ML20155E783
| ML20155E783 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 09/15/1998 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20155E776 | List: |
| References | |
| PROC-980915, NUDOCS 9811050178 | |
| Download: ML20155E783 (123) | |
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O JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES iNFORMATION COPY, THIRD INHiRVAL PLAN Revision 2
<- Effective Date IO'A'-CY i Prepared by: aN< 2 Date: 9-N-7I T. Zwick(I)T Engineer Reviewed by: k Date: 9 - E - 16 U
Approved by ( , da Date: 9[I4[$f d s[ Tech Services Dept. Manager or ey L- Date: 9 i /P 1
M. Colomb / Shqjxecutive Officer
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Table of Contents
- 1. I NTR O D U CTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 i-
- 2. APPLICABLE DOCUMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
- 3. SYSTEM CLASSIFICATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
- 4. INSERVICE TESTING PROGRAM FOR PUMPS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
- 5. INSERVICE TESTING PROGRAM FOR VALVES . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
- 6. SYSTEMS SUBJECT TO TESTIN G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 sd
) APPENDIX A - PUMP TESTING PROGRAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 APPEN DIX B VALVE TESTING PROGRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 d
APPENDIX C -
SUMMARY
OF CHANGES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 120 J
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NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT p INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES
1.0 INTRODUCTION
Revision 2 of the James A. FitzPatrick ASME Inservice Testing (IST) Program will be in effect through the end of the third interval unless changed and re-issued for reasons other than the routine update required at the start of the fourth interval in accordance with 10 CFR 50.55a(f). The fourth inspection interval begins in September of 2007.
This document outlines the IST Program for J.A. FitzPatrick based on the requirements of Section XI of the ASME Boiler and Pressure Vessel Code,1989 Edition (the Code).
The 1989 edition of the Code specifies that the rules for the inservice testing of pumps and valves are stated in the ASME/ ANSI Operations and Maintenance (OM) Standards, Part 6, " Inservice Testing of Pumps in Light-Water Reactor Power Plants," and Part 10
" Inservice Testing of Valves in Light-Water Reactor Power Plants." An exception was taken in 10 CFR 50.55a to OM-10 related to leakage rate testing of containment isolation valves. References in this document to OM-1, OM-6, and OM-10 correspond to the 1987 ASME/ ANSI OM Standard Parts 1,6, and 10, respectively, unless otherwise noted. For OM-6 and OM-10, the applicable edition includes the 1988 OMa addenda.
2.0 APPLICABLE DOCUMENTS This IST Program was developed in accordance with the requirements of the following l documents: i
. Title 10, Code of Federal Regulations, Part 50 l
. Final Safety Analysis Report, J.A. FitzPatrick Nuclear Power Plant e J. A. FitzPatrick Technical Specifications
. ASME Boiler and Pressure Vessel Code,Section XI,1989 Edition
. ASME/ ANSI Operations and Maintenance Standard, Parts 1, 6, 10, 1987 Edition including the 1988 OMa addenda
- Other documents used for guidance in the development of the IST Program are listed below:
. NRC Regulatory Guide 1.26, " Quality Group Classifications and Standards for Water , Steam , and Radioactive-Waste- Contaminating Components of Nuclear Power Plants" i
}/ . Standard Review Plan NUREG 0800, Section 3.9.6, " Inservice Testing of Pumps and Valves" Rev. No. 2 Page .L of123 l
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NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES
. NRC Generic Letter 89-04, " Guidance on Developing Acceptable Inservice Testing Programs"
. NRC Minutes of the Public Meetings on Generic Letter 89-(M e NUREG-1482, " Guidelines for Inservice Testing at Nuclear Power Plants"
. Safety Evaluation of Certain Relief Requests from Section XI of the American Society of Mechanical Engineers Code for the James A. Fitzpatrick Nuclear Power Plant, dated May 2,1991.
3.0 SYSTEM CLASSIFICATION l
In the NRC Safety Evaluation dated May 2,1991 for the J.A. FitzPatrick Section XI l
pressure test program, the NRC evaluated the deletion of certain Class II-augmented l air / nitrogen systems from the inservice inspection program. These systems included the Drywell Inerting, CAD, and Purge system, the Containment Differential Pressurization system, the Breathing, Instrument, and Service Air system, the Containment Hydrogen Monitoring system, and the Standby Gas Treatment system. The NRC's evaluation found, based on a review of the regulations, the ASME Code, and regulatory guides, that there is no basis for requiring inservice inspection of these particular systems.
Although this finding related only to the hydrostatic testing of these systems, the basis for classification of these systems would also be applicable to the IST program. Therefore, l in accordance with NUREG-1482, components in these systems are not required to be in l the IST program. They may be included in the IST program and designated as non-Code or augmented componcnts. Relief requests for non-Code components may be implemented without NRC evaluation and approval.
Containment isolation valves in the systems listed above have been included as Category l A valves in the IST program. Other safety-related components in those systems have also been included in the IST Program and identified as augmented components. In addition to the systems listed above, portions of the Main Steam leakage Control System contain valves that are not within the scope of 10 CFR 50.55a. These valves have also been l classified as augmented in the J.A. FitzPatrick IST Program.
Similarly, the Diesel Generator system is a non-Code Class system as identified in Regulatory Guide 1.26. The J.A. FitzPatrick ISI Program has classified the following Diesel Generator subsystems as augmented Class III:
. Emergency Diesel Generator Fuel Oil Transfer
. Emergency Diesel Generator Fuel Oil Service Rev. No. 2 Page _4_ of 123
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT r INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES V] i
- Emergency Diesel Generator Combustion Air
- Emergency Diesel Generator Cooling Water
- Emergency Diesel Generator Air Start These subsystems also meet the definitions for skid-mounted components and enponent subassemblies as discussed in NUREG-1482. In NUREG-1482, the NRC has determined that the testing of the major component is an acceptable means for verifying the operational readiness of the skid-mounted and component subassemblies. This is acceptable for both Code Class and non-Code Class components. Therefore, based on the NRC position in NUREG-1482 and the existing Technical Specification requirements, operability tests, preventative maintenance activities and design redundancy, the components in the six Emergency Diesel Generator subsystems listed above, will not be included in the IST Program.
4.0 INSERVICE TESTING PROGRAM FOR PUMPS f
b> 4.1 Code Compliance This IST Program is based on the requirements of OM-6 as referenced by Subsection IWP of the 1989 Code edition. Where these requirements have been determined to be impractical, conformance would cause unreasonable hardship without any compensating increase in safety, or an alternative test provides an acceptable level of quality and safety, relief from Code requirements is requested pursuant to the requirements of 10 CFR 50. 55a (f)(6)(i).
4.2 Allowable Rances of Test Ouantities The allowable ranges for test parameters as specified in OM-6 Table 3 will be used for all measurements of pressure, flow, and vibration except as provided for in specific relief requests.
4.3 Testine Intervals The test frequency for pumps included in the IST Program will be as set forth in OM-6, Section 5.1. A band of V 25 percent of the test interval may be applied to a test schedule as allowed by the J.A. FitzPatrick Technical Specifications to provide for operational flexibility.
n Rev. No. 2 Page .1_ of122
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES 4.4 Pumn Procram Table Appendix A lists those pumps included in the IST Program with references to parameters to be measured and applicable requests for relief.
4.5 Relief Reauests for Pumn Testine Appendix A includes relief requests related to pump testing.
5.0 INSERVICE TESTING PROGRAM FOR VALVES 5.1 Code Compliance This IST Program is based on the requirements of OM-10 as referenced by Subsection IWV of the 1989 Code edition. Where these requirements have been determined to be impractical, conformance would cause unreasonable hardship without any compensating increase in safety, or an alternative test provides an acceptable level of quality and safety, relief from Code requirements is requested pursuant to the requirements of 10 CFR 50. 55a (f)(6)(i).
5.2 Testin2 Intervals The test frequency for valves included in the IST Program will be as set forth in l OM-10, Section 4.2, 4.3, and 4.4. A band of V 25 percent of the test interval i may be applied to a test schedule as allowed by the J.A. FitzPatrick Technical Specifications to provide for operational flexibility. Where quarterly testing of valves is impractical, testing may be performed during cold shutdown or refueling outage periods as permitted by OM-10, Sections 4.2.1.2 and 4.3.2.2.
5.3 Stroke Time Acceptance Criteria The acceptance criteria for the stroke times of power-actuated valves will be as set forth in OM-10 Section 4.2.1.4 and 4.2.1.8 and NUREG-1482 Section 4.2.7.
5.4 Check Valve Testine Full-stroke exercising of check valves to the open position using system flow requires that the maximum required accident condition flow be used and measured. Deviations to this requirement must satisfy the requirements of Generic Letter 894)4.
O Rev. No. 2 Page _6._ of 123
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT Q
O INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES 5.5 Containment Isolation Valves Containment isolatirn valves which do not provide a reactor coolant system pressure isolation function are tested in accordance with OM-10 Section 4.2.2.2.
In addition, as required by 10 CFR 50.55a(b)(2)(vii), containment isolation valves are analyzed in accordance with OM-10 Section 4.2.2.3(e) and corrective action is applied in accordance with OM-10 Section 4.2.2.3(f).
- 5.6 Valve Procram Table Appendix B lists those valves included in the IST Program with references to .
required testing, respective test intervals, applicable requests for relief and cold I shutdown and refueling outage justifications.
5.7 Relief Reauests for Valve Testine Appendix B includes relief requests, cold shutdown justifications, and refueling outage justifications related to valve testing.
6.0 SYSTEMS SUBJECT TO TESTING SYSTEM # SYSTEM NAME DRAWING #
01-125 Standby Gas Treatment FM-48A 02-2 Reactor Water Recirculation FM-26A 02-3 Nuclear Boiler Instrumentation FM-47A 03 Control Rod Drive FM-27B i 07 Neutron Tip Monitors FM-119A 10 Residual Heat Removal FM-20A,B 11 Standby Liquid Control FM-21 A l 12 Reactor Water Cleanup FM-24A 13 Reactor Core Isolation Cooling FM-22A i 14 Core Spray FM-23A 15 Reactor Building Closed Loop Cooling FM-15A,B
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16-1 Leak Rate Analyzer FM-49A lV e
19 Fuel Pool Cooling FM-19A i
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NEW YORK POWER AUTHORITY l
JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES SYSTEM # SYSTEM NAME DRAWING #
20 Radioactive Waste FM-17A 23 High Pressure Cooling Injection FM-25A 27 Containment Atmosphere Dilution FM-18A,B,D 29 Main Steam FM-29A 34 Feedwater FM-34A 39 Breathing, Instrument & Service Air FM-39A 46 Service & Emergency Service Water FM-46A,B 66 Reactor Building Service Ventilation FM-10H (Service Water) 70 Control Room Service & Chilled Water FB-35E O
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NEW YORK POWER AUTHORITY -
JAMES A. FITZPATRICK NUCLEAR POWER PLANT 1 l
l O INSERVICE TESTING PROGRAM FOR PUMPS AND VAINES APPENDIXA l
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PUMP TESTING PROGRAM O
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JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIXA PUMP TESTING PROGRAM Table of Contents Pump Table Explanation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ................11 Pu mp Tabl e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 Relie f Requests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 PRR-01: Ge ne ric . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 PRR-02R1: Standby Liquid Control . . . . . . . . . . . . . . . . .........................15 PRR-03: Standby Liquid Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 PRR-04: Co re S p ray . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 PRR-05R1: Emergency Service Water . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 PRR-06: RHR/ Emergency Service Water . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 t
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l NEW YORK POWER AUTHORITY l' JAMES A. FITZPATRICK NUCLEAR POWER PLANT l p INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES l V APPENDIXA PUMP TABLE EXPLANATION Summary ofInformation Provided The Pump Table provides the following information:
System
- Individel pump identifier Class The drawing on which the pump appears Drawing coordinates Speed (", if variable Differential pressure ("
Discharge pressure (" (positive displacement pumps)
Flow rate"'
Vibration ("
Test interval
("
These parameters are each addressed with either an "X" indicating the parameter is measured, an "X" with a PRR notation indicates relief is requested to modify or eliminate measurement of the parameter. A blank indicates that measurement of the respective parameter is not applicable.
Pump Relief Reauests PRR-XX refers to relief requests for the Pump Testing Program. Each pump request for relief provides the following information:
System Individual pump identifier Code Classification Safety Function Code test requirement for which relief is requested
- Basis for relief
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- Proposed alternate testing Rev. No. 2 Page 11 of 123
NEW YORK POWER AUTHORITY JAMES A FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES PUMP TABLE DRAWING DWG DIFFERENTIAL DISCHARGE FLOW PdPSECTION PUMP ID CLASS NUMBER CO-ORD SPEED PRESSURE PRESSURE RATE V18 RATION FREQUENCY 10P-1 A 3 F M-208 86 X PRRM A X 14UARTERt.Y 10P-1 B 3 FM-2LB B4 X PRRM X X 14UARTERLY 10P-1C 3 FM-208 C4 X PRRM X X 14UARTERLY 10P-1D 3 F M-208 C4 X PRR M X X 14UARTERLY l
10P-3A 2 FM 20A C-7 X PRR41 X X 14UARTERLY 10P-38 2 FM-20A C4 X PRR41 X X 1-QUARTERLY 10P-3C 2 FM 20A C-7 X PRR41 X X 14UARTERLY 10P-30 2 FM-20A C4 X PRR41 X X 14UARTERLY 11P-2A 2 FM-21A D4 X X PRRC2A1 X PRR43 1-QUARTERLY 11P-28 2 FM-21A B4 X X PRR02A1 X PRR43 14UARTERLY l
l 14P-1 A 2 FM-23A C4 X PRR41 X X 14UARTERLY PRR44 l
l 14P-19 2 FM-23A C-3 X PRR41 X X 14UARTERLY PRR44 l
23P-19 2 FM-25A E4 X X X 14UARTERLY 23P-1M 2 FM 25A E-4 X X X X 14UARTERLY 46P-2A 3 FM468 D4 X PRR45R1 X X 140ARTERLY PRRM 46P-2B 3 FM44 C4 X PRR45R1 X X 14UARTERLY PRRM 12 OF 123 REV NO
! I NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT I 1
O INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES I APPENDIXA l
Pumo Relief Reauests 1
1 l PRR-01 1 1 i l
SYSTEM: VARIOUS PUMPS: Various l CLASS: Various FUNCTION: This is a generic relief request.
TEST REQUIREMENT: OM-6 Section 4.6.2.1, if the presence or absence of liquid m a gage line could produce a difference of more than 0.25% in the indicated value of the measured pressure, means shall be provided to assure or determine the presence or absence of liquid as required for the static correction used.
3 (Q BASIS FOR RELIEF: In accordance with OM-6 Section 4.6.2.2, the pump differential pressure may be determined by the difference in the pressure at a point in the inlet pipe (suction pressure) and the pressure at a point in the discharge pipe (discharge pressure). When the requirements of OM-6 Section 4.6.2.1 are applied to the measurement of pump suction pressure, the 0.25% limit is overly restrictive since the pump suction pressures are typically at relatively low levels.
Compliance with this requirement could complicate venting procedures and introduce unnecessary health physics risks associated with handling and disposing of radioactive contaminated water with no commensurate gain or improvement of test reliability.
In most cases, the pump discharge pressure exceeds the suction pressure by at least a factor of five (5). This being the case, a 0.25% error introduced into the suction pressure measurement results in an error of 0.0625% in the differential pressure calculation. This is insignificant in light of the potential 6% error (2% full scale accuracy and full scale range of three times the reference value) allowance applied to both the suction and discharge pressure measurement in OE6 Section 4.6.
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NEW YORK POWER AUTIIORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX A l
Pumn Relief Reauests i
f PRR-01 (Continued)
ALTERNATE TESTING: If the presence of absence of liquid in a gauge line used for sensing f
- pump suction pressure could produce a difference of more than 0.25% in the calculated value of the pump differential pressure, means shall be provided to ensure or determine the presence or absence ofliquid as required for the static correction used.
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APPENDIXA
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Pumo Relief Reauests i 1 1 PRR-02R1 l
. SYSTEM: STANDBY LIQUID CONTROL (SLC) !
PUMPS: 11P-2A, B l CLASS: 2 FUNCTION: These pumps inject borated water into the reactor vessel as an alternate means for negative reactivity addition and reactor shutdown.
I TEST REQUIREMENT: .OM-6 Section 4.6.5, specifies the use of a rate or quantity meter installed in the pump test circuit when measuring flow rate.
fq OM-6 Section 4.6.1.1 specifies the instntments used for flow rate Q measurement must be accurate to within i 2% of full scale reading on the instrument.
BASIS FOR RELIEF: The SLC test loop is not equipped with flow instrumentation and the only practical means of determining flow rate is to monitor the change of level in a test tank to which water is being pumped. The installed tank has a capacity of only 210 gallons and - is capable of acconunodating less than 5 minutes of pump operation at rated conditions (2 50 GPM).
Due to limitations of pumping time and human factors related to measuring the change in test tank water level, the accuracy of flow rate determination cannot be verified to be within i2% as required by tne Code. Historically, the calculated flow lates are within 0.95 to 1.10 of reference flow rate (54.5 gpm).
ALTERNATIVE TESTING: The flow rate of the SLC pumps will be determined by measuring the change in water level in the test tank during a period of pump operation at the reference discharge pressure over a period of at least two (2) minutes. The level change will be converted to flow rate and evaluated in r]
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accordance with analysis and evaluation criteria specified in OM-6, Section 6, as applicable.
Rev. No. 2 Page 15 of123 l
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIXA Pump Relief ~Reauests PRR-03 SYSTEM: STANDBY LIQUID CONTROL (SLC)
PUMPS: 11P-2A, B CLASS: 2 FUNCTION: These pumps inject borated water into the reactor vessel as an alternate means for negative reactivity addition and reactor shutdown.
TEST REQUIREMENT: OM-6 Section 4.6.1.6, the frequency response range of the vibration measuring transducers and their readout system shall be from one-third minimurn pump shaft rotational speed to at least 1000 Hz.
BASIS FOR RELIEF: The nominal speed of the SLC pamps is 520 RPM, which correlates to a rotational frequency of 8.67 Hz. OM-6 Section 4.6.1.6 requires the frequency response range of the vibration measuring transducers and their readout system to be accurate to i 5% full scale over the range of 2.89 - 1000 Hz.
The Authority has instruments for use during surveillance testing with certified accuracy ofi 5% full scale over a range of 5-2000 Hz. Calibration is verified accurate using a system test methodology over a range of 10-1000 Hz in units of displacement (mils p-p) and 6.5-1000 Hz in units of velocity (ips peak). The system test verification is limited by the capability of the calibration shaker system to accurately sustain vibration at meaningful amplitudes outside the tested frequencies. The certified calibration i 5% range is arrived at through addition of individual transducer and meter inaccuracies over the stated frequency range.
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NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT f~ INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES i.
APPENDIXA Pumo Relief Reauests 1
I PRR-03 (Continued) l The instrument lower frequency response limits are a result of l high-pass filters installed to eliminate low frequency elements associated with the input signal from entering the process of single and double integration. These filters prevent low frequency electronic noise from distoning reading in the resultant units (ips, l mils). As a side effect, any actual vibration occurring at low l frequencies is filtered out. This is a necessary trade-off, as 1 my of l electronic noise at 2.5 Hz translates to approximately 62.6 mils p-p with the accelerometer used with these instruments, at a nominal sensitivity of 50 mv/g.
l The Authority has extensively researched this issue concernmg Code compliance and intent, and strongly feels that, for these pumps, procurement of equipment capable of meeting the Code y required accuracy is impractical with little or no benefit.
Instrumentation capable of meeting the Code for these pumps is cumbersome, difficult to operate, prone to human error, costly to purchase and extensive to calibrate. The number of vendors that supply instrumentation accurate at these frequencies is limited, and there are even fewer vendors capable of performing the required ;
calibration services. Most standard qualified calibration laboratories provide calibration services only to a minimum of 10 Hz.
In addition to the impracticality of procuring the instruments, the Authority feels that the instmments presently used are adequate to assess the condition of these pumps. The manufacturer of these pumps, Union Pump Company, Battle Creek, Michigan, has stated that these pumps, being of a simplified reciprocating design, have no failure mechanism that would be revealed at frequencies less than shaft speed. Union Pump has stated that all failure modes of this pump resulting in increasing vibration will be manifested at shaft speed frequency or harmonics thereof. In light of the information provided by Union Pump, monitoring sub-synchronous vibration for these pumps is not needed, but super-synchronous (x readings will provide meaningful infonnation in the detection of imminent machinery faults.
Rev. No. 2 Page 17 of123
NEW YORK POWER AUTHORITY ;
JAMES A. FITZPATRICK NUCLEAR POWER PLANT ;
INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES l
APPENDIX A Pump Relief Requests PRR-03 (Continued)
A search of the INPO NPRDS database has revealed only one l failure reported for pumps of this or similar design whose discovery mentioned increased vibration levels. The cited cause of the failure was improper endplay set leading to gearing failure.
Failures of this type would normally be detected at running (shaft)
I speed frequency, harmonics thereof, or non-harmonic super-
! synchronous bearing defect frequencies. It should also be noted that these are standby pumps that are normally operated only during pump and valve testing. In the unlikely event this system is
- required to fulfill its design function, only one of the two redundant pumps need operate for a period of 23 to 125 minutes.
In addition to vibration monitoring performed for the IST Program, l
these pumps are included in the Authority's Rotating Equipment Monitoring Program. Vibration spectral data is periodically collected and analyzed for the pump and gear motors in addition to those required by the Code. The equipment used by the Rotating Equipment Program is certified accurate to i 5% over a frequency range of 5-2000 Hz and is also limited by high-pass integrating l
filters, but allows for discrete frequency analysis and trending i using FFTs. Vendor specifications state that this equipment should provide fairly accurate data down to 2 Hz in units of acceleration l
(g peak) by using the raw transducer signal, negating the need for integration. Study of low frequency spectra taken in g peak with l
these instruments has revealed no distinct sub-synchronous peaks above the noise floor acceleration signal.
In light of their rigorous testing and limited design run time, it is not likely that a minor mechanical fault would prevent these pumps from fulfilling their design function and unlikely that development of a major fault would go unnoticed.
O Rev. No. 2 Page 18 of123
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT
- w INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES l
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APPENDIXA I Pump Relief Reauests i
PRR-03 (Continued)
In conclusion, the Authority feels that the use of high quality, commercially available vibration monitoring equipment calibrated to be at least accurate to 5% full scale over a range of 6 Hz to
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500 Hz (nominal shaft speed - 8.67 hz) is an appropriate method of l monitoring the mechanical condition of the SLC pumps. Such l instruments will provide meaningful and useful measurements over the frequency range in which the pump faults will develop and manifest. This meets the intent of the Code and certainly will neither adversely impact system reliability nor the health and safety of the general public. In addition, it relieves the Authority of the burden and . expense involved in the procurement, calibration, training and certification associated with obtaining new equipment I that is simply not needed to adequately assess the condition of the p SLC pumps, j O ALTERNATE TESTING: The vibration measurements will be taken using instrumentation i accurate to i 5% full scale over a frequency response range of 6 Hz to 500 Hz. The data will be evaluated per OM-6 Section 6.
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1 NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIXA Pump Relief Requests l PRR-04 SYSTEM: CORE SPRAY (CSP)
PUMPS: 14P-1 A, B l
CLASS: 2 FUNCTION: Pump cooling water from the suppression pool to the reactor in the event of a LOCA.
l TEST REQUIREMENT: OM-6 Section 4.6.1.2(a), the full scale-range of each analog instrument shall be not greater than three times the reference value.
BASIS FOR RELIEF: The differential pressure for the Core Spray pumps is calculated using the installed suction and discharge pressure gauges. The l
suction pressure gauge is designed to provide adequate suction i pressure indication during all expected operating conditions. The full-scale range,60 psig, is sufficient for a post-accident condition when the torus is at the maximum accident pressure. This, however, exceeds the range limit for the suction pressure under the test condition (approximately 5 psig).
The installed suction pressure gauge and discharge pressure instrumentation loop are calibrated to within 2% full scale accuracy. The full-scale range of the pump discharge pressure l instrumentation loop is 500 psig. Pump discharge pressure during testing is typically 300 psig. Thus the maximum variation due to inaccuracy in measured suction pressure is i 1.2 psi and in measured discharge pressure is i 10 psi. Thus, the differential pressure would be 295 i 11.2 psi or an inaccuracy of 3.8%. If the full scale range of the suction pressure gauge was within the Code allowable of 3 times the reference value or 15 psig, the resulting ,
differential pressure measurement would be 295 10.3 psi or an '
inaccuracy of 3.5%. Thus the increase in inaccuracy of 0.3% is insignificant and does not warrant the additional manpower and exposure required to change the suction pressure gauge for test i purposes.
Rev. No. 2 Page 20 of 123 l
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT
,q INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES U APPENDIXA Pump Relief Reauests PRR-04 (Continued)
In addition, the Code would allow a full-scale range for the discharge pressure measurement of 900 psig. This would translate into a differential pressure measurement of 295 i 18.3 psig or an inaccuracy of 6.2%. The existing measurement is significantly better than the maximum Code allowable inaccuracy.
ALTERNATE TESTING: The existing installed plant suction pressure gauges will be used to determine the pump differential pressure for testing of the Core Spray pumps.
I' O
Rev. No. 2 Page 21 of123
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIXA Pump Relief Reauests PRR-05RI SYSTEM: EMERGENCY SERVICE WATER (ESW)
PUMPS: 46P-2A, B CLASS: 3 FUNCTION: These pumps provide cooling water for safety-related heat loads during > loss-of-coolant design basis accident.
TEST REQUIREMENT: OM-6 Section 5.2(b), the resistance of the system shall be varied until the flow rate equals the reference value. The pressure shall then be determined and compared to its reference value.
Alternatively, the flow rate can be varied until the pressure equals the reference value and the flow rate shall be determined and compared to the reference flow rate value.
BASIS FOR RELIEF: Emergency Service Water (ESW) systems are designed such that the total pump flow cannot be adjusted to cne finite value for the purpose of testing without adversely affecting the system flow balance and Technical Specification operability requirements.
These pumps must be tested in a manner that the service water loop remains properly flow balanced during ar.d after the testing and each supplied load remains fully operable per Technical Specifications to maintain the required level of plant safety during plant operation.
The ESW water system loops are not designed with a full flow test line with a single throttle valve. The flow therefore cannot be throttled to a fixed reference value every time. Total pump flow rate can only be measured using the total system flow indication ;
installed on the common supply header. Only the flows of the j serviced components can be individually throttled. Each load is j throttled to a FSAR required flow range which must be satisfied l for the load to be operable. All loads nre aligned in parallel, and l 11 receive ESW flow when the associated ESW pump is running, 4:gardless whether the served component is in service or not.
Rev. No. 2 Page 22 of 123 i
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT r'~T INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES V
APPENDIX A Pumn Relief Reauests PRR-05R1 (Continued)
During power operation, all loops of ESW are required to be operable per the Technical Specifications. A loop of ESW cannot be taken out of service for testing without entering a limiting Condition for Operation (LCO). With each loop of ESW balanced a requirement to quarterly adjust ESW loop flow to one specific flow value for inservice testing conflicts with system design and operability requirements (i.e. flow balance) as required by Technical Specifications.
ALTERNATE TESTING: As discussed in the basis for relief it is extremely difficult or impossible to retum to a specific flow rate or differential pressure for testing these pumps. Multiple reference points could be established according to the Code, but it would be hp impossible to obtain reference values at every possible point. An alternative to the testing requirements of OM-6 is to base the acceptance criteria on a reference curve. Flow rate and discharge pressure are measured during inservice testing in the as found condition and compared to an established reference curve. The following elements are used in developing and implementing the reference curves.
- 1) A reference pump curve has been established for each pump from empirical data obtained during tests when these pumps were known to be operating acceptably. These pump curves represent pump performance consistent with the original pump test data.
- 2) The reference points were used to develop the pump curves were measured using plant instruments that were calibrated to verify their accuracy prior to performing the tests. In addition to the plant instruments portable UT flow instrumentation was installed during reference testing.
i r
d l Rev. No. 2 Page 23 of12J l
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIXA
, Pump Relief Requests PRR-05R1 (Continued)
- 3) The reference pump curves are based upon the manufacturers pump curves that were validated during preoperational testing and in 1990. Performance Engineering report JPEM-91-001 provides the correlation of data developed during the five tests used to establish the reference pump curve.
- 4) The points utilized were beyond the flat portion of the curve and demonstrated that flow was within the acceptable design limits.
- 5) The acceptanu criteria bases are documented in calculation JAF 91-96 Rev.1. The limits established do not conflict with the Technical Specifications or the Final Safety Analysis Report operability criteria.
- 6) Review of vibration data trend plots indicates that the change in vibration reading do not vary significantly over the narrow range of pump curves being used. Based upon this reference values are uniform and recorded at the upper motor bearing housing location in three directions.
- 7) After any maintenance or repair that may affect the established reference pump curve, a new reference pump curve shall be determined or the existing pump curve revalidated by an inservice test.
O Rev.16,. 2 Page 24 of123
NEW YORK POWER AUTHORITY L%iES A. FITZPATRICK NUCLEAR POWER PLANT g INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES U
APPENDIXA l
Pump Relief Reauests l
PRR-06 SYSTEM: RHR SERVICE WATER / EMERGENCY SERVICE WATER PUMPS: 10P-1A, B, C, D & 46P-2A, B CLASS: 3 FUNCTION: These pumps provide cooling water for safety-related heat loads during a loss-of-coolant design basis accident.
l TEST REQUIREMENT: OM-6 Section 4.6.1.1 specifies the instruments used for prez.. ure l measurement must be accurate to within i 2% of full-scale reading of the instrument.
1 BASIS FOR RELIEF: The RHRSW and ESW pumps are of a vertical submerged open
,e line shaft design. There is no installed instrument for direct
(
measurement of the inlet pressure. Instead, the minimum pumping level is monitored to ensure adequate NPSH is available for pump operation. Since the forebay water level is not expected to change significantly during the testing of these pumps, only one measurement per test is required.
During each test, the difference in elevation between the forebay water level and the pump discharge pressure gauge will be
- determined by measurement. This value will be verified to be l less than or equal to the value corresponding to the minimum water level required for pump operation and will also be used to l calculate pump differential pressure. This calculation method is in accordance with OM-6, Section 4.6.2.2, and NUREG-1482, Section 5.5.3.
Due to limitations of human factors related to measuring the elevation between the forebay water level and the pump discharge pressure gauge, the accuracy of differential pressure l calculation (s) cannot be verified to within i 2% as required by the Code.
1 Rev. No. 2 Page E_ of 123
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX A Pump R:'.tf Reauests PRR-06 (Continued)
ALTERNATE TESTING: In accordance with the guidance provided in NUREG-1482, Section 5.5.3, Differential Pressure for the RHRSW and ESW pumps will be measured as follows:
For each pump, the pump correction value v ;" Se determined by measuring the difference in elevation betwt ; , se forebay water level and the pump discharge pressure gauge, ..nd then calculated in accordance with the procedure. The discharge pressure of the pump will be recorded and then added to the pump correction value to determine the Differential Pressure. This value will be recorded during the performance of each test and then evaluated in accordance with analysis and evaluation criteria specified in OM-6, Section 6, as applicable.
P O
Rev. No. 2 Page 26 of 123 l y
NEW YORK POWER AUTHORITY '
JAMES A. FITZPATRICK NUCLEAR POWER PLANT A. INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES (j
APPENDIX B VALVE TESTING PROGRAM O
Rev. No. 2 Page 27 of123
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PIANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B VALVE TESTING PROGRAM Table of Contents Valve Table Explanation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........................30 Val v e Sy mbol s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3 V alve Types . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3 Val ve A ctuator Types . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3 Te st M e thod . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .,i . . . . . . . 34 I
Test Requirement . . . . . . . . . . . . . . . . . . . . . . . . .... ......... ................. ... . .................34 !
Test Freque ncy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 V al v e Table . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .i . . . . . . . 3 5 i
i Cold Shutdown Justification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 88 i CSJ-01: Reactor Water Recirculation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 88 CSJ-02: Control Rod Drive Hydraulics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 88 CSJ-03: Residual Heat Removal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 89 CSJ-04: Residual Heat Removal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ....................90 CSJ-05: Reactor Core Isolation Cooling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 90 CSJ-06: Reactor Core Isolation Cooling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 91 CSJ-07: Reactor Building Closed Loop Cocling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 91 CSJ-08: Reactor Building Closed Loop Cooling....... . ......................92 CSJ-09: High Pressure Coolant Injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 92 CSJ-10: High Pressure Coolant Injection ................ . .......................93 l CSJ-11: High Pressure Coolant Injection ............ . .............................93 )
CSJ-12: Containment Atmosphere Dilution ........ ...........................94 CSJ-13: M a i n S team . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 94 CSJ-14: M a in S te am . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 5 CSJ-15: Feed w ate r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 95 CSJ-16: Containment Atmosphere Dilution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 96 l 9 ;
1 l
Rev. No. 2 Page 28 of 123 l l
l
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT
/~' INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES i
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A PPENDIX B VALVE TESTING PROGRAM Table of Contents m=
Refueling Outage Justification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 97 ROJ-01: Generic - Excess Flow Check Valves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 97 ROJ-02: Reactor Water Recirculation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 98 ROJ-03: Reactor Water Recirculation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 98 ROJ-04: Automatic Depressurization................ . . ............................99 ROJ-05: Residual Heat Removal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 99 ROJ-06: Residual Heat Remova1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 101 ROJ-07: Standby Liquid Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 102 ROJ-08: Reactor Core Isolation Cooling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 103 ROJ-09R1: Core Sp ray . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 104 ROJ-10: Co re S p ray . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 105 n ROJ-11: Reactor Building Cooling Water . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 105 V ROJ-12:
ROJ-13:
H igh Pressure Coolant Injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 106 High Pressure Coolant Injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 107 ROJ-14: High Pressure Coolant Ir.jection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 107 ROJ-15: High Pressure Coolant Injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 108 ROJ-16: H igh Pressure Coolant Injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 108 ROJ-17: H igh Pressure Coolant Injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 109 ROJ-18: High Pressure Coolant Injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 109 ,
ROJ-19: M ain S team . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 10 ROJ-20: Feed wate r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 10 ROJ-21: Instrument A ir . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 1 ROJ-22: Emergency Service Water . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 112 I
Relie f Requests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 13 VRR-01: Automatic Depressurization/Mam Steam . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 113 VRR-02: Automatic Depressurization/ Main Steam . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 114 VRR-03: Traversing In-Core Probe . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 116 VRR-04: High Pressure Coolant Injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 117 VRR-05: Withd raw n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
VRR-06R1: Service Water / Emergency Service Water. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 118 A
t )
'ud Rev. No. 2 Page 29 of123
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B VALVE TABLE EXPLANATION Summary of Information Provided The Valve Table is sorted by system number, then drawing number, and provides the following informat.on:
- Individual valve identifier
- Drawing coordinates Code Class
- IST Category
- Nominal size
- Valve type
- Actuator type
- Test required
- Relief request (RR)/ cold shutdown (CS) justification / refueling outage (RO) justification Alternate test
- Remarks O
Rev. No. 2 Page 30 of 123
l l
NEW YORK POWER AUTIIORITY l l JAMES A. FITZPATRICK NUCLEAR POWER PLANT l INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES l APPENDIX B l Cold Shutdown Justification CSJ-XX refer to cold shutdown justifications which provide the justification for testing affected components at cold shutdown instead of every three months. The Cold Shutdown Justifications provide the following information System
- Individual valve identifier Valve category Safety function Justification i
(~N I V Refueline Outage Justification l 1
)
ROJ-XX refer to refueling outage justifications which provide the jusufication for testing affected components at refueling outages instead of every three monthi or at cold shutdown. The Refueling Outage Justifications provide the following information:
System Individual valve identifier Valve category Safety function Justification i O
Rev. No. 2 Page 31 of 123
NEW YORK POWER AUTIIORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Valve Relief Requests VRR-XX refer to relief requests for the Valve Testing Program. Each valve request for relief provides the following information:
- System
- Individual valve identifier
- Valve caJegory
- Code Classification
- Safety Function
- Code test requirement for which relief is requested
- Basis for relief
- Proposed alternate testing I
i i
Rev. No. 2 Page 32 of 123
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PIANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES
-v APPENDIX B Valve Symbols Valve Types 3W Three-way valve AN Angle valve BF Butterfly valve BK Ball check BL Ball valve CK Swing check GA Gate valve GL Globe valve LK Lift check NK Non-return valve PG Plug valve RD Rupture disk
- p. RV Relief valve l 4,.
SC Stop check SK Spring check TK Testable check WK Wafer check XP Explosive valve Valve Actuator Tvoes AO Air operator EH Electro-hydraulic HO Hydraulic operator MA Manual operator MO Motor operator PA Pilot actuated SA Self actuated SO Solenoid operator SP Spring operator SQ Squib actuator
(%
Rev. No. _2_ Page 33 of 123 I
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NEW YORK POWER AUTHORITY LuiES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Test Method Test Reauiremen_1 OM-10 Section PIT Valve position indication 4.1 ETO Exercise test to open position 4.2.1.2 ETC Exercise test to closed position 4.2.1.2 PEO Partial exercise to open pcsition 4.2.1.2 PEC Partial exercise to closed position 4.2.1.2 STO Full stroke time measured to open position 4.2.1.4 STC Full stroke time measured to close position 4.2.1.4 FSO Fail safe test to the open position 4.2.1.6 FSC Fail safe test to the closed position 4.2.1.6 LKJ Leak test per 10 CFR 50 Appendix J 4.2.2.2 LKO Leak test for other than containment isolation valve 4.2.2.3 RLF Relief valve test 4.3.1 VBT Vacuum breaker operability test 4.3.1 FFT Check valve forward flow verification test 4.3.2.2 RFC Check valve reverse flow closure test 4.3.2.2 PFT Check valve partial flow test 4.3.2.2 MME Check valve exercise using manual mechanical exerciser 4.3.2.4(b)
DIS Check valve disassembly and inspection 4.3.2.4(c)
XPT Explosively actuated valve test 4.4.1 RDT Rupture disk test 4.4.2 XVD Explosive valve internal inspection Tech Spec j Test Frequency
-1 Quarterly -6 10 CFR 50 Appendix J
-2 Cold Shutdown -7 OM-1 Section 1.3.3
-3 Refueling -8 OM-1 Section 1.3.4
-4 6 months -9 OM-10 Section 4.4.1
-5 2 years -10 OM-10 Section 4.4.2
-11 Tech Spec Requirement l 9
Rev. No. 2 Page 34 of 123
l' .r b b" NEWYORK POYuER AUTHOR!TY JAMES A FITZPATRICK NUCLEAR POWIER PLANT INSERVICE TESTNG PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM. Staney Gas T. ; - SYSTEM D 01-125 DRA)MNG FM4 DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVE ID COCRO CLASS CATEGORY SIZE (IN) TYPE TYPE POSITION REQ'TS CSJ#tOJ REQUEST TEST REMARKS Ot-12iasOV touA C4 2A 8 4 00 SF MO Osc STO-1 AUGMENTED STC-1 PIT 4 01-125MOV-1008 F4 2A B 4 00 BF MO OfC STO-1 AUGMENTED STC-1 PfT4 01-125MOV-11 G4 2A 8 24 00 BF MO O STO-1 AUGMENTED PtT4 01 125MOV-12 F4 2A B 24 00 BF MO O STO-1 AUGMENTED PIT 4 01-125MOV 14A D4 2A 8 24 00 SF MO OIC STO-1 AUGMENTED STC-1 P1T4 01-125MOV-148 E4 2A 8 24 00 BF MO OdC STO-1 AUCL SITED STC 1 PIT 4 01-125MOV 15A D-3 2A 8 24 00 SF MO O STO-1 AUGMENTED PIT 4 01-125MOV-158 F-3 2A B 24 00 SF MO O STO-1 AUGMENTED PrT4 2
4 PAGE 35 OF 123 REV NO: 2
NEW YORK POWER AUTHORITY JAMES A FITZPAT; CK NUCLEAR POWER Pt. ANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEP Automaec Cm w. System. SYSTEM 0: 02 DRAWING. FM-29A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVE O CO-ORD CLASS CATEGORY SIZE ON) TYPE TYPE POSITION REQTS CSJ.ROJ REQUEST TEST REMARKS G2AOV-1 F GF 1 B 1 00 GL AO C Pt !-S PASSIVE 02AOV-18 G7 1 8 1 00 GL AO C pit-5 PASSIVE 02RV-1 07 2 C 3 00 CK SA OC ETO-1 ROJ44 MME-3 ETC-1 MME-3 02RV-2 D-7 2 C 3 00 CK SA OC ETO-1 RO104 MME-3 ETC-1 MME-3 02RV-3 D-7 2 C 3 00 CK SA OC ETO-1 ROJ04 MME-3 ETC-1 MME-3 02RV4 47 2 C 3 00 CK SA OC ETO 1 ROJ44 MME-3 ETC-1 MME-3 02RV-5 D7 2 C 3 00 CK SA OC ETO-1 ROJ-04 MME-3 ETC-1 MME-3 I
l 02RV4 47 2 C 3 00 CK SA OC ETO-1 ROSC4 MME-3 ETC-1 MME-3 j 02RV-7 D7 2 C 3 00 CK SA OC ETG1 ROJoe MME-3 ETC-1 MME-3 l
02RV-8 0'7 2 C 3 00 CK SA OC ETG1 ROJO4 MME-3 l
ETC-1 MME-3 02RV-9 D-7 2 C 3 00 CK SA OC ETO-1 ROJ44 MME-3 ETC-1 MME-3 f
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JAaAES A FIT 2PATINCK NUCLEAR POWEsd PLANT seSEfMCE TESTWeG PROGRAM FOR PUMPS AND VALVES VALwE taste SYSTEas Auseweec t __ Sysamm,SYSTEnB Ek 02 DRAngeG Fis-26A DWG VALVE VALVE ACTUATOR SAFETY TEST REUEF ALTERNATE -
VALVE ID C048tO CLASS CATEOORY St2E (pg TYPE TYPE Po$lTION REQTS CSJ4tOJ REQUEST TEST REGAARES 02Rw-n l 47 2 C 3 00 CK SA QC ET4s RCu.os aftAE-3 ETC-t tense 4 i
02RV-71A G4 1 SC 6 00 RV SA AO OC STot VetR41 ET43 STO.1 VUtR42 ETC-3 RLF-7 02RV-718 G4 1 BC 6 00 stV SA. AO OC ST41 VRR01 ET43 -
STC-s VRR42 ET43 RLF 7 02RV-71C G4 9 BC 6 00 RV SA AO OC STot VRR41 ET43 ST41 VItR42 ETC4 RLF-7 G2RV-71D F4 8 BC 6 00 RV SA. AO OC STot VRR41 EY43 4 STCf VRR42 ETC 3 J
PLF-7 02RV-71E F7 1 BC 6 00 RV SA, AO OC STot VRR41 ET43 ;
STC1 VRR42 ETC 3 .
RLF-7 02RV-71F F7 1 BC 6 00 RV SA, AO OC ST49 VRR41 ET43 l STC-1 VRR42 ETC-3 RLF-7 02RV-71G F-7 1 BC 6 00 RV SA AO OC ST41 Vetit41 . ET43 I STC-t VRR42 ETC-3 RLF-7 1
1 02RV-71H C-7 1 BC 6 00 RV SA. AO OC ' STot VUtR41 ETSS STC-1 VitR42 ETC4 <
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NEWYORK PorKR AUTHORITY JAMES A FITZPATRICK NUCLEAR POWER Pt. ANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM Automa$c Degwosswtzamen System - SYSTEM 0: 02 DRAWING' FM-29A DWG VALVE VALVE ACTUATOR SAFETY TEST REUEF ALTERNATE VALVE O C40RD CLASS CATEGORY SIZE ON) TYPE TYPE POSITION REQ'TS CSJROJ REQUEST TEST REMARKS 024&3 C-7 2 C 10 00 CK SA OIC EIO-1 ROJ44 MME-3 ETC1 MME-3 02VB4 C-7 2 C 10 00 CK SA O/C ETO-1 ROJoe MME-3 ETC-1 MME-3 02VB-5 C-7 2 C 10 00 CK SA OtC ETOL1 ROJoe MME-3 ETC-1 MME-3 02VB 6 C-7 2 C 10 00 CK SA OC ETO-1 ROJ44 MME-3 ETC-1 MME-3 02VB-7 C-7 2 C 10 00 CK SA OC ET41 ROJ04 MME-3 ETC-1 MME-3 02VB 8 C-7 2 C 10 00 CK SA OC ET41 ROJ44 MME-3 ETC-1 MME 3 02VS9 C-7 2 C 10 00 CK SA OC ETO-1 ROJ44 MuE-3 ETC-1 MME-3 02VB-10 C-7 2 C 10.00 CK SA OC ET41 ROJO4 MME-3 ETC-1 MME-3 02VB 11 C-7 2 C 10 00 CK SA OC ETot ROJ44 MME-3 ETC-1 MME-3 l
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SYSTEM Reecsor Wmeer Recendaten. SYSTEM D 02-2 DRAWING FM-26A DWG VALVE VALVE ACT5JATOR SAFETY TEST RELIE' ALTERNATE VALVE O CO-ORO CLASS CATEGORY SIZE ON) TYPE TYPE FUNCTION RECrTS CSJROJ REQUEST TEST REMARKS 02-2AOva9 E-4 1 A 0 75 GA AO C S TC-1 FSC-1 ,.
PtT4 LKJ4 02-2AOV-40 E4 1 A 0 75 CA AO C STC-1 FSC-1 PfT4 ;
LKJ4 6
02-2EFV#S-129A B4 1 AJC 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE 1SOtATES ON EXCESS FLOW LKO4 LKO-3 02-2EFVPS-1288 84 1 AtC 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE tSOLATES ON EXCESS FLOW LKO-5 LKOL3 ,
02-2EFVPT-24A C4 1 A/C 1 00 BK SA C ETC-1 ROJo1 ETC4 VALVE ISOLATES ON EXCESS FLOW '
LKO-5 LKO-3 1
02-2EFVPT-248 C4 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKO4 LKO'3 02-2EFVPT-25A C-3 1 A/C 1 30 BK SA C ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKO5 LKO4 02-2EFVPT-258 C4 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKO-S LKO-3 C4 A/C 0 75 SK SA C RFC-1 ROJ42 RFC4 [
02-2RWR-13A 1 tKJ4 02-2RWR-138 C4 1 A/C 0 75 SK SA C RFC-1 ROJ42 RFC-3 LKJ4 02-2RWR41A D4 1 AC ' O 75 SK SA C RFC-1 ROJ43 RFC4 LKJ4 1
02-2RWR-418 D4 1 AC 0 75 SK SA C RFC1 ROJ03 RFC4 022EFV1CPT111 A E3 1 AlC 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKO4 LKO4 PAGE 30 OF 123 REV NO . 2 i
NEW YORK POINER AUTHORITY JAMES A FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES l
SYSTEM Reactor Water Recro.dahon - S'fSTEM ID 02-2 DRAWING FM-26A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVEID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE POSITION REQTS CSJtROJ REQUEST TEST REMARKS l 022LF V14PI1118 E4 1 AIC 1 00 BK SA C E T C-1 ROJ41 E T C-3 VALVE tSOLATES ON EXCESS FLOW LKO-5 LKO-3 1
022EFV1#T110A F4 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE tSOLATES ON EXCESS FLOW LKO4 LKO4 022EFV14T110C D4 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE ISOt ATES ON EXCESS FLOW LKO-5 LKO4 i 022EFVITT110E F4 1 AIC 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE LSOLATES ON EXCES$ FLOW LKO4 LKO-3 l
022EFv14T110G De 1 A!C 1 00 BK SA C ETC 1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW f LKO-5 LKO4 022EFV24PT111 A E-3 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE RSOLATES ON EXCESS FLOW LKO4 LKO-3 022EFV2-DPT1118 E4 1 AIC 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE ISOUTES ON EXCESS FLOW LKO4 LKO4 022EFV2TT110A F-3 1 A/C 1 00 BK SA C ETC 1 ROJ41 ETC4 VALVE tSOLATES ON EXCESS FLOW LKO4 LKG3 022EFV2fT110C D4 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE RSOLATES ON EXCESS ELOW LKO4 LKO4 022EFV2TT110E F4 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE tSOLATES ON EXCESS FLOW i LKO-5 f ra :
l d22EFV2#T110G D4 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKO4 LKO-3 02MOV-53A C4 1 9 28 00 GA MO C STC-1 CSJ41 STC 2 PIT 4 02MOV 538 C4 1 B 2800 GA MO C STC-1 CSJ41 STC4 PfT4
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JAMES A FIT 2 PATRICK 98UCLEAR POWER PLANT INSERVICE TEST #eO PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM kh Soder Vessed enerumeras SYSTEM O O24 DRAWudG. FE47A 5 DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVEID COORD CLASS CATEGORY S4ZE (W) TYPE TYPE POSITION REQTS CSJtROJ REOUEST TEST REMARKS 02-3EFV-11 F-7 1 AfC 1 00 SK SA C ETC-1 ROJ41 E T C-3 VAa.VE tSOLATES ON E.XCES$ FLOW LKO4 LKO-3 ,
024EFV-13A E-T 1 AJC 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE ISOLATES ON EXCESS FLOW f LKO-5 LKO4 ,
02-3EFV-138 E4 1 AIC 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE tSOLATE3 ON EXCESS FLOW {
LKO-5 LKO-3 02-3EFV-15A E-7 1 AIC 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE tSOLATES ON EXCES$ FLOW LKO4 LKO-3 02-3EFV-158 E-4 1 A/C 1 00 BK SA C ETC 1 ROJ41 ETC-3 VALVE ISOLATES ON EXCESS FLOW LKO4 LKO4 ,
02 3fFV-15N B7 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKO4 LKO4 i
02-3EFV-17A D-7 1 AfC 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKO S LKO-3 02-3EFV-178 D4 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE ISCLATES ON EXCESS FLOW LKO4 LKO4 02-3EFV-19A D-7 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE ESOLATES ON EXCESS FLOW ,
LKO4 LKO4 t
02-3EFV-198 04 1 AJC 1 00 SK SA C ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW ;
LKO4 LKO-3 .
02-3EFV-21A 0+5 1 AJC 1 00 9K SA C ETC-1 ROJ41 ETC-3 VALVE ISOt.ATES 04 EXCESS FLOW
, LKO4 LKO4 I
02-3EFV-218 C-7 1 A/C 1 00 8K SA C ETC-1 ROJ41 ETC-3 V+"* T ATES ON EXCESS FLOW i LKO4 LKO4 02-3EFV-21C C-4 1 A/C 1 00 SK SA C ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKO4 LKO4 '!
02-3EFV-210 H4 1 A/C 1 00 BK SA C ETC1 ROJ41 ETC4 VALVE RSOLATES ON EXCE3S FLOW t LKO4 LKO-3 02-3EFV,23 F-7 1 AJC 1 00 SK SA C- ETC1 ftOJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW j LKO4 LKO4 i
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J#MES A FIT 2 PATRICK NUCLEAR 80WER PLANT INSERVICE TESTNG PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM ><*==r Boder VesseiInstruments - SYSTEM O 02,3 DRAWING FM47A DWG VALVE VALVE ACTUATOR SAFETY TF.ST REllEF ALTERNATE VALVE O CO&D CLASS CATEGORY StZE WJ TW TYPE POSITION REOTS CSJ/ROJ REQUEST TEST REMARKS O2-3LF V-23A H4 1 AK; 1 00 BK SA C E T C-1 ROJ41 ETC4 VALVE ISOLA T ES ON EXCES$ F LOW LKO4 LKO4 02-3EFV-238 D-7 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE tSOLATES ON EXCESS FLOW LKO4 LKO4 02-3EFV-23C D4 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE ISOLATES ON EXCESS FLOW LKO4 LKO-3 02-3EFV-23D C-7 1 A/C 1 00 BK SA C ETC-1 ROJ 01 ETC-3 VALVE tSOLATES ON E.XCESS TOW LKO4 LKO-3 02-3EFV 25 C-7 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC 3 VALVE fSOLATES ON EXCESS FLOW LKO-5 LKO-3 02-3EFV-31 A H5 1 AtC 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE tSOLATES ON EXCESS FLOW LKO4 LKO4 02-3EFV-3tB H4 1 AtC 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVERSOLAlm A EXCESS FLOW LKO4 LKG3 02-3EFV-31C H4 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE tSOLATES ON EXCESS FLOW LKO4 LKO-3 02-3EFV41D M4 1 A/C 1 OC BK SA C ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKO4 LKO-3 02-3EFV-31E D-7 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKO4 LKO-3 024EFv41F M4 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE ISOLATES ON EXCESS FLOW LKO4 LKO4 02-3EFV41G O-5 1 AtC 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO4 024EFV-31H G4 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE PSOLATES ON EXCESS FLOW LKO4 LKO4 02-3EFV41J M4 1 A/C 1.00 BK SA C ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKO4 LKO-3 02 3EFV4tK H4 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE 3SOLATES ON EXCESS FLOW LKO4 LKO4 REV NO 42 OF 123
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- NEWYORK POWER AUTNORITY JAMES A FfT2 PATRICK NUCLEMt POIAER PLANT WSERVICE TESTueG PROGRAM FOR PUMPS AND VALVES f VALVE TABLE SYSTEM pm Bader vesset imineents - 3YSTEM C 02-3 DRANNG FM-47A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVEID COORD CLASS CATEGORY SIZE (W) TYPE TYPE POSrTION RFOTS CSJROJ REQUEST TEST REMARKS 024EFv,31L es4 1 AC 1 00 BK SA C ETC-1 P.OJ41 ETC-3 VALVE tSOLATES ON EACESS FLOW LKO-5 LKO-3 024EFV-31M D4 1 AC 1 00 8K SA C ETC-1 ROJ41 ETC-3 VALVE ISOLATES ON EXCESS FLOW ,
LKO-5 - LKO4 I 024EFV-31N H4 1 AC. 1 00 SK SA C ETC1 ROJ41 ETC4 VALVE tSOLATES ON EXCESS FLOW I
LKO-5 tx04 024Efv41P H4 1 AC 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE RSOLATES ON EXCESS FLOW LKO4 LKO-3 024EFV-31R G4 1 AC 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE tSOLATES ON EXCESS FLOW LKO-5 LKO-3 7 02-3EFV-31 S G4 1 A/C 1 00 BK SA C FTC-1 ROJ41 ETC-3 VALVE tSOtATES ON EXCESS FLOW LkO-5 LKO-3 024EFV43 B-4 1 AIC 1 00 8K SA O ETC-1 ROJ41 ETC4 VALVE tSOLATES ON EXCESS FLOW 1.KO-5 LKO-3 I
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NEWYORKPOWER AUTHCRITY JAMES A FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTNG PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM CortclRodDme - SYSTEM O, 03 DRAYANG FM-278 OWG VALVE VALVE ACTUATOR SAEETY TEST RELIEF ALTERNATE VALVE O CO{RD CLASS CATEGORY SIZE (IN) TYPE TYPE POSITION RESTS CSJIROJ REQUEST TEST REMA9KS OJAOV-126 C-4 2 8 1 (X) GL AO O $10-1 E10-3 3 CRAM im4E TEST FSO-1 GL8944 POSITON 7 03AOV-127 D4 2 8 1 00 GL AO O STO 1 ETO-3 SCRAM TIME TEST FSO-1 GLBSO4 POSITION 7 C3AOV-32 H4 2 8 1 00 GL AO C STC-1 FSC-1 PfT4 03AOV-33 F4 2 B 2 00 GL AO C STC-1 FSC-1 PIT 4 03AOV-34 H4 2 8 1 00 GL AO C STC-1 FSC-1 PtT-5 03AOV 35 F-4 2 8 2 00 GL AO C STC-1 FSC-1 PfT4 03AOV-36 M4 2 8 1 00 GL AO C STC-1 FSC-1 PfT4 03AOV-37 F4 2 8 2 00 GL AO C STC-1 FSC-1 PfT4 03AOV-38 H4 2 8 1 00 GL AO C STC-1 FSC-1 PfT-5 03AOV49 F4 2 8 2 00 GL AO C STC-1 FSC-1 PfT4 03HCIA114 04 2 C 0 75 BK SA O FFT-1 FFT4 SC8 TAM TIME TEST GL8944 POSITION 7 03HCU 115 C-4 2 C 0 50 BK SA C RFC-1 CSJ42 RFC-2 03HCIA138 C4 2 C C 50 BK SA C RFC-1 REVERSE FLOWTESTED VLA RCD MOTION REV NO 44 OF 123
NEWYORK POMER AUTHORITY [
- JAAES A FfT2 PATRICK NUCtEAR POWER Pt. ANT
- INSEfMCE TESTING PROGRAM FOR PUMPS AND VALVES 7 VALVE TA8LE i SYSTEM CareulRodtw - SYSTEM O. 03 DRAMmeG FM-278 2 DWG VALVE VALVE ACTUATOR SAFETY TEST REUEF ALTERNATE VALVE ID COCRD CLASS CATEGORY SIZE 000 TYPE TYPE POSmON REO'TS CSSROJ REQUEST TEST REMARKS 0350v-120 C4 2 8 O SO GA AO C SIC 1 ETC-3 SCRAM TIME TEST FSC-1 GL9944 POSITION 7 03SOV-121 C-4 2 8 0 50 CA AO C STC-1 ETC-3 SCRAM TasE TEST FSC-1 Gle944 POSITION 7 y SCRAM TIME TEST !
03SOV-122 C4 2 8 0 50 GA AO C STC-1 ETC-3 FSC-1 GulS44 POSmON 7 0350V-123 C4 2 8 1.50 GA AO C STC-1 ETC-3 SCRAM TIME TEST FSC-1 GL8944 POSITION F t
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NEWYORKPOWER AUTHOR 4TY JAMES A FITZPATRICK NUCLEAR POWER PLANT INSERVICE TEST!NG PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM Travereng in<, ore Pree SYSTEM ID 07 ORAWING FM-119A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVE ID CO-ORD CLASS CATEGORY StZE ON) TYPE TYPE POSfTION REQTS CSJtROJ REQUEST TEST REMARKS 07EV-104A F4 2A D 0375 XP SQ C KPT4 AUGMENT ED 07EV-1048 F-4 2A D 0375 XP SQ C XPT-9 AUCMENTED 07EV-104C F-4 2A D 0375 XP SQ C XPT4 AUGMENTED 07SOV-104A F-5 2A A 0 375 BL SO C STC-1 VRR43 AUGMENTED FSC-1 PtT-5 LKJ4 07SOV-1048 F-4 2A A 0 375 BL SO C STC-1 VRR43 AUGMENTED FSC-1 PtT4 LKJ4 07SOV-104C F-4 2A A 0375 BL SO C STC-1 VRR 43 AUGMENTED FSC-1 PfT 5 LKJ4 REV NO 4 OF 123
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DWG VALVE VALVE ACTUATOR SAFETY ' TEST REUEF ALTERNATE ,
VALVE ID CO4RD CLASS CATEGORY S:ZE (M) TYPE TYPE POSITION REOTS CSJstOJ REQUEST TEST REMARKS' suAOV48A F4 1 AtC 24.00 TK SA. AO Orc F F T-1 CSJ43 F F T-2 RFC-1 CSJ43 - RFC-2 LKO4 [
r 10AOV488 F4 1 A/C 24 00 TK SA. AO OC FFT-1 CSJ43 FFT4 ;
RFC-1 CSJ43 RFC-2 LKO-5 10MOV-13A 84 2 8 20 00 GA MO OfC STO-1 STC-1 PfT4 e 10MOV-138 C-4 2 8 20 00 GA MO OC STO-1 I STC-1 PIT 4
'.t 10MOV-13C C4 2 8 20 00 GA MO OfC STO-1 STC-1 PtT-5 ,
t 10MOV 13D C4 2 9 20 00 GA MO OC STO-t STC-1 PIT 4 i
10MOV 15A C4 2 8 20 00 GA MO C STC-1 PIT 4 10MOV-158 C-4 2 8 20 00 GA MO C STC1 PtT4 MO C STC-1 ,
10MOV-15C C4 2 8 20 00 GA PtT4 1 10MOV-153 C4 2 8 20 00 GA MO C STC-1 PIT-5 r
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NEWYORKPOWER AJTHORITY JAMES A FITZPAT5UCK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM Resummi Heat Removal . SYSTEM ID 10 DRAWING- FM-20A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVE ID CO N CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REQTS CSJtROJ REQUEST TEST REMARKS 10W)v- st8 W 2 8 4 00 GA MO OtC SIO-1 STC-1 PIT 4 t0MOV-17 D4 1 A 20 00 CA MO C STC-1 CSJ44 STC-2 P(T4 LKC4 SATISFIED BY LKJ3 LKO4 LKJ3 PER JAF4ALCMtSC40554 LKJ4 10MOV-18 E-5 1 A 20 00 GA MO C STC-1 CSJ44 STC-2 JAF4E-96417 PtT4 LKO4 10MOV-21A E4 2 8 4 00 GA MO C PIT-5 PAS $lVE 10MOV-218 E4 2 8 4 00 GA MO C PIT 4 PASSIVE 10MOV-25A F4 1 A 24 00 GA MO OfC STO-1 LKO4 SATtSFIED BY LKJ4 STC-1 PER JAF4ALCMISC40554 PIT 4 LKO-5 LKJ4 LKJ4 10MOV-258 F-3 1 A 24 00 GA MO Q1C STO-1 LKO4 SATISFIED BY LKJ-3 STC-1 PER JAF4ALC#tSC40554 PfT4 LKO4 LKJ-3 LKJ4 10MOV-26A G-7 2 A 10 00 CA MO OfC STO-1 JAF-SE45417 STC-1 PIT 4 10MOV-268 G4 2 A 10 00 GA MO OC STO-1 JAF-SE46417 STC-1 PfT4 1CMOV-27A F-8 1 A 18 00 AN MO OC STO-1 [[::JAF-SE-96417|JAF-SE-96417]] STC-1 PIT 4 REV NO 48 OF 123
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- SYSTEM O- to DRAWING: FM-204 i
DWG VALVE VALVE ACTLe. TOR SAFETY TEST REUEF ALTERNATE VALVE O CO-ORD CLASS CATEGORY SIZE CN) TYPE TYPE FUNCTION REOTS CSJ5tOJ REQUEST TEST REMARKS 10 MOW 4/8 F-3 1 A 18 00 AN MO OtC SiO-1 JAF4E46417 STC-1 PtT4 i
10MOV-31A G4 2 A 10 00 GL MO OfC STO-t t STC-1 Pf74 LKJ4 ,
10MOV-318 G4 2 A 10 00 GL MO OfC JTO-1 STC-1 PIT 4 LKJ4 4
10MOV-34A E-T 2 8 14 00 GL MO Otc STO-1 STC-1 PIT 4 i
10MOV-348 E4 2 8 14 00 GL MO orc STO-1 -
STC-1 i PfT4 10MOV-38A E-7 2 A 4 00 GL MO OfC STO-1 ,
STC-1 PIT 4 LKJ4 10MOV-388 E4 2 A 4 00 GL MO OC STO-1 STC-1 PIT 4 tx14 i
10MOV4!hA E4 2 A 16 00 GL MO OfC STO-1 [[::JAF-SE-06|JAF-SE-06]] 017 STC-1 PtT4 t
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NEW YORK POWER AUTHORITY l
JAMES A FITZPATRICK NUCLEAR POWER PLANT
! WSERVICE TESTING PROrAAM FOR PUMPS ANO VALVES l VALVE TABLE
[ SYSTEM R=h Heat Remowat - SYSTEM ID to DRAWING FM20A l
l OWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE l VALVEID CO M CLASS CATEGORY SIZE ON) TYPE TYPE FUNCTION REOTS CSJiROJ REQUEST TEST REMARKS 1L%IOvetA 04 2 8 20 00 GL MO OfC SIO-1 l
STC-1 PfT-5 10MOV468 D-3 2 8 20 00 GL MO OC STO-1 STC-1 PfT-5 10RHR-262 H-3 2 C 4 00 CK SA C RFC-1 10RHR-277 G4 2 C 4 00 CK SA C RFC-1 10RHR42A C4 2 C 16 00 CK SA OfC FFT-1 RFC-1 10RHR428 C4 2 C 16 00 CK SA OfC FFT-1 RFC-1 10RHR42C C4 2 C 16 00 CK SA OdC FFT-1 RFC-1 i
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10RHR420 C3 2 C 16 00 CK SA OfC FFT-1 RFC-1 10RHR44A C4 2 C 3 00 CK SA OC FFT-1 ROJ45 PFT-1 AT LEAST ONE VALVE PER OUTAGE RFC-1 DIS-3 WITH ALL VALVES N GROUP INSPECTED AT LEAST ONCFJ6 YRS.
10RHR448 C4 2 C 3 00 CK SA OC FFT-1 ROJ45 PFT-1 AT LEAST ONE VALVE PER OUTAGE RFC-1 DtS4 WITH ALL VALVES IN GROUP INSPECTED AT LEAST ONCES YRS l
10RHR44C D4 2 C 3 00 CK SA OC FFT-1 ROJ45 PFT-1 AT LEAST ONE VALVE PER OsJTAGE RFC-1 DtS4 WITH ALL VALVES N GROUP INSPECTED AT LEAST ONCES YRS.
10RHR440 D-3 2 C 3 00 CK SA OC FFT-1 ROJ45 PFT-1 AT LEAST OidE VALVE PER OUTAGE RFC-1 DIS-3 WITH ALL VALVES N GROUP NSPECTED AT LEAST ONCE2 YRS REV N0 f4 OF 123
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DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVE 10 CO&D CLASS CATEGORY StZE (RN) TYPE TYPE FUNCTION REOTS CSJfROJ REQUEST TEST REMARKS 13EF V41A F-7 1 AtC 1 00 BK SA C E T C-1 ROJ41 ETC4 VALVE 850LATES ON EXCESS FLOW LKO4 LKO-3 i
13EFV418 F 1 A/C 1 00 BK SA C ETC-1 ROJ41 - ETC4 VALVE ISOLATES ON EXCESS FLOW 5 LKO4 LKO4 13EFV42A G-7 1 AlC 1 00 8K SA C ETC1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW i LKO-5 LKO4 13EFV4?B F-7 1 AC 1 00 8K SA C cETC-1 ROJ41 ETC 3 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKLe3 13MOV-15 F-7 1 A 3 00 GA MO C STC-1 ,
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- N NEW YORK POWER AUTHORITY JAMES A FITZPATRICK NUCLEAR POWER PLANT M3ERVICE TEST:NG PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM Care Spray - SYSTEM O *4 DRAtMNG FM43A DWG VALVE VALVE ACTUATOR SAFETY TEST RELEF ' ALTERNATE VALVE O COOtD ' CLASS CATEGORY SIZE (IN) - TYPE TYPE FUNCTION REQ'TS CSAROJ REQUEST TEST REMARKS 14MOV-12A F4 1 A 10 00 GA MO OtC STG1 STC 1 PIT 4 LKO-5 LKJ LKO-S SATtSFED BY LKJ-3 LKJ4 PER JAF<,ALC4stSC40554 14MOV-12B F-4 t A 10 00 GA MO OC STO-1
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NEWYORK POWER AUTHORITY JAMES A FITZPATRICK NUCLEAR POWER PLANT NSERVICE TESTNG PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM Leak Rate Anatper - SYSTEM C 16-1 DRAWING FM49A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVE D COMD CLASS CATEGORY StZE (N) TYPE TYPE FUNCTON REQ'TS CSJtROJ REQUEST TEST REMARKS 16-1 AOV-101 A D-7 2A A 0 375 GA AO C STC-1 FASI ACilNGVALVE FSC-1 AUCMENTED PIT 4 LKJ4 16-1 AOV 1018 E-7 2A A 0 37S GA AO C STC-1 FAST ACTNG VALVE FSC-1 AUGMENTED PfT4 LKJ4 16-1 AOV-102A O-7 2A A 0 375 GA AO C STC-1 FAST ACTNGVALVE FSC-1 AUGMENTED PIT 4 LKJ4 16-1 AOV-1028 C-7 2A A 0375 GA AO C STC-1 FAST ACTNG VALVE FSC-1 AUGMENTED PfT-S LKJ4 i
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NEWYORKPonER AlffMORfTI JAMES A FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PFOORAM FOR PUMPS AMO VALVES VALVE TABLE SYSTEM Fust Pact Coolme - SYSTEM O. 19 ORAWING FM-19A DWG VALVE VALVE ACTUATOR SAFETY TEST REUEF ALTERNATE VALVE O COOT 0 CLASS CATEGORY SIZE (WO TYPE TYPE FUNCTION REQ'TS CSJROJ REQUEST TEST REMARMS 19VB-1 A G-5 3A C 1 50 RV SA C RLF 4 AUGMENTED 19VB-18 G-5 3A C 1 50 RV SA C RLF4 AUGMENTED 1
4 1
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MN JAMES A FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM Radweste - SYSTEM D 20 DRAWING FM-17A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVEID CO M CLASS CATEGORY St?E (tN) TYPE TYPE FUNCTION REQ'TS CSJtROJ REQUEST TEST REMARKS 20AOV43 F4 2A A 3 00 BL AO C SIC-1 F AST ACleeG VALVE FSC-1 AUCMENTED PfT4 LKJ4 20AOV-95 C4 2A A 3 00 BL AO C STC-1 FAST ACTINGVALVE FSC-1 AUGMENTED PIT 4 LKJ4 20MOV42 F-7 2A A 3 00 GA MO C STC-1 AUGMENTED PIT-5 LKJ4 l
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NEWYORKPOWER AUTHORITY JAMES A FITZPATRICK NUCLEAR POIAER PLANT NSERVICE TESTING PROORAM FOR PUMPS AND VALVES 5
VALW TABLE 1 SYSTEM hgh Pressure Coolert kycten . SYSTEM D 23 DRAWING FM45A -
DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVE ID COORD
- CLASS CATEGORY SIZE UN) TYPE TYPE FUNCTION REQTS CSJROJ REQUEST TEST REMARKS 23AOV-a2 G-2 2 B 1 00 GA AO C S T C-1 FASI ACilNGVALVE FSC-1 ,
PIT 4 i 23EFV41A G4 1 AJC 1 00 BK SA C ETC-1 ROJ48 ETC4 - VALVEISOLATES ON EXCESC FLOW LKO4 LKO4 23EFV41B G-7 1 AIC 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKO-S LKO-3 J
23EFV42A G-7 1 A/C 1 00 BK SA ' C ETC-1 ROJ41 FTC4- VALVE tSOLATES ON EXCESS FLOW l LKO4 LKO4 23EFV42B G-7 1 . A/C 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FL OW LKO-5 LKO 23HOV-1 F-3 2 B 10 00 GA HO OC STO-1 FAST ACTINGVALVE v STC-1 ,
PfT4 23HPk12 C4 2 AtC 16 00 LK SA OIC FFT-1 ROJ-12 RFC-1 RFC4 LKJ4 =
23HPI-13 C-7 2 C 2 00 SC SA. MA OC FFT-1 RO113 DES 4 RFC-1 CSJ49 RFC,2 23HPI-130 C4 2 C 2 00 SK SA OC FFT-1 ROJ-17 DIS 4 RFC-1 PFT-1 23HPl-131 C4 2 C 2 00 SK SA C RFC-1 ROJ-15 Disa ,
23HPk18 F-7 1 C 14 00 CK SA O FFT-1 CSJ-10 MME-2 F
23HP1-32 G4 2 C tiLOO CK SA C RFC-1 ROJ-14 DIS 4 ?
23HPl402 E-7 2A C 2 00 CK SA OC FFT-1 CSJ-11 VRR44 FFT-2 AUGMENTED COMPONENT RFC-1 RFC 2 VERIFIED CLOSEO As PAIR WITH 23HPk403 !
23HP1403 E-7 2A C 2 00 CK SA OC FFT-1 CSS 11 VRR44 FFT,2 AUGMENTED COMPONENT RFC-1 RFC-2 VERIFED CLOSED AS PAIR WITH )
I 23 Ml402 1
i PAGE 65 OF 123 REV NO : 2 i
NEW YORK POWER AUTHORITY JAMES A FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS ANO VALVES VALVE TABLE SYSTEM Mgh Presaso Coolart Irgecten - SYSTEM D 23 DRAWING FM 25A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVE ID COGD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REQTS C!JtROJ REQUEST TEST REMARKS 23MPI-56 C4 2 C 2 00 SK SA O F F 1-1 ROJ-13 LMS-3 23HPl41 8-7 2 C 16 00 CK SA O FFT-1 RO115 OtS4 PFT4 23HPl42 F-4 2 C 4 00 CK SA O FFT-1 RO116 DIS-3 23HPl45 C4 2 AC 20 00 LK SA OJC FFT-1 ROS12 RFC-1 RFC-3 LKJ4 23MOV-14 F-3 2 8 10 00 GA MO O STO-1 PIT 4 23MOV-15 F4 1 A 10 00 GA MO OdC STO-1 STC-1 PIT 4 LKJ4 23MOV-16 F-7 1 A 10 00 GA MO OC STO-1 STC-1 PfT4 LKJ4 23MOV-17 G4 2 8 16 00 GA MO C STC-1 PIT 4 23MOV-19 F4 1 A 14 00 GA MO OC STO-1 STC-1 PIT 4 LKJ4 23MOV-20 F4 2 8 14 00 GA MO O STO-1 PfT4 23MOV 21 G4 2 8 8 00 GL MO C STC-1 PfT4 23MOV 25 F4 2 8 4 00 GL MO OC STO-1 STC 1 Pff4 23MOV47 F4 2 8 16 00 GA MO O STO-1 PfT4 66 OF 123 REV NO
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NEWYORKPOWER AUTHORITY JAMES A FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM Contsevnert Anneephenc DMxm - SYSTEM O 27 DRAWING FM-18A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVE O CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REQTS CSJtROJ REQUEST TEST REMARKS 2 /AOV-126A G4 2A B 1 00 GL AO O MO-1 AUGMENitD FSO-1 FAST ACTING VALVE PIT 4 PIT-3 27AOV-1268 F-5 2A B 1 00 GL AO O STO-1 AUGMENTED FSO-1 FAST ACTINGVALVE PIT-5 PIT-3 27AOV-128A G4 2A B 1 50 GL AO OtC STO-1 AUGMENTED STC-1 FAST ACTINGVALVE FSO-1 PfT4 PIT-3 27AOV 1288 E-4 2A B 1 50 GL AO OIC STO-1 AUGMENTED STC-1 FAST ACTINGVALVE FSOwt PtT-5 PfT 3 27AOV-129A F-4 2A B 1 00 GL AO OtC STO-1 AUGMENTED i
STC-1 FAST ACTINGVALVE l FSO-1 PtT-5 PtT-3 I 27AOV-1290 F-4 2A B 1 00 GL AO OIC STO-1 AUGMENTED
! STC-1 FAST ACTING VALVE FSO-1
! PfT4 PtT-3 27 CAD-19A G4 2A C 2 00 CK SA O FFT-1 AUGMENTED l
27 CAD-198 C4 2A C 2 00 CK SA O FFT-1 AUCMENTED j
27RD-1A F-7 2A D 2 00 RD SA C RDT-10 AUGMENTED l
27RD-18 C-7 2A D 2 00 RD SA C RDT-10 AUGMENTED 27RD-2A F4 2A D 2 00 RD SA C RDT-10 AUGMENTED C4 2A D 2 00 RD SA C RDT-10 AUGMENTED 27RD-28 27$v-114A G4 2A C 1 00 RV SA C RLF4 AUGMENTED 275V-1148 Do 2A C 1 00 RV SA C RLF4 AUGMENTED l
l l 68 OF 123 REV
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NEW YORK PONER AUTHORITY JAMES A FITZPATRICK NUCLEAR POWER PtANI INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM Contamwerd Atmosphwc Duas . SYSTEM D- 27 DRA1MIC FM-18A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVE D CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REQTS CSJ4tOJ REQUEST TEST REMARKS 27sv-115A G-4 2A C 0 50 RV SA C RLF4 AUGMENTED 27SV-1156 E-4 2A C 0 50 RV SA C RLF4 AUGMENTED 27SV-118A G4 2A C 0 50 RV SA C RLF4 AUGMENTED 27SV-1188 C4 2A C O SO RV SA C RLF4 AUGMENTED 27SV-119A F7 2A C 0 50 RV SA C RLF4 AUGMENTED 27SV-1198 C7 2A C 0 50 RV SA C RLF4 AUGMENTED s
27SV-201A F-3 2A C 1 00 RV SA C _ RLF4 AUGMENTED ,
-t 27sv-2018 F-3 2A C 1 00 RV SA C RLF4 AUGMENTED 27SV-202 H-3 2A C 10) RV SA C RLF4 AUGMENTED ,
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NEW YORK PCMNER AUTHORITY JAMES A FITZPATRICK NUCLEAR POWER PLANT INSERW,E TESTING PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM Cemtawwnent Atmosphenc Deuhon SYSTEMID- 2T DRAWING. FM 188 DN3 VALVE VALVE ACTUATOR SAFETY TEST RELtEF ALTERNATE VALVEID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTI(* - RE&TS CSJtROJ REQUEST TEST REMARKS C4 2A A 20 00 BF AO OC STO 1 AUGMEN T ED 27AOV-101 A STC-1 FSC-1 PtT-5 LKJ4 27AOV-101B C4 2A A 20 00 BF AO OdC STO 1 AUGMENTED STC-1 FSC 1 PIT-5 LKJ4 C-2 2A A 24 00 BF AO C STC-1 CSS 12 STC-2 AUCEENTED 27AOV-111 FSC-1 FSC-2 l
PfT-5 LKJ4 2A A 24 0, BF AO C STC-1 CS112 STC-2 AUGMENTED 27AOV-112 C-3 FSC-1 FSC-2 PIT-5 LKJ4 27AOV-113 04 2A A 24 PO BF AO C STC-1 CS512 STC-2 FSC-1 FSC-2 AUGMENTED PIT-5 I LKJ4 i
J l
D4 2A A 24 00 BF AO C STC-1 CSS 12 STC-2 AUGMENTED 27AOV-114 FSC-1 FSCc2 PtT-5 LK16 I
24 A 20 00 BF AO C STC-1 CS512 STC-2 AUGMENTED
( 27AOV-115 C-2 FSC-2 F SG-1 1
1 PtT-5 LKJ4 l
C-3 2A A 20 00 BF AO C STC 1 CS112 STC-2 AUGMENTED 27AOV-115 FSC-1 FSC-2 PIT-5 LK16 70 OF 123 REV N
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JAMES A FIT 2PATRtCK NUCLEAR POWIER PLANT '
INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES VALVE TABLE Contanmere Abncephere Deuhon - SYSTEM D 27 DRAWING FM-188 SYSTEM DWG VALVE VALVE ACTUATOR SAFETY TEST .
RELIEF ALTERNATE CATEGORY StZE (IN) T*E TYPE FUNCTION REQTS CS.WROJ REQUEST TEST REMARKS VALVE ID COORD CLASS B4 2A A 20 00 BF AO C SIC-1 MJGMENTED 2 TAOv-11F FSC-1 PIT 4 LKJ4 84 2A A 20 00 BF AO C STC-1 AUGMENTED 27AOV-118
' FSC-1 PIT 4 LKJ4 C4 2A A 1 50 GL AO OC STO-1 AUGMENTED
- 27AOV-131A STC-1 FSC-1 PfT4 LKJ4 C-3 2A A 1 50 GL AO OC STO-1 AUGMENTED 27AOV-1318 STC-1 FSC-1 PIT-5 LKJ4 A 1 50 GL AO OC STO-1 AUGMENTED 27AOV-132A C-4 2A STC-1 i
FSC-1 PfT4
' LKJ4 2A A 1 50 GL AO OC STO-1 AUGMENTED
, 27AOV-1329 C-3 STC-1 FSC-1 PtT4 LKJ4 JVC 1 50 SK SA OC FFT-1 AUGMENTED 27 CAD 47 C-4 2A RFC-1 LKJ4 AtC 1 50 SK SA OC FFT-1 AUGMENTED 27 CAD 48 C4 2A RFC-1 LKJ4 2A A/C 1 50 SK SA OC FFT-1 AUGMENTED 27 CAD 49 C-3 RFC-1 LKJ4 1
i PAGE 71 OF 123 REV NO 2
NEW YORK POWER AUTHOR 1TY JAMES A FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM Contaavnert Almosphenc DOhan - SYSTEM O 27 DRAWING FM-18B DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVE O CO4RD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTON REQ TS CSJIROJ REQUEST TEST REMARKS 2/CAO-10 C-3 2A Arc 1 50 SK SA OtC F F T-1 AUGMEN T ED RFC-1 LKJ4 2?MOV-113 C4 2A A 3 00 BF MO Orc STO-1 AUGMENTED STC-1 PIT-5 LKJ4 27MOV 117 B4 2A A 3 00 BF MO OC STO-1 AUGMENTED STC-1 P!T-5 LKJ4 27MOV-120 H4 2A B 12 00 BF MO O STO-1 C$ ate STC-1 AUGMENTED PIT-5 27MOV-121 H4 2A B 6 00 BF MO O STO-1 AUGMENTED PtT-5 27MOV-122 C4 2A A 3 00 GL MO O/C STO-1 AUGMENTED STC-1 PIT-5 LKJ4 27MOV-123 B6 2A A 3 00 GL MO OC STO-1 AUGMENTED STC-1 PIT-5 LKJ4 27SOV-125A F-5 2A A 1 00 GL SO C STC-1 AUGMENTED FSC-1 FAST ACTINGVALVE PtT-5 LKJ4 27SOV-1258 F-4 2A A 1 00 GL SO C STC-1 AUGMENTED FSC-1 FAST ACTING VALVE PIT-5 LKJ4 2750V-125C F-5 2A A 1 00 GL SO C STC-1 AUGMENTED FSC-1 FAST ACTINGVALVE PIT-S LKJ4 REV NO 72 OF 123
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NEW YORK POWER AUTHORITY JAMES A FITZPATRICK NUCLEAR POWIER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM Contanment AnnospnencDeuten - SYSTEM 10 27 DRAWING FM-188 OWG VALVE VALVE ACTUATOP SAFETY TEST ' REttEF ALTERNATE-VALVEID CO M CLASS CATEGORY SIZE UN) TYPE TYPE FUNCTION RE(YTS CSJfROJ REOVEST TEST - REMARKS 2/SOv-1250 F -4 2A A 1 00 GL SO C SIC-1 AUGMENTED FSC-1 FAST ACTINGVALVE PIT 4 LKJ4 27SOV-135A E4 2A A 1 00 GL SO C STC-1 AUGMENTED FSC-1. FAST ACTINGVALVE PtT4
-LKJ4 27SOV-1358 F4 2A A 1 00 GL SO C STC-1 AUGMENTED FPC-1 FAST ACTING VALVE Pf74 LKJ4 27SOV-135C E4 2A A 1 00 GL SO C STC-1 AUGMENTED FSC-1 FAST ACTING VALVE PtT4 LKJ4 27SOV-135D F4 2A A 1 00 GL SO C STC1 - AUGMENTED FSC-1 FAST ACTINGVALVE PIT 4
'LKJ4 27VB-1 C4 2A A;C 30 00 CK SA OIC ETOL1- MME-1 AUGMENTED ETC-1 MME-1 PtT4 LKO4 LKO-3 27VB-2 C4 2A AfC 3000 CK SA OfC ETO-1 MME-1 AUGMENTED ETC-1 MME-1 PIT 6 LKO4 LKO 3 27VB-3 C4 2A AAC 30 00 CK SA OlC ETO-1 MME-1 AUGMENTED ETC-1 MME-1 PIT 4 LKO4 LKO-3 27VB-4 C4 2A AAC 30 00 CK SA OtC ETO-1 MME-1 AUGMENTED ETC-1 MME-1 PIT 4 i LKO-5 LKO-3 PAGE 73 OF 123 ,
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NEWYORKPChWR AUTHOPJTY JAMES A FITZPATRICK NUCLEAR POWER Pt. ANT INSERv1CE TESTING PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM Cer.tanmort Atmospherc Ddubon . SYSTEM ID 27 DRAWING FM-188 DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE &MTION REO7S CSJIROJ REQUEST TEST REMARKS 2/vB-$ C4 2A AC 30 00 CK $A OvC E IO-1 MME-1 AubMLNIED ETC-1 MME-1 PIT-5 LKO4 LKO-3 27VB 4 C4 2A AIC 20 00 CK SA OtC ETO-1 MME-t AUGMENTED ETC-1 MME-1 PfT-5 LKJ4 27VB-7 C4 2A NC 20 00 CK SA OtC ETO-1 MME-1 AUGMENTED ETC-1 MME-1 PfT-5 LKJ4 I
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, PAGE 74 OF 123 REV NO 9 O O
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NEWYCRKPOWER AUTHORtTY JAMES A FITZPATRICK 'NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES VALVE TABLE Contenment AtmosphencDemon - SYSTEMID 27 DRAWING FM-180 SYSTEM VALVE VALVE ACTUATOR SAFETY TEST RELEF ALTERNATE DWG TYPE TYPE FUNCTION RECrTS CSJ4tOJ - REQUEST TEST REMARKS VALVE ID COLORD CLASS CATEGORY SIZE (IN)
A 0375 GL SO C STC-1 AUGMENTED 21SOV-119t1 C-7 2A FSC-1 FAST ACTING VALVE PIT 4 LKJ4 0375 GL SO C STC-1 AUGMENTED '
27SOV-119E2 C4 2A A
' FSC-1 FAST ACTING VALVE PIT 4 LKJ4 0-4 2A A 0 375 GL So C STC-1 AUGMENTED
, 27SOV-119F1 FSC-1 FAST ACTINGVALVE PIT 4 LKJ4 A 0375 GL SO C STC-1 AUGMENTED s 27SOV-119F2 C4 2A FSC-1 FAST ACTING VALVE PIT 4 LKJ4 i
F4 2A A 0375 GL SO C STC-1 AUGMENTED 27SOV 120E1 FSC-1 FAST ACTING VALVE PIT 4 LkJ4 A 0375 GL SO C STC-1 AUGMENTED 27SOV-120E2 F4 2A FSC-1 FAST ACTING VALVE PfT4 LKJ4 A 0 375 GL SO C STC-1 ' AUGMENTED 27SOV-12CF1 F-4 2A FSC-1 FAST ACTING VALVE PIT-5 LKJ4 F-4 2A A 0375 GL SO C- STC 1 AUGMENTED 27SOV-120F2 FSC-1 FAST ACTING VALVE PIT 4 i
LKJ4 F4 2A A 0375 GL SO C STC-1 AUGMENTED i 27$0V-122E1 FSC-1 FAST ACTING VALVE PIT 4
' LKJ4 i
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NEWYORKPOWER AUTHORITY JAMES A FITZPATRICK NUCLEAR POWER PLANT WSERVICE TESTNG PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM Catmenmere Atmcachere Deuhon - SYSTEM O 27 DRAWING FM-180 DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVE 10 COORD CLASS CATEGORY SIZE (lN) TYPE TYPE FUNCTION REQ'TS CSJtROJ REQUEST TEST REMARKS 27SOV-122E2 F4 2A A 0 375 GL SO C STC-1 AUGMENT LD FSC-1 FAST ACTNG VALVE PIT 4 LKJ4 27SOV-122F1 G4 2A A 0375 GL SO C STC-1 AUGMENTED FSC-1 FAST ACTINGVALVE PIT 4 LKJ4 27SOV-122F2 G4 2A A 0375 GL SO C STC-1 AUGMENTED FSC-1 FAST ACTNG VALVE PtT4 LKJ4 27SOV-t23E1 E4 2A A 0375 GL SO C STC 1 AUGMENTED FSC-1 FAST ACTING VALVE PtT4 LKJ4 27SOV-123E2 E4 2A A 0 375 GL SO C STC-1 AUGMENTED FSC-1 F AST ACTNG VALVE PIT-5 LKJ4 27SOV-123F1 F-4 2A A 0375 GL SO C STC-1 AUGMENTED FSC-1 FAST ACTNG VALVE PIT 4 LKJ4 2750V-123F2 F-4 2A A 0375 GL SO C STC-1 AUGMENTED FSC-1 FAST ACTNGVALVE PIT -5 LKJ4 27SOV-124E1 C4 2A A 1 00 GL SO C STC-1 AUGMENTED FSC-1 FAST ACTINGVALW PfT-5 LKJ4 27sOV-124E2 C-4 2A A 1 00 GL SO C STC-1 AUGMENTED FSC-1 FAST ACTNGVALVE PtT4 LKJ4 76 OF 123 REV N
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i NEW YORK PCMER AUTHORITY JAMES A FITZPATRfCK NUCLEAR POWER Pt. ANT MSERVICE TESTNG PROGRAM FOR PUMPS AND VALVES . ,
j VALVE TASLE SYSTEM Ceremnmort Atmosphenc Dduren - SYSTEM O 27 DRAWING FM-100 f F
DWG VALVE VALW ACTUATOR SAFETY TEST RELIEF . ALTERNATE .. _'
VALVEID CO4RO CLASS CATEGORY SIZE ON) TYPE TYPE FUNCTION REQ'TS CSJ4tOJ REQUEST -TEST REMARKS 5 AUGMENTED ;
2?SOV-124F 1 C-4 2A A 0 375 GL SO - C STC-1 FSC-1 FAST ACTNG VALVE
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DWG VALVE . VALVE ACTUATOR SAFETY TEST RELIEF ALTERMATE VALVE ID CO-ORD CLASS CATEGORY S12E (W) TYPE TYPE FUNCTION REQ'TS CSJfROJ REQUEST TEST REMARKS-29AOV40A E4 1 A 24 00 GL AO C STC-1 ROJ-19 FSC-1 FSC-3 PIT 4 ;
LKJ4 p 29AOV408 D-5 1 A 24 00 GL AO .C STC-1 RO119 FSC-1 FSC-3 PfT-5 .
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29AOV46A G4 1 A 24 00 GL AO C STC-1 CSJ-13 FSC 1 FSC 2 Li Pfr-5 ,
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REV NO ; 2 PAGE 79 OF 123 ~ [
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NEW YORK POrKR AUTHORITY JAMES A FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM Man Stearn - SYSTEM ID 29 DRAWING FM-29A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVE O CO M CLASS CATEGORY StZE UN) TYPE TYPE FUNCTION REQTS CSJtROJ REQUEST TEST REMARKS 29LF v4GA F4 1 AIC 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKO4 LKO-3 29EFV-308 F4 1 AC 1 00 BK SA C ETC 1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKO4 LKO-3 29EFV-30C F-5 1 AIC 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE RSOLATES ON EXCESS FLOW LKO4 LKO4 29EFV-30D F-5 1 AtC 1 00 BK SA C ETC-1 ROJ41 E TC-3 VALVE ISOLATES ON EXCESS FLOW LKO4 LKO4 29EFV44A F4 1 A/C 1 00 BK SA C ETC 1 ROJ41 ETC4 VALVE RSOLATES ON EXCESS FLOW LKO4 LKO-3 29EFV 348 F4 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE tSOLATES ON EXCESS FLOW LKO-5 LKO4 29EFV-34C F4 i AC 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE POLATES ON EXCESS FLOW LKO4 LKO4 l
29E FV44D F4 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE RSOLAYES ON EXCESS FLOW l
LKO4 LKO-3 29EFV43A E4 1 At 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE RSOLATES ON EXCESS FLOW LKO4 LKO4 29ETV 538 E4 1 AIC 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKO4 LKO-3
, 29EFV-53C E4 1 AtC 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE rSOLAIES ON EXCESS FLOW LKO-S LKO-3
( 29EFV $3D E4 1 AJC 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE tSOLATES ON EXCESS FLOW LKO4 LKO-3 29EFV-54A E-5 1 AIC 1 00 BK SA C ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 29EFV-548 E4
- AtC 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE RSOLATES ON EXCESS FLOW LKO4 LKO-3 29EFv44C E4 1 A/C 1 00 BK SA C ETC-1 ROJ41 ETC-3 VALVE ISOLATES ON EXCESS FLOW LKO4 LKO-3 l
I REV NO 80 OF 123 l
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JAMES A FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRME FOR PUMPS AND VALVES VALVE TA8LE . _
SYSTEM Man Semen . SYSTEM O 29 ORAVWNG Fes-29A i DWG VALVE . VALVE ACTUATOR SAFETV TEST REUEF ALTERNATE VALVE 10 COCRD CLASS CATEGORY S:ZE (tN) TYPE TYPE FUNCTION REQTS CSJ440J -REQUEST TEST REMARKS 29EF V-640 E4 1 AC I LO SK SA C. E TC-1 ROJ01 EIC-3 VALVE tSOLATES ON EACESS FLOWW.
LKO4 LKG3 i 29MOV-200A C4 2A B 1 00 GL MO O STG1 AUGMENTED PIT 4 1
84 2A B 1 00 GL MO' O STG1 AUGMENTED 1 29MOV 2008
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i RT4 B4 2A B 1 DD GL - MO OC STOL1 AUGMENTED 29MOV-2028 STC 1 RT4 H-3 2A B 1 00 GL MO C STG1 CSA14 STO-2 AUGMENTED j 29MOV 203A PIT 4 ';
I i 29MOV-203B H3 2A B 1 00 GL MO O STot CSA14 STO-2 AUGMENTED i PfT4 i AUGMENTED f
- 29MOV-204A C4 2A B 1 00 GL MO C STC-1
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NEW YORK POWER AUTHORITY JAMES A FtTZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS ANO VALVES VALVE TABLE SYSTEM Emergency See Weser - SYSTEM O #'m ORAWING F8 35E DWG VALVE VALVE ACTUATOR SAFETY TEST REUEF ALTERNATE VALVE O COORD CLASS CATEGORY St2E (INp TYPE TYPE FUNCTION REQTS CSJ5tOJ REQUEST TEST RFMARKS 46t 10f tsW-toi G4 3 8 4 00 GA MA O E141 Roa22 EI43 46(70)ESW102 C4 3 8 4 00 GA MA O ETO1 ROJ-22 ETO3 46(70ESW-103 F4 3 8 4 00 GA MA O ET41 ROJ 22 EYG3 46(70)ESW-104 C4 3 8 4 00 GA MA O ETot ROJ 22 ET43 46(70pSWS-101 H.S 3 C 6 00 CK SA C RFC,t 46(70N102 H-8 3 C 6 00 CK SA C RFC-t 46(70)SWS t3 H4 3 8 0 00 GL MA C ETC1 a6(70)SWS-14 E4 3 8 6 00 GL MA C ETC-1 70TCV-120A F-7 3 8 2 00 3W AO O ST41 VRR46R1 FSG1 7DTCV-1208 C4 3 8 2 00 3W AO O SY41 VRR>J6R1 FSGI 70TCV-12 t A F4 3 8 2 00 3W AO O ST41 VRR46Rt FSG1 I 70TCV-1218 C7 3 8 2 00 3W AO O STG1 VRR46R1 FSG1 70WAC-12A F4 3 8 4 00 GA MA C ETC-1 70WAC-128 C4 3 8 4 00 GA MA C ETC1 70WAC-5A F-2 3 8 4 00 GA MA C ETC1 70WAC-58 42 3 8 4 00 GA MA C ETC-t GE e4 OF 123 REV NO
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< . JAMES A FtTZPATRICK MUCLEAR POWER Pt. ANT INSERWICE TESTING PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM Emergency Sennoe WMer - SYSTEM O. 46 DRAWtNG FM e DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE VALVE 10 COMO CLASS CATEGORY SIZE (#6 TYPE TYPE FUNCTION REOTS CSSROJ REQUEST TEST REMARKS 46ESW-19A 84 3 C 2 00 SK SA O FFT-t 46ESW-208 84 3 C 2 00 SK SA O FFT-1 i 84 3 C 2 00 SK SA O FFT1 46ESW-218 -
46ESW M 8-7 3 C 2 00 SK SA O FFT-1 46SWS47A B4 3 C 3 00 CK SA . C RFC-1 46SWS478 8-7 3 C 3 00 CK SA C RFC-1 46SWS48 84 3 C 3 00 CK SA C RFC-1 46SWS49 84 3 C 3 00 . CK SA C RFC-1 6
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PAGE 85 OF 123 REV NO 2 a
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NEW YORK POWER AUTHORITY JAMES A FITZPATRKX NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES VALVE TABLE SYSTEM Emergency Serwce Water - SYSTEM O 46 DRAWING FM-F,8 OWG VALVE VALVE ACTUATOR SAFETY TEST RELNEF ALTERNATE VALVE O CO-ORD CLASS CATEGORY SIZE ON) TYPE TYPE FUNCTION REQ'TS CSJtROJ REOUEST TEST REMARKS 4t>LSW-13A E4 3 C 3 00 CK SA O F F I-1 46ESW 138 C-2 3 C 3 00 CK SA O FF T-1 46E SW-1A E-7 3 C 12 00 CK SA O FFT-1 46ESW-1B D-T 3 C 12 00 CK SA O FFT-1 46ESWM E4 3 C 1 00 CK SA C RFC-1 46ESW40B E-4 3 C 1 00 CK SA C RFC-1 t
46ESW-?A E4 3 C GfM CK SA O FFT-1 46ESWJB E4 3 C 6 00 CK SA O FFT-1 4FESW-9A E-4 3 C 8 00 CK SA O FFT-1 46ESW-98 04 3 C 8 00 CK SA O FFT-1 46MOV-101 A E4 3 8 10 00 GA MO O STO-1 P1T-5 46MOV-t01B C4 3 8 10 00 GA MO O STO-1 PIT 4 46MOV-102A E4 3 8 8 00 GA MO C STC-1 PET 4 46MOV-1028 D4 3 8 8 00 GA MO C STC-1 PIT 4 46RV-112A G-7 3 C 6 00 RL SA C RLF4 46RV-1128 F4 3 C 6 00 RL SA C RLF 4 46RV 112C F-7 3 C 6 00 RL SA C RLF4 46RV-1120 G4 3 C 6 00 RL SA C RLF4 REV NO P 86 OF 123 O' O
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. I NEWYORK POWWER AUTHORITY [
JAMES A FITZPATRICK NUCLEAR POWER PLANT
. INSERVICE TESTesG PROGRAM FOR PUMPS AND VALVES ;
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i' VALVEID COCRO CLASS CATEGORY ' Size (IN) TYPE TYPE FUNCTION REOTS CSJROJ REQUEST - TEST' REMARKS -
46SWS40A . C-6 3 C . 4 00 CK SA C RFC-1 1 46SWS408 C-5 3 C 4 00 ' CK~ SA C RFC-1 f 1 .
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NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Cold Shutdown Justifications CSJ-01 SYSTEM: REACTOR WATER RECIRCULATION (RWR)
COMPONENTS: 02MOV-53A, B CATEGORY: B SAFETY FUNCTION: These valves close, on low reactor pressure to isolate the faulted loop coincident with initiation of the RHR System in the LPCI mode, to prevent diversion of LPCI flow.
JUSTIFICATION: To exercise these valves, the respective recirculation pump must be secured. Securing either pump (single loop operation) is limited by Technical Specification requirements and is not prudent. Single loop operation also requires a reduction in power.
These valves will be tested during cold shutdown and each refueling ot. 3e when Reactor Water Recirculation Pumps can be secured in accordance with OM-10 Section 4.2.1.2(f) and (g).
CSJ-02 SYSTEM: CONTROL ROD DRIVE HYDRAULICS (CRD)
COMPONENTS: 03HCU-115 (Typical for 137 HCUs) CATEGORY: C SAFETY FUNCTION: These valves close on initiation of a scram to prevent diversion of scram drive water into a depressurized charging header.
JUSTIFICATION: Exercising these valves during operation would require depressurization of the charging header with the potential for a loss of scram function.
These valves will be tested during cold shutdown and each refueling outage in accordance with OM-10 Section 4.3.2.2(f) and (g).
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Rev. No. 2 Page 88 of123 l
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NEW YORK POWER AUTHORITY l
' JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Cold Shutdown Justifications l CSh0) l SYSTEM: RESIDUAL HEAT REMOVAL (RHR)
COMPONENTS: 10AOV-68A, B CATEGORY: A/C SAFETY FUNCTION: These valves open to provide flowpaths for LPCI injection to the reactor vessel. They close for pressure isolation from the reactor vessel.
JUSTIFICATION: With the reactor at operating pressure, the RHR pumps cannot develop sufficient discharge pressure to open these valves. The installed air operators are designed to open these valves at zero differential pressure, which is not practical with the reactor at operating pressure. Therefore, i these valves cannot be full or part stroke exercised during nornal plant operation.
l Since there is no position indication for these valves, closure verification j must be done by backflow testing. Such testing during plant operation is
! impractical due to personnel safety concerns related to the potential release l of radioactive steam at high pressure.
These valves will be tested during cold shutdown and each refueling l l out. ge in accordance with OM-10 Section 4.3.2.2(f) and (g).
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Rev. No. 2 Page 89 of 123
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l NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Cold Shutdown Justifications CSJ-04 SYSTEM: RESIDUAL IIEAT REMOVAL (RIIR)
COMPONENTS: 10MOV-17 & 10MOV-18 CATEGORY: A SAFETY FUNCTION: These valves remain closed to protect the RHR System piping and components from overpressurization during plant operation and inadvertent drain down events while in cold shutdown. 10MOV-17 also performs a containment isolation function.
JUSTIFICATION: With the reactor pressure greater than 75 psig, these valves are prevented from opening by an electrical interlock.
These valves will be tested during cold shutdown and each refueling outage in accordance with OM-10 Section 4.2.1.2(f) and (g).
CSJ45 SYSTEM: REACTOR CORE ISOLATION COOLING (RCIC)
COMPONENTS: 13RCIC-7 CATEGORY: A/C SAFETY FUNCTION: This valve opens to allow condensate drainage from the steam exhaust piping to the suppression chamber. It closes for containment isolation.
JUSTIFICATION: Closure verification for this valve is accomplished by performing a back flow test where the drain line is isolated from the steam exhaust line.
Placing the RCIC system in this configuration during plant operation is l undesirable and could adversely affect the plant's response in the event of a transient.
This valve will be tested during cold shutdown and each refueling outage in accordance with OM-10 Section 4.3.2.2(f) and (g).
O Rev. No. 2 Page 90 of 123
1 NEW YORK POWER AUTIIORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B '
1 Cold Shutdown Justifications CSJ46 SYSTEM: REACTOR CORE ISOLATION COOLING (RCIC) l COMPONENTS: 13RCIC-37 & 13RCIC-38 CATEGORY: C SAFETY FUNCTION: These valves open to eliminate any differential pressure that could force .
water from the suppression chamber into the RCIC steam exhaust piping l when the suppression chamber pressure is greater than atmospheric.
1 JUSTIFICATION: Verifying proper operation of these valves involves a test that requires isolation of the vacuum breakers for an extended period of time. During this test, the RCIC system is considered to be inoperable. Due to operational concerns associated with the plant's response to possible transients without an operable RCIC system, it is considered to be q imprudent to test these valves while the plant is operational.
V These valves will be tested during cold shutdown and each refueling outage in accordance with OM-10 Section 4.3.2.2.(f) an (g).
CSJ-07 SYSTEM: REACTOR BUILDING CLOSED LOOP COOLING (RBC)
COMPONENTS: 15AOV-130A, B; 15AOV-131 A, B 15AOV-134A CATEGORY: A SAFETY FUNCTION: These valves close to provide containment isolation.
JUSTIFICATION: During normal plant operation, these valves must remain open to provide cooling water to the Drywell coolers and Drywell equipment drain sump cooler. Closing these valves during plant operation could cause a spike in drywell pressure due to the loss of cooling water flow, which may result in a reactor scram and plant shutdown.
These valves will be tested during cold shutdowns and each refueling outage in accordance with OM-10 Section 4.2.1.2(f) and (g).
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Rev.No. 2 Page 91 of 123
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEA.R POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Cold Shutdown Justifications CSI-08 SYSTEM: REACTOR BUILDING CLOSED LOOP COOLING (RBC)
COMPONENTS: 15AOV-132A, B; 15AOV-133A, B CATEGORY: A SAFETY FUNCTION: These valves close to provide containment isolation.
JUSTIFICATION: During normal plant operation, these valves must remain open to provide cooling water to the recirculation pump motor and seal coolers. Closing these valves would result in damage to the recirculation pumps.
These valves will be tested during cold shutdowns and each refueling outage in accordance with OM-10 Section 4.2.1.2(0 and (g).
CSJ-09 SYSTEM: IIIGII PRESSURE COOLANT INJECTION (HPCI)
COMPONENTS: 23HPI-13 CATEGORY: A/C SAFETY FUNCTION: This valve opens to allow condensate drainage from the steam exhaust piping to the suppression chamber. It closes for contair. ment isolation.
JUSTIFICATION: Closure verification for this valve is accomplished by performing a back flow test where the drain line is isolated from the steam exhaust line and the torus is vented to atmosphere. Placing the HPCI system and containment in this configuration during plant operation is undesirable and could adversely affect the plant's response in the event of an accident.
This valve will be tested during cold shutdowns and each refueling outage in accordance with OM-10 Section 4.3.2.2(f) and (g).
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Rev. No. 2 Page 92 of123
NEW YORK POWER AUTHORITY l JAMES A. FITZPATRICK NUCLEAR POWER PLANT
(~) INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES V
APPENDIX B i Cold Shutdown Justifications l
CSJ-10 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)
COMPONENTS: 23HPI-18 CATEGORY: C SAFETY FUNCTION: This valve opens to provide a flowpath for the HPCI system injection to l the reactor vessel.
JUSTIFICATION: With the reactor at operating pressure, the HPCI pump can develop sufticient discharge pressure to open this valve, however HPCI injection of cold water to the reactor vessel during critical operation could result in an undesirable reactivity excursion and thermal transient to the piping <
components. During plant operation, the differential pressure developed across the valve disc could be in excess of 1000 psid - precluding manual manipulation of the valve. Therefore, these valves cannot be exercised Os during normal plant operation.
This valve will be tested during cold shutdown and each refueling outage in accordance with OM-10 Section 4.3.2.2(f) and (g).
CSJ-11 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)
COMPONENTS: 23HPI-402 and 23HPI-403 CATEGORY: C SAFETY FUNCTION: These valve open to eliminate any differential pressure that could force water from the suppression chamber into the HPCI exhaust piping when the suppression chamber pressure is greater than atmospheric. They close to prevent HPCI exhaust stem from entering the suppression chamber air space, thus bypassing the yten.hing action of the torus.
JUSTIFICATION: Operation of the HPCI pump turbine does not prove operability of these valves and special testing is required. This testing necessitates isolation of the vacuum breaker piping, which results in the inoperability of the HPCI system for the duration of the test. Due to the importance of the HPCI system function and the lack of a redundant HPCI train, it is not ph considered prudent to perform this testing during plant operation at power.
Rev. No. 2 Page 93 of123
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Cold Shutdown Justifications CSJ-11 (Continued _1 These valves will be tested during cold shutdown and each refueling outage in accordance with OM-10 Section 4.3.2.2(f) and (g).
CSJ-12 SYSTEM: CONTAINMENT VENT & PURGE (CAD)
COMPONENTS: 27AOV-111,112,113 CATEGORY: A 27AOV-114,115,116 SAFETY FUNCTION: These valves close to provide a containment isolation function.
JUSTIFICATION: Due to NRC concerns that these valves will not close under Design Basis Accident conditions, they will not be opened whenever primary containment is required except for safety-related reasons. For this reason, these valves will be tested during cold shutdown and each refueling outage in accordance with OM-10 Section 4.2.1.2(f) and (g).
CSJ-13 SYSTEM: MAIN STEAM (MSS)
COMPONENTS: 29AOV-86A, B, C, D CATEGORY: A SAFETY FUNCTION: These valves close to provide containment isolation.
JUSTIFICATION. Performance of the fail close test for the MSIVs requires entry into the Steam Tunnel. This cannot be done during normal operation.
These valves will be tested during cold shutdown and each refueling outage in accordance with OM-10 Section 4.2.1.2(f) and (g).
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Rev. No. 2 Page 94 of 123
l l NEW YORK POWER AUTHORITY l JAMES A. FITZPATRICK NUCLEAR POWER PLANT l
l INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES I APPENDIX B Cold Shutdown Justifications CSJ-14 l SYSTEM: MAIN STEAM (MSS)
COMPONENTS: 29MOV-203A, B CATEGORY: .B l SAFETY FUNCTION: These valves open to provide flowpaths for post-accident MSIV packing leak-off to the Standby Gas Treatment System.
JUSTIFICATION: Opening these valves during power operation could subject downstream piping to pressures in excess of its 150 psig design pressure.
These valves will be tested during cold shutdown and each refueling outage in accordance with OM-10 Section 4.2.1.2(0 and (g).
fm CSJ-15
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SYSTEM: FEEDWATER (FWS)
COMPONENTS: 34NRV-111 A, B CATEGORY: A/C SAFETY FUNCTION: These valves close to provide containment isolation and to prevent diversion of HPCI flow into the feedwater system.
JUSTIFICATION: Exercising these valves during operation would require isolation of feedwater flow to the reactor vessel. This is neither prudent nor practical without a plant shutdown.
These valves will be tested during cold shutdown and each refueling outage in accordance with OM-10 Section 4.3.2.2(0 and (g).
Rev. No. 2 Page 95 of123
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Cold Shutdown Justifications CSJ-16 SYSTEM: CONTAINMENT VENT & OURGE (CAD)
COMPONENTS: 27MOV-120 CATEGORY: B SAFETY FUNCTION: This valve is closed to provide isolation for one path of contaimnent purge
- to the Standby Gas Treatment System to ensure purge flow doesn't exceed filter capacity. The valve is opened to connect either the drywell atmosphere or the torus atmosphere to SBGT for normal containment venting and purging when primary containment is not required. The valve maybe required to be opened to vent primary containment to SBGT under severe accident conditions.
JUSTIFICATION: This valve is required to be closed whenever primary containment is required (Tech Spec Amendment 154).
These valves will be tested during cold shutdown and each refueling outage in accordance with OM-10 Section 4.2.1.2(f) and (g).
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Rev. No. 2 Page 96 of123
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, NEW YORK POWER AUTHORITY . I
- JAMES A. FITZPATRICK NUCLEAR POWER PLANT f
O INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Refueline Outane Justifications l
ROJ-01 S'YSTEM: VARIOUS 1
I COMPONENTS: Excess Flow Check Valves CATEGORY: A/C (Listed Below)
SAFETY FUNCTION: These valves close to isolate the respective instrument lines in the event of '
a pipe break downstream of the valves.
- JUSTIFICATION
- Exercising these valves requires isolation of their associated safety-related instrument, which could place the plant in an unsafe condition. In addition, the induced hydraulic transients resulting from establishing flow and subsequent valve closure would most likely result in an engineered safety feature actuation. -During such testing, radiation doses to test personnel would be high due to the location of these valves and reactor i'
water effluent during the test.
l These valves.cannot be tested during cold shutdown since the reactor I vessel is not pressurized.
l These valves will be tested during refueling outages during the primary system inservice pressure test in accordance with OM-10 Section 4.3.2.2(e) and (h).
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02-2EFV-PS-128A,B 02-3EFV-19A,B 14EFV-31 A,B 02-2EFV-PT-24A,B 02-3EFV-21A,B,C,D 23EFV-01A,B 02-2EFV-I'T-25A,B 02-3EFV-23A,B,C,D 23EFV-02A,B l 02-2EFV1-DPT-111 A,B 02-3EFV-23 29EFV-30A,B,C,D 02-2EFV1-FT-110A,C.E.G 02-3EFV-25 29EFV-34A,B,C,D 02-2EFV2-DPT-111 A,B 02-3EFV-31 A,B,C,D 29EFV-53A,B,C,D 02-2EFV2-FT-110A,C,E,G 02-3EFV-31E,F,G,H 29EFV-54A,B,C,D 02-3EFV-11 02-3EFV-31J,K,L,M 02-3-EFV-13A,B 02-3EFV-31N,P,R,S
, .c ' 02-3EFV-15A,B . 02-3EFV-33
( 02-3EFV-15N 13EFV-01A,B 02-3EFV-17A,B 13EFV-02A,B
- Rev. No. 2 Page 97 of123 l
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NEW YORK POWER AUTIIORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT 9
INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Refueline Outace Justifications ROJ-02 SYSTEM: REACTOR WATER RECIRCULATION (RWR)
COMPONENTS: 02-2RWR-13A, B CATEGORY: A/C SAFETY FUNCTION: These recirculation pump seal water injection valves close to provide containment isolation.
JUSTIFICATION: Exercising these valves during normal operations or cold shutdown requires securing the Recirculation pumps and entering containment to check the valves closed by using a back-leakage test. Testing during operations is therefore impossible.
Testing during cold shutdown by performing back-leakage tests would require extensive time for test equipment set-up and place an undue burden on the plant staff. In addition, entry into the containment may be prohibited if the drywell remains inerted.
Back-leakage testing and leakrate testing will be performed during each refueling outage in accordance with OM-10 Section 4.3.2.2(e) and (h).
ROJ-03 SYSTEM: REACTOR WATER RECIRCULATION (RWR)
COMPONENTS: 02-2RWR-41 A,B CATEGORY: A/C SAFETY FUNCTION: These recirculation pump seal purge check valves close to provide containment isolation.
JUSTIFICATION: Closing these valves any time Reactor Water Recirculation Pumps are running subje:ts the pump seals to thermal transients and pressure fluctuations, thereby, shortening seal life. Pressure fluctuations and oscillations can degrade the pressure-retaining ability of either or both seal stages. Additionally, securing seal purge flow while the Reactor Water Recirculation Pumps are running introduces reactor coolant and associated corrosion products into the seal cavity, which also shortens seal life.
These valves will be tested during each refueling outage during leak testing performed per 10CFR50, Appendix J, in accordance with OM-10 Section g 4.3.2.2(e) and (h). W i l
Rev. No. 2 Page 98 of123
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT p INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES t.] APPENDIX B Refuelinn Outane Justifications ROJ-04 SYSTEM: AUTOMATIC DEPRESSURIZATION (ADS)
COMPONENTS: 02RV-1 through 02RV-11 02VB-1 through 02VD-11 CATEGORY: C SAFETY FUNCTION: These valves remain closed to prevent steam from an open safety / relief valve (SRV) from entering the drywell. They open following closure of an SRV to prevent the formation of a water column within the downcomer that could cause toms damage during subsequent lifting of the same SRV.
JUSTIFICATION: Exercising these valves requires local manipulation of each valve and thus entry into the containment. During plant operation at power, and on occasion while in cold shutdown, the containment atmosphere is maintained in a nitrogen-inerted condition. During such periods, entry into the containment is not practical due to personnel safety concerns.
Testing will be performed during each refueling outage in accordance with OM-10 Section 4.3.2.2(e) and (h).
- ROJ-05 SYSTEM: RESIDUAL HEAT REMOVAL (RHR)
COMPONENTS: 10RHR-64A, B, C, D CATEGORY: C SAFETY FUNCTION: These valves open on forward flow to provide minimum flow protection for the RHR pumps and close on reverse flow to prevent diversion of flow through an idle parallel pump. i JUSTIFICATION: These valves are exercised open every three months by flow during pump testing. However, quantitative flow measurements as a means of verifying these valves open has been determined to be impractical. 1 i
There is no installed flow instrumentation in the minimum flow line thus I attempts at flow measurements are being made with a strap on ultrasonic I flow meters. Due to the minimum flow line configuration and operating conditions, there is a high amount of cavitation / turbulence in the line I O ;
i Rev. No. 2 Page 99 of.l_21
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Refueling Outace Justifications ROJ-05 (Continued) causing the ultrasonic flow meter to go into fault. Attempts have been made at different locations and with different size transducers, and faults still occur.
This test method requires the RHR pumps to be operated repeatedly (three to four times) at minimum flow conditions for the maximum time period allowed by procedure. Running at this condition is undesirable, particularly for a test method that frequently does not yield meaningful results. NRC Information Notice 89-08 documented concerns about pump damage by operating at low flow conditions. When this test is performed with no flow measurements being taken, the time spent at minimum pump flow is short.
In addition, this testing must be performed in a radiation area, which has caused increased exposure to personnel while multiple test attempts and transducer repositioning are accomplished. It is concluded that con efforts with this method are not practical.
Attempts were made to distinguish the check valve opening impact on the valve bonnet using a seismic vibration probe. Meaningful results could not be obtained again due to the high background noise and vibration associated with a pump start at minimum flow.
The method of using process flow and pressure instrumentation in the main line to infer the flow in the minimum flow line was investigated.
Ilowever, the small flow rate through the minimum flow line in comparison with the main line flow would not be discernable within the )
accuracy of the process instrumentation.
In accordance with Generic Letter 89-04, Position 2, during each refuel .
outage at least one (1) valve will be disassembled, inspected, and verified f operable. The acceptance criteria as stated in the Generic Letter is I provided in the maintenance procedure used for check valve disassemble. l If any valve is found to be inoperable, the remaining valves will be disassembled and inspected prior to startup. The inspection schedule will be such that all four (4) valves in tne group are inspected at least once l every six (6) years.
O Rev. No. _2_ Page 100 of 123
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NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT l
n INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES i
l V APPENDIX B Refuelina Outace Justifications ROJ4)6 SYSTEM: RESIDUAL HEAT REMOVAL (RHR)
COMPONENTS: 10RHR-95A,B CATEGORY: C SAFETY FUNCTION: These valves close to prevent reverse flow from the torus.
JUSTIFICATION: These are simple check valves with no means of determining disc position without performing a back leakage test. Performing such a test during plant operations would require setting up a test rig and performing a hydrostatic test. As discussed in NUREG 1482, section 4.1.4, the NRC has determined that the need to set up test equipment is adequate justification to defer backflow testing of a check valve until a refueling outage.
During cold shutdown, the system lineup changes and the effort involved ,
with setting up test equipment would constitute an unreasonable burden on '
the plant staff, )
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These valves will be verified to close each refueling outage during a hydrostatic leak rate test in accordance with OM-10 Section 4.3.2.2(e) and (h).
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Rev. No. 2 Page 101 of 123
l NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Refueling Outage Justifications ROJ-07 SYSTEM: STANDBY LIQUID CONTROL (SLC)
COMPONENTS: 11SLC-16 & 11SLC-17 CATEGORY: A/C SAFETY FUNCTION: These valves prohibit backflow from the reactor vessel to the SLC System and provide for containment isolation. They open to permit SLC System flow to the reactor vessel.
JUSTIFICATION: Full or partial-stroke exercising these valves requires that flow be established through the subject check valves. The only practical means of initiating flow through these valves requires actuation of the SLC system and pumping from the SLC Tank to the reactor vessel. During normal plant operation, this would introduce boron into the reactor w.ssel resulting in unacceptable reactivity and chemistry transients. Testirg during cold i
shutdown would result in chemistry transients and undue burden on plant staff with respect to maintenance of the SLC pump explosive valves.
Testing will be conducted during each refueling outage and as required by Technical Specifications, by injecting water into the reactor vessel by use of the Standby Liquid Control pumps. Following the exercise open test, the valves will be verified to close by means of a back-leakage test.
O Rev. No. 2 Page 102 of 123
NEW YORK POWER AUTIIORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Refueline Outane Justifications ROJ-08 SYSTEM: REACTOR CORE ISOLATION COOLING (RCIC)
COMPONENTS: 13RCIC-04 and 13RCIC-05 CATEGORY: A/C i SAFETY FUNCTION: These valves close to provide containment isolation.
JUSTIFICATION: There is no provision on either of these valves that provides position l indication of the disc. As a result, valve closure must be verified by back- l leakage testing. In order to verify valve closure by the back-leakage technique, the RCIC exhaust line must be isolated for the duration of the l test causing the RCIC system to be inoperable, l The potential safety impact of voluntarily placing the RCIC system in an inoperable status during plant operation at power is considered to be O4 imprudent and unwarranted in relation to any apparent gain in system reliability derived from the dosure verification. In addition, the valves are located approximately twenty (20) feet from the floor necessitating erection of a large scaffold in the vicinity of the RCIC pump. This also is considered to be undesirable from the aspect of potential damage to RCIC system components should the scaffold be subjected to structural failure.
Based on the foregoing discussion, testing of these valves during plant operation at power is considered to be impractical. During cold shutdowns, erection of the scaffold in addition to other activities related to test performance would place an extreme burden on the plant staff and would likely result in unwarranted extensions to all forced outages with the added negative impact on plant performance and availability.
These valves will be verified to close by performing a back-leakage test at l each refueling outage in accordance with OM-10 Section 4.3.2.2(e) and l (h). l O
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NEW YORK POWER AUTIIORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B
}1efuelit1_c Outane Justifications ROJ-09R1 l SYSTEM: CORE SPRAY (CSP)
COMPONENTS: 14AOV-13A,B CATEGORY: A/C SAFETY FUNCTION: These valves open to provide flowpaths from the Core Spray System to the reactor vessel. They close for pressure isolation protection of the low pressure core spray piping.
JUSTIFICATION: There is no mechanism by which these valves can be full-stroke exercised without injecting water from the core spray pumps to the reactor vessel.
During plant operation, the core spray pumps cannot produce suflicient discharge pressure to overcome reactor vessel pressure and provide flow into the vessel.
The installed air operators are capable of exercising the valves, providing there is not differential pressure across the valve seat. During plant operation, there is a significant differential pressure across the valve seat.
During cold shutdown, injecting into the reactor vessel requires a major effort to establish the prerequisite conditions and realignment of the Core Spray system to allow supplying water from the Condensate Storage Tank.
Torus water cannot be used since it does not meet the chemistry requirements for reactor grade makeup. It is estimated that such a test would take about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to perform and would result in a significant burden on the plant operating staff. In addition, there is a potential for overfilling the reactor vessel and flooding the main steam lines. This could adversely affect the performance of the main steam safety / relief valves (SRVs) since a contributing factor to the historically poor performance of the SRVs is water contamination of the operators.
During cold shutdowns, each of the valves will be exercised using the installed air operators (considered a partial-stroke). This test satisfies the exercising of both safety positions.
Each of the valves will be full-stroked exercised during each refuel outage in accordance with OM-10 Section 4.3.2.2(e) and (h) by injecting full accident flow into the reactor vessel. The closed position is leak tested every 24 months per OM-10 Section 4.2.2.3(a). This position complies with the guidance of NUREG-1482, Section 4.1.4.
Rev. No. 2 Page 104 of 123
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT
/~' INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES i
l APPENDIX B :
Refueline Outace Justifications ROJ-10 SYSTEM: CORE SPRAY (CSP)
COMPONENTS: 14 CSP-62A,B CATEGORY: C SAFETY FUNCTION: These valves close to prevent reverse flow from the torus.
JUSTIFICATION: There are no position indicators or other means to verify closure of these valves. As a result, valve closure must be verified by back-leakage testing. Performing such a test during plant operations would require setting up for and performing a hydrostatic test. As discussed in NUREG 1482, section 4.1.4, the NRC has detennined that the need to set up test
, equipment is adequate justification to defer backflow testing of a check i
valve until a refueling outage.
During cold shutdown, the system lineup changes and the effort involved with setting up test equipment would constitute an unreasonable burden on the plant staff.
These valves will be verified close each refueling outage in accordance with OM-10 Section 4.3.2.2(e) and (h) during a hydrostatic leak rate test.
ROJ-11 SYSTEM: REACTOR BUILDING CLOSED LOOP COOLING (RBC)
COMPONENTS: 15RBC-214 CATEGORY: C SAFETY FUNCTION: This valve closes to prevent flow diversion when the Emergency Service Water system is supplying cooling water to RBC heat loads.
JUSTIFICATION: There is no provision on this valve that provides position indication of the disc. There are no test taps and block valves to enable a back-leakage test to verify closure. OM-10, Section 4.3.2.4(c) allows disassembly each p refueling outage to verify operability as an alternative to quarterly testing.
V Rev. No. 2 Page 105 of L2]
i NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES l 4
APPENDIX B Refueline Outane Justifications ROJ-12 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)
COMPONENTS: 23HPI-12 and 23HPI-65 CATEGORY: A/C SAFETY FUNCTION: These valves close to provide containment isolation.
JUSTIFICATION: There is no provision on either of these valves that provides position indication of the disc. As a result, valve closure must be verified by back-leakage testing. In order to verify valve closure by the back-leakage technique, the HPCI exhaust line must be isolated for the duration of the test causing the HPCI system to be inoperable. The potential safety impact of voluntarily placing the HPCI system in an inoperable status during plant operation at power is considered to be imprudent and unwarranted in relation to any apparent gain in system reliability derived from the closure verification. In addition, the valves are located approximately twenty (20) feet from the floor necessitating erection of a large scaffold in the vicinity of the HPCI pump. This also is considered to be undesirable from the aspect of potential damage to HPCI system components should the scaffold be subjected to structural failure.
Based on the foregoing discussion, testing of these valves during plant operation at power is considered to be impractical. During cold shutdowns, erection of the scaffold in addition to other activities related to test performance would place an extreme burden on the plant staff and would likely result in unwarranted extensions to all forced outages with the added negative impact on plant performance and availability. These valves will be verified to close by performing a back-leakage test at each refueling outage in accordance with OM-10 Section 4.3.2.2(e)and (h).
O Rev. No. 2 Page 106 of 123
1 NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT j l INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES l APPENDIX B Refueline Outace Justifications l l
ROJ-13 !
SYSTEM: HIGH PRESSURE COOLANT INJECTION OIPCI)
COMPONENTS: 23HPI-13 and 23HPI-56 CATEGORY: C l l
SAFETY FUNCTION: These valves opens to permit HPCI turbinc condensate to drain to the torus.
JUSTIFICATION: There are no means for exercising these valves to the open position where positive indication of acceptable valve performance is verified. OM-10, Section 4.3.2.4(c) allows disassembly each refueling outage to verify operability as an alternative to quarterly testing.
ROJ-14 SYSTEM: HIGH PRESSURE COOLANT INJECTION OIPCI)
COMPONENTS: 23HPI-32 CATEGORY: C l SAFETY FUNCTION: This valve closes during the suction swap from the Condensate Storage Tank to the torus to prevent diversion of the torus flow from the HPCI pump suction.
JUSTIFICATION: There is no provision on this valve that provides position indication of the i disc. There are no block valves between this valve and the suction of the !
HPCI pump to enable a back-leakage test to verify closure. OM-10, Section 4.3.2.4(c) allows disassembly each refueling outage to verify l operability as an alternative to quarterly testing.
O Rev. No. 2 Page 107 of122
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Refueline Outace Justifications ROJ-15 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)
COMPONENTS: 23HPI-61 CATEGORY: C SAFETY FUNCTION: This valve opens to provide a flowpath from the torus to the suction of the HPCI booster pump.
JUSTIFICATION: The only practical method available to full flow exercise this valve is to pump water from the torus into the reactor vessel. Due to the lack of suitable water quality in the torus, this option is not practical. OM-10, Section 4.3.2.4(c) allows disassembly each refueling outage to verify operability as an alternative to quarterly testing. In addition, this valve will be partial-flow tested once per operating cycle.
ROJ-16 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)
COMPONENTS: 23HPI-62 CATEGORY: C SAFETY FUNCTION: This valve opens to provide a flowpath for minimum flow from the HPCI main pump.
JUSTIFICATION: Due to the configuration of the minimum flow motor operated valve control logic, fully developed flow cannot be achieved through this check valve. Additionally, full-stroke exercising cannot be verified with existing instrumentation. OM-10, Section 4.3.2.4(c) allows disassembly each refueling outage to verify operability as an alternative to quarterly testing.
O Rev. No. _2_. Page 108 ofJ23,
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT l q INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B i
l Behteline Outace Justifications ROJ-17 SYSTEM: HIGII PRESSURE COOLANT INJECTION (HPCI) l COMPONENTS: 23HPI-130 CATEGORY: C SAFETY FUNCTION: This valve opens to provide a flowpath for cooling water circulation through the HPCI turbine lube oil cooler and closes to prevent flow diversion.
JUSTIFICATION: This valve has no means of determining disc position or flowrate and, thus there is no mechanism for verifying full accident flow. In addition, there are no test taps and block valves to enable a back leakage test to verify closure. OM 10, Section 4.3.2.4(c) allows disassembly each refueling outage to verify operability as an alternative to quarterly testing.
ROJ 18 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)
COMPONENTS: 23HPI-131 CATEGORY: C SAFETY FUNCTION: This valve closes to prevent flow diversion from the HPCI booster pump.
JUSTIFICATION: There is no provision on this valve that provides position indication of the disc. There are no test taps and block valves to enable a back-leakage test to verify closure. OM-10, Section 4.3.2.4(c) allows disassembly each refueling outage to verify operability as an alternative to quarterly testing.
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NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Befueling Outace Justifications ROJ-19 SYSTEM: MAIN STEAM (MSS)
COMPONENTS: 29AOV-80A,B,C,D CATEGORY: A SAFETY FUNCTION: These valves are normally open to provide steam to the main turbine generator and auxiliaries. They close to isolate steam flow and for containment isolation.
JUSTIFICATION: Fail safe exercising these valves requires local manipulation of valves located inside containment. During plant operation at power, and on occasion while in cold shutdown, the containment atmosphere is maintained in a nitrogen-inerted condition. During such periods, entry into the containment is not practical due to personnel safety concerns.
These valves will be verified to fail safe close at each refueling outage in accordance with OM-10 Section 4.2.1.2(e) and (h).
ROJ-20 SYSTEM: FEEDWATER (FWS)
COMPONENTS: 34FWS-28A, B CATEGORY: A/C SAFETY FUNCTION: These valves close to provide containment isolation upon cessation of feedwater flow during accident conditions.
JUSTIFICATION: There is no provision on either of these valves that provides position indication of the disc. As a result, valve closure must be verified by back-leakage testing. During plant operation at power, these valves cannot be closed without precipitating a plant shutdown.
During cold shutdowns, performing a back-leakage test requires entry into the containment vesses and extensive system preparations, including draining of the main feedwater piping from the outlet of the sixth point feedwater heaters to the reactor vessel isolation valves (approrimately 2000 gallons per line). Furthermore, testing of 34FWS-28B requires shutdown of the cleanup system. It is estimated that testing either of these Rev. No. 2 Page 110 of121
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NEW YORK POWER AUTIIORITY '
JAMES A. FITZPATRICK NUCLEAR POWER PLANT l
! INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Refueline Outace Justifications l
BOJ-20 (Continued) ,
l valves would require up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and demand significant staff resources. Also, entry into the containment at cold shutdown with the l containment inerted is a personnel safety concern. l Closure of these valves will be demonstrated during each refuel outage in accordance with OM-10 Section 4.3.2.2(e) and (h) by conducting a back-leakage test.
ROJ-21 SYSTEM: INSTRUMENT AIR (IAS)
COMPONENTS: 391AS-22 & 391AS-29 CATEGORY: A/C SAFETY FUNCTION: These valves open to provide nitrogen to the MSIVs and the SRV y accumulators inside the containment. They close for containment isolation.
JUSTIFICATION: Exercising these valves open is performed by charging the bleed-down header following MSIV testing. During plant operation at power, this is impractical since closure of the MSIVs would cause a plant trip. Also performing such a test requires entry into the containment vessel and local manipulation of test connections located inside the drywell.
During plant operation at power and, on occasion, while in the cold shutdown mode, the containment atmosphere is maintained in a nitrogen-inerted condition. During such periods, entry into the containment is not practical due to personnel safety concerns.
These valves will be tested open at each refueling outage in accordance with OM-10 Section 4.3.2.2(e) and (h).
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NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Refueling Outace Justifications ROJ-22 SYSTEM: EMERGENCY SERVICE WATER (ESW)
COMPONENTS: 46(70)ESW-101,102,103,104 CATEGORY: B SAFETY FUNCTION: These valves are manually opened to provide ESW flow to Control and Relay Room air handlers to ensure continued cooling in the event the normal chilled water system is rendered inoperable.
JUSTIFICATION: These valves provide isolation between the raw ESW System and the i glycol / water mixture in the chilled water system. Opening these valves I will cause contamination of the glycol / water solution. Therefore, it is not !
practical to test these valves during plant operation.
During cold shutdown, extensive time would be required to orain the ;
glycol from the system to prevent contamination. This would constitute an l unreasonable burden on the plant staff.
I These valves will be exercised open during each refueling outage in accordance with OM-10 Section 4.2.1.2(e) and (h).
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I NEW YORK POWER AUTHORITY
- JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Valve Relief Reauests !
VfPR-01 SYSTEM: AUTOMATIC DEPRESSURIZATION (ADS)/ MAIN STEAM COMPONENTS: 02RV-71A,B,C,D,E,F,G,H,J,K and L CATEGORY: B/C CLASS: 1 1
FUNCTION: These valves open when actuated by a manual switch to reheve reactor i pressure during an accident or transient condition. Valves 02RV-71 A, B, C, D, E, G, and H open on receipt of ADS actuation signal.
TEST REQUIREMENT: OM-10, Section 4.2.1.4 - stroke time for power operated valves O
V BASIS FOR RELIEF: These valves are fast-acting valves and do not have position indication.
Therefore, stroke time cannot be effectively measured.
When testing these valves, a reactor pressure of at least 50 psig is needed for operdng by the pilot assembly and a minimum reactor pressure of 940 psig is specified to minimize potential damage to the pilot valve and disc surfaces. Testing at each startup from a cold shutdown would produce additional stress cycles, which may lead to a low cycle fatigue failure.
ALTERNATE TESTING: Following each refuel outage or once each operating cycle with reactor pressure at least 940 psig, these valves will be exercised in accordance 4 with the operational test requirements set forth in the JAF Technical Specifications. SRV tailpipe temperatures and acoustic monitors will be used to verify valve opening.
A) b Rev. No. 2 Page 113 of 123
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Valve Relief Reauests VRR-02 SYSTEM: AUTOMATIC DEPRESSURIZATION (ADS)/ MAIN STEAM COMPONENTS: 02RV-71A,B,C,D,E,F,G,H,J,K and L CATEGORY: B/C CLASS: 1 FUNCTION: These valves open to relieve reactor pressure during an accident or transient condition.
TEST REQUIREMENT: OM-1, Section 3.3.1.1 - Periodic testing of Class 1 Pressure Relief Valves BASIS FOR RELIEF: Currently during refueling outages, the SRV pilot assembly is removed and transported to a certified valve testing facility for performance of the following tests: setpoint (lift pressure), rescat (reclosing pressure),
and pilot stage seat tightness. A main body slave is used to test each pilot. ANSI /ASME OM-1 states, "No maintenance, adjustment, disassembly, or other activity which could affect as found set pressure or seat tightness data is permitted prior to testing." Since main body seat leakage is monitored continuously during normal plant operation, its seat tightness as found determination is satisfied prior to the pilot assembly removal.
ANSI /ASME OM-1 also states, " Tests prior to maintenance or set pressure adjustment, or both, shall be performed in the following sequence: (a) visual examination; (b) seat tightness determination; (c) set pressure determination; (d) determination of compliance with the Owner's set tightness criteria; (e) determination of electrical characteristics and pressure integrity of solenoid valves; (f) determination of pressure integrity and stroke capability of air actuator; (g) determination of operation and electrical characteristics of position l
indicators; (h) determination of operation and electrical characteristics l of bellows alarm switet.; and (i) determination of actuating pressure of auxiliary actuating device sensing element, where applicable, and g electrical continuity". W Rev. No. 2 Page 114 of 123
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT l
,q INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES V
APPENDIX B Valve Relief Requests VRR-02 (Continued)
Strict adherence to the sequence cannot be satisfied by testing the pilot assembly only. Currently, the plant's test practices ensure that l applicable tests specified in ANSI /ASME OM-1 Section 3.3.1.1, Main Steam Pressure Relief Valves with Auxiliary Actuating Devices, are performed and the entire valve operability is verified in accordance l with Technical Specifications, but not in the sequence specified by OM-1 Section 3.3.1.1.
Common industry practice is to test the Target Rock safety / relief SRV pilot assemblies as separate units. Therefore, removal of the entire valve assembly for testing would create hardship by (1) extending plant i outages for the removal and installation process, (2) cost increase and l
schedule delays for decontamination, and (3) increased shipping l n expenses. These hardships are not warranted since there is no )
Q compensating increase in the level of quality and safety. The as found test data is not affected and all applicable tests required by ANSI /ASME OM-1 are performed.
ALTERNATE TESTING: SRV pilot assemblies will be tested using a slave main valve body to comply with ANSI /ASME OM-1, Periodic Testing requirements.
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NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Valve Relief Reauests VRR-03
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SYSTEM: TRAVERSING IN-CORE PROBE (TIP)
VALVES: 07SOV-104A, B, C CATEGORY: A CLASS: 2 Augmented FUNCTION: These valves close to provide containment isolation.
TEST REQUIREMENT: OM-10, Section 4.2.1.4 - stroke time for power operated valves l
l BASIS FOR RELIEF: The computer control system for the TIP system includes a provision l for measuring valve cycle time (opened and closed) and not closure l time alone. The sequence opens the subject valve (stroke < 2 l seconds), maintains it energized for 10 seconds (including the cpening stroke), and de-energizes the valve solenoid allowing the valve to stroke closed (< 2 seconds). The total ahpsed time is specified to be 5; 12 seconds.
ALTERNATE TESTING: The overall cycle time (opened and closed) for these valves will be measured and evaluated in accordance with OM-10 Section 4.2.1.8.
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L NEW YORK POWER AUTHORITY-L JAMES A. FITZPATRICK NUCLEAR POWER PLANT
\p ~ INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES lQ APPENDIX B Valve Relief Reauests I
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' VRR-04 I SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)
VALVES: 23HPI-402,23HPI-403 1
CATE' GORY: C-CLASS: 2 Augmented FUNCTION: These valves open to eliminate any differential pressure that could force water from the suppression chamber into the HPCI exhaust piping when the suppression chamber pressure is greater than atmospheric. They close to prevent HPCI exhaust steam from entering j the' suppression chamber air space, thus bypassing the quenching action l
of the suppression pool.
TEST REQUIREMENT: OM-10, Section 4.3.2.2 - each check valve shall' be exercised or examined in a manner which verifies obturator travel to the closed, I full-open or partially open position required to fulfill its function.
BASIS FOR RELIEF: There are no position indicators on these valves or other means for verifying valve closure, thus the only practical means of verifying closure is to perform a back-leakage test. Since the valves are installed in series with no intermediate test tap, verifying the each individual valve closes is not practical.
To perform the specified safety function in the closed direction, only ,
one valve of the pair needs to close. Thus in accordance with l NUREG-1482 Section 4.1.1, verifying that either valve closes is adequate to demonstrate reliable operation of the pair.
I ALTERNATE TESTING: These valves will be exercised open and the pair (at least one valve) ;
will be verified to close during cold shutdown and each refueling outage in accordance with OM-10 Section 4.3.2.2(f) and (g). In l- accordance with NUREG-1482, if the closure test of the pair of valves i fails, then corrective action will be applied to both valves prior to returning the system to operability.
O Rev. No. _2_ Page 117 of 123
NEW YORK POWER AUTHORITY !
JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES ,
1 APPENDIX B l Valve Relief Requests VRR-06t'1, SYSTEM: SERVICE WATER / EMERGENCY SERVICE WATER COMPONENTS: 70TCV-120A,B, 70TCV-121 A,B, 67PCV-101 l
CATEGORY: B CLASS: 3 f
FUNCTION: The normal function of the temperature control valves 70TCV 120A & B and 121A & B are modulation to limit the flow of chilled water to maintain discharge air temperature and relative humidity to maintain a temperature of 75 degrees F in the Operations Office, Control Room, and Relay Room.
l j Moisture elements provide a control signal to keep the valves in the full
- open position when the relative humidity rises above 50%. The safety function of these valves is the same as above except that failure of the valve actuator mechanism results in valve movement to the maximum
- cooling water flow position (full open). Emergency Service Water (ESW) can also be circulated through the unit coolers if the chiller units become l inoperable.
l The normal function of valve 67PCV-101 is to maintain a backpressure at the common service water return header for the cable tunnel and electric bay coolers. The safety function of this valve is to fail open upon the loss of air.
TEST REQUIREMENT: OM-10, Section 4.2.1.4 - stroke time for power operated valves BASIS FOR RELIEF: These valves have no position indication or manual control switches.
Valve operation is controlled by temperature switches or pressure controllers. Stroke timing these valves would be extremely difficult and require an abnormal system configuration to obtain consistent stroke time results. Performing a stroke time test of these valves is impractical without a compensating level of quality and safety.
O Rev. No. 2 Page 118 of 123
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT
<w INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES i
ALTERNATE TESTING: In accordance with the guidance provided in NUREG-1482 adequate !
assessment of the operational readiness of these valves is achieved as I follows:
1 All valves are fail safe tested on a quarterly frequency. Prior to the test l
the valves are verified to not be in the full open position. During conduct of the test the valve air or electrical control is interrupted and the valve operation is observed locally to verify proper operation and movement to the fail safe full open position.
Valves 70TCV-121A,B are also stroked once per operating cycle per )
Technical Specification 4.11.B.2 during the calibration of their i associated instrumentation control loop.
Valves 70TCV-120A,B are also stroked once per operating cycle during the calibration of their associated instmmentation control loop. 1 I
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NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX C l
SUMMARY
OF CHANGES l
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I NEW YORK POWER AUTHORITY l JAMES A. FITZPATRICK NUCLEAR POWER PLANT
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! INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX C Pump Chances PAGE PUMP ID(s) CIIANGE REASON l 5 of 123 NA Deleted reference to code Editorial interpretations in paragraph 4.1 11 of 123 Note 1 Revised for clarification Editorial 11 of 123 ALL Added system, pump class NUREG 1482 and eliminated test type 12 of 123 10P-1A-D Corrected relief request Typos l re er n es 10P-3A-D l
14P-1A/B 11P-2A/B 46P-2A/B 15 of 123 PRR-02 R1 Revised Relief Request RAI dated April 30,1998 22 of 123 PRR-05 R1 Revised Relief Request RAI dated April 30,1998 25 of 123 PRRM) New Relief Request Address Water Level Measurement i
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NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX C Valve Chances
- PAGE VALVE ID(s) CHANGE REASON Valve Table All Added safety function Required by OM Code j Valve Table Various Corrected Dwg Coord Editorial 6 of 123 NA Deleted reference to do Editorial code interpretation in j paragraph 5.1 ;
29 of 123 NA Revise to Table of Reflect changes to ROJ &
Contents RR's 30 of 123 NA Changed valve category Editorial to IST category )
33 of 123 NA Changed RL to RV for Consistency relief valve designator 34 of 123 Test Frequency Added Test Frequency TS change 241 added SLC No.11 valve testing to IST Program 34 of 123 Test Added XVD test TS change 241 added SLC Requirements requirement to table valve testing to IST Program 36,37,38,57, 02RV-1 thru 11, Deleted Relief Valve Valves are not relief valves 65 of 123 02VB-1 thru 11, Test Requirement 23HPI-402 & 403, 13RCIC-37&38 _
37 of 123 02RV-71A thru Added STC-1 IST Requirement L
50 of 123 10RHR-52A Deleted from IST Appendix J Testing no rogram longer required 10RHR-52B 73 of 123 27VB-1 thru 5 Added Test ETC-1 & Show test requirements for M M E-1 both safety funtion 82 of 123 34FWS-28B Deleted FFT-2 Re-evaluation of safety function APPENDIX C 9
Rev. No. 2 Page 122 of121
i NEW YORK POWER AUTHORITY
! JAMES A. FITZPATRICK NUCLEAR POWER PLANT i i
INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES )
v Valve Chances *
(continued) l PAGE VALVE ID(s) CHANGE REASON 87 of 123 46SWS-911 & Deleted Valves Exempt per OM-10,1.1 916 116 &117 of 123 All Added augmented after 2 Editorial I
for Class l 116 of 123 27AOV101- Deleted Relief Request Revised Testing A&B "9"##*"""
VRR-05 27VB-6&7 118 of 123 66PCV-101, Revised VVR06R1 to Address RAI dated April delete system 66 valves 30,1998 and to remove 67TC%107C and expand the function, valves not required to be 66TCV-107F basis for relief, and tested l
^ # #8 *8
,p 70TCV-120A, B l 70TCV-101 Various Table All Relief Change all type RL To be consistent with valve Valves Valves to RV IST database
- changes to the valve tables are not indicated with a revision bar since the entire table was revised to evaluate the safety positions for valves.
O Rev. No. 2 Page 123 of 123