21 of Enclosure 2.
ITEM 4.
Evaluate the qualification of your Class 1E Electrical equipment against the guidelines provided in Attachment 4., " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," provides supplemental information to be used with these guidelines.
For the equipment identified as having " Outstanding Items" by Attachment 3, provide a detailed " Equipment Qualification Plan."
Include in this plan specific actions which will be taken to determine equipment qualification and the schedule for completing the actions.
RESPONSE
4.
Evaluation of qualification of equipment against the guidelines provided in Attachment 4 of IE 79-OlB will be provided with the evidence of qualification.
ITEM 5.
Identify the maximum expected flood level inside the primary containment resulting from postulated accidents.
Specify this flood by elevation such as the 620 foot elevation.
Provide this information in the format of Attachment 3.
RESPONSE
5.
The maximum flood level inside containment is given on page 22 of enclosure l
2.
ITEM 6.
Submit a " Licensee Event Report" (LER) for any Class lE electrical equipment item which has been determined as not being capable of meeting environmental qualification requirements for service intended.
Send the LER to the appropriate NRC Regional Office within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of identification.
If plant operation is to continue following identification, provide justification i
--. Enclosure 1 TLL 111 ITEM 6 Con't.
for such operation in the LER. Provide a detailed written report within 14 days of identification to the appropriate NRC Regional Office.
Those items which were previously reported to the NRC as not being qualified per IEB-79-01 do not require an LER.
RESPONSE
6.
At this point, no such items have been identified.
i U
i 4
i THREE MILE ISLAND NUCLEAR STATION UNIT 1 I&E BULLETIN 79-OlB PROGRAM PRELIMINARY REPORT i
TABLE OF CONTENTS Systems List......................................Page 1 Legend for Master List.........................Page 2
Major Components Lists by System...............Pages 3 - 18 Service Condition Profiles........................Page 19 Maximum Flood Level Inside Containment............Page 22 l
.m-__ _ _. -
Paga 1, 4
Systems List Systems required to function for and subject to resultant environments of postulated accidents (LOCA/HELB inside Containment, HELB outside Containment).
System Page Number Reactor Building Isolation System 3, 4 Reactor Building Emergency Cooling System 5
Reactor Building Emergency Cooling River Water System 6
Makeup and Purification System (HPI and Isolation) 7 Decay Heat Removal System (LPI) 8 Decay Heat Closed Cooling Water System 9
Core Flood System 10 Reactor Building Spray System 11 Nuclear Services Closed Cooling Water System 12
'Rin Steam System 13 Emelgency Feedwater System 14 Hydregen Recombiner System 15 Reactor Protection System 16, 17 Engineered Safeguards Actuation System 18 j
{.
l
Page 2.
t Legend for Master List Locations (List of Structures Subject to Accident Environments)
CODE STRUCTURE S1 Reactor Building S2 Auxiliary Building S3 Fuel Handling Building S4 Intermediate Building Accident Conditions (For Which Equipment is Required to Function)
CODE STRUCTURE Al LOCA Inside Containment A2 MSLB Inside Containment A3 FWLB Inside Containment A4 MSLB Outside Containment A5 FWLB Outside Containment
M ASTI:lt I.!ST OF CI. ASS IE I I'. M. i'
. Page 3.
Mv EI.E CI'lll CA I. EQl:
IIMQt'IltM D TO Ft'NCTION IINDER POSTUI.ATED ACCIDMNT PJNDITIONS r BuMng Isoladon System (RBS a
SYSTEM:
-,,m.m Accident ant 1. D.
umber Description Location Condition Comments AII-V1 A R. B. Furge Iso.
SI A1,2,3 to D CA-V4A, Stm. Gen. F Wtr. Iso.
S2 Al 3,3 B
CA-V159 Dcmin. Wtr. R. B. Iso.
S2 i., '4, 3 C A-V5 A.
Stm. Gen. F Iso.
S2 A1,2,3 B
CA-V1 Press.Stm. Space Sampic S2 A1,2,3 Iso.
CA-V2 RCS Sampic Cont. Iso.
S2 A1,2,3 CA-V3 Press. Wtr. Space Sampic S2 A1,2,3 Iso.
CA-V13 IlCS Letdown Sample Iso.
S2 A1,2,3 CM-V1 Cont. MonitoringIso.
S4 A1,2,3 CM-V2 Cont. Monitoring Iso.
S4 S4 A1,2,3 C M-V3 Cont. Monitoring Iso, l
A1,2,3 S4 CM-V4 Cont. Monitoring Iso.
. S1 -.
A1,2,3 CF-V2A, Core Flood Tani: Sampic Iso.1 A1,2,3 B
1 C F-VID A,
C. V. Iso. V:2 ve S2 A1,2,3 -
B CF-V20A, Core Flood Tank Sample Iso. S2 A1,2,3 B
IC-V2
- 1. C. Closed Loop Iso.
Sl A1,2,3 IC-V3
- 1. C. Retu rn Iso.
S2 A1,2,3
!C-V4 I. C. Supply S2 A1,2,3 l
I s...
s..
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.s
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Paga 4.
6 MASTER LIST OF CIASS IE (CON'T)
Reactor Building Isolation System (RBIS)
" ant I. D.
Accident
' mber
' Description Location Condition Comments
.i IC-V6 CRD Cooling Penet. Iso".
A1,2,3 MU-V2 A, Letdown Iso.
S1 A1,2,3 B
MU-V3 Letdown Iso.
S2 A1,2,3 MU-V18 Charging Iso.
S2 A1,2,3 MU-V25 RCP Seal Letdown S1 A1,2,3 MU-V2G RCP Seal Letdown S2 A1,2,3 NS-V4 RCS Pump Cooler Disch.
S2 A1,2,3 NS-V15 RCS Pump Cooler Inlet S2 A1,2,3 NS-V35 RCS Pump Cooler Disch.
S2 A1,2,3 RB-V2A RBECS Normal Cool Inlet A1,2,3 RB.V7 RBECS Normal Cool Outlet A1,2,3 WDG-V3 RB Vent Header Iso.
S1 A1,2,3 WDG-V4 RB Vent Header Iso.
S2 A1,2,3 l
WD L-V RCS DRN Tank Outlet Iso.
S1 A1,2,3 303 WDL-V RCS DRN Pump Outlet Iso.
S2 A1,2,3 304 WDL-V RB-Sump Outlet Iso.
S2 A1,2,3 534 WDIeV RB Sump Outlet Isa.
S2' A1,2,3 535 l
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&\\STEli I.IST OF Cf. ASS IE Paga 5.
E I.E UI'!!! CA 1. HQl'1l'.'.! ENT IWOI!!!iE D TO FIINCTinN IJNDER I)CSTUI.ATED ACCIDI NT ('ONDITIONS 4
l SYSTE31:
Reactor Building Emergency Cooline System (RBECS)
PlantI.D.
Accident Number Description 1,ocation Co'ndition Comments AII-EIA R.B. Air Recirc & Cooling S1 A 1, 2, 3 i
IJnit AH-ElB R.B. Air Recire. & Cooling S1 A 1, 2, 3 Unit Ali-E1C R.B. Air Recirc. & Cooling S1 A1,2,3 Unit e
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MASTE!! I,IST OF CI. ASS TE Page 6.
E I.E CTiti CA I, EQl!Il'M Eh"I' REQUlltCD TO FUNCilON UNDER POSTULATP.D ACCIDENT CONDITIONS SYSTEM:
Reactor Building Emergency Cooling River Water System (RR)
Plant I. D.
Accident Number Description Location Condition Comments RR-V3A,B,C R.B. Emerg. Cool Coil S2 All Inlet RR-V4A,B,C R.B. Emerg. Cool Coil S2 All D
Outlet RR-VG RBECC Press Control S4 All RR-V5 RR-VG Reg. Bypass S4 All e
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M ASTEft LIST Ol' CLASS lE Page 7.
F.I.1:CI'I:t CA 1 MO!'I PM ENT IIEQt'!!:E D TO l't:NCTION l'NDEll PCSTUI.ATED ACCIDENT CONDITIONS SYSTEM: Makeun and Purification Svstem (MU PS) HPI & ! solation Plant I. D.
Accident Number _
Description Location Condition Comments CU-P1A,B,C 31UPS Pumps A,B, & C' S2 A1,2,3,4 IU-P2A,B, C MU-P1A,B, C Aux. Oil S2 A1,2,3,4 Pumps i
IU-P3 A, B,C MU-P1A,B, C Main Oil S2 A1,2,3,4 Pumps I U-P4 A,B,C MU-P1A, B, C Gear Oil S2 A1, 2, 3, 4 Pumps
.fU-V2A,B Letdown Cooler Outlet S1 A1,2,3,4 Contained in 11 BIS 1U-V3 Letdown Isolation S2 A1,2,3,4 Contained in IlBIS
.1 U-V14 A & B BWST to MUPS S2 A1,2,3,4
~,1 U-V 1GA, B,
liPI Iso at Containment S2 A1,2,3,4 C, D
.IU-VIS Char;;ing Iso at Containment S2 A1,2,3,4 Contained in 11 BIS
.lU-V20 RCP Seal Wtr. Iso S2 A1,2,3,4 Isolated by Operator c.1 U -V25 RCP Seal Letdown Iso.
S1 A1,2,3,4 Contained in 11 BIS
.1U-V2G HCP Seal Letdown iso.
S3' A1,2,3,4 Contained in It BIS
- ,1 U-V3 G MUPS Pump Recirc. Iso.
S2 A1,2,3,4
" 1 L'- V3 7 31 CPS Ihtmp Recire. Iso.
S2 A1, 2, 3,4 A1,2,3,4
.',1 C-V 12 MU-T1 Disch. Iso.
~
S2 A1,2,3,4 This valve should have thc i ability to close on !!PI If tank inventory is depleted Power supply to MU-V12 is Non-ES
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MASTElt I.IST OF CI. ASS IE_
Page 8.
I'I.E CTI(I CA 1. EQt*Il' Al ENT REOllittED TO FUNCTION ITNDER POSTULATED ACCIDENT CO.NDITIONS SYSTEM:
Decay Heat Removal System (DIIRS) LPI
~~
-.~
i Plant I. D.
Accident.
~
Number Description Location Condition Comments DH-PIA &B DIIRS Pump A&B S2 A1,2,3,4 DII-V4A &B DH-P1A&lB disch. Iso. V1vs.
S2 A1,2,3,4 Dil-VGA &B RB Sump to DIIRS IIdr.
S1 A1,2,3,4 Dil-VSA&B Borated Wtr. to DIIRS S2 A1,2,3,4 DII-V1 DII Suction from Loop B S1 A1,2,3,4 DII-V2 DII Suction from Loop B S1 A1,2,3,4 DII-V3 DH Suction Cont. Iso.
S2 A1,2,3,4 DH-V7A &B Dll Exch. to MUPS S2 A1,2,3,4
~
DII-VG1A &B Caustic Pump Disch. to DlI S2 A1,2,3,4 0
e 8
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31ASTI:11 T.!ST OF Cf. ASS IE E LE C1'I:I CA L EQUI P.it ENT Page 9.
IIEQt'!!ti: D TO FL'NCTION UNDER POSTULATED ACCIDENT CONDITIONS Decay Heat Closed Cycle Cooling Water System (DC)
SYSTESI:
'~
'~
lant I. D.
Accident s' umber Description Location Condition Comments DC-PIA, DHCCW Pumps A&B S2 ALL B
D C-V10 A, DC Surge Tank Slakeup S3 ALL Air Operated B
D C-V2 A,
D!!RS Heat Exch. Inlet
' S2 ALL Air Operated B
D C-VG5 A,
DHRS Cooler Bypass S2' ALL Air Operated B
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MASTEIt I.!ST OF CIJSS IE Page 10.
E I.C Cfl:! CA l. EQI'l l'.TI E NT llEOt:II ED TO Ft'NCTION l'NDER POSTULATED ACCIDENT CONDITIONS I
SYSTE31:
Core Flood System (CF)
Plant I. D.
Accident Number Descriotion Location Condition Cominents F-V20A CF Tank Sampic & Iso Valve '
S2 A1,A2,A3 Contained in RBIS, Close l
on ESAS OF-V20B CF Tank Sampic & Iso Valve S2 A1,A2,A3 Contained in RBIS, Close on ESAS CF-VIDA CV Isolation for Makeup to S2 A1,A2,A3 Contained in RBIS, Close CF Tank on ESAS CF-V19 B CV Isolation for Makeup to S2 A1,A2,A3 Contained in RBIS, Close CF Tank on ESAS CF-V2A CF Tank Sample Isolation Sl' A1,A2,A3 Contained in RBIS, Cic30 on ESAS CF-V2n CF Tank Sampic Icolation S1 A1,A2,A3 Contained in RDIS, Close on ESAS CF -V1A CF Tank Iso. Valve S1 A1,A2,A3 Valve Opened and Breake Tagged out.
'1 A1, A2, A3
Valve Opened and Breake CF-V1B CF Tank Iso Valve S
Tagged out.
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M ASTEf t i.!ST OF CI. ASS IE, Page 11.
E1.E CI'I:I CA L EQl'IE.TIENT
_REQt:liti:D TO Ft'NCflON (!NDER l'OSTUI. ATE D ACCIDENT CONDIT!ONS 4
SYSTEM:
Reactor Building Spray System (RBSS) i
'lant I. D.
Accident l
Number Description Location '
Condition Comments BS-P1A&B RBSS Pumps A & B S2 A 1, 2, 3 BS-VIA&B BS-PIA & B Disch. Valves S2 A1,2,3 BS-V2A&B NAOII to BS-P1A &B S2 A1,2,3 BS-V3A&B RBS Pump Suction Iso.
S2 A1,2,3 e
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A1 AST!:lt LIST O F CL.SSS IE E LE CT!!I CA I. I:QtlI!).\\1ENT Page 12.
REQl'!!!E D TO Fl'NCTION !!NDER POSTULATED ACCIDENT COhDITIONS-'
S E T E SI:
Nuclear Services Closed Cycle Cooling. Water System (NS)
'lant I. D.
Accident Number Description Location Condition Comments NS-P1 A, NS Pump A, B & C S2 ALL B,C NS-V4 RCP Coolers Disch.
S2 ALL Contained in RBIS
~
NS-VIS RCP Coolers Inlet S2 ALL Contained in. RBIS l
NS-V32 Inlet Hdr. to Non-Nuc.
S2 ALL l
Equipt.
NS-V35.
RCP Cooler Disch.
S1 ALL Contained in RBIS NS-VS2A, RBECS Units Fan Cooler S4 ALL B,C Inlet NS-V53 A, RBECS Units Fan Cooler S4 ALL B,C Outlet I
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- MASTElt I.lST OF Cl. ASS IE Page 13.
E l.1-: UI' ! :1 CA _1. 1
- 01 : 1 l'.'.l l: N~l' ggjlil:1:D TO ITNC*llON l'N!W.Il POSTUI.ATI:D ACCrol.:NT CONDITIONS SYSTEM:
Main Steam System (MS)
Plant I. D.
Accident Number Description Location Condition Comments MS-V1A Main Steam Isol. Valve S4 A2,A4 MS-V1B Main Steam Isol. Valve S4 A2,A4 MS-VIC Main Steam Isol. Valve S4 A2,A4 MS-V1D Main Steam Isol. Valve
$4 A2,A4 MS-V2A, B Stm. Supply to EFW S4 All Pump Turbine MS-V10A, B Stm. Supply to EFW S4 All Pump Turbine MS-V13 A, B Sim. Supply to EFW Pump S4 All Tu rbinc NS-VG Sim. Supply to EFW Pump S4 All Turbine l
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MASTEli I.lST O F Cf. ASS IF_
Page,14.
E l.1:(.T!!!('A I. EQt'II'.il E NT 1 EQITIRE D TO FI'N(r!*!O.': !!NDER POSTUI.ATI'D ACCIDENT CONDITIONS SYSTEM:
Emergency Feedwater System (EFWS) t Plant I.D.
Accident Number Description Location Condition Comments 3 F-Pl.
EFWS Pump (Turb)
S4 All 3F-P2A EFWS Pump (Motor)
S4 All SF-P2B EFWS Pump (Motor)
S4 All E F-VlA &B EFWS Pump Suction IIdr.
S4 All E F-V2A&B EFWS Pump Disch. IIdr.
S4 All CF-V4 Emer. River Wtr. to EFWS S4 All E F-V5 Emer. River Wtr. to EFWS S4 All E F-VS A, B. Min. Flow Valves S4 All
&C E F-V00 AT:b E FWS Control Valves S4 All Air Operated Control Valves e
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M ASTI:It 1.1ST OF CI. ASS IE Page 15.
EI.ECrit!CAEL: Qt'll'. lIG, r i
REQI.'lliM D TO Ft.'NCTION liNDER POSTUI.ATED ACCIDMST CONDITIONS SYSTEM:
. Ilydrogen Recombiner System ( HR)
'. ant I. D.
Accident
. umber Descrintion Location Condition Comments HR-R-1 Hydrogen S4 Al 303'EL in Leak Rate Test Recombiner Equipment Area Recombiner Control S4 Al 295'EL There are 2 Console Redundant Control Consoles HR-VOL Sol. Op. Isol. Va S1 Al l'E 125V DC Power Supply HR-V22 Sol. Op. Isol. Va S1 Al 1E 125V DC Power Supply.
HR-V03 Sol. Op. Isol. Va S1 Al 1E 125V DC Power Supply llR-V23 Sol. Op. Isol. Va S1 Al 1E 125V DC Pow'er Supply Note: One hydrogen recombiner will be installet ':
prior to restart. The sccor (redundant) recombiner ucci not be installed, however, t piping system, electrical power supplies and structur,
provisions shall be installec and available. The second hydrogen recombiner shall be installed after an accider within the time period l
available before they need
-o be operational.
I l
Page 16.
ATAST!'R I.IST OF CI. ASS IE t
E I.F CTIM CA I. EOl'I PM EN"O RI'OI'!1:C D TO I'l'NC~l' ION IrNI)E R POST (?LATCD ACCiDCNT CONDITIONS SETEM:
Reactor Protection System (RPS)
Plant 1. D.
Accident Number Description Locatio'n_,, Condition Comments R C 14 A-RC Flow Measure S1 All Input to Power - Flow dFr4 Monitor Logie i
RC14B-RC Flow Measure S1 All Input to Power - Flow d Fr 4 Monitor Logic R C 14 A-RC Flow Measure S1 All Input to Power - Flow d Fr 3 Monitor Logic R C 14B-RC Flow Measure S1 All Input to Power - Flow d Fr 3 Monitor Logic RC14A-RC Flow Measure S1 All Input to Power - Flow d FF 2 Monitor Logic R C 14 B-IIC Flow Measure S1 All Input to Power - Flow Monitor Logic d PT 2 R C 14 A-RC Fluw Measure S1 All Input to Power - Flow Monitor Logic d Ier 1 R C 14 Il-IIC Flow Measure S1 All Input to Power - Flav' Monitor Logic d Fr 1 ItC 311-Irr2 ItC Pressure Measure S1
.All input to RC Tentp-Pressure Logic NI-5
' Flux Scusers S1 All Over Power Trip Protec:
NI-G Flux Sensors S1 All Over Power Trip Protec-NI-7 Flux Sensors S1 All Over Power Trip Protet NI-S Flux Sensors S1 All Over Power Trip Protec-
-_.2'
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Page 17.
i MASTER LIST OF CLASS lE (CONT'D) l Reactor Protection System (RPS)
Plant I.D.
Accident Number Description Location Condition Comments RC3A-PT 2 RC Pressure Measure S1 All Input to RC Temp-Pressure Logic RC3B-PT 1 RC Pressure Measure S1 All Input to RC Temp-Pressure Logic RC3A-PT 1 RC Pressure Measure S1 All Input to RC Temp-Pressure Logic RC4B-TE 3 RC Temperature Measure S1 All Input to RC Temp-Pressure Logic RC4A-TE 3 RC Temperature Measure S1 All Input to RC Temp-Pressure Logic RC4B-TE 2 RC Temperature Measure S1 All Input to RC Temp-Pressure Logic RC4A-TE 2 RC Temperature Measure S1 All Input to RC Temp-Pressure Logic RC-MIS RC Pump Monitor S1 All Input to Power-Flow P 1A1 Monitor Logic RC-MIS RC Pump Monitor S1 All Input to Power-Flow P 1A2 Monitor Logic RC-MIS RC Pump Monitor S1 All Input to Power-Flow P 1B1 Monitor Logic RC-MIS RC Pump Monitor S1 All Input to Power-Flow P 1B2 Monitor Logic PS-672,673 674,6675 RB Pressure Switches S1 All High.RB Pressure Trip 4 psig i
l
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Page 18.
MASTE11 LIST OF CLASS IE ELC CTI:t CA1. eof'IIGIPNT-REQUII:EI) TO Fl'NCTION NDEll POSTULATED ACCIDENT CONDITIONS SYSTE31:
Encincered Safecuards Actuation System (ESAS) i Plant I.D.
Accident Number Descrintion I.ocation Condition Comments RCSA-PT3 Pressure Trausmitter S1 ALL T. ansmits RC Pressuri i
ESAS ALL R C3A-PT4 Pressure Transmitter S1 Transmits RC Pressure ESAS RC3B-Fr3 Pressure Trans:nitter S1 ALL Transmits RC Pressure ESAS Fr 2S2 a P. Transmitter S2 A1, 2,3 Transmits RB pressure Info. to ESAS PT 285 Zh P. Transmitter S2 A1, 2, 3 Transmits RB Pressu r.
info. to ESAS Fr 288 o P. Transmitter S2 A1,2,3 Transmits RB Pressure Info. to ESAS PS 283 Pressure Switches S2 A1,2,3 Actuate at 30 psig IIB Pressure PS 284 Pressure Switch'es
.S2 A 1, 2, 3' Actuate at 30 psig RB Pressurc PS 2SG Pressure Switches S2 A1, 2,3 Actuate at 30 psig HD i
i Pressure PS 287 Pressure Switches S'2 A1,2,3 -
Actuate at 30 psig R11 Pressure PS 260 Pressure Switches S2 A 1, 2, 3 Actuate at 30 psig HH Pressure PS 290 Pressure Switches S2 A 1, 2, 3 Actuate at 30 psig IIH Pressurc e
u 4
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Page 19.
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Service (bndition Profiles 1.0 Service Conditions Inside Containment for a Loss of Coolant Accident 1.1 Temperature and Pressure Beactor building temperature and pressure profiles subsequent to a IOCA are shown on FSAR Figures ' 4-55, 14-56, 14-59 through 14-63, and 14-66. %e maximum reactor cilding pressure is defined on Figure 14-66 as 50.6 psig, with maximum temperature indicated as 2750F on Figure 14-63A. Wese Figures are included in this Enclosure on Attachments A and B.
1.2 Padiation I&E Bulletin 79-OlB suggests a guideline for gartma radiation as 2 x 107 Pads.
1.3 Sutmergence A maximum flood level, equivalent to the 286.94 foot building elevation, has been determined for the worst case line break within containment.
1.4 Chemical Spray spiprent within containment may be expsed to a chenical spray environment subsequent to a IdCA.
%e composition of the spray will be made up of borated water from the Borated Water Storage Tank (EWST), sodium hydroxide from the reactor building spray system and the reactor coolant which exits the break.
% e borated water is maintained at 2270 pga boron. We sodium hydroxide raises the pH of the boratcxi water into the alkaline range to approximately 9.5.
Pago 20.
2.0 Service Cbnditions for PWR Main Steam Line Break Inside Containment 2.1 Tanperature and Pressure Section 4.2.1 of I&E Bulletin 79-OlB allows that equigent qualified for the IOCA envimmi:nt can be considered qualified for a main steam line break inside containment provided the plant design inw1.porates an automatic spray system not subject to disabling single conponent failures.
The UMI-l design incorporates the reactor building emergency cooling systen and reactor building spray systen which, in conjuction, provide a single failure proof mechanism to limit peak reactor building pressure and tauperature.
It is considered that the TMI-1 design meets the intent of Section 4.2.1 of I&E Bulletin 79-OlB, and that equignent qualified for the IOCA environment are also considered qualified for a main steam line break environment.
2.2 Padiation 6 Rads.
I&E Bullecin 79-OlB suggests a conservative gartna dose of 2 x 10 2.3 Submergence The same as Section 1.3 of this Enclosure.
2.4 Chenical Sprays The same as Section 1.4 of this Enclosure.
3.0 Service Cbnditions for PWR Feedwater Line Break Inside Cbntainment The environmental conditions that result from a MSIB inside containment are more severe and envelope those that result from a EWIB inside containment. As stated in Section 2.O, equipent qualified for IDCA environr:ent will be considered qualified for a MSIB inside containment. Therefore, equipent qualified for IOCA environment will also be considered qualified for a Eh1B inside containment.
Pags 21.
4.0 Service Cbnditions Outside (bntainnent 4.1 Areas Subject to a Severe Environment as a Result of a HEW For the purpose of response to I&E Bulletin 79-OlB at this time, the credible HEW that subjects areas to severe environments are a MSLB and a EWG in the Intenvvlinte Building. 'Ihe environmental conditions that result frm a MSLB in the InternM1 ate Building are nore severe and envelope those that result frm a EWIB in the InternMinte Building.
'Iherefore, equiptent qualified for a MSG environment in the Intermediate Building will also be considered qualified for a EWG in the Internediate Btilaing.
4.1.1 Temperature and Pressure Intermediate Biilding tcrnperature and humidity profiles following a steam line break are indicated on Attachments C, D and E to this Enclosure. 'Ihe pressure is considered as per Supplement 2,Part IX 'of the FSAR. 'Ihese profiles show a maximum temperature of 3230F and a maxrnum relative humidity of 100 percent.
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t Paga 22.
P Maxinum Flood Ievel Inside Containment A maxinum flood level of 5.94 feet (286.94 foot building elevation) has been determined for the worst case line break within containment.
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