ML20100N426
| ML20100N426 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 10/19/1984 |
| From: | LONG ISLAND LIGHTING CO. |
| To: | |
| References | |
| OL-A-008, OL-A-8, NUDOCS 8412130133 | |
| Download: ML20100N426 (63) | |
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I SHNPP FSAR b
-3.11 ENVIROM4 ENTAL DESIGN OF ELECTRIC AND MECKANICAL EQUIRiENT
'3.11.0 GENERAL l
Equipment that is relied on to perform a necessary safety function must be demonstrated to be capable of maintaining functional operability under all service conditione postulated to occur during.its installed life for the time it is required to operate. This requirement, which is embodied in General Design Criteria 1, 2, 4, and 23 of Appendix "A" and Sections III and XI of j
Appendix "B" to 10CFR50, is applicable to equipment located inside and outside containment. More detailed requirements and guidance relating to the methods and procedures for demonstrating this capability have been set forth in
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10CFR50.49.
4 16 s
The purpose of this section is to provide information on the environmental conditions and design bases for which safety related electrical and mechanical equipment is designed to ensure compliance with the above.
In addition, this section describes the applicants' environmental qualification program and methodology for compliance with NUREG-0588 Category II guidelines and therefore 10CFR50.49.
This section consists of a written description, tables, figures, appendices, and data references describing the equipment qualification f or safety-related Class IE components used in the plant.
Descriptions of these tables, figures, appendices, and ref erences are as follows:
Table 3.11.0 This table lists the NSSS supplied safety-related equipment with the applicable qualification reference indicated.
I Table 3.11.0 This table lists the Ebasco supplied safety-related equipment.
i Table 3.11.0 This table lists the design criteria used in safety-related equipment for non-seismic vibrations.
Table 3.11.1 This table defines the location codes used in the Master List.
Figure 3.11.1 This figure provides the format and legend for the SENPP i
" Master List."
Figure 3.11.1 This figure provides a legend for the SENPP Component Evaluation Sheet.,
Appendix 3.11A - This appendix co'ntains the NUREG-0588 Comparison.
16,
Apendix 3.115 - This appendix contains the Containment and Reactor Auxiliary Building Zone Maps for temperature and radiation.
Figures 3.118-1 through 3.118-19 are the Post Accident Temperature in Spaces Cooled by ESF KVAC 16 Systems.
Figures 3.118-20 through 3.118-29 are the zone maps for,the Integrated Radiation Doses to Equipment During Normal and Post Accident Environments.
i 3.11.0-1 Amendment No.16
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SHNPP FSAR I:
Appendix 3.11C - This appendix contains supplemental analyses and their results used to demonstrate the thermal response of saf ety-related equipment located inside Containment, and the subsequent ability to survive and operate during and af ter the design basis accident 16 WCAP 8587, Supplement No. 1 - This qualification reference indicates the individual qualification details for each particular type of eqaipment,
meeting IEEE 323-1974, supplied by the NSSS Vendor, Westinghouse.
This WCAP and supplement are not contained in the FSAR and are generic reference documents for all NSSS supplied IE equipments meeting IEEE 323-1974.
WCAP-7410-L, WCAP-7744 and the Westinghouse Environmental Supplemental Qualification Testing Program (see Westinghouse Letter NS-CE-692, C. Eicheidinger to D. B. Vassallo, July 10, 1985, and NRC Letter from D. 8.
Vassalo to C. Eicheidinger, November 19, 1975) - This qualification reference indicates the qualification details for equipment suppied by the NSSS Vendor, j
Westinghouse which meets IEEE 323-1971.
These WCAPs and the Supplemental Program are not contained in the FSAR and are generic reference documents for all NSSS supplied IE equipment meeting IEEE 323-1971.
I The design environmental criteria for safety-related electrical and mechanical 16 f
equipment are based on equipment location.
Radiation Environment for qualification of electrical and mechanical equipment is based on radiation doses calculated using source terms and methodology discussed in NUREG-0588, I
NUREG-0588 Rev.1, and Section II-8.2 of NUREG-07 37.
As far as practical, equipment for these systems is located outside the Containment Building or other areas where high radioactivity levels or adverse environmental conditions could exist under normal, test, or accident conditions.
I 16 l Safety-related equipment are capable of perforaing their intended functions under the following specified environmental conditions:
l a)
All safety-related components are capable of meeting their rated performance specifications under the environmental service conditions expected as a -result of ncrual operating requirements, including the range of expected minimum and maximum environmental conditions.
16 l b)
All safety-related equipment are capable of completing their functions under the environmental service conditions related to the design basis accident.
The environmental service conditions related to a design basis accident are specified to include:
normal operating conditions existing before the event, conditions generated by the event, and conditions which exist subsequent to the event for such time as is required f or the protective actions to be carried to completion.
l 3.11.0-2 Amendment No. 16
i, -.
SENFP FSAR TABLE 3.11.0-1 NSSS SUPPLIED SAFETY-RELATED EOUI1 MENT Hodel or Qualification (c) Qualification PER/IEEE-323 Drawing Equipnent Manufacturer Number Reference (1971-1974)
Wide-Range W
763 ESE-1A 1974 16 Reactor Trr sarton (Group A)
Coolant Pressure Transmitter Wide-Range W
76PH2 ESE-18 1974 Reactor Teritrak Coolant Pressure 16 Transmitter Pressurizer W
763 ES E-1 A 1974 Pressure TIT Barton Transmitter Pressurizer W
7 64 ESE-3A 1974 16 Level IIT Barton (Group A)
Transmitter Steam W
7 64 ESE-3A 1974 Generator ITT Barton 76DP2 ESE-3B 1974 Level WR&NR Veritrak (Group A)
Transmitter Reactor W
752 ESE-4A 1974 Coolant TIT Barton (Group B)
Flow 16 Transmitter Steam Flow W
7 64 ESE-3A 1974 Transmitter TIT Barton (Group A)
Narrow-Range RdF 21204 ESE-5 1974 Reactor-Coolant Temperature Detectors Wide-Range RdF 21205 ESE-6 1974 16 Reactor-Coolant Temperature Detectors 3.11.0-3 Amendment No. 16
SHNPP FSAR TABLE 3.11.0-1 (Cont'd)
NSSS SUPPLIED SAFETY-RELATED EQUIR4ENT Model or Qualifleation l 16 Drawing Qualification PER/IEEE-323 Equipment Manufacturer Number Reference (1971-1974)
Containment W
351
, ESE-21 1974 l 16 Pressure LTT Barton Sensor Electric W Sturtevant Model A
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Hydrogen 02D0448 Rev. O I
Recombiners (HAL.)
! 10 and Equipment
'Halamar CP-6070-1 Rev. 2 SP-1 1974 Shaffer (SHAF.)
Eng.
CP-4070-1 Rev. 1 (SHAF.)
Excore W - IGTD 24154 ESE-8A 1974 Detectors Stem Mounted NAMCO EA-180 NAMCO Tes t 1971 Limit Switches Report 11/21/77 Inside Containment Valve Motor Limitorque SMB WCAP-7410-L 1971 Operators Class H WCAP-7744 Inside NS-CE-692 Containmen:
Valve ASCO Various NS-CE-755 1971 Solenoid Operators Inside coneainmenc Steam W
763 ESE-1A 1974 16 Pressure ITT Barton (Group A)
Transmitter Turbine W
753 ESE-2 1974 Pressure TTI Barton (Group B)
Transmitter Containment W
752 ESE-4 1974 Pressure TTT Barton (Group B)
Transmi).:t 9r PAM W - RID VX252 ESE-14 1974 Indicators 3.11.0-4 Amendment No. 16
.. SHNPP FSAR' TABLE 3.11.0-1 (Cont'd)
NSSS SUPPLIED SAFETY-RELATED EQUIIM ENT Model or Qualification 16 Drawing Qualification PER/IEEE-323 Equipment Manufacturer Number Reference (1971-1974)
PAN W - CID Optimal ESE-15 1974 Recorders 100 16 Refueling W
752 ESE-4 1974 l
Water TTT Barton (Group B)
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Storage Tank Lw el Transmitter l16 Feedvater W
752 ESE-4 1974 Flow TTT sarton (croup a)
Transmitter Component W
753 ESE-2 1974 16 Cooling Heat Trr sarcon (croup a)
Exchanger Discharge Pressure Process WISD 7300 ESE-13 1974 Protection System Nuclear W NICD 1054E26 ESE-10 1974 16 Instrumen-Rw. D tation Systen 16 Solid W NICD 2-Train ESE-16 1974 State Protection System Main W
1139E34 WCAP-10469 1974 16
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Control 1139E35 WCAP-10369 Board 1139E36 Main Termination Cabinets Safeguards W NICD 2-Train ESE-16 1974 16 Test Cabinets 3.11.0-5 Amendment No. 16
SHNPP FSAR TABLE 3.11.0-1 (Cont'd)
NSSS SUPPLIED SAFETY-RELATED EQUIR1ENT Model or Qualification ( ) Qualification l16 Drawing PER/IEEE-323 Equipment Manuf acturer Number Reference (1971-1974)
RVLIS W - ID MULT 1 ESE-50C(d) 1974 (Reactor Vessel Lev el Insertamentation 1.6 System)
Instrument W PEDS 7-1/2 KVA,
ESE-18 1974 Static 1 Phase t
Inv erter 60 Cycle l
Inst. Bus Inv erter Reactor W LVSD Type DS416 ESE-20 1974 16 Trip RTS Switchgear Stem Mounted NAMCO EA170 Fisher Control 1971 Limit D-2400X Test #72AR73 16 Switches and 1529 (Outside Containment)
Valv e Limitorque Various Limitorque 1971 Motor Report #800 Operators 6/7/76 (Outside Containment)
Limitorque 16 Report # 600456 11/22/74 l
l Valv e ASCO FT831654 NS-CE-755 1971 Solenoid HT8300854-RF l
Operators (Outside Containment)
Componenc W IMD 8249D36 AE-2 1974 16 Cooling Water Pwsp Motor 3.11.0-6 Amendment No. 16
SHNPP FSAR i
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TABLE 3.11.0-1 (Cont'd)
I L
NSSS SUPPLIED SAFETY-RELATED EQUIIM ENT i.
Model or Qualification (c) Qualification PER/IEEE-323 16 Drawing Equipment Manufacturer Number Reference (1971-1974)
Residual WIMD 8246D34 AE-2 1974 16 Heat Rosa al Pump Motor Centrifugal W IMD 8241D38 AE-2 1974 16 Charging Pump Motor Boric Acid Champump 862239 Champump 1974 16i Transfer Report A18187 Pump Motor l
NOTES:
(A) This note has been deleted.
(B) This equipment is not required to perform its function under severe post-accident ew ironmental condition. Qualification type testing for this equipment is described in WCAP-7410-L, WCAP-7744 and the Westinghouse Ewironmental Supplement Qualification Testing Program.
(C) EQDP's from WCAP-8587 Supplement 1.
(D) This uabreila report references EQDPs ESE-4, 15, 42, 48, 49, and 53 as 16 identified in WCAP S387.
3.11.0-7 Amendment No. 16
SiiNPP FSAR TABLE 3.11.0-2 EBASCO PURCHASED SAFETY RELATED EQUIPMENT GUALIFICATION MODEL PER IEEE-323 EQUIPMENT SUPPLIER NUMBER (1971 OR 1974) 6.9 kV Metal Clad Siemens-Allis F1.500 Al 1974 i
Switchgear (15 kV Class) 480 V Metal Gould-Brown Type LK 1974 Enclosed Boveri Switchgear l
Motor Gould-Brown Series 5600 1974 Control Boveri Centers Einergency Transamerica DSRV-16-4 1974 Diesel DeLaval Engine and Generator 3
Energency Transamerica N/A 1974 Diesel DeLaval Generator Control Panel 15 kV Pouer Anaconda N/A 1974 Cable 600 V Power &
Kerite 4/A 1974 Control Cable 300 V American N/A 1974 Instrumentation, Insulated Communication, Wire and Computer Input Cable Thermocouple Samuel Moore N/A 1974 Cable
& Company
- lower, Anaconda N/A 1974 Control and Ins trumentation Cable t
Amendmen t No. I
SHNPP FSAR TABLE 3.11.0-2 (Continued)
QUALIFICATION MODEL PER IEEE-323 EQUIPMENT SUPPLIER NUMBER (1971 OR 1974)'
Emergency Diesel Transamerica N/A 1974 Generator - Engine DeLaval Control Panel N/k Emergency Diesel-Transamerica 1974 Motor Controller DeLaval Control and American Insulated N/A 1974 Instrumentation Wire Corp.
Cable Triaxial Cable Boston Insulated N/A 1974 Wire & Cable Co.
125 V DC Gould-Brown FC-1 1974 Distribution Boveri Panels 3
125 V Batteries C & D Batteries LC-19 1971 125 V DC Battery.
C & D Batteries ARR 130 HK 150F 1971 Chargers Containment Electrical
'Jes tinghouse WX-33527 1974 Penetrations:
WX-33528 Low-Voltage Power.
WK-33881 Control and WX-33453 througt Instrumentation ifX-33487 Containment Electrical Westinghouse WX-33452 1974 Penetrations Medium Voltage Electronic Rosemount 1153 Series 8 1974 Instrumentation Thermocouple WEED Instrument N/A 1974
.tssemblies and Test Thermowells 1
Auxiliary Control Reliance Electric N/A 1974 Panel Local Instrument Mercury Company of N/A 1974
' Cabinets & Racks Norwood, Inc.
Auxiliary h lay Systems Control N/A 1974 Cabinets & Racks Corp.
^*"d*"'***'I 3.11.0-9
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SHNPP FSAR TABLE 3,11.0-2 (Continued)
OUALIFICATIO*i MODEL PER IEEE-323 EOUIPME!rr SUPPLIER NUMBER (1971 OR'1974)
, Isolation. Panels Consolidated Control
' N/A 1974 Corp.
Transfer Panels Systems Control Corp.
N/d 1974 Sequencer Panels Systems Control Corp.
N/A 1974 Level Switches Magnetrol A-153F 1974 International 17-7-75 Level Transmitters Transamerica RE-36562 1974 DeLaval XM-36495 Low Range Otfferential FLO-TEK NS-10RA-2A2 1974 Pressure Transmitters Low Range Flow Switches Fluid Components FR-72-4 1974 FR-72-4R 3
Containment Fan American-Air Filter N/A 1974 Coolers Water Chillers York 1974 Chilled Water Goulds Pumps 1974 Circulating Pumps Air Mandling Units Bahnson 1974 Centrifugal Fans Barry Biower 1974 Axial Flow Fans Joy Manufacturing 1974 In-Line Fans Joy Manufacturing 1974 Electric Heating Ooils Brasch Manufacturing 1974 Dampers bskin Manufacturing 1974 Air Cleaning Units CTI - Nuclear 1974 Tornado Protection Qun11ty Air Design N/A Dampers
\\
e Amendmen t No. 1
SHNPP FSAR' TABLE 3.11.0-2 (Continued)
QUALIFICATI0*i MODEL PER IEEE-323 EOUIPMENT SUPPLIER NUMBER (1971 OR 1974)
Butterfly Valves BIF - Unit of 1974 General Signal Auxilitry Steam Inge rsoll-Rand 3 HMTA-9-Pump 1974 Generator Feed 5008P39-Motor Pumps & Motor Frame Auxiliary Steam Ingersoll-Rand 4x9-NH-7-Pump 1974 Generator Feed GS-2N Turbine Pumps & Turbine Service Water Goulds Pumps 3405L-Pump 1974 Booster Pumps 447TS-Motor
& Motor Frame Spent Fuel Pool Goulds Pumps 3405L-Pump.
1974 Cooling Pump 445TS-Motor
& Motor Frame 3
Diesel Oil Goulds Pumps 3196ST-Pump 1974 Transfer Pump 213T-Motor
& Motor Frame Containment Ingersoll-Rand 8 x23WDF-Pump 1974 Spray Pump 5008P39-Motor
& Motor Frame Feedwater Bo rt-Wa rne r 16"x16" 1974 Isolation Flex 1 wedge Gate and Check Valves 2 1/2" & Larger Pacific Valves 180-7-WE-X 1974 C.S. Valves 150-7-WE-X 160-7-WE-X 2 1/2" & Larger Rockwell 4" & 6" N/A C.S. Valves 900d International 1911 RHJTY Diaphragm Valves ITT Grinnell Valve 1974 Motor & Manual Jamesbury Corp.
8229, 6226 1974 Operated Butterfly Valves j
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3.11.0-11 Amendmen t No. 3
SHNPP FSAR TABLE 3.11.0-2 (Continued)
OUALIFICATION MODEL PER IEEE-323 EQUIPMEW SUPPLIER NUMBER (1971 OR 1974)
Main Steam Power Control Components OX69-XS-X8BW-10BW 1971*
Operated Relief Valves Self Cleaning R. F. Adams Co.
MD5S-80 1974 Strainers Mis cellaneous ITT/Hassel Dahl Valves: MD/C-1974 Control Valves 830, 672, 502 and Accessories Operators:
ITT G/C; E/H-NH-90, 92 Misc. Control Masonellan Inc..
48-40411 1974 Valves and Accessories Emergency Mayward Tyler 1974 Service Water Pump Co.
Pumps & Motors 3
Solenoid Target Rock Corp.
1021010 1974 Operated Globe 1032110 Valves ESW Intake Allis-Chalmers 10'x8' 1974 Structure Rectangular Butterfly Valves Streassaal Emergency Service Crane-Deming 3065-AIO-Pump 1974 Water Screen Wash Reliance 254T-Motor Pumps and Motors Frame 2 1/2-Inch & Larger Anchor / Darling 1974 Motor Operated Valves 2-Inch & Smaller Rockwell 1974 Motor Operated Valves Plug Valves Tufline 1974 Packless Globe Valves Kerotest 1974 Traveling Water Screens Envirex 1974
(*)Under Negotiation To Up-Date To IEEE 323-74 G
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SMNPP FSAR P
TABLE 3.11.0-3 3-SAFETY RELATED EOUIPMENT DESIGN FOR NONSTISMIC VIBRATION Equipment Standard or Requirement
_'A.
Containment Spray Pump Pump bearing housing and pump shaf t Component Cooling Water Pumps
-vibration double amplitude is limited RER Pumps to.003 inches Charging Pumps Chilled Water Pumps Emergency Service Water Pumps Bearing housing and shaft vibration limited to.005 inches double amplitude B.
Auxiliary Feedwater Pumps Pump bearing housing and pump shaf t vibration peak to peak is limited to 0.0021 inches at speeds up to 110 percent rated speed C.
All other Safety Related Pumps API 610 or better 9.
All Electrical Motors NEMA Standard 9GI-12-05 E.
Containment Fan Coolers Fan bearing housing vibration limited to.002 inches double amplitude F.
All other HVAC Equipment ASHRAE Systems handbook t
3.11.0-13
SHNPP FSAR 3.11.1 EQUIPMENT IDENTIFICATION AND ENVIRONMENTAL CONDITIONS 3.11.1.1 Equipment Identification The methodology to determine which equipment important to safety is to be environmentally qualified is based on the IE Bulletin 79-01B approach of reviewing plant systems which perform safety functions. The equipment within such systems, which are necessary for the performance of the safety function, are identified and qualified environmentally to demonstrate acceptable performance throughout its installed life.
~
Plant safety related systems _are identified in FSAR Table 3.2.1-1.
The specific equipment, within safety related systems, which is environmentally qualified, is identified on separate master lists submitted to the NRC.
Figure 3.11.1-l'and applicable notes provide the format and legend for the Shearon Harris Nuclear Power Plant-master list for electrical safety related equipment. All equipment defined in the scope of 10CFR50.49 is included in the Shearon Harris EQ Program.
16 3.11.1.2 Environmental Conditions Normal and accident environmental conditions are explicitly identified in various FSAR sections. The figures contained in FSAR Section 3.11B show general plant areas. Superimposed on these figures, in tabular form, are the environmental conditions used for qualification-purposes. The individual Component Evaluation Sheets (CES) for qualified safety related equipment summarize the environmental conditions to which a specific Leem is qualified.
Figure 3.11.1-2 and applicable notes provide the format and legend for the Shearon Harris Nuclear Power Plant CES for electrical safety related equipment.
CES are included, for each piece of equipment, in the appropriate environmental qualification documentation package, which substantiates qualification in detail.
3.11.1-1 Amendment No. 16
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TABl.E 3.11.1-1 g5 SHEARON liARRIS NUCLEAR POWER PLANT SAFETY RELATED EQUIP' TENT LOCATION CODES IDEN PC AREA EXCL XttlN X? TAX Y}lIN Y) TAX ZttlN ZitAX RCYL WPII I WASTE PROC EL 21I i192.00 1382.90 1655.00 1945.00 211.00 234.00' O.00 WP21 2 WASTE PROC EL 236 1192.00 1382.90 1655.00 1945.00 235.00 259.00 0.00 WP21 2 WASTE PROC EL 236 EXCL 1288.00 1382.90 1766.00 1834.00 234.00 260.90 0.00 WP22 3 WP CTL Rtt & VAULT 1288.00 1382.90 1766.00 1834.00 235.00 259.90 0.00 AB52 4 RAB EL 305-CONT Rtl 1537.00 1764.00 1513.00 1699.90 304.00 323.00 0.00 AB52 4 RAB EL 305-CONT Rtt EXCL 1737.00 1764.00 1572.00 1699.90 304.00 330.00 0.00 AB52 4 RAB EL 305-CONT Rtt EXCL 1537.00 1565.00 1570.00 1699.90 303.00 324.00 0.00 AB52 4 RAB EL 305-CONT Rtt EXCL 1537.00 1592.00 1640.00 1699.90 303.00 324.00 0.00 AB01 5 RAB EL 190 1383.00 1650.00 1513.00 1699.90 190.00 215.00 0.00 AB01 5 RAB EL 190 EXCL 1500.00 0.00 1700.00 0.00 189.00 450.00 65.00 w
AB21 6 RAB EL 236 1383.00 1740.00 1513.00 1699.90 235.00 259.90 0.00
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AB21 6 RAB EL 236 EXCL 150'.00 0.00 1700.00 0.00 190.00 450.00 65.00 5
0 AB31 7 RAB EL 261 1383.00 1710.00 1513.00 1699.90 260.00 284.90 0.00 I
AB31 7 RAB EL 261 EXCL I476.00 1525.00 1570.00 1647.00 259.00 310.00 0.00 C
AB31 7 RAB EL 261 EXCL' 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 AB31 7.RAB EL 261 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 AB32 8 RAB EL 261-305 1476.00 1525.00 1570.00 1647.00 260.00 303.00 0.00 AB32 8 RAB EL 261-305 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 Tall 9 TANK AREA IINIT I 1319.00 1382.90 1513.00 1654.90 230.00 340.00 0.00 SWil 10 SEC WASTE EL 216 1383.00 1453.00 1700.00 1900.00 216200 234.90 0.00 SWil 10 SEC WASTE EL 216 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 SW21 11 SEC WASTE EL 236 1383.00 1453.00 1700.00 1900.00 235.00 259.90 0.00 SW21 11 SEC WASTE EL 236 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 TB31 12 TURB ELS 240-261 1278.00 1710.00 1345.00 1513.00 240.00 2H4.90 0.00 g
TB41 13 TURB ELS 266-314 1278.00 1710.00 1345.00 1513.00 285.00 340.00 0.00 Fillt 14 FUEL llDLG EL 216 1435.00 1917.00 1700.00 1900.00 216.00 234.90 0.00 Fill t 14 FUEL liDLC EL 216 EXCl.
1500.00 0.03 1700.00 0.00 190.00 450.00 65.00 AB43 15 RAB EL 286-SWCR Rtt 1383.00 1589.90 1513.00 1602.00 285.00 303.90 0.0o AB43 15 RAB EL 286-SWCR Rtt EXCL 1565.00 1589.90 1513.00 15 ?).00 2H4.00 310.00 0.00 SF AB43 15 RAB EL 286-SWGR R}t EXCL I476.00 1525.00 1570.00 1602.00 284.00 310.00 0.00 AB41 16 RAB EL 286 1383.00 1650.00 1513.00 1699.90 285.00 103.90 0.00 C
AB41 16 RAB EL 286 EXCL I476.00 1525.00 1570.00 1647.00 259.00 310.00 0.00
TABLE 3.11.1-1 (Cont 'd )
SHEARON HARRIS NUCLEAR POWER PLANT 15 SAFETY RELATED EQUIP.'!ENT LOCATION CODES IDEN PC AREA EXCL XttlN XttAX Yt!IN Y!!AX ZillN ZtlAX
- RCYI, AB41 16 RAB EL 286 EXCL 1383.00 1565.00 1513.00 1602.00 284.00 310.00 0.00 AB41 16 RAB EL 286 EXCL 1565.00 1589.90 1570.00 1602.00 284.00 310.00
'O.00 AB41 16 RAB EL 286 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 FH21 17 FUEL HDLC EL 236 1453.00 1917.00 1700.00 1900.00 235.00 259.90 0.00 FH21 17 FUEL HDLC EL 236 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 ABil 18 RAB EL 216 1383.00 1749.00 1513.00 1699.90 215.00 234.90 0.00 ABil 18 RAB EL 216 EXCL 1590.00 1710.00 1610.00 1699.90 214.00 236.90 0.00 ABil 18 RAB EL 216 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 WP31 19 WASTE PROC EL 261 1192.00 1382.90 1655.00 1945.00 260.00 289.90 0.00 WP31 19 WASTE PROC EL 261 EXCL 1192.00 1382.90 1766.00 1905.00 274.00 290.90 0.00 SW31 20 SEC WASTE EL 261 1383.00 1453.00 1700.00 1900.00 260.00 284.90 0.00 m
u SW31 20 SEC WASTE EL 261 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 FH31 2I FUEL HDLC EL 26I I453.00 2016.00 1700.00 1900.00
'260.00 284.90 0.00 e--
Fil31 21 FUEL HDLG EL 261 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 m
I XY31 22 XFilR YARD 1380.00 1620.00 1145.00 1346.90 250.00 300.00 0.00 Db WT31 23 WATER TREAT BLDG 750.00 1042.00 2060.00 2310.00 250.00 350.00 0.00 WT31 23 WATER TREAT BLDC EXCL 935.00 1026.00 2060.00 2106.90 249.00 351.00 0.00 DF31 24 DEISEL F0 STR BLDG 2140.00 2240.00 1852.00 1948.00 240.00 300.00 0.00 BA31 25 AUX BLR AREA 724.00 1025.90 2012.00 2475.00 250.00 300.00 0.00 BA31 25 AUX BLR AREA EXCL 823.90 1025.90 2400.00 2475.00 249,00 301.00 0.00-BA31 25 AUX BLR AREA EXCL 724.00 1025.90 2l06.90 2399.90 249.00 101.00 0.00 BA31 25 AUX BLR AREA EXCL
.724.00 949.90 2032.00 2l06.90 249.00 301.00 0.00 IE31 26 INTAKE STR-EllER SW 44.00 173.00 1475.00 1685.00 230.00 300.00 0.00 SS31 27 Ett SCREEN STRUCT 160.00 300.00 2130.00 2200.00 230.00 300.00 0.00 y
IS31 28 INTAKE STRUCT-SW 1560.00 1750.00 415.00 525.00 250.00 300.00 0.00 AB51 29 RAB EL 305 1458.00 1545.00 1513.00 1654.00 304.00 325.00 0.00 ABSI 29 RAB EL 305 EXCL 1537.00 1545.00 1513.00 1570.00 303.00 321.00 0.00 y
WP41 30 WASTE PROC EL 276 1192.00 1382.90 1766.00 1905.00 275.00 289.90 0.00 WP51 31 WASTE PROC EL 291 1192.00 1382.90 1658.00 1945.00 290.00 340.00 0.00 Fil41 32 FUEL HDtJ: EL 286 1411.00 2016.00 1700.00 1900.00 285.00 303.90 0.00 FH41 32 FUEL llDLG EL 286 EXCL 1500.00 0.00 1700.00 0.00 190.00
'450.00 65.00 C
CBit 33 RCB EL 221 1500.00 0.00 1700.00 0.00 190.00 235.00 65.00
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.;]
TABLE 3.II.1-1 (Cont 'd )
15 SHEARON HARRIS NUCLEAR POWER PLANT SAFETY RELATED EQUIP?!ENT LOCATION CODES IDEN PC AREA EXCL XHIN X) TAX YHIN YMAX ZHIN ZtlAX HCYL CB21 34 RCB EL 236 1500.00 0.00 1700.00 0.00 236.00 260.00 65.00 CB31 35 RC8 EL 261 1500.00 0.00
-1700.00 0.00 261.00 285.00 65.00 CB41 36 RC8 EL 286 1500.00 0.00 1700.00
.0.00 286.00 450.00 65.00 IC31 37 INTAKE STRUCT-CW 1220.00 1335.00 620.00 780.00 250.00 300.00 0.00 CT31 38 COOL'C TOWER 1210.00 1500.00 420.00 150.00 250.00 650.00 0.0d15 DC31 39 DIESEL CEN BLDG 1573.00 1727.00 1057.00 1180.00 240.00 350.00 0.00 2S31 40 230KV SWYD 330.00 790.00 525.00 1I80.00 250,00 280.00 0.00 FH51 41 FUEL HDLC EL 305 1411.00 20I6.00 1700.00 1900.00 304.00 322.90 0.00 FH".; 41 FUEL HDLC EL 305 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00 FHal 42 FUEL HDLC EL 324 1411.00 2016.00 1700.00 1900.00 323.00 340.00 0.00 FH61 42 FUEL HDLG EL 324 EXCL 1500.00 0.00 1700.00 0.00 190.00 450.00 65.00
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CF01 43 INT STR-CPE FR RIV 7425.00 7575.00 7725.00 7875.00 137.00 230.00 0.00 5
C HD01 44 HAIN DAH SPILLWAY 5425.00 5575.00 5725.00 5875.00 195.00 270.00 0.00 SB31 45 SERVICE BLDG 950.00 1055.00 1700.00 1955.00 250.00 350.00 0.00 E
CS31 46 CAS STOKAGE BLDG 965.00 1040.00 2400.00 2575.00 250.00 305.00 0.00
$i 15 YD31 49 Y D R AREA #1 390.00 1650.00 60.00 1800.00 240.00 275.00 0.00 YD32 50 Y D R AREA #2 1650.00 2600.00 60.00 1800.00 240.00 275.00 0.00 YD33 51 Y D R AREA #3 1650.00 2600.00 1800.00 3400.00 240.00 275.00 0.00 YD34 52 Y D R AREA #4 390.00 1650.00 1800.00 3400.00 240.00 275.00 0.00 ao.
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SENPP FSAR 3.~ 11. 2 QUALIFICATION TESTS AND ANALYSIS
- Environmental qualification testing and/or analysis based on tests are performed on safety related equipment located in a harsh environment.
The results are evaluated for compliance with the Category II NUREG-0588 guidelines.
Nuclear Steam Supply System (NSSS) Class 1E equipment is qualified under the Westinghouse environmental qualification program as stated in Westinghouse Topical Report WCAP 8587. This report describes the basic methodology on which the Westinghouse qualification program is based and includes qualification methods used for harsh environment Class 1E equipment.
The NRC has reviewed and accepted the generic qualification methodology described in Westinghouse Topical Report 8587.
The applicants review the report to verify applicability to Shearon Karris.
f Specifically, all reviews consider but are not LLaited to the following:
a)
^ Assurance that the test report is applicable to SHNFP. This is accomplished by assuring that the project name, purchase order and equipment specification as a minimum are identified on or traceable to the report.
t)
A comparison of the test sample is made to assure that the equipment i
tested is identical to or representative of the purchased equipment.
c)
The aging (radiation, humidity, temperature, electro-mechanical 16 cycling, etc., as required) simulation is evaluated to detensine if the test equipment has been placed in a condition which simulates its expected end of qualified life condition prior to design basis accident testing.
Process temperatures, when applicable, are addressed.
d)
The design basis accident environmental test conditions (temperature, pressure, chemical spray, etc.) are evaluated to determine if they envelop the Shearon Harris expected environmental conditions in the unlikely event of a design basis accident.
e)
Anomalies observed during qualification testing are evaluated.
In addition, other items such as test sequence, margin, interf aces are also addressed during the environmental qualification report review process.
Compliance with the various NRC Regulatory Guides and General Design Criteria is described in FSAR Sections 1.8 and 3.1, respectively.
4 i
8 I
3.11.2-1 Amendment No. 16 i
e-s.
515PP FSAR 3.11.3 QUALIFICATION TEST RESULTS A summary of the harsh environment qualification test results for each type of qualified safety related equipment is provided in the individual Component Evaluation Sheet (CES) for each equipment.
Refer to FSAR Section 3.11.1.2 for a discussion of CES. The CES identifies the applicable environmental qualification documentation package which substantiates qualification in detail. Documentation packages are prepared for equipment groups by type (e.g., all Target Rock Solenoid Operator harsh environment qualification documents are contained in a single documentation package).
Typical documents which are included in the environmental qualification documentation packages are:
a)
Equipment Functional Description and Summary,
b)
Component Evaluation Sheets, c)
Equipment Specifications, 16 d)
Environmental Qualification Report (s),
e)
Supplementary Review / Analysis Sheets which provide analysis /
calculations performed to demonstrate qualification to each applicable environmental parameter, f)
Review Cuidelines and Checklist which discuss environmental conditions, testing, aging and replacement, interfaces and maintenance considerations, g)
Drawing (s) showing equipment details, and h)
Open Items.
The various documentation packages are-permanently stored and maintained at the Shearon Harris Nuclear Power Plant.
l J
~
Amendment No. 16 s
.d:
SENPP FSAR 3.11.4 LOSS OF VENTILATION :
3.11.4.1 Equipment Qualification Plant areas lcontaining safety related equipment and their support systems are temperature controlled to provide a controlled environment during normal and most severe DBA conditions. _During normal plant operating conditions, plant area environments are less than or equal to those shown in Appendix 3.115.
16 Safety related toeperature controlling equipment, located in a harsh environment is environmentally qualified in the same manner as other safety related equipment in the same plant areas.
3.11.4.2 Air Conditioning Systees The Seismic Category I and Safety Classes 2 and 3 Air Conditioning Systems, are powered from Class 1E electrical power supplies and are provided for the locations described in Section 9.4.
They are designed such that the single failure of an active component, after a design basis accident, cannot impair the ability of the systems served by the air-conditioning equipment to fulfill their safety fur.ctions. Should the air-conditioning unit in one of the rooms in a Seismic Category I, Safety Class 2 or 3 system becosa inoperative during normal operation, sufficient equipment is still available to mitigate the consequences of a design basis accident.
3.11.4.3 Ventilation Systems Two redundant Safety Class 3, Seismic Category I air handling units are provided in the Reactor Auxiliary Building for the Control Room envelope.
The system design assures that proper ambient temperature is maintained at all times.
It is not considered cre'dible that simultaneous loss of the two units could occur.
Humidity is not controlled during accident conditions in most areas, except in the Control Room, and 100 percent humidity is assumed unless otherwise indicated.
3.11.4.4 Design Basis Temperatures The maximum temperatures considered in the sizing of ventilation and cooling systems serving safety-related systems were determined by quantitative analysis of the following factors:
a)
Maximum outdoor design temperatures for the geographical area of the plant (both wet-bulb and dry-bulb readings) per ASHRAE standards.
j b)
Maximum internal piping thermal loads, if applicable, for the particular area or room, using maximum operating temperatures for the pipe l
contents and maximum footage of active pipe for each mode of operation.
3.11.4-1 Amendment No. 16 O
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SHNPP FSAR c)
Maximum internal electrical load, assuming full lighting for the room and using, if applicable, the maximum control and equipment resistance losses for.each mode of operation.
d)
Maximum heat trahsfer for miscellaneous equipment surfaces.
e)
Maximum heat transfer from the surfaces of open pools and tanks, using the maximum operating temperature of the contents.
f)
Maximum heat transfer from the surfaces of the room, including walls,
2 floor and ceiling or roof.
3.11.4.5 Temperature conditions Inside containment and timin Steam Tunnel During/ Af ter a Design Basis Accident The temperature conditions inside the Containment or Main Steam Tunnel resulting f rom a design basis accident are a function of time. The following FSAR figures show these conditions for the various postulated line breaks considered:
4 Figure 3.11.4-1 DBA Temperature Profile Inside Containment 16 (combined LOCA/MSLB)
Figure 3.11.4-2 DBA Tumperature Profile Inside Containment (LOCA)
Figure 3.11.4-3 DBA Temperature Profile Inside Containment (MSLB)
Figure 3.11.4-4 DBA Temperature Profile Inside Main Steam Tunnel (MSLB)
I l
3.11.4-2 Amendment No. 16 l
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' SHNPP FSAR 6
' 3.11.5:
ESTDitTED CHElICAL AND RADIATION ENVIRO?HENT 4
3.11.5.1 Chemical Env ironment 1
Safety Related ' Systems.are designed to perform their safety-related 16!
functions in the temperature, pressure, and humidity conditions discussed in i
Section 3.11.1 and in Section 6.2.
In addition, components of ESF systems inside the Containment are designed to perform their safety-related functions in a long-term contact with boric acid and sodium hydroxide solutions,
recirculated through.the Safety Inj ection System (SIS) and Containment Spray System (CSS).
f The pH time history of the water. both in the cont _sinnent spray and in the i
i containment sump, as well as the boron concentrat'.on in the Reactor Coolant System, is discussed in Section 6.5.2..
i The containment atmosphere is maintained below 4 volume percent hydrogen consistent with the recommendations of Regulatory Guide 1.7.
The extent to F
which this and other recommendations of Regulatory Guide 1.7 are followed are discussed in FSAA Section 6.2.5.
i The boron inj ection portion of the Safety Inj ection System (SIS) is designed for 12 weight percent boric acid.
The CVCS, SIS, and CSS are designed. for both the maximum and long-tera boric acid concentration of 2000-2100 ppe at a 16i pH of 8.5 to 11.0.
l 3.11.5.2 Radiation Environment Safety related systems and components are desidned to perform their safety l 16 related functions af ter the normal 40 year operational exposure plus one accident exposure.
The normal operational exposure is based on the design
'i source terms presented in Section 11.1.nd Section 12.2.1.
Post accident 16:
system and component radiation exposures are dependent on equipment location. Source terms and other accident parameters are presented in Section 12.2.1 and in Chapter 15. For' safety related systems, normal j
operational exposure and post accident radiation exposures are listed in 16 :
1 Appendix 3.118.
The degree to which the recommendations of Regulatory Guide 1.4, " Assumptions i
Used for evaluating the Potential Radiological Consequences of a Loss of j
Coolant Accident for Pressurized Water Reactors," has been used in determining the source terms used in evaluating radiation exposure is detailed in Section 1.8.
l 16 The design radiation exposures are based on gamma and beta radiation. The j
ef fects of beta radiation are effectively attenuated by small amounts of j
shielding, such as conduits for cable and casings for equipment.
Organic sacerials which are located inside the Containment are identified in I
Section 6.1.2.
1 i
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3.11.5-1 Amendment No. 16 1
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.SENPP FSAR 3.11.6 PRESSURE ENVIRONMENT Normal operating pressure inside containment is 14.7 psia as indicated in FSAR Section 6.2.1.5.2, as well as in all other plant areas.
Design basis accident pressure conditions inside containment and the main steam tunnel are a function of time as shown in the following FSAR figuress FSAR Figure 3.11.6-1 Pressure Profile Inside Containment (Combines LOCA/MSLB) 6 FSAR Figure 3.11.6-2 Pressure Profile Inside Containment (MSLB)
FSAR Figure 3.11.6-3 Pressure Profile Inside Main Steam Tunnel (HSLB)
In all other plant areas, during a design basis accident inside containment or the main steam tunnel, the pressure remains at 14.7 psia.
3.11.6-1 Amendment No. 16
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f-SHNPP FSAR NOTES TO FIGURE 3.11.1-1 SHNPP QUALIFICATION PROGRAM MASTER LIST LEGEND The interpretation of each entry in the Master List is as follows:
1 1.
TAG NO.
Specific device alpha-numeric designation which identifies plant equipment.
2.
COMPONENT NAME A brief title or description of the item being qualified (Note:
components are the smallest breakdown of equipment types or categories for qualification purposes).
4 3.
FUNCTION &
This is a description of the service which the SERVICE component peeforas.
4 HANUFACTURER Specific vendor who manufactured the component, but not necessarily supplier of the component.
For example, limitorque manuf actures valve operators for a valve vendor who in turn supplies the entire valve-operator assembly to che utility.
5.
MODEL/ SERIAL NO.
Specific vendor designation for a family or group of like components.
6.
PLANT LOCATION The general plant area location where the component is located.
(See Table 3.11.1-1) 7.
LOCATION The coordinates locate the equipment within the "X",
"Y", "Z"
" plant location" and more importantly within the environmental zones.
(See Table 3.11.1-1) 8.
ENVIRONMENTAL Classification of each component's environment as CATEGORY:
H/M
" harsh" or " mild".
9.
FUNCTION CATEGORY A single entry is made herein for the applicable category listed in NUREG-0588, Appendix E, paragraph 2.
The categories are "A",
"B",
"C",
or.
"D".
10.
SAFETY FUNCTION Major plant safety functions indicative of safety function performed by the system in which the equipment is a part of.
Examples of safety functions are: Containment Isolation (CI), Emergency Reactor Shutdown (ERS), Reactor Core Cooling (RCC),
Containment Heat Removal (CHR), Core Reci. dual Heat Removal (CRHR), Prevention of Significant Release of Radioactive Material to the Environment (PRRM),
Supporting Systems (SS).
Amendment No. 16 Page 1 of 2
SiciPP FSAR NOTES TO FIGURE 3.11.1-1 (cont'd) 11.
REG l.97 An asterisk (*) is inserted whenever the component is a Category 1 or 2 post-accident monitoring instrument in accordance with Regulatory Guide 1.97.
16 12.
CES No.
The unique identification number identifying the Component Evaluation Sheet which sunuaarizes the required environmental conditions and provides other qualification summary information for the given component.
13.
REV. #
The revision number applicable to a Component Evaluation Sheet.
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i Amendment No.16 Page 2 of 2
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SENPP FSAR NOTES TO FIGURE 3.11.1-2 EXPLANATION AND LEGEND FOR COMPONENT EVALUATION SHEETS 1.
Terms f ound below " Equipment Description" designation are explained as follows:
TAG NO.
Specific device alpha-numeric designation which identifies plant equipment.
EQUIPMENT TYPE Designation of equipment into categories (e.g., sensors, motors) which correspond to NUREG-0588 Appendix E. Section 1.d categories.
COMPONENT A brief title or description of the item being qualified (Note:
components are the smallest breakdown of equipment types or categories for qualification purposes).
MANUFACTURER Specific vendor who manufactured the component, but not necessarily the supplier of the component.
For example, limitorque manuf actures valve operators for a valve vendor who in turn supplies the entire valve-operator assembly to the utility.
The vendor supplying the equipment of the utility if MAJOR SUPPLIER other than the manuf acturer of the equipment.
MODEL AND Specific vendor designation for a family or group of SERIAL NO.
like co^mponents.
16 FUNCTIONAL This is a description of the service which the DESCRIPTION component performs.
& SERVICE ACCUR SPEC This is either the requirement for accuracy used in Station Safety Analysis or the standard manufacturer's limits used in generic testing or instruments, whichever requires greater accuracy.
ACCUR DEMON This is a value which is equal to or better than the accur. spec entry. Value is for the long-term stable operation of instruments.
SPECIFICATIONS This entry is the equipment specification the equipment is designe.1 to meet.
PURCHASE ORDER This is the purchase oraer used during equipment NO.
procurement.
PLANT LOCATION The general plant area location where the component is located.
Amendment No. 16
SHNPP FSAR NOTES TO FIGURE 3.11.1-2 (cont'd)
COORDINATES The coordinates locate the equipment within the "X",
"Y', "Z"
" Plant Location" and more importantly within the erwironmental parameter zones.
4 INSTALLED An indication of installation status to minimize an YES/NO attempt to audit an installation when equipment is not installed.
i INSTAL. REF.
The source of data for installation status.
QUALIFICATION Entry (rarely ende) to indicate equipment need not ba
}
EED1PTION qualified by use of CES.
For example, a mechanical only device may be on the Master List and is not. to be qualified.
If this is so, entry of notes in the 1
reference section of CES is expected.
QUALIFICATION An indication of the emrironmental qualification STATUS status of the equipment.
a.
Qualified - Without Exception - This category is based upon the existing qualification 16 i
documentation demonstrating that the equipment will be capable of performing its intended safety function at any time during its qualified life, plus post-accident duration as required.
Total
{
compliance with the requirements has been fully documented.
b.
Qualified - Awaiting Confirmatory Data - This category is used when most of the qualification report review and analysis, to demonstrace qualification, has been completed; but some open 4
items, which are identified, must be resolved.
i In all cases, there is a high degree of confidence that the open items will be resolved satisf actorily, thus enabling this status to be i
upgraded to Qualified - Without ' Exception.
l l
c.
Qualified - For Interia Cperatior. - This category l
is used primarily when qualification testing has not been completed, but there is a high degree of confidence that the equipment can be qualified, thus permitting interim, operation.
In addition, the criteria in Enclosure 1 of Policy Issue SECY-82-51 issued 2/4/82 for j ustification for interia operation is used and documented.
Page 2 of 6 Amendment No. 16
SHNPP FSAR r
NOTES TO FIGURE 3.11.1-2 (cont'd) d.
Relocate Equipment - This category is selected when the equipment is not dersonstrated to be qualified for its initial installation location. This equipment must be relocated to a new location where qualification can be demonstrated for the new conditions.
(See e, f, g, h, below.)
e.
Shield Equipment - T is category is used when equipment is not demonstrated to be qualified f or its installed location and simple shielding (e.g., from beta radiation) can assure adequacy of qualification.
f.
Retest Equipment - This category is used when equipment is undergoing recesting to demonstrate qualification as required.
g.
Replace Equipment - This category is used when qualification cannot be demonstrated and it is prudent to replace the equipment with a suitable, qualified replacement.
f h.
Qualified - Awaiting liinor Analysis.
(See b above.)
16 1.
Deson. NUREG-0588C - This category is used for equipeant in the scope of the definition of Category "C" as stated in NUREG 0588, Appendix E.
- j. Requires Maj or Analysis - This category is used when there are significant concerns related to the qualification status of the equipment and a e4 or effort is required to qualify the equipment.
2.
Terms found below Environment Parameter Actual (Column 1) are described 2
below OPERABILITY Requirements for operation which may be in time Nore/ Test (hours, days, sonths, years), cycles, or both.
DBA a
TD1PERATURE Normal and DBA temperature conditions at the equipment location.
PRESSURE Normal (generally atmospheric) and D8A press.re conditions at the equipment location.
4 i
Page 3 of 6 Amendment No.16
~. _ _ _ _, _ _. - _ _ _ _
~,
SHNPP FSAR NOTES TO FIGURE 3.11.1-2 (cont 'd)
RELATIVE Ambient normal and DBA relative humidity conditions HUdIDITY at t!?e equipment location.
CHEMICALSPRAI Values for the chemical composition of the chemical spray (containment ' spray) utilized in a post-DBA 2
event in containment.
l R GAMMA GAMMA is the sua total of the 40 year normal plus A BETA applicable (1 yr, 1 ao, 1 day) DBA gamma dose. Beta D 8 SHIELD is the applicable (1 yr, 1 mo, I day) DBA Beta Dose.
S T.I.J.
- 8. Shield is the credit (10-100%) permitted, to reduce the Beta dose, due to enclosures, material coverings, and thicknesses. TID (Total Integrated Dose) is the sus total of CAMMA plus Beta (after sheilding) applicable to the equipment.
ACE-INST LIFE The goal or requirement for equipment life, usually (per 323 1974 DEF) 40 years.
SUBHERCED LEVEL Maximum plant elevation (Ft.) reached during flood conditions. Generally, equipment should be located above this level.
l i
3.
Data below Environment Parameter DEM. QUALIF. (column 2) are the actual values the equipment is qualified to which corresponds on a "one-to-one" basis with the actual parameter (column 1).
4.
Data below Documentation Actual (column 3) is the reference source 16 (generally FSAR Environmental Zone Maps, etc.) which identifies the l
requirements in column 1.
j 5.
Data below Documentation DEM. QUALIF (column 4) is the reference source l
(environmental qualification test reports, engineering analysis, etc.)
which substantiates the information in column 2.
6.
Data below Qualification Method (column 5) is the actual methodology used to demonstrate qualification. The most likely entry is " Combined Test and Supplementary Review" to indicate that the qualification method is a type test supplemented by analysis / review.
7.
Data entered below the H/M (column 6) is the indication if a zone's environmental parameter valaes are harsh or mild.
8.
Data entry below Outstanding Items (column 7) would be for significant items of concern which do not allow an item to be classified as qualified. Minor items just requiring confirmation will not be considered outstanding items.
l l
Amendment No. 16 Page 4 of 6
SKNPP FSAR NOTES TO FIGURE 3.11.1-2 (cont 'd) 9.
Data below the rightmost columns is as follows:
REPLACEMENT Special requirements to replace items not normally replaced during normal maintenance as a condition of qualification, if the equipment or component therein is not qualified to 40 years.
MAIMTENANCE As for replacement, only' conditions related to qualification are entered.
i SUBCOMPONENTS This is an entry that may be used by the utility to l
help locate components enveloped within a larger qualification package.
For example, relays may be included here.
SAFETY FUNCTION Major plant safety functions indicative of the safety function performed by the system in which the equipment is a part of.
Examples of safety functions are Containment Isolation (CI), Emergency Reactor Shutdown (ERS), Reactor Core Cooling (RCC),
Containment Heat Ramoval (CHR), Core Residual Heat Removal (CRHR), Prevention of Significant Release of Radioactive Material to the Environment to the Environment (PRRM), Supporting Systems (SS).
10.
Data Entered in the Parameter - Suppl. Review Box (lower mid-left side) 16 PARAMETER A list of all parameters (operability through submergence) being reviewed.
SUPPL. REVIEW Identifies the Supplementary Review sheets, included in the documentation packages, for each parameter (operability, temperature, pressure, relative humidity, chemical spray, radiation, aging, submergence) which justifies qualification to each parameter.
NUREG-0588 A single entry is made herein of the applicable APPENDIX E NUREG-0588, Appendix E category listed in CATEGORY paragraph 2.
The categories are "A", "B",
"C", or "D".
- 11. Data entered in the lower middle box is as follows:
FOR PUNCHLIST ITEMS References the documentation package number (usually same) where Punchlist (EQ outstanding) items may be found.
1 REFERENCES References Qualification Information sources applicable to the equipment being qualified.
Amendment No. 16 Page 5 of 6
SHNPP FSAR NOTES TO FIGURE 3.11.1-2 (cont 'd) 12.
Data entered in the lower right hand box is as follows:
QUALIFICATION References the documentation package number (usually SIGN OFF the same) where the names and signatures of individuals preparing / checking the documentation package may be found.
CES #
The unique four-digit number which identifies the individual Component Evaluation Sheets.
16 REVISION #
The revision number applicable to a component Evaluation Sheet.
DATE The date the applicable revision to the Component Evaluation Sheet was made.
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AMENOMENT NO.19 4
SHEAMON HARRIS FIGURE NUCLEAR POWER PLANT DBA TEMPERATURE PROFILE INSIDE MAIN STEAM Caroline TUNNEL (MSLBI 3.11.4 4 Power & Light Company FOR ENVIRONMENTAL QUALIFICATIOfl FINAL SAFETY ANALYSIS REPORT
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C SH'iPP FSAR l
APPENDlX 3.11A NUREG-0588 COMPARISON FOR 16 SHEARON HARRIS NUCLEAR POWER PLANT 3.11A-1 Amendment No. 16
e SHNPP FSAR
,i t
APPf N0lx 3.llA NUREG-0588 COMPARISON CATEGORY ll Applicante to Equipment Quallflod in Accordance with IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Program 1
ESTAB.lSWENT OF THE QUALIFICATION PARAMETERS FOR DESIGN SASIS EVENTS 1.1 Temperature and Pressure Conditions inside Contaltunent - Loss-of-Coolant Accident (LOCA (1) The time-dependent temperature and 1.1 (1) Time dependent temperature and pressure, established for the design pressure LOCA profiles are used.
of the containment structure and Ref er to figures in FSAR Sections found acceptacle by the staf f, may 3.11.4 and 3.11.6 and Appendix 3.118 be used f or environmental quellfication of equipment.
(2) Acceptable methods for calculating (2) Mass and energy release rates are and establishing the containment consistent with those sunnerized in pressure and temperature envelopes NtREG-0588 Appendix A.
Ref er to FSAR to which equipment should be Section 6.2.1.3 for details.
16 quellfled are summarlzed below.
Acceptable methods for calculating mass and energy release rates are summarized in Appendix A.
Pressurized Water Reactors (PWRs)
Ory Containment - Calculate LOCA CONTEWT-LT26 is used in calculating contain:nont environment using the post-LOCA containment environment.
CONTEMPT-LT or equivalent Industry Ref er to FSAR Section 6.2. l. f.3.2 codes. Additional guidance is provided in Standard Review Plan (SF)
Section 6.2.1.1.A, NUREG-75/087. The assumption of partial reveporization will be allowed. Other assumptions that reduce the temperature response of the containment will be evaluated on a-case-by-case basis.
Ice Condenser Contaltunent - Calculate SHNPP does not have an Ice condenser LOCA containment environment using containment; therefore, this is not LOTIC or equivalent industry codes.
appl i cable.
Additional guidance is provided in SW Section 6.2.1. l.8, NtREG-75/087.
- 3. IIA-2 Amendment No. 16
h SEGP FSAR APPEN0lX 3.1IA NUREG-0588 COMPARISON CATEGORY Ii Applicable to Equipment Quallflod in Accordance wi th IEEE Std. 323-1971 Shearon Harris Nuclear Pcasor Plant Progran Bolling Water Reactors (BWRs)
Merk I, 11, and t il Contalrunent -
SHNPP is a PWR; therefore, this is not Calculate LOCA environment using applicable.
methods of GESSAR Appendix 38 or equivalent industry codes. Additional guidance is provided in SW Section 6.2.1.1.C, NUREG-75/087 (3) In llou of using the plant-specific (3) SHNPP is a dry containment PWR; containment tosporature and pressure therefore, this is not applicaole.
design profiles for BWR and Ice condenser types of plants, the generic envelope shown in Appendix C may be used for quellfication testing.
(4) The test profiles included in (4) Plant-speci fic containment temperature 16 Appendix A to lEEE Std. 323-1974 and pressure profiles are used. Refer should not be considered an to f igures Ir. FSAR Sections 3.11.4 and acceptable alternative in lieu of 3.11.6 and Appendix 3.118.
using plant-specific contaltenent tegerature and pressure design, profiles unless plant-specific l
analysis is provided to verify the adequacy of those profiles.
1.2 Temperature and Pressure Conditions inside Containment - Main Steam Line Break (MSLS)
(1) more quellfication has not been 1.2 (1) A plant-specific analysis consistent completed, the environmental with the requirements of NtREG-0588, parameters used for equipment utilizing CONTEWT-LT as described in quellfication should be calculated FSAR Section 6.2.1.3.3, has been used using a plant-specific model based to determine the temperature and on the statf-epproved assumtions pressure conditions inside containment discussed in Itan I of Appendix 8.
for a MSI.B.
(2) Other models that are acceptable for (2) See 1.2 (1) above, caloalating containment parameters are listed in Section 1.1(2).
3.11A-3 Amendment No. 16
t SHNPP FSAR APPENDIX 3.11A NUREG-0580 COMPMiSON CATEGORY II Applicable to Equipment Quallflod in Accordance with IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Progran (3) In lieu of using the plant-specific (3) SHNPP is a dry containment PWR; containment teaperature and pressure therefore, this is not applicable.
. design profiles for SWR and Ice See 1.1 (1) above.
condenser plants, the generic envelope shamn in Appendix C may be used.
(4) The test profiles included in (4) Plant-specific containment temperature Appendix A to IEEE Std. 323-1974 and pressure design profiles are used.
should not be considered an Ref er to 1.1 (l) above.
acceptable alternative in lieu of using plant-specific contaltunent tosperature and pressure design profiles unless plant-specific analysis is provided to verify the adequacy of those profiles.
(5) mero qualification has been cosoleted (5) in general, combined MSLS/LOCA but only LOCA conditions were considered, profiles are utilized for time-then it must be demonstrated that the dependent temperatures and pressures LOCA qualification conditions exceed or (Refer to FSAR Figures 3.11.4-1 and l
are equivalent to the maximum calculated 3.11.6-1, respectively) regardless of g
MSLB conditions. The following technique less stringent quellfication require-Is acceptable monts; heuever, in those cases wnere the test condition profile does not (a) Calculate the peak teasorature envelope the applicable Shearon Harris from an MSla using a model based profile, the folloeing technique is on the statf's approved used:
assumptions discussed in item 1 of Appendix B.
, Additional justification (e.g.,
component thermal lag analysis)
(b) Show that the peak surface is provided to demonstrate that temperature of the component to be the equipment can maintain its quellflod does not exceed the LT.A required functional operability or quellfication temperature by the requalification testing is per-method discussed in item 2 of I
Appendix B.
formed with appropriate mergins, or qua.llflod physical protection may (c) If the calculated surf ace temperature exceeds the quellfl=
be provided to assure that the cation temperature, the staf f equipment experlences only the requires that (I) additional condiflons for which it is justification be provided to q uel l f led.
demonstrate that the equipment 3.11A-4 Amendment No. 16 1
~..
1 !
SHNPP FSAR APPENQlX 3.llA NUREG-0588 COMPARISON CATEGORY l1 Applicable to Equipment Qualified in Accordance with IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Program can maintain its required functional operability If its surf ace temperature reaches the calculated value or (11) requalification testing be performed wIth appeopeIate margins, or (III) quellfled physical protection be provided to assure that the surface temperature will not exceed the actual quellfication temperature.
1.3 Effects of Chenleal Spray The of fects of caustic spray should be 1.3 The most severe containment spray addressed for the equipment quellfication.
environment (boron concentration and pH The concentration of caustics used for level) is used for environmental quellfleetion should be equivalent to or qualification. The actual (calculated) n,ye severe than those used in the plant spray environment bounds any postulated i
gg containment spray system, if the chemical single failure.
composition of the caustic spray can be af fected by equipment malfunctions, the most severe caustic spray environment that results from a single failure in the spray systen should be asstaned. See SRP Section 6.5.2 (NUREG-75/087), paragraph 11, item (e) for caustic spray solution guidelines 1.4 Radiation Conditions inside and Outside Containment The radiation environment for 1.4 For quellfication purposes, re&ctions quellfication of equipment should be based in air dose due to spray washout and on the normally expected radiation platecut are not used in calculating environment over the equipment quellflod the post-accident radiation environments.
life, plus that associated with the most Therefore, radiation doses used in severe design besls accident (08A) during quellfication are maximum total Integrated or folicwing which that equipment must dose calculated over the equipment remain functional. It should be assumed quellflod life, plus that associated with that the GBA related environmental condi-the most severe design basis accident.
tions occur at the end of the equipment quellflod life.
l 3.11A-5 Amendment No. 16 E
SHNPP FSAR APPEN0lX 3.11A NUREG-0588 COMPARISON CATEGORY II Applicable to Equipment Quallflod in Accordance with IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Proqram The sample calaslations in Appendix D and the following positions provide an acceptable approach for estabilshing radiation Ilmits for quellfication.
Additional radiation mergins identitled in Section 6.3.1.5 of IEEE Std. 323-1974 for quellfication type testing are not required If these methods are used.
(1) The source term to be used in (1) The source term used in all cases in determining the radiation environment determining the radiation environment associated with the design basis LOCA is that 100 percent of tne noble should be taken as an Instantaneous gases, 50 percent of the lodines, and release f rom the f uel to the atmosphere I percent of the remaining fission of 100 percent of the noble gases, products are released Instantaneously 50 percent of the lodines, and 1 per-from the f uel to the containment cent of the remaining fission products.
atmosphere.
Ib For all other non-LOCA design basis accident conditions, a source term Involving an Instantaneous release from the fuel to the atmosphere of 10 percent of the noble gases (except Kr-85 for which a release of 30 per-cent should be assumed) and 10 percent of the lodines is acceptable.
(2) The calculation of the radiation (2) Time-dependent transport of released environment associated with design fission products within various basis accidents should take lato regions of containment and auxillary account the time-dependent transport structures is assumed in the calcula-of released fission proects within tion of the radiation environment j
various regions of containment and associated with design basis j
auulliary structures.
accidents.
(3) The Initial distribution of activity (3) The Initial distribution of activity within the containment should be within the containment is based on a based on a mechanistically rational mechanistically rational assunction as assump tion. Hence, for compartmented described in FSAR Section 12.2.
Since I
containments, such as in a IWR, a the internal structures of the large portion of the source should oe contsinnent were designed to provide assumed to be initially contained in vertical compartments around eacn of 3.11A-6 Amendment No. 16
l SCIPP FSAR APPEN0lx 3.11A NUREG-0588 COMPARISON CATEGORY ll Applicable to Equipment Quallfled in Accordance wi th IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Program the drywell. The asseption of the steen generators and the reactor uniform distribution of activity vessel and since the Containment Spray throughout the containment at time and/or the containment ventilation and zero is not appropelate.
filtration systems provide mixing for the containment atmosphere,
.3 determination was made to assume a uniform distribution of activity througnout the containment.
(4) Effects of ESF systems, such as (4) Credit for the removal of aleborne contalrunent sprays and containment activity by ESF systens has Deen ventilation and filtration systems, taken. In addition, the distribution which act to remove airborne activity of activity is taken into account as and redistribute activity within described in (3) above and by (5) contelnment, should be calculated
- belos, using the same assumptions used in the calculation of of f site dose.
See SRP Section 15.6.5 (NUREG-75/087) and the related sections ref erenced in the Appendices to that section.
16 (5) Natural deposition (l.c.s plate-out)
(5) The SHNPP model assumes zero removal of alroorno activity should be for plate-out; however, the contain-l determined using a mechanistic model mont sumo source terms are developed and best estimates for the model by assuming dilution of 50 percent of parameters. The asstanotion of 50 the core Inventory of halogens and percent Instantaneous plate-out of 1 percent of other nuclides with the the lodine released f rom the core combined volumes of tne Reactor should not be made. Removal of Coolant, Accumulators, Boron injection l
lodine f rom surf aces by steen Surge Tanks and the Refueling Water condensate flee of washof f by the Storage Tank. The resulting Initial containment spray may be assumed if siano (diluted coolant) activity is such effects can be justified and given on FSAR Table 12.2.1-26.
quantitled by analysis or superiment.
(6) For unshielded equipment located in (6) The ganse dose and dose rate used in the containment, the gamme dose and dose quellfication for equipment located l
rate should be equal to the dose and dose insido containment is calculated for f
rate at the centerpoint of the con-various zones utilizing distance and tainment plus the contribution from shielding credits. Refer to FSAR location dependent sources such as Appendix 3.118 for applicaDie doses in the sump water and plate-out, unless various zones, it can be shown by analyres that location and shielding of the equip-ment reduces the dose and dose rate.
Amendment No. 16
SmiPP FSAR i'
APPEN0lx 3. IIA NUREG-0588 COMPARISON CATEGORY II Applicable to Equipment Quellfled in Accordance with IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Program I
i (7) For unshielded equipment, the beta (7) For unshleided equipment, the beta l
doses at the surface of the equip-dose is calculated at the most ment should be the sum of the alrborne conservative locatice for all and plate-out sources. The aleborno appropriate contributors of beta doses beta dose should be taken as the beta including alrturne, and suspended dose calculated for a point at the sources.
containment center.
(8) Shielded components need be quellfled (8) Components are quellfled, by exposure only to the gamme radiation levels to gamme radiation only, to the total required, provided an analysis or (numerical) Integrated dose required.
test shows that the sensitive portions The total dose includes gamme and beta of the component or equipment are not radiation and appropriate shleiding exposed to beta radiation or that the credits with adequate justification.
ef f ects of beta radiation heating and lonization have no deleterious effects on component performance.
(9) Cables arranged in cable trays in the (9) See 1.4 (8) above. in addition, the containment should be assumed to be beta dose at the equipment may be egosed to half the beta radiation dose reduced by equipment covering meterial calculated for a polnf at the center of (i.e., cable jackets, boxes, etc.) and gg the containment plus the gamme ray dose thickness as permitted by Section calculated in accordance with Section 4.1.2 of l&E Bulletin 79-018 In 1.4(6).
This reduction in beta dose is these cases, justification is a
allowed because of the localittd
- provided, shleiding by other cables plus the cable tray itself.
j (10) Paints and coatings should be asstaned (10) Paints and coatings are assumed to to be exposed to both beta and gamme be exposed to both beta and gamme rays I
rays in assessing their resistance to in assessing their resistance to radiation. Plate-out activity should radiation. Plate-out activity is be assumed.to remain on the equl.wnt assisned to remain on the equipment.
surface unless the effects of the l
removal mechanisms, such as spray wash-off or steam condensate flow, can be justified and quer.tifled by analysis or experiment.
(11) Components of the emergency core cool-(11) Cosponents of the Residual Heat Ing system (ECCS) located outside con-Removal System and the Containment taltenent (e.g., pumps, valves, seals Spray System located outside con-and electrical equipment) should be tainment are quellflod to withstand quellflod to withstand the radiation trae radiation equivalent to that 3.11A-8 Amendment No. 16
e SENPP FSAR l
APPENOlX 3.11A NUREG-0588 COMPARISON f
CATEGORY 11 Applicable to Equipment Quallflod in Accordance with IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Program equivalent to that penetrating the penetrating the containment plus the containment, plus the exposure from exposure from the suas fluid. See 1.4 the sump fluid using assumptions con-(5) above.
sistent with the requirements stated in Appendix K to 10 CFR Part 50.
(12) Equipment that may be exposed to (12) Equipment exposed to radiation doses 4
radiation doses below 10 reds should at any level :re not considered to be not be considered to be exempt f rom exempt from radiation quellfication, radiation quellfication, unless analy-unless analysis supported by test sis supported by test data is provided data and/or operating experience is to vorify that these 1oveis wIIl not prowIdod to verify that these ieveIs degrade the operability of the equip-will not degrade the operability of mont below acceptable values.
the xquipment below acceptable values. Otherwise equipment is quallfled to their required doses.
See 1.4 (8) and (9) above.
l 16 (13) The statt will accept a given compo-(In The applicants' environmental quellfl-neot to be qu'tilflod provided it can cation program complies with the be shown that the component has been guidellnes previously described in quellfled to Integrated beta and ganme Itan 1.1 (1) through (12).
doses which are equal to or higher than those levels resulting f rom an analy-sls slallar in nature and scope to that included in Appendix 0 (which uses the source term given in item (t: e vel, and that the component incorporates appropriate factors pertinent to the plant design and operating charac-teristics, as given in these general I
guidelines.
(14) when a conservative analysis has not (14) A conservative analysis has been been provided by the applicant for provided by the applicant in the staf f review, the staf f will use the _
FSAR sections referenced above.
radiation environment guidelines con-tained in Appendix D, suitably corrected for the differences in reactor power level, type, containment size, and other aggropriate factors.
3.11A-9 Amendment No. 16 l
i SiciPP FSAR APPEN0lX 3. IIA NUREG-0588 COMPARISON CATEGORY iI Applicaole to Equipment Quellffed in Accordance alth IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Program
- 1. 5 Environmental Conditions for Outside Contairment (1) Equipment located outside containment I.5 (I) Equipment located outside contain-that could be subjected to high-ment is quellflod to operate following energy pipe breaks shoulo be quell-a high-energy pipe break as described fled to the conditions resulting from in FSAR Section 3.6 and Appendix 3.6A.
the accident for the duration required.
Only that equipment necessary to miti-The techniques to calculate the environ-gate or monitor the consequences of monte: parameters described in Sections the postulated Hela accident is quell-1.1 through 1.4 (Cate, gory 11) above fled to the respective HELS shoul d be appl i ed.
conditions.
(2) Equipmen. located in general plant (2) Equipment located in general plant ar'as outside containment where equlpa areas outside Containment are qual-ment is not subjected to a design Ifled for the maximum normal and basis accident environment should be abnormal range of environmental con-quellflod to the normal and abnormal ditions postulated in the equipment range of environmental conditions area. Ref er to the figures In FSAR postulated to occur at the equipment Appendix 3.118 for applicaDie location.
environmental parameters in these general plant areas.
(3) Equipmeni not served by Class IE (3) Equipment served by Class IE environmental support systems, or environmental support systems that served by, Class IE support systems that' may be secured during plant operation may be secured during plant operation or shutdown will be quellfied for the or shutdown, should be quellfled to the limiting Anticipated Operation limiting environmental conditions that Occurrence (A00) environmental condi-are postulated for the location, assue-
- lons assuming loss of the environ-Ing a loss of the environmental support montal support systen, but such system; or, there may be designs where conditions are considered to be within l
a loss of the environmental support the A00 temperature envelope of mild system may expose some equipment to environment.
environments that exceed the quellflod limits. For these designs, appropriate monitoring devices should be provided to alert the operator that abnormel conditions exist and to permit an assessment of the conditions that j
occurred in order to determine If corrective action, such as replacing any af f ected equipment, is warranted.
3.11A-10 Amendment No. 16
\\
~1 SHNPP FSAR l
1 APPEN0lx 3. IIA NUREG-0588 COMPARISON CATEGORY ll Applicable to Equipment Quallfled in Accordance wi th IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Progren 2.
QUALIFICATION METH005 2.1 Selection of Methods (1) Quellfication methods should conform 2.1 (1) Qualification methods conform to the to the requirements defined in IEEE guidelines of IEEE Std. 323-1971; Std. 323-1971 however, much of the equipment has been upgraded to meet fdtC Regulatory Guide 1.89 Revision 0 and its adopted standard IEEE Std. 323-1974 as described in FSAR Section 1.8.
Refer to FSAR Tables 3.11.0-l and 3.11.0-2 for quellfled equipment which has been upgraded.
(2) The choice of the methods selected Is (2) In general equipment located in a largely a metter of technical judgment harsh environment is quellflod for the and availability of Information that time required by type test in an 16 supports the conclusions reached.
accident test environment. Su ppl e-Experience has shown that quellfication mentary review and analysis is of equipment subjected to an accident necessary to demonstrate that the test environment without test data is not environmental conditions exceed or are adequate to demonstrate functional equivalent to the applicable Shearon operability, in general, the statt Harris conditions. Functional opera-will not accept analysis in llou of bility is required during qualifica-test data unless (a) testing of the tion testing.
cogonent is impractical due to size limitations, and (b) partial type test data is provided to support the ana-lytical assug tlons and conclusions reached.
(3) The environmental qualification of (3) The environment qualification of equipment exposed to CBA environ-ionts equipment Ic,cated in a harsh environ-should conform to the following post-ment conforms to the following:
tions. The bases should be provided j
f or the time Interval required for operability of this equipment. The i
operability and f ailure criterla should be specified and the safety margins defined.
(a) Equipment that must f unction in (a) Equipment that must function in order to mitigate any accident order to mitigate or monitor any should be quellflod by test to accident is quellflod as stated l
3.11A-11 Amendment No. 16
48 '
SHNPP FSAR APPEN0lx 3.11A NUREG-0588 COMPARISON CATEGORY Ii Applicable to Equipment Quallfled in Accordance wi th IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Program i
demonstrate its operability for in 2.1 (2) above, to demonstrate the time required in the environ-operability for the time mental conditions resulting from required.
that accident.
(b) Any equipment (safety-related (Is) Non-safety related equipment in or non-saf ety-related) that this category has been upgraded need not f unction in order to to Class 1E status. Safety-altigate any accident, but that related equipment is quellflod must not f all In a manner detrl-as described in 2.1 (2) above.
mental to plant safety should be quellflod by test to demonstrate its capability to withstand any accident environment for the time during which it asst not fall.
(c) Eculpnent that need not function (c) This equipment is quellflod for in order to mitigate any accident a mild environment as described and whose f ailure in any mode in in 10CFRSO.49 The applicant any accident environment is not complies with this requirement detrimental to plant safety need with respect to saf ety-related 16 on ly be quel l f l od f or i ts non-equipment. (NtftEG-0588 Is only accident service environment.
appilcable to safety-related equipment.)
Althougi actual type testing is preferred, other methods when justifled may be found accep-table. The bases should be provided for concluding that such equipeent is not required to function in order to mitl=
gate any accident, and that its failure in any mode in any acci-dont environment is not detrl-I mental to plant safety.
l (4) For envircamental quellfication of (4) When the environment f rom such an I
equipment subject to events other than event (e.g., loss of offsite power) a DBA, which result in abnonnel is enveloped by the environment from environmental conditions, actual type anticipated operational occurrences testing is preierred. However, analy-rather than significant design basis sls or operating history, or any event changes, the area is defined as appl icable combination thereof, a mild environment area; therefore, 3.11A-12 Amendment No. 16
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SICPP FSAR APPEN0lX 3. I1A NUREG-0588 COMPARISON CATEGORY 11 Applicable to Equipment Quallf f ed in Accordance wi th IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Program coupled with partial type test data the equipment is quellflod under mild may be found acceptable, subject tc environmental condiflons.
the applicability and detail of Infor-motion provided.
1
- 2. 2 QJalification er Test (1) The f ailure criterla should be esta-
- 2. 2 (1) in lieu of f ailure criteria, the bilshed prior to testing.
Appilcant has insured that the quall-fications by test include an acceptance criteria. Completed testing which did not include a specific acceptance criteria are analyzed or verlflod acceptable for their apollcation.
(2) Test results should demonstrate that (2) Refer to Section 3 for details on the equipment can perform its requirsJ
- mergin, function for all service conditions postulated (with margin) during its 16 in,,,,,,,
l (3) The Items described in Section 5'.2 of (3) SHNPP utilizes these guidelines for IEEE Std. 323-1971 supplemented by establishing test procedures. In items (4) through (12) below consti-addition, equipment upgraded to the tute acceptable guidelines for esta-1974 standard utilizes the guidelines bilshing test procedures.
of Section 6.3 of IEEE Std. 323-1974 as applicable supplemented by itents (4) through (12) beloe.
(4) When estabilshing the simulated (4) SHNPP utillzes a simulated combined environeental profile for quellfying MSLS/LOCA environmental profile for equipment located inside containment, equipment inside containment as it is preferred that a singse pectile shown on FSAR Figures 3.11.4-1 and be used that envelops the environ-3.11.6-1 The preferred method of mental conditions resulting from any quellfication is to assure that this design basis event during any mode profile is enveloped by the environ-of plant operation (e.g., a profile mental test profile to which the that envelops the conditions pro-equipmeint Is qualltled.
duced by the main stomallne break and loss-of-coolant accidents).
(5) Equipment should be located above (5) in general, equipment is located above flood level or protected against the maximum flood level. Equi pmen t submergence by locating the equip-required to be located below the max-3.11A-13 Amendment No. 16 w
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SENPP FSAR APPENDlX 3.I1A f
NUREG-0588 COMPARISON CATEGORY ll Appilcable to Equipment Quallfled in Accordance with lEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Program ment in quellflod watertight en-laum flood level is quellfled to closures. Where equipment is operate in a submerged condition or located in watertight enclosures, justification is provided to demon-quellfication try test or analysis strate that the equipment can perform should be used to demonstrate the its safety f unction for the duration adequacy of such protection. Where required before being submerged and equipuent could be submerged, It subsequent f allure wIi1 not af f oct the should be Identitled and demonstrated accomplishment of safety function.
to be quellflod by test for the duration required.
(6) The temperature to which equipment (6) The temperature to which equipment is is quellfled, when exposed to the quellfled is monitored throughout the simulated accident evirotunent, should test to assure that it was exposed to be defined lyr thermocouple reading on the bulk temperature equivalent to or or as close as practical to the sur-more severe than that temperature f ace of the cogonent being quellfled.
assumed in the bounding envelope if there were no thermocouples located dorlwed f rom the accident analysis.
near the equipment during the tests, in some cases, this monitoring is heat transf er analysis should be used based on using the steam tables and to determine the temperature at the the measured steam pressure to obtain g
component. (Acceptable heat transfer the saturated steam temperature, analysis methods are provided in Appendix B.)
(7) Performance characteristics of equip-(7) Equipment performance characteristics ment should be verifled before, after, are monitored before, during, and and periodically during testing after testing. The degree of equip-throughout its range of required ment monitoring (i.e., periodic or operability.
continuous) is based on equipment function, failure modes, and practicality of testing.
(8) Caustic spray should be incorporated (8) During simulated event testing, a during simulated event testing at the caustic spray is used. Spray system maximum pressure and at the tempera-actuation is delayed so as to simulate fure conditions that would occur when the required conditions as closely as the onsite spray systems actuate.
possible.
(9) The operability status of equipment (9) See 2.2 (7) above.
should be monitored continuously during testing. For long-term testing, hor-ever, monitoring at discrete Intervals should be justlfled I f used.
I 3.11A-14 Amendment No. 16 l
6 SHNPP FSAR APPEN0lX 3.11A Nt. REG-0588 COMPARISON CATEGORY l1 Applicable to Equipment Quallflod in Accordance with IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Program (10) Expected extrenes in power supply (10) During simulated event environmental voltage range and frequency should be test application of voltage / frequency applied during simulated event extremos may not be fossible. Post environmental testing.
test is the point at which extremos of voltage /f requency are considered.
Voltage / frequency tolerance is typically enveloped by industry standards which is the design constraint f or the design of the power distribution syst6m as described in FSAR Section 8.
Design optimization l
ls vorifled for voitage, frequency, etc. This ensures the adequacy of equipment and distribution system.
(11) Dust environments should be addressed (11) Equipment susceptability to dust is 16 when establishing quellfication considered in the plant maintenance service conditions.
procedures or by the use of protective covers.
J (12) Cobalt-60 is an acceptable gamme (12) Cobalt-60 or an equivalent source radiation source for environmental is used.
quellfication.
2.3 Test Secuence (1) Justification of the adequacy of the 2.3 (1) Justification for the test sequence is test sequence selected should be provided. In addition, the test provided.
environmental conditions are reviewed to assure that they simulate as close I
as practicable the postulated environment.
(2) The test should slamalate as closely as (2) Environmental service conditions practicable the postulated environment.
expected to occur are enveloped by the test simulation environment and/or by supplementing analysis and review.
(3) The test procedures should conform to (3) See 2.2 (3) above.
the guidelines descelted in Section 5 of IEEE Std. 323-1971 l
3.llA-15 Amendment No. 16
a SHNPP FSAR APPEN0lX 3.llA NUREG-0588 COMPARISON CATEGORY 11 Applicable to Equipment Quallflod in Accordance with IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Program (4) The staf f considers that, for vital (4) In general, equipment which sust electrical equipment such as penetre-perform a safety function in a harsh trations, connectors, cables, valves environment is quellflod by subjecting and motors, and transmitters located
" sample" equipment to the test condl-Inside containment or exposed to tions. Where this is impractical hostile steem environments outside (e.g., due to size limitations) containment, separate ef fects testing justification is provided for separate for the most part is not an acceptable effects testing. Sequential testing quellfication method. The testing of is the standard method of test with such equipment should be conducted in a exceptions documented and justified.
manner that subjects the same piece of equipment to radiation and the hostile steam environment sequentially.
2.4 Other Qualification Methods Quallfication by analysis or operating 2.4 in general, supplementary review and experience implemented, as described in analysis is used to evaluate test data to IEEE Std. 323-1971 and other ancillary demonstrate qualification. Testing is standards, may be found acceptable. The generally employed to quality the adequacy of these methods will be equipment.
ig evaluated on the basis of the quality and detall of the Information submitted in support of the assumptions made and the specific f unction and location of the equipment. These methods are most suitable for equipment where testing is precluded by physical size of the equip-ment being quellfled. It is required that when these methods are employed some partial type tests on vital components of the equipment be provided in support of these methods.
3.
MARGINS (1) Quantifled mergins should be applied 3.
(1) The applicant has utilized the NRC to the design permeters discussed in staff acceptable approach of Section I to assure that the postula-demonstrating that the temperature, ted accident conditions have been pressure, and radiation conditions enveloped during testing. These are derived using the NtREG-0588 t
3.llA-16 Amendment No. 16
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l SFCIPP FSAR r
APPEN0lx 3. llA NUREG-0588 COMPARISON CATEGORY li Applicante to Equipment Quallflod in Accordance with IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Program mergIns should be applied in addition methodolog; which is suf ficiently to any margins (conservatism) applied conservative such that margin need during the derivation of the specified account only for inaccuracles in the plant parameters.
test equipment. See Resolution of Comment 70 in NtREG-0588, Rev. 1 (2) The mergins provided in the design (2) See 3 (1) above.
will be evaluated on a case-by-case basis. Factors that should be con-sidered in quantifying mergins are (a) the environmental stress levels induced during testing, (b) the duration of the stress, (c) the number of itses tested and the numoer of tests performed in the hostile environment, (d) the performance characteristics of the equipment while subjected to the environmental stresses, and (e) the 16 specified function of the equipment.
l (3) teien the quellfication envelope in (3) Appendix C ls appilable to BWR and ice Appendix C is used, the only required condenser containments. SHNPP is a margins are those accounting for the dry containment PWR; therefore, quell-Inaccuracles in the test equlpment.
fication to Appendix C is not l
Suf ficient consvvetism has already appl i cable.
been included to account for uncertaintles such as production errors and errors associated with defining satisf actory performance (e.g., when only a small number of units are tested).
{
(4) Some equipment may be required by the (4) Equipment procured f or short-term design to only perform its safety operation has been reviewed to assure f unction within a short time period that it is quellflod for the time into the event (i.e., within seconds required to operate with additional or minutes), and, once Its f unction is margin.
complete, subsequent failures are shown not to be detrimental to plant safety.
Other equipment may not be required to perform a safety f unction but naast not f all within a short time period into the 3.11A-17 Amendment No. 16 w,,, - y
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e SHNPP FSAR in APPEN0lX 3. I1A NUREG-0588 CCNPARISON CATEGORY I1 Applicable to Equipment Quallflod in Accordance with IEEE Std. 323-1971 Shearon Harels Nuclear Power Plant Program event, and subsequent f ailures are also shown not to be detrimental to plant safety. Equipment in these categories is required to remain functional in the accident environment for a period of at least one hour in excess of the time assumed in the accident analysis. For all other equipment (e.g., post-accident monitoring, recombiners, etc.), the 10 percent time mergin Identitled in Section 6.3.1.5 of IEEE Std. 323-1974 may be used.
4 AGING (1) Qualification programs that are 4
(1) The ef f ects of aging are considered committed to conform to the require-for the qualification programs that ments of IEEE Std. 382-1972 (for are committed to conform to the valve operators) and IEEE Std.
requirements of IEEE Std. 382-1972 334-1971 (for motors) should consider (for valve operators) and IEEE Std. 16 the effects of aging. For this 334-1971 (for motors),
equipment, aging of fects, regardless of its location in the plant, should be considered and included in the qualification program.
(2) For other equipment, the quellfication (2) Aging effects on all Class IE equip-programs should address aging only to ment located in a harsh environment the extent that equipment that is com=
are considered. Specific maintenance /
posed, in part, of meterials susceptible, surveillance requirements are to aging ef f ects should be Identitled, referenced in the Equipment Qualifica-and a schedule for periodically replacing tion Documentation Package. Where it the equipment and/or materials should be has been determined that quellflod established. During Individual case equipment or sub-components must be reviews, the staf f will require that replaced / maintained, due to aging the effects of aging be accounted for effects, it will be so noted on the on selected equipment i t operating Component Evaluation sheet for the experience or testing Indicates that the af f ected equipment. This information equipment may exhibit deleterlous aging will be incorporated into the appli-mechanisme.
cent's maintenance / surveillance l
progran.
3.11A-18 Amendment No. 16
1 I
4 SHNPP FSAR APPENDIX 3.I1A NUREG-0588 COMPARISON CATEGORY II Applicable to Equipment Quallflod in Accordance with IEEE Ctd. 323-1971 Shearon Harris Nuclear Power Plant Program THE FOLLGilNG CATuiGtY I PORTIONS OF SECTION 4 ARE APPLICABLE FOR THE QUALIFICATION PROIRAfeS THAT ARE CGeHTTED TO COWGtt TO THE REQUIRBeENTS OF BEEE STD. 382-1972 (FOR VALVE OPERATOR $) A80 IEEE STD. 334-1971 (FOR NOTORS).
4.1 Aging ef fects on all equipment, 4.1 The ef fects of aging are considered for all regardless of its location in the safety related equipment located in a harsh plant, should be considered and environment.
Included in the quellfication program.
4.2 The degrading influences discussed in 4.2 The degrading influences discussed in Sections 6.3.3, 6.3.4, and 6.3.5 of Sections 6.3.3, 6.3.4, and 6.3.5 of IEE IEEE Std. 323-1974 and the electrl-Std. 323-1974 and the electrical and cal and mechanical stresses associated mechanical stresses associated with cyclic with cyclic operation of equipment operation of equipment are considered and should be considered and included as included as part of the Equipment Qualiti-
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part of the aging programs.
cation Progree.
16 4.3 Synergistic ef fects should be con-4.3 Syneralstic ef f ects are considered and are sidered in the accelerated aging a part of SHNPP's ongoing Environmental prograus. Investigation should be Qualification Progree, performed to assure that no known synergistic ef fects have been identi.
fled on materials that are included in the equipment being quellfled.
Where synergistic etfacts have been Idontifled, they shoutd be accounted for in the quellfication programs.
Refer to NLREG/CR-0276 (SAND 78-0799) and NUREG/CR-0401 (SAND 78-1452),
"Quellfication Testing Evaluation i
Quarterly Reports," for additional Information.
4.4 The Arrhenius methodology is 4.4 in general, Arrhenius methodology and other considered an acceptable method of aging methods (when used) are supported oy addressing accelerated aging. Other type tests and supplementary analysis.
l aging methods that can be supported by type tests will be evaluated on a l
case-by-case basis.
l 3.11B-19 Amendment No. 16
f SHNPP FSAR D.
APPEN0lx 3. IIA NUREG-0588 COMPARISON CATEGORY II Applicable to Equipment Quellflod in Accordance wi th IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Program e.5 Knoen meterial phase changes and 4.5 Knoen material phase changes are evaluated reactions should be defined to if necessary, during quellfication to insure that no knosn changes occur Insure that no known changes occur within within the extrapolation limits.
the limits of qualification.
l
- 4. 6 The aging acceleration rate used
- 8. 6 The aging accoloration rate used during j
during quellfication testing and the quellfication testing and the basis for basis upon which the rate was estab-the rate is described and Identitled in lished should be descelbed and the Equipment Qualification Documentation juatifIed.
Package.
4.7 Periodic surveillance testing under 4.7 In general, Class IE equipment located in a normal service conditions is not harsh environment is quellflod by testing, considered an acceptable method for Periodic surveillance testing is not used ongoing quellfication, unless the plant as a method of quellfication.
cosign includes prowlsions for sub-Jocting the equipment to the limiting Ib service environment conditions (spect-fled in Section 3(7) of IEEE Std. 279-1971) during such testing.
4.8 Ef fects of relative humidity need 4.8 SHNPP compiles with this recommendarlon not be considered in the aging of electrical cable Insulation.
4.9 The qualifled iIfe of the equipment 4.9 The qualifled iIfe of the equipment and (and/or component as applicable) and the basis for its selection is included the basis for its selection should be in the specific Equipment Qualification I
defined.
Documentation Package.
4.10 Quellflod life should be established 4.10 Quellfled life is established as on the basis of the severity of the described.
testing performed, the conservatisms employed in the extrapolation of data, the operating history, and in ott.or methods that may be reasonably assumed, coupled with good engineering judgnent.
Em 0F APPt.lCAILE CATE90RY I PORTIONS OF SECTION 4 3.11A-20 Amendment No. 16
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SICIPP FSAR l
e APPEN0lx 3.11A NUREG-0588 COMPAR1 SON i
CATEGORY 11 fi Applicable to Equipment Quallfled in Accordance with IEEE Std. 323-1971 Shearon Harris Nuclear Power Plant Program 5
QUALIFICATION 00CtNENTATl0N (1) The staf f endorses the requirements 5
(1) The mein purpose of the quellfication stated in IEEE Std. 323-1974 that, documentation is to previde auditable "The qualification documentation evidence that each type of equipment shall verify that each type of is quellflod for its appilcation and electrical equipment is quellflod meets its specified performance for its application and meets Its requirements. Section 3.11 of the specified perforinence requirements.
SHNPP FSAR provides Information on The basis of quellfication shall be the type of documentation generated 16
- '*3"*d to show the relationship as evidence of quelltleation.
/
of all facets of proot needed to support adequacy of the conclete equipment. Data used to demonstrate the quellfication of the equipment shall be pertinent to the application and organized in an auditable form."
(2) The guidelines for documentation in (2) Refer to Section 3.11 of the SHNPP IEEE Std. 323-1971 when fully imple-FSAR for a description of the mented are acceptable. The docu-documentation generated to demonstrate montation should include sufficient qualification.
Information to address the required information identitled in Appendix E.
A certificate of conformance by itself is not acceptable unless it is acconcenled by test data and Informa-tion on the quellfication program.
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- "A Amendment No. 16 i
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