ML20053E755
| ML20053E755 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 06/07/1982 |
| From: | Bayne J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| IEB-79-01B, IEB-79-1B, JPN-82-49, NUDOCS 8206100015 | |
| Download: ML20053E755 (200) | |
Text
_ _ - _ _
POWER AUTHORITY OF THE STATE OF NEW YORK 10 COLUMBUS CIRCLE NEW YORK. N. Y.10019 GEORGEY BERRY (2528 397-6200 o*E R ATING OFFICER TRUSTEES f JOHN BOSTON JOHN S.DYSON C"*'""'" pnEstDENTMROCEDU RES aEORoE L.iNoALLS JOSEPH R, SCHMIEDER wtC E CnannuaN
, Etion T 6 CHIEF RICH ARD M FLYNN '""""
JPN-82-49 LEROY W. SINCLAIR ROBE n f 8. MILLO nil 4C EF FIN NC 4L F R E D E RIC K R. CLARK THOM AS R FREY Director of Nuclear Reactor Regulation a'EUla'6U"N.<'
U.S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Mr. Domenic B. Vassallo, Chief Operating Reactors Branch No. 2 Division of Licensing
Subject:
James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Environmental Qualification of Safety-Related Electrical Equipment Request for Additional Information
Reference:
PASNY letter, J.P. Bayne to D.B. Vassallo (NRC) dated June 2, 1982 (JPN-82-87), same subject.
Dear Sir:
The referenced letter was inadvertently transmitted without its enclosure. That enclosure, entitled " Revised Environmental Parameters and TMI Related Equipment", is hereby submitted for your use.
If you have any questions, please contact Mr. J.A. Gray, Jr.
of my staff.
Very truly yours, 0
~
- 4. . Bay
-Senior Vice President Nuclear Generation w/ attached: "*
w/o attached:
ON 8206100015 820607
{DRADOCK 05000333 PDR
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w/attachemnts: Mr. Cyril J. Crane Franklin Research Center The Parkway at Twentieth Street Philadelphia, PA 19103 w/o attachemnts: Mr. J. Linville Resident Inspector U.S. Nuclear Regulatory Commission i P.O. Box 136 Lycoming, New York 13093 Mr. Ron Barton-United Engineers & Constructors, Inc.
30 S.-17th Street Philadelphia, PA 19101 i
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l RESPONSE TO NRC IE BULLETIN 79-01B JAMES A. FITZPATRICK NUCLEAR POWER PLANT POWER AUTHORITY OF THE STATE OF NEW YORK b
D REVISED ENVIRONMENTAL PARAMETERS TMI RELATED EQUIPMENT l May 1982
Rev. 3 PREFACE
~N, s
This report is in response to the NRC issued I&E Bulletin 79-OlB (January 19, 1980) requiring all owners of Nuclear Power Plants to perform a detailed review of the environmental qualification of Class lE electrical equipment to ensure system operability under accident conditions.
The environmental parameters not addressed during the James A. Fitz-Patrick FSAR environmental qualification program, but presently re-quired such as submergence (Containment and Reactor Building), water spray, aging and High Energy Line Break (HELB) outside primary con-tainment are now included.
The qualification analysis performed utilized criteria and guidelines as established by the subject bulletin, the SER, subsequent NRC regional meetings and the NRC meeting held in Bethesda, Maryland on July 7-10, 1981.
Following a design basis loss of Coolant Accident (LOCA) within the primary reactor containment, the Class lE electrical equipment re-quired to isolate and mitigate the accident will perform its intended function within the " harsh environment" as detailed in the previously submitted " Response to IE Bulletin 79-OlB, October 1980" and within this report.
/
(\ , Similarly for a High Energy Line Break (HELB) outside the primary t ! containment analysis shows that all necessary isolation valves will
\ perform their function within the required time frame while sub-jected to the environment of the accident.
Action plans include equipment replacement, testing or engineering analyses. Although recommendations have been made for equipment re-placement, this does not indicate equipment failure. There is a high probability that the equipment could be qualified by test or analysis, including supportive documentation, if either method was chosen. Rather, replacement is the most cost and time effective method to establish environmental qualifications.
The HELB analysis included with the October 1980 submittal resulted in highly conservative results as credit v as not taken for the ef fects of heat sinks. The HELB analysis pres--tsd herein includes heat sinks, long term heat sources, and modifications resulting from the fire hazard analysis, thereby providing a more realistic trarsient response.
In addition to the HELB reanalysis, the following ef forts were per-formed subsequent to the October 1980 submittal:
a) Field survey to reflect the "As-Built" condition of the plant.
I b) A functional analysis to determine which electrical conponents x
i previously submitted are not essential for safe shutdown. These "i items have been marked " Delete" on the systems component evalu-ation worksheet.
Rev. 3 O
x' c) Analysis providing integrated radiation doses based on equip-ment operating times.
d) Conducted a plant radiation survey to extablish actual radi-ation exposure rates for Class lE equipment during the 40 year normal environment.
e) Analysis providing long term post-HELB and post-LOCA temperature in the Reactor Building.
f) Analysis providing humidity profile post-HELB and post-LOCA in the Reactor Building, g) Analysis providing radiation values post-LOCA in the RHR Heat Exchanger Rooms and Reactor Water Cleanup Fump Rooms.
h) Equipment Qualification Action Plan committments resulting from Safe' Shutdown System Analsyis.
- 1) Analysis providing temperature and pressure profiles in the Turbine Building.
The technical summary and assumptions used in establishing the HELB environmental envelope are being included to provide a basis for the f
^s implementation of the follow-up qualification program as well as equipment purchase for future modifications.
A. HELB Transients The Reactor Building transients due to High Energy Line Breaks were based on Reference 151. The peak area temperatures resulting from the HELB occur rapidly and are shown in Section 2 of this report.
Long term effects from equipment heat loads above the Crescent area do not affect the peak calculated values.
A secondary long term effect is seen from equipment operation in a few areas of the Reactor Building. The areas containing the RHR heat ex-changers and the RHR piping (nodes 272-5, 272-11, 272-1 and 272-2) will have a significant long term temperature rise from RHR system operation, and the area containing the inverter (71INV-3B) in long term temper-ature. In general, however, post-HELB equipment operation increases the overall building temperatures approximately 2'F in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The figures for nodes 272-5, 272-11, 272-1 and 272-2 have been plotted with the Suppression Pool temperature transient from FSAR and figure 14.6-7 superimposed. After a HELB or after a LOCA the RHR equipment will be cooling the Suppression Pool water, and these areas will there-fore follow the temperature transient of the pool.
The inverter area on the north side of elevation 344 ' is not well ven-tilated'by the SGTS, and slowly rises a few degrees above the rest
\ of the building in the long term. A straight line at 110*F has been added to the figure for 344-0 to envelope this effect.
Rev. 3 p
k ,j s B. Long Term Post-HELB Temperatures Although Reference 151 did not account for the equipment heat loads post-HELB above the Crescent area, e.g. MCC's, inverters, emergency lighting, and piping, neither did it take credit for heat losses through the exterior walls and siding, or for operation of the Standby Gas Treatment System (SGTS). After including equipment heat loads transmission heat gain from the Drywell, Suppression Chamber Ceiling, and Fuel Pool Walls (and Fuel Pool surface load), the losses roughly matched the gains. References 148 and 150 showed the slight (2*F) temperature rise which resulted. A slightly high-er rise occurs in the inverter area on El. 344' as shown in Refer-ence 149.
C. Long Term Post-LOCA Temperatures These heat loads are the same as post-HELB except that the Drywell is also assumed to stay at an average of 175'F for 1 month, af ter which both the Drywell and Suppression pool heat loads drop out.
Normal spent fuel pool cooling was assumed.
D. Seasonal Variations Any 6 month qualification period will encompass both winter and summer outside ambient conditions. Depending on the time of year of the postulated accident, the resulting long term building temp-
,/' sg
, erature will rise or fall over a period of 6 moi.hs. Our calcu-
\s / lated values provide an upper bound over this period. Based on average monthly norms presented in FSAR Table 2.21, the minimum long term Reactor Building temperature would be approximately 46'F in January with an outside ambient temperature of 25.l*F. The " thermal inertia" of the building will help damp the day to day variations, but an extended period of cold weather lower than 25'F could drop the Reactor Building temperature below the normal operating minimum temperature of 40*F.
E. Humidity Analysis Relative Humidity Above El. 272' The R.H. transients are presented in Section 2 for the RHR, HPCI, and RCIC breaks (steam line breaks), and for the RWCU breaks (liquid line breaks).
The transients are based on the site design conditions of 93*F dry bulb and 73*F wet bulb (65' dew point). The steady state relative humidity is very sensitive to the fuel pool temperature and to the external atmosphere relative humidity. However, according to Ref. 152 the ambient wet bulb at Oswego will be above 73*F between 2-1/2 and 5% of the cooling season (4 months). Therefore, the transients provided plus a 10% R.H. margin should be acceptable
( g for long term.
_ _ - - _ - . - . - - - - - . _ _ . _ - _ = . - - - . . - .
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Normal spent fuel pool cooling (SFPC) was assumed to be lost with i '
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! loss of offsite power. Operator action to cool the fuel pool via l the RHR system would require access to the Reactor Building approxi-
- mately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the HELB.
Assuming normal spent fuel pool cooling is available, the post-LOCA -
relative humidity will be enveloped by the post-HELB conditions. t
! Relative Humidity in the Crescent Area i
j The Crescent area relative humidity is conservatively assumed to
- remain at 100%.
Since the moisture removed from the Reactor Building by the demis-
) ters in the Standby Gas Treatment System (SGTS) is returned to the
- l equipment drains sump in the East Crescent, the water level will rise slowly (approximately 1/2" per day, see Refs. 35 and 96) in that area until sump pumps can be made available to pump it out or the STGS operation is terminated.
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lE BULLETIN 79-01B REV. 4 l JAMES A. FITZPATRICK NUCLEAR POWER PLANT I
l DOCKET NO. 50-333 PAGE 1 0F 2 ;
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- INDEX I
f SECTION TITLE l >
L I
j Reference Document Master List l
j 01 Equipment Qualification Plan / Master index Cross Reference List 1
i 02 Service Conditions i
i 03 Residual Heat Removal System (LPCI Mode) RHR t
! 04 Reactor Protection System (RPS) 05 Automatic Depressurization System (ADS) i 1 06 Standby Gas Treatment System i l l 07 Core Spray System f 08 Primary Containment & Reactor Vessel isolation Control System (PCRVICS) i 09 Primary Containment Atmospheric Control System 10 Reactor Building isolation & Control System 11 Emergency Service Water System 12 High Pressure Coolant injection System (HPCl) l l
1 13 Reactor Building Ventilation System i
j 14 Reactor Core isolation Cooling System (RCIC) i f 15 Electrical Penetrations j
16 Cables 17 Motor Control Center 18 Instrument Racks -
19 Junction Boxes j
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f IE BULLETIN 79-01B REV. 4 i JAMES A. FITZPATRICK NUCLEAR POWER PLANT j DOCKET NO. 50-333 PAGE 2 0F 2 i
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_.l N D E X i
- SECTION TITLE 1
l 20 Terminal Blocks l 21 Power Supplies 22 Batteries
! 23 Hiscellaneous f 25 Notes and Tables 26 Post-Accident High Range Stack / Vent Effluent Monitoring System; NUREG 0737, Sec. ll.F.1.1 i 27 Containment and Reactor Pressure and' Level Accident Monitoring j System; NUREG 0737, Sec. fl.F.1.4 and ll.F.1 5 I i i 28 Post-Accident Containment Radiation Monitoring System; l l NUREG 0737, Section ll.F.1 3 and ll.E.4.2 7 i
l 29 Post-Accident Sampling System; j NUREG 0737, Section ll.B.3.2 I
30 Primary Containment Hydrogen Sampling System; NUREG 0737, Section ll.F.1.6 '
31 Primary Containment and Leak-Rate Analyze System; NUREG 0661, - Torus Temperature Monitoring System .
3'2 indicating Lights / Meters / Recorders 33 Relays 34 Control Swltches 35 Splices and Connectors 36 Safety-Relief Valve Position Indication System; NUREG 0737, Section ll.D.3.1. '
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lE BULLETIN 79-01B Jamss A. FitzPatrick Nuclear Powar Plant REV. 4 Docket No. 50-333 i
REFERENCE DOCUMENT MASTER LIST
- 2. Pressure Transient Curves (See Section 2 - List of HELB Curves) 1 3 Radiation Specifications for JAF Response to 79-01B, SWEC Report,
! Issued Oct. 1980
- 4. FSAR Vol. IX Supp. 25, Special Report - Effects of a High Energy Piping System Break Outside of Primary Containment, item 20.
5 Franklin Institute Research Laboratories Report F-C4033-1, " Tests of Raychem Flamtrol Insulated and Jacketed Electrical Cables Under Simultaneous Exposure to Heat, Gamma Radiation, Steam and Chemical Spray while Electrically Energized," January 1975
- 6. Raychem letter, April 29, 1975.
t 7 Raychem Test Report EM-517, "The Effects of Radiation and Aging on
, Flamtrol Insulated Wire," April 8, 1972.
~ 8. Radiation International Report, " Irradiation of 3 Cable Samples, "
March 13, 1972.
j 9 G.E. letter, March 27, 1975 and G.E. Report No. MT-3-72, " Nuclear incident Qualification Tests on improved Vulkene Flame Resistant Control Cables," March 17, 1972.
l 10. The'0konite Co. letter, April 9, 1980, " Insulation and Jackets for l
Control and Power Cables in Thermal Reactor Nuclear Generating Stations," IEEE Transactions on Power Apparatus and Systems, Vol.
PAS-88, No. 5, May 1969.
- 11. The Okonite Co. letter, April 9, 1980, " Electrical Characteristics of Non-Filled XLPE Insulated Wire in 90C Water"
- 12. The Okonite Co. letter April 9,.1980, " Electrical Characteristics of Carbon Black Filled XLPE Insulated Wire in 90C Water" 13 The Okonite Co. letter, April 9, 1980, " Qualification of Okonite Ethylene-Propylene Rubber Insulation for Nuclear Plant Service"
- 14. Specification APO-76 (Okonite Co.), "600V Power Cables" 15 Specification APO-88B (Cerro Wire & Cable), "600/1000V Control cables" l s l
IE BULLETIN 79-OlB James A. Fi tzPatrick REV* 4 Nuclear Pow 2r Plant Docket No. 50-333
- 16. Specification APO-88C (G.E. Co.), "600/1000V Control Cables" 17 Specification APO-89 (Cerro Wire & Cable), "300V Instrumentation i Cables"
- 18. Specification APO-98 (Cerro Wire & Cable), " Thermocouple Extension Wires" 19 Specification APO-101 (Raychem), " Coax Cables" i 20. Specification AP-27 (G.E. Co.), " Electrical Penetration Assemb11es"
- 21. Specification APO-13A (Wm. Powell), " Cast Iron Valves with Motor Operators"
- 22. Specification APO-138 (Velan Engineering), " Cast iron Valves with 4 Motor Operators" 23 Specification APO-15 (Velan Engineering), " Carbon Steel Valves with Motor Operators"
, 24. Specification APO-70 (Fisher) 25 Specification AP-1 (G.E. Co)
/'~ 26. Lim 1 torque Test Report 600198, " Test of Limitorque Valve Operators,"
January 2, 1969 ,
- 27. Franklin Institute Research Laboratories Final Report F-C2232-01,
" Test of a Limitorque Valve Operator under a Simulated Reactor Containment Post-Accident Steam and Chemical Environment," November 1968.
- 28. Limitorque Test Report B0003, " Qualification Type Test Report Limitorque Valve Actuators for Class IE Service Outside Containment in Nuclear Power Station Service." Test performed November 13, 1974 to January 23, 1975 29 FSAR Table Q7 3-2, " qualification Test conditions"
- 30. FSAR Table Q7.3-3, " Qualification Test Requirements"
- 31. Rockwell Dept. 40-8648, Report No. 2792-03-02, Rev. I
- 32. Specification AP-20 l 33 FSAR Table 7.2-2 l
( 34. GE Qual. Test for F01 Electrical Assembly by RM Schuster, 4-30-71 l and PASNY's response letter to NRC, letter JAFP-79-60B1, 11/2/79 on '
! IE Bulletin 79-01 v
IE BULLETIN 79-OlB Jam 2s A. FitzPatrick Nuclear Powar Plant REV. 4 Docket No. 50-333
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ss </ 35. Stone & Webster Calculation (12966-PE(N)-022-0)
" Reactor Building Relative Humidity Transient for Equipment Qualification Following A High Energy Line Break", (See Section 2, Ref. 3A & 3B, for R. H. Transients)
- 36. Installation Procedure by GE (GEK-34666) 37 Prototype Testing Summary Tests EPAZ-007, EPAQ-008, EPAQ-009, EPAQ-01D
- 38. The Franklin Institute Research Laboratories Final Report F-C2857,
" Test of Electrical Cables Under Simulated Post-Accident Reactor Containment Service," September 1970, prepared for Cerro Wire and Cable Co.
l 39 Cerro Wire & Cable Co. Technical Report No. NP-04, " Tests of Electrical Cables after Simulated Post Accident Reactor Containment Service,"
March 1971, Supplement to Franklin Institute Report F-C2857
- 40. Rockbestos Co. letter, April 29, 1980 and Report, " Qualification of Firewall lli Class 1E Electric Cables (cross-linked insulation),
July 7, 1977, revised November 26, 1979
- 41. Masonellan letter, April 29, 1980 g Ref: PASNY Purchase Order No. APO-54B Stone & Webster Job Order No. 11825
. Masonellan Wo'rk Order No. 4-56236 Masonellan letter dated March 17, 1980 Stone & Webster letter dated April 1, 1980 Subj ect: Solenoid Valve and Limit Switch Environmental Qualification
- 42. Atwood & Morrill Co., Inc., dated March 14, 1980 James A. FitzPatrick Nuclear Station Power Authority of the State of New York P.O. No. APO-21-1 dated December 3, 1970 Stone & Webster J.O. No. 11825 I
Atwood & Morrill Order No. 12141 Qualification for ASCO Solenoid Valves and Limit Switches
, 43 Atkomatic Valve Co., Inc. dated March 20, 1980 l
Stone & Webster P.O. APO-62 dated April 14, 1971 l Our Invoice V-125523
- 44. General Electric Document No. 22A2928 Rev. 2 BUR Equipment Environmental Data, issued April 16, 1971
, 45. Stone & Webster Letter PAS-23699 dated May 7, 1980 l
Purchase Order No. NY0-6-77-29 Technical Services NRC IE Bulletin 79-01B l
Environmental qualification of Class lE Equipment i
James A. FitzPatrick Nuclear Power Plant Power Authority of the State of New York I
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IE BULLETIN 79-OlB JEmes A. FitzPatrick REV. 4 Nuclear Power Plant 4
Docket No. 50-333 k
- 46. BlF Letter dated May 6, 1980 Solenoid Valves and Position Switches Environmental Qualification of Class 1E Equipment James A. FitzPatrick Nuclear Power Plant BlF S/N 14340-1&2 (66A0V-100A&B) 14341-1&2 (66A0V-101 A&B)
BlF Proposal 02215-680418 47 Taylor Instrument Company, Division of Sybron Corporation, letter dated April 25, 1980 Environmental Qualification of Equipment James A. FitzPatrick Nuclear Power Plant
- 48. Specification APO-90 (Cryenco), " Nitrogen Supply and Storage System" 49 Transamerical Delaval Inc., Barksdale Controls Division, letter j dated April 25, 1980
! Pressure Switches l NRC IE Bulletin 79-01B Environmental Qualification of Class 1E Equipment James A. FitzPatrick Nuclear Power Plant
- 50. Specification APO-58 (RTD's, Thermocouples and Wells), " Electric f~s Thermometer Trinity, Inc."
- 51. Beckman instruments Inc. letter dated May 5, 1980 Environmental Qualification of Class 1E Equipment James A. FitzPatrick Nuclear Power Plant J.O. No. 12966.76, PAS-23563
- Beckman Contract 648781
- 52. Nuclear Measurements Corporation letter dated April 17, 1980 i ,
Radiation Monitors and Panels
, NRC IE Bulletin 79-01B Environmental Qualification of Class IE Equipment James A. FitzPatrick Nuclear Power Plant ,
- 53. Johnson Controls, Inc. letter dated May 6, 1980 Environmental Qualification of Equipment
- 54. Franklin institute Research Laboratories Final Report F-C3271
" Qualification Test of a Limitorque Valve Actuator in a Steam Environment," February 1972 55 ITT Barton letter dated May 6, 1980
. Differential Pressure Switches NRC IE Bulletin 79-01B Environmental Qualification of Class IE Equipment James A. FitzPatrick Nuclear Power Plant
IE BULLETIN 79-01B James A. FitzPatrick Nuclear Power Plant REV. 4 Docket No. 50-333 .
/' 'T 56. Limitorque Corporation letter dated May 12, 1980 Qualification
(
) Information James A. FitzPatrick J.0. 12966.76 PAS-23670
- 57. Figure 14.6-6 from FSAR
" Transient Results f rom LOCA Drywell Temperature", copy inserted in Section 2.
- 58. Figure 14.6-8 from FSAR
" Transient Results from LOCA - Containment Pressure," copy inserted in Section 2.
59 Stone & Webster letter PAS-23630, dated April 15, 1980 to Rockwell international. J.O. No. 12966. Solenoid Valves and Position ,'
Switches NRC IE Bulletin 79-01B Environmental Qualification of Class 1E Equipment, James A. FitzPatrick Nuclear Power Plant
- 60. Fisher Controls Company letter dated June 6, 1980 ,
Solenoid Valves and Positions Switches /
Environmental Qualification of Class 1E Equipment James A. FitzPatrick Nuclear Power Plant
- 61. Stone & Webster letter PAS-23562 dated April 1, 1980 to Atkomatic Valve Co., J.O. No. 12966.76 Solenoid Valves d NRC IE Bulletin 79-01B Environmental Qualification of Class 1E Equipment James A. FitzPatrick Nuclear Power Plant
- 62. Okonite letter dated June 24, 1980 with attached "LOCA Qualification
' of Okoguard Insulated Cables and T-95 and No. 35 Splicing Tapes" James A. FitzPatrick Nuclear Power Plant
, 63 Okonite letter dated August 25, 1980, 600V Electric Cables James A.
FitzPatrick Nuclear Power Plant
- 64. Exide letter dated June 19, 1980 with attached " qualification Test Report 44478-2 Battery Cells", James A. FitzPatrick Nuclear Powee Plant 65 Limitorque Test Report B-0027 "Limitorque Valve Actuator, Temperature Related to HigS Superheat Ambient Temperatures." James A. FitzPatrick Nuclear Power Plant l 66. Stone & Webster letter PAS-22936 dated November 3, 1978 to Power l Authority of the State of New York Task 12966.08-00022 Drywell i inspection and documentation of Class 1E Equipment ISE Circular 78-08 environmental qualification. James A. FitzPatrick Nuclear Power Plant v
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IE BULLETIN 79-01B REV. 4 James A. fitzPatrick Nuclear Power Plant Docket No. 50-333 ',
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) 67 Power Authority of the State of New York letter dated February 10, 1978 to the Nuclear Regulatory Commission. Written Response to IE Bulletin 78-02 Docket No. 50-333, Terminal Block environmental qualification, James A. Fit,xPatrick Nuclear Power Plant.
, 68. Limitorque letter dated Augur,t 28, 1980. Qualification information f
P/0 E-PP-1039, James A. FitzTatrick Nuclear Power Plant.
69 Franklin Institute Research Laboratories final report F-C3441,
" Qualification Test of Limitorque Valve Operators in a Simulated Reactor Containment Post-Accident iteam Environment."
< 70. Exide Form 7894R1-9.76E, BI:kup, power systems 400KW (500KVA) Uninterruptible Power System. James A. Fit 2 h t' rick Nuclear Power Plant.
- 71. Stone & Webster calculation (12966.76, PE(N)-006-0) " Maximum Flood Level Inside Primary Containment During DBA". James A. FitzPatrick
'f) Nuclear Power Plant.
- 72. Limitorque Corporation letter dated October 9, 1980 Qualification Information for James A. FitzPatrick P/0 E-PP-1039, 0/N 3E9919 73 Limitorque Corporation letter dated February 11, 1981 Qualification information for James A. FitzPatrick P/0 E-PP-1039, 0/N 3E9919 A LimitorqueCorporationketterdatedMarch 18, 1981 Qualification 74.
(V) information for James.A. FitzPatrick P/0 E-PP-1039, 0/N 3E9919 75 General Atomic High Range Radiation Monitor Energy Response Test and Dose Rat 6 Calibration Report No. 255978,
- p. <
- 76. G.E.TestNo.EPAQ-46,kpoxyRadiationTest
- 7
- 77. G.E. Test'No. EPAQ-4f,pCross-Linked Polyethylene insulation Radiation Test ^
l
! 78. G.E. Test No. EPAQ-11, 4ted April 21, 1968, Thermocycling Test l-l 79 Franklin ,Research Center Test Report No. F-C5285-1, dated May 2, 1980, qual. Test of Electrical Cables in a simulated LOCA Environment.
- 80. Mercold Corporation letter dated August 19, 1981 l
- 81. Connecticut Yankee Atomic Power Company letter dated March 29, 1978, Summary of Franklin Institute Environmental Qualification Test Program. Terminal Block / Box Combinations.
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IE BULLETIN 79-01B Jamas A. FitzPatrick REV. 4 Nuclear Power Plant Docket No. 50-333
- 82. SWEC Materials Engineering Division Technical Guidelines Number METG-8.1-0, Selection of Plastics and Elastomers for Radiation Environments 83 Specification AP-26 (Buffalo Forge Company) " Unit Space Coolers"
- 84. SWEC Response to NUREG 0578, item 2.1.6.b James A. FitzPatrick Nuclear Pcwer Plant, dated May 1980 85 Final Safety Analysis Report, Volume Ill, Table 7 2.2, James A.
FitzPatrick Nuclear Power Plant
- 86. Stone & Webster Calculation (12966-C-76-1), "Arrhenius Analysi s to Verify 40 Year Aging of General Electric 600V and 1000V Control Cable", James A. FitzPatrick Nuclear Power Plant 87 General Electric Letter, September 21, 1981
- 88. Stone s webster Calculation (12966-C-76-2), "Arrhenius Analysis to Verify 40 Year Ag:ng of Okonite RHR and CS Cable", James A. FitzPatrick Nuclear Power Plant 89 Stone & Webster Calculation (el2966-RP-76-009), " Post-LOCA 1-hr to 6-month Air immersion Doses in Primary and Secondary Containment",
James A. FitzPatrick Nuclear Power Plant
- 90. Stone & Webster Calculation (12966-RP-76-14), " Post-LOCA Dose Levels in Certain HELB Nodes, ELS 272 and 300", James A. FitzPatrick Nuclear Power Plant
- 91. Stone & Webster Calculation (12966-RP-76-13), " Accumulated Doses for i Hour to 6 Months Exposures to Pipe with Post-LOCA Water Activity", James A. FitzPatrick Nuclear Power Plant
- 92. Stone & Webster Calculation (12966-RP-76-008), " Post-LOCA 1 Hour to 6 Month Doses in Reactor Bldg. from Activity in Piping and Equipment", James A. FitzPatrick Nuclear Power Plant 93 NAMCO Controls Letter, August 12, 1980, Switches for Use in Nuclear Power Plants
- 94. Transamerica Delaval Letter, July 24, 1980, Pressure Switches, James A. FitzPatrick Nuclear Power Plant 95 Xomox Corporation Telex, December 15, 1981, Stone & Webster APO-40, Power Authority of the State of New York
- 96. Stone & Webster calculation (12966-PE(N)-020-0) , " Flooding Levels of Reactor Building Floors Resulting from High Energy Line Breaks",
James A. FitzPatrick Nuclear Power Plant
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IE BULLETlH 79-01B James A. FitzPatrick REV. 4 Nuclear Powar Plant Docket No. 50-333 I
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97 Stone s webster Calculation (12966-B-76-1), " Determination of Reactor Building Elev. 272'-0", 300'-0" and 344'-6", Post-LOCA Temperatures", Superseded by Ref's. 148, 149, & 150, James A.
FitzPatrick Nuclear Power Plant
- 98. General Electric Confirmation 99 Automatic Switch Company Letter, March 9, 1982 100. Target Rock Corp. Report No. 2199A dated January 9, 1979 and Addendum i dated November 26, 1979 " Qualification Test Report, Pneumatic Assembly; Air Operator Assembly P/N 7567F-000-19 and Solenoid Assembly P/N 1/2 SMS-A-01-2.
101. JAFNPP Component Functional Analyses for Determination of Environmental Qualification Requirements for NRC Bulletin 79-01B.
102. NRC Bulletin 79-01B, DOR Guidelines, Section 7 Aging Requirements.
103. JAFNPP Turbine Bldg. Steam Line Break Analyses dated 6/19/81, 12966 PEN 019-0.
104. JAFNPP Reactor Bldg. Ventilation System Design Basis and Calculations dated 2/12/71, 11825-66-20.
O t 105. JAFNPP FSAR, Figure 14-6-7 Temperature Profile: " Transient Results form LOCA Suppression Pool Temperature".
106. JAFNPP FSAR, Section 14.6 " Analysis of Design Basis Accidents".
107. SWEC Calculation 12966PE(N)-006-0 dated September 30, 1980 " Maximum Flood Level inside Primary Containment During the DBA."
108. Bailey Environmental Qualification Test E227 File #145C3007 GEMAC 50-555 109. General Electric Environmental Qualification Test Summary NSE80372 Instrument Dept. - Laboratories Operation Report 4/28/70, GEMAC 50-555.
110. G.E. NED0-24286 Results on Qualification Data Search for JAFNPP Table A-4 Item #1, GE SK6339XC90A.
111. G.E. Environmental Qualification Test Summary NSL80365 Ogden Technology Laboratories test procedure #70528, dated 4/23/71, Static-0-Ring, 12A-BB-NX.
112. BWR Equipment Qualification Summary Report #QSR034-A-01, 10/8/80, Static-0-Ring, 12N-BB-NX.
113. G.E. Environmental Qualification Test Summary NSE80366, Ogden Technology Laboratories Test Report #70528, 4/23/71, Static-0-ring,
N) w' 12N-BB4-NX.
- m. __ _ . . _ . , , _ m , .
IE BULLETIN 79-01B Jamas A. FitzPatrick REV. 4 Nuclear Powar Plant Docket No. 50-333 0
i 114. Same as 113, except NSE80367, Static-0-Ring, 12N-BB-NX. j 115. G.E. Environmental Qualification Test Summary NSE80368, Ogden Technology Laboratories Test Report #70526, Static-0-Ring, SN-AA3-X.
116. BWR Equipment Qualification Summary #034-A-01, Static-0-Ring, SN-AA3-X.
117. G.E. Environmental Qualification Test Summary NSE80360, Approved Engineering Test Laboratories Report #596-0398, 8/13/75, Barksdale B2T-M2355 118. G.E. Environmental Qualification Test Summary NSE80355, Yarway Environmental Test Procedure, Lockheed Electronics Company Inc.
Test Report #2755-4755, 8/4/71, Yarway, 4418EC.
119. G.E. Environmental Qualification Test Summary NSE80350, Test Report Approved Engineering Test Laboratories Report, Number 596-398, dated 8/13/75, Barksdale, B2T-M23SS.
120. G.E. Environmental Qualification Test Summary NSE80351, Addendum to TP-199-4 dated 4/27/71, Barton 288A.
121. G.E. Environmental Qualification Test Summary NSE80356, Yarway
/ Environmental Test Procedure for Fig. 4418EC, Lockheed Electronics Company Inc., 2755-4755, 8/4/71, Yarway, 4418EC.
122. BWR Equipment Qualification Summary, QSR 031-A-01 SEE Cl - 032 -
Environmental qualification Table E12-1, Static-0-Ring, SN-AA3-SX.
123. BWR Equipment Qualification Summary, QSR-029-A-01, Environmental qualification Table E12-3, Barton, 289-2419 124. G.E. Environmental Qualification Test Summary NSE80167, Ogden l Technology Laboratories Test Procedure 70528, 4/23/71, Static Ring.
125. G.E. Environmental Qualification Test Summary NSE80357, ITT Barton TP-199-4, 5/24/71, 288A.
126. G.E. Environmental qualification Test NSE80353, Yarway 4418C.
l l
, 127. G.E. Environmental Qualification Test Summary NSE80359, Viking j Laboratories Test Letter Report #30411, dated 8/16/76, Fenwal 17002-40.
128. G.E. Environmental Qualification Test Summary NSE80175, ITT Barton TP-199-4, 5/24/71, Barton 289 O
\g 129. G.E. Environmental Qualification T9st Summary NSE80382, Approved Eng. Test Lab Report #596-0398, Barksdale B2T-M12SS.
130. Deleted
IE BULLETIN 79-01B James A. FitzPatrick REV. 4 Nuclear Powar Plant Docket No. 50-333 A
131. BWR Equipment Qualification Summary QSR-630-A-01, Environmental Qualification Table C11-1, Magnetrol 5.0-751.
132. Deleted 133. BWR Equipment Qualification Summary QSR-116-A-01, Robertshaw 83842-A8.
134. Limitorque Test Report #600376A dated May 13, 1976; " Nuclear Power Station Qualification Type Test Report - Limitorque Valve Actuators for BWR Service", 10MOV-18.
135. G.E. Environmental qualification Test Summary NSE80389, Wyle Laboratories Report #12625, dated 6/5/62. Barksdale D24-A150SS.
136. G.E. Environmental Qualification Test Summary NSE80391, Ogden Technology Laboratories Test #5N-AN3-SX, Static-0-Ring Procedure 70526, dated 5/4/71, Static-0-Ring, SN-AA3-SX.
137. BWR Equipment Qualification Summary 048-A-03, Table E-32-6, Rosemont 1151GP.
138. BWR Equipment Qualification Summary 044-E-01, Test Report #228, 50-551, GEMAC 551-032EKZZ2.
139. G.E. Environmental Qualification Test Summary NSE80373, Ogden Technology Laboratories Test Procedure 70526, dated 5/4/71, Static-0-Ring, 6N-aAA21-V.
140. Deleted 141. BWR Equipment Qualification Summary QSR-078-A-01, GE AKD-5 142. Normal Reactor Building Radiation Levels in the Vicinity of Class 1E Equipment for NRCB 79-01B, September 3, 1981.
143. Limitorque Test Report #B0058, dated 1/11/80, Limitorque Valve Actuator Qualification for Nuclear Power Station Service.
144, Qualification Test of ASCO Solenoid Valves, Test #AQS21678/TR -
Rev. A.
145. NAMCO Controls Test Report No. QTR 111, " Qualification of EA740 Series Limit Switches for Use in Nuclear Power Plants in Compliance with IEEE Standards 323-1974, 382-1972 and 344-1975", October 1, 1981 146. Stone & Webster Calculation (12966-B-76-2), " Determination of East Pipe Tunnel Post-LOCA Space Temperatures", James A. FitzPatrick Nuclear Power Plant 147. Stone & Webster Calculation (12966-PE(N)-001-3) , " Revision of b Reactor Building Subcompartment Model to Account for Fire Hazard Modifications", James A. FitzPatrick Nuclear Power Plant
IE BULLETIN 79-018 )
REV. 4 James A. FitzPatrick Nuclear Pownr Plant Docket No. 50-333 0, 148. Stone & Webster Calculation (12966-PE(N)-024-0), " Reactor Building Long Term Temperature Post-LOCA & Post-HELB", James A. FitzPatrick Nuclear Power Plant 149. Stone & Webster Calculation (12966-PE(N)-025-0), " Reactor Building Inverter Room Long Term Steady State Temperature, Post-LOCA & Post-HELB", James A. FitzPatrick Nuclear Power Plant 150. Stone & Webster Calculation (12966-PE(N)-026-0), "Effect of Equipment Heat Loads on the Reactor Building for the First 24 Hours Post-LOCA
& Post-HELB", James A. FitzPatrick Nuclear Power Plant 151. Stone & Webster Calculation (12966-PE(N)-003-5) , " Revised Results of High Energy Line Break Analysis in the Reactor Building including Fire Hazard Modifications", James A. FitzPatrick Nuclear Power Plant 152. Evaluated Weather Data for Cooling Equipment Design, Addendum N. 1 Winter and Summer Data, Fluor Products Inc. , Copyright 1964, Pg. 40 153. Stone & Webster Calculation (12966-PE(N)-010-0), " Heat Balance and Heat Up Rate of the Spent Fuel Pool Before/After the Loss of the SFPC System with RHR Providing Emergency Cooling, Case til of JPN-80-19", James A. FitzPatrick Nuclear Power Plant 154. Stone & Webster Dwg. No. Il825-FB-4A-10, Reactor Building Floor Drainage El. 227'-6" & 256'-6" 155. Stone & Webster Dwg. No. Il825-FB-4B-10, Reactor Building Floor Drainage El. 272'-0" 156. Stone & Webster Generic Calculation (Generic-PE-219-0), " Condensate Film Thickness After a LOCA" 157. Response to FSAR qusetion 7.3, Item 8 - FSAR Volume Vill 158. Philadelphia Electric Company Test "A" Report Summary Qualification Test of Electrical Terminal Blocks and Cable Splice Insulation, dated 1/17/79 159. Environmental Qualification Test Report of Raychem Nuclear Cable Brakeout and End Sealing Kits, Report #58442-2, dated April 3, 1981 160. Qualification Report of NAMCO EA180 Limit Switches Report, #QTR-105 161. Rockbestos Co. Report (qualification of Firewall ill Class IE Electric Cables (IR Radiation Cross-Linked Insulation)," June 7, 1978 i
o
IE BULLETIN 79-OlB James A. FitzPatrick REV. 4 Nuclear Power Plant Docket No. 50-333
)
}
162. Rockbestos Co. Report QR #1804 " Qualification of Firewall EP Class IE Electric Cables", April 6, 1981 163. JAFNPP, FSAR Section 4.8.6 - RHR System (Containment Spray Mode)
Design Basis 164. IEEE 323-1974 and NUREG 0588, Cat. 1 Aging Requirements 165. Eaton Co. Qualification Report (Isometics Test Report, June 1978),
LOCA Nuclear Qualification of EPDM/Hypalon Composit Cable 166. Deleted 167. Stone and Webster Letter PAS-25475, dated April 2, 1982 and PAS-25482, dated April 12, 1982 to PASNY. J.O. No. 12966.76. Technical Summary and Enveloping Relative Humidity curves for the Reactor Building Post-LOCA/HELB.
! 168. JAFNPP (SWEC) System Design Description - Motor Generator Room Ventilation (21-7.3) - System 66, dated Nov. 28, 1973 169. JAFNPP (SWEC) System Design Description - Heating, Ventilating &
Cooling for the Turbine Building (21-8.1) - System 67, dated Maj 2, 1978.
170. JAFNPP (SWEC) System Design Description - Control Room Air Conditioning System (21-10.A2) - System 70, dated July 19, 1974.
171. JAFNPP (SWEC) System Design Description - Relay Room Air Conditioning l System (21-10. Al) - System 70, dated July 19, 1974.
V
IE BULLETIN NO. 79-OlB James A. FitzPatrick REV. 4 Nuclear Power Plant Docket No. 50-333 l
d 200. Final Report on the Evaluation of Raychem WCSF Cable Splices, PEl-TR-82-4-2 (Reactor Building) 201. Final Report on the Evaluation of Anaconda FR EP Instrumentation Cable, PEl-TR-82-4-3 202. Final Report on the Evaluation of General Electric EB-25 Terminal Blocks, PEl-TR-82-4-4 203. Final Report on the Evaluation of Trombetta Solenold, G206, OEl-TR-82-4-28 204. Final Report on the Evaluation of Electroswitch, 2401C and 2402C, PEl-TR-82-4-16 205. Final Report on the Evaluation of Rockbestos Firewall 111, SIS, PEl-TR-82-4-9 206. Final Report on the Evaluation of GE CR2940 Switch PEl-TR-82-4-10 207. Final Report on the Evaluation of GE CR120B Relay PEl-TR-82-4-14 208. Final Report on the Evaluation of NAMCO EA740-80100 Limit Switch PEl-TR-82-4-18 209. Final Report on the Evaluation of Belden Wire Type 83348 PEl-TR-82-4-5 210. Final Report on the Evaluation of Buchanan Terminal and Fuse Blocks, PEl-TR-82-4-13A 211. Final Report on the Evaluation of Buchanan Terminal and Fuse Blocks, PEl-TR-82-4-13B w/in Hoffman NEMA 4 Enclosure 212. Final Report on the Evaluation of Boston Insulated Wire Signal Cable, XLPE-CSPE, PEl-TR-82-4-12 213. Final Report on the Evaluation of Conax RTD W/ Seal Assy.
PEl-TR-82-4-1 214. Final Report on the Evaluation of Conax EPA PEl-TR-82-4-31 215. Final Report on the . Evaluation of Thomas & Betts Tefzel Connectors PEl-TR-82-4-45 216. Final Report on the Evaluation of Magnetrol Level Switch BCS-751-EP, PEl-TR-82-4-17 217. Final Report on the Evaluation of Raychem WCSF Cable Splices, PEl-TR-82-4-37 (Primary Containment) hg 218. Final Report on the Evaluation of Champlain Switchboard Wire s
v / AWM 3271vw-l PEl-TR-82-4-35
i lE BULLETIN 79-01B James A. FitzPatrick REV. 4 Nuclear Power Plant Docket No. 50-333 219. Final Report on the Evaluation of General Electric Vulkene Flame Resistant Control Cable PEl-TR-82-4-27 .
l 220. Final Report on the Evaluation of Raychem insulatino End Caps i PEl-TR-82-4-37A (Reactor Building)
- 221. Final Report on the Evaluation of Raychem insulating End Caps with WCSF Outer Sleeves PEl-TR-82-4-37B (Primary Containment) 222. Final Report on the Evaluation of Limitorque Valve Operator SMB-1, PEl-TR-82-4-42 1 223. Interim Report on the Evaluation of Struthers Dunn 219 Relay with CX3965 Base, PEl-TR-82-4-8 224. Final Report on the Evaluation of Rockbestos Cable RSS-6-104-1981, PEl-TR-82-4-34 225. Final Report on the Evaluation of Target Rock Solenoid Valve 81JJ-004, PEl-TR-82-4-36 i
a
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, . c - - - - - - - _ - _ _ _ _
Rev. 3 EQUIPMENT CUALITICATION PLAP Page 1 of 2 s
- A. The scope cf work within this report includes
! 1. Master Index Cross Reference Lict
- 2. System Component Evaluation ' Work Sheets i 3. L:st of source information refercnces
- 4. - Notes '
- 5. LOCA & HELB Environmental Curves Te establish qualification of the NSSS supplief Class IE equip-ment, the General Electric Design Specification 22A2928 was used as the source document. However, General Electric has been re-quested to provide additional infermation with regard to qualifi-l cation and qualification methods for Class II equipment within
! their scope of supply (Ref.: Letter G-E?l-0-52 dated April 11, 1980).
i S. WORK SHEET CLARIFICATIONS i
i l 1. To utilize the Cross-Reference Master List, the System Section Number should be determined from the Index Sheet i in front of the report. This section number will be found in the third column of the Master List along with the i appropriate page of the work sheet. In addition a Rack /
Instrument Cross-Reference Index and a listing of instru-ments by HELB Nodes is available to locate instruments by i Rack or by HELB Node.
["N t 2. Where the HELB curves are referenced on the worksheets,
)
- \s_, the appropriate curve /s number is tabulated in the work- i sheet " Document Reference Specification Column." The l required curve /s will be found in Section 2.
i 3. Ref erence numbers tabulated in the " Document Ref erence -
Qualification" column of the work sheets ref er to documents listed in Section 24, " Reference Document Master List."
- 4. An asterisk indicates a parameter not previously considered at the time of license issuance. The resolution of all para-meters with an asterisk is found in 'the " Outstanding Items" column.
- 5. The use of the word none in the " Outstanding Items" column indicates that the qualification test environment was equal to or more stringent than the expected HELB/LOCA accident enviro nment.
- 6. The use of the word none in the " Qualification Method" column indicates that no qualification tests were performed for that parameter.
- 7. Numbered notes, i. e. Note 5, Note 7, etc. are tabulated in Section 25.
- 8. " Delete" indicates that a functional analysis has determined the equipment is not essential to safe shutdown.
j Note: These sheets have not been updated to include the latest environmental and qualification data.
. _ _ _ _ , -~ . _ , _ - .. - - _ _ . ~ _ _ . _ . - _ - _ _ . _ , _ . - _
Rev. 3 Page 2 of 2 C. REACTOR CORE ISOLATION COOLING SYSTEM (RCIC)
[\ The RCIC is referenced in the " Emergency Operation Procedures" for
\ ,) initiation in the event of a LOCA. Although available, it has not been considered in the safety analysis of the plant and therefore is not required to mitigate the consequences of an accident. We have included it in this report on an " informational basis" and hence does not preclude the original analysis.
D. STANDBY LIQUID CONTROL SYSTEM (SLC)
The SLC system although considered a "special safety system" is needed only in the improbable event that not enough control rods can be inserted in the reactor core to accomplish shutdown and cooldown in the normal manner.
The system is expected not to be needed for unit safety nor re-quired to mitigate a LOCA or HELB.
E. AGING This item was not addressed during the FSAR environmental qual-ifications analysis.
F. HIGH ENERGY LINE BREAK (HELB)
HELB analysis for outside primary containment previously relied on separation and redundancy of safety related equipment to insure safe shutdown of the reactor and therefore did not re-quire environmental qualification- for HELB. A HELB analysis
,,/) has been performed and the results factored into the qualification Refer to Section 02 and approp-
- requirements for the equipment.
riate work sheets.
G. SUBMERGENCE Maximum Flood Level within Primary Containment and the Reactor Building is 258'-6" and 227'-11" respectively. (See ref. 71 and 96.) All safety related equipment is located above this level.
! H. QUALIFICATION PROGRAM Each System Component Evaluation Work Sheet with an outstanding item has an explanatory note. The future qualification activities indicated by these notes can be summarized as follows:
1 l 1) Equipment lacking complete environmental qualification will be replaced. Environmentally qualified replacement equip-ment will be purchased and installed in accordance with the .
(
schedule submitted in the " Response to Safety Analysis Report".
- 2) Equipment with a limited qualified life will be entered into a scheduled replacement program.
- 3) For equipment needing additional environmental qualifications,
[, the BWR Equipment Qualification Owner's Group will provide
\ _,,} the necessary qualification program.
I
[
IE BULLETIN 79-OlB REV. 3 JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333
! i SERVICE CONDITIONS )
l l For equipment identified as Inside Primary Containment in the " Master" List, the accident service conditions and typical accident qualification
- test requirements are given in Table 1 and Attachments 1 and 2. For i
equipment identified as Outside Primary Containment, the accident service conditions are given in HELB temperature and pressure transient curves.
l Detailed environmental test specifications and qualification documentation i
are given on the " System Component Evaluation Work Sheets."
y
'N I
/
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,_m..-....., ..~ . - - _ _ . - - , - . - . . . . - . . , _ . . - - _ _ _ - - - - - -----,.-.----m.,m.-~. .~
l Page 1 of 2 Rev. 3 l
1 IE BULLETIN 79-01B JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCUMEFf NO. 50-333 DOCUMENT REFERENCE NO.
i TEMPERATURE PRESSURE DESCRIPTION OF NODE NODE NUMBER
{ IA 2A West Crescent 227-1 IB 2B East Crescent 227-2 1C 2C HPCI Room 272-1 ID 2D RCIC Room ~
272-2 1E 2E Outside RCIC Room 272-3
, IF 2F Southwest CRD Area 272-4 IG 2G RHR Heat Exchanger Room A 272-5 1H 2H Northwest CRD Area 272-6 IJ 2J Northwest Corner 272-7 IK 2K CRD Repair Area 272-8 IL 2L Southside Room 272-9 IM 2M Railroad Bay 272-10 IN 2N RHR Heat Exchanger Room B 272-11 IP 2P Southeast CRD Area 272-12 IQ 2Q Northeast CRD Area 272-13 j . IR 2R Neutron Monitoring TIP Room 272-14 IS 2S Access Area 272-15 I 2T IT Outside + Above CRD Repair 272-16 l fs IU 2U RWCU Pump Room B 300-1 l r IV 2V RWCU Pump Room A 300-2 l 1W Open Bay
- 2W 300-3 IX 2X Small Hatch Area 300-4 1Y 2Y North Passage 300-5 IZ 2Z Northeast Stairway Area 300-6 IAA 2AA RWCU Heat Exchanger Room 300-7 1AB 2AB Phase Separator Tank Room 300-8 l 1AC 2AC Large Hatch Area 300-9 1AD 2AD Sludge Pump Room 300-10 1 1AE 2AE Combined 326'-9" Elevation 326-0 l 1AF 2AF Combined 344'-6" Elevation 344-0 l 1AG 2AG Combined 369'-6" Elevation 369-0 l IAH 2AH Lower S-W Torus Room 227-3 l
1AJ 2AJ Lower S Torus Room 227-4 1AK 2AK Lower S-E Torus Room 227-5 1AL 2AL Lower E Torus Room 227-6 1AM 2AM Lower N-E Torus Room 227-7 1AN 2AN Lower N Torus Room 227-8
, IAP 2AP Lower N-W Torus Room 227-9
( 1AQ 2AQ Lower W Torus Room 227-10 1AR 2AR Upper S-W Torus Room 244-1 1AS 2AS Upper S Torus Room 244-2 1AT 2AT Upper S-E Torus Room 244-3 1AU 2Ad Upper E Torus Room 244-4 1AV 2AV Upper N-E Torus Room 244-5 1AW 2AW Upper N Torus Room 244-6 1AX 2AX Upper N-W Torus Room 244-7 1AY 2AY Upoer W Torus Room 244-8
~. .._ . _ _ _ - _ . _ _ _ . _ _ _ . ._ _ . . . . . _ . . . . _ . . . . _ _ _ . _ . . - _ _ _ _ . . . . - . _ _ _ _ _ _ _ _ _ _ . . _ _ ___ _ _
Page 2 of 2 Rev. 3 IE BULLETIN 79-OlB I JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCUMENT NO. 50-333 DOCUMENT REFERENCE NO.
TEMPERATURE PRESSURE DESCRIPTION OF NODE NODE NUMBER 1AZ 2AZ RCIC Fire Enclosure --
IBA --
'Small Hatch Area, North Hallway & Long Term Envelope 344-364 1BB -- Post LOCA & Post HELB Long Term Analysis 344-3&4 IBC --
Post LOCA & Post HELB Long Term Analysis RB DOCUMENT REFERENCE NO.
RELATIVE HUMIDITY DESCRIPTION OF NODE NODE NUMBER l
3A Post HELB Dehumidification, RHR, HPCI, RCIC Breaks RB i 3B Post HELB Dehumidification, RWCU Breaks RB s
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"' ft.AN EL 344' C FIG.12.3-5 REACTOR BUILDING PLAN EL.344*-6" JAMES A.FITZFRTRICK NUCLEAR POWER PLANT POWER AUTHORITY OF THE STATE OF NEW YORK FINAL SAFETY ANALYSIS REPORT
IE BULLETIN 79-OlB REV. 3 ATTACIMENT #1 (1 OF 3) O . . . I , . 250 - d 200 - c F _ Y ^ 150 r x o !
- 100 -
I 50
- a a FULL RMR CAPACITY 4 RHR PUMPS,2HX, ICS W/ CONT. $ PRAY ,,.
D a t LOOP,2 RMR PUMPS,l HX,1CS W/ CONT. $ PRAY c = 1 LOOP I RMR PUMP,IHX,103 W/ CONT.$ PRAY
- 8 = 1 LOOP. ! RMR PUMP,1 HX,1C3 W/o CONT. SPRAY -
0 0 I 2 3 5 10"I 10 10 10 10 10' 10 lo s TlWE (~9ec) FSAR FIG.14.6-6 TRANSIENT RESULTS FROM LO C A-DRYWELL TEMPERATURE A JAMES A.FITZPATRICK NUCLE AR' POWER PLANT POWER AUTHORITY OF THE STATE OF NEW YORK FIN AL SAFETY ANALYSIS REPORT
- + - - - - - = - - - ~' ' ' ' ' ' ' '
' ' = ' * *
- IE BULLETIN 79-OlB REV. 3 ATTACHMENT vil (2 OF 3) s 250 g g g g g 4 c,8 200 -
6 a 150 C
*l Y
B - i
.J g i. -
s _ a a FULL RMR CAPACITY 4 RMR PUMP 3. 2 MX,1 CS, W/ CONT. $ PRAY t al LOOP,2 RMR PtJMPS,I MX,1 C3 W/ CONT. SPRAY c = 1 LOOP,1 RNR PUMP,1 MX,1 C3 W/ CONT. SPRAY d a t LOOP,I RMR PUMP,1 MX,1 C3 W/0 CONT. $ PRAY . 30 _ e
~
10 10 to 10 10 10 10 to TIME FSAR i FI G.14.6- 7 TRANSIENT RESULTS FROM LOC A-PRESSURE SUPPRESION POOL TEMPERATURE
/
JAMES A.FITZPATRICK NUCLE AR POWER PLANT POWER AUTHORITY OF THE STATE OF NEW YORK '
/
FIN AL SAFETY ANALYSIS REPORT
IE BULLETIN 79-OlB REV. 3 ATTACHMENT #1 (3 0F 3) g 4 4 4 4 a 3 a = FULL RMR CAPACtTY 4 RMR PUMPS,2 HX,1 CS W/ CCNT. $ PRAY b = 1 LOOP,2 RMR,1 HX,1 C$ W/ CONT. $ PRAY
. c = 1 LOOP,i RMR,1 MX,1 C$ w/ CONT. $ PRAY d = 1 LOOP,1 RMR, i HX,1 CS w/o CONT. $ PRAY 40 - -
ORYwELL S a - _ E 1 5'
=
g - - A IE 20 - g - u d c
- 8 10 -
TORRUS a 0 - ' ' ' ' ' ' 10 to 10 10 10 10 10 10 TIME (~sec) FSAR FIG '14. 6-8 TRANSIENT RESULTS FROM LOC A-( CONTAINMENT PRESSURE JAM ES A. FITZ PATRICK NUCLE AR POWER PLANT POWER AUTHORITY OF THE STATE OF NEW YORK FIN AL SAFETY ANALYSIS REPORT
ATTACHMENT #2 (1 OF 2) Rev. 3
, IE Bulletin 79-OlB JAFNPP FSAR
! TABLE Q.7.3-2 i QUALIFICATION TEST CONDITIONS Test Valve- Condition
- Relief valves 1 HPCI steam line isolation valve 2 RCIC steam line isolation valve 2 Reactor vessel head spray isolation valve 2 Main steam line isolation valves 3 1
Main steam line drain isolation valve 3 RHR shutdown cooling isolation valve 2 Reactor water cleanup isolation valve 2 I . l l l
- Test conditions refer to qualficiation test requirements listed on Table Q. 7. 3-3 1 of 1 l
c m- T g.
\bu fAClit!Effr 2 (2 OF 2) k JAFNPP 7. 3 PSAR IE BULLETIN 79-OlB TAnts o.7.3-3 , QUALIFICATION TEST REQUIREMENTS Test' Condition Supplier components Must- Be Omrable Under The Following Con 11tions 1 GE Temperature 340 F 340 F 3 20 F 253 F 200 F Pressure -2 to 56 psig -2 to 35 psig -2 to 35 psig 0 to 25 psig 0 to 20 psig Rel. humidity 100% 100% 100% 103% 100%
Duration 45 sec 3 hours 6 hours 1 day 100 days Radiation
- 1) The total integrated dose during the design basis LOCA added to the long-term operating doce will be 2 x 107 rads.
2 SCW Temperature 304 F 275 to 304 F 200 to 275 F 210 F Pressure 56 psig 56 psig 56 psig 56 psig ! Rel, humidity 100% 100% 100% 100% Duration 20 sec 200 sec 600 see 24 hr Radiation 600 RAD /hr 600 RAD /hr 600 RAD /hr 600 RAD /hr gar.ma gamma gamma gamma
- 0. 5 PAD /hr 0.5 RAD /hr 0.5 R AD/hr 0.5 R AD/hr neutron neutron neutron neutron i
3 SCW Temperature 310 F Pressure 62 psig Rel. humidity 100% Duration 30 see Radiation ., 3 x 10s ny fast 21 mey 75 rad /hr gamma i Electrical cables must be operable under the following conditions: l Temperature 340 F 281 F 160 F Pressure 62 psig 56 psig 56 psig Rel. humidity >90% >905 >90%
- 2 Duration 15 min 2 hr 24 hr Radiation
- 1) The total integrated dose during the design basis LOCA added to the long-term operating dose will be 6 x 107 rads.
l 1 of 1 Supplement 0
IE BULLETIN 79-OlB . REV. 3 JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 LIST OF REFERENCES REF. NO. 1' FSAR Figure 14.6-6 " Transient Results From LOCA - Drywell Temperature" 2 FSAR Figure 14.6-7 " Transient Results From LOCA- Pressure Suppression Pool Temperature" l 3 FSAR Figure 14.6-8 " Transient Results From LOCA - Containment Pressure" 4 Response to AEC Question 7.3 of 1/12/72 (Supplement 8) 5 Response to AEC Question 7.22 of 8/11/72 (Supplement 12) 6 FSAR Appendix I, Figure 1.1-1 " General Arrangement, Drywell and Suppression Chamber" 7 FSAR Appendix I, Figure 1.3-1 "Drywell Elevation" 8 Special Report " Effects of a High Energy Piping System Break Outside of Primary Containment," July 22, 1974. (FSAR Volume IX, Supplement 25) 9 Response to AEC question 6.3 of 11/29/71 (FSAR Supplement 7) 10 Response to AEC Question 9.12 of 8/11/72 (FSAR Supplement 12) e l V '
1
% V b l
t 4 IE BULLETIN 79-01B REV. 3 JAMES A. FITZPATIIICK tlUCLt'AR POtlER PIANT DOCKET NO. 50-333 Tm I
, ACCIDENT SERVICE CONDITIONS AND QUAL.lFICATION TEST REQUIREMENTS 1 ?
i l 1. INSIDE PRIMARY CONTAlletENT i Pa ramet e r Accident Service Condition Typical Qualificatdon Test Requirements Temperature ( F) Transients are given in References I, 2 and 3 Transient qualif ication test requi rement s a re and for the Drywell and Suppression Cimmber given in tatular form in Ref erence 4 and justi-i Pressure (Pstg) (See Attactusent 1) fication for these values is provided in Ref e-rence 5. (See Attachment 2) 1 i Relative Humidity 100% 1001 (See Attactement 2) j (%) 9 Chemical Spray Not applicable. Demineralized water sprays in 8 See Ref erence 5 for typical equignent quellfication the containment and suppression clamber may be tested for sprays. manually actuated by the operator Imt are not an engineered safety feature. Radiation Radiation accident service conditions not Qualification test requirements are given in calculated in FSAR Reference 4 (See Attachment 2) Sutasergence Drywelli Approxiantely to EL. 261' tased on Not required in FSAR centerline of vents (Reference 7 ) Suppression clamber EL. 244'6)" (Reference 6 ) i j I1. OUTSIDE PRIMARY CONTA1NMENT i' Pa ramet e r Accident Service Condition Typical Q mtification Test Requisements Temperature ( F) Steam Tunnel: Pressure and Temperature t ran. Transient qualificat ion test requi rement s f or the HSIV i and sients given in tabular form in Reference 8 are given in tabular form in Ref erence 4 (See Attach-Pressure (Psig) reagense to item 20 ment 2). Crescent Area: 130 F. 0.1 psig - given in Acceptable wittu>ut qualification tests, see Reference 8 Reference 8. response to item 20 response to items 18 and 13. Seg* ration and tenundency of safety related equipent considered sufficient. I I } l
} !
l i 1- i j - , l J ,$ s t-- e
- IC DULLETIN 79-01B REV. 3 JAMES A. FITZPATRICK HUCLI AR POtJER PLANT TAB 12 1 (Continued)
DOCKET NO. 50-333 p t ~. ! 4 d it. OUTSIDE PRIMARY CONTA1191Elff V f Pa ramet e r Accident Service Condition Typical Qualiftration Test Requi rement s [ l* l Relative Humidity 1001 Same as for temperature and pressure in Crescent Area j (1) l l e Chemical Spray Not Applicable Not Applicable l s 1 Radiation Radiation ' accident service conditions not Same as for temperatuse and pressure in Crescent Area f calculated in FSAR Sutmergence Crescent Areat EL. 237'9 on one side of the Same as for temperature and pressure in Ceescent Area I water tight hulkhead only (either side, Reference 9. '._ Screenwellt EL. 255', Reference 'O. Safety equipment in screenwell ateve flood level. t i I 8 p I
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TMI Mod. No. FI-80 28 NUREC 0737. Sectlen ll.E.4.1.2 1E BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION SECTION 9 JAVES A FITZPATRICK NUCLEAR POWER PLA?lT WORK SHEET po 70 00cKET NO.50-333 oF ENVIRONMENT + D OCUM E NTATION REFERENCE QU AllFIC ATIO N OUT STANDING EQUIPMENT OESCRIPTION SPECIFICATION QUALIFICATION METHOD ITEMS PARAMETER SPECIFIC ATIO N l QUALIFICATION l OPER ATIN G yiMg 180 days >l80 Jays 101 222,143 Simultaneous Eme SYST E M PCACS PLANT 10 NO 27MOV-122 27MOV-123 T Eurt R TuR E COMPONENT Valve Operator (Order #3C1052A) PRESSURE PSIA I4.7 15.0 25 44 222.143 Simultaneous None MANUFACTURER Limi torque RELATIVE M ODEL NO. ! SMB-l NUMIDITY % 20-90 90-100 100 44.167 222 A N %m % F UNCTION lsolation Valve OEMNER ALIZE D WATER SPRAY ( Not Applicable D M O'N . T. l . D . AC CIDEN T SE RVICE: Primary Containment Isolation & Cont. Purgin, R ADI ATION 7 8.2x10g 6.4x10g 1.0x10 3 222,143 Sequential S me Rads LOCATION. Reactor Building A9lNe 40 years >40 years 164 222.143 Sequential None 2 - 3 7* 7 FLOOD LEVEL ELEV 227'-6" SU8 MERGENCE Not Applicable - - ABOVE FLOOD LEVEL YES E NVIRONME NT 1. Harsh accident environment.
- 2. Required after a containment LOCA for containment isolation and containment purging.
NO SPECIFICATION 3 Qualify to the LOCA environment as experienced in the component Reactor Bldg. locations. BASIS Installed Feb., 1982 PASNY P.O. #81-1390 AglNg Valve Serial No's PROGRAM 27MOV-122 #286560 INPUT 27HOV-123 #286561 OT E S: N gF NCE SECTION TID.(TOTAL INTEGRATED DOSE le 40 VR. NORM AL + ACCIDEN T AL 705 x Rev. O N /A NOT A PPLIC ABLE 1.8x10%
THI Mod. No. FI-81-03 bAME A T ATR CK JUCLEAR POWER PLANT SYSTEM COMPONENT EVALUATION SEC IOtJ DOCKET NO.50-333 WORK SHEET PG 57 OF ENVIRONMENT + DOCUME NTATION REFERENCE QU ALIFIC ATIO N OUTSTANDING EQUIPME NT DESCRIPTION METHOD ITEMS PARAM ETER SPECIFICATION QUALIFICATION SPECIFICATION QUALIFICATION OPE R ATING TIME 24 hours >24 hours 101 203 Simultaneous None SfstEM RCIC NORMAL AC NT PLANT so NO 13TS-1 h0-104 II0 212 44.104 203 Simultaneous None COuPONENT. Trip Solenoid 8 ^ * * * *" " MONUFAC TURER Trombetta (Terry Turbine) , M ODEL NO: c206 20-90 90-100 100 44,167 203 Simultaneous None F UNCTION Turbine Control DEMNER ALIZE D / N WATER SPRAY N Not Applicable g M O'N . T. I . D . AC CIDE NT i SE RVICE: RCic T bine Trip 6 7 Rads g , g6 gam 3 m SymM N LOCATION: Reactor Building West Cre. scent, RCIC Turb. ASING El. 233 40 years 40 years 164 203 Sequential None FLOOD LEVEL ELEV. 227'-6" SU8 MERGENCE Not Applicable ABOVE FLOOD LEVEL E NVIRONMENT l. liarsh accident environment. NO SPE CIFICATION BASIS ASING Installed Feb. ,1982 PROGRAM PASNY P.O. #81-3397 INP UT Rev. 0 SN EF NOE SE CTION T f D. ( TOTAL INTEGRATED DOSE )e 40 VR. NORM AL + ACCIDENT AL DOSE N /A NOT A P PLIC ABL E 7x10 0
TMt Mod. No. F1-81-03 NUREG 0737, Section ll.K.3.13.8 IE BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION sECTioN '4 JAMES A FITZPATRICK NUCLEAR POWER PLANT pc sa DOCKET NO.50-333 WORK SHEET of i
' l ENVIRONMENT + DOCUM E N TATION REFERENCE QU ALIFIC ATIO N OUTSTANDING l EQUIPME NT DESCRIPTION METHC0 ITEMS PARAM ETER SPECIFIC ATION QUALIFICATION SPECIFICATION QUALIFICATION f
i OPERATING ,M 180 days 101 TIME SYS T E M RCIC PLANT ID NO 13A-K49 T E M PE R TUR E '* eF 40-120 ,4 44,370,373 COMPONENT DC Contactor PRES $URE 14.7 N/A PSIA 44'170'373 4 MANUFACTURER General Electric RELATIVE 10-60 N/A 44,170,t73 g MODEL NO. : IC2800-1607 HUMIDITY % ($g7, 3} F ONC TION p, pay i DEMINER ALIZE D WATER spray Not Applicable DEMON. T. I . D . A CCIDE N T f SE RVICE; Various R ADI ATION 700 N/A 3 g_ Rads LOCATION. Control Room / Relay Room Admin Bidg. El. 300'/286 ASING 40 years 164 s. FLOOD LEVEL ELEV N/A SUBMERGENCE ( Not Applicable A8 0VE FLOOD L E VI'L h E NVIRCNMENT SPE CIFICATION
- 1. Mild environment go BASIS AGING Installed Feb., 1982 PROGRAM l PASNY P.O. #82-2814 gwpyy i
I PCE SECTION T I D. ( TOTAL INTEGRATED DOSE )s 40 VR. NORM AL + ACCIDEN TAL DOSE t Rev. O EF l N /A NOT A PPLIC ABLE
d J Section 14, Page 58
- 1. This component will be qualified under an existing qualification program which is testing Motor Control Centers for a harsh environment.
- 2. At the time of purchase, no existing qualified DC contactor could be identified. Therefore, a component which is already used in other 9
Class IE applications was selected for interim use. This component is considered the "best available" at the time of installation. Schedule: MCC testing presently scheduled for completion by December, 1982 or verify existing information is adequate for the mild environment in which this device ! l is located. l O O l I i
i TMt Mod. No. F1-81-05 l NUREG 0737, Section ll.K.3.22.B l c BULLETIN 79- 01B SYSTEM COMPONENT EVALUATION SECTION '4 l LAMES A FITZPATRICK NUCLEAR POWER PLANT WORK SHEET pc s9
)OCKET NO.50-333 OF
! ENVIRONMENT + DOCUMENTATION REFERENCE QU ALIFIC ATIO N OUTST ANDING l t DESCRIPTION METHOD ITEMS PARAMETER SPECIFI C ATIO N QUALIFICATION SPECIFICATION QUALIFICATION l EQUIPMENT
~
OPE R % TIN G i TlWE 2 hours 24 hours 101 216 Simultaneous None SYSTEM RCIC l NORMAL A i IgNT PLANT tD NO LS-076A,B LS-077A,8 T EM PER TUR E COMPONENT Level Swltch PRESSURE 14.7 N/A N/A
- 216 Simultaneous None MONUF ACTU RE R Hagnetrol PSIA M OD E L NO.:BCS-751 -EP/vPx-S lHD400 uM oi y 40-90 N/A N/A
- 216 Simultaneous None F UNC TION Suction Valve Control DEMMER All2E D / Not Applicable WATER SPRAY N D MON. (L Diff) T. 1. 0 . ACCIDE N T SE RVICE: Storage Tank Level g Switchover to Torus R ADI ATION 7x10 'N/A 1.0x10 7 3 216 Sequential None Rads LOCATION: Condensate Tank Pit El. 262' ASING 40 years 40 years 164 216 Sequential None FLOOD LEVEL ELEV. N/A SUBWERGENCE / Not Applicable )
l OBOVE FLOOD LEVEL E NViR ONME NT 1. Mild Environment go SPECIFICATION BA$l8 l AGING PROGRAW Installed Feb., 1982 INPUT . PASNY P.O. #81-4801 T.lO (TOTAL INTEGRATED DOSE )s 40 VR, NORM AL + ACCl0EN T AL DOSE EF NCE SE CTION Rev. O N /A NOT A PPLIC ABLE
TMt Mod. No. F1-81-05 NUREG 0737. Section II.K.3.22.8 IE BULLETIN 79-018 'YSTEM COMPONENT EVALUATION JAVES A FITZPATRICK NUCLEAR POWER PLANT SEC1 TON 14 DOCKET NO.50-333 WORK SHEET pc 60 OF ENVIRONMENT + DOCUME N TATION REFERENCE QU AllFIC ATI O N OUT ST ANDING I EQUIPMENT DESCRIPTION METHOD ITEMS l PARAM ETER SPECIFI C ATION QUALIFICATION SPECIFICATION QUALIFICATION OPER A TIN G 24 hours >24 hours 301 208 N/A None TIME SYSTEM RCIC PLANT ID NO 13PNS-055A B
" 40-120 N/A N/A
- 208 Simultaneous None COMPONENT.
Position Switch PRESSURE MONUFAC TURE R NAMCO PSIA 14.7 N/A N/A
- 208 Simultaneous None R E L ATIVE N/A N/A M OD E L NO. : EA740-80100 40-90
- 208 Simultaneous None (wi th Conau Condui t Seal HUMIDITY %
Assembly) FUNCTION Suction Valve Position DEMINER ALIZE D WATER SPRAY Not Applicable 9 D M o'N. T. 1. D . ACCIDE N T SE RVICE: Storage Tank Level Switchover to Torus RA TION 7x10 N/A l.0x10 3 208 Sequential None L OC ATiON. Condensate Tank Pit El. 262' AetMG 40 years 40 ) ears 164 208 Sequentlal None FLDOD LEVEL ELEV N/A SUBMERGENCE ( Not Applicable ---- - - --
)
ABOVE FLOOD LE VE'L E NVIR ONME NT 1. Mild Environment NO SPECIFICATION B A SIS AGING Installed Feb., 1982 PASNY P.O. #81-4316 INPUT NCE SECTION T I D ( TOTAL INTEGRATED DOSE le 40 YR. NORMAL 4 ACCIDENTAL DOSE Rev. O F N /A NOT A PPLIC ABLE
THI Hod. No. F1-80-16 NUREG 0717, Sec tion li.F.I.3 f, ll.E.4.2.7 UCLEAR POWER PLANT SYSTEM COMPONENT EVALUATION SECTION M hAME A TZ ATR CK WORK SHEET pg i3 or DOCKET NO.50-333 ' + DOCUM E N TATION REFERENCE ENVIRONMENT QU ALIFIC ATIO N OUTS T ANDING EQUIPMENT DESCRIPTION QUALIFICATION METHOD ITEMS PARAMETER SPECIFI C ATIO N QUALIFICATION SPECIFICATION T E >l80 days > 180 days 101 214 S&multaneous None SYSTEM Primary Containment System NORMAL AC{D NT PLANT ID NO-X-100D TE M PE R TUR E op 135-150 308 340 4. 57 214 Simultaneous None COMPONENT Electrical Penetratlos PRES $URE gg,y.16.7 PSfA 59.7 90 44.58 214 Simultaneous None MJ.N UFA ci d RE R Conax 44,57,58 MODEL NO.: 7251-20000-01 40-90 100 100 214 Simultarcous None NUM OITV*4 FUNCTION Primary Containment " Electrical Penetration Yes Yes 106 214 Simultaneous N<me R P D 4 0'g. N/A T. 1. D . A C CIDE N T SE R VICE ' Instrumentation reed-thru R ADI ATION g g 8 LCCATION: Drywell El.290' AGING 40 years 40 years 164 214 Sequential None FLOOD LEVEL ELEV 258'-6" SusMERGENCE Not Applicable ABOYE FLOOD LEVEL h E NVIRONME NT 1. liarsh accident environment. gp
- 2. Equipment required post-LOCA for communication of electrical instrument signals from containment sensors.
B A SIS AGING PROGRAW Installed Feb., 1982 INPUT PASNY P.O. #80-EF NCE SE CTION T 1.D ( TOTAL INT EGR AT ED DOSE )e 40 YR. NORM AL + ACCIDEN T AL DOSE Rev. O N /A NOT A PPLIC ABL E 8.5x!06
TH! Mod. No. F1-80-16
! IE BULLETIN 79-018 NUREG 0737. Section ll.F.I.3 s li.t.4.2.7 JAMES A FITZPATRICK NUCLEAR POWER PLANT SYSTEM COMPONENT EVALUATION SEcTION 16 DOCKET NO.50-333 WORK SHEET pc iu o ,-
ENVIRONMENT
+ DOCUME NTATION REFERENCE OU ALIFIC ATI O N OUTSTANDING EQUIPMENT DESCRIPTION PARAM ETER SPECIFI C ATIO N QUALIFICATION SPECIFICATION QUALIFICATION METHOD ITEMS OPER ATIN G TIME 180 days >l80 days 101 223 Simultaneous None SYSTEM Ceneric eak
{ 135-150 308 340 44.57 223 Simultaneous None T[RTuRE COMPONENT Coaxial Cable PRESSURE M ANuFACTURE R Rockbestos PSIA Ik.2x16.7 59.7 118 7 44.58 223 simultaneous None MODEL NO' R55-6-104-1981 , , ' ,4 40-90 100 100 44.57.58 223 simultaneous None F ONCTION Instrument Cable OEM *':R ALIZE D WATER SPRAY yes YES M 223 d b"" D MON ' T . 1. D . ACCIDE N T SE RVICE: Various Signal R AD A ON 1.09x10 0 1.0x108 2.0x10 3 223 Sequential Em L OC ATION Drywell AtlNG 40 years >40 years 164 223 Sequentlal None FLOOD LEVEL ELEV 258'-6" SusMERGENCE Not Applicable - ABOVE FLOOD LEVEL @ E NVIRONME NT 1. 2. Harsh accident environment. Equipment required to provide display instrumentation signal for high containnent radiation. Also NO SPECIFICATION provides a primary containment isolation signal. 8 A SIS 3 Equipment required for post-accident nnnitoring. AGING PROGRAM Installed Feb., 1982 INPUT EF NCE SE CTION T.l D. ( TOTAL INTEGRATED DOSE le 40 VR. NOAMAL + ACCIDENT AL DOSE ev. O N /A NOT A P PLIC A8 L E 6 8.5x10
w w ""' b a r..u=uss - - - - :
<- e TMl Mod. No. F1-80-16 NUREG 0737. Section ll.F.1.3 & li.I 4.L 7
*E BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION 16 JAMES A FITZPATRICK NUCLEAR PO.WER PLANT SECTION DOCKET No.50-333 WORK SHEET PG 11 or E N VI R O NMO 4 T . + DOCUM E NTATION REFERENCE QU ALIFIC ATIO N OUTSTANDING .
~ EQUIPMENT DESCRIPTION HETHOD ITEMS PA RAM ETER SPECIFI C ATIO N QUALIFICATION SPECtFICATION QUALIFICATION O ATING Slieul t aneous 180 days ;>l80 days 101 218 Analysis None
- CT/S T E M Generic .
NORMAL AC{lD NT > PLA'? ID NO N/A i T E R TU8L E 40-120 N/A 385 44,170,171 H8 Simultanema None COMPO4 NT Switchboard Wire I f PRESSURE MANUFA CTURE R Champlain pgga 14.7 N/A 127.7 44,170,171 218 Simultauecos None (General Atomic) -
, MODEL 4: ,osH '32 71 vw-l 10-60 N/A 100 44,170,171 218 Simultarcous Nor.c 12 "C , 600V - NUMIDin %
PU DN Control Wiring
)
DEMNER AUZE D WATER SPRAY ( -- - Not App?Icable - o g og, " ^ .s.D. A C CIDE NT I
; SERVICE: Various -
i "
- DI 10N g , Kp N/A 2.0=100 3 218 Sequential None
^ ^
'-' ~
LOCATION. ' ,. / Control Room / Relay Room Admin Bldg. El. J00*/286 heiNo 40 yea p 40 years 164 Seqwn t i a l Nonc f/ _ j R18 1 _ . FLOOD LEVEL CLEY ' gjA SU8 MERGENCE Not Applicable -
,D ABOVE FLOOD LEVFL @ E Nv40*iME NT 1. Hild E.tvironment go SPECIFICATION 848tS AGING PROGRAhl installed Feb., 13f12 INPUT EF NCE SECTIC?f T L D ( TOTAL INT E GR AT ED DOSE )e 40 VR NORMAL 4 ACCIDENTAL DOSE Rev. O N /4 NOT A PPLIC ABLE
. O TH Mod No. F1-80-01 NUREG 0737. Section ll.D.3.I IE BULLETIN 79-OIB SYSTEM COMPONENT EVALUATION 16 JAVES A FITZPATRICK NUCLEAR POWER PLANT SECTION DOCKET NO.50-333 WORK SHEET pc 12
_ OF ENVIRONMENT + DOCUM ENTATION REFERENCE QU ALIFIC ATIO N OUTSTANDING EQUIPMENT DESCRIPTION METH9D ITEMS PARAMETER SPECIFI C ATI O N QUALIFICATION SPECIFICATION QUALIFICATION OPER ATING TA0 days TIME 24 Hours 101 165 Simultaneous None SYS T E M Nuclear Boller PLANT ID NO- Generic COMPONENT Instrument Cable PRESSURE M1NUFACTU RE R Eaton PSIA 14.7 25,7 139,7 44,2 B A 165 Simultanenus None REL E MODEL NO. : 1952-68310 20-90 90- 00 100 .I 7 15 Si m itaneous None F U NC TION Instrument Signal DEMNERALIZED < NOT APPLICABLE - >- WATER SPRAY DEMON. N/A T. 1. D . ACCIDEN T SERVICE; Relief Valve Position Indication R ADI AT10N 6 RADS 6.9x10 6.8x10 6 2.0x10 3 Sequentlal 165 None LOCATION: Reactor Bldg - Generic AtlNG 40 Years >40 years 164 Sequential None 165 FLOOD LEuEL ELEY 227'-6" SUBMERGENCE e NOT APPLICABLE - >- A8OVE FLOOD LEVEL h E NVIRONME NT 1. Harsh accident environment. yo SPECIFICATION BASIS AGING PROGRAM Installed Dec., 1980 INPUT PASNY P.O. #79-17009 OM 4
+ REF NCE SE CTION T I D. ( TOTAL INTE9 RATED DOSE le 40 Vp. NORMAL + ACCIDEN TAL DOSE N /A NOT A PPLIC ABL E 7x10
rzr= . V)
/
TM1 Mod. No. F 1-80-01 NUREG 0737, Section 11.0.3.1
- E BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION SECTION 16 LAMES A FITZPATRICK NUCLEAR POWER PLANT WORK SHEET
)OCKET NO.50-333 pc i3 or ENVIRONMENT + DOCUMENTATION REFERENCE QU AllFIC ATIO N OUT ST ANDING EQUIPMENT DESCRIPTION METHOD ITEMS PARAM ETER SPECIFI C A TIO N QUALIFICATION SPECIFICATION QUAllFICATION OPER ATIN G TIME 24 Hours 80 days 101 165 Simultaneous None SYSTEM Nuclear Boller NORMAL AC "
rat 3 PLANT ID NO Generic T PE R TUR E 135-150 308 340 44.57 165 Simultaneous None
,CouPONENT Instrument Cable M ANUFACTURE R faton 3 A U. ' Itaneous Nne RE MODEL NO: 1952-68310 100 44,57,58 U DITV '/o 40-90 100 165 Simultaneous None F UNCTION instrument Signal WATER SPRAY Yes yes 106 165 Simultaneous None Mo[A T.1. D . ACCfDENT SERVICE: Relief Valve Position Indication R ADI ATION l.09x10 8
1.0x10 8 2.0x10 g 3 RADS 165 Sequential None LOCATION; Drywell (SIx m nths) 40 Years >40 years 164 165 Sequential None AtlNG l I FLOOD LEVEL ELEV 258'-6" SUBMERGENCE e NOT APPLICABLE - # 1 ABOVE FLOOO LEVEL h E NVIRONME NT 1. Harsh accident environment. NO SPE CIFICATION B A SIS AtlNG Installed Dec., 1980 PROGRAM PASNY P.O. #79-17089 gypyy T.I D. ( TOTAL INTEGRATED DOSE le 40 VR, NORMAL 4 ACCIDENTAL DOSE REF NCE SECTION N /A NOT A PPLIC ABLE 8.5xt0
TM1 Mod. No. fI-81-47 LAME A TZ ATR CK JUCLEAR POWER PLANT SYSTEM COMPONENT EVALUATION SECTION 16
)OCKET NO.50-333 WORK SHEET po _ 14 oF ENVIRONMENT + DOCUMENTATION REFERENCE QUALIFIC ATI O N OUTST ANDING DESCRIPTION PAR AM ETER SPECIFICATION QUALIFICATION SPECIFICATION QUALIFICATION METHOD ITEMS l EQUIPMENT ATINS 180 days >180 days 101 201 Simultaneous None SYS T E M Ceneric ea T EMPE R TUR E 40-104 292 385 44,1B0 201 Simultaneous None
*F
' CouPONEN T Instrument Cable MONUFACTURER Anaconda 8 8
- i M ODEL NO.
- FR EP 20-90 90-100 100 44,167 201 Simultaneous None F UNCTION Instrument Cable OEMNER ALIZE D j WATER SPRAY N Not Applicable )
M ON. T . 1. D . ACCIDE N T R ADI ATION 6.87x10 6.8x10 2.0x10 3 201 Sequential None Rads LOCATION: Reactor Bldg-Generic A0 LNG 40 years >40 years 164 201 Sequential None FLOOD LEVEL ELEy 227'-6" SUBMERGENCE Not Applicable ) C8 0VE FLOOD LEVEL E NVIRONME NT 1.' Harsh accident environment.
- 2. Qualified for generic Reactor Bldg. enveloping environnnnts for LOCA/HELB.
g , BASIS AGING Install Dec., 1982 ROGRAM INPUT ev. O NS yfREF NCE SECTION T.10. ( TOTAL INTEGRATED DOSE le 40 VR. NORMAL + ACCIDENTAL DOSE l N /A NOT A P PLIC ABL E 7x10
=aw -
IMI Mod. No. F1-81-47 UCLEAR POWER PLANT SYSTEM COMPONENT EVALUATION S cTION 16 JAME A TZ ATR CK WORK SHEET accKET NO.50-333 PG ls or ENVIRONMENT + DOCUME N TATION REFERENCE QU ALIFIC ATI O N OUTSTANDING I DESCRIPTION METHOD ITEMS l EQUIPME NT PARAM ETER SPECIFICATION QUALIFICATION SPECIFICATION QUALIFICATION O ATING 180 days N/A 101 N/A N/A N/A SfSTEM Ceneric NORMAL AC CIDE N T PLoNT D NO N/A TE M PE R TUR E 40-120 N/A N/A 44,170.171 209 Simultaneous None
'F C OM PONE N T Instrument Cable PRESSURE 14.7 N/A N/A 44,170.17I 209 Simultaneous None MONUFACTU RE R Belden PStA NODEL NO. : C/N 83348 10-60 N/A N/A 44.170,171 209 Simultaneous None N DI Y '4 a
FUNCTION Panel instrument Cable CEMINERALIZED e WATER SPRAY ( lbt Applicable ) I GCCURACV- SPEC. N/A DE M O't T. I . D . ACCIDE N T f RAD TION N/A 1.7x10 700 3 209 Sequential None LOCATION: Control Room Admin. Bldg. El. 300' A0 LNG 40 years >40 years 164 209 Sequential None FLOOD LEVEL ELEV N/A SUBMERGENCE Not Applicable
)
AB OVE FLOOD L EVE'L h E NVIR ONME NT SPECIFICATION
- 1. Hild Environment NO B A SI S ASING PROGRAM Installed Feb., 1982 INPUT NCE SE CTION T.I D. ( TOTAL INTEGRATED DOSE )= 40 VR, NORM AL + ACCIDENT AL DOSE Rev. O EF N /A NOT A PPLIC ASL E
TMI M0J. No. FI-Generic hAME A T ATR CK UCLEAR POWER PLANT SYSTEM COMPONENT EVALUATION SECTION 16 30CKET NO.50-333 WORK SHEET PG '6 or ENVIRONMENT + DOCUME N TATION REFERENCE QU ALIFIC ATIO N OUTSTANDING EQUIPME NT DESCRIPTION METHOD ITEMS PARAMETER SPECIFICATION QUALIFICATION SPECIFICATION QUALIFICATION ME 180 days >180 days 101 205 Simultaneous None SYS T E M Generlc NORMAL ACCIDENT PLANT iD NO N/A T PE R TUR E 40-120 N/A 3f6 i 44,170,171 205 s lmul t aneous None C OM PON E N T. Switchboard Wire PRESSURE MONUFACTURE R Rockbestos PSIA 14.7 N/A 127.7 44,170,171 205 Simultaneous None RELATIVE MODEL NO. : Firewall SIS 80-60 N/A 100 44,167 UMIDITV '4 205 Simultaneous None F U NC TION Control Wlring DEMNERALIZED WATER SPR4Y Not Applicable
}
D MON. T. l . D . AC CIDE N T SE RVICE: Various RAD Afl0N
., 700 N/A 2.0x10 3 205 Sequential None LOCATION:
Control Room / Relay Room A0 LNG 40 years >40 years 164 205 sequential None Admin Bldg. El.300'/286' FLOOD LEVEL ELEV N/A SU8 MERGE NCE Not Applicable
-)
08 0VE FLOOD LEVEL E NVIRONME NT l. Hild Environment NO SPE CIFICATION BASIS AGING Installed Feb., 1982 PROGRAM INPUT Rev. O N EF N /A NOT A PPLIC ABLE NCE SECTION T l D. ( TOTAL INTEGRATED DOSE l a 40 VR. NORM AL + ACCIDENT AL DOSE
O TMI Hod. No. Fl-Generic
"" #37 SYSTEM COMPONENT EVALUATION SECTION 16 JAME A T PATR CK UCLEAR POWER PLANT WORK SHEET '7 OOcKET NO.50-333 PG OF ENVIRONMENT + D OCUM E NTATION REFERENCE QUALIFIC ATIO N OUTSTANDING EQUIPMENT DESCRIPTION NETHOD ITEMS PARAMETER SPECIFI C ATIO N QUALIFICATION SPECIFICATION QUAllFICATION OPE R ATIN G TIME 180 days >I80 days 103 205 Simultaneous None SYSTEM Generic P'eak PLANT :D NO N/A T PE R TUR E 40 104 292 346 44.lBD 205 simultaneous None COMPONENT: Cable 8
- MANUFACTURER Rockbestos '
t R ELATIVE l MODEL NO: Firewall 111 0-90 90-100 100 44.167 205 simultaneous NUMIDITY % None F ONC TION Control Cable OEMINER ALIZE D WATER SPRAY
/
N Not Applicable
)
EMON. T. B . D . A CCIDE N T
"!RVICE: Various R ADt ATION 6 6 8 Rads 7x10 6.8x10 2.0x10 3 205 sequen t l a l None LOCATt0N. Reactor Bldg-Generic A0 LNG 40 years 40 years 164 205 sequential None FLOOD LEVEL ELEV. SUBMERGENCE 227'-6.. Not Applicable h
l ABOVE FLOOD LEVEL E NVIR ONME NT 1. Harsh accident environment gg $PECIFICATION B A SIS ASING Installed Feb., 1982 PROGRAM INPUT NOT E S: ST Rev. O REF NCE SE CTION T I D ( TOTAL INTEGRATED DOSE )e 40 VR. NORM AL + ACCIDENT AL DOSE N /A MOT A P PLIC AB L E 4
TMl No.8 No. F1-Generic T N L M ON SECTION 16 JAME A TZ ATR CK JUCLEAR POWER PLANT DOCKET NO.50-333 WORK SHEET pg la or 1 ENVIRONMENT +00CUMENTATION REFERENCE QU ALIFIC ATIO N OUTST ANDING EQUIPMENT DESCRIPTION SPECIFICATION QUALIFICATION M E T l40 D ITEMS PARAM ETER SPECIFICATION QLIALIFICATION 180 days >l80 days 101 212 Simultaneous None o[ ATING SYSTEM Generic eak g TEM PE R TUR E 40-104 292 300 44,18D 212 Simultaneous None
'F
!+ COMPONENT. Cable PRESSURE M ANUFA CTURE R Boston Insulated PSIA 14.7 25 7 64.7 44.2BA 212 Simultaneous None Wire R T1VE M ODEL NO: 14538-H-006 (XLPE ) 20-90 90-100 100 44,167 212 Simultaneous None g 9,,y , F ONCTION Signal Cable WATER SPRAY D M O'N . T. 1. D . ACCIDE N T SE RVICE: Various f RA TSON 6.87x10 6.8x10 1.0x10 3 212 Sequential None LOCATION: Reactor Bldg.-Generic ASIM9 40 years >40 years 164 212 Sequent lal None FLOOD LEVEL ELEV 227'-6" SUBMERGENCE Not Applicable ABOVE FLOOD LEVEL E NVIRONME NT 1. Harsh accident environment. NO SPE CIFICATION BASIS AGING Installed Feb., 1982 PROGRAM INPUT NOTEL g N T.l D. ( TOTAL INTEGRATED DOSE ) m 40 VR. NORM AL -+ ACCIDEN T AL DOSE Rev. O tF NCE SECTION N /A NOT A PPLIC ABLE 7x10
TMl Mod. No. F1-81-03 NUREG 0737, Sect ion ll.K.3.13.8 IE BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION 16 JAMES A FITZPATRICK NUCLEAR POWER PLANT SECTION DOCKET NO.50-333 WORK SHEET pc in Or ; i ENVIRONMENT + DOCUM E NTATION REFERENCE QU ALIFIC ATIO N OUT ST ANDING EQUlPNENT DESCRIPTION METHOD ITEMS PARAMETER SPECIFI C ATIO N QUALIFICATION SPECIFICATION QUALIFICATION OE '"' y E 2 hours 110 Days 101 219 Simultaneous kw SYSTEM RCIC COMPONENT. Cable
. 1 . .BA 19 $1multaneous &w M ANUFACTURE R General Electric $ A R EL ATIVE M OD E L NO : B.Vulkene Flame Resis- 20-90 90-100 100 44,167 219 Simultaneous nie tant Control Cable 1000V HUMIDITY */o C 8AWG DEMINER ALIZE D WATER SPRAY Not Applicable h
M dN. T . 1. D . AC CIDE N T i SE RVICE: RCIC Turbine Trip 6 6
""D',"' "
1.17x10 1.tx10 2.2x10 3 219 sequential nw ds l LOCATION: Reactor Building - A0 LNG 40 years >feo years 164 219 sequential N< w l FLOOD LEVEL ELEV. 227'-6" $UBMERGENCE ( Not Applicable ABO #E FLOOD LEVE'L E NVIRONME NT 1. Harsh accident environment. NO SPECIFICATION BASIS AGING Install Dec., 1982 PROGRAM PASNY P.O. SW88C-120 tupuy REF NCE SE CTION T 10. ( TOTAL INTEGRATED DOSE le 40 VR. NORM AL 4 ACCIDENTAL DOSE Rev* 0 N /A NOT A PPLIC ABLE 7x10 l
1MI Hod. No. F1-Generic NUREG 0737 iE BULLETIN 79-010 SYSTEM COMPONENT EVALUATION JAMES A FITZPATRICK NUCLEAR POWER PLANT SECTION 20 DOCKET NO.50-333 WORK SHEET pc 8 og ,
' ENVIRONMENT + DOCUM EN TATION REFERENCE QU AllFIC ATIO N OUTSTANDING EQUlPMENT DESCRIPTION METHOD ITEMS PARAMETER SPECIFICATION QUALIFICATION SPECIFICATION QUALIFICATION OPE A ATIN G 180 days N/A N/A N/A 101 N/A TIME SYSTEM Generic ,
e p TE ptRTURE 40-120 N/A 346 44,170,171 210 Simultaneous None op COMPONENT Fuse Block CENUFACTURER Buchanan SIA ' "^ ' ' ' ' $'*"#"' "$ " "* MODEL NO' NQO361006 fy 10-60 N/A 100 44,170,171 210 Simultaneous None F U NC TION C{rcuit Protection OEMINER ALIZE b j \ WATER SPRAY N Not Applicable / D MON. T. 1. D . ACCIDE NT SE R VIC E - Control Systems R ADI ATION 8 7 NA . 3 210 Sequential None Rads L OC A TION; Control Room / Relay Room Admin Oldg. Ei.300'/286' A83"8 40 years >40 years 164 210 Sequential Nnne FLOOD LEVEL ELEy N/A SU8 MERGE N CE Not Applicable
--)
08 0V E FLOOD LEVEL h E NVIRONME NT SPECIFICATION
- 1. Hild Environment un B A SIS AGING Installed Feb., 1982 PROGRAM INPUT REF NCE SE CTION T I D. ( TOTAL INTEGRATED DOSE ).40 VR. NORMAL 4 ACCIDENTAL DOSE Rev. O N /A NOT A PPLIC ABL E
s TH I Mod. No. F l-Gener i c IE BULLETIN 79-OIB SYSTEM COMPONENT EVALUATION NuRcc 0737 20 JAMES A FIT 2 PATRICK NtlCLEAR POWER PLANT SECTION DOCKET NO. 50- 333 WORK SHEET PG 9 OF ENVIRONMENT + DOCUM E NT ATION REFERENCE QU ALIFIC ATIO N OUTSTANDING l EQUIPMENT DESCRIPTION METHOD ITEMS l PARAM ETER SPECIFI C ATIO N QUALIFICATION SPECIFICATION QUALIFICATION 180 days N/A 101 N/A N/A N/A E SYST E M Ceneric PLANT #0 No N/A COMPONENT Terminal Block PRE 35URE PSIA 14.7 N/A 127 7 44,170,171 210 Simultaneous None MAD .e ACTURER Buchanan I M OD EL M5! NQBil2, NQB106 y. 10-60 N/A 100 44,170,171 210 Simultaneous None y o NQ511082 F UNC TION Wire Spilce OEMINER ALIZE D j WATER SPRAY N Not Applicable h ! MON. T. 1. D . AC CIDE NT SERVICE: Control Systems
"" *' N/A 2.0x10 210 Sequen t i a l None Rads " 700 3 LOCATION:
Control Room / Relay Room ASING 40 years 40 years 164 210 Sequential None , Admin Bldg. El. 300'/286' I FLOOD LEVEL ELEV N/A SUBMERGENCE - Not Applicable ) A20VE FLOOD LEVEL h E NVIRONME NT SPE CIFICAf t04
- 1. Mild Environment NO B A SIS AGING Installed Feb., 1982 PROGRAM INPUT Rev. O N[R EFE R NCE SECTION T 10. ( TOTAL INTEGRATED DOSE le 40 VR. NORM AL + ACCIDEN T AL DOSE N /A NOT A PPLIC ABLE
IMi Mod. No. F1-Generic NUREG 0737 E BU LLETIN 79- OlB SYSTEM COMPONENT EVALUATION 20 JAVES A FITZPATRICK NUCLEAR POWER PLANT SECTION
)OcKET NO.50-333 WORK SHEET pG 10 Or ENVIRONMENT + DOCUMENTATION REFERENCE QUALIFIC ATIO N OUTST AN DING E60lPMENT DESCRIPTION METHOD ITEMS PA RAM ETER SPECIFI C ATIO N QUALIFICATION SPECIFICATION QUALIFICATION ATIN S 180 days >l80 days 101 211 Simultaneous None SYS T E M Generic Pe k
. PLANT 10 NO N/A TE RME 44,lBD j 40-104 292 346 211 Simultaneous None
, COMPONENT-Terminal Block (In NEMA 4 Enclosure)
M ANUFACTURE R Buchanan PS A . M.7 127.7 M.M 211 U multanmus None RELATIVE M ODEL NO! NQBil2, NQB106, 44,167 NUMIDITY % 20-90 90-100 100 2tl Simultaneous None NQ511082 F U NCTION Wire Splice OEMNER ALIZE D / \ N Not Applicable
/*
WATER SPRAY D MON. T . 1. D . A CCIDE N T
% RVICE ~ Control Systems 6 6 8 g
Rads LOCATION. Reactor Bldg.-Generic AGING 40 years >40 years 164 Sequential gone 211 Flood LEVEL ELEY 22 7 e .6s e SUBMERGENCE Not Applicable ) ABOVE FLOOD LEVEL E NV'R ONME NT 1 Harsh accident environment. yo SPECIFICATION BASIS i AetNG Installed Feb., 1982 PROGRAM INPUT Rev. O
"* N $ g"F N NCE TlD.(TOTAL INTEGRATED DOSE )e 40 VR. NORM AL + ACCIDENT AL DOSE SECTION N /A NOT A PPLIC A8LE 7x10%
TMI Mod. No. F1-Generic SYSTEM COMPONENT EVALUATION SECTION 21 JAME AFT PATR CK JUCLEAR POWER PLANT DOCKET NO.50-333 WORK SHEET PG s or ENVIRONMENT + D OCUM E N TATION REFERENCE QU ALIFIC ATIO N OU TST ANDING
! EQUIPMENT DESCRIPTION METHOD ITEMS PARAMETER SPECIFIC ATION QUALIFICATION SPECIFICATION QUAllFICATION OPE R A TIN G g TIME 180 days 101 SYSTEM Generic Pc' PLANT ID NO' N/A T PE R TUR E 40-120 N/A 44.170,17f COMPONENT Circuit Breaker (Molded Case) g,3 14.7 N/A 44,170,171 M ANUFACTURE R Ceneral Electric M ODEL NO: TEBil1015 10-60 N/A 44,170,171 +
NuM DiTY */. F UNC TION Circult lSolation and DE MINER All2 E D Protection WATER SPRAY ( Not Applicable p D MOk T. f . D . AC CIDE N T SE RVICE : Various
" * "g",'d s 700 N/A 3 5
LOCATION. Control Room / Relay R e AelNG 40 years i Admin Bldg. El.300'/286' 164 FLOOD LEVEL ELEV N/A SUBMERGENCE __ ABOVE FLOOD LEVEL h E NVIRONME NT SPE CIFICATION
- 4. Hild Environment NO BA$1S l
AGING installed Feb., 1982 PROGRAld INPUT
+ R EF NCE SE CTION T f D. ( TOTAL INTEORATED DOSE )e 40 VR. NORM AL + ACCIDEN T AL DOSE Rev. 0 N /A NOT A PPLIC ABL E
j () Section 21, Page 5 Component - Molded Case Circuit Breaker
- l. This component will be qualified under an existing qualification program which is testing motor control centers for a harsh environment.
s
- 2. The Authority is presently evaluating qualification information on similar circuit breakers from the same manufacturer.
3 This item is installed into an existing power distribution panel. The design of the distribution panel necessitates the use of a breaker consistent with the panel design. Identical breakers in the panel are already employed in a Class 1E application. l 4. At the time of purchase, no qualified circuit breaker meeting the t installation requirements was available. This item was considered the "best available" at time of installation considering its location in a mild environment. Schedule: Complete MCC testing by December 1982. e f i i
- l
lH1 M..d. No. F1-Cencric NUREG 0737 iE BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION SECTION 21 JAVES A FITZPATRICK NUCLEAR POWER PLANT 30CKET NO.50-333 WORK SHEET pc 6 OF , ENVIRONMENT + DOCUME N TATION REFERENCE QU ALIFIC ATIO N OUTSTANDING l EQUIPME NT DESCRIPTION PARAMETER SPECIFI C ATIO N QUALIFIC ATION SPECIFICATION QUALIFICATION METHOD ITEMS t OPER ATING x TIME 180 days 101 '
' system Generic Pr k T PE R TUR E 44,1BC l10 c0MPONENT (ircuIt Breaker (Molded Case)
[Rg E,S,SUR E 15.0 44 g 14.7 M ANUFACTURE R General Liectric M OD EL N0' THQB1120 20~90 90'i00 44 U lT 84 i F UNCTION Circuit ISolatlon and Protection OEMINER ALIZE D N Not Applicable _.
/
! WATER SPRAY D MON.
"^ T . 1. D . AC CIDE N T SERVICE: Various R ADI ATION h 3.4x10 3.3x10 3 Rads L OCATION- Reactor Bldg.-Generic for E l . 2 72' & 300' &
AGING 40 years 164 FLOOD LEVEL ELEy 227'-6" SUBMERGENCE / Not Applicable N h / ABOVE FLOOD LEVE'L E NVIRONME NT 1. Harsh accident environment SPECIFICATION
- 2. Quall cation provW for postWA envir .wnts in tk Reactor B14. on einations 2N f t. and 300 f t.
go B A SI S AGING PROGRAM Install Dec., 1982 INPUT Rev. 0 SN EF NCE SE CTION T ID ( TOTAL INTEGRATED DOSE le 40 VR. NORM AL + ACCIDENTAL DOSE k N /A NOT A PPLIC ABL E 7,to
4 Section 21, Page 6 () Component - Molded Case Circuit Breaker Installation incanplete
- 1. This breaker will be Installed in existing 120 VAC power distribution panels. The design of the distrubution panel necessitates the use of a breaker consistent with the panel design. Installation schedule completion by December 1982.
- 2. At the time of purchase, no quallfled breaker meeting the installation requirements was available.
3 Also investigating system design to determine if power source can be relocated to a distribution panel in a mild environment area. Schedule: Qualification issue or redesign by December 1982. O I ) O
, e- g w _- a _m n,w-mn, _ .- .n._n_- ...-- -. __..-,_.-. , , , ,,., , - _ ,-, y ---,,, ,- ,. , - - -,-
Section 26 - High Range Noble Gas Effluent Monitors NUREG 0737, item II.F.1.1
- 1. NUREG 0578 and later clarifying letters stated the design requi rements for a high range noble effluent gas monitoring system.
These documents stated that this system was not required to be qualified.
- 2. In an attempt to implement the THI Action Plan in the accelerated schedule required by the Commission, the Authority investigated various monitoring systems based on numerous technical aspects and most importantly schedule constraints. The JAF system, which was selected was procurred from the General Electric Company, consists of Victoreen radiation monitoring equipment and Leeds and Northrup recorders. This equipment has not undergone qualification testing.
It was at a later date upon the issuance of NUREG 0737 that the equipment qualification requirement was stated. 3 The Authority is investigating two possible methods of meeting the later qualification requirements: . 1) Conduct a qualification program on the components of the existing system (possibly joining other utilities who also have their equipment), or
- 2) Replace the equipment with a new system if suitably quallfled equipment is available.
- 4. A review of the design specifications of the Victoreen and Leeds
(, Northrup equipment, confirms that the equipment is being used in the manner and environment for which it was originally designed. This provides justification for its Interim use until its qualification can be confirmed. I I
- O
TMl Mod. do. F1-80-14 E BULLETIN 79- OIB **" SYSTEM COMPONENT EVALUATION 737' SE CTION"" F ' ' JAMES A FITZPATRICK NUCLEAR POWER PLANT WORK SHEET JoCKET NO.50-333 PG 1 OF ' ENVIRONMENT + DOCUMENTATION REFERENCE QU ALIFIC ATIO N OUTS T ANDING I EQUlPMENT DESCRIPTION SPECIFICATION QUALIFICATION METHOD ITEMS PARAMETER SPECIFI C ATIO N QUALIFICATION l (u "" '"* 180 days 101 3 NORMAL ACC ENT PLANT 8D NO- 17RE-053A,8 TEM PER TUR E f A-
*F 40-100 40-100 COMPONENT Radiation Detector PRESSURE ggg gg,7 g MONUFACTURER Victoreen (General Electric)
RELATIVE g g M OD EL NO.: gg7A_y 40-90 40-90 NUMIDITY % F uMCTION Hi Range Effluent DEMMER ALIZE D l Monitor \ WATER SPRAY Not Applicable 7 D M O'N . t T. 1. D . AC CIDE N T SERVICE; Post Accident . R ADl AT10N Monitoring of Stack Effluent p , 3 5x103 3.oxlo' 3 Cases LOCATION: Stack El. 282'-6" 40 years 164 FLOOD LEVEL ELEY N/A SU8 MERGENCE ( Not Applicable ) ABOVE FLOOD LEVEL E NVIRONME NT , NO SPECIFICATION 2. Equipment located in stack, so does not experience external acci(ent (LOEA) environments. BASIS AGING PROGRAM Installed Feb., 1982 INPUT PASNy P.O. #80-17553 EF NCE SECTION T.l D. ( TOT AL INT EGR AT ED DOSE ) e 40 T R. NORM AL + ACCIDEN T AL DOSE Rev. O N /A NOT A PPLIC ABLE 3 5xt03
TMI Hod. No. F 1 14 NUREG 0737, Section ll.F.1.8 JAME A F TZPATR CK JUCLEAR POWER PLANT SYSTEM COMPONENT EVALUATION SECTION 26 DOCKET NO.50-333 WORK SHEET PG 2 or ENVIRONMENT + DOCUME N TATION REFERENCE QU AllFIC ATIO N OUTSTANDING EQUIPMENT DESCRIPTION METHOD ITEMS PARAM ETER SPECIFICATION QUALIFICATION SPECIFICATION QUALIFICATION OPER ATIN G a TIME 180 days 101 SYSTEM Process Radiation NORMAL AC D NT PLANT D NO'17RT-053A,8 40-100 40-100 E
- COMPONENT Preampli fier PRESSURE gg,7 gg,7 g MANUFACTURER Victorcen PSIA (General Electric)
" M MODEL NO1 847A-1 40-90 40-90 f I M DI Y '4 F ONC Tl0N Hi Range Effluent Honitor DEMNERALtZED WATER SPRAY Not Appilcable -
)
D MON. m) T. I . D . ACCIDE N T SERVICE: Post Accident i Honitoring of Stack Effluent R ADI ATION , 3.0x10 g Gases Rads 3 56x10' 3 LOCAT404: Stack El. 282'-6"
" "' 40 years 164 FLOOD LEVEL ELEV. N/A SUBMERGENCE _ _ _ _
ABOVE FLOOD LEVEL h E NVIRONME NT
- 1. Equipment required for post-accident monitorinej of stack ef fluents.
yo SPECIFICATION
- 2. Equipment located in stack, so does not experience external accident (LOCA) environments.
BASIS l AGING PROGRAM installed Feb., 1982 lupuy PASNY P.O. #80-17553 Rev. O N REF ENCE SE CTION N /A NOT A P PLIC ABL E T I O. ( TOTAL INTEGRATED DCSE le 40 VR NORMAL + ACCIDEN TAL DOSE 3 5x103
1Fi Mod. No. F1-80-14 NURE G 0737, Sec tion ll.F.1.1 iE BULLETIN 7 9 - 010 SYSTEM COMPONENT EVALUATION 26 JAMES A FITZPATRICK NUCLEAR POWER PLANT SECTION 30CKET NO.50-333 WORK SHEET po 3 or ENVIRONMENT + DOCUMENTATION REFERENCE QU ALIFIC ATIO N OUTSTANDING I EQUIPME NT DESCRIPTION METHOD ITEMS l PARAM ETER SPECIFI C ATI O N QUALIFIC ATION SPECIFICATION QUAllFICAT ION OPER ATING a-TIME 180 days 101 SfSTEM Frocess Radiation Monitoring NORMAL ACCl ENT PLANT ID N0'I/RE-434A,B T E MPE R TURE 40-104 N/A 168 4
'F COMPONENT. Radiation Detector MANUFACTURER Victoreen (General Electric)
MODEL NO' 847A-2 HUM DITV '/o 20-90 N/A 168 ' F U NC TION Hi Range Effluent DEMINER ALIZED Monitor WATER SPRAY ( Not Applicable - ) M o'N , m T. 1. D . AC CIDE NT SERVICE: Post Accident R ADI ATION Honitoring of Turbine Building 765 65 3 Effluent Gases Rads LOCATION:HG Set Room El. 300'-0" (8Y) AGING 40 years 164 Y FLOOD LEVEL ELEV N/A SUBMERGENCE ( Not Applicable -
)
C8 0VE FLOOD LEVI'L h E NVIRONMENT
- 1. Equipment required for post-accident monitoring of Turbine Building Ventilation.
NO SPECIFICAT104
- 2. Equipment located remote from the area being monitored. Post LOCA-there will be a slight increase in BASIS radiation levels.
i A0149 PROGRAM installed Feb., 1982 INPUT PASNY P.O. #80-17553 - EE NCE SECTioN T.I D. ( TOTAL INTEGRATED DOSE 1 40 YR. NORM AL + ACCIDENT AL DOSE Rev. O N /A NOT A PPLIC ABL E 700
TMl Mod. No. fl-80-14 SYSTEM COMPONENT EVALUATION EcTi 26 JAME A T ATR CK UCLEAR POWER PLANT WORK SHEET pg 4 oF DOCKET NO.50-333 ENVIRONMENT + DOCUM E N TATION ftEF ERENCE QUALIFIC ATI O N OUTST ANDING EQUlPMENT DESCRIPTION SPECIFICATION QUAllFICATION METNOD ITEMS P RAMETER SPECIFI C ATIO N QUALIFICATION OPER ATIN G 180 days 101 TIME M orin9 NORMAL ACCIDENT PLANT ID NO 17RT-434A,B 40-104 N/A Ib8 COMPONENT Preampli fier PRESSURE #- PSIA 14.7 N/A M ANUFACTURE R Victoreen 168 (General Electric) RELATIVE " MODEL NO: 847A-1 HUMIDITY */, 20-90 N/A 168 FUNCTION Hi Range Ef fluent DEMNERALIZED / Monitor WATER SPRAY \ Not Applicable - > EMON. m) T . 1. D . ACCIDE NT SE RVICE: Post Accident Monitoring of Turbine Building R ADI ATION g Effluent Gases Rads 765 65 3 LOCATION: MG Set Room E L. 300 '-0" (8Y) w ASING 40 . years 16k FLOOD LEVEL ELEY N/A SUBMERGENCE Not Applicable -
)
A8OVE FLOOD LE VI'L E NVIRONME NT
- 1. Equipment required for post-accident monitoring of Turbine Building Ventilation.
NO SPECIFICATIOs
- 2. Equipment located remote from the area being monitored. Post LOCA-thtre will be a slight increase in B A SIS radiation levels.
l i ASING PROGRAM Installed Feb., 1982 INPUT PASNY P.O. #80-17553 REF NCE SECTION T 10. ( TOTAL INTEGRATED DOSE l s 40 YR. NORM AL + ACCIDENT AL DOSE Rev. O 700 N /A NOT A P PLIC ABLE -
. - ~ - - - -..
TMI Mod. No. F1-80-14 NURFG 0737, Sec tion ll.F.1.1 E BULLETIN 79-018 SYSTEM COMPONENT EVALUATION 26 JAMES A FITZPATRICK NUCLEAR POWER PLANT SECTION
)OCKET NO.50-333 WORK SHEET PG s OF E NVIR O NM ENT + DOCUM E NTATION REFERENCE QU ALIFIC ATIO N OUTSTANDING EQUIPMENT DESCRIPTION METHOO ITEMS PARAM ETER SPECIFICATION QUAllFICATION SPECIFICATION QU ALIFICAT ION OPER ATING _
TIME 80 days 101 H ori 9 NORMAL ACCIDE N T PL AN T ID NO 17RE-463A,8 l T ERTURE 40-100 N/A 44,169 COMPONENT Radiation Detector PRESSURE . MONUFACTURER Victorcen PSIA 14.7 N/A 44,169 (General Electric) M OD EL NO' 847A-2 20-90 N/A 44,169 NUM DITY */. F U NC TION Hi Range Effluent Honstor oEMmER ALIZE D / N
- \
WATER SPRAY / 0 M 0'N . m T. 1. D . ACCIDENT SERVICE: Pos t Accident Monitoring of Radwaste R ADI ATION I.ox10 3 3 's. Rads 75x103 Building Effluent Cases LOCATION: Turbine Building El. 300'-0" (232) AglNg 40 years 164 FLOOD LEVEL ELEV SUBMERGENCE Not N/A Applicable
)
ABOVE FLOOD LEVEL h E NVIR ONME NT
- 1. Equipment required for post-accident moni ng of the Radwaste Bldg. ventilation exhaust.
NO SPE CIFICATION
- 2. Equipment located remote from the area beli monitored. P)st LOCA-there will be no significant change BASIS in ambient conditions.
AGING PROGRAM Installed Feb., 1982 INPUT PASNY P.O. #80-17553
+ EFE NCE SECTION T ID (TOTAL INTEGRATED DOSEle 40 VR. NORN AL + ACCIDENTAL DOSE av. 0 N /A NOT A PPLIC ABLE 1750
TMI Mod. No. FI-80-14 NUREG 0737, Section li.F.1.1 JAME A T ATR CK Jt; CLEAR POWER PLANT SYSTEM COMPONENT EVALUATION SECTION 26 DOCKET NO.50-333 WORK SHEET po 6 or ENVIRONMENT + DOCUME N TATION REFERENCE QU AllFIC ATIO N OUTST ANDING EQUlPMENT DESCRIPTION METHOD ITEMS PARAMETER SPECIFICATION QUALIFICATION SPECIFICATION QUALIFICATION OPE R ATIN G 180 days 301 M TIME SYS T E M Process Radiation
" " " ' "9 NORMAL AC NT !
PLANT ID NO'17RT-463A,8 T RWRE 44,169 40-100 N/A COMPONENT Preampll fler PSIA 14.7 N/A 44,169 h MONUFACTURER Victoreen (General Electric) M ODEL NO ' 874A- 1 N M Di Y 20-90 N/A 44,169 N F UNCTION Hi Range Effluent DEMNERALIZED Monitor WATER SPRAY < - No t Applicable /, D MON. m T . 1. D . ACCIDE N T SERVICE: Post Accident Monitoring of Radwaste R ADI ATION 75x103 1.0x10 3 3 Building Effluent Gases Rads LOCATION: furbine Building El. 300'-0" (232) AgtNG 40 years 164 h FLOOD LEVEL ELEV N/A SU8 MERGENCE ( Not Applicable ) 080VE FLOOD LEVEL h E NVIRONME NT
- 1. Equipment required for post-accident monitoring of the Radwaste Bldg. ventilation exhaust.
SPECIFICATION NO 2. Equipment located remote from the area being monitored. Post LOCA-there will be no significant change B A SIS in anhient condi tions. A0 LNG PROGRAM installed Fe., 1982 INPUT PASNY P.O. #80-17553
+ NCE SECTION T.l D. ( TOTAL INTEORATED DOSE le 10 VR, NOftM AL i- ACCIDEN T AL DOSE Rev. 0 F N /A NOT A PPLIC A8LE 1750
THI Mod. No. F1-80-14 SYSTEM COMPONENT EVALUATION Eci 26 JAME A TZ ATR CK JUCLEAR POWER PLANT WORK SHEET DOCKET NO.50- 333 PG 7 of ENVIRONMENT + D OCUM E N TATION REFERENCE QU AllFIC ATIO N OUTST ANDING j EQUIPMENT DESCRIPTION METHOD ITEMS PARAMETER SPECIFIC ATIO N QUALIFICATION SPECIFICATION QU AllFICATION l 4 SYSTEM #
- toring NORMA AU "
p PLANT iD NO 17RR-053A.B 7RR-434A B TEustRTURE 17RR-463A,8 #F 40-120 N/A 44.170 COMPONENT Strip Recorder PRESSURE 14.7 N/A 44,170 g MANUFACTURER Leeds & Northrup PSIA (General Electric) 400 EL NO : Speedomax Hark lli 10-60 N/A 170 ? HU D ITV '/. Serial #C81-36883-1-6 F U NC TION Hi Range Effluent DEMNER ALIZE D / Recorder Not Applicable WATER SPRAY \ D MON. ( m) T. I . D . AC CIDE N T SE RVICE : Post Accident Recording of effluent gases R ADI ATION r 700 N/A 3 Display Ins t runen t a t ion Rads LOCATION: Control Room Admin. Bldg. El 300' ASING 40 years 164 W FLOOD LEVEL ELEV N/A SUBMERGENCE ( Not Applicable -
)
CROVE FLOOD LEVEL h E NVIRONME NT 1. Hild Environment
" "' ' ' *" " '"9 " * " " 9* "" # '$ 9 '"'"
yO SPE CIFIC ATION gases BASIS AGING Installed Fe., 1982 PROGRAM PASNY P.O. #80-17553 INPUT T I D. ( TOT AL INTEGRATED DOSE j e 40 VR. NORMAL + ACCIDENTAL DOSE Rev. O EF NCE SECTION N /A NOT A PPLIC ABLE
)
d TMI Mod. No. F1-80-14 NUpf G 0737, Sec tion ll.F.I.1 1E BULLETIN 7 9 - 010 SYSTEM COMPONENT EVALUATION JAVES A FITZPATRICK NUCLEAR POWER PLANT sEcTloN 26 DOCKET NO.50-333 WORK SHEET pc s oF , ENVIRONMENT + DOCUMENTATION REFERENCE QU ALIFIC A TIO N OUTSTANDING
' E Q UIPMEN T DESCRIPTION METHOD ITEMS F ON OPE R ATING T 180 days 01 ri NOR W AL AC NT N
PLANT ID NO I7RM-053A.B 17RM-434A.B T E M PE R TUR E Y 17PM-463A,B og 40-120 N/A 44,170 COMPONENT Readout Module PRESSURE w MANUFACTURER Victoreen PSIA 14.7 N/A 44,170 (General Electric) Rft TlVC 10-60 N/A 44,170 y MOC EL NO: 846-2 l'.*144 DITY */o F U NC TION Hi Range Effluent Monitor Ctw ER AL1 ZED mATEM SPRAY
/
N Not Applicable ) D 404. m) T. 1. D . AC CIDE N T SE RVICE . Post Accident Monitoring of effluent gases P. ADI ATION 700 N/A 3 Rads LOCATION: Control Room Admin. Bldg. El. 300' A9 LNG 40 years 164 3. - FLOOD LEVEL ELEV N/A SUBWERGENCE Not App!! cable ) . A80VE FLOOD LEVEL h E NVIRONME NT 1. Hild Environment SPE CIFICATION 2. Equipment required for post-accident monitoring of Stack, Turbine Bldg. and Radwaste Bldg. effluent NO gases 84518 AGING Installed Feb., 1982 PROGRAW PASNY P.O. 480-17553 input REF NCE SECTION T.l D ( TOTAL INTEGRATED DOSE le 40 VR. NORM AL + ACCIDEN TAL DOSE Rev. O N /A NOT A PPLIC ABLE f
l 1 1 Section 27 - Containment and Reactor Pressure / Level Monitoring System ; NUREG 0737, Sections li.F.1.4 and ll.F.1 5 O
- 1. The instrumentation package for this item was procured from the General Electric Company and is described on the attached sheets.
- 2. The status of the qualification documentation for this equipment is as follows:
- a. Foxboro (power supplies, summer, recorder, signal conditioner, ano distrubution module) - All of this equipment is installed in a mild environment. Qualification testing is cunplete and the Authority expects to have the test documentation by July 15, 1982.
- b. Barton (pressure transmitters) - This equipment is in the final stages of qualification testing. Qualification
' test parameters exceed the JAF accident envi ronments.
Based on the present schedule for completion of the test program, it is expected that the Authority will receive the test documentation by August 15, 1982. 3 At the time of procurement, the equipment which was obtained and installed was the "best available" and is considered acceptable in the interim pending confirmation of qualification. n' ~- . O 4
TMl Hod. No. FI-80-15 NUREG 0737, Section ll.F.I.46ft.F.1.5 iE BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION SECTION 77 JAMES A FITZPATRICK NUCLEAR POWER PLANT WORK SHEET po 1 og DOCKET NO.50-333 ENVIRONMENT + DOCUM E N TATION REFERENCE QU ALIFIC ATI O N OUTSTANDING EQUIPMENT DESCRIPTION QUALIFICATION SPECIFICATION QUAllFICATION METHOD ITEMS PA RAMETER SPECIFI C ATION OPER A T!NG 180 days 101 TIME evel A ident H tor ng NORMAL ACglDENT PL AN T ID NO (see below) TEM PE R TUR E 40.y04 220 44,lA,1W A er COMPONENT Pressure-level Transmitter PRES $URE g 44,2A,2W MONUFACTURER ITT Barton (General Electric)
" ' N M ODEL NO. : 763 20-90 90-100 44,167 NUM Dl Y '4 F U NCTION Press / Level Honitor OEMhER ALIZE D \
WATER SPRAY Not Applicable f
'* T. I . D . A C CIDE N T DE M ON.
SE RVICE; Accident Monitoring g Display instrumentation R ADI ATION 6.8x106 '- Rads 6.87x106 3 LOCATION: Reactor Building (as shown below) ASING 40 years 164 N FLOOD LEVEL ELEV 227'-6" SU8 MERGENCE ( Not Applicable )
- 1. Harsh accident environment.
ABOVE FLOOD LEVEL h E NVIRONME NT
- 2. Post accident environmental conditions based on a worst case Primary Containment LOCA as experienced in NO SPE CIFICATION the Reactor Dldg. or a Reactor Bldg. HELB for the equipment loca t i tm s .
3 Equipment provides sensor signal to display instrumentation located in the Control Ronm. Installed Feb., 1982 B A SI S PASNY P.O. f80-19148 23LT-203A1,BI------344'(5.5W) A0 LNG 2 7P T-I l 5A1, A2, B I , B2-34'e ' (5. 5W) PROGRAM 06PT-61 A ,618-----300' (3R) . (5.5W) meur 23LT-203A2,82 - 227'(4.5A),(4.5D. REFE NCE SECTION T I D. ( TOT AL INTE4 RATED DOSE)s4 Y . NORM AL -t ACCIDEN T AL DOSE i Rev. O I N /A NOT A PPLIC ABLE
mus .
. - a TMI Mod. No. F1-80-15 NUREG 0737, Sect ion ll.F.I.fi t, ll.F.1.5 iE BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION JAMES A FITZPATRICK NUCLEAR POWER PLANT SECTION 27 DOCKET NO.50-333 WORK SHEET pG 2 og ENVIRONMENT + DOCUM ENTATION REFERENCE QU ALIFIC ATIO N OUTST ANDING
! EQUIPMENT DESCRIPT!ON METHOD ITEMS PARAM ETER SPECIFI C ATION QUAllFICATION SPECIFICATION QUALIFICATION OPER ATING 180 days 301 -e TIME SYS T E M Cont. t, Reactor Press /
PLA T ID NO T 0 COMPONENT Pressure-Level Transmitter PRESSURE PSIg 14.7 15 5 44,2A MANUFACTURER ITT Barton (General Electric) MODE. NO: 764 20-90 90-100 44,167 h HUMIDITY */. F UNCTION Press / Level Monitor DEMINERAll2ED / Not Applicable WATER SPRAY N DE M O'N . T . 1. D . ACCIDE N T SERVICE: Accident Moni torIng Display Instrumentation R ADI ATION g f 4 Rads LOCATION: Reactor Building El.227'(4.5A),(4.5D) 40 years AGING 164 g 227'*6" FLOOD LEVEL ELEV SU8 MERGENCE Nnt Applicable - ) l A8 0VE FLOOD LEVEL @ E NVIRONME NT 1. Harsh accident environment. j
- 2. Post accident environmental conditions based on worst case Containment LOCA, or Reactor Bldg. HELB NO SPECIFICATION for the equipment locations.
B A SIS 3 Equipment provides sensor signal to display Instrumentation located in the Control Room AGING t Installec. Feb., 1982 PROGRAM PASNY P.O. #80-19148 INPUT l EFE NCE SE CTION T f D. ( TOTAL INT E GR AT ED D OSE l e 40 VR. NORM AL + ACCIDEN T AL OOSE Rev. O 0 N /A NOT A P PLIC AB L E 7a10
\
T M l Mod ._1+3. F1-80-15 NUREG 0737. Section II.F.I.4 & fl.r.1.5
.E BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION 27 1AMES A FIT 7 PATRICK NUCLEAR POWER PLANT SECTION
)OCKET .NO.50-333 WORK SHEET PG 3 or l -
thiPONMENT + DOCUM E N TATION REFERENCE QUALIFib T IO N OUT ST ANDING-
! EQUIPMENT DESCRIPTION ; -
, ITEh*S SPECIFI C ATIO N s QUALIFICATION SPECIFIC AT 60N QUAllFICATION Y '/,H O D PAPAW}TER
- I .
1 , OPER ATINS TIME ISO days -
~' Wiz -
Y - .a SfSTEM Contal.nnent & Reactor ' - + Press / Level Accident Monitoriri : NORMAL ACCIDENT h , PLA T ID NO (%.e be low) i S
- s *0-120 P/A 44,170 ' ' F' COMPONENT. Recorder _
/ _'
PRESSURE 44,170 . pSgA 14.7 N/A . MONUFACTURE R Foxboro 7, (General Electric) ~ l ' - r - M
,r .,.
j MooEL NO. : 226S 10-60 "N/A ' 44,170 *
/ Y/
NUMIDITY % , j -j, , j f ,-.C v. r FJ4CTION pg ess/ Level Recording ' ^ DEMINER ALIZED
- N' WATER SPRAY ( .
- Not Appilcable - - _
ACCuqACv- EC uog. 0.5% T. 1. D . A CCet N T ' g-SERVICE: Accident Monitoring R ADI Afiole # Display Ins t runen ta t ion ?N N/A 3
'~
Rads LOCATION: Control Roorn Admin. Bldg. El 300' 164 1 FLOOD LEVEL ELEV N/A SUSMERGE N CE ( Not Applicable - - A80VE FLOOD LEVfL @ E?PM o*vMENT 1. 2. Hild Environnent Equipment required for post-accident nw>nitoring of containment level and pressure, and reactor pressure. SPE CIFICATiog NO SASIS Installed Feb., 1982 agegg PASNY P.O. #80-19148 PROGRAM 270R-il5AI,Bl.A2,82 INPUT 23LR-202A,2028,203A 2038 Rev. O N REF N /A NOT A PPLIC ABLE NCE SECTION T I D. ( TOTAL INTEGRATED DOSE ).40 VR. NORM AL + ACCIDEN T AL DOSE
~. .
tMI Mod. No. F1-80-15 iE BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION # ' # * * * ' ' ' " ' #SECTION ~ ' ' ' ' ' # 2' ' '7 JA?/ES A FITZPATRICK NUCLEAR POWER PLANT WORK SHEET Pc 6 or aoCKET NO.50-333 ENVIRONMENT + DOCUM E N TATION REFERENCE QU ALIFIC ATIO N OUTST ANDING DESCRIPTION
~
METHOD ITEMS r PA RAM ETER SPECIFIC ATIO N QUALIFICATION SPECIFICATION ' QUAllFICATION l EQUIPMENT [," "I '"' 180 days 101 SYS T E M Containment & Reactor ~ Press / Level Accident Monitoring PLANT 10 NO (see below) COMPONENT $lgnal Condi tioner PRESSURE 'W pgga 1 7 N/A 4fe ,171 C3tNUFAC TURE R Foxboro (General Electric) RELAT E N/A 44,171 #- M00EL NO: N-2Al-12V 10-60 FUNCTION Pressure / Level Monitoring TER SPRAY Not Applicable - [ oCCuRAc" [og, 10.5% T. s . o . ACCioE N T 9E R VtCE . Accident Monitoring s R A01 AT10N Rads 700 N/A 3 LOCATION: Relay Room Admin. Bldq. El. 286' aging 40 years 164
- FLOOD LEVEL ELEV N/A SU8 MERGENCE bt Applicable -
080VE FLOOD LEVEL E NVIRONME NT 1. Mild Environment
- 2. Equipment required for post-accident monitoring of containment level and pressaire, aml reac or pressure g ,
B ASep Installed Feb, 1982 PASNY P.O. 180-19148 ASINS 235CM-202A.202B,203A,2038 PROGRAM 275CM-IISA.Il5B '" 05sCM-61A,61B Y 10 ( TOTAL INTEGRATED DOSE l s 40 VR. NORM AL + ACCIDEN T AL DOSE Rev. O R EF NCE SE CTION N /A NOT A PPLIC ABL E 1 - - - - - - _ _ _ _ _ _ _ _ _ _ _ _
l TN Mw %. ri-80-15 Nunt G 0 73 7, sec t i on ll.r.l.4 e li.r.l.s 018 JAME A T ATR CK NUCLEAR POWER PLANT SYSTEM COMPONENT EVALUATION SECTION 27 DOCKET NO.50-333 WORK SHEET pc s or ENVIRONMENT + DOCUMENTATION REFERENCE QU ALIFIC ATIO N OUT ST ANDING EQUIPMENT OESCRIPTION PARAME TER SPECIFI C ATIO N QUALIFICATION SPECIFICATION QUALIFICATION METHOD nEMS OPER ATIN G # TIME 180 days 301 SYSTEW Eontainment & Reactor NORMAL ACC NT PL A*8 T ID NO (see below) TE M PtR TURE ! ep 40-120 N/A 4k+I73 COedPONE N T. Signal Distribution PRESSURE
' lis . 7 N/A N 4%.I73 MANUFACTURE R Foxboro (General Electric)
RELATIVE NODEL NO.
- N-2AX*D10 10-60 N/A 44,17I NUMIDITT %
F O NC T LOM Pressure / level OEMDeER ALIZE D j Not Applicable -
/
ACCURACY- EC 8 g 05% T. .D. ACCIDE N T SE RVICE: Accident Monitoring R ADi ATION , Rads 700 N/A 3 LOCATION: Relay Room Admin Bldg. El.286' AsiNe
- Q gn lg FLOOD LEVEL ELEV N/A SUBWERGENCE . _ . Not Applicable -
ABOVE FLOOD LEVEL @ E dVIRONME NT 1. 2. Mild Environment Equi m nt required for post-accident moni toring or containment seve and pressure, ana reactor pressure. BASIS Installed Feb., 1982 PASHY P.O. #80-19148 ,,,q, 23SDM-203A,8 23SDM-Il5A,B INPUT 06SDM-61A,B NO T E S: N y qF NCE SE CTION T ID ( TOTAL INTEGRATED DOSE ) 40 VR. NORMat, + ACCIDEN TAL DOSE ev. O N /A NOT A P PLIC A8 L E
a w w 1MI Mod. No. FI-SO-15
" " " " 737' ' '
IE BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION 5'
" ' ' $cTIo'N S -E-JAMES A FITZPATRICK NUCLEAR POWER PLANT WORK SHEET PG 6 oF
. DOCKET NO.50-333 l ENVIRONMENT + 00 CUM ENTATl0N REFERENCE QU ALIFIC ATIO N OUTST ANDING EQUlPMENT DESCRIPTION QUALIFICATION METHOD ITEMS PARAM ETER SPECIFICATION QUALIFICATION SPECIFICATION OPERATINR 180 days 101 TIME Syst E M Containment (, Reactor
*'^"'*" NORMAL AC NT
; g PLANT tD NO' 2 35UN-203A,B T RTURE 44,171 Y 40-120 N/A COMPONENT. Summer Module PR SSURE
14.7 N/A 44,171 MANUFACTURER Foxboro PSIA (General Electric)
"' 10-60 N/A 4%,171 f-MOD EL NO: N-2AP-SUM NUMIDITY '/.
F U NC TION P res s ure/Le ve l DEMesERALIIED
/ Not Applicable -'
ACCURACY EC +0 5% ,g,p, ACCIDE N T SE R Vie.E : Accident Monitoring . R ADI ATION N/A 3 700 Rads LOCATION: Relay Room Admin. Bldg. El.286 ASING 40 years 164 FLOOD LEVEL ELEV N/A SUBMERGENCE Not Applicable - ( ABOVE FLOOD LEVEL @ E NVIRONME NT 1. 2. Nild Environment Equipment required for post-accident monitoring of containment levei and pressure, and reactor pressure. NO SPE CIFICATION 8 ASIS AGING PROGRAM Installed Feb., 1982 INPUT PASNY P.O. #80-19148 i< EF NCE SECTION T ID ( TOTAL INT EGR AT ED 00SE le40 VR. NORNAL + ACCIDENTAL DOSE Rev. O N /A NOT A PPLIC ABL E
TMI Mod. No. F1-SO-15 E BULLETIN 7 9 - 010 """ G '3 7 ' JAMES A FITZPATHICK NUCLEAR POWER PLANT SYSTEM COMPONENT EVALUATION 5 3 "" " #SECTION
' ' ' ' # ! >5-
)OCKET NO.50-333 WORK SHEET PG. 7 oF ENVIRONMENT + DOCUME N TATION REFERENCE QU ALIFIC ATIO N OUTST ANDING I DESCRIPTION METHOD ITEMS PARAM ETER SPECIFI C ATIO N QUAllFICATION SPECIFICATION QUALIFICATION l EQUIPMENT OPH AUN G
- 180 days 101 TIME SYST E M Eontainment & Reactor j NORMAL AC DENT PLANT ID NO 23 PDM-200A,B TEM PER TUR E Y op 40-120 N/A 44,l71 COMPONENT Power Dis t ributlon Module PRES $URE PSIA 14.7 N/A 4t,178 i
4 M ANUFA C TU RE R Foxboro (General Electric) MODEL NO : N-2ANU+DP10 10-60 N/A 44,171 Nu DT *4 F U NC TION Pressure / Level Monitoring . OEMINER ALIZED ' Not ApplIcabie \ WATER SPRAY % f 0 M O'N. ?0 5% T. I . o . ACCioENT 9E R VICE : Accideet Monitoring R ADI ATION Rads 700 N/A 3 LOCATION: Relay Room Admin. Bldg. EI.286' ASING 40 years 164 FLOOD LEVEL ELEV N/A SUBMERGENCE ( - Not Applicable ABOVE FLOOD LEVEL h E NVIRONME NT I. 2. Hild Environment Equipment required for post-accident monitoring of containment level and pressure, and reactor pressure. go SPECIFICATION BASIS ! l AGING PROGRAM Installed Feb., 1982 INPUT PASNY P.O. #80-19148 NCE SE CTION T ID ( TOTAL INTEGRATED DOSE ).40 VR. NORM AL + ACCIDENT AL DOSE Rev. O EF N /A NOT A PPLIC ABLE
TMl Mod. No. F1-80-15 SYSTEM COMPONENT EVALUATION "'* '" ' ' # SE
' jT O b[
AME A TZPATR CK UCLEAR POWER PLANT ')OCKET NO.50-333 WORK SHEET pc 8 og ENVIRONMENT + DOCUMENTATION REFERENCE QUALIFIC ATION OUTST ANDING EculPMENT OESCRIPTION QU AllFICATION METNOD ITEMS PARAM ETER SPECIFICATION QUAllFICATION SPECIFICATION OPER ATIN G 108 SfSTEM Eontainment f. Reactor PLA T D 2 / A,8 T RWRE .- 40-120 N/A 4%.178 COMPONENT Multinest Power Supply "E *: pg,SSURE g 14.7 N/A 44.371 MANUFACTURE R Foxboro (General Electric) R ELATIVE MODEL NO N-2ARPS-A6 ST.D # HuMIDITV % 10-60 N/A 44,l78 F UNC TION Pressure / Level Hrmi torin9 DEMINERallIED / Not Applicable
\
WATER SPRAY \ / ACCHAC M S C N/A o uoq. T. 1. D . A CCIDE N T SE 9VICE- Accident Monitoring W Rads LOCATION: Relay Room Admin. Bldg. El.286' AGING y. 40 years 164 FLOOD LEVEL ELEV y/g SUBMERGENCE Nog App]Icable - ABOVE FLOOD LEVEL @ E NVIRONME NT l. 2. Mild Environment Equipment required for post-accident monitoring of containmetc evel and pressure, and reactor pressure. SPECIFWM BASIS ASING PROGRAW Instal led Feb. , 1982 INPUT PASNY P.O. #80-19148 NOTES:
- T I D. ( TOTAL INTEGRATED DOSE l e 40 VR. NORW AL + ACCIDEN T AL DOSE
+ EF NCE SE CTION ev. O N /A NOT A P PLIC ABL E
i Section 28 - Primary Containmsnt High Range Radiation Monitoring System NUREG 0737, Sections ll.F.1.3 and ll.E.4.2.7 O i
- 1. The qualification test reports for the General Atomics equipment l listed on the following sheets are presently.in final review by l' the Authority. This review is expected to be completed by June 30, 1982.
- 2. A preliminary assessment of these reports has confirmed that the JAFNPP accident environmental parameters have been enveloped by the testing performed.
I O 4 i l i e I i ( l
-w-a-5:-e- y- -wg -..pym.em-,paeg,-yws mo ee--hy< m- p_gp,y,-%~ rg tjre ,gw m -aw+ .y waer-rl-e'.p-g4+ Pe- 9ep*Prm*,t--9sr-Wwyyy. g $q--wwmpgur-- we g v g -~ 7T6'F*y?W
TNi Mod. No. FI-80-16 NUREG 0737. Section ll .F.I.3 & ll.E.4.2.7 JAME A TZ ATR CK vCLEAR POWER PLANT SYSTEM COMPONENT EVALUATION SE CTION 28 JoCKET NO.50-333 WORK SHEET PG ' Or ENVIRONMENT + DOCUM E NTATION REFERENCE l iEQUIPMENT DESCRIPTION QU ALIFIC ATIO N OUT ST ANDING f PARAM ETER SPECIFICATION QUALIFICATION SPECIFICATION QUALIFICATION NETHOD ITEMS T 180 days 101 A Sainp NORMAL ACCIDEN T l PLANT 10 NO 27RE-104A.8 l i T E MPE R TUR E g
*F 135-150 308 %.57
- COMPONENT Radiation Detector Element PRESSURE PSIA 14.2-16.7 A MINuFACTURER Ceneral Atomics 59 7 %.58 N00EL NO* RD-23
- HW T 'A 40-90 100 44.57.58 F UNCTION Honitoring Containnent Radiation OEMNERPLIZED j WATER SPRAY Yes 106 0 MON. T. I . D . ACCIDE N T SERVICE: Display Instrunentation
& P.C. Isolation Signal R ADI ATION 1.0x108 1.09x108 3 Rads LOCATION. Drywell El.290' A0 LNG 40 years 164 x r FLOOD LEVEL ELEY 258'-6" SUBMERGENCE Not Applicable
)
ABOVE FLOOD LEVEL h E NVIRONME NT 1. 2. tiarsh accident environment. Equipment required to provide display ins trumentation signal for high contalmnent radiation. Also NO SPE C'FICATION provides a primary containment isolation signal. B A SIS 3 Equipment required for post-accident monitoring. AGING PROGRAM installed Feb.. 1982 INPUT PASNY P.O. #80-17552 Rev. O R EF NCE SE CTION T 10. ( TOTAL INTEGRATED DOSE le 40 VR. NORM AL 4 ACCIDEN TAL DOSE l N /A NOT A PPLIC ABL E 8 5x10 6
THt Mod. No. F1-80-16 NUREG 0737, Section ll.F.I.3 & I I .E .t i.2. 7 JAI /EA T PATR CK UCLEAR POWER PLANT SYSTEM COMPONENT EVALUATION SECTION 78 DOCKET NO.50-333 WORK SHEET PG 2 OF ENVIRONMENT + DOCUM E N TATION REFERENCE QUALIFIC ATIO N OUISTANDING
! EQUIPMENT DESCRLPTION METNOD ITEMS l PARAMETER SPECIFICATION QUALIFICATION SPECIFICATION QUALIFICATION OPE R ATIN G TIME 180 days I0I s in NORMAL AC IDENT PLANT iD No 27RM-104A,B T E M PE R TUR E 40-120 N/A i
'F 4%.870 COMPONENT Radiation Readout Module and Power Supply PRESSURE ,
"^ N MANUFACTURER Ceneral Atomics P8' '7
N00EL NO* RP-2C /RP-23 10-60 HMD Y */. N/A 44,170 F U NC TIONContaInment Radiation DEMINERALIZED WATER SPRAY k Not AppIlcable --
)
D M O'N. T. 1. D . AC CIDE N T SERVICE Display Instrumentation R ADI Af t0N p 700 N/A 3 Rads LOCATION: Control Room Admin. Bldg. El.300'
,,,, y-FLOOD LEVEL ELEV N/A SUBMERGENCE Not Applicable - ->
ABOVE FLOOD LEVEL @ E NVIRONME NT 1. Hild Environment NO SPECIFICATION 2. Equipment required for post-accident monitoring of primary containment radiation. 84818 AGING PROGRAN Inttalled Feb., 1982 INPUT PASNY P.O. #80-17552 Rev. 0 + EF NCE SE CTION T 4 D. ( TOTAL INTEGRATED DOSE ) = 40 VR. NORM AL + ACCIDEN T AL DOSE N /A NOT A PPLIC ABLE
i i l Section 29 - Post-Acciden). Sampling System
- NUREG-0737, Section ll.B.3.2 i
I O I
- 1. The majority of equipment associated with this system is not required
) to be environmentally qualified. Per Regulatory Guide 1.97, Rev. 2, components of a "high commercial grade" should be utilized. A listing of the "high commercial grade" equipment to be utilized is l listed on Sheet 3 3 }
- 2. Sheets 1 and 2 are components associated with an alternate power j supply for the post-accident sampling system that interfaces with j Class IE electrical buses in the plant.
1 4 i i 4 i a !, O ) i i i i I i l i i 1 1 4
wk -- . . _ _ ___ O O TMl Nod. No. FI-80-19 NUMfG 0737. Section ll.B.3.2 SYSTEM COMPONENT EVALUATION SECTION 29 IA E A T PATR CK JCLEAR POWER PLANT
>OCKET NO.50-333 WORK SHEET PG i or ENVIRONMENT + D OCUMEN TATION REFERENCE QU AllFIC ATIO N OUTST ANDING l
DESCRIPTION METHOD ITEMS PARAM ETER SPECIFI C ATION QUALIFICATION SPECIFICATION QUALIFICATION lEQUlPMENT O ATIN G 180 days >l80 days 101 20% N/A (see notes 2 & 3 SfSTEM Post Accident Sampling NORMAL D AGeaK ! PLANT ID NO' - TEMPERTURE 40-120 N/A N/A 44,171 204 Simultaneous None COMPONENT Transfer Switch PRESSURE Ils . 7 N/A N/A 44.171 204 $1multaneous None MANUFACTURER Electroswitch PSIA R E L ATIVE MODEL NO. : 240lc 10-60 N/A N/A 44,171 204 Simultaneous None NUMIDITV '4 F U NC TION Electrical Power / \j DEMNER ALIZE D Not Transfer Switch \ Applicable WATER SPRAY MO'N. N/A T. 1. D . AC CIDE N T SERVICE: Power Supply to Sample System R ADI ATION 700 N/A 1.0x104 3 204 Sequential None Rads LOCATION: Relay Room Mmi n. Bldg. El.286' AtlNG 40 years (see notes 2r.3) 164 204 (see notes 2r.3)
- FLOOD LEVEL ELEY N/A SUBMERGENCE Not Applicable
)
i ABOVE FLOOD LEVEL h E NVIRONME NT ", g y,f NO SPECIFICATION 3. Electroswitch Corporation has begun a new test program which will qualify 24 series switches for 40 years. BASIS Test program expected to be completed by the end of year 1982. A0lNS PROGRAM install Dec., 1982 INPUT PASNY P.O. #81-6131 Rev. 0 + EFE NCE SE CTION T f D ( TOTAL INTEGRATED DOSE ) = 40 VR. NORM AL' e ACCIDENT AL DOSE l N /A NOT A PPLIC ABL E
TM1 Mod. No. F1-80-19 NUREG 0737. Section li.B.3.2 1E BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION 29 JAMES A FITZPATRICK NUCLEAR POWER PLANT SECTION DOCKET NO.50-333 WORK SHEET pc 2 or ENVIRONMENT + DOCUME N TATION REFERENCE QU ALIFIC ATIO N OUTSTANDING EQUlPMENT OESCRIPTION METHOD ITEMS PARAMETER SPECIFICATION QUALIFIC ATION SPECIFICATION QU AllFICATION OPER ATIN G 180 days >t80 days 101 204 N/A TIME (see notes IE2) SYST E M Post Accident Sampling NORMAL AC{D NT PLANT ID NO - T ERTURE 40-104 N/A N/A 168 204 Simultaneous None COMPONENT Transfer Switch PRESSURE M,tNUFA CTU RE R Electroswitch PSIA I4.7 N/A N/A 168 204 Simultaneous None HU DITY % 20-90 N/A N/A 168 204 Simultaneous N me F ONC TION Electrical Power DEMNER ALIZE D Transfer Switch WATER SPRAY
- ^
Not Applicable - l 4 0'N T. I . O . ACCIDE NT SE RVICE; Power Supply to Sample System R ADI ATION 765 65 1.0x10% 3 204 Sequential None Rads LOCATION: MG Set Room El.300'-O' 40 years (see notes It2) 164 204 (see notes 162)
- FLOOD LEVEL ELEV N/A SUBMERGENCE Not Applicable - -- ---
-)
ABOVE FLOOD LEVEL ES E NVIRONME NT
- 1. Test specimen demonstrated less than ,0 years of life.
NO SPECIFICADON 2. Electroswitch Corporation has begun a new test program which will qualify 24 series switches for 40 years. B A SIS Test program expected to be completed by the end of year 1982. AsiNe PROGRAM Install Dec., 1982 INPUT PASNY P.O. #81-613l Rev. O EE NCE SE CTION T ID ( TOTAL INTEGRATED 003E le 40 VR NORM AL + ACCIDENTAL DOSE N /A NOT A PPLIC ABLE 700
O O O Section 29 - Post-Accident Sampling System Page 3 List of High Grade Commercial Components No. Component Manufacturer /Model No. Location Comment
- 1. GE BWR Post Acc. GE/ Spec. C5474-SPI MG Set Room 300' Sample Station Rev. 1
- 2. Indicating Lights GE/ET16 Relay RM SIP Panel See Sh. 32, pg. I 3 Terminal Block GE/EB25Al2W Relay RM SIP Panel See Ref. 202 Terminal Block GE/EB25A06W MG Set Room 300' 4
- 4. Control Switch GE/CR2940-US203E Relay RM SIP Panel See Ref. 206 Control Switch GE/CR2940-UA202B Relay RM SIP Panel 4
- 5. Circuit Breaker GE/THQB1120 Relay RM SIP Panel See Sh.21 , pg.6 Circui t Breaker GE/THQBil30 RB 272' (3Y) I
- 6. Relay GE/CR2810A14AT2 Relay RM SIP Panel Relay GE/CR2810A14DH2 Relay RM SIP Panel Relay GE/CR2810A14AK2 MG Set Room 300' Relay Agastat/7012AF MG Set Room 300' 7 Solenoid Valves Target Rock /81JJ-001 RB Generic See Sh. 30, pg. I Solenoid Valves Target Rock /81JJ-005 RD Generic Solenoid Valves Target Rock /81JJ-002 RB Generic Solenoid Valves Target Rock /81JJ-006 RB Generic Solenoid Valves Target Rock /81JJ-003 RB Generic Solenoid Valves Target Rock /81JJ-004 RB Generic
- 8. Exhaust Fan LAU Ind./HPR10 MG Set Room 300' 9 Flow Switch Drying Sys/955R MG Set Room 300'
- 10. Ala rm Malory Sonalert/SC110D MG Set Room 300'
1 M I Mod . M. F1-BC-20 NUREG 073 7. Sec t ion ll F.1.6 SYSTEM COMPONENT EVALUATION SECTION 30 hAME A T ATR CK UCLEAR POWER PLAl.T WORK SHEET i 30cKET NO.50-333 PG OF ENVIRONMENT + D OCUM EN TATION REFERENCE QU ALIFIC ATIO N OUTST ANDING l I EQUIPME NT DESCRIPTION METHOD ITEMS i PARAMETER SPECIFIC ATIO N QUALIFICATION SPECIFICATION QUALIFICATION "I'"' 180 days >l80 los 225 Simultancous None ' SYSTEM E in ent Atew> spheric - - PLANT 10 40 (see below)
" Siaultaneous Une 40-104 110 385 44,148,166 225 COMPONENT Solenoid Valve M ANUFA CTURER Target Rock SIA 7 G.0 77.7 % M S imi t aneous k M OD EL NO: BlJJ-00le 20-90 90-100 100 44,167 225 Simultaneous *me D TY '4 F UNC TION H2 Sample OEMhER AllIE D y WATER SPRAY N Not Applicable -
)
0 N 0'N. N/A T. I . D . ACCIDE N T SERVICE: Primary Containment Isolation Valve R ADI ATION g,g7,gg6 Nme 6.8x106 2.27x10 3 225 Sequential Rads LOCATION: Reactor Building (as shown below) AGING 40 years >40 164 225 Sequential None FLOOD LEVEL ELEV 227'-6" SUBMERGENCE __ Not Applicable ABOVE FL O OD LEVEL ES E NVIR ONME NT
- 1. Harsh accident environment.
- 2. Required to operate post-LOCA/MSLB occuring in primary containment, but located in Reactor Building.
NO SPECI'lCATION 3 Accident environment is long term post-LOCA for Reactor Building. Install Dec., 1982 8 A S18 4. The Ethylene-Prophylene gasket and the polyamidc-imide disc in the subject valve will be replaced PMNY P _0_ #80- 7 7 D(i7 during the next outage with the sillcone rubber gaskets and discs as used in the test specimen in 2 7S0V-120El .120E2, ll9Fl . Il9F2, order to extend aualified lifetime. 122El.122E2 - El.326' (SR) 2750V-123EI.E2 - El.300' (SR) AMNO 6650V-210A,B - El.300' (2T,6R) 2750V-120F1, 120F2, 122FI, 122F2 , PROGRAM 123F1, 123F2 - El. 272' (3W) INPUT 2750V-119El il9E2, 124E1, 124E2 , 124Fl. 124F2 - El. 227' (SP) T I D ( TOTAL INTEORATED DOSE)e4 Y NORM AL + ACCIDENTAL DOSE Rev. O EF NCE SECTION N /A NOT A PPLIC ABLE
Ms . TMI Mod. No. F1-80-20 NURIG 0737, Section ll F.I.6 iE BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION 30 JAMES A FITZPATRICK NUCLEAR POWER PLANT SE CTION . DOCKET NO.50-333 WORK SHEET pc 2 og j i ENVIRONMENT + DOCUME N TATION REFERENCE QU ALIFIC ATIO N OUTST ANDING EQUIPMENT DESCRIPTION METHOD ITEMS PARAMETER SPECIFI C ATIO N QUAUFICATION SPECIFICATION QUAllFICATION I l OPER ATIN G 4 180 days 101 TIME SYSTEM Containment Atmospheric Sampling NORMAL ACCIDENT PLANT D No 23HAZ-102A,B TE M PE R TUR E f,0- 10 4 110 44,188 4
'F COMPONENT Hg Analyzer Panel PRESSURE M ANUFA CTURE R Comsip N00EL NO* KllI 20-90 90-100 44,167
U OlTY '4 t i FUNCTION Gas Analyzing OEuedERALIZED \ WATER SPRAY Not Applicable /-' o O'N . T. 1. D . ACCIDE N T 9E R VICE: Primary Containment H2 Monitoring R ADI ATION 2.2x105 2.0x105 fc 3 Rads LOCATION. Reactor Bulldlng El. 300' (2T),(6R) A0 LNG 40 years 164 p' FLOOD LEVEL ELEV 227'-6" SUBMERGENCE ( Not Applicable - - - -
--- h C8 0VE FLOOD LEVEL h E NVIRONME NT SPECIFICATION 1.
2. Harsh accident environment. Required to operate post-LOCA/MSLB occuring in primary containment, but located in Reactor Building. NO
- 3. Accident environment is long term post-LOCA for Reactor Building.
Ba SIS ASING PROGRAM Install Dec., 1982 INPUT
- PASNY P.O. #80-19794 T l D. ( TOTAL INTEGRATED DOSE ) e 4 0 VR. RM AL 4 ACCIDEN TAL DOSE Rev. EF NCE SE CTION i
O N /A NOT A PPLIC ABL E I.8x10
. _ _ . ,. . - - - - . - - _ . . - . . ~ . . -..- - .- - . -. ~. - .......- .. --. - - -.-- _ .. - - - . . -.
t t 1 l Section 30, Page 2 1 4 j 1. The Authority is presently reviewing the qualification test reports for the Comsip Kill hydrogen analyzer and is refining the aging analysis. The review of this report-will be finalized i by July 15, 1982.
- 2, This equipment will be installed by December,1982.
{ f
- i 1
lO i ) a 1 . i i I ) i ) L
, l I !
- i
!0 4 4 t i I 2 :
O TMl Hod. No. F1-80-20 NiJREG 0737, Section ll.F.I.6 E BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION SECTI0N M-JAMES A FITZPATRICK NUCLEAR POWER PLANT JOCKET NO.50-333 WORK SHEET PG s OF ENVIRONMENT + D OCUME NTATION REFERENCE QU ALIFIC ATIO N OUTST ANDING l EQUIPMENT DESCRIPTION PARAM ETER SPECIFIC ATIO N QUALIFICATION SPECIFICATION QUALIFICATION METHOD ITEMS f OPER ATIN G 180 days 10; + TIME SYSTEM Con inment Atmospheric PLONT 10 N0' 27HAX-102A,B T E M PE R TUR E 40-120 N/A 44,170 ~~4
'F COMPONENT.2 H Analyzer Panel (Remote Indicator) PRESSURE 14.7 N/A 44,170 MONUFACTURE R Comsip NODEL NO: Kill HUMIDITY % 10-60 N/A 44,170 Y F UNC TION Primary Containment OEMINERAllIED y N H2 Monitorin9 WATER SPRAY Not Applicable #
D M 0'N. T. 1. D . AC CIDE N T SE RVICE; pi gp lay I ns t rumen t a t ion Rads LOCATION: Control Room Admin. Bldg. El.300' N AGlNG 40 years 164 FLOOD LEVEL ELEV. N/A SUBMERGENCE Not Applicable h I C8 0VE FLOOD LEVEL h E NVIR ONME NT SPE t iFICATION 1. 2. Hild Environment Equipment required for post-accident monitoring of containment H2 concentration. go BASIS AGING PROGRAM Install Dec., 1982
,,d P U T PASNY P.O. #80-19794 EF NCE SECTION T I D. ( TOTAL INTEGRATED 00SE )e 40 VR. NORM AL 4 ACCIDENTAL DOSE ev. O N /A NOT A PPLIC ABLE
1 Section 30, Page 3- ; i. !O
- j. 1.
The Authority is presently reviewing the qualification test reports for the Comsip Kill hydrogen analyzer and is i refining the aging analysis. The review of this report ! will be finalized by July 15, 1982. I i l ) 2. 'This equipment wl11 be Installed by December, 1982. 1 ! l } I i 1 l r !O . f r f I- __. - . . _ _ _ _ _ _ _ . _
0 0 TMI Med. O FI-FO-20 NUREG 073 7. Sect ion ll.F.I.6 t E BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION S E CTIO,J_20 LAMES A FITZPATRICK NUCLEAR POWER PLANT MCKET NO.50-333 WORK SHEET pc 6 or ENVIRONMENT + D OCUM E N TATION REFERENCE QU ALIFIC ATIO N OUTST ANDING I DESCRIPTION METHOD ITEMS PARAMETER SPECIFIC ATIO N QUALIFICATION SPECIFICATION QUALIFICATION l EQUIPMENT l OPERATiNe 180 day, 101
- TIME syst E M C inment Atr.ospheric -
PLANT 10 N0'N/A TE M PER TUR E 40 104 110 44,les E CowPONENT Heat Trace Control Station e PRESSURE PSIA 14.7 15.0 ,4 Mt.NUFACTURER Thermon RELATIVE M ODTL NO! - (unb rwun at this 20-90 90-100 44,167 4 time) HUMIDITY % FUNCTION Heat trace control WATER SPRAY D 4 0'N . T.1. D . A CCIDEN T SE RVICE: 2H Gas Sample Line Heat Tracing R AD Ai' " 2.2x105 2.0x105 3 k
,ds LOC 4 TION. Reactor Building
- 1. 300' AelMe 40 years 164 N Flo0D LEVEL ELEV 227'-6" SU8 MERGENCE
( Not Applicable Harsh accident environments f.8 0V E FLOOD LEVEL h E NVIRONME NT SPECIFICATION 1.
- 2. Equipment required for post-accident monitoring r.,f containment H concentrations.
B A SIS ASING PROGRAW Install Dec., 1982 PASNY P.O. #81-4994 h[REF NCE SE CT10N T I O. ( TOTAL INTEGRATED 00SE )e 40 VR. NORM AL + ACCIDEN T AL DOSE Rev. O N /A NOT A PPLIC ABLE I.8=l0%
i i. 4 Section 30, Pages 4, 5, 6 4 l : i
- 1. The qualification program for the heat trace system for the ,
hydrogen monitoring system is in progress at Southwest !
! Research Laboratories. This program is expected to be completed !
l In the same time frame as final system installation presently l scheduled for December, 1982. l - i 3 l i i i I l 1 ' l l 4 1 i 1 i l ! l I l 1 , ,i ) i O I t t f 1 i 5 f 1 i ) , i 4 e i l 7 O : 5 i ! e 5 i l
--.-,,,,c,.--,n-,.---
O O TH I PSM . No. F1-80-20
#AkiG 0737, Section ll F.1.6 E BULLETIN 79- OIB SYSTEM COMPONENT EVALUATION JAMES A FITZPATRIOK NUCLEAR POWER PLANT SECTiON 30-
)OCKET NO.50-333 WORK SHE. i pc s of ENVIRONMENT + DOCUM E NTATION REFERENCE QU ALIFIC ATI O N OUTST ANDING
[EQUlPMENT DESCRIPTION PARAM ETER SPECIFI C ATIO N QUALIFIC ATION SPECIFICATION QUALIFICATION METHOD ITEMS
]
OPER ATIN G
- 180 days 101 TIME SYSTEM C nment Atmospheric - -- -- -
PLANT ID NO N/A
; T E M PE R TUR E 40-104 110 44.188 l 'F l COMPONENT RTD MEssuRE 'Y PSIA 14.7 15.0 44 LA1NUFACTU RE R Thermo Sensor N00EL NO ' Platinum RTD 20-90 90-I00 44.167 #
u lTV '4 F ONCTION Heat Trace Temp.
/
Control DEMhER ALIZE D WATER SPRAY Not Arplicable h p god. tem) T . 1. D . ACCIDE N T SERVICE. Hg Gas Sanpling Heat Tracing g g Y Rads 3 LOCATON- Reactor Bldg. U*"*'IC ASING 40 years 164 i' FLOOD LEVEL ELEV 227'-6" SUBMERGENCE Not Applicable ABOVE FLOOD LEVEL h E NVIR ONME NT SPECIFICATION
- 1. Harsh accident environment.
NO B ASIS A0140 Install Dec., 1982 PROGRAM PASNY P.O. #81-4934 INPUT Oh T ID ( TOTAL INTESRATED DOSE le 40 VR. NORM AL + ACCIDENTAL DOSE EP NCE SE CTION b Rev. O N /A NOT A P PLIC AgLE 7x10
_ _ . - . . . - - - - _ _ _ _ . _ = - - - - _ _ _ _ . _ _ _ -. - - i Section 30, Pages 4, 5, 6 l I
- 1. The qualification program for the heat trace system for the hydrogen monitoring system is in progress at Southwest l Research Laboratories. This program is expected to be completed !
In the same time frame as final system installation presently i scheduled for December, 1982. ' l l l l i s ( I I l l l i I f I i f i l i
- w THI Mod. No. F1-30-20 kUREC 0737. Section li.F.1.6 E BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION SE CTi0,4 30 JAVES A FITZPATRICK NUCLEAR POWER PLAtJT 3OCKET NO.50-333 WORK SHEET PG 6 or ENVIRONMENT + DOCUME NTATION REFERENCE QUALIFIC ATIO N OUTST ANDING EQUlPNENT DESCRIPTION METHOD ITEMS PARAM ETER SPECIFICATION QUALIFICATION SPECIFICATION QUALIFICATION OPE R A TING TIME 180 days 101 '
SrSTEM Con inwnt Atmospheric - - - PLANT 10 NO N/A TEM PER TUR E 40-104 110
'F 44.188 COMPONENT Heat Trace Cable
'8 "
MONUFACTURER Thermon R EL ATIVE N00EL N0' SSK-Mineral Insul. 20-90 90-100 44.167 NUMID 47Y '/. F UNC TION Heat tracing DEMINERALIZED WATER SPRAY ( Not Applicable
)
o NON. T. I . D . ACCIDEN T SERVICE; 2H Gas Sample Heat R ADI ATION g' Tracgng Rads 6.87x106 6.8x106 3 LOCATsON: Reactor Building Generic AtlNG 40 years , FLOOD LEVEL ELEV 227'-6" SUSMERGENCE Not Applicable
-h 180VE FLOOD LEVEL h E NVIRONME NT SPECIFICATiog 1.
2. Harsh accident environment. Equipment required for post-accident monitorina of containmant H e ncentrations. 2 BASIS A0lNS PROGRAM Install Dec., 1982 14PUT PASNY P.O. #81-4994 T ID ( TOTAL INTEGRATED DOSE le 40 VR. (RMAL + ACCIDEN TAL DOSE l Rev. O REF NCE SECTION N /A NOT A PPLIC ABL E 7x10
i l Section 30, Pages 4, 5, 6 l
@ 1. The qualification program for the heat trace system for the l
hydrogen monitoring system is in progress at Southwest j Research Laboratories. This program is expected to be completed i in the same time frame as final system Installation presently j scheduled for December, 1982. , ! i l t i I ! f ( I i
. I
~
l l 9 ; 1
_ <c TMl Mod. No. F1-81-47 NUREG 0661 SYSTEM COMPONENT EVALUATION SECTION 38 JAME AFT ATR CK JUCLEAR POWER PLANT MCKET NO.50-333 WORK SHEET pc i or ENVIRONMENT + DOCUMENTATION REFERENCE QU ALIFIC ATIO N OUTST ANDING !EQUlPMENT DESCRIPTION SPECIFICATION QUAllFICATION METHOD ITEMS PARAM ETER SPECIFICATION QUALIFICATION 0[,ER ATIN G 180 days >l80 days 108 213 Simultaneous None SYST E M Primary Containment and Lea ate NORMAL ACgNT PLONT #D NO 1(r1RTD-131 thru 146 T E M PE R TUR E 40-104 204 432 I AH+1 AQ 213 Simultaneous Nme
'F COMPON E NT RfD Assembly with Seal Assembly PRESSURE gg,7 gg,3 p ,7 74g g 79 MONUFACTURER Conax PSIA M ODEL NO. ! 01 with NU DiTV'4 20-90 90-100 100
- 213 Simultaneous None FCNCTION Temperature Monitoring OEMMER ALtZED WATER SPRAY Not Applicable )
ACCURACY- EC +l Fahrenhelt GE RVICE: Torus Temp. Monitoring D' 1.0x10 7 Rads 1.0x10 7 2.0x10 3 213 Sequential None LOCATION: Torus Room El. 234' AtlNG 0 years $40 years 164 213 Sequential None FLOOD LEVEL ELEV 227'-6" SUBMERGENCE Not Applicable -) ABOVE FLOOD LEVE'L E NVIRONME NT
- 1. Harsh accident environment.
NO SPECIFICATION 8 A SIS ASING Installed Feb. 1982 PROGRAM PASNY Change Order #5 to INPUT Teledyne Agreement Rev. O N EFE NCE SE CTION N /A NOT A PPLIC ABLE T.10 (TOTAL INTEGRATED 073E le 40 VR. NOh4 AL + ACCIDENT AL DOSE 7xl0
G O O TMI Mod. No. F1-61-47 NUREG 0661
.E BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION SECTION -"
JAMES A FITZPATRICK NUCLEAR POWER PLANT WORK SHEET pc 2 DOCKET NO.50-333 oF , ENVIRONMENT + DOCUM E N TATION REFERENCE QUALIFIC ATIO N OUTST ANDING EQUIPMENT DESCRIPTION QUALIFICATION METHOD ITEMS PARAMETER SPECIFIC ATION QUAUFICATION SPECIFICATION OPE R ATIN G 180 days 101 Y TIME SYST E M P wy Containment and _ - PLANT iD NO'1(r1TM-133 COMPONENT Datalogger I'f"'""' 40-120 N/A 44.170 sga 14.7 N/A 44,170 M ANUFACTURE R Acurex , MODEL NO : Autodata Ten /5 mum DiTY *4 10-60 N/A 44,170 Y F UNCTION Data Logging and Averaging Calculation oEMmERALiz[D WATER SPRAY ( Not Applicable
)
O M O'N. T . 1. D . AC CIDE NT S SE RVICE Torus Temperature Monitoring R AD A 10N 700 N/A 3 LOCATION: Control Room Admin. Bldg. El. 300' # ASING 40 years 164 FLOOD LEVEL ELEV N/A SUBMERGE NCE Not Applicable ' EBOVE FLOOD LEVEL E NVIR ONME NT I. Hild Environment NO SPE CIFICATION BASIS AGING PROGRAM Installed Feb., 1982 INPUT PASNY P.O. #80-NOTES: T I D. l TOTAL INTEGRATED DOSE )s 40 VR. NORM AL + ACCtDENT AL DOSE Rev. O SN EFENNCE SE CTION N /A NOT A P PLIC ABL E
Section 31, Page 2 O t
- 1. There was no requirement in NUREG 0661 that this equipment be Class IE qualified. Further review by the Authority is required in order to establish the qualification requirement for this equipment based on its function during a post-accident situation.
4 O l O l i
Section 32_- Indicating Lights and Meters
- 1. The two General Electric items listed on the attached sheets are presently undergoing review by the Authority. In June, 1982 the Authority will conduct a technical audit of General Electric's qualification records to confirm qualification for the mild environments in which these items are located.
- 2. At time of procurement, these components were determined to be the "best available" and are presently used in other Class 1E mild environment applications at JAFNPP. Utilization of these components are on an interim basis.
3 The application and environment in which these components are installed are in accordance with component's design specifications. 4 O l J l I I i l O
l w- - TMl Hod. No. FI-Generic JAME A TZPATR CK UCLEAR POWER PLANT SYSTEM COMPONENT EVALUATION SECTION 32 DOCKET NO.50-333 WORK SHEET pg , _-_or ENVIRONMENT + DOCUMENTATION REFERENCE QUALf FIC ATIO N OUT STANrj r NG EQUIPMENT DESCRIPTION METHOD i PARAMETER SPECIFICATION l QUALIFICATION SPECIFICATIGN QUALIFICATION ITEMS T 180 days 101 # SYS T E M Generic NORMAL AC NT l PLANT ID NO~ 8*!/A 1 i PE R TUR E N 40-120 N/A 44,170,I71 COMPONENT Indicating Llgist , , PRiBSURE gg,7 yf4 PteA 44.979.979 M ANUFACTURE R General Electric RELATIVE MTEL NO ET 16 10-60 ,N/A 44,170,171 k NUMIDITY , F U NC TION Indication DEMmER ALlZE D
, ,p ,y (- - - -
Not Applicable - - - - - - -- -
---)
C MON T. l . D . ACCIDE N T SE RVICE : Display indication , R ADI ATION Rads 700 N/A 3 LOCATION: , Ad n i E O 6 ASING 40 years 164 # FLOOD LEVEL ELEV h/A SusuERGENCE / Not Applicable N N / i ABOVE FLOOD LEVEL h E NVIRONME NT SPECIFICATION
- l. Mild Environme.it NO l B 4 518 l
AGvNe PROGRAM installed Feb., 1982 p NOT E S: ND Rdv. O gO tFE NCE SECTION T I D. ( TOTAL INTEGRATED DOSE l e 40 VR. NORM AL -+ ACCIDENT AL DOSE N /A NOT A PPLIC ABLE
TM1 Mod. No. F1-Generic SYSTEM COMPONENT EVALUATION s CTioN_32 AME A TZ ATR CK I4UCLE A R POWER PLANT JoCKET No.50-333 WORK SHEET PG 2 or ENVIRONMENT + D OCUM E N TAT 10A REFERENCE QU ALIFIC ATIO N OUTS T ANDING DESCRIPTION METHOD ITEMS l EQUIPMENT PARAM ETER SPECIFI C ATION QUALIFICATION SPECIFICATION QU AllFICATION OPE R A TING 180 days 10I TIME
'F SYSTEM Generic NORMAL ACCIDE N T PLANT ID NO N/A T E M rt R TUR E 40-120 N/A 44,I70,171 A 8F COMPONENT Vertical Indicating Heter PRES $URE PSIA yg 7 g gg,gg ,g79 (-
MONUFACTURE R General Electric MM L NO ! 180 HU DI Y % 10-60 N/A 44,170,171 F UNCTION Press / Level and Temperature indication OEM86ER ALIZE D WATER SPRAY
/
N Not Applicable p DE M ON. T. 1. D . A C CIDE N T SE RVICE Di splay Instrumentation R ADI ATION 700 N/A .Y 3 Rads LOCATION: ControlRoom/RelayRoom 40 years Admin Bldg. El. 300'/286 ASING 164 g_ FLOOD LEVEL ELEV N/A SU8 MERGENCE Not Applicable - ' - - ABOVE FLOOD LEVEL h E NVIR ONME NT SPECIFICATION
- 1. Mild Environment NO B ASIS AGING Installed Feb., 1982 PROGRAM PASNY P.O. #80-19148 INPUT Rev. O N REF NCE SECTION N /A NOT A PPLIC ABLE T I D ( TOTAL INTEGRATED DOSE ) e 40 VR. NORM AL -+ ACCIDEN T AL DOSE
l ' TM t Pbd. No. F 1 -Gene r t SYSTEM COMPONENT EVALUATION SE TION 33 hAME A TZ ATR CK UCLEAR POWER PLANT accKET NO.50-333 WORK SHEET pc i or ENVIRONMENT + DOCUME N T ATION REFERENCE QUAltFIC ATION OUT ST ANDING EQUIPMENT DESCRIPTION QUALIFICATION METHOD ITEMS PARAM ETER SPECIFI C ATION QUALIFICATION SPECIFICATION 0 ATIN G 180 days N/A 101 N/A N/A N/A SYS T E M Ceneric b[ PLANT ID NO N/A TE M PE R TUR E or 40-120 N/A N/A 44.170.171 207 Simultaneous None COMPONENT Control Relay PRESSURE MONUFACTURE R Ceneral Electric PSIA I4.7 N/A N/A 44,170,171 207 Simultaneous None RE AT M OD EL NO: CRl20B 10-60 N/A N/A 44,170,171 207 Simultaneous None F UNC TION Control Logic WATER SPRAY Not AppIIcable -
)
D M 0'N. T. I . D . AC CIDE N T SERVICE: Control Systems
" 6
" $'ds 700 N/A 1.0x10 3 207 Sequential None L OC AT ION:
ld 00'/ . A 40 years % years M 207 kquentlat he FLOOD LEVEL ELEV N/A SUBMERGENCE - N st Applicable - ABOVE FLOOD LEVEL h E NVIRONME NT I, HIld Environment NO SPECIFICATION BASIS AGING Installed Feb., 1982 lNPUT N TEL T I D ( TOTAL INTEGRATED DOSE le 40 VR. NORMAL + ACCIDENTAL DOSE Rev. O EF NCE SECTION N /A NOT A PPLIC ABLE
TMl Mod. No. Fl* Generic SYSTEM COMPONENT EVALUATION SECT.ON 33 1AME A T ATR CK JUCLEAR POWER PLANT WORK SHEET
)OCKET NO.50-333 pg , or
_' ENVIRONMENT + DOCUM E N TATION REFERENCE QU ALIFIC ATI O N OUTSTANDING J EQUlPMENT DESCRIPTION QUALIFICATION METHOD ITEMS { PARAM ETER SPECIFI C ATIO N QUALIFICATION SPECIFICATION
'" 180 days 4 101 T
SYSTEM Generic PLANT ID NO N/A TE M PER TUR E y COMPONENT Control Relay PRESSURE M PSIA 14.7 N/A 44,170,171 M ANUFACTURE R General Electric RELATI E 10-60 N/A 44,170,171 NODEL NO' 12HFA151A2F X F UNC TION Control Relay OEMWER ALIZE D / WATER SPRAY N Not Applicable ____p N/A QCCURACV- SPEC'N. DE M O T. l . D . ACCIDE N T SERVICE; Control Systems 700 N/A iY
"
- E',"d s 3 LOCATION:
ControlRoom/RelayRoom 40 years + AGING 164 Admin Bldg. El.300'/286: FLOOD LEVEL ELEV N/A SUBMERGENCE - Not Applicable _ _ _ _ .
)
080VE FLOOD LEVE'L h E NVIRONME NT SPECIFICATION
- 1. Hild Environment NO B A548 AGING Installed Feb., 1982 INPUT REF NCE SE CTION T ID ( TOTAL sNTEGRATED DOSE ) = 40 VR. NORM AL + ACCIDEN T AL DOSE Rev. O N /A NOT A PPLIC ABLE
Section 33, pages 2, 3, 4
- 1. Three General Electric relays listed on the Indicated sheets are presently undergoing review by the Authority. In June 1982, the Authority will conduct a technical audit of General Electric's qualification records to confirm qualification for the mild environments in which these components are installed.
- 2. At time of procurement, the components were determined to be the "best available" and are presently used in other Class IE mild environment appilcations at JAFNPP.
3 The appilcation and environment in which these components are installed are in accordance with the component's design spect-fications. O
.,W _ .
mw - ' _ - . . . . - - - - . . . . . . _ ,.. O TM 1 tbd No . F 1 -Gene r i c IE BULLETIN 79- OlB NUREG 0737 JAMES A FITZPATRICK NUCLEAR POWER PLANT SYSTEM COMPONENT EVALUATION SECTION 33 30CKET NO.50-333 WORK SHEET pc 3 og ENVIRONMENT + DOCUMENTATION REFERENCE QU ALIFIC ATIO N OUTSTANDING
! EQUIPMENT DESCRIPTION METHOD ITEMS i PARAM ETER SPECIFICATION QUALIFICATION SPECIFIC ATION QUALIFICAT ION OPE R ATING j TIME 180 days 101 ?
SYSTEM Generic PLANT ID NO N/A COMPONENT Control Relay If """ 40-120 N/A 44,170,171 PRESSURE 14.7 N/A 44,170,171 e PSIA M ANUFACTURE R General Electric M OD E L NO.
- CR2811A218L gg y, 10-60 N/A 44,170,171 >
F U NC TION Control Logic AT R SPRAY Not Applicable MON. T. I . D . A CCIDE N T SERVICE: Various R ADI ATION 700 N/A 3 P Rads LOCATION: Control Rrxwn/ Relay Rm a 40 years 164 V-AGING Admin Bldg. El .300'/t%' FLOOD LEVEL 'stEV N/A SUBMERGENCE ( Not Applicable
)
- 1. Hild Environment ABOVE FLOOD LEVEL QD E NVIRONME NT NO SPECIFICATION BASIS AGING Installed Feb., 1982 PROGRAM INPUT EF NCE SECTION T 10. ( TOTAL INTEGRATED DOSE )e 40 VR. NORM AL + ACCIDENT AL DOSE Rev. O N /A NOT A F PLIC A8LE
O Section 33, Pages 2, 3, 4
\
1
- 1. Three General Electric relays listed on the Indicated sheets are !
presently undergoing review by the Authority. In June 1982, the Authority will cond,uct a technical audit of General Electric's qualification records to confirm qualification for the mild environments in which these components are installed.
- 2. At time of procurement, the components were determined to be the "best available" and are presently used in other Class 1E mild environment applications at JAFNPP.
3 The application and environment in which these components are installed are in accordance with the component's design spect-fications. i i i i j
TMI Mod. No. F1-Generic SYSTEM COMPONENT EVALUATION SECTION 33 JAME A T ATR CK JUCLEAR POWER PLANT WORK SHEET po 1 og DOCKET NO.50-333 ENVIRONMENT + DOCUMENTATION REFERENCE QU ALIFIC ATIO N OUTSTANDING EQUIPMENT DESCRIPTION METHOD ITEMS PARAMETER SPECIFI C ATIO N QUALIFICATION SPECIFICATION QUAllFICATION OPE R ATIN G *- 180 days 101 TIME SYST E M Ceneric NORMAL ACCIDENT PLANT 10 NO' N/A 40-120 N/A 44,170,171 T TQPE R TURE COMPONENT. Trip Relay PRESSURE gg,7 gfg gg,97g y7, g M ANUFACTdRE R General Electric PSIA MODEL NO : 12HGAlllJ 10-60 N/A 44,170,17I & HU DiTY'4 F UNC TION Control Relay WATER SPRAY Not Applicable
)
, D MON. T. 1. D . AC CIDE N T SE RVICE: Various R ADI ATION 700 N/A 3 y Rads LOCATION:
Control Room / Relay Room Admin Bldg. El.300'/286e AGING 40 years 164 T FLOOD LEVEL ELEV N/A SUBMERGENCE - Not Applicable ABOVE FLOOD LE VE'L E NVIR ONME NT I. Hild Environment NO SPE CIFICA7t 04 , BASIS I i AGING Installed Feb., 1982 PROGRAM INPUT EF NCE SECTION T.8 0 ( TOTAL INTEGRATED DOSE )s 40 VR. NORM AL + ACCIDENTAL DOSE Rev. O N /A NOT A PPLIC ABLE
l Section 33, Pages 2, 3, 4 l
- 1. Three General Electric relays listed on the indicated sheets are presently undergoing review by the Authority. In June 1982,.the Authority will conduct a technical audit of General Electric's qualification records to confirm qualification for the mild environments in which these components are installed.
- 2. At time of procurement, the components were determined to be the "best available" and are presently used in other Class 1E mild environment applications at JAFNPP.
3 The application and environment in which these components are installed are in accordance with the component's design speci-fications.
. - - - - .._7, - - - , , _ _ , , , - - - - - _ _p--. _ , - ---- , , - - ,---,---,-,-y.,, . - ,-y--_
TM1 Mod. No. F1-Generic E BULLETIN 7 9 - 010 SYSTEM COMPONENT EVALUATION "" 737 JAMES A FITZPATRICK NUCLEAR POWER PLANT SE CTION_3 3 JocKET NO.50-333 WORK SHEET po__ s or ENVIRONMENT + DOCUME NT ATION REFERENCE QU ALIFIC ATIO N OUT ST ANDING
' E Q UIPME NT DESCRIPTION METHOD ITEMS
! PAR AM ETER SPECIFI C ATIO N QUALIFICATION SPECIFICATION QUAllFICATION OPER ATIN G TIME 180 days See Note 2 101 See Note 3 See Note 2 None SYSTEM Ceneric NORMAL ACCIDE N T l PL ANT ID NO N/A T PE R TUR E 40-120 N/A 222.8 44,170,171 223 S imul t aneous None COMPONENT Control Relay
"" " None PSIA 14.7 N/A 14.7 44,170,171 223 Simultaneous MONUFA CTURE R Struthers Dunn R E L ATIVE M OD E L NO: See betw 10-60 N/A 44,170,171 Simultaneous None NUMIDITY */, 98 223 F U NC TION Control logic DEMINER ALIZE D / Not Applicable N WATER SPRAY N #
EMON T. l . D . AC CIDE N T SE RVICE : Control Systems
''8" None 700 N/A 1.7x10 3 223 Sequential LOCATION:
Control Room / Relay Room Ne me ASING 40 years >40 years 223 Se4uential Admin Bldg. El.300'/286' 164 FLOOD LEVEL ELEV N/A SUBMERGENCE - Not Applicable - CBOVE FLOOD LEVEL E NVIR ONME NT l. Hild Environment SPE QFOTION
- 2. The entry for the qualification values / methods are pending successful completion of the IEEE 323-74. Test NO program in progress, see appendix I of Ref. 223.
Installed Feb., 1982 Basis 3 Reference 223 is adequate justification for continued use until successful completion of the IEEE 323-74. Hodel No's: AGINS 219XDX167NE 219BDX104NE PROGRAM 219FXXIl4NE INPUT 219XXF103NE 219XDX166NE NO T E S: ST A Rev. O REF NCE SECTION T ID ( TOTAL INTEGRATED DOSE )e 40 VR. NORM AL + ACCIDENTAL DOSE N /A NOT A PPLIC ABL E
~ ..
_ . - .O O IMI Mod. No. F1-Generic
" #3
.' E A TZ ATR CK UCLEAR POWER PLANT SYSTEM COMPONENT EVALUATION SECTION %
DOCKET NO.SO-333 WORK SHEET po I or E N VIR ONMEN T + DOCUME NTATION REFERENCE QU ALIFIC ATIO N OUTSTANDING EQUIPMENT DESCRIPTION METHOD ITEMS PA R AM ETER SPECIFI C ATIO N QUALIFICATION SPECIFICATION QUAllFICATION OPE R ATING TIME 180 days N/A 101 206 N/A N/A SYSTEM Generic
, 9 T[R WR E 40-120 N/A N/A 44,570,I71 206 Simultaneous one COMPONENT Contro' Switch
$ A I '7 N 'IN'IN b M M ANUFACTURE R General Electric M OD E L NO: CR29f;0-YS203E 10-60 N/A N/A 44,170,171 206 Simultaneous Nonc H %
F O MC TION Control Logic i OEMNER ALIZE D
- pp icable _ . _ _ ___
WATER SPRAY DE M 0'N . T . 1. D . AC CIDE N T SE RVICE - Various R ADI ATION 700 N/A 1.0x10 3 206 Sequential None Rads LOCATION: Control Room / Relay Room AGING Admin Bldg. El. 300'/286' 40 years >40 years 16A 206 Sequential None FLOOD LEVEL ELEV N/A SU8 MERGE NCE Not Applicable - - - - - - ~ - - 7 ABOVE FLOOO LEVEL E NVIR ONME NT 1. Hild Environment NO SPECIFICATION 84 SIS h AG' 'O Installed Feb., 1982 PROGRAM INPUT EF NCE SE CTION 7 I D. ( TOTAL INTEGRATED DOSE )s 40 VR. NORM AL + ACCIDEN TAL DOSE Rev* 0 N /A NOT A PPLIC ABL E
TN I Me d. W. F1-Generic NUSEG 0737 IE BU LLETIN 79- OlB SYSTEM COMPONENT EVALUATION SECTion " JAMES A FITZPATRICK NUCLEAR POWER PLANT WORK SHEET occKET NO.50-333 PG 2 OF ENVIRONMENT + D OCUM E N TATION REFERENCE QUALIFIC ATI O N OUTST ANDING l EQUlPME NT DESCRIPTION METHOD ITEMS PARAM ETER SPECIFI C ATIO N QUALIFICATION SPECIFICATION QUALIFICATION OPER ATING TIME 180 days N/A 101 206 N/A N/A SYS T E M Generic eak PL AN T ID NO N/A T PE R TUR E 40-120 N/A N/A 44,370,I71 206 Simultaneous None COMPONENT Keylock Switch l PRESSURE 44,170,171 PSIA 14.7 N/A N/A 106 Simultaneous None M ANUFA CTURE R General Electric RELATIVE MODEL NO.
- ER2940-UN200D NUMIDITY % 10-60 N/A N/A 44,170.171 206 Simultaneous None F UNC T!04 Eontrol Logic
/ Not Applicable WATER SPRAY g MON T. l . D . AC CIDE N T SE RVICE : Various R ADI ATION 700 N/A 1.0x10 3 206 Sequential None Rads L OC ATION.
Control Room / Relay Room Admin Bldg. El.300'/206' A83N8 40 year >40 years 164 206 Sequential None FLOOD LEVEL ELEY. N/A SUBMERGENCE / Not Applicable \
% /
ABOVE FLOOD LEVEL h E NVINONME NT SPE CIFICATION
- 3. Hild Environment l NO BAS 1S AGINS ROGRAM Installed Feb., 1982 INPUT
" T I D. l TOTAL INTEGRATED DOSE l e 40 VR. NORM AL + ACCIDENT AL DOSE Rev. O $N EFN!NCE SECTION N /A NOT A PPLIC ABLE
7 _ _ _ - _ _ _ _ _ - _ - _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ l O O O TH1 Mod. No. F1-Generic NUREG 0737 & 0661 UCLEAR POWER PLANT SYSTEM COMPONENT EVALUATION SE CTION 35 JAYE A TZPATR CK WORK SHEET pc i _or DOCKET NO.50-333 ENVIRONMENT 4 DOCUMENTATION REFERENCE QU ALIFIC ATIO N OUT ST ANDING EQUIPMENT DESCRIPTION QUAUFICATION SPECIFICATION QUAllFICATION METHOD ITEMS PA RAM ETER SPECIFI C ATIO N OPE R ATIN G 180 days >180 days 101 200,217 Simultaneous None TIME SYS T E M Ceneric e _ PLANT 10 NO N/A T E M PE R TUR E 135-150 308 400 44.57 200,217 Simultaneous None
'F C OM PONE N T insulating Sleeve M ANUFAC TURER Raychem 8 A
" ' 100 100 44,57.58 200,217 Simultaneous None N00EL NO: WCSF 40-90 F UNCTION Wire Splice OEM ERAL insulation Yes Yes 106 200,217 Simultaneous None DE M ON . T . 1. D . AC CIDE N T SE RVICE; Various R ADI ATION 8 8 8 Rads 1.09x10 1.0x10 2.0x10 3 200,217 Sequential None L O C ATION . Primary Containment, Reactor Bldg & Turbine Bldg ASING 40 years >40 years 164 200.217 Sequential None f FLOOD LEVEL ELEV 258'*6" SUBMERGENCE Not Applicable )
ABOVE FLOOD LEVEL E NVIRONME NT 1. Harsh accident environment ' gg SPE CIFICATION 8 A S18 A0 LNG OGRAM installed Feb., l'382 INPUT Rev. O NOT E S:
- ST A" T ID ( TOTAL INTEGRATED 00SE j e 40 VR NORM AL + ACCIDEN TAL DOSE EF NCE SECTION N /A NOT A PPLIC ABLE 8.5x106
U O TMl Mod. No. FI-Ceneric SYSTEM COMPONENT EVALUATION SECTION 35 bAME A T ATR CK JUCLEAR POWER PLANT WORK SHEET DOCKET NO.50-333 PG 7 oF ENVIRONMENT + DOCUM E NTATION REFERENCE QU ALIFIC ATIO N OUTSTANDING EQUIPMENT DESCRIPTION METHOD ITEMS P RAMETER SPECIFI C ATION QUALIFICATION SPECIFIC AT ION QUAllFICATION 180 days >180 days 7[ ^ " ' 101 215 Simultaneous None SYST E M Generic
' ACCIDENT NORMAL PLANT ID NO N/A T RWRE 135-150 308 340 44.57 215 Simultaneous None C OM PON E N T- Connector MANUFACTURER Thomas & Betts 8 ^ ~ '
" Simultaneous MODEL NO ! Te f ze l 40-90 100 100 4 .57.58 215 None F U NCTION Wire Connector OEM ER l ED ye5 106 Simultaneous None p ,y y,, 215 D M O'N T . 1. D . AC CIDE N T SERVICE: Various R ADI ATION 1.09x10 1.0x10 2.0x10 3 215 Sequential None Rads LOCATION: Primary Containment, Rea r Building Turbine AtlNG 40 years 40 years 164 215 Sequential None i
FLOOD LEVEL ELEV 227'-6" SU8 MERGENCE Not Applicable -h ABOVE FLOOD LEVEL E NVIRONME NT 1. Harsh accident elvironment NO SPECIFICATION B A SIS AGING Installed Feb., 1982 PROGRAM INPUT Rev. O N REF NCE SE CTION T 10. ( TOTAL INTEGRATED DOSE )= 40 VR.pRM AL + ACCIDEN TAL DOSE 8.5x10 N /A NOT A PPLIC ABLE
THI Mod. No. Fl-Generic JAME A TZ ATR CK JUCLEAR POWER PLANT SYSTEM COMPONENT EVALUATION s cTION 35 DOCKET NO.50-333 WORK SHEET PG t oF ENVIRONMENT + DOCUM E NTATION REFERENCE QU AllFIC ATI O N OUTST ANDING j EQUIPMENT DESCRIPTION METHOD ITEMS PARAM E TER G'ECIFI C ATIO N QUALIFICATION SPECIFICATION QUALIFICATION OPER ATIN G TIME 180 days >l80 days 301 158,159 Simultaneous None SYS T E M Ceneric f5 L COMPONENT Insulating End Cap
" 44,2BA MANUFACTURER Raychem SIA 1 7 25.7 66 158,15g Simultaneous None RE T1 E M ODEL No. : 10lA0ll thru 034-52 ,
20-90 90-100 100 44,I67 158,159 S unu t t aneous None DEMNER IZED Not Applicable
)
D MON. T. l . D . AC CIDE N T SERVICE. Various d 6.87x10 6.8x10 2.0x10 3 158,159 Sequential None L OC ATION Reactor Building
& Turbine Bldg. Generic ASING 40 years >40 years 164 158,159 Sequential tbne FLOOD LEVEL ELEV 227'-6" SU9 MERGENCE Not Applicable ABOVE FLOOD Lt.VE'L E NVf RONME NT 1. Harsh accident environment.
N0 3PECIFICAT104 2. Additional testing will proceed Power Authority revew will begin following receipt of Test Report. SASIS Completion estimeted by August, 1982. AGING Installed Feb., 1982 PROGRAM INPUT
- T I D. ( TOTAL INTEGRATED DOSE j e 40 VR. NORM AL + ACCIDEN T AL DOSE Rev. O N R EF NCE SE CTION N /A NOT A PPLIC ABLE 7x10 0
THI Mod. No. F1-Cenerlc NUREC 0737 & 0661 hAME A TZPATR CK UCLEAR POWER PLANT TEM COWONENT EVAWATION SECTION 35 accKET NO.50-333 WORK SHEET pg 5 og ENVIRONMENT + DOCUME N TATION REFERENCE QUALIFIC ATIO N OUT ST ANDING EQUIPME NT DESCRIPTION METHOD ITEMS PARAM ETER SPECIFIC ATIO N QUALIFICATION SPECIFICATION QUALIFICATION OPERATING TIME 180 days >180 days 108 220 Simultaneous Nme SYST E M Generic PLANT ID NO N/A COMPONENT. Insulating End Cap with WCSF Outer Sleeve PRESSURE CANUFACTURE R Raychem PSIA 14.2-16.7 59 7 80.7 44.58 220 Simultaneous Nme R ELATIVE N0 DEL NO: 101A0ll thru 094-52 NuMiDiTy % 40-90 100 100 44.57.58 220 Simultaneous None with WCSF Sleeve F UNC TION Wire Splice Insulation OEMINERALIZED WATER SPRAY yes yes 106 220 Simultaneous None i i EMDN. T. 1. D . ACCIDE N T 9ERVICE. Various R ADi AT' " g,ds 1.09x10 1.0x10 2.0x10 3 220 Sequential None LOCATION: Primary Containment i ASING 40 years >40 years 164 220 Sequential None l i , FLOOD LEVEL ELEV 258'-6" SU8 MERGENCE Not Applicable - 7 l Q caOvE FLODD LEVEL E NVIRONME NT 1. Harsh accident environment NO SPECTICATION BASIS AGING PROGRAW Installed Feb.. 1982 IN P tJ T l Rev. O N [ REFE NCE SE CTION T.l D. I TOTAL INTEGRATED DOSE ) = 40 VR. KRM AL + ACCIDEN TAL DOSE l N /A NOT A P PLIC ABL E 8.5x10 l
J-i i i Section 36 - Safety Relief Valve Position Monitoring System NUREG 0737, Section iI.D.3 ! 1. The Babcock and Wilcox Valve Monitoring System is presently undergoing qualification testing administered by a multi-utility task l
- group and the Babcock and Wilcox Power Generation Group. Preliminary scoping envi ronmental tests were completed in 1981. The present test schedule shows test completion and q.111fication report to be issued in October, 1982.
i 4 l 4 4 ) i i O t 1 h i. 4 l r 1 l 4 t i li i i
THI 1. No. F1-80-01 NURE. 0737, Section 18.D.3 1 E BULLETIN 79-018 SYSTEM COMPONENT EVALUATION SECTION 36 JAMES A FITZPATRICK NUCLEAR POWER PLANT WORK SHEET pc i
)OCKET NO.50-333 OF l
ENVIRONMENT + DOCUM ENTATION REFERENCE QUALIFIC ATIO N OUT S T ANDING l EQUlPMENT DESCRIPTION SPECIFI C ATIO N QUAllFICATION SPECIFICATION QUALIFICATION M'E T H O D ITEMS PA RAM ETER i oPERATINe 24 Hours l TIME 101 X-SYSTEM Nuclear Boiler PLANT #D NO 02VMC-071A*L TEM PER TUR E 40-120 N/A 44.171 ,(- er COMPONENT Signal Conditioning j Amplifier and Alarm Module PRESSURE 14.7 N/A 44.171 M-ca.NUFACTURE R Unhol t z-Dic kle PSIA (Babcock & Wilcox) R LAT E MODEL NO. : P22MHA-1 and HA 10-60 N/A 44.171 Y F UNC TION Data Acquisition DEMINER AllIE D - Instrument ^0 ' WATER SPRAY D MON. T. 1. D . AC CIDE N T (full Scale) SE RVICE: Relief Valve Position Indication R ADI Afl0N 700 N/A 3 Y RADS LOCATION: Relay Room Admin. Bldg. El. 286' A91MG 40 years yg f_ FLOOD LEVEL ELEy N/A =0 NOT APPLICABLE ?- SU8 MERGENCE A80VE FLOOD LEVEL h E NVIRONME NT 4. Hild environment. NO SPECIFICATION l 84568 l \ ASING ' PROGRAW Installed Dec., 1980 gypyy PASNY P.O. #79-17033 l NO T E S: A P NCE SECTION T I D ( TOTAL INTEGRATED DOSE )s 4 0 VR. NORM AL 4 ACCIDEN TAL DOSE N /A NOT A PPLIC ABLE
O O O_ TMI Mod. No. F1-80-06 NUREG 0737, Section ll.D.3.1 E BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION SECTION 36 JAMES A FITZPATRICK NUCLEAR POWER PLANT JOCKET NO.50-333 WORK SHEET P o. 2 or ENVIRONMENT + 00CUMENTATION REFERENCE QU ALIFIC ATIO N OUTST ANDING J EQUIPMENT DESCRIPTION QU ALIFICATION METHOD ITEMS PARAMETER SPECIFIC ATIO N QUALIFICATION SPECIFICATION l i j OPE R ATIN O 24 lburs S "I SYSTEM Nuclear Boiler PLANT s0 NO. 02VMY-07tA+L 02VHY-07]Al+Lt T EMPER TURE ' Remote Charge 'F 135-150 308 44. 57 COMPONENT. Preamplifler PRESSURE Unholtz-Dickle PSIA # 14 2-16.7 59.7 44. 58 MANUFAChR a Rcock & Wilcox) 22CA-2TR RELATIVE g MODEL NO.: -90 100 44. 57. 58 HUMIDITY */s, F U NCTION lmpedance Converters DEMINERALIZED /- WATER SPltAY Yes 106 ACCURACY- SPEC. +2% F.M ON. - T. I . C . ACCIDE N T SE RVICE : Relief Valve Position Indicaton R ADI ATION 1.0x10 g g 3 LOCATION: Drywell 295' A9tNG 40 Years 164 k FLOOD LEVEL ELEV 258'-6" SusMERGENCE : HOT APPLICABLE - 3 ABOVE FLOOD LEVE'L h E NVIRONME NT l. Harsh accident environment. NO SPECIFICATION BASIS A0lNS PROGRAM Installed Dec.. 1980 PASNY P.O. #79-17083 INPUT NOT E S:
- ST A "
T I D. ( TOTAL INTEGRATED DOSE )e 40 VR. NORM AL + ACCIDENTAL DOSE EF NCE SE CTION 6 N /A NOT A PPLIC ABLE 8.5x10
T M t Nd . No . F1-80-01 NUREG 0737, Section II.D.3.1 E BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION SECTION 36 1AMES A FITZPATRICK NUCLEAR POWER PLANT JoCKET NO.50-333 WORK SHEET po i or ENVIRONMENT + DOCUME N TATION REFERENCE QU ALIFIC ATIO N OUTSTANDING DESCRIPTION QUAllFICATION METNOD ITEMS f EQUIPMENT PA RAMETER SPECIFI C ATION QUALIFICATION SPECIFICATION t OPE R ATIN G 24 Hours 101 Y TIME syST E M Nuclear Boiler "O FE PLANT ID NO Generic I TEurERTURE g. i C OM PON E N T. Coaxial Slgnal Cable (Hardline) , PRESSURE PSIA 14.2-I6.7 59 7 44, 58 N M ANUF ACTU RE R Unholtz-Dickle (Babcock & Wilcox) M OOCL NO: 22CA-2TR 0-90 100 44, S', 58 X UM D TY '4 F U N C TION Accelerometer Signal Output Yes 106 ' WATER SPRAY Cable D MON. N/A T . 1. D . ACCIDE N T SERVICE: Relief Valve Posit ion Indicaton R ADI ATION 1.0x10 I ADS 1.09x10 3 (six months) LOCATION; Drywell 40 Years 164 FLOOD LEVEL ELEY 258'-6" SUBMERGENCE < NOT APPLICABLE
- ABOVE FLOOD LEVEL h E NVIR ONME N T I. Harsh accident environment.
wo SPECIFICATION BASIS AGING PROGRAM installed Dec., 1980 INPUT PASNY P.O. #79-17083 NO T E S: 4 N T O D. ( TOTAL INTEGRATED DOSE l e 40 VR. NORM AL 4 ACCIDEN TAL DOSE
+ tF NCE SECTION 6 N /A NOT A PPLIC ABLE 8,$xgo
. O TMl Mod. No. F1-80-Ol O
l NUREG 0737. Section 11.0.3.1 1E BULLETIN 79- OlB SYSTEM COMPONENT EVALUATION SECTION w JAVES A FITZPATRICK NUCLEAR POWER PLANT WORK SHEET 6 DOCKET NO.50-333 PG OF I l ENVIRONMENT + DOCUME N TATION REFERENCE QU AllFIC ATI O N OUT ST ANDING EQUIPNENT DESCRIPTION SPECIFICATION QUAllFICATION METHOD ITEMS PARAMETER SPECIFICATION QUALIFICATION OPER ATING ? TIME 24 Hours 108 SYS T E M Nuclear Boller PLANT ID NO 02VME-07tA+L 02VME-071Al+L1 , T E M PE R TUR E #
*F 135-150 308 44. 57 COMPONENT. Accelerometer
4 SIA 14.2-16.7 59.7 44. 58
. M ANUFACTURE R Endevco (B &W)
RELATIVE , MODEL NO: 2273AM20 40-90 100 44. 57. 58 + NUMIDITY % F UNC TION Acoustic Valve g DEMNER allied Monitor Yes 106 WATER SPRAY i ACCURACY- SPEC. 12mv/g DE M ON . T . 1. D . ACCIDEN T SERVICE: Relief Valve Position Indication R ADI ATION g g s 1.09sl0 1.0x10 3 RADS Drywell (six months) L OC AT10N: AelNG 40 Years 164 4 4 NOT APPLICABLE -w FLOOD LEVEL ELEY 258'-6" SU8 MERGENCE A8 0VE FLOOD LEVEL h E NVIRONME NT l. Harsh accident environment. yn SPECirlCATION 8 ASIS AGING PROGRAM Installed Dec. 1980 PASNY P.O. #79-17083 INPUT NOTEt N T l D. ( TOTAL INTEGRATED DOSE )e 40 VR. N0hM AL + ACCIDEN T AL DOSE tF NCE SE CTION N /A NOT A PPLIC ABLE 8.5xio '
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