ML20217K584
| ML20217K584 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 09/30/1997 |
| From: | Boyer J, Michael Colomb, Edler F POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20217K581 | List: |
| References | |
| PROC-970930, NUDOCS 9710280326 | |
| Download: ML20217K584 (130) | |
Text
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Attachment I to JAFP 97.xxx JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES THIRD INTERVAL
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i REFSHE\T;;;g)gy JAhtES A, FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUhiPS AND VALVES TillRD INTERVAL Revision 1 Effective Date 9/2d973a g 4 W r' Prepared by: A e% Date: 9-13 O
@/ IS%cer Reviewed by: Date: ) IE Approved by: dE Date: 9. A 4*. t 7 F. EdlerY Tech Services Dept, hianager P
or eb'
- Date: 9 h /N7 hi, Colomb / SiteMixecuti.c Officer /
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Table of Contents 1.
I NTRO D U CTIO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
- 2. APPLICAB LE DOC UM ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
- 3. SYSTEM CLA SSI FICATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
- 4. INSERVICE TESTING PROGRAM FOR PUMPS . ........................... . .. .. .... . .. ..... 5
- 5. INSERVICE TESTING PROGRAM FOR VALVES . . ..... .. .................. .. . ..... . . ...... 6
- 6. SY STEM S S UBJ ECT TO TESTING. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
APPENDIX A - PUM P TESTIN G PROG RAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ,
APP EN DIX B - VALV E TESTING PROG RAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
APPEN DIX C - S UM M ARY OF Cll ANG ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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NEW YORK POWER AUTilORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES
1.0 INTRODUCTION
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Revision 1 of the James A. FitzPatrick ASME Inservice Testing (IST) Program will be in effect through the, end of the third interval unless cianged 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 documents:
- Title 10 Code of Federal Regulations, Part 50 i e Final Safety Analysis Report, J.A. FitzPatrick Nuclear Power Plant e J. A. FitzPatrick Technical Specifications
, o ASME Boller and Pressure Vessel Code,Section XI,1989 Edition 1
o 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:
1 e NRC Regulatory Guide 1.26, " Quality Group Classifications and Standards for l Water , Steam ,' and Radioactive Waste Contaminating Components of Nuclear Power Plants" e Standard Review Plan NUREG 0800, Section 3.9.6, " Inservice Testing of Pumps and Valves" Rev No. 1 Page .L of _120
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NEW YORK POWER AUTHORITY:
JAMES A. FITZPATRICK NUCLEAR POWER PLANT >
INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES i
e NRC Generic Letter 89-04, " Guidance on Developing Acceptable Inservice Testing Programs" i
e NRC Minutes of the Public Meetings on Generic letter 89-04 '
- - e NUREG 1482, " Guidelines for Inservice Testing at Nuclear Power Plants"
. e Safety Evaluation of Certain Relief Requests from Section XI of the American
) Society of Mechanical Engineers Code for the James A. Fitzpatrick Nuclear i i Power Plant, dated May 2,1991.
- 3.0 SYSTEM CLASSIFICATION i
In the NRC Safety Evaluation dated May 2,1991 for the J.A. FitzPatrick Section XI pressure test program, the NRC evaluated the deletion of certain Class II augmented I air / nitrogen systems from the inservice inspection program. These systems included the Drywell Inerting, CAD, and Purge system, the Containment Differential Pressurization L system, the Breathing, Instrument, and Service Air system, the Containment Hydrogen
- Monitoring system, and the Standby Gas Treatment system. The NRC's eva'uation
" 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 -I 1
classification of these systems would also be applicable to the IST program. Therefore, in accordance with NUREG 1482, components in these systems are not required to be in "
the IST program. They may be included in the IST program and designated as non-Code or augmented components.
implemented without NRC evaluation and approval.
Relief requests for non-Code . components mayJ be i
' = Containment isolation valves in the systems listed above have been included as Category 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 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 i- Diesel Generator subsystems as augmented Class III
li e Emergency Diesel Generator Fuel Oil Transfer i
e Emergency Diesel Generator Fuel Oil Service L
L Rev. No. 1 Page ._4 of.129 L
NEW YORK POWER AUTHORITY
- JAMES A. FITZPATRICK NUCLEAR POWER PLANT +
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INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES i-
! o Emergency Diesel Generator Combustion Air i
- i e Emergency Diesel Generator Cooling Water j e Emergency Diesel Generator Air Start -
These subsystems also meet the dermitions for skid mounted components and component
[ subassemblies as discussed in NUREG 1482. In NUREG 1482, the NRC has determined i that the testing of the major component is an acceptable means for- verifying the L- 1 operational readiness of the skid mounted and component subassemblies. This is i 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, preventativr. maintenance activities and design redundancy, the i components in the six Emergency Diesel Generator subsystems listed above, will not be included in the IST Program.
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~4,0 INSERVICE TESTING PROGRAM FOR PUMPS
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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 and any Code interpretations. 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 Rannes 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 Intervah
- The test frequency for pumps included in the IST Program will be as set forth in OM-6, Section 5.1. A band of 25 2 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 ficxibility.
- Rev No.- 1 Page.i_of129
NEW YORK POWER AUTIIORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES t 4.4 Pumo 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 Recuests for Pumo Testing Appendix A includes relief requests related to pump testing.
5.0 INSERVICE TESTING PROGRAM FOR VALVES 5.1 Code Comollance This IST Program is based on the requirements of OM 10 as referenced by Subsection IWV of the 1989 Code edition and any Code interpretations. W1cre 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 Testine Intervah The test frequency for valves included in the IST Program will be as set forth in OM 10, Section 4.2, 4.3, and 4.4. A band of 2 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. 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 Critetig 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 4 Generic I_etter 89-04.
Rev. No. 1 Page h of.120
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NEW YORK POWER AUTilORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES .
5.5 Conalnment Isolation Valves Containment isolation valves which do not provide a reactor coolant system pressure isolation function are tested in accordane with OM 10 Section 4.2.2.2.
In addition, as required by 10 CFR 50.55a(b)(2)(vil), 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 Program Tablg Appendix B lists those valves included in the IST Program with references to required testing, respective test intervals, applicable requests for relief and cold shutdown and refueling outage justifications.
- 5.7 Relief Requests for Valve Testine Appendix B includes relief requests, cold shutdown justifications. and refueling outage justifications related to valve testing.
6.0 SYSTEMS SUBJECT TO TEST.ltLQ 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 07 Neutron Tip Monitors FM 119A 10 ResidualIIeat Removal FM 20A,B 11 Standby Liquid Control FM 21 A 12 Reactor Water Cleanup FM 24A 13 Reactor Core Isolation Cooling FM-22A 14 Core Spray FM 23A 15 Rt actor Building Closed Loop Cooling FM 15A,B 16-1 leak Rate Analyzer FM-49A 19 Fuel Pool Cooling FM 19A 4
Rev. No. 1 Page _L of _120
NEW YORK POWER AUTilORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT I
INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES SYSTEM # SYSTEM NAME DRAWING #
20 Radioactive Waste FM 17A 23 liigh 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 Ventilt:lon FM 10ll (Service Water) 70 Control Room Service & Chilled Water FB 35E Rev. No. 1 Page .B_ of _120 m ---.____.-___._ ______
NEW YORK POWER AUTilORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIXA PUAfP 'ITSTING PROGRAM 5
Rev. No. 1 Page 1 of _qg
NEW YORK POWER AUTIIORITY JAMES A FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIXA PUMP TESTING PROGRAM Table of Contents
. Pump Table Explana t io n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1 PumpTable.....................................................................................................12 Re l le f Req uests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 PRR-01: Oe ne ric . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
PRR-02: S tandby Liquid Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 PRR-03: Stand by Liquid Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 PRR-04: Co r e S p ray . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 0 PRR-05: Emergency Service Water . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 I
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Rev.No. 1 Page 10 ofjlQ 1
NEW YORK POWER AUT110RITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIXA PUMP TABLE EXPLANATION Summary ofInformation Provided The Pump Table provides the following information:
Individual pump identifier Test type " Design" refers to tests where design or substantial flowrate is achieved.
The drawing on which the pump appears Drawing coordinates
- Speed"', if variable Differential pressurei "
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 note number indicating the parameter is measured but with some exception to the Code, or by a note number indicating relief is requested to eliminate measurement of the parameter. A blank indicates that measurement of the respective parameter is not applicable.
Pumn Reh gyqggg PRR-XX refer to relief requests for the Pump h. ; 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
- Proposed alternate testing Rev. No. 1 Page _1.L of _120
NEW YORK POWER AUTIlORITY l JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIXA gggm Relief Reauests PRR-01 SYSTEM: VARIOUS PUMPS: Various 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 in 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.
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 pressu;es are typically at relatively low levels.
Complisnce with this requirement could complicate venting procedures and introduce unnecessary health physics risks associated with handling and disposing of radioactive contaminate eater 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 OM-6 Section 4.6.
Rev. No. 1 Page_11_.of.R0
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NEW YORK POWER AUTilORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIXA Pumo Relief Reauests PRR-01 (Continued)
ALTERNATE TESTING: If the presence of absence of liquid in a gauge line used for sensing 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 ens are or determine the presence or absence of liquid as required for the static correction used.
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NEW YORK POWER AUTilORITY JAMES A. RTZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUhtPS AND VALVES APPENDIXA INnus Relief Reauests PRR-02 SYSTEM: STANDHY LIQUID CONTROL (SLC)
PUMPS: 11P 2A. B CLASS: 2 FUNCTION: These pumps inject borated water into the reactor vessel as at I alternate means for negative reactivity addition and reactor shutdown.
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.
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 from which water is being pumped.
ALTERNATE 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.
Rev,No, 1 Page _lL of.120
NEW YORK POWER AUTilORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSEkVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX A hp Relief Reauestg PRR-03 SYSTEM: STANDRY LIQUID CONTROL (SLC)
PUMPS: llP 2A, B CLASS: 2 FUNCTION: These pumps inject borated water into the reactor vessel as an alternate means for nw ' e reactivity additicn and teactor 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 minimum pump shaft rotational speed to at least 1000liz.
BASIS FOR RELIEF: The nominal speed of the SLC pumps is 520 RI'M, which correlates to a rotational frequency of 8.67 Ifz. 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 2 5% full scale over the range of 2.89 1000 liz.
The Authority has instmments for use during surveillance testing with certified accuracy of 2 5% full scale over a range of 5 2000 liz. Calibration is verified accurate using a system test methodology over a range of 10-1000 liz in units of displacement (m!!s p-p) and 6.5100011z la units of velocity (ips peak). The system test veri 0 cation 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.
Rev. No. 1 Page _15_.of.129
NEW YORK POWER AUT110RITY JA> !.S A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND YAINES APPENDIX A LNimLRelltf Reqqq.sn PRR 03 (Continued)
The instmment lower frequency response limits are a result of high pass nlters installed to climinate low frequency elements associated with the input signal from entering the procest of single and double integration. These Siters prevent low frequency electronic noise from distorting reading in the resultant cnits (ips, mils). As a side effect, any actual vibration occurring at low frequencies is filtered out. This is a necessary trade off, as 1 my of electronic noise at 2.5 llz translates to approximately 62.6 mils p-p with the accelerometer used with these instmments, at a nominal sensitivity of 50 mv/g.
The Authority has extensively researched this issue concerning Code compliance and intent, and strongly feels that, for these pumps, procurement of equipment capable of meeting the Code 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 instmmentation 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 llz.
In addition to the impracticality of procuring the instmments, the Authority feels that the instruments 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 readings will provide meaningful information in the detection of imminent machinery faults.
Rev. No. _L Page 17 ofILO
NEW YOR}i POWER AUTIIORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS ANRJALVFJ APPIMDIX A Pumo Relief Reauests PRR-03 (Continued)
A search of the INPO NPRDS database has revealed only one failure reported for pumps of this or similar design whose discovery mentioned increased vibration levels. The cited cause of the failure was improper end play set leading to gearing failure.
Failures of this type would normally be detected at running (shaft) speed frequency, harmonics thereof, or non-harmonic super-synchronous bearing defect frequencies. It should also be noted that these are standby pumps which are normally operated only during pump and valve testing. In the unlikely event this system is required to ful0ll 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, 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 certined accurate to , 5% over a frequency range of 5 2000 liz and is also limited by high-pass integrating filters, but allows for discrete frequency analysis and trending using FFI's. Vendor speci0 cations state that this equipment should provide fairly accurate data down to 2 liz in unit; of accelyation (g peak) by using the raw transducer signal. nyating the nemt for integration, Study of low frequency spect *.i takn in g p4 with these instruments has revealed ne distinct W.ynchronM A'iks 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.
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Rev. No. _L Page _1L ofILQ
NEW YORK POWER AUT110RITY JAMES A. FITZPAT!UCK NUCLEAR POWER PLANT INSERVICE TEST!NG PROGRAM FOR PUMPS AND VALVES APPENDIXA Pumo Relief Recuests PRR-03 (Continued)
In conclusion, the Authority feels '.it the use of high quality, commercially avaliable vibration monitoring equipment calibrated to be at least accurate to i 5% full scale over a range of 6 Ilz to 500 llz (nominal shaft speed . 8.67 hz) is an appropriate method of monitoring the mechanical condition of the SLC pumps. Such 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 arsi expense involved in the procurement, calibration, training and certification associated with obtaining new equipment which is simply not needed to adequately assess the condition of the SLC pumps.
ALTERNATE TESTING: The vibration measumnents will be taken using instrumentation accurate to 2 5% full scale over a frequency response range of 6 l{z to 500 liz. The data will be evaluated per OM-6 Section 6.
Rev. No. 1 Page _11.,of 120
NEW YORK POWER AUTilORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT t INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIXA Pumn Relief Reauests PRR44 4 SYSTEM: CORE SPRAY (CSP)
PUMPS: 14P 1 A, B CLASS: 2 FUNCTION: Pump cooling water from the suppression pool to the reactor in the event of a LOCA.
TEST REQUIREMENT: OM-6 Section 4.6.1.2(a), the ful' scale range of each analog instniment 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 suction pressure gauge is designed to provide adequate suction pressure indication during all expected operating conditions. The i
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 instatmentation loop are calibrated to within 2 2% full scale accuracy. The full scale range of the pump discharge pressure 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 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 i 10.3 psi or an 4
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 purposes.
Rev. No. 1 Page 20_ of _12.Q
1 NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALYES APPENDIXA Pumo Relief Recueitj PRR-041 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 6.25 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.
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NEW YORK POWER AUTilORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VAQ'ES i APPENDIXA Pumo Relief Reauests PRR-05 l
SYSTEM: EMERGENCY SERYlCE WATER (ESW) i PUMPS: 46P 2A, B i CLASS: 3 i
FUNCTION: These pumps provide cooling water for safety related heat loads j during a 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.
Ahernatively, the flow rate can be varied until the pressure equalt the reference value and the flow rate shall be determined and compared to the reference flow rate value.
BASIS FOR RELIEF: The Emergency Service Water pumps are vertical turbine type pumps which are submerged in and take suction from Lake Ontario. It is impra:tical to establish a single reference point as flow rate and differential pressure depend on multiple nonrepeatable parameters Lake level, strainer differential pressure, individual heat exchanger throttle valve positions, and system fouling levels all affect the point on the curve at which each pump operates at any single point in time. There is no overall system flow control available that would make it practical to establish a single repeatable reference point.
Compliance with this requirement is not practical. An alternate approach can be used which provides an equivalent means of monitoring the pumps for degradation.
Rev. No. 1 Page .22._ of.110
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NEW YORK POWER AUT110RITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT Ib' SERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIXA Mn Relief Reauests PRR 05 (Continued)
ALTERNATE TESTING: The alternate testing for 46P-2A is described as follows:
Data Validation The ESW Pump A maintenance history indicates that no pump maintenance has been performed other than a clearance adjustment in 1983. As a result, the original pump curve was analyzed which indicated a need to perform a speed correction to the present motor RPM. This was performed and the results were used to verify the pump mechanical condition. Data was collected in 1991 was compared to the speed corrected curve and the results indicated that the pump was perfomiing very near to the new speed corrected curve and was considered to be operating acc:ptably.
Methodoloey The methodology used to calculate the ESW Pump A acceptance criteria was to determine the slope between two points that are closest to the selected pump operating range and use this line as the ,
design line. Design points ar each end of the operating range are calculated by linear interpolation along the design line. Once the '
end points are known.. the code acceptance criteria lines are calculated on these end points and lines drawn to bound the i
acceptable and alert ranges. The OM-6 Table 3b limits for vertical line shaft pumps are used.
l During testing, the pump differential pressure (head) is calculated I- based on screenwell level and the pump discharge pressure. The pump flow is determined by using the mean ESW loop flow from a computer based trend. Using the Total Developed Pump Ilead and the mean flow, acceptable performance is verified by comparing these values to the design operating range as described above.
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l NEW YORK POWER AUTilORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIXA Pumo Relief Reauests PRR 05 (Continued)
The alternate testing for 46P 2B is described as follows:
Data Validation The ESW pump B was refurbished in 1979, Testing was performed by the vendor mirrored the installed conditions and no corrections were necessary. Data was collected in 1991 and compared to the vendor test data. . The results indicated that the pump was operating at approximately 2 feet above the pump curve.
The differr.nce was attributed to the higher flow instmmentation accuracy and a pump speed slightly higher than the 1979 vendor test.
Methodo!ony The methodology used to calculate the ESW pump B acceptance criteria was to determine the slope of the design line between the two points that are closest to the selected pump operating range and use this line as the design line, llowever, because the data collected in 1991 was greater than that on the design line, the test value was used ss the baseline. The acceptance criteria was determined by using the design line slope and the test values from 1991. Design points at each end of the operating range are calculated by linear interpolation along the design line, Once the end points are known, the code acceptance criteria lines are calculated on these end points and lines drawn to bound the acceptable and alert ranges. The OM-6 Table 3h limits for vertical line shaft pumps are used.
During testing, the pump differential pressure (head) is calculated -
based on screenwell level and the pump discharge pressure. The pump flow is determined by using the mean ESW loop flow from a computer based trend Using the Total Developed Pump Ilead and the mean flow, acceptable performance is verified by comparing these values to the design operating range as described above.
Rev. No. 1 Page 24 of_120
NEW YORK POWER AUTHORITY JAMES A FITZPATRICK NUCLEAR POWER PLANT INSERVLCE TESTING PROGRAM FOR PUMES AND VALVES APPENDIX B VALVE TESTING PROGRAM Rev. No. 1_
Page 25 of 120 l
NEW YORK POWER AUTHORI'lY
) JAMES A. FITZPATRICK NUCLEAR POWER ria 'IT INSERVICE TESTING PROGRAM FOR PUMPS Ahtd%LVES r
{
t APPENDIX B VALVE TESTING PROGRAM l
Table of Contents Valve Table Explanation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 8 V alv e Sy mbol s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1 Val ve Types . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1 Val ve Actua tor Types . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 Te st M e thod . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Test Requ irement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2 Test F reque ncy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
.........................32 Val v e Table . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Cold Shutdown Justification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
CSJ-01: Reactor Water Recirculation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 86 CSJ-02: Control Rod Drive Hydraulics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 86 CSJ.03: Residual Heat Remc a1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 87 CSJ-04: Residual Heat Removal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 8 8 CSJ-05: Reactor Co re Isolation Cooling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 88 CSJ-06: Reactor Core Isolation Cooling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 89 CSJ-07: Reactor Building Closed Loop Cooling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 89 CSJ-08: Reactor Building Closed Loop Cooling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 90 CSJ-09: . High Pressure Coolant injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 90 CSJ-10: High Pressure Coolant Injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 91 CSJ-11. High Pre ssure Coolant Injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 91 CSJ-12: Containment Atmosphere Dilution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 92 CSI-13: M a i n S team . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
CSJ 14:
M a in Steam . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
CSJ-15:
Feed wa te r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
CSJ-16: Containment Atmosphere Dilution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 94 Rev. No. 1 Page 26 of 120
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM COR PUMPS AND VALVES
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A PPENDIX B VALVE TFSUNG PROGRAM Table of Contents R e fueling Outage Justification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 95 ROJ-01: Generic - Excess Flow Check Valves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 95 ROJ-02: Reactor Water Recirculation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . , 96 ROJ-03: Reactor Water Recirculationn . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . %
ROJ-04: Au tomatic Depressurization. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 97 ROJ 05: Residual Heat Removal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 97 ROJ-06: Residual Heat Removal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 99 ROJ-07: Standby Liquid Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ,100 ROJ-08: Reactor Core Isolation Cooling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 101 ROJ-09: Co re Sp ray . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 102 ROJ-10: Co re Sp ra y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 103 ROJ-l1: Reactor Building Cooling Wate r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ,103 ROJ-12: H igh Pressure Coolant Injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 104 ROJ-13: H igh Pressure Coolant Injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 105 ROJ-14: H igh Pressure Coolant Injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 105 ROJ-15: H igh Pressure Coolant Injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 106 ROJ-16: H igh Pressure Coolant Injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 106 ROJ-17: H igh Pressure Coolant Injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 107 ROJ-18: High Pressure Coolant Injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 107 ROJ-19: M a i n S team . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
ROJ-20: Feed wa te r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 108 ROJ 21: I ns trume nt A ir . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 109 ROJ-22: Emergency Service Water . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 10 Rel ie f Requests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1 1 VRR-01: Automatic Depressurization/ Main Steam . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I i 1 VRR-02: Automatic Depressurization/ Main Steam . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 12 VRR-03: Traversing In-Core Probe . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 14 VRR-04: High Pressure Coolant Injection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 15 VRR-05: Containment Atmosphere Dilution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 16 VRR-06: Service Water / Emergency Service Water . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 17 Rev. No. 1 Page 27 of 120
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
- informelon
Individual valve identifier l Drawing coordinates Code Class Valve Category Nominal size Valve type Actuator type Test required Relief request (RR)/ cold shutdown (CS) justification / refueling outage (RO) justification Alternate test Remarks Rev. No. 1 Page 28 of 120
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NEW YORK POWER AITI110RITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B 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 Refueline Outane Justification ROJ XX refer to refueling outage justifications which provide the justification for testing affected components at refueling outages instead of every three months or at cold shutdown. The Refueling Outage Justifications provide the following information:
System Individual valve identifier Valve category Safety function Justification Rev. No. 1 Page 29 of_120
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Valve Relief Recuests 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 category Code Classification Safety Function-Code test requirement for which relief is requested
- Basis for relief Proposed alternate testing l
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Rev. No. 1 Page 30 of 120
NEW YORK POWER AUTilORITY l JAMES A. FITZPATRICK NUCLEAR POWER PLANT i
INSERVICE TESTING PROGRAM FOR PUMPS AND VALYSS APPENDIX B Valve Symbols Valve Tvoes 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 clak NK Non-return valve PG Plug valve RD Rupture disk RL Relief valve 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. 1 Page _11_ of 120
NEW YORK POWER AUTIIORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Test Method Test Reauirement 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 position 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 Test Freauency
-l 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 Rev. No. 1 Page 32 of 120
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NEW YORK POMR AUTHORITY JAMES A FIT 2PATRtCK NUCLEAR POWER PLANT lNSUMCE TESTING PROGRAM FOR PUMPS ANO VALVES VALVE TABLE SYSTEM Stans, Gas Tressment - SYSTEM D 01-125 ORAWING FMe DWG VALVE ' VALVE - ACTUATOR - TEST RELIEF ALTERNATE VALVEID CO-ORD CLASS CATEGORY SIZE tte TYPE TYPE RECTS CSDROJ REOUEST TEST REteMtm3 01-125MOV 100A C4 2A 8 4 00 SF MO STOwl AUMNTED STC1 PfT4
..01-125MOV-1006 F4 2A 8 4 00 er MO' STO 1 AUOaAENTED
' STC-1 PfT4 01-125MOV-11 G4 2A 8 24 00 . 8F MO STO'1 AUOMENTED pit 4 01-125MOV-12 F4 2A 8 24 00 SF MO STO-1 AUGMENTED P T-s 01 125MOV-14A 04 2A 8 24 d0 SF MO STO-1 AUGMENTED STC-1 Pff-5 01-125MOV-148 E4 2A 8 2400 SF RAO . STO 1 AUGMENTED STC-1 Pff4 01-125MOV-15A D-3 2A 8 24 00 - BF MO STO 1 AUOMENTED PtT-5 <
Di-i25MOv-iS8 r.3 2A 8 2400 Br MO STO-1 AUceAENTEo PfT4 {
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NEW YORK POMER AUT6fM
' JAMES A FIT 2 PATRICK NUCLEAR POWER PLANT WSERVICE TESTesG PROGRAAs FOR PUMPS AND VALVES VALVE TA8LE SYSTEM Ausomasc Ospressurirmamm Syseem . SYSTEM O 02 ORAngeG. FafL2tA DWG - VALVE VALVE ACTUATOR TEST RELIEF ALTERNATE VALVEID CO4RD CLASS CATEOORY SIZE (W) TYPE TM RECrTS CSJ4tOJ REQUEST TEST ret 44RKS 02AOV-11 47 1 8 1 00 GL AO Pf14 PASSNE D2AOV-18 G-7 1 8 1 00 CL AO PIT 4 PASSNE 02RV-1 H-T 2 C 3 00 CK SA ETO-1 ROJ4s ease 4
' ETC-1 hasE-3 RLF4 tedE4 02RV-2 R7 2 C 1 00 CK SA ETO-1 ROJ44 tesE4 ETC-1 tesE4 RLF4 ResE4 02RV4 47 2 C 3 00 CK SA ETG1 ROJos AmsE4 ETC-1 tesE4 RLF4 hasE-3 02RV4 47 2 C 3 00 CK SA ETO-1 ROJos tesE4 ETC-1 test 4 ALF4 teksE4 02RV4 HF 2 C 3 00 CK SA ETG1 ROJ44 tasE4 ETC.1 ansE4 RLF4 taasE4 02RV4 47 2 C 3 00 CK SA ETO-1 ROJ44 tesE4 ETC-1 tame 4 RLF4 teME-3 02RV-7 H7 2 C 3 00 CK SA ETG1 ROJ44 teME4 ETC-1 hasE-3 RLF4 tesE-3 02RV4 47 2 C 3 00 CK SA ETO-1 ROJos amE-3 ETC-1 ashsE-3 RLF4 hasE-3 02RV4 HF 2 C 3 00 CK SA ETOL1 ROJ44 hasE-3 ETC-1 tesE4 RLF4 sesE4 02RV-10 H7 2 C 3 00 CK SA ETO-1 ROJ44 masE-3 ETC-1 tesE4 RLF4 tesE4 REV NO . 1 PAGE 34 OF 120
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L NEW YORK POWER AUTHORITY
- JAAES A FITZPATRICK NUCLEAR POMER PLANT INSERVICE TESTseG PROORAM FOR PuedPS AND VALVES _
l VALVE TAetE SYSTEM Ausamese Depresounrenan Sysema SYSTEta D 02 ' DRMtENG FE20A DWWG VALVE ' VALVE ACTUATOR TEST RELEF ALTEftenATE i VALVE D COORD CLASS CATEGOprY SIZE (es) TYPE TYPE REQ'TS CSAROJ REOLEST TEST REtenftKS (
-.Gb ' hmE4 . ,
02RVJ1A G4 1 SWC 6 00 RL SA, AO STO WIWtet ETG3 RLF-7 WRR42 ETC4 6 4
02RVJ18 G4 1 BC 6 00 RL SA AO STO 1 vmR41 ETO-3 RLF-7 vpste ETC4 ' if I
d2RV-71C G4 1 SC 6 00 ftL SA AO ST41 VRR41 ET43 '
l ftLF-7 WRRM ETC-3 I
, t 02RVJ1D F4 '1 BC 6 00 RL '
RLFJ WRRM ETC-3 -!
02RVJ1E F-7 1 BC 6 00 RL SA AO STO-1 VRR41 ETO-3 RLF-7 VRRM ETC4 I
02RVJ1F FJ 1 BC 6 00 RL SA AO $701 WRR41 ET43 !
ALFJ WRRM ETC-3 i t
' 02RV-71G GJ 1 BC 6 00 ftL SA AO STO-1 - Wputet ' ETO4 ftLF-7 VfutM ETC4 ;
02RVJ1H G-7 1 BC 6 00 RL SA AO STG1 VRR41 ETO-3 i
i stLFJ VftRM ETC4 '
i 02RV-.'1J GF 1 BC 6 00 RL SA AO ST41 vRR41 ETO4 f
RLFJ VftRM ETC4 i 02RVJ1K G4 BC 6 00 RL SA AO Vfut41 i
1 STOwt ETO3 j, ItLFJ WWWtc ETC4
'i '
1 I O2RVJ1L G-T 1 SC 6 00 RL SA AO STO-1 VRft41 ETG3 [
RLFJ wputM ETC4
02VB-1 M 2 C 10 00 CK SA ETO-1 ftOJoe enAE4 ETC 1 tesE4 ;
ftLF4 taE4 !
i 02VB-2 He 2 C 10 00 CK SA E TO-1 ROJoe amE 3 5 ETC-1 taE4 -
[
RLF4 taE4 i t
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NEWYORK POWER AUTHORITY +
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WSERVICE TESTING PROGRAM Fort PUMPS AND VALVES VALVE TABLE SYSTEM . Reactor Waner Reara%-SYSTEM O 02-2 DRAWeeG FE2eA OWG VALVE VALVE ACTUATOR TEST MLEF ALTERNATE VALVE D C04RD CLAS4 CATEGORY SaZE (W) TYPE TYPE 'PECTS CSJROJ REQUEST TEST REesARKS 02-2EFV14JPT-1119 E4 1 NC 1 00 SK SA ETC-1 ROJ41 ETC4 VALWE ISOLATES ON EJtCESS FLOW LKGS LES3 02,2EFV1fT-1104 F-3 1 NC 1 00 SK SA ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EMCESS FLOW LKG5 LK43 02 2EFV1ET-110C D3 1 AtC 1 00 SK SA ETC1 ROJ41 ETC4 VALVE e90LATES ON EXCESS FLOW
, LKO4 LK43 02-2EFV1TT 110E F4 1 A/C 1 00 SK SA ETC-1 ROJ41 ETC-3 VALVE 8SOLATES ON EMCESS FLOW LKO4 LK43 02-2EFV1FT-110G D4 1 A/C 1 00 SK SA ETC-1 ROJ41 ETC4 VALVE RSOLATES ON EACESS FLOW LKO4 LKO4 02-2EFV24PT-111A E-3 1 AC 1 00 SK SA ETC-1 ROJ41 ETC4 VALVE ISOLATES OM ERCESS FLOW LKoS LKO-3 02-2E FV24PT-1118 E4 1 AtC 1 00 SK SA ETC-1 R(M41 ETC4 VALVE sSOLATES ON EACESS FLOW -
LKO-5 LKO-3 02 2EFV2fT-110A F-3 1 A/C 1 00 SK SA ETC-1 ROJ41 ETC4 VALVE ISOLATES O90 EXCESS FLOW LKO4 LK43 02-2EFV2fT 110C D-3 1 AtC 1 00 SK SA ETC-1 ROJ41 ETC-3 VALVE aSOLATES ON EMCESS FLOW LKOS LK43 02-2EFV2TT-110E F4 1 A/C 1 00 SK- SA ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKOL5 LKO-3 02-2EFV2fT-110G D4 1 AtC 1 00 SK SA ETC-1 ROJ 01 ETC4 VALVE ISOLATES ON ERCESS FLOW LKa$ LKG3 03sOV43A C-3 1 .9' 2e 00 GA mao STC-1 CSJ41 STC 2 PIT 4 02MOV 'A C4 1 9 ' 26 00 GA MO STC-1 CSJ41 STC4 PIT 4 REV NO : 1 PAGE 30 OF 120
NEW YORK POvuER AUTHORITY JAMES A FIT 2PATRK;K NUCLEAR POWtER PLAs(T StSERVICE TESTING PROGRAM FOR PUMPS AND VALWES VALVE TA8LE SYSTEM 96h Boder Vesosi kurumeras - SYSTEM D 02-3 n** m FM47A DWG VALVE VALVE ACTUATOR TEST RELEF - ALTERDMTE VALVE ID COORD CLASS CATECORY SIZE (IN) TYPE TYPE RECYTS CS mtOJ REQUEST TEST nFeenRKS 02-3E F V-11 F -7 1 NC 1 00 SK SA EiC-1 ROJ41 ETC4 . vmWE ISOLATES ON ERCESS FLOW LKO4 LKO4 02 3EFV-13A E-7 1- A/C 1 00 SK SA ETC-1 ROJ41 ETC-3 VALVE ISOLATES ON EACESS FLOW LKO'S LKO-3 02-3EFV-138 E4 1 AfC 1 00 SK SA ETC-1 ROJ41 ETC-3 VALWE IBOLATES sN EACES$ FLOW LKO4 LKO-3 02-3EFV-15A E-7 1 A#C 1 00 8K SA ETC1 ROJ41 ETC-3 VALVE ISOLATES ON ERCE.*S FLOW LKO-5 LKa3 02-3EFV *58 E4 1 A#C 1 00 8K SA ETC-1 ROJ41 ETC4 - VALVE ISOLATES OM ERCESS FLOW LKO4 LKS3 02-3EFV 15N S-7 1 AfC 1 00 SK SA- ETC-1 ROJ41 ETC4 VALVE ISOLATES ON ERCESS FLOW LKO4 LMO-3 02 3EFV-17A O-7 1 A/C 1 00 SK SA ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EMCESS FLOW LKO-5 LK43 02-3EFV-178 D4 1 A/C 1 00 8K - SA ' ETC-1 ROJ41 ETC-3 VALwE ISOLATES ON EXCESS FLOW LKO-S LKO4 02 3EFV-19A DF 1 NC 1 00 SK SA ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 02-3EFV-198 D-4 1 A/C 1 00 8K SA ETC-1 ROJ41 ETC4 VALVE ISOLATES ON ERCES$ FLOW LKO-5 LKO4 02-3EFV-21A H-5 1 A/C 1 00 OK SA ETC1 ROJ41 ETC4 VALwE ISOLATES ON EMCESS FLOW LKG5 LKO 3 02-XFV-218 C-7 1 AIC 1 00 8K SA ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EMCES$ FLOW LKO4 LKG3 02 3EFV 21C C4 . AfC 1 00 SK SA ETC 1 ROJ41 ETC-3 VALVE ISOLATES ON ERCESS FLOW LKSS LKO4 02 3EFV-210 H4 1 A/C 1 00 'OK SA ETC1 h* ETC4 VALVE ISOLATES ON EACESS FLOW LKO-5 LKO-3 02-3EFV-23 F-7 1 A/C 1 00 SK SA ETC-1 ROJ01 ETC4 . VALVE sSOLATES ON EACES$ FLOW LKO-5 LKO4 REV NO : t PAGE 30 OF 120
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SYSTEM Mm Boder Vessed bueuneres - SYSTEM O 024 ORAngeG FM47A ;
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LKOS LKO4 '
02-3EFV-230 C-7 1 AC 1 00 SK SA ETC-1 ROJ41 ETC4 VAtwE ISOt.ATES ON EXCESS FLOW LKO-6 LK43 02-3EFV-25 C-F . 1 AC ' 1 00 SK SA ETC-1 ROJ41 ETC4 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO4 t 02-3EFV-31A R$ 1 AC 1 00 SK S4 ETC-1 ROJ41 ETC4 VALWE ISOLATES ON EXCESS FLOW LKO4 LK43 l I
02-3EFV-319 HS 1 AC 1 00 SK SA ETC-1 ROJ41 ETC4 VALWE ISOtATES ON EXCESS FLOW l tKO-5 LKa3 02-3EFV41C HS 1 AC 1 00 SK SA' ETC-1 ROJ41 ' ETC4 VALWEISt 4TES ON EMCESS FLOW LK45 LKS3 02-3EFV41D R$ s AC 1 00 SK SA ETC1 ROJ41 ETC-3 vat * . sSOLATES ON EXCESS FLOW LKO4 LKO4 02-3EFV41E D-T 1 AC 1 00 SK SA ETC-1 ROJ41 ETC4 VAtwE tSOLATEC Oes EXCESS FLOW LKOS LKO4 02-3EFV41F H$ 1 AC 1 00 SK SA ETC-1 ROJ41 ETC4 VALVE ISOLATES ON ERCESS FLOW !
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VALVE D- 'C04RD CLASS CATEGORY SIZE (108) TYPE TYPE RE&TS CSJNtOJ NEOUEST TEST REtammstS 02-JEFv41L M4 1 NC 1 00 SK SA ETC1 ROJ4' ETC-3 tetLWE GBOLATES ON EJtCESS FLOW LMO4 LMO4 02-3EFV41M D4 1 WC 1 00 SK SA ETC1 ROJos ETC-3 VALVE ISOLATES ON EMCESS FLOW LNGS LMO-3 024EFV41N 04 4 1 WC 1 00 SK SA ETC.1 ter*M . ETC-3 VALVE ISOLATES 088 EJECESS FLOW LMOS LKO'3
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NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVFS 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 isoitte 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 Specilication 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 outage when Reactor Water Recirculation Pumps can be secured in accordance with OM-10 Section 4.2.1.2(0 and (g).
CS]-0?
SYSTEM: CONTROL ROD DRIVE HYDRAULICS (CRD)
COMPONENTS: 03HCU-l15 (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 wil: be tested during cold shutdown and each refueling outage in accordarm e with OM-10 Section 4.3.2.2(0 and (g).
Rev. No. _]_ Page 86 ofILQ
NEW YORK POWER AUTIIORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES 1
APPENDIX B Cold Shutdown Justifications CSJ-03 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 v'essel. 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 t'ese 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, these valves cannot be full'or part stroke exercised during normal plant operation.
Since there is no position indication for these valves, closure verification must.be done by backflow testing. Such testing during plant operation is impractical dae to personnel safety concerns related to the potential release of radioactive steam at high pressure.
These valves will be tested during cold shutdown and each refueling .
outage in acccrdance with OM 10 Section 4.3.2.2(f) and (g).
Rev. No. 1 Page 87 of_J20
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Cold Shutdown Justifications CSL-04 SYSTEM: RESIDUAL HEAT REMOVAL (RHR)
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.
4 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).
CSJ-05 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 -
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).
Rev. No. 1 Page 88 of _129 4
NEW YORK POWliR AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND._Y. ALVES APPENDIX B Cold Shutdown Justifications CSJ-06 SYSTEM:
REACTOR CORE ISOLATION COOLING (RCIC)
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 when the suppression chamber pressure is greater than atmospheric.
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 imprudent to test these valves while the plant is operational.
These valves will be tested during cold shutdown and each refueling outage in accordance with OM-10 Section 4.3.2.2.(0 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(0 and (g).
Rev. No. I Page 89 of 120
n NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT I
INSERVICE TESTING PROGRAM FOR PUhfPS AND VALVES APPENDIX B Cold Shutdown Justifications CS1-08 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(f) and (g).
CSJ-09 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)
COMPONENTS: 23HPI-13 CATEGORY: AIC SAFETY FUNCTION: This valve opens tr allow condensate drainage from the stern 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 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).
Rev.No. 1 Page 90 of 120 i
e NEW YORK POWER AUTIIORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX R Cold Shutdown Justif'ications CSJ-10 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)
COMPONENTS: 2311PI-18 CATEGORY: C SAFETY FUNCTION: This valve opens to provide a flowpath for the HPCI system injection to the reactor vessel.
JUSTIFICATION: With the reactor at operating pressure, the HPCI pump can develop sufficient 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 valvn cannot be exercised during normal plant operation.
This valve will oe tested during cold shutdown and each refueling outage in accordance with OM-10 Section 4.3.2.2(f) and (g).
CSJ-l1 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 steam from entering the suppression chamber air space, thus bypassing the quenching 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 considered prudent to perform this testing during plant operation at power.
Rev. No. 1 Page 91 of 120
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TFSTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Cold Shutdown Justifications CSJ-11 (Continued)
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 10 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).
Rev. No. _L. Page 92 of__120
a NEW YORK POWER AUTIlORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Cold Shutdown Justifications Ghl4 SYSTEM: MAIN STEAM (MSS)
COMPONENTS: 29AOV-203A, B CATEGORY: B 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 vfith OM-10 Section 4.2.1.2(0 and (g).
CS]-15 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 ie, dwater 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. 1 Page 93 of_120
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Cold Shutdown Justifications CS]-16 SYSTEM: CONTAINMENT VENT & OURGE (C/J)
COMPONENTS: 27MOV-120 CATEGORY: B SAFETY FUNCTION: This valve is closed to provide isolation for one path of containment 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).
N Rev. No. 1 Page 94 of 120
n NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT I
INSERVICE TESTING PROGRAM FOR PUMPS AND VALVFE APPENDIX B Refueline Outane Justifications ROJ-01 SYSTEM: VARIOUS COMPONENTS: Excess Flow Check Valves CATEGORY: A/C (l.isted 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 water effluent during the test.
These valves cannot be tested during cold shutdown since the reactor vessel is not pressurized.
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).
EXCESS FLOW CHECK VALVES 02-EFV-PS-128A,B 02-3EFV-19A,B 14EFV-31A,B 02-2EFV-PT-24A,B 02-3EFV-21A,B C.D 23EFV-01A,B 02-2EFV-PT-25A,B 02-3EFV-23A,B,C,D 23EFV-02A,B 02-2EFV1-DFI'-111 A,B 02-3EFV-23 29EFV-30A,B,C,D 02-2EFVI-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 02-3EFV-15A,B 02-3EFV-33 02-3EFV-15N 13EFV-01A,B 02-3EFV-17A,B 13EFV-02A,B Rev. No. 1 Page 95 of 120
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NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRMi FOR PUMPS AND VALVES P APPENDIX B Refueline outane Justifications ROJ-02 SYSTEM: REACTOR WATER RECIRCULATION (RWR)
COMPONENTS: 02 2RWR-13A, B CATEGORY: A/C SAFETY FUNCTION: These recirculation pump seal water injecion 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 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 subjects 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 4.3.2.2(e) and (h).
Rev. No. 1 Page 96 of_11Q
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM TOR PUMPS AND VALVES APPENNX B Refueline Outace Justifications ROJ-04 SYSTEM: AUTOMATIC DEPRESSURIZATION (ADS)
COMPONENTS: 0?RV-1 through 02RV-11 02VB-1 through 02VB-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 torus 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.
JUST1FICATION: 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.
There is no installed flow instrumentation in the minimum flow line thus attempts at flow measurements are being made with a strap on ultrasonic flow meters. Due to the minimum flow line configuration and operating conditions, there is a high amount of cavitation / turbulence in the line Rev. No. 1 Page 97 of 120
- g NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Refueline Outane 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 conditians. 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 continued efforts with this method are not practical.
Attempts were made to distmguish 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 instmmentation in the main line to infer the flow in the minimum flow line was investigated.
However, 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 894%, Position 2, du . ig each refuel outage at least one (1) valve will be disassembled, inspected, and verified operable. The acceptance criteria as stated in the Generic letter is provided in the maintenance procedure used for check valve disassemble.
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 the group are inspected at least once every six (6) years.
Rev. No. 1 Page 98 of 120
.- ._. = - -
NLW YORK POWER AUTilORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX R Refueline Outace Justincations ROJ-06 SYSTEM: RESIDUAL HEAT REMOVAL (RHR)
COMPONENTS: 10RIIR-95A,B CATEGORY: C SAFETY FUNCTION: These valves close to prevent reverse Dow 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.
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) .md 4
(h).
\
Rev. No. 1 Page 99 of 120
v NEW YORK POWER AUTilORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT i
INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Refueline Outace Justificatiom ROJ 07 SYSTEM:- STANDBY LIQUID CONTROL (SLC)
COMPONENTS: llSLC-16 & llSLC-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 110w to the reactor vessel, JUSTIFICATION: Full or partial-stroke exercising these valves requires that flow be estLblished 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 vessel resulting in unacceptable reactivity and chemistry transients. Testing during cold shutdown would result in chemistry transients and undue burden on the 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.
Rev. No. 1 Page 100 of 120
NEW YORK POWER AUTIIORITY JAMES A. FITZPnTRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Refuelinn Outace Justifications ROJ-08 SYSTEM: REACTOR CORE ISOLATION COOLING (RCIC)
COMPONENTS: 13RCIC-04 and 13RCIC-05 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 RCIC exhaust line must be isolated for the duration of the test causing the RCIC system to be inoperable.
The potential safety impact of voluntarily placing the RCIC 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 RCIC pump. This also is corsidered 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 each resueling outage in accordance with OM-10 Section 4.3.2.2(e) and (h).
Rev. No. 1 Page 101 of 120
NEW YORK POWER ALTTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT 1
-JNSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX R Refueline Outare Justifications ROJ-09 -
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 sufficient 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 overfillmg 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 v.lves will be exercised using the installed air operators (considered a partial-stroke).
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.
Rev. No. 1 Page 102 of 120
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- NEW YORK POWER AUTIIORITY -
JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX R 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 i valves. As a result, valve closure must be verified by back-leakage testing. Perfonning such a test during plant operations would require setting t.p for 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.
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.
ROI-l1 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 refueling outage to verify operability as an alternative to quarterly testing.
Rev. No. 1 Page 103 of 120
i NEW YORK POWER AUTI{ORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Refueline Outane Justifications ROJ 12 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)
COMPONENTS: 2311PI 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 stmetural 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).
Rev. No. 1 Page 104 of_1_;LQ
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT
-INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Refueline Outane Justifications ROJ-13 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)
COMPONENTS: 23HPI 13 and 23HPI-56 CATEGORY: C SAFETY FUNCTION: These valves opens to permit HPCI turbine 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, f
ROJ-14 l SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)
COMPONENTS: 23HPI-32 CATEGORY: C SAFETY FUNCTION: This valve closes during the suction swap from the Condensate Storage Tank to the torus to prevent diversion of the toms flow from the HPCI pump suction.
JUSTIFICATION: There is no provision on this valve that provides position indication of the 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 operability as an alternative to quarterly testing.
1 Rev. No. 1 Page 105 of 120
n NEW YORK POWER AUTilORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B \
Refueling Outane 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.
Rev. No. 1 Page 106 of 120
ur-NEW YORK POWER AUTHORITY
- JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B I
Refuelinn Outane Justifications ;
ROJ-17 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCD 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 additi;;- there are no test taps and block valves to enable a back-leakage test w 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)
J 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.
Rev. No. 1 Page 107 of_120
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAh! FOR PUMPS AND VALVF3 l
APPENDIX B Rsfueline Outane Justifications ROJ-19 SYSTEM: MAIN STEAh! (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 nuintained 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 vessel and extensive system preparations, including draining of the main feedweter piping from the outlet of the sixth point
~
feedwater heaters to the reactor vessel isolation valves (approximately 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. 1 Page 108 of 120
"l NEW YORK POWER AUTIIORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VAIATS
\ APPENDlX B l Refueline Outane Justifications ROJ 20 (Continued) 4 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 widi the containment inerted is a personnel safety concern.
4 Closure of these valves will be demonstrated during each refuel outage in accordance with OM 10 Section 4..s.2.2(e) and (h) by conducting a back-i leakage test.
L 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 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 irnpractical since closure of the MSIVs would cao: a plant trip. Also
- performing s"ch a test requires entry into the containment vessel and local 3 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).
Rev. No. 1 Page .l.00_ of 110
NEW YORK POWER AUT110RITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX R Refuelinn Outare Justifications ROJ 22 SYSTEM: EMERGENCY SERVICE WATER (ESW)
COMPONENTS: 46(70)ESW.101,102,103,104 CATEGORY: B SAFETY FUNCTION: These vahes 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 glycol / water mixture in the chilled water system. Opening these valves 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 drain the glycol from the system to prevent contamination. This would constitute an unreasonable burden on the plant staff.
These valves will be exercised open durir.g each refueling outtge in accordance with OM 10 Section 4.2.1.2(e) and (h).
i 1
1 Rev. No. 1 Page lifL of120
S NEW YORK POWER AUTIIORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B i
Valve Relief Reauests VRR-01 '
SYSTEM: AUTOMATIC DEPRESSURIZATION (ADS)/ MAIN STEAM COMPONENTS: 02RV-71 A,B,C,D,E,F,0,ll,J.K and L CATEGORY: B/C CLASS: 1 FUNCTION: These valves open when actuated by a manual switch to relieve reactor pressure during an accident or transient condition. Valves 02RV 71 A, B, C, D, E, G, ar.d H open on receipt of ADS actuation signal.
TEST REQUIREMENT: OM-10, Section 4.2.1,4 stroke time for power operated valves 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 opening 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 had 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 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.
Rev. No. 1 Page 111 of.120
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NEW YORK POWER AUTiiORITY 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 71 A,B,C,D,E,F,0,II,J.K and L i
CATEGORYt B/C CLASS: 1 FUNCTION: These valves oper 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 DASIS FOR RELIEF: Currently during refueling outages, the SRV pilot acsembly is removed and transported to a certified valve testing facility for performance of the following tests: setpoint (lift pressure), rescat (reclosing pressure),
and pilo; stage seat tightness. A main lxxly slave is used to test each pilot. ANSI /ASME OM 1 states, "No maintenance, adjusunent, disassembly, or other activity which could affNt as found set pressure or seat tightness data is pe,mitted 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 indicators; (h) determination of operation and electrical characteristics of bellows alarm switch; and (i) determination of actuating pressure of auxiliary actuating device sensing element, where applicable, and electrical continuity".
Rev. No. 1 Page 112 of_110
c.
1 NEW YORK POWER AUTIIORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX R 2
Valve Relief Reauests 9RR-02 (Continued)
' Strict adherence to the sequence cannot be satisfied by testing the pilot assembly only. Currently, the plant's test practices ensure that applicable tests specified in ANSI /ASME OM 1 Section 3.3.1.1, Main Steam Pressure Relief Valves with Auxiliary Actuating Devices, are perfonned and the entire valve operability is verified in accordance with Technical Specifications, but not in the sequence specified by OM 1 Section 3.3.1.1.
Common indusuy 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 outages for the removal and installation process, (2) cost increase and schedule delays for decontamination, and (3) increased shipping expenses. These hardships are not warranted since there is no 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.
Rev. No. I page .11)L of _{29
NEW YORK POWER AUT110RITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX R Valve Relief Reauests VRR-03 SYSTEM: TRAVERSING IN CORE PROBE (TIP)
VALVES: 07SOV 104A, B, C CATEGORY: A CLASS: 2 FUNCTION: These valves close to provide containment isolation.
TEST REQUIREMENT: OM-10, Section 4.2.1.4 stroke time for power operated valves BASIS FOR RELIEF: The computer control system for the TIP system includes a provision for measuring valve cycle time (opened and closed) and not closure time alone. The sequence opens the subject valve (stroke < 2 seconds), maintains it energized for 10 seconds (including the olvning stroke), and de-energizes the valve solenoid allowing the valve to stroke closed (< 2 seconds). The total clapsed time is specified to be s 12 seconds.
ALTERNATE TESTINO: 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.
Rev No. - 1 Page _Lli. of _120
m NEW YORK POWER AUTI{ORITY JAMES A. ITTZPATRICK NUCLEAR POWER PLANT l
INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES !
APPENDIX B ,
Valve Relief Reauests VRR-04 SYSTEM: IIIGH PRESSURE COOLANT INJECTION (ilPCI)
VALVES: 231{PI-402,2311PI-403 CATEGORY: C CLASS: 2 FUNCTION: These valves open to climinate any differential pressure that could force water from the suppression chamber into the IIPCI exhaust piping when the suppression chamber pressure is greater than atmospheric. They close to prevent IIPCI exhaust steam from entering the suppression chamber air space, thus bypassing the quenching action 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, 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 i one valve of the pair needs to close. Thus in accordance with NUREG 1482 Section 4.1.1, verifying that either valve closes is adequate to demonstrate reliable operation of the pair.
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(0 and (g). h accordance with NUREG-1482, if the closure test of the pair of valvea fails, then corrective action will be applied to betti valves prior to returning the system to operability.
Rev. No. 1 Page _111.of_120
NEW YORK POWER AUTIlORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Valve Relief Reauests i ,
vrses SYSTEM: CONTAINMENT ATMOSPHERE DILUTION (CAD)
VALVES: 27AOV 101 A,27AOV 101B,27VB-6, and 27VB 7 CATEGORY: A/C CLASS; 2 FUNCTION: These valves open to equalize pressure in the torus with pressure in the reactor building and close to provide containment isolation.
- TEST REQUIREMENT
- OM 1, Section 1.3.4.3 (b) - leak test every two years BASIS FOR RELIEF: The requirements for leak testing of containment isolation valves are '
covered in OM 10 Section 4.2.2.2. Compliance to OM 1 Section 1.3.4.3 (b) would treat these containment isolation valves differently than all other containment isolation valves. As stated in OM 10 Section 4.2.2.2, containment isolation valve leak testing shall be in accordance with 10 CFR 50 Appendix J.
ALTERNATE TESTING: Leak test the valves in accordance with OM 10 Section 4.2.2.2.
s
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Rev. No. 1 Page 116 of_120
NEW YORK POWER AUTIIORITY JAMES A. FIT 7 PATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Valve Relief Reauests VRR-06 SYSTEM: SERVICE WATER / EMERGENCY SERVICE WATER COMPONENTS: 66PCV-101,66TCV-107E,66TCV 107F,70TCV-120A,B, 70TCV-121A,B,67PCV 101 CATEGORY: B CLASS: 3 FUNCTION: These valves are the control valves for safety related ventilation coolers. They regulate the flow of service water (nonnal plant conditions) and the flow of emergency service water (accident conditions) to the East / West Crescent Area Unit Coolers, the East / West Cable Tunnel Cooling Coils, the Electric Bay Coolers, the Relay Room Air llandling Units, and the Control Room Air Handling Units.
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.
These valves are controlled by temperature switches or pressure controllers. It would be extremely difficult to obtain accurate stroke times and thus compliance with this requirement is impractical.
ALTERNATE TESTING: Adequate assessment of the operational readiness of these valves is achieved as follows:
Valves 66TCV-107E,F, and 70TCV-121A,B are stroked once per operating cycle per Technical Specification 4.11.B.2 during the calibration of their associated instmmentation conuol loop.
Valves 70TCV-120A,B are also stroked once per operating cycle during the calibration of their associated instmmentation control loop.
Operation of valves 66PCV-101 and 67PCV-101 is verified on a quarterly basis during the surveillance testing of the entire Emergency Service Water system.
Rev. No. 1 Page 117_ of-JLQ
NEW YORK POWER AUTIlORITY JAMES A. FIT 2 PATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX C i
SUMMARY
OF Cl{ANGES Rev. No. 1 Page _LL6_ of _]20
NEW YORK POWER AUTilORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT 3
INSERVICE TFSTING PROGRAM FOR PUMPS AND VALVES APPENDIX C Pumo Chances PAGE PUMP ID(s) CHANGE REASON _
l 1
0 H
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J Rev. No. I Page 119 oflLQ J
- a,-,,.,. , e .,,.a- - - - - . . - - . - -. - . - - - - , , , - .n , - , .- n.---. . . . ----e . .-- -,-- - - e
NEW YORK POWER AUTHORITY :
JAMES A, FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX C yalve Channu i
- PAGE VALVE ID(s) CHANGE REASON
! 70,94 27MOV 120 Added Cold Shutdown Valve can't be opened
! Justification, CSJ 16 when primary containment is required t
1 1
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9 a !
F 4
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4 4
4 i "
i Rev. No. ._1_ Page _120_ of _120 i
i SECOND INTERVAL TO TillRD INTERVAL COLD SilUTDOWN JUSTIFICATION, REFUEL OUTAGE JUSTIFICATION, AND RELIEF REQUEST VALVE CROSS REFERENCE SECOND TillRD COMMENT INTERVAL INTERVAL CSI CSJ 01 No change CS2 CSJ 03 No change CS3 CSJ 04 No change CS4 CSJ 08- No change CS$ CSJ-07 No change CS6 CSJ 10 No change CS7 CSJ-02 do Change CS8 -
This Cold Shutdown Justification was deleted in the second inten'al.
CS9 CSJ 15 No change CS10 CSJ 12 No change CS11 -
This Cold Shutdown Justification was deleted in the second interval.
CS12 CSJ-13 No change CS13 CSJ 14 No change CS14 -
This Cold Shutdown Justification was deleted in the second interval.
CSIS -
This Cold Shutdown Justification was deleted in the second inten al.
CS16 CSJ-09 No change CS17- CSJ-M No change CS18 CSJ 05 No change-CS19 -
This Cold Shutdown Justificatian was deleted in the second inten>al.
- ROJ1 ROJ 10 No change
_. 1
SECOND INTERVAL TO TillRD INTERVAL COLD SilUTDOWN JUSTIFICATION, REFUEL OUTAGE JUSTIFICATION, AND RELIEF REQUEST CROSS REFERENCE I
SECOND TillRD COMMENT INTERVAL INTERVAL ROJ2 ROJ-03 No change CSJ 16 This cold shutdown justification was initiated due to addition of new valve.
NOTE V1 ROJ-02 Changed Relief Request to Refueling Outage Justification in accordance with OM 10 Section 4.2.1.2 and 4.3.2.2 which allows deferral of testing until a refueling outage.
NOTE V2 -
No relief required since testing is in accordance with Generic Letter 89-N Position 7.
NOTE V3 - This Relief Request was deleted in the Second Interval.
NOTE V4 -
This Relief Request was deleted in the Second Interval.
NOTE V5 ROJ-07 Changed Relief Request to Refueling Outage Justification in accordance with OM-10 Section 4.2.1.2 and 4.3.2.2 which allows deferral of testing until a refueling outage.
NOTE V6 ROJ-08 Chaaged Relief Request to Refueling Outage Justification in accordance with OM-10 Section 4.2.1.2 and 4.3.2.2 which allows deferral of testing until a refueling outage.
NOTE V7 -
This Relief Request was deleted in the Second Interval.
NOTE V8 - This Relief Request was deleted in the Second Interval.
NOTE V9 ROJ-15 Changed Relief Request to Refueling Outage Justi fication in accordance with OM-10 Section 4.2.1.2 and 4.3.2.2 which allows deferral of testing until a refueling outage.
Also changed frequency of disassembly and inspection from once every 6 years to once per refueling in accordance with Generic Letter 89-N Position 2.
NOTE V10 -
This Relief Request was deleted in the Second Interval.
NOTE Vll - This Relief Request was deleted in the Second Interval.
NOTE _V12 ROJ-20 Changed Relief Request to Refueling Outage Justification in accordance with OM-10 Section 4.2.1.2 and 4.3.2.2 which allows deferral of testing until a refueling outage.
2
i SECOND INTERVAL TO TillRD INTERVAL COLD SilUTDOWN JUSTIFICATION, REFUEL OUTAGE JUSTIFICATION, AND RELIEF REQUEST CROSS REFERENCE SECOND TillRD COhiMENT
, INTERVAL INTERVAL NOTE V13 - This Relief Request was deleted in the Seebd Interval.
NOTE V14 ROJ 21 Changed Relief Request to Refueling Outage Justification in accordance with OM 10 Section 4.2.1.2 and 4.3.2.2 which allows deferral of testing until a refueling outage.
NUREG-1482 Section 3.1.1.3 states that valves may be tested during : fueling outages if they would otherwise tx, tested during cold shutdown outages that require the containment to be de inerted for performance of the testing. The NRC staff determined that maintaining a separate test schedule was not warranted.
NOTE V15 -
This Relief Request was deleted in the Second Interval.
4 NOTE V16 -
This Relief Request was deleted in the Second Interval.
NOTE V17 CSJ ll Changed the portion of the Relief Request dealing with VRR-04 testing interval to a Cold Shutdown Justification.
4 Alternate testing (testing a pair of valves together) is in accordance with NUREG-1482, Section 4.1.1 but relief is still required.
NOTE V18 -
This Relief Request was deleted in the Second Interval.
NOTE Vl9 -
This Relief Request is deleted. OM-10 Section 4.2.2.2 references 10 CFR 50 Appendix J for leak testing. OM-10 Section 4.2.2.3 allows for testing Category A valve in groups.
NOTE V20 -
This Relief Request was deleted in the Second Interval.
NOTE V21 - This Relief Request was deleted in the Second Interval.
NOTE V22 ROJ-13 Changed Relief Request to Refueling Outage Justification in accordance with OM-10 Section 4.2.1.2 and 4.3.2.2 which allows deferral of testing until a refueling outage.
NOTE V23 -
This Relief Request was deleted in the Second Interval.
NOTE V24 -
This Relief Request was deleted in the Second Interval.
NOTE V25 -
This Relief Request was deleted in the Second Interval.
3
SECOND INTERVAL TO THIRD INTERVAL COLD SilUTDOWN JUSTIFICATION, REFUEL OUTAGE JUSTIFICATION, AND RELIEF REQUEST CROSS REFERENCE SECOND TillRD COMMENT INTERVAL INTERVAL NOTE V26 --
This Relief Request was deleted in the Second Interval.
NOTE V27 -VRR 01 VRR-01 addresses stroke time testing and VRR-02 VRR-02 addresses relief valve testing NOTE V28 - ROJ-01 Changed Relief Request to Refueling Outage Justification in accordance with OM 10 Section 4.2.1.2 and 4.3.2.2 which allows deferral of testing until a refueling outage.-
NOTE V29 -
This Relief Request is d-leted. OM 10 Section 4.2.1.8(e) allows fast acting valves to be exempted from a a acceptance criteria if maximum limiting stroke is set at 2 seconds.
NOTE V30 -
This Relief Request was deleted in the Second Interval.
- NOTE V31 -
This Relief Reque". was deleted in the Second Interval.
NOTE V32 ROJ-05 Changed Relief Request to Refueling Outage Justification in accordance with OM 10 Section 4.2.1.2 and 4.3.2.2 which allows deferral of testing until a refueling outage.
Generic 1.etter 89-04 Position 2 allows vahe groupings of up to 4 valves with one valve disassembled and inspected each refuel outage.
NOTE V33 - -
This Relief Request was deleted in the Second Interval.
NOTE V34 ROJ 12 Changed Relief Request to Refueling Outage Justification in accordance with OM 10 Section 4.2.1.2 and 4.3.2.2 which allows deferral of testing until a refueling outage.
NOTE V35 ROJ 13 Changed Relief Request to Refueling Outage Justification -
in accordance with OM-10 Section 4.2.1.2 and 4.3.2.2 which allows deferral of testing until a refueling outage.
-NOTE V35 -
This Relief Request was deleted in the Second Interval.--
NOTE V37 -
This Relief Request was deleted in the Second Interval.
NOTE V38 -
This Relief Request was deleted in the Second Interval.
NOTE V39 -
This Relief Request was deleted in the Second Interval.
4 -
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j , !
j SECOND INTERVAL TO TillRD INTERVAL COLD SilUTDOWN JUSTIFICATION, REFUEL OUTAGE JUSTIFICATION, ;
AND RELIEF REQUEST CROSS REFERENCE ,
i l
l SECOND TillRD COMMENT j INTERVAL INTERVAL !
NOTE V40 - This Relief Request was deleted in the Second Interval.
NOTE V41 - This Relief Request was deleted in the Second Interval.
! NOTE V42 - This Relief Request was deleted in the Second Interval.
i NOTE V43 -
This Relief Request was deleted in the Second Interval.
l- NOTE V44' -
Th} Relief Request was deleted in the Second Interval.
NOTE V45 -- fnis Relief Request was deleted in the Second Interval.
! NOTE V46 -
This Relief Request is deleted. OM 10 does not require -
trending of containment isolation valve leakage rates for any size valve. ,
NOTE V47 - ROJ-16 Changed Relief Request to Refueling Outage Justification
- in accordance with OM-10 Section 4.2.1.2 and 4.3.2.2 1
which allows deferral of testing until a refueling outage, t !
. NOTE V48 - This Relief Request is deleted. OM-10 Section 4.2.2.3 allows for testing Category A valve in groups and does not require trending of Category A valve leakage rates.
I NOTE V49 ROJ 22 Changed Relief Request to Refueling Outage Justification in accordance with OM 10 Section 4.2.1.2 and 4.3.2.2 which allows deferral of testing until a refueling outage.
l NOTE V50 VRR-03 No change
! NOTE V51 - This Relief Request is deleted. This provision is a part of OM 10, Section 4.2.1.2(g) and 4.3.2.2(g).
e NOTE V52 - This Relief Request was deleted in the Second interval, p
i- NOTE V53 - This Relief Request was deleted in the Second Interval.
NOT8 V54 ROJ-17 Changed Relief Request to Refueling Outage Justification in accordance with OM-10 Section 4.2.1.2 and 4.3.2.2
- which allows deferral of testing until a refueling outage.
! . NOTE V55 -
This Relief Request was deleted in the Second Interval, i
i 5
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SECOND INTERVAL TO TillRD INTERVAL COLD SilUTDOWN JUSTIFICATION, REFUEL OUTAGE JUSTIFICATION, AND RELIEF REQUEST CROSS REFERENCE SECOND TillRD COMMENT INTERVAL INTERVAL NOTE V56 ROJ M Changed Relief Request to Refueling Outage Justification in accordance with OM 10 Section 4.2.1.2 and 4.3.2.2 ;
which allows deferral of testi,7 until a refueling outage.
NOTE V57 ROJ 09 Changed Relief Request to ketueling Outage Justincation ,
in accordance with OM-10 Section 4.2.1.2 and 4.3.2.2 which allows deferral of testing until a refueling outage.
NOTE V58 ROJ 04 Changed Relief Request to Refueling Outage Justification in accordance with OM 10 Section 4.2.1.2 and 4.3.2.2 which allows deferral of testing until a refueling outage.
NUREG-1482 Section 3.1.1.3 states that valves may be tested during refueling outages if they would otherwise be tested during cold shutdown outages that require the containment to be de-inerted for performance of the testing. The NRC staff determined that maintaining a separate test schedule was not warranted.
NOTE V59 ROJ 19 Changed Relief Request to Refueling Outage Justincation in accordance with OM 10 Section 4.2.1.2 and 4.3.2.2 which allows deferral of testing until a refueling outage.
NUREG 1482 Section 3.1.1.3 states that valves may be tested during refueling outages if they would otherwise be tested during cold shutdown outages that require the containment to be de-inerted for performance of the testing, The NRC staff determined that maintaining a separate test schedule was not warranted.
NOTE V60 CSJ-03 Changed Relief Request to Cold Shutdown Justification.
Valves can be exercised using two RIIR pumps in parallel in the shutdown cooling mode during cold shutdown.
- ROJ 11 This Refueling Outage Justification is being added for a new valve to the IST Program (15RBC-214) to cover disassembly and inspection on a refuel outage basis.
4 6
[
SECOND INTERVAL TO TillRD INTERVAL COLD SilUTDOWN JUSTIFICATION, REFUEL OUTAGE JUSTIFICATION, AND RELIEF REQUEST CROSS REFERENCE SECOND TillRD COMMENT INTERVAL INTERVAL ROJ14 While the forward flow verification for valve 2311PI 32 was deleted, a reverse tiow closure test was added. This Refueling Outage Justification is added to cover disassembly and inspection on a refuel outage basis.-
ROJ 18 This Refueling Outage Justification is being added for a new valve to the IST Program (2311PI 131) to cover disassembly and inspection on a refuel outage basis.
VRR 05 OM-1 requires that Primary Containment Vacuum Breakers be leak tested on a 2 year basis. This relief request proposes that they be leak tested on the same frequency as all other containment isolation valves.
VRR-06 NUREG 1482 Section 4.2.9 states that control valves with a fall safe function are required to be tested to meet all Code requirements for Category H valves which includes stroke time testing. This relief request proposes that other testing is adequate to monitor these valves for degradation.
7
SECOND INTERVAL TO TillRD INTERVAL PUMP RELIEF REQUEST CROSS REFERENCE SECOND TillRD COMMENT-INTERVAL INTERVAL NOTE P1 - This relief request was deleted in the second interval.
NOTE P2 - Relief request not required for third interval. OM-6 does not require measurement of inlet pressure.
NOTE P3 - This relief request was deleted in the second interval.
NOTE P4 -
This relief request was deleted in the second interval.-
NOTE PS -
Relief request not required for third interval. OM 6 does not require observation of lubrication level or pressure.
NOTE P6 - This relief request was deleted in the second interval.
NOTE P7 PRR-02 Portion of relief request related to duration of testing is no longer required. Relief for method of determining flow rate remains.
NOTE P8 -
This relief request was deleted in the second interval.
NOTE P9 -
Relief request not required for third interval. OM 6 allows vibration measurement on upper motor bearing for vertical line shaft pumps.
NOTE P10 -
This relief request was deleted in the second interval.
NOTE P11 PRR-04 No change.
NOTE P12 -
Relief request not required for third interval. OM 6 does not require measurement of inlet pressure, For positive displacement pumps, only need measurement of discharge pressure.
~ NOTE P13 PRR-03 No change.
NOTE P14 -
Relief request not required for third interval. OM-6 does not require measurement of bearing temperatures.
NOTE P15 -
This relief request was deleted in the second interval.
NOTE P16 PRR-01 No change.
NOTE Pl7 -
Relief request not required for third interval. OM-6 allows use of digital instmmentation.
8
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J
SECOND INTERVAL TO ' dlRD INTERVAL PUMP RFLIEF REQUES 'OSS REFERENCE SECOND TillRD COMMENT INTERVAL INTERVAL NOTE Pl7 -
Relief request not required for third interval. OM 6 allows use of digital instrumentation. .
1 NOTE P18 - This relief request was deleted in the second interval.
- PRR-05 Added relief request for the method of testing ESW pumps using a pump curve.
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