ML20090J946

Applicant Exhibit A-30,consisting of Forwarding Response to 831219 Request for Addl Info Re Environ Qualification Program

ML20090J946
Person / Time
Site:	Limerick
Issue date:	04/09/1984
From:	Kemper J PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	Schwencer A Office of Nuclear Reactor Regulation
References
OL-A-030, OL-A-30, NUDOCS 8405230397
Download: ML20090J946 (29)

Text

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r O L-PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET 00CMETfD P.O. BOX 8699 PHILADELPHI A. PA.19101 24 APR 24 p1;ja

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Mr. A. Schwencer, Chief Docket Nos. 50-352 [RE

Licensing Branch No. 2 50-353 Division of Licensing U. S. Nuclear Regulatory Constission Washington, D.C. 20555

Subject:

Limerick Generating Station, Units 1 and 2 Environmental Qualification Report Reference Letter from A. Schwencer to E. G. Bauer, Jr.

dated December 19, 1983 File GOVT 1-1 (NRC)

Dear Mr. Schwencer:

The reference letter transmitted requests for additional information (RAI) from the Environmental Qualification Section. Enclosed are our responses to these RAI. As an aid to your review of the response to RAI 270.7, we have enclosed the report referenced in the response, Drywell Temperature Response g a Small Steam Break.

'Ittese responses will be included in Revision 1 of tha Environmental Qualification Report scheduled to be submitted in April, 1984.

Sincerely,

$l$f JLP/gra/010984435 Enclosures ggy gMfR RE A10RY COMMissl0N Copy to: See Attached Service List '-

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Judg3 Lawren a Br:nnOr (w/On:l cur 0) l','

' , '3f Judge Peter A. Morris (w/ enclosure)

Judge Richard F. Cole (w/ enclosure)

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Troy B. Conner, Jr., Esq. (w/ enclosure)

Ann P. Hodgdon, Esq. (w/ enclosure)

Mr. Frank R. Romano (w/ enclosure)

Mr. Robert L. Anthony (w/ enclosure Mr. Marvin I. Lewis (w/ enclosure) ,

Charles W. Elliot, Esq. (w/ enclosure)

Zori G. Ferkin, Esq'. (w/ enclosure)

] t'e. 3 Thomas Carusky' (w/ enclosure)

. Director, Penna. Emergency (w/ enclosure) f Manag'ement Agency.

Mr. Steven P.' Hershey (w/ enclosure)

Angus Inve, Esq. '

(w/ enclosure)

I ,/, , - Mr. Joseph H. White,'III (w/ enclosure)

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David Wersen, Esq. (w/ enclosure)

Robert J. Sugarman, Esq. , (w/ enclosure)

'7 Sporte W. Perry, Esq. (w/ enclosure)

,y Jay M. Gutierrez, Esq. (w/dnclosure)

Atomic Safety & Licensing (w/ enclosure) i Appeal Board Atomic Safety & Licensing (w/ enclosure)

Board Panel Docket & Service Section (w/ enclosure)

Marth'a W. Bush, Esq. (w/ enclosure) g James'Wiggins 1 (w/ enclosure)

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e Additional Information Requiro' *. ,

Limerick Environmental Qualification l'ro;3am Q 270.1 Correlate the systems listed in Table 3.2-1 of the FSAR (SRP 3.11) with the systems listed in Appendix A, " List of Systems

.(c Important to Safety," of the environmental qualification (EQ) program submittal of October 1983.

Provide justification for any system listed in Table

} 3.2-1 which is, excluded from Appendix A (e.g., all components of the system are located in a mild

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environment, etc.). Identify the Class lE function for any systems which are added to Appendix A.

Response

k The requested systems correlation is shown on the following

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Que: tion 270.1 ,

SYSTEM CORRELATION * "

TABLE 3.2-1 SYSTEMS E!!VIRONMENTAL QUALIFICATION REPORT JUSTIFICATION FOR EXCLUSION I NSSS A. Reactor System Not included Does not contain electrical equipment.

B. Nuclear Boiler System Nuclear Boiler System Not applicable.

Nuclear Boiler Instrumentation Nuclear Steam Supply Shutoff System C. CRD Hydraulic System Same Not applicable.

D. Recirculation System Reactor Recirculation System Not applicable.

E. Reactor Water Cleanup System Same Not applicable.

F. Traversing Incore Probe Not included except TIP drive isolation Not required to mitigate the (TIP) System valves which are included in Primary effects of a Design Basis Event.

Containment Instrument Gas System.

II Engineered Safety Features A. Reactor Core Isolation Same Not applicable.

Cooling (RCIC) System D. Residual Heat Removal Same Not applicable.

System .

C. Core Spray System Same Not applicable.

D. High-Pressure Coolant Same Not applicable.

i Injection (HPCI) System E. Standby Liquid Control System Same Not applicable.

III Fuel Storage and Handling: Not included except RIIR intertic Not required to mitigate the Reactor Vessel Servicing valves which are included in Residual Effects of a Design Basis Event.

IIcat Removal (RIIR) .

4 IV Radioactive Waste Not included. Not required to mitigate the Management Effects of a Design Basis Event.

Que: tion 270.1 ,

TABLE 3.?-l SYSTEMS ENVIRONMENTAL QUALIFICATION REPORT JUSTIFICATION FOR EXCLUSION -

V Water Systems A. Service Water System Not included All equipment located in a mild environment area.

B. Emergency Service Water Same Not applicable.

System C. RHR Service Water System Residual Heat Removal (RHR) Not applicable.

D. Reactor Enclosure Cooling Reactor Enclosure Cooling Not applicable.

Water System Water - Iso Valves E. Turbine Enclosure Cooling Not included. All equipment located in a Water System mild environment area.

F. Circulating Water System Not included. All equipment located in a mild environment area.

VI Diesel Generator System Not included. All equipment located in a mild environment area.

VII Heating, Ventilating, and Air Conditioning Systems A. Control Structure Not included. All equipment located in a mild environment area.

B. Reactor Enclosure and Refueling Arca

1. Reactor Enclosure and Reactor Enclosure HVAC Not applicable.

refueling Area Recirc. Mode

2. Refueling Floor HVAC Not included. Not required to mitigate the System (Normal Operation) Effects of a Design Basis Event.

3. Reactor Enclosure Air Reactor Enclosure HVAC Not applicable.

Recirculation System Recire. Mode.

4. Standby Gas Treatment Same Not applicable.

System

5. PHR, HPCI, RCIC and ECCS Pump Room HVAC Not applicable.

CS Rooms HVAC

Ou22 tion 270.1 _4_

TABLE 3.2-1 SYSTEMS ENVIRONMENTAL QUALIFICATION REPORT JUSTIFICATION FOR EXCLUSION C. Primary Containment

1. Drywell Cooling System Drywell llVAC Not applfcable.

2. Purge System Containment Atmospheric Control Not applicable.

System (CAC)

3. Ilydrogen Recombiner Added Note 1

4. Vacuum Relief System Not included. Does not contain safety-related electrical equipment.

D. Radwaste and Offgas Not included. All equipment located in a Enclosure mild environment area.

E. Turbine Enclosure Not included. All equipment located in a mild environment area.

F. Diesel-Generator Enclosure Not included. All equipment located in a mild environment area.

G. Spray Pond Pump Structure Not included. All equipment located in a mild environment area.

H. Miscellaneous Pump Not included. Does not contain safety-Structures Schuylkill, related electrical equipment.

rerkiomen, Circulating Water I. Miscellaneous Structures Not included. Does not contain safety-(Auxiliary Boiler, Fuel related electrical equipment.

Oil Transfer, Water Treatment, Sewage Treat-ment) l J. Administration Building Not included. Does not contain safety-related electrical equipment.

K. Ilot Maintenance Shop Not included. Does not contain safety-related electrical equipment.

i Note 1 - Although the manufacturer has conducted an environmental qualification test for this system, the BWR Owners Group has demonstrated that the Hydrogen Recombiners are not required to serve any safety function in BWR's with inerted containments. This position is expressed in BWR Owners Group letter to D. G. Eisenhut (NRC) dated 8/13/82 and NtJSCO letter to W. J. Dircks (NRC) dated 8/6/82. The NRC Staf f is continuing its consideration of this issue (ref. SECY-83-292). The Ilydrogen Recombiners may be deleted from the Environmental Qualification Report Appendix A & B lists subsequent to an NRC staif decision.

Questien 270.1 .

TABLE 3.2-1 SYSTEMS ENVIRONMENTAL QUALIFICATION REPORT JUSTIFICATION FOR EXCLUSION .

VIII Main Steam and Power Not included All equipment located in a Conversion System mild environment area.

IX ' Instrumentation and Control Systems A. Reactor Protection (Trip) Nuclear Boiler System Not applicable.

i System B. Engineered Safety Features System

1. Emergency core cooling Systems:

Ifigh pressure coolant Same Not applicable.

injection.

Automatic depressuriza- Nuclear Boiler System Not applicable.

tion system.

Core spray Same Not applicable.

Low pressure coolant Residual lleat Removal (RilR) Not applicable.

injection (RIIR)

2. Primary containment and Not included All equipment location in a reactor vessel isola- mild environment area, tion control system.

3. Class lE power system 4 kV Power Not applicable.

440 V Load Centers and MCC's

4. RilR containment spray Residual Heat Removal (RHR) Not applicable.

mode.

5. Service water systems:

RHR service water Residual lleat Removal (RilR) Not applicable.

Emergency service water Same Not applicable.

6. Containment atmospheric control systems:

Combustible gas control Containment Atnospheric Control System Not applicable.

system Combustible gas monitor- Containment Atmospheric Control System Not applicable.

ing system Primary containment Containinent Atmosi.heric Control System Not applicable.

vacrum relief r;yste:m ,

Question 270.1 .

TABLE 3.2-1 SYSTEMS ENVIRONMENTAI, QUALIFICATION REPORT JUSTIFICATION FOR EXCLUSION

7. Main steam line isolation Main Steam Isolation Valve-Leakage Not applicable.

valve leakage control system.

8. RHR suppression pool Residual Heat Removal (RHR) Not applicable.

cooling system.

9. Reactor enclosure re- Reactor Enclosure IIVAC-Recire. Mode Not applicable.

circulation system.

10. Reactor enclosure Reactor Enclosure ifVAC-Recirc. Mode Not applicable.

isolation system.

11. IIabitability and control Not included. All equipment located in a room isolation mild environment area.

12. Standby gas treatment Not included. All equipment located in a filter room and access mild environment area, area unit coolers.

13. Diesel-generator enclo- Not included. All equipment located in a sure ventilation system. mild environment area.

14. Spray pond pump structure Not included All equipment located in a ventilation system. mild environment area.

15. ESF switchgear and Not included All equipment located in a battery rooms cooling mild environment area, system.

16. Emergency core cooling ECCS Pump Room ifVAC Not applicable.

system pump compartment

unit coolers.

17. Drywell unit coolers Drywell HVAC Not applicable.

18. Control enclosure Not included. All equipment located in a chilled water system. mild environment area.

19. Auxiliary equipment room Not included. All equipment located in a ventilation system. mild environment area.

20. Standby gas treatment Same Not applicable, system. .

Que2 tion 270.1 ,

TABLE 3.2-1 SYSTEMS ENVIRONMENTAL Qt!ALIFICATION REPORT JUSTIFICATION FOR EXCLUSICC7 . .

C. Safety-Related Display Various Systems - Not applicable.

Instrumentation. Identified by Note O in EQR.

D. Systems Required for Safe Shutdown.

1. reactor core isolation Same Not applicable.

cooling system.

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2. Standby liquid control Same Not applicable.

system.

3. Reactor shutdown cooling Residual lieat Removal (RilR) Not appliable, mode of the RilR system.

4. Remote shutdown system Not included. All equipment located in a mild environment area.

E. . Control Systems Not Not included. Not required to mitigate the Required for Safety. Effects of a Design Basis Event.

1. Reactor pressure vessel instrumentation.

2. Reactor manual control system.

3. Recirculation control system.

4. Feedwater control system.

5. Pressure regulator and turbine generator system.

6. Neutron monitoring system ,

Traversing incore probe Rod block monitor Source range monitor s

7. Reactor water . cleanup system, i

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Que: tion 270.1 ..

TABLE 3.2-1 SYSTEMS FMVIRONMENTAL QUALIFICATION REPORT JUSTIFICATION FOR EXCLUSION - -

8. Spent fuel pool cooling and cleanup system.

9. Radwaste system Gaseous radwaste system Liquid radwaste system Solid radwaste system

10. Area radiaion monitoring system.

11. Process computer.

12. Containment instrument gas system.

13. Refueling interlocks.

14. Leak detection system.

15. Fire protection and suppression system.

16. Non-Safety-Related equipment area cooling ventilation systems.

17. Process radiation monitoring system.

18. Rod sequence control system.

X Electric Systems A. Engineered Safety Features AC Equipment 1.;4 kV switchgear, in- 4 kV Power Not applicable.

cluding safeguard bus feeder breakers, pro-tective relays, control panels.

Que tion 270.1 ,

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TABLE 3.2-1 SYSTEMS ENVIRONMENTAL QUALIFICATICN REPORT JUSTIFICATION FOR EXCLUSION

2. 440 V load centers, 440 V Ioad Centers and MCC's tat applicable.

including 4160/440 V transformers, pro-tective relays, control panels. .

3. 440 V motor control 440 V Ioad Centers and MCC's Not applicable.

centers.

D. Engineered Safety Features DC Equipment

1. 125 V and 125 V/250 V Not included All equipment located in a station batteries and mild environment area.

racks, battery chargers.

2. Motor control center Safeguard DC Power Not applicable and distribution panels, including protective relays.

C. 120 V Vital AC System Equipment

1. 120 V distribution panels Not included. All equipment located in a mild environment area.

D. Electric Cables for Safety-Related Equipment

1. 5 kV power cables Class IE Power Interfaces Not applicable.

2. 600 V power cables, Class IE Power Interfaces Not applicable, including all de power cables.

3. Control and instrumen- Class IE Power Interfaces Not applicable.

tation cables.

E. All Other Instrumentation Systems Required for Safety

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1. Process radiation Same Not applicable, monitoring system. '

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Que2 tion 270.1 .

TABLE 3.2-1 SYSTEMS ENVIh0NMENTAL QUA!.IFICATTO*l M PORT JUSTIFICATION FOR EXCLUSION -

2. Neutron monitoring system Not included Required for RG 1.97, however, it performs its safety function prior to exposure to harsh environment. System exempted in its entirety. Not required to mitigate the Effects of a Design Basis Event.

3. Safety relief valve Same Not applicable, position indication.

4. Leak detection systems Plant Leak Detection Not applicable.

f Nain steam line leak detection RCIC system leak detection RNCU system leak datection HPCI system leak detection

5. Containment instrument Primary Containment Instrument Gas Not applicable.

. gas system-ADS control

6. Deleted

7. Nigh-pressure / low- Residual lleat Removal (RHR) Core Spray Not applicable.

pressure systems interlock.

8. Safeguard piping full Same Not applicable.

system.

F. Niscellaneous Electrical

1. Primary containment Class IE Power Interfaces Not applicable. ,

enclosure electrical penetration assemblies.

2. Raceway systems, safety- Not included Environmental qualification related. not required for metallics.

3. Emergency lighting Not included. Not required to mitigate the related Effects of a Design Basis Event.

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L Que2 tion 270.1 . .

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E BLE 3.2-1 SYSTEMS FNVIRONMENTAL QUALIFICATION REPORT JUSTIFICATION FOR EXCLUSION

4. Emergency lighting Not locluded. Not required to mitigate the systems Effects of a Design Basis Event.

5. Emergency comemnications

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Not included. Not required to mitigate the systems Effects of a Design Basis Event. -

6. Motors, non safety- Not included. Not required to mitigate the related Effects of a Design Basis Event.

7 Inverters

8. Valve operators Miscellaneous Systems Not applicable.

G. Offsite Power Systems Not included. Not required to mitigate the Effects of a Design Basis Event.  !

XI Auxiliary Systems i

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A. Safeguard Piping Fill Safeguard Piping Fill System Not applicable.

System Including Feed-

water Fill System.

B. Suppression Pool Cleanup Same Not applicable.

4 System.

i C. Domineralized Water Makeup Not included. Not required to mitigate the

. System. Effects of a Design Basis Event.

i D. Drywell Chilled Water Drywell Chilled Water - Isolation Valves Not applicable.

System.

E.- Control Structure Chilled Not included. All Equipment located in a Water System. mild environment area.

4 F. Compressed Air and Not included. Not required to mitigate the Instrument Gas System. Effects of a Design Basis Event.

G. Sampling System. Containment Atmospheric Control Not applicable.

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H.. Equipment and Floor Drains Same Not aplicable.

I. Fire Protection System Not included. Nokrequiredtomitigatethe Effects of a Design Basis Event.

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Question 273.1 .

TABLE 3.2-1 SYSTENS ENVIRONNENTAL QUALIFICATION REPORT JUSTIFICATION FOR EXCLUSION E

J1. Nitrogen System Containment Atmospheric Control Not applicable.

System (CAC)

J2. Cenerator External Hydrogen Not included. Not required to mitigate the System Effects of a Design Basis Event.

K. Post-Accident Sampling Containment Atmospheric Control Not appl!. cable.

System System (CAC)

XII Enclosures Not included. Does not ck cain electrical equipment.

XIII- Spray Pond Not included. All equipment located in a mild environment area.

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Q 270.2 Identify, by categories listed in NUREG-0737, the (SRP 3.11) components included in the qualification program in response to TMI Action Plan Requirements.

Response

The NUREG-0737 components included in the qualification program are sunearized on the following pages.

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Q 270.2 (SRP 3.11)

NURZG 0737 Component-Plant Clarification Iter I.D. Number II.B.1 PSV-41-F013A PSV-41-F013H PSV-41-F013J PSV-41-F013N PSV-41-F013S HV-41-1F001*

HV-41-1F002*

HV-49-1F007 HV-49-1F008 HV-55-1F002 HV-55-1F003 II.D.3 ZE-41-115A-1 o 3-1 C-1 D-1 E-1 F-1 G-1 H-1 J-1 K-1

- L-1 M-1 N-1 S-1 ZT-41-115A-1 B-1 C-1 D-1 E-1 F-1 G-1 H-1 J-1 K-1 L-1 1

II.F.1 LT-42-115A 115B IN085A*

PT-42-101 103A*

1038*

PT-57-101 121

l Q270.2 (SRP 3.11)

> NLAEG 0737 Component-Plant Clarification Item I.D. Number s II.F.1 (cont'd.) AE-57-151 AIT-57-151 AE-57-188 AIT-57-188 II.F.2 TE-42-104A*

104B* '

• 104C* -

104D*

105 *

• PT-49-1NO35A*

II.I.e 1NO35E*

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'These items were not identified in the 10/83 revision of the EQR, however, all are qualified and will be incorporated within the next revision of the EQR.

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4 Q 270.3 Provide a statement that IE equipment located in areas (SRP 3.11) which experience a significant increase in radiation during a LOCA has been reviewed for possible damage to solid state devices.

Response l

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All areas which experience a significant increase in radiation j during a LOCA are considered a harsh environment by the Equipment i Qualification Progres. Accordingly, all Class 1E equipment located in l these areas including solid state devices have been reviewed for

. possible damage due to radiation.

Q 270.4 Indicate your compliance with a one hour time margin for (SRP 3.11) equipment with operability times less than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, or provide justifications for reduced margins.

Response

The Environmental Qualification Report indicates compliance with the one hour time margin for equipment with operability times less than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> except for the P-300 series Environmental Qualification Review Records (EQRR). Justification for this exception is based on 1) an adequate degree of. margin with respect to the specified operating time requirement (see response to question Q 270.10), 2) the subsequent failure of these instruments will not cause other safety-related equipment to malfunction and 3) the subsequent failure of these instruments will not mislead the plant operator.

C 270.5 Final rule, 10 CFR 50.49. states that equipment required (SRP 3.11) to remain functional during and following design basis events be included in the qualification program.

Indicate that all design basis events (e.g. moderate -

energy line breaks, fuel handling accident, etc.) as defined in the rule, 'have been considered in the development of the list of systems and equipment submitted.

Respense All design basis events as defined in 10CFR 50.49 have been considered in the development of the LGS Environmental Qualification Report Appendix A systems list and Appendix B equipment list. Section 2.0 has been changed to provide additional information.

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l 2.0 EQUIPMENT REQUIRING ENVIRONMENTAL QUALIFICATION l

Equipment important to safety as defined in 10CFR50.49 includes  !

L both safety-related and non-safety-related equipment plus i post-accident monitoring equipment. Safety-related equipsaant is defined as that equipment which is relied upon to remain functional during and following design basis events to ensure ,(a) the integrity of the reactor coolant pressure boundary, (b) the capability to shut i down the reactor and maintain it in a safe shutdown condition, and (c) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the

. guidelines of 10CFR Part 100. Also identified as important to safety

is non-safety-related equipment whose failure under postulated ' '

environmental conditions could prevent the satisfactory accomplishment of required safety functions by safety-related equipment. Components required for display information and to perform post-accident sampling and monitoring and radiation monitoring (Regulatory Guide 1.97,

" Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident", Rev. 2, Category I and II Equipment) and TMI upgrades (NUREG-0737, " Clarification of TMI Action P1'an Requirements", Rev. 1, Equipment) have been included to the extent required therein. Specifically included, in accordance j with NRC guidance, ara those systems required to achieve or supports 4

1. Emergency Reactor Shutdown i 2. Containment Isolation j 3. Reactor Core Cooling '

4. Containment Heat Removal

5. Core Residual Heat Removal

6. Frevention of Significant Release of RadLoactive Material to' the Environment.

Limerick has established a comprehensive, systematic program for j identifying electrical equipment required to be environmentally

!. qualified. As discussed above, safety-related equipment is identified j according to the safety function objectives of 10CFR50.49(b)(1), and

is placed on the Limerick Project Q-List.

I l With respect to non-safety-related electrical equipment whose failure could prevent achieving these safety function objectives (see 4

Paragraph (b) (2) of 10CFR50.49), a review of systems interactions has been performed to ascertain which components fall into this category.

This systems interactions review took into account the following j various studies and analyses:

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a. Separation Review Program (1) High Energy Line Break. analysis

, (2) Moderate Energy Line Break and Flooding study

(3) Fire Hazard (Appendix R) Safe Shutdown study i (4) Electrical Equipment Separation per Regulatory Guide 1.75 i

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b. Cosunon Sensor Failure study

c. Control Systems Failure study

d. Reactor Vessel Water Level Instrumentation study

e. Nuclear Safsty Operational Analysis i f. Control Room Design Review f In addition, a study of the effects of high energy line breaks on control systems is in progress in response to IE Information Notice 79-22. Any components identified by these studies whose failure could prevent attainment of the safety function objectives are included on the project Q-List as appropriate. .

A separate program to verify the equipment important to safety has been completed. This program, the PECO Component Classification Program, was initiated as a sub-program of the LGS Environmental Qualification Program and it addressed all Nuclear Steam Supply (NSS) and Balance of Plant (BOP) systems and components. The Classification Program involved a re-review of the following documents:

> 1. FSAR

2. QAD's

3. P&ID's

4. Instrument Index

5. Equipment Index

6. Electrical Drawings

7. System Descriptions and Operating Manuals The postulated event analyses in Chapter 15 of the FSAR were reviewed to identify systems which have a safety-related function or support in any manner a safety-related function. A matrix of required'systemt versus postulated events was prepared and non-safety-related systems were identified. A classification sheet for each safety-related component was prepared on a system basis. Each of the component's functions in support of the Chapter 15 events were identified on the classification sheets and a five digit classification code was determined for each function. Then an overall component code was derived based on hierarchical order.

During the course of identifying the equipment which requires qualification, equipment important to safety was identified which is both subject to a harsh environment and for which exception is taken with respect to qualification to that harsh environment. In these instances the equipment meets one or more of the following exceptions:

. 1. Equipment is not required to perform its safety function to mitigate the effects of any design basis accident 'in the harsh environment, and equipment failure in ' the harsh environment will not adversely impact safety '. unctions or mislead the operator.

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2. Equipment is required to perform its safety function to mitigate the effects of a specific DBA, but is not subjected to a harsh environment as a result of that DBA.

3. Equipment performs its function before its exposure to the harsh environment, and the adequacy of the time margin provided is justified subsequent failure of the equipment

as a result of the harsh environment will not degrade other safety functions or mislead the operator.

4. The safety function can be accomplished by some other designated equipment that has been adequately qualified and satisfies the single-failure criterion. ,

Appropriate justification for the determination of one of the above categories is provided on an equipment-specific basis.

i Subsequent to identifying equipment requiring qualification, equipment locations were identified using design drawings and later verified by field inspection. All equipment locations are identified by architectural room numbers which have defined boundaries.

Appendix A, Table 1 contains the Master Systems List categorized by system safety function objective. Appendix B contains the list of electrical equipment encompassed by the Environmental Qualification Program. The equipment list identifies equipment by system included in Electrical Equipment Qualification Program. Additionally Regulatory Guide 1.97 equipment is identified on the equipment list by I

a note 0.

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l Q 270.6 Verify that your EQ program in'cludes, without exception, (SRP 3.11) all Reg. Guide 1.97, Rev. 2 Category 1 and 2 equipment in a harsh environment which is currently installed or which will be installed prior to fuel load.

Response

l All Reg. Guide 1.97, Rev. 2 Category 1 and 2 equipment located in a harsh environment is included in the EQ program. 'Ihis equipment is identified by a Note 0 in Appendix B of the Environmental Qualification Report except the Neutron Monitoring System which is exempted as noted in the response to Question 270.1 (SRP 3.11).

Q 270.7 The temperature profile, Figure 3C, shown on page C-30 (SRP 3.11) of your October 1983 submittal is not consistent with that in the FSAR. Also, the method and results of the temperature calculation in the FSAR are expected to be amended in accordance with Section 6.2.1.3.3 of the Limerick SER (NUREG-0991). Revise Figure 3C in the EQ

> program and applicable portions of the FSAR to be consistent with the temperature profile accepted by the NRC staff in NUREG-0991.

Response

The temperature profile, Figure 3C, shown on page C-30 cf the Environmental Qualification Report is consistent with both the transient analyses for the main steam line break, recirculation line break, and intermediate size line break as discussed in section 6.2.1.1.3.3 and the transient analysis for small primary steam break as discussed in the report "Drywell Temperature Response to a Small Break, A Realistic environmental Qualification Envelope, Limerick Generating Station, Units 1 and 2", General Electric Company, November 1983. As discussed in the EQR, this composite temperature profile envelopes these primary containment transients. FSAR Sections 3.11 and 6.2 are being changed to rLference this report for environmental qualification considerations following a small steam break.

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7747 LGS FSAR 3 (. within the normal limits which are shown in the EQR. FSAR Section 9.4 describes the NVAC system.

The total integrated radiation doses (TID) for normal operation l

for 40 years of equipment life have been calculated assuming a 100% load factor and rated power. The doses are based on the design radiation source terms of the radiation sources within each plant area. *

' l Aging effects on all equipment are considered in the -

i qualification program to conform to the requirements of Section,4 of.NUREG-0588. Components susceptible to aging effects are' ~ * .* .

i identified, and refurbishment and/or replacement is incorporated into the Limerick Preventive Maintenance / Surveillance Program.

Known susceptibility to aging degradation, results of inspections and manufacturer's recommendations are factored into the l Maintenance / Surveillance Program.

Effects of known normal vibratory loads on equipment are considered in the EQ program when significant.

3.11.3.2 Accident Environmental Conditions J(

j l Operability duration requirements have been determined based on 1

the length of time the equipment must maintain its ability to perform its safety function.

T'eh primary containment time dependent pressure and temperature j

profiles calculated for the spectrum of postulated LOCAs and MSLas have been calculated using NRC approved ERC' methodology.

Information on the genera on of these profiles is contained in FSAR Section 6.2J& ged 5.ll-E , i I

Temperature and pressure conditions resulting from a HELB outside containment have been determined using plant specific profiles.

These profiles bound accident environments caused by other events. FSAR Sections 3.6, 6.2, and 9.4 describe the analyses used in generating these profiles. Additional information is l

contained in the EQR. ,

l

i Post-LOCA radiation doses inside primary and secondary containments were calculated in accordance with NUREG-0737, item 11.B. 2. The source terms are consistent with those l I 3.11-3 'R,ev. 21, 06/83

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770 7 tas rsAR 3.11.9 SPARE AND REPLACEMENT PARTS q

Safety-related equipment and spare and replacement parts are being ordered to meet or exceed the original specifications. For replacement parts being procured from the original specification and which are identical to the originally supplied equipment, a certificate to support of conforaance isHowever, qualification. considered sufficient documentation if identical replacement parts are not available, environmental qualification for the new replacement parts will be demonstrated.

.11.10 MILD ENVIRONMENT QUALIFICATION Mild environment qualification is a quality assurance function and is sddressed Guide 1.33 criteria. using 10 CFR 50 Appendia 8 and Regulatory Q ,, ,

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(W^ s 6.2.1.1.3.3.5 small size Breaks h ,1 ;

6.2.1.1.3.3.5.1 Reactor System Blow r. ' ns' This section discusses the contai e ans nt s'ssociated with small primary system line breaks The sizes of primary system ruptures in this category are those that do not result in reactor depressurisation due either to loss of reactor coolant or automatic operation of the ECCS equipment. Following the occurrence of a break of this size, it is assumed that the ,

reactor operators will initiate an orderly plant shutdown and ;

depressurization of the reactor system. The thermodynamic - ' '

process associated with the blowdown of primary system fluid from such a break is one of constant enthalpy. If the primary system break is below the water level, the blowdown flow will consist of reactor water. Blowdown from reactor pressure to the drywell I pressure will flash approximately one-third of this water to steam and two-thirds will remain as liquid. Both phases will be at saturation conditions corresponding to the drywell pressure.  !

Thus, if the drywell is at atmospheric pressure (for example) the steam and liquid associated with a liquid blowdown would be at 2120F.

l If the primary system rupture is located above the RPV water level so that the blowdown flow consists of reactor steam only, the resultant steam temperature in the containment is significantly higher than the temperature associated with liquid blowdown. This is because the constant enthalpy depressurization of high-pressure, saturated steam results in superheated conditions. For example, decompression of 1000 psia saturated steam to atmospheric pressure results in 2980F superheated steam (860F of superheat).

A small reactor steam leak (resulting in superheated steam) imposes the most severe tempetature conditions on the drywell structures and the safety equipment in the drywell. For larger steam line breaks, the superheat temperature is nearly the same as for small breaks, but the duration of the high temperature condition is less for the larger break. This is because the larger breaks depressurize the reactor more rapidly than the orderly reactor shutdown that is assumed to be initiated for the small break.

3 6.2.1.1.3.3.5.2 Containment Response i

For drywell design considerations, the following sequence of events is assumed to occur. With the reactor and containment operating at the maximum normal conditions, a small break occurs that allows blowdown of reactor steam to the drywell. The resulting pressure increase in the drywell leads to a high drywell pressure signal that scrams the reactor and activates the containment isolation system. The drywell pressure continues to 6.2-16

-2N 370.7 LGS FSAR k Evaluation of NUREG-0737, Item II.E.4.2(7)," dated June 14, 1982.

6.2-23 Tonical Report OCF-1, Nuclear Containment " solation Evstem Owens - Corning Faberglas Corporat;,on (January l

1979) 6.2-24 J.E. Krueger and R.C. Sansone, " Purge and Vent Valve Operability Dualification Analysis, Report 6-06-83, prepared for Philadelphia ElectrJe Co. Limerick Generating Station Unit 1", Clow corporation (June 1983).

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! Q 270.8 Provide information on the specific maintenance / surveil-l (SRP 3.11) lance activities to be performed on 1) Cables located

! inside containment, 2) Limitorque valve operators, 3)

ASCO solenoid valves, 4) Conax electrical penetrations,

5) temperature / pressure / level sensors and transmitters.

! Response

1) Part of the environmental qualification program for cables includes pre-aging the test specimens at elevated temperatures r i for a duration based on Arrhenius calculations. The pre-aging j simulates at least 40 years service toeperature exposure within i i the IAS primary containment. Based on the successful completion

{ of testing, periodic replacements of cables are not required due

to age related degradation. It should also be noted that the i testing is conservative in view of the fact that the pre-aging j was performed in an air environment whereas the actual 3

environment will normally be inerted with nitrogen. 'Although

! maintenance of cable is not indicated based on the qualification test results, the electrical equipment in primary containment is 4 periodically tested via surveillance tests and is periodically maintained as required by the plant maintenance program. The '

i results of the surveillance tests prove the operability of not only the equipment, but also the electrical cable and interface devices. The equipment maintenance work also enables the

. maintenance personnel to physically inspect the cable and interface devices. Degradation which is detected during these l inspections will be evaluated and appropriate actions taken.

i 2) Besed on an evaluation of the environmental qualification of j Limitorque valve operators and Peach Bottom operating experience, PECO initiated a comprehensive inspection and rework program for the operators. This program has been in-progress during Limerick construction and will be completed before plant start-up. It is

, our judgment that this program will preempt the necessity for any l unusual maintenance beyond normal lubrication requirements which

are identified in the plant maintenance procedures. The valve operators are periodically tested via surveillance tests and the

l. opening and/or closing time of critical valves is measured thus '

{ providing a mechanism to detect changes in the operator j characteristics which bear investigation. In addition, following i

certain maintenance activities, the drive motor load current is  !

j measured thereby providing another means to evaluate operability and identify unusual conditions which warrant investigation.

j

3) ASCO solenoid valves have also been environmentally qualified by a program which included thermal pre-aging. Component j replacements are based on an evaluation of the in plant service conditions as compared to the test bases, pequirements which are l identified by the comparison are identified in the EQR and then

! incorporated into the computerised plant maintenance program. In l general, plant surveillance tests periodically exercise the

solenoid valves and prove operability.

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4) The conan penetrations have been analysed for susceptibility to ,

thermal degradation based on material tests and the Limerick i specific service conditions. The result of the analysis is that the penetrations have a qualified life in excess of 40 years 1 therefore specific qualification related surveillance tests have -

not been developed, however, as indicated in the preceding response pertaining to cables in primary containment, equipment surveillance tests prove the operability of electrical interface devices in addition to the equipment itself.

5) Instrumentation has been environmentally qualified in a program which includes thermal pre-aging. Maintenance requirements for all equipment in the Limerick Environmental Qualification Program  !

are provided on the individual Environmental Qualification Review l Records. This information is reviewed by plant personnel who ,

then incorporate the specific requirements into the computerised plant maintenance program. Instrumentation operability is checked ;

periodically via surveillance tests. l t

Q 270.9 A number of EQRR sheets state that deficiency (SRP 3.11) resolutions are to be completed by December 1,1983.

Please provide the revised qualification status and EQRR for those component items for which the i deficiencies have been resolved. i

Response

Revisions have been made to the EQRR shee4.s for which i deficiencies have been resolved. In general most deficiencies are expected to be resolved by February 13, 1984. An exception will be one test program on Yeam/Litton in line plug connectors (ref. SQRR 44) which will be in progress with an espected completion of 4/1/84. (

Q 270.10 A number of discrepancies have been noted in the EQRR l (sRP 3.11) sheets, for example: i EQRR 1, pages 764 through 777, states that the required  !

accuracy is to be determined yet the equipment is listed as qualified with no outstanding items.

EQRR 142,'pages 495 through 568, reference 119 is used r as the qualification document for Asco solenoid release model x8018A4. This model' cou}d not be located in reference 119.

t BQRR 1812, pages 569 through 588 and 1246 through 1255, ,

the accuracy requirement is not specified yet the  ;

equipment is listed as qualified with no outstanding }

items, i k

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I EQRR P311, page 1095, shows inadequate margin for operating time and radiation. EQRR P106, page 1300, specified radiation is not enveloped.

EQRR pages 815 and 853 show deficiencies in temperature and/or pressure but no resolutior; is stated.

These and all other similar discrepancies must be eliminated. Revised EQRR sheets should be resubmitted prior to the site audit to reflect any changes in the qualification status.

Response

The EQRR sheets which contain discrepancies or are incomplete have been changed to provide current status.

Q 270.11 Appendix C, table 1 of the EQ submittal references (SRP 3.11) notes in superscript letters appearing in columns titled " pressure", " temperature" and " humidity". These reference notes appear to have been omitted. Please provide reference notes a through q. 'j

Response

only superscript letters a through g are used to identify the ;j particular line break accident resulting in the values of temperature, pressure and relative humidity given in Table 1. The identification of these breaks is provided as Note 4 of Table 1 on page C-11 of the ECR. <

These same letters (except capitals) are used on the Appendix E EQRR ,

sheets.

Q 270.12 Indicate what actions will be taken for equipment items (SRP 3.11) whose " qualified life" is less than 40 years.

Response

Qualification maintenance requirements are identified on the individual EQRR sheets. This information is reviewed by plant operating personnel who incorporate the specific requirements into the computerized plant maintenance program. This computerized program ensures that required maintenance is automatically brought to the attention of responsible plant personnel so that appropriate refurbishment can be performed to maintain the equipment's qualified life.

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