JAFP-07-0116, Fourth Interval Inservice Testing Program for Pumps and Valves

Fourth Interval Inservice Testing Program for Pumps and Valves

ML072750479
Person / Time
Site:	FitzPatrick
Issue date:	09/26/2007
From:	Jim Costedio Entergy Nuclear Northeast, Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
JAFP-07-0116 JAF-RPT-MULTI-03365, Rev 10
Download: ML072750479 (123)

Text

Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.

James A. Fitzpatrick NPP P.O. Box 110 Lycoming, NY 13093 Tel 315 342 3840 September 26, 2007 JAFP-07-0116 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555

Subject:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 License No. DPR-59 Fourth Interval Inservice Testing Program for Pumps and Valves

Dear Sir or Madam:

This letter submits the Fourth Interval Inservice Testing (IST) Program for Pumps and Valves, JAF-RPT-MULTI-03365 Revision 10, for the James A. FitzPatrick Nuclear Power Plant.

There are no new commitments contained in this letter.

Questions concerning this submittal may be addressed to Mr. Jim Costedio at (315) 349-6358.

Very truly yours, XinTmCostedio Licensing Manager JC:ed Enclosure AN-7

cc: Mr. Samuel Collins Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406-1415 Mr. John Boska I Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop O-8-C2 Washington D.C. 20555 Resident Inspector's Office U.S. Nuclear Regulatory Commission James A. FitzPatrick Nuclear Power Plant P.O. Box 136 Lycoming, New York 13093

Engineering Report No. JAF-RPT-MULTI-03365 Rev 10 Page 1 of 121 ENTERGY NUCLEAR go Entergy Engineering Report Cover Sheet Engineering Report

Title:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES FOURTH TEN YEAR INTERVAL Engineering Report Type:

New F] Revision 0 Cancelled n] Superseded El Applicable Site(s)

ANOI El IP2 El IP 3 El JAF 0 PNPS El vy DI wpo El AN02 El ECH I-1 GONS El RBS El WF3 F1 EC No. E*N/A: L-3053 Report Origin: Z Entergy [] Vendor Vendor Document No.:

Quality-Related: Z Yes F1 No Prepared by: John M. Scranto

• Date: 2/447 RespongbkSgineer (Print Name/Sign)

Design Verified/ N/A Date:

N/A Design Verifier (if requ ) (Print Name/Sign)

Reviewed by:

William C.Pelzer / Date:

T--7

• Reviewer(Print NamZSi **

Reviewed by*: N/A Date:

ANII (if reouired) (Print Name/Sign)

Approved by. Daniel Vandermark "7 &16 Supervisor (Print Name/Sign)  !

INSERVICE TESTING PROGRAM PLAN For The 4th Ten Year Interval Effective October 1, 2007 James A. FitzPatrick Nuclear Power Plant Commercial Service Date: October 17, 1974 Facility Name:

James A. FitzPatrick NPP P.O. Box 110 Lycoming, NY 13093 Owner:

Entergy Nuclear Operations, Inc.

P.O. Box 110 Lycoming, NY 13093

TABLE OF CONTENTS

1. IN T R OD U CT IO N ......................................................................................... .......................... 4

2. REGULATORY BASIS AND SCOPE ............................................ ................................ 4

3. APPLICABLE DOCUMENTS ......................................................................................... 6

4. SYSTEM CLASSIFICATION ...................................................................................... 6

5. INSERVICE TESTING PROGRAM FOR PUMPS ........................................................... 7

6. INSERVICE TESTING PROGRAM FOR VALVES ....................................................... 8

7. SYSTEMS SUBJECT TO TESTING ....................................................................................... 9

8. AUGMENTED TESTING .............................................................................................. 10

9. IST PROGRAM COMPONENT BASIS DOCUMENT ................................................. 10 APPENDIX A - PUMP TESTING PROGRAM ................................................................................ 11 APPENDIX B - VALVE TESTING PROGRAM ................................. 27

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES

1.0 INTRODUCTION

1.1 General This document outlines Fourth Ten-Year Interval Program Plan for In-service Testing (IST) of Pumps and Valves at the James A. FitzPatrick (JAF) Nuclear Power Plant. This Program Plan was prepared in accordance with the rules of the ASME Code for Operation and Maintenance of Nuclear Power Plants, ASME OM Code-2001, through the ASME OMb Code-2003 Addenda (OM-2001 through Omb-2003 - "The Code").

1.2 Commercial Operation Date and IST Intervals The James A. FitzPatrick (JAF) Nuclear Power Plant began commercial operation on October 17, 1974, and the First Ten-Year ISI/IST Interval began on that date. The Third Interval start date was extended from October, 1996 to October, 1997 (JPN-94-037, JPN-96-019, and JPN-97-015). This extension used the allowable 12 months total extension of intervals permitted by the Code.

The Third-Interval IST Program was applicable for the interval from October 1, 1997 through September 30, 2007. Therefore, the 4th 10-Year IST Interval is applicable from October. 1, 2007 through September 30, 2017.

1.3 Applicable Codes The Fourth 10-Year Interval Program Plan complies with the OM-2001 Edition through the OMb-2003 Addenda, which was incorporated by reference into 10CFR50.55a via Federal Register/Vol 69, No. 190 on October 1, 2004. The Third 10-Year Interval IST Program Plan complied with ASME Section XIL 1989 Edition.

2.0 REGULATORY BASIS AND SCOPE 2.1 10CFR50 The fundamental requirement for the testing of pumps and valves comes from 10CFR50.55a(f),

which requires, in part, that:

"Throughout the service life of a boiling or pressurized water cooled nuclear power facility, pumps and valves which are classified as ASME Code Class .1, Class 2, and Class 3 must meet the inservice test requirements.. .set forth in Section XI of Editions of the ASME Boiler and Pressure Vessel Code and Addenda that... are incorporated by reference in paragraph (b) of this section..."

Pump and valve inservice testing is also required by 10CFR50 Appendix A, "General Design Criteria. For Nuclear Power Plants," GDC 1; and 10CFR50, Appendix B, "Quality Assurance Criteria For Nuclear Power Plants And Fuel Reprocessing Plants," Criterion XI.

Rev. No. 10 Page 4 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES Appendix A GDC 1 states in part, "Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed."

Appendix B Criterion XI, "Test Control," states in part, "A test program shall be established to assure that all testing required to demonstrate that structures, system, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. The test program shall include, as appropriate ... operational tests during nuclear power plant operation of structures, system, and components. Test procedures shall include provisions for assuring that all prerequisites for the given test have been met, that adequate test instrumentation is available and used, and that the test is performed under suitable environmental conditions."

2.2 ASME Boiler and Pressure Vessel Code The specific regulatory basis for the IST program is iOCFR50.55a(f), "In service Testing Requirements." This section of 10CFR50 requires the following:

The testing performed during the second (and successive) 120-month interval'must comply with the requirements of the Code Edition incorporated by 10CFR50.55a(b) 12 months prior to the start of the interval.

For FitzPatrick, the Fourth 120-month interval began on October 1, 2007. Therefore, the Code Edition of interest is the one endorsed by NRC in 10CFR50.55a via Federal Register/Vol 69, No. 190 on October 1, 2004. The Code Edition in effect on October 1, 2006 was the OM 2001 Edition through OMb 2003.

2.3 OM 2001 and OMb 2003 Addenda The organization of the new Code is significantly different than the Code used for the Third Interval IST Program Plan. This Program Plan is written to conform generally to the outline structure of the new Code. The new Code contains the following major sections:

• ISTA: General Requirements

• ISTB: In service Testing of Pumps

• ISTC: In service Testing of Valves

• Appendix 1: In service Testing of Pressure Relief Devices Rev. No. 10 Page 5 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES 3.0 APPLICABLE DOCUMENTS This IST Program was developed. in accordance with. the requirements of the following documents:

• Title 10, Code of Federal Regulations, Part 50

• Final Safety. Analysis Report, J.A. FitzPatrick Nuclear Power Plant

• J.A. FitzPatrick Technical Specifications Other documents used for guidance in the development of the IST Program are listed below:

• Standard Review Plan NUREG 0800, Section 3.9.6, "Inservice Testing of Pumps and Valves"

• NRC Generic Letter 89-04, "Guidance on Developing Acceptable Inservice Testing Programs"'

• NUREG-*1482 Revision 1, "Guidelines for Inservice Testing at Nuclear Power Plants" 4.0 SYSTEM CLASSIFICATION 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 11-augmented air/nitrogen systems from the inservice inspection program. These systems included the Drywell Inerting, CAD, and Purge system, the Containment Differential Pressurization system, the Breathing, Instrument, and Service Air system, the Containment Hydrogen Monitoring system, and the Standby Gas Treatment system. The NRC's evaluation found, based on a review of the regulations, the ASME Code, and regulatory guides, that there is no basis for requiring inservice inspection of these particular systems.

Although this finding related only to the hydrostatic testing of these systems, the basis for classification of these systems would also be applicable to the IST program. Therefore, 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 (see section 8.0 Augmented Testing). Relief requests for non-Code components may be implemented without NRC evaluation and approval.

Containment isolation valves in the systems listed above have been included as Category 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.

Rev. No. 10 Page 6 of 121

• JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES 4.1 IST Technical Position Papers Certain systems and components have been evaluated for inclusion in the IST Program based

• on the above and other industry guidance. IST Program Technical Position Papers have been prepared to document this evaluation to ensure a thorough thought process was applied and documented. The following Position Papers are references to the JAF IST Program:

IST-02-01 RHIRSW MOV Testing (10MOV-89A/B, 148A/B, 149A/B)

IST-02-02 Reactor Core Isolation Cooling System Testing IST-02-03 Emergency Diesel Generator System Testing 5.0 INSERVICE TESTING PROGRAM FOR PUMPS 5.1 Code Compliance This IST Program is based on the requirements of ASME OM Code-2001, through the ASME OMb Code-2003 Addenda. Where these requirements have been determined to be impractical, conformance would cause unreasonable hardship without any compensating increase in safety, or an alternative test provides an acceptable level of quality and safety, relief from Code requirements is requested pursuant to the requirements of 10 CFR 50. 55a (f)(6)(i).

5.2 Allowable Ranges of Test Quantities The allowable ranges for test parameters as specified in ASME OM Code-2001, through the ASME OMb Code-2003 Addenda will be used for all measurements of pressure, flow, and vibration except as provided for in specific relief requests.

5.3 Testing Intervals The test frequency for pumps included in the IST Program will be as set forth in ASME OM Code-2001, through the ASME OMb Code-2003 Addenda. A band of +/- 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.

5.4 Pump Program Table Appendix A lists those pumps included in the IST Program with references to parameters to be measured and applicable requests for relief.

5.5 Relief Requests for Pump Testing Appendix A includes relief requests related to pump testing.

Rev.. No. 10 Page 7 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES 6.0 INSERVICE TESTING PROGRAM FOR VALVES 6.1 Code Compliance This IST Program is based on the requirements of ASME OM Code-2001, through the ASME OMb Code-2003 Addenda. 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).

6.2 Testing Intervals The test frequency for valves included in the IST Program will be as set forth in ASME OM Code-2001, through the ASME OMb Code-2003 Addenda. A band of +/- 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 ASME OM Code-2001, through the ASME OMb Code-2003 Addenda.

6.3 Stroke Time Acceptance Criteria The acceptance criteria for the stroke times of power-actuated valves will be as set forth in ASME OM Code-2001, through the ASME OMb Code-2003 Addenda..

6.4 Check Valve Testing 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 from this requirement must satisfy the requirements of NUREG-1482 revision 1.

6.4.1 Non-intrusive Check Valve Testing and Condition Monitoring The use of non-intrusive check valve testing methods, such as Acoustics, Ultrasonics and Eddy Current are employed at JAF. These check valve testing methodologies will be utilized to supplement existing forward and reverse flow check valve testing and enhance the ability to monitor and predict check valveperformance.

6.5 Containment Isolation Valves Containment isolation valves that do not provide a reactor coolant system pressure isolation function are tested in accordance with ASME OM Code-2001, through the ASME OMb Code-2003 Addenda. In addition, as required by 10 CFR 50.55a(b)(2)(vii), containment isolation valves are analyzed in accordance with ASME OM Code-2001, through the ASME OMb Code-2003 Addenda and corrective action is applied in accordance with ASME Rev. No. 10 Page 8 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES OM Code-2001,. through the ASME OMb Code-2003 Addenda.

6.6 Valve Program Table Appendix B lists those valves included in the IST Program with references to required testing, respective test intervals, applicable requests for relief and cold shutdown and refueling outage justifications.

6.7 Relief Requests for Valve Testing Appendix. B includes relief requests, cold shutdown justifications, and refueling outage justifications related to valve testing.

7.0 SYSTEMS SUBJECT TO TESTING SYSTEM # SYSTEM NAME DRAWING #

01-125 Standby Gas Treatment FM-48A 02 Automatic Depressurization FM-29A 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-I 19A 10 Residual Heat Removal FM-20A,B 11 Standby Liquid Control FM-21A 12 Reactor Water Cleanup FM-24A 13 Reactor Core Isolation Cooling FM-22A 14 Core Spray FM-23A 15 Reactor Building Closed Loop Cooling FM-15A,B 16-1 Leak Rate Analyzer . FM-49A 20 Radioactive Waste FM-17A 23 High Pressure Cooling Injection FM-25A 27 Containment Atmosphere Dilution FM-18A,B,D 29 Main Steam FM-29A Rev. No. 10 Page 9 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES E SYSTEM # SYSTEM NAME DRAWING #

34 Feedwater FM-34A 39 Breathing, Instrument & Service Air FM-39A 46 Service & Emergency Service Water FM-46A,B 66 Reactor Building Service Ventilation FB-10H (Service Water) 70 Control Room Service & Chilled Water FB-35E 8.0 AUGMENTED TESTING Pumps and valves that are not within the designated James A. Fitzpatrick Code Class 1, 2, or 3 boundaries are not under the jurisdiction of ASME OM Code-200 1, through the ASME OMb Code-2003 Addenda and are considered augmented components. Additionally, there may be components within the Code Class 1, 2, or 3 boundaries with safety functions outside the licensing basis of the plant. These components may have safety functions that are classified as augmented. Such components or safety functions are considered augmented in the IST program plan and associated testing may not necessarily meet all requirements established in ASME OM Code-2001, through the ASME OMb Code-2003 Addenda. Relief requests, cold shutdown justifications, and refueling outage justifications for augmented components are provided for information only and do not necessarily require approval.

9.0 IST PROGRAM COMPONENT BASIS DOCUMENT - JAF-RPT-MULTI-04406 The IST Program Component Basis Document provides the basis for component inclusion and exclusion. This is a living document and may be revised and/or have pending changes exclusive of the IST Program Plan. Administrative control for the Basis Document will be documented in AP-19.05.

Rev. No. 10 Page 10 of 121

JAMES A. FHTZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX A PUMP TESTING PROGRAM Table of Contents Pump T able E xplanation ............................................................... .................................................. 12 P um p T ab le ....................................................................................................................................... 13 R elief R equests ....... .................................................. ........................................................................ 14 PRR-01: Standby Liquid Control ........ .............................................. 14 PR R-02: C ore Spray ................................................................................................. 16 PRR-03: Emergency Service Water ................................. 18 PRR-04: RHR Service Water and Emergency Service Water ................................ 21 Rev. No. 10 Page 11 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX A PUMP TABLE EXPLANATION Summary. of Information Provided The Pump Table provides the following information:

" System

" Individual pump identifier 0 Class 9 Group 0 The drawing on which the pump appears

• Drawing coordinates

• Speed), if variable

• Differential pressure0')

• Discharge pressure0') (positive displacement pumps)

• Flow rate()*

0 Vibration(l) e Test interval

• Relief Request e Test Procedure (1) These parameters are each addressed with either an "X" indicating the parameter is measured, an "X" with a PRR notation indicates relief is requested to modify or eliminate measurement of the parameter. A blank indicates that measurement of the respective parameter. is not applicable.

Pump Relief Requests PRR-XX refers to relief requests for the Pump Testing Program. Each pump request for relief provides the following information:

• System

• Individual pump identifier Code Classification Safety Function Code test requirement for which relief is requested

• Basis for relief Proposed alternate testing Rev. No. 10 Page 12 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 13 of 121 INSERVICE TESTING PROGRAM PUMP TABLE DWG DIFF DISCH FLOW INPSECTION TEST SYSTEM NOMENCLATURE PUMP ID CLASS GROUP DWG No. CO-ORD SPEED PRESS PRESS RATE VIBE FREQUENCY RELIEF REQUEST PROC RHR Service Water 10P-1A 3 A FM-20B B-6 X - X X Quarterly PRR-04 ST-2XA X X X 2YR Comp Pump Test PRR-04 RHR Service Water lOP-lB 3 A FM-20B B-5 X X X Quarterly PRR-04 ST-2XA X . _ X X 2YR Comp Pump Test PRR-04 RHR Service Water lOP-iC 3 A FM-20B C-6 X X X Quarterly PRR-04 ST-2XA X X X 2YR.Comp Pump Test PRR-04 -

RHR Service Water 10P-iD 3 A FM-20B C-5 X X X Quarterly PRR-04 ST-2XA X X X 2YR Comp Pump Test PRR-04 Residual Heat Removal 1OP-3A 2. A . FM-20A C-7

• X X X Quarterly ST-2AL LPCI Pump X X X 2YR Comp Pump Test Residual Heat Removal 1OP-3B 2 A FM-20A C-4 X X X Quarterly ST-2AL LPCI Pump X X X 2YR Comp Pump Test Residual Heat Removal 1OP-3C 2 A FM-20A C-7 X X X Quarterly ST-2AM LPCI Pump X X X 2YR Comp Pump Test Residual Heat Removal 1OP-3D 2 A FM-20A C-4 X X X Quarterly ST-2AM LPCI Pump X X X 2YR Comp Pump Test Standby Liquid Control Pump 11P-2A 2 B FM-21A D-4 X Quarterly ST-6HA X X X 2YR Comp Pump Test PRR-01 Standby Liquid Control Pump 11 P-2B 2 B FM-21A B-4 X Quarterly ST-6HB X X X 2YR Comp Pump Test PRR-O1 Core Spray Pump .14P-1A 2 B FM-23A C-8 ._X X Quarterly PRR-02 ST-3PA X X X 2YR Comp Pump Test Core Spray Pump 14P-1B 2 B FM-23A C-3 X X Quarterly PRR-02 ST-3PB X X X 2YR Comp Pump Test High Pressure Coolant 23P-1B 2 B FM-25A E-5 X X Quarterly ST-4N Injecetion Booster Pump X X X _2YR Comp Pump Test High Pressure Coolant 23P-1M 2 B FM-25A E-4 X X X Quarterly ST-4N Injecetion Main Pump X X X X 2YR Comp Pump Test Emergency Service 46P-2A 3 B FM-46B D-8 X X Quarterly PRR-03 PRR-04 ST-80 Water Pump X X X 2YR Comp Pump Test PRR-03 PRR-04 Emergency Service 46P-2B 3 B FM-46B C-8 X X Quarterly PRR-03 PRR-04 ST-8O Water Pump X X X 2YR Comp Pump Test PRR-03 PRR-04

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX A Pump Relief Requests PRR-01 System:

-STANDBY LIQUID CONTROL (SLC)

ASME Code Components Affected:

11P-2A, B Component/System Function:

These pumps inject borated water into the reactor vessel as an alternate means for negative reactivity addition and reactor shutdown.

Applicable Code Edition and Addenda

ASME OM Code-2001 including 2003 Addenda OM Code Category:

Group B

Applicable Code Requirement

ISTB-3500, "Data Collection", 3510, "General", 3510(e), "Frequency Response", 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 1000 Hz.

Reason for Request

The nominal speed of the SLC pumps is 520 RPM, which correlates to a rotational frequency of 8.67 Hz. Table ISTB-3510-1, "Required Instrument Accuracy", requires the frequency response range of the vibration measuring transducers and their readout system to be accurate to +/- 5% full scale over the range of 2.89 - 1000 Hz.

FitzPatrick Nuclear Station has instruments for use during surveillance testing with certified accuracy of +/- 5%. full scale over a range of 5-2000 Hz. Calibration is verified accurate using a system test methodology over a range of 10-1000 Hz in units of displacement (mils p-p) and 6.5-1000 Hz in units of velocity (ips peak). The system test verification is limited by the capability of the calibration shaker system to accurately sustain vibration at meaningful amplitudes outside the.

tested frequencies. The certified calibration +/- 5% range is arrived at through addition of individual transducer and meter inaccuracies over the stated frequency range.

The instrument lower frequency response limits are a result of high-pass filters installed to eliminate low frequency elements associated with the input signal from entering the process of single and Rev. No. 10 Page 14 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES double integration. These filters prevent low frequency electronic noise from distorting reading in the resultant units (ips, mils). As a side effect, anyactual vibration occurring at low frequencies is filtered out. This is a necessary trade-off, as 1 mv of electronic noise at 2.5 Hz translates to approximately 62.6 mils p-p with the accelerometer used with these instruments, at a nominal sensitivity of 50 mv/g.

FitzPatrick Nuclear Station has extensively researched this issue concerning Code compliance and intent, and 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, proneto human error, costly to purchase and expensive to calibrate. The number of vendors that supply instrumentation accurate at these frequencies is limited, and there are even fewer vendors capable of performing the required calibration services. Most standard qualified calibration laboratories provide calibration services only to a minimum of 10 Hz.

Proposed Alternative and Basis for Use:

In addition to the impracticality of procuring the instruments, FitzPatrick Nuclear Station 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.

A search of the EPIX (formerly 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 endplay 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. that are normally operated only during pump and valve testing. In the unlikely event this system is required to fulfill its design function, only one of the two redundant pumps need operate for a period of 23 to 125 minutes.

In addition to vibration monitoring performed for the IST Program, these pumps are included in the FitzPatrick Nuclear Station Rotating Equipment Monitoring Program. Vibration spectral data is periodically collected and analyzed for the pump and gear motors in addition to those required by the Code. The equipment used by the Rotating Equipment Program is certified accurate to +/- 5%

over a frequency range of 5-2000 Hz and is also limited by high-pass integrating filters, but allows for discrete frequency analysis and trending using FFTs (Fast Fourier Transforms). Vendor specifications state that this equipment should provide fairly accurate data down to 2 Hz in units of acceleration (g peak) by using the raw transducer signal, negating the need for integration. Study of low frequency spectra taken in g peak with these instruments has revealed no distinct sub-synchronous peaks above the noise floor acceleration signal.

Rev. No. 10 Page 15 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES 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.

Proposed Alternative Testing:

The vibration measurements will be taken using instrumentation accurate to +/- 5% full scale over a frequency response range of 6.5 Hz to 1000 Hz. The data will be evaluated in accordance with ISTB-6000, "Monitoring, Evaluation, and Analysis".

Duration of Proposed Alternative:

The proposed alternative identified in this IOCFR50.55a Request shall be utilized during the Fourth Ten Year IST Interval.

Precedents:

This 10CFR50.55a Request was previously approved for the Interval 3 IST Program in NRC SER dated November 17, 1998 (TAC No. MA0096). The circumstances and basis for the previous NRC approval have not changed.

References:

None Rev. No. 10 Page 16 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX A Pump Relief Requests PRR-02 System:

CORE SPRAY (CSP)

ASME Code Components Affected:

14P-1A, B Component/System Function:

Pump cooling water from the suppression pool to the reactor in the event of a LOCA.

Applicable Code Edition and Addenda

ASME OM Code-2001 including 2003 Addenda OM Code Category:

Group B

Applicable Code Requirement

ISTB-3500, "Data Collection", 3510, "General", 3510(b), "Range", the full scale-range of each analog instrument shall be not greater than three times the reference value.

Reason for Request

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 full-scale range, 60 psig,.is sufficient for a post-accident condition when the torus is at the maximum accident pressure. This, however, exceeds the range limit for the suction pressure under the test condition (approximately 5 psig).

The installed suction pressure gauge and discharge pressure instrumentation loop are calibrated to within +/- 2% full scale accuracy. The full-scale range of the pump discharge pressure 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 +Iý 1.2 psi and in measured discharge pressure is +/- 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 +/- 10.3 psi or an inaccuracy of 3.5%.

Rev. No. 10 Page 17 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES Proposed Alternative and Basis for Use:

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.

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 +/- 18.3 psig or an*

inaccuracy of 6.2%. The existing measurement is significantly better than the maximum Code allowable inaccuracy.

Proposed Alternative Testing:

The existing installed plant suction pressure gauges will be used to determine the pump differential pressure for testing of the Core Spray pumps.

Duration of Proposed Alternative:

The proposed alternative identified in this 10CFR50.55a Request shall be utilized during the Fourth Ten Year IST Interval.

Precedents:

This 10CFR50.55a Request was previously approved for the Interval 3 IST Program in NRC SER dated November 17, 1998 (TAC No. MA0096). The circumstances and basis for the previous NRC approval have not changed.

References:

None Rev. No. 10 Page 18 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX A Pump Relief Requests PRR-03 System:

EMERGENCY SERVICE WATER (ESW)

ASME Code Components Affected:

46P-2A, B Component/System Function:

These pumps provide cooling water for safety-related heat loads during a loss-of-coolant. design basis accident.

Applicable Code Edition and Addenda

ASME OM Code-2001 including 2003 Addenda OM Code Category:

Group B

Applicable Code Requirement

ISTA-3130, "Application of Codes Cases", ISTA-3130(b) states, "Code Cases shall be applicable to the edition and addenda specified in the test plan."

ISTB-5222(b), "The differential pressure or flow rate shall then be determined and compared to its reference value."

ISTB-5222(c), "System resistance may be varied as. necessary to achieve the reference point."

ISTB- 5223(b), "The resistance of the system shall be Varied until the flow rate equals the reference point."

Reason for Request

Emergency Service Water (ESW) systems are designed such that the total pump flow cannot be adjusted to one finite value for the purpose of testing without adversely affecting the system flow balance and Technical Specification operability requirements. These pumps must be tested in a manner that the service water loop remains properly flow balanced during and after the testing and each supplied load remains fully operable per Technical Specifications to maintain the required level of plant safety during plant operation.

Rev. No. 10 Page 19 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES The ESW water system loops are not designed with a full flow test line with a single throttle valve.

The flow therefore cannot be throttled to a fixed reference value every time. Total pump flow rate can only be measured using the total system flow indication installed on the common supply header.

Only the flows of the serviced components can be individually throttled. Each load is throttled to a FSAR required flow range which must be satisfied for the load to be operable. All loads are aligned in parallel, and all receive ESW flow when the associated ESW pump is running, regardless whether the served component is in service or not.

During power operation, all loops of ESW are required to be operable per the Technical Specifications. A loop of ESW cannot be taken out of service for testing without entering a Limiting Condition for Operation (LCO) Action Statement. With each loop of ESW balanced a requirement to quarterly adjust ESW loop flow to one specific flow value for inservice testing conflicts with system design and operability requirements(i.e. flow balance) as required by Technical Specifications.

It is extremely difficult or impossible to return to a specific flow rate or differential pressure for testing these pumps. Multiple reference points could be established according to the Code, but it would be impossible to obtain reference values at every possible point. An alternative to the testing requirements of ISTB is to base the acceptance criteria on a reference curve.

ISTA-3130, "Application of Codes Cases", ISTA-3130(b) states, Code Cases shall be applicable to the edition and addenda specified in the test plan.

NUREG-1482, Revision 1, Section 5.2 states "ASME introduced Code Case OMN-9, "Use of Pump Curves for Testing.... which the NRC staff subsequently included in RG 1.192. NUREG 1482, Section 4.2.5 further states; "The use of OMN-9 requires relief because OMN-9 is only applicable to the ASME OM Code 1990 through ASME OMb Code 1992. Licensees with a Code of record that is not applicable to the acceptance of this Code Case may submit a request for relief to apply the Code Case consistent with the indicated conditions to provide an acceptable level of quality and safety. The Code of record for FitzPatrick Nuclear Station's Fourth 10-Year IST Interval is ASME OM Code-2001 Edition w/2003 Addenda. Code Case OMN-9, as stated in RG 1.192, is applicable to the ASME OM Code 1990 Edition through OMb Code 1992 Addenda.

Proposed Alternative and Basis for Use:

Flow rate and Total Developed Pump Head (in accordance with NUREG-1482 section 5.5.3) will be measured during inservice testing in the as-found condition and compared to an established reference curve developed in accordance with Code Case OMN-9 and the additional conditions as prescribed in RG 1. 192 FitzPatrick Nuclear Station requests approval to use the guidelines set forth in Code Case OMN-9, "Use of a Pump Curve for Testing," including the associated conditions prescribed in RG 1.192, Operation and Maintenance Code Case Acceptability, ASME OM Code, in lieu of the ASME OM Code paragraphs ISTB-5222 and ISTB-5223 requirements for ESW pumps 46P-2A and 46P-2B Code Case OMN-9 should be considered acceptable for use with OM Code-2001 Edition w/2003 Rev. No. 10 Page 20 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES Addenda as the Code of record. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), FitzPatrick Nuclear Station requests relief from the specific ISTB Code requirements identified in this relief request Proposed Alternative Testing Flow rate and Total Developed Pump Head will be measured during inservice testing in the as-found condition and compared to an established reference curve developed in accordance with Code Case OMN-9 and the additional conditions as prescribed in RG 1.192.

Duration of Proposed Alternative:

The proposed alternative identified in this 10CFR50.55a Request shall be utilized during the Fourth Ten Year IST Interval.

Precedents:

This 10CFR50.55a Request was previously approved for the Interval 3 IST Program in NRC SER dated November 17, 1998 (TAC No. MA0096). The circumstances and basis for the previous NRC approval have not changed.

References:

Code Case OMN-9 , "Use of a Pump Curve for Testing" Regulatory Guide 1.192, "Operation and Maintenance Code Case Acceptability, ASME OM Code",

Table 1, "Acceptable OM Code Cases" OM Code-2001 w/ 2003 Addenda, Paragraph ISTA-3130, "Application of Code Cases" Rev. No. 10 Page 21 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX A Pump Relief Requests PRR-04 System:

RHR Service Water (System 010)

Emergency Service Water (System 046)

ASMIE Code Components Affected:

Smooth Running Pumps in the IST Program 10P-1A 1OP-IB lOP-iC 10P-1D 46P-2A 46P-2B Component/System Function:

Provide cooling water to RHR Heat Exchangers during a design basis event Provide cooling water to Emergency Diesel Generators and Essential Cooling Water loads

Applicable Code Edition and Addenda

ASME OM Code-2001 including 2003 Addenda OM Code Category:

Group A and Group B

Applicable Code Requirement

ISTB-3300, "Reference Values", ISTB-3300(a), states-, Initial reference values shall be determined from the results of testing meeting the requirements of ISTB-3100, "Preservice Testing", or from the results of the first inservice test.

ISTB-3100(d), states; Reference values shall be established at a point(s) of operation (reference point) readily duplicated during subsequent tests.

ISTB-3300(f), states; All subsequent test results shall be compared to these initial reference values or to new reference values established in accordance with ISTB-33 10, ISTB-3320, or ISTB-6200(c).

Rev. No. 10 Page 22 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES ISTB-5120, "Inservice Testing", ISTB-5121, "Group A Test Procedure", ISTB-5121(e) and ISTB-5123(e), "Group A Test and Comprehensive Test Procedure", states; All deviations from the reference values shall be compared with the ranges of Table ISTB-5121-1 and corrective action taken as specified in ISTB-6200. Vibration measurements shall be compared to both the relative and absolute criteria shown in the alert and required action ranges of Table-ISTB-5121-1. For*

example, if vibration exceeds either 6Vr, or 0.7 in./sec, the pump is in the required action range.

ISTB-5220, "Inservice Testing", ISTB-5221, "Group A Test Procedure", ISTB-5221(e) and ISTB-5223(e), "Group A Test and ComprehensiveTest Procedure", states; All deviations from the reference values shall be compared with the ranges of Table ISTB-5221-1 and corrective action taken as specified in ISTB-6200. Vibration measurements shall be compared to both the relative.

and absolute criteria shown in the alert and required action ranges of Table-ISTB-5221-1. For example, if vibration exceeds either 6Vr, or 0.7 in./sec, the pump is in the required action range.

ISTB-5320, "Inservice Testing", ISTB-5321, "Group A Test Procedure", ISTB-5321(e) and ISTB-5323(e), "Group A Test and Comprehensive Test Procedure", states; All deviations from the reference values shall be compared with the ranges of Table ISTB-5321-1 or 5321-.2, as applicable and corrective action taken as specified in ISTB-6200. Vibration measurements shall be compared to both the relative and absolute criteria shown in the alert and required action ranges of Table-ISTB-5321-1 or 5321-2 as applicable. For example, if vibration exceeds either 6Vr, or 0.7 in./sec, the pump is in the required action range.

Reason for Request

The smooth running pumps in the FitzPatrick Nuclear Station IST Program have at least one vibration reference value (Vr) that is currently less than 0.05 in/sec. A small value for Vr produces a small acceptable range for pump operation. The OM Code Acceptable Range limit for pump vibrations from Table ISTB-5121-1, Table ISTB-5221-1, Table ISTB-5321-1 and Table ISTB-5321-2 for both the Group A test and Comprehensive test is <=2.5 Vr. Based on a small acceptable range, a smooth running pump could be subject to unnecessary corrective action if it exceeds this limit.

ISTB-6200(a), "Corrective Action - Alert Range", states; If the measured test parameter values fall within the alert range of Table ISTB-5121-1, Table ISTB-5221-1, Table ISTB-5321-1 or Table ISTB-5321-2, as applicable, the frequency of testing specified in ISTB-3400 shall be doubled until the cause of the deviation is determined and the condition is corrected.

For very small reference values for vibrations, flow variations, hydraulic noise and instrument error can be a significant portion of the reading and affect the repeatability of subsequent measurements.

Also, experience gathered by the FitzPatrick Nuclear Station Predictive Maintenance (PdM) Group has shown that changes in vibration levels in the range of 0.05 in/sec do not normally indicate significant degradation in pump performance.

In order to avoid unnecessary corrective actions, a minimum value for Vrof 0.05 in/sec is proposed.

This minimum value would be applied to individual vibration locations for those pumps with reference vibration values less than 0.05 in/sec.

Rev. No. 10 Page 23 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES Therefore, the smallest OM Code Acceptable Range limit for any IST pump vibration location would be no lower than 2.5 times Vr, or 0.125 in/sec, which is within the "fair" range of the "General Machinery Vibration Severity Chart" provided by IRD Mechanalysis, Inc. Likewise, the smallest OM Code Alert Range limit for any IST Pump vibration location for which the pump would be inoperable would be no lower than 6 times. Vr, or 0.300 in/sec.

For comparison purposes, ASME XI, Table IWP-3100-2, "Allowable Ranges of Test Quantities"',

specifies a vibration Acceptable Range limit of 1.0 mil for a displacement reference value <=0.5 mils. In velocity units, a displacement reference value of 0.5 mils is equivalent to 0.047 in/sec for an 1800 rpm pump and 0.094 in/sec for a 3600 rpm pump.The effective minimum reference value.

proposed (0.05 in/sec) for smooth-running pumps is roughly equal to the ASME XI IWP reference value for an 1800 rpm pump and more conservative than the reference value for a 3600 rpm. pump.

Without this relief, the Acceptable Range limit for some extremely smooth running pumps is reduced by as much as a factor of 10..

In addition to the requirements of ISTB for IST, the pumps in the FitzPatrick Nuclear Station IST Program are also included in the FitzPatrick Nuclear Station PdM Program. The FitzPatrick Nuclear Station PdM Program currently employs predictive monitoring techniques suchas:

vibration monitoring and analysis beyond that required by ISTB,* bearing temperature trending, oil sampling and analysis, and/or thermography analysis as applicable.

If the measured parameters are outside the normal operating range or are determined by analysis to be trending toward an unacceptable degraded state, appropriate actions are taken that may include:.

a Condition Report (CR) initiated, increased monitoring to establish a rate of change, review of component specific information to identify cause, and removal of the pump from service to perform maintenance.

It should be noted that the pumps in the IST Program will remain in the FitzPatrick Nuclear Station PdM Program even if certain pumps have very low vibration readings and are considered to be smooth running pumps.

Proposed Alternative and Basis for Use:

In lieu of applying the vibration acceptance criteria ranges specified in Table ISTB-5121-1, Table ISTB-5221-1, Table ISTB-5321-1 or Table ISTB-5321-2, as applicable, smooth running pumps with a measured reference value below 0.05 in/sec for a particular vibration measure location will have subsequent test results for that location compared to an Acceptable Range limit of 0.125 in/sec and an Alert Range limit of 0.300 in/sec (based on a minimum reference value 0.05 in/sec). These proposed ranges shall be applied to vibration test results during both Group A tests and Comprehensive tests.

In addition to the Code requirements, pumps in the FitzPatrick Nuclear Station IST Program are included in and will remain in the FitzPatrick Nuclear Station PdM Program regardless of their smooth running status.

Rev. No. 10 Page 24 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES Using the provisions of this 10CFR50.55a Request as an alternative to the specific requirements of ISTB identified above will provide adequate indication of pump performance and continue to provide an acceptable level of quality and safety without unnecessarily imposing corrective action since changes in vibration levels in the range of 0.05 in/sec do not normally indicate significant degradation in pump performance.

Proposed Alternative Testing:

Smooth running pumps with a measured referencevalue below 0.05 in/sec for a particular vibration measure location will have subsequent test results for that location compared to an Acceptable Range limit of 0.125 in/sec and an Alert Range limit of 0.300 in/sec (based on.a minimum reference value 0.05 in/sec). These proposed ranges shall be applied to vibration test results during both Group A tests and Comprehensive tests.

Using the provisions of this relief request as an alternative to the vibration acceptance criteria ranges specified in Table ISTB-5121-1, Table ISTB-5221-1, Table ISTB-5321-1 or Table ISTB-5321-2 provides an acceptable level of quality and safety since the alternative provides reasonable assurance of pump operational readiness and the ability to detect pump degradation. Therefore, pursuant to 10 CFR 50.55a(f)(6)(i), FitzPatrick requests relief from the specific ISTB Code requirements identified in this 10CFR50.55a Request.

Duration of Proposed Alternative:

The proposed alternative identified in this 10CFR50.55a Request shall be utilized during theFourth Ten Year IST Interval.

Precedents:

None at FitzPatrick Nuclear Station.

For similar relief, refer to Beaver Valley Power Station, Unit 2, Docket No.50-412, SER Date December 27, 2004, Evaluation of Inservice Testing Pump Relief Request PRR-8, (TAC No.

MC3241)

For an additional similar relief request, refer to Diablo Canyon Power Plant, Unit Nos. 1 and 2 -

Approval of Relief Requests P-RR 1, P-RR2, and P-RR3 for the Third 10-year Pump and Valve Inservice Testing Program Interval (TAC Nos. MC6632 AND MC6633) SER Dated January 30, 2006.

References:

NUREG-1482, Rev. 1, Section 5.4, "Monitoring Pump Vibration in Accordance with ISTB" General Machinery Vibration Severity Chart provided by IRD Mechanalysis, Inc.

Beaver Valley Power Station, Unit 2, Docket No.50-412, SER Date December 27, 2004, Evaluation of Inservice Testing Pump Relief Request PRR-8, (TAC No. MC3241 Rev. No. 10 Page 25 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES Diablo Canyon Power Plant, Unit Nos. 1 and 2 - Approval of Relief Requests P-RR1, P-RR2, and P-RR3 for the Third 10-year Pump and Valve Inservice Testing Program Interval (TAC Nos.

MC6632 AND MC6633) SER Dated January 30, 2006.

Rev. No. 10 Page 26 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B VALVE TESTING PROGRAM Rev. No. 10 Page 27 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B VALVE TESTING PROGRAM Table of Contents.

V alve Table Explanation ...................................................................................... . .................. 30 V alve S ym b ols ...................................................................................................................................... 33 V alve T ypes ............................................................................................................................... 33 V alve Actuator T ypes ...................................................................................................... 33 T est M eth o d .......................................................................................................................................... 34 Test Requirem ent ................................................................................................................. 34 T est Frequency ..................................................................................................................... 34 V alve Tab le ....................................................................................................................................... 35 Cold Shutdown Justifications ............................................... 87 CSJ-01: Reactor Water Recirculation .................................................................... 87 CSJ-02: Reactor Core Isolation Cooling ............................................................... 87 CSJ-03: High Pressure Coolant Injection ........,.............................. 88 CSJ-04: Containment Vent and Purge ................... ..................... 88 CSJ-05: Main Steam .......................................... 89 C SJ-06: M ain Steam .............................................................................................. 89.

CSJ-07: Reactor Water Cleanup ................................... 90 CSJ-08: Main Steam ............................................................. 90 CSJ-09: Reactor Core Isolation Cooling ............................................................. 91 Rev. No.. 10 Page 28 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B VALVE TESTING PROGRAM Table of Contents Refueling Outage Justifications ..................................................................................................... 92 ROJ-01: Generic - Excess Flow Check Valves ....... ........................... 92 ROJ-02: Reactor Water Recirculation ............................... 93 ROJ-03: Reactor Water Recirculation ............................... 93 ROJ-04: Automatic Depressurization ..................................... 94 ROJ-05: Residual Heat Rem oval ............................................................................ 94 ROJ-06: Residual Heat Removal ................................... 96 ROJ-07: Standby Liquid Control ........................................................................... 96 ROJ-08: Reactor Core Isolation Cooling .............................................................. 97 ROJ-09: Core Spray ......................................................................... 98 ROJ-10: Core Spray ......... .................................. 99 ROJ- 11: Reactor Building Cooling Water ............................................................... 99 ROJ-12: High Pressure Coolant Injection ............................ 100 ROJ-13: High Pressure Coolant Injection ................................................................ 101 ROJ-14: High Pressure Coolant Injection ..................................................................... 101 ROJ-15: High Pressure Coolant Injection .................................................................. 102 ROJ-16: High Pressure Coolant Injection .................................................................. 102 ROJ-17: High Pressure Coolant Injection .............................................................. 103 ROJ-18: High Pressure Coolant Injection ................................................................ 103 ROJ-19: Main Steam .... ................................. ..... 104 R O J-20: Feedw ater .................................................................................................... 104 R O J-21: Instrum ent Air ................................................... ........................................ 105 ROJ-22: Core Spray ....... .... . ................................ 106 ROJ-23: High Pressure Coolant Injection .................................................................. 107 ROJ-24: Reactor Core Isolation Cooling ........................ 107 ROJ-25: Reactor Core Isolation Cooling ................................................................... 108 ROJ-26: High Pressure Coolant Injection ....................................... ........................ 108 ROJ-27: Control Rod Drive Hydraulics ..................................................................... 109 R OJ-28: Residual H eat Rem oval ............................................................................... 109 R OJ-29: Residual H eat Rem oval ..................................................................... ........... 110 ROJ-30: Feedwater ..................... ............................................. 110 Valve R elief R equests ................. .................................. *................................................................. 111 VR R -0 1: Withdraw n ................................................................................. ................ 111 VRR-02:. Traversing In-Core Probe.......................................................................... 112 VRR-03: Various Excess Flow Check Valves ........................................................... 114 VRR-04: High Pressure Coolant Injection .................................................................. 117 VRR-05: Ventilation System Control Valves ............................................................. 119 Rev. No. 10 Page 29 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 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 information:

" Individual valve identifier

• Drawing coordinates

" Code Class

• IST Category

" Nominal size

• Valve type

• Actuator type

'Test required

• Relief request (RR)/cold shutdown (CS) justification/ refueling outage (RO) justification

" Alternate test

" Test Procedure.

" Remarks Rev. No. 10 Page 30 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 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 Refueling Outage 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. 10 Page 31 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Valve Relief Requests VRR-XX refer to relief requests for the Valve Testing Program. Each valve request for relief provides the following information:

System Individual valve identifier Valve category Code Classification Safety Function

• Code test requirement for which relief is requested Basis for relief.

• Proposed alternate testing.

Rev. No. 10 Page 32 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Valve Symbols Valve Types 3W. Three-way valve AN Angle valve BF Butterfly valve BK Ball check BL Ball valve CK Swing check GA Gate valve GL Globe valve LK. Lift check NK Non-return valve PG Plug valve.

RD Rupture disk RV Relief valve SC Stop check SK Spring check TK Testable check WK Wafer check XP Explosive valve Valve Actuator Types AO *Airoperator 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. 10 Page 33 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Test Method Test Requirement PIT Valve position indication ETO Exercise test to open position ETC Exercise test to closed position PEO Partial exercise to open position PEC Partial exercise to closed position STO Full stroke time measured to open position STC Full stroke time measured to close position FSO Fail safe test to the open position*

FSC Fail safe test to the closed position LKJ Leak test per 10 CFR 50 Appendix J LKO Leak test for other than containment isolation valve RLF Relief valve test VBT Vacuum breaker operability test FFT Check valve forward flow verification test RFC Check valve reverse flow closure test PFT Check valve partial flow test MME Check valve exercise using manual mechanical exerciser DIS Check valve disassembly and inspection XPT Explosively actuated valve test RDT Rupture disk test XVD Explosive valve internal inspection.

NIT Non Intrusive Check Valve Test Test Frequency

-1 Quarterly -6 10 CFR 50 Appendix J

-2 Cold Shutdown -7 Appendix I Section 1-1320

-3 Refueling -8 Appendix I Section 1-1350

-4 1 year -9 ISTC-5260

-5 2 years -10 Appendix I Section 1-1360

-11 Tech Spec Requirement Rev. No. 10 Page 34 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 35 of 121 INSERVICE TESTING PROGRAM

- VALVE TABLE SYSTEM: Standby Gas Treatment DRAWING: FM-48A DWG CODE CLASS VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID/NAME CO-ORD ACT/PASS. CATEGORY SIZE (IN) TYPE TYPE POSITION REOTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 01-125MOV-100A C-6 2A 8 4.00 BF MO O/C STO-1 **ST-7E AUGMENTED ACTIVE STC-1 ST-7E PIT-5 ST-41D 01-125MOV-100B F-6 2A B 4.00 BF MO O/C STO-1 ST-7E AUGMENTED STC-1

• ST-7F PIT-5 ST-41 D 01-125MOV-11 G-8 2A 8 24.00 BF MO 0 STO-I ST-7E AUGMENTED PIT-5 ST-41D 01-125MOV-12 F-8 2A B 24.00 BF MO 0 STO-1 ST-7F AUGMENTED PIT-5 ST-41D 01-125MOV-14A D-6 2A B 24.00 BF MO O/C STO-1 ST-7E AUGMENTED STC-l ST-7E PIT-5 ST-41D 01-125MOV-14B E-6 2A B 24.00 BF MO O/C STO-1 ST-7F AUGMENTED STC-1 ST-7F PIT-5 ST-41D 01-125MOV-15A D-3 2A B 24.00 BF MO 0 STO-1 ST-7E AUGMENTED PIT-5 ST-41D 01-125MOV-15B F-3 2A B 24.00 BF MO 0 STO- I ST-7F AUGMENTED PIT-5 ST-41D

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 36 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Automatic Depressurization System DRAWING: FM-29A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE(IN) TYPE TYPE POSITION REQ"TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS ST-41 K PASSIVE 02AOV-17 1 B 1.00 GL AO C PIT-5 ST-41K PASSIVE 02AOV-18 1 B 1.00 GL AO C PIT-5 ST-41K

• PASSIVE 02RV-1 2 C 3.00 CK SA O/C ETO-1 ROJ-04 MME-3 ST-68B ETC-1 MME-3 ST-68B 02RV-2 2 C 3.0O CK SA O/C ETO-1 ROJ-04 MME-3 ST-68B ETC-1 MME-3 ST-68B 02RV-3 2 C 3.00 CK . SA O/C ETO- 1 ROJ-04 MME-3 ST-68B ETC-1 MME-3 ST-68B 02RV-4 2 C 3.00 CK SA O/C ETO-1 ROJ-04 MME-3 ST-68B ETC-1 MME-3 ST-68B 02RV-5 2 C 3.00 CK SA O/C ETO-1 ROJ-04 MME-3 ST-68B ETC-1 MME-3 ST-68B 02RV-6 2 C 3.00 CK SA O/C ETO-1 ROJ-04 MME-3 ST-68B ETC-1 MME-3 ST-68B 02RV-7 2 C 3.00 CK SA 0/C ETO-1 ROJ-04 MME-3 ST-68B ETC- I MME-3 ST-68B 02RV-8 2 C 3.00 CK SA O/C ETO-1 ROJ-04 MME-3 ST-6BB ETC-1 MME-3 ST-68B 02RV-9 2. C 3.00 CK SA O/C ETO-1 ROJ-04 MME-3 ST-68B ETC-1 MME-3 ST-68B 02RV-10 2 C 3.00 CK SA O/C ETO-1 ROJ-04 MME-3 ST-68B ETC-1 MME-3 ST-68B 02RV-11 2 C 3.00 CK SA O/C ETO-1 ROJ-04 MME-3 ST-68B ETC-I MME-3 ST-68B

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 37 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Automatic Depressurization System DRAWING: FM-29A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE POSITION REOTrS CSJ/ROJ REOUEST TEST PROCEDURE REMARKS 02RV-71A G-6 1 B/C 6.00 RV SA, AO 0 STO-1 ETO-3 ST-22B STC-1 RLF-7 MST-102.05 02RV-71B G-6 1 B/C 6.00 RV SA. AO 0 STO-1 ETO-3 ST-22B STC-1 RLF-7 MST-102.05 02RV-71C G-6 1 B/C 6.00 RV SA, AO 0 ,. STO-1 ETO-3 ST-22B STC-1 RLF-7 MST-102.05 02RV-71 D F-6 1 B/C 6.00 RV SA, AO 0 STO-1 ETO-3 ST-22B STC-1 RLF-7 MST-102.05 02RV-71E F-7 1 B/C 6.00 RV SA. AO 0 STO-1 ETO-3 ST-22B STC-1 RLF-7 MST-102.05 02RV-71F F-7 1 B/C 6.00 RV SA, AO 0 STO-1 ETO-3 ST-22B STC-1 RLF-7 MST-102.05 02RV-71G F-7 1 B/C 6.00 RV SA,AO 0 STO-1 ETO-3 ST-22B STC-1 RLF-7 MST-102.05 02RV-71H G-7 1 B/C 6.00 RV SA, AO 0 STO-1 ETO-3 ST-22B STC-1 RLF-7 MST-102.05 02RV-71J G-7 1 B/C 6.00 RV SA. AO 0 STO-1 ETO-3 ST-22B STC-1 RLF-7 MST-102.05 02RV-71K G-6 1 B/C 6.00 RV SA, AO 0 STO-1 ETO-3 ST-22B STC-1 RLF-7 MST-102.05 02RV-71L F-7 1 B/C 6.00 RV SA, AO 0 STO-1 ETO-3 ST-22B STC-1 RLF-7 MST-102.05 02VB-1 C-7 2 C 10.00 CK SA O/C ETO-1 ROJ-04 MME-3 ST-68B ETC-1 MME-3 ST-68B 02VB-2 C-7 2 C 10.00 CK SA O/C ETO-1 ROJ-04 MME-3 ST-68B ETC-i MME-3 ST-68B

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 38 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Automatic Depressurization System DRAWING: FM-29A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE POSITION REQaS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 02VB-3 C-7 2 C 10.00 CK SA O/C ETO-1 ROJ-04 MME-3 ST-68B ETC-1 MME-3 ST-68B 02VB-4 C-7 2 C 10.00 CK SA O/C ETO-1 ROJ-04

• MME-3 ST-68B ETC-1 MME-3 ST-68B 02VB-5 C-7 2 C 10.00 CK SA O/C ETO-1 ROJ-04 MME-3 ST-68B ETC-1 MME-3 ST-68B 02VB-6 C-7 2 C 10.00 CK SA O/C ETO-I ROJ-04 MME-3 ST-68B ETC-1 MME-3 ST-68B 02VB-7 C-7 2 C 10.00 CK SA O/C ETO-1 ROJ-04 MME-3 ST-68B ETC-1 MME-3 ST-68B 02VB-8 C-7 2 C 10.00 CK SA O/C ETO-I ROJ-04 MME-3 ST-68B ETC-1 MME-3 ST-68B 02VB-9 C-7 2 . C 10.00 CK SA O/C ETO-I ROJ-04 MME-3 ST-68B ETC-1 MME-3 ST-68B 02VB-10 C-7 2 C 10.00 CK SA O/C ETO-I ROJ-04 MME-3 ST-68B ETC-i MME-3 ST-68B 02VB-11 C-7 2. C 10.00 CK SA O/C ETO-1 ROJ-04 MME-3 ST-68B ETC-1 MME-3 ST-68B

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 39 of 121 INSERVICE TESTING PROGRAM.

VALVE TABLE SYSTEM: Reactor Water Recirc DRAWING: FM-26A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REQTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 02-2AOV-39 E-4 1 A 1.00 GA AO C STC-1 ST-IC FSC-1 ST-iC PIT-5 ST-41K LKJ-6 ST-39B-X41 02-2AOV-40 E-3 1.00 GA AO C STC-i ST-iC FSC-i ST-iC PIT-5 ST-41K LKJ-6 ST-39B-X41 02-2EFV-PS-128A B-6

• A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-2EFV-PS-128B B-6 A/C 1.00 BK SA C ETC-1 ROJ-01. VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-2EFV-PT-24A C-3 A/C 1.00 BK. SA C ETC-1. ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3  : ISP-i 02-2EFV-PT-24B C-8 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-i VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-2EFV-PT-25A C-3 A/C .1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-2EFV-PT-25B C-8 A/C 1.00 BK SA C ETC-1 ROJ-01. VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-2RWR-13A C-3 A/C 0.75 SK SA C RFC-i ROJ-02 RFC-3 ST-39B-X31AC LKJ-6 FFT-1 ISTC-3550 02-2RWR-13B C-8 A/C 0.75 SK SA C RFC-1 ROJ-02 RFC-3 ST-398-X31BC LKJ-6 FFT-1 ISTC-3550 02-2RWR-41A D-3 A/C 0.75 SK SA C RFC-1 ROJ-03 RFC-3 ST-39B-X31AC LKJ-6 FFT-1 . ISTC-3550 02-2RWR-41B D-8 A/C 0.75 SK SA

• C RFC-i ROJ-03 RFC-3 " ST-39B-X31BC LKJ-6 FFT-1 ISTC-3550 A/C 1.00

• BK SA C ETC-1 ROJ-01

• VRR-03 ETC-3 ISP-1 02-2EFVl-DPTI 1lA E-3 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 40of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Reactor Water Recirc DRAWING: FM-26A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE POSITION RE0TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 02-2EFVI-DPT111B E-8 1 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 . LKO-3 ISP-1 02-2EFV1 -FT110A F-3 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-2EFV1-FT1 i0C D-3 NC 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-2EFV1-FT110E F-8 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-2EFV1-FT1lOG D-8 A/C 1.00 BK SA C ETC-i ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-2EFV2-DPT111A E-3 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-2EFV2-DPT1 11B E-8 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-2EFV2-FT1 10A F-3 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-2EFV2-F1 I0C D-3 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-2EFV2-FT1 10E F-8 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON.EXCESS FLOW LKO-5 LKO-3 ISP-1 022EFV2-FTll0G D-8 NC 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-2MOV-53A C-3 B 28.00 GA MO C STC-1 CSJ-01 STC-2 ST-27A PIT-5 ST-41K 02-2MOV-53B C-8 B 28.00 GA MO C STC-1 CSJ-01 $STC-2 ST-27A PIT-5 ST-41K

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 41 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Nuclear Boiler Instrumentation DRAWING: .FM-47A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE POSITION REO'TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 02-3EFV-1 1 F-7 1 A/C 1.00 BK SA C ETC-1 . ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATE ON EXES FLOW LKO-5 LKO-3 ISP-i 02-3EFV-13A E-7 1 A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-13B E-4 1 A/C 1.00 BK ETC-I ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-15 A E-7 1 NC 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-15B E-4 1 A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-15N B-7 A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-i 02-3EFV-17A D-7 1 A/C 1.00 BK ETC-1 ROJ-01 VRR-03 *ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-178 D-4 1 A/C 1.00 BK ETC-I ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-19A D-7 I A/C i.00 BK ETC-I ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-19B D-4 1 A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-21A H-5 1 A/C 1.00 BK ETC-1 . ROJ-01 VRIl.03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-21B C-7 1 A/C 1.00 BK ETC-i ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-21C *C-4 1 A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-i VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-21D H-4 1 A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-I VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-23 F-7 I A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page.42 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Nuclear Boiler Instrumentation DRAWING: FM-47A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE 1JD CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE POSITION REOTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 02-3EFV-23A H-5 1 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1

• VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-23B A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-23C A/C 1.00 BK ETC-i ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-23D A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-25 A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-31A A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-i 02-3EFV-31B A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-i 02-3EFV-31C A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-31D A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-31E A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-31F A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-31G A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-31H A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-31J A/C 1.00 BK ETC-I ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-31K A/C 1.00 BK ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 43 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Nuclear Boiler Instrumentation DRAWING: FM-47A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE POSITION REQ'TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 02-3EFV-31L H-4 1 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-31M D-4 I A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-31N H-4 1 A/C 1.00 BK SA C ETC-i .ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-i 02-3EFV-31P H-4 1 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-31R G-4 1 NC 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-i VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-31S G-4 I NC 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 02-3EFV-33 B-4 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 44 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Control Rod Drive DRAWING: FM-27B DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE(IN) TYPE TYPE POSITION REO'TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS O3AOV-126 C-4 2 B 1.00 GL AO 0 STO-1 ETO-3 RAP-7.4.1 SCRAM TIME TEST FSO-1 RAP-7.4.1 GL89-04.POSITION 7 O3AOV-127 D-4 2 B 1.00 GL AO 0 STO-1 ETO-3 RAP-7.4.1 SCRAM TIME TEST FSO-I RAP-7.4.1 GL89-04 POSITION 7 03AOV-32 H-4 2 B 1.00 GL AO C STC-1 ST-20MB FSC-1 ST-20MB PIT-5 ST-41D 03AOV-33 F-4 2 B 2.00 GL AO C STC-1 ST-20MB FSC-1 ST-20MB PIT-5 ST-41D 03AOV-34 H-4 2 B. 1.00 GL AO C STC-1 ST-20MB FSC- I ST-20MB PIT-5 ST-41D 03AOV-35 F-4 2 B 2.00 GL AO C STC-1 ST-20MB FSC-1 ST-20MB PIT-5 ST-41D 03AOV-36 H-6 2 B 1.00 GL AO C STC-1 ST-20MA FSC-1 ST-20MA PIT-5 ST-41D 03AOV-37 F-6 2 B 2.00 GL AO C STC-1 ST-2OMA FSC- I ST-2OMA PIT-5 ST-41D 03AOV-38 H-6 2 B 1.00 GL AO C STC-1 ST-20MA FSC-1 ST-20MA PIT-5 ST-41D 03AOV-39 F-6 2 B 2.00 GL AO C STC-1 ST-2OMA FSC-1 ST-20MA PIT-5 ST-41D 03HCU-114 D-4 2 C 0.75 BK SA 0 FFT-1 FFT-3 RAP-7.4.1 SCRAM TIME TEST GL89-04 POSITION 7 RFC-l Skid Mounted 03HCU-115 C-4 2 C 0.50 BK SA C RFC-1 ROJ-27 RFC-3 ST-68A FFT-1 ISTC-3500 03HCU-138 C-4 2 C 0.50 BK SA C RFC-1 ST-20C RFC VIA ROD MOTION FFT-1 ISTC-3500

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 45 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Control Rod Drive DRAWING: .FM-27B DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE(IN) TYPE TYPE POSITION REQ'TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 03SOV-120 C-4 2 B 0.50 GA SO C STC-1 ETC-3 ST-20C SCRAM TIME TEST FSC-1 ST-20C GL89-04 POSITION 7 O3SOV-121 C-4 2 B 0.50 GA SO C STC-1 ETC-3 ST-20C. SCRAM TIME TEST FSC-1 ST-20C GL89-04 POSITION 7 03SOV-122 C-4 2 B 0.50 GA SO C STC-1 ETC-3 ST-20C SCRAM TIME TEST FSC-1 ST-20C GL89-04 POSITION 7 03Sov-123. C-4 2 B 0.50 GA SO C STC-1 ETC-3 ST-20C SCRAM TIME TEST FSC-1 ST-20C GL89-04 POSITION

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 46 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Traversing Incore Probe DRAy VING: FM-119A DWG VAALVE VALVE ACTUATOR SAFETY TEST

• RELIEF ALTERNATE VALVE ID CO-ORD CLASS CATIEGORY SIZE (IN) TYPE TYPE POSITION REOTS CSJ/ROJ REQUEST

• TEST REMARKS 07EV-104A F-5 2 D 0.375 XP "S C XPT-9 ISP-7-2 07EV-104B F-4 2 D 0.375 XP So C XPT-9 ISP-7-2 07EV-104C F-4 2 D 0.375 XP SQ C XPT-9 ISP-7-2 07SOV-104A F-5 2 A 0.375 BL sO C STC-1 VRR-02 ST-lC FSC-1 ST-IC PIT-5 ST-41K LKJ-6 ST-39B-X35 07SOV-104B F-4 A 0.375 1EL SO C STC-1 VRR-02 ST-lC FSC-1 ST-lC PIT-5 ST-41K LKJ-6 ST-39B3-X35 07SOV-104C F-4 A 0.375 EL- SO C STC-1 VRR-02 ST-lC

• FSC-1 ST-lC PIT-5 ST-41 K LKJ-6 ST-39B-X35

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 47 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Residule Heat Removal DRAWING: FM-20A DWG VALVE VALVE ACTUATOR SAFETY . TEST . RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE POSITION REQTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 10AOV-68A F-6 1 A/C 24.00 TK SA, AO O/C FFT-1 ROJ-29 FFT-3 ST-2AG RFC-1 ROJ-29 LKO-5 ST-39J LKO-5 ST-39J 1OAOV-68B F-5 1 A/C 24.00 TK SA. AO O/C FFT-1 ROJ-29 FFT-3 ST-2AH RFC-1 ROJ-29 LKO-5 ST-39J LKO-5 ST-39J 10MOV-13A B-6 2 B 20.00 GA MO O/C STO-1 ST-2AL

-STC-1 ST-2AL PIT-5 ST-41K 10MOV-138 C-4 2 B 20.00 GA MO O/C STO-1 ST-2AM STC-1 ST-2AM PIT-5 ST-41K 10MOV-13C C-6 2 B 20.00 GA MO O/C STO-1 ST-2AL STC-1 . . ST-2AL PIT-5 ST-41K 10MOV-13D C-5 2 B 20.00 GA MO O/C STO-1

• ST-2AM STC-1 ST-2AM PIT-5 ST-41K 10MOV-15A C-6 2 B 20.00 GA MO C STC-1 ST-2AL PIT-5 ST-41K 10MOV-15B C-4 2 B 20.00 GA MO C STC-1 ST-2AM PIT-5 ST-41K IOMOV-15C C-6 2 B 20.00 GA MO C STC-1 ST-2AL PIT-5 ST-41K 10MOV-15D C-4 2 B 20.00 GA M0 C STC-1 " ST-2AM PIT-5 ST-41K 1OMOV-16A D-8 2 B 4.00 GA MO . O/C STO-1 ST-2AL STC-1 ST-2AL PIT-5 ST-41 K

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 48 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Residule Heat Removal DRAWING:. FM-20A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE(IN) TYPE TYPE FUNCTION REO`TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 10MOV-16B D-3 2 B 4.00 GA MO O/C STO-1 ST-2AM STC-1 ST-2AM PIT-5 ST-41K 10MOV-17 D-5 A 20.00 GA MO C STC-1 ROJ-28 STC-3 ST-1S PIT-5 ST-41K LKO-5 SATISFIED BY LKJ-3 LKO-5 LKJ-3 TST-121 PER JAF-CALC-MISC-00554 LKJ-6 ST-39B-X12 10MOV-18 E-5 A 20.00 GA MO C STC-1 ROJ-28 STC-3 ST-IS JAF-SE-96-017 PIT-5 ST-41K LKO-5 ST-2AS 1OMOV-25A F-8 A 24.00 GA MO O/C STO-1 ST-2AL LKO-5 SATISFIED BY LKJ-3 STC-1 ST-2AL PER JAF-CALC-MISC-00554 PIT-5 ST-41K LKO-5 LKJ-3 ST-2JA LKJ-6 ST-39B-X13A 10MOV-25B F-3 A 24.00 GA MO O/C STO-1 ST-2AM LKO-5 SATISFIED BY LKJ-3 STC-1 ST-2AM PER JAF-CALC-MISC-00554 PIT-5 ST-41K LKO-5 LKJ-3 ST-2JB LKJ-6 ST-39B-X13B 10MOV-26A G-7 2 A 10.00 GA MO O/C STO-1 ST-2AL JAF-SE-96-017 STC-1 ST-2AL PIT-5 ST-41K 1OMOV-26B G-4 2 A 10.00 GA MO O/C STO-1 ST-2AM JAF-SE-96-017 STC-1 ST-2AM PIT-5 ST-41K 1OMOV-27A F-8 A 18.00 AN MO O/C STO-1 ST-2AL JAF-SE-96-017 STC-1 ST-2AL PIT-5 ST-41K

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 49 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Residule Heat Removal DRAWING: FM-20A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST.

VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE . FUNCTION REO'TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 10MOV-27B F-3 1 A 18.00 AN MO O/C STO- 1 ST-2AM JAF-SE-96-017 STC-.1 ST-2AM PIT-5 ST-41K IOMOV-31A G-6 2 A 10.00 GL MO O/C STO-1 ST-2AL STC-1 ST-2AL PIT-5 ST-41 K LKJ-6 ST-3913-X39A IOMOV-31B G-5 2 A 10.00 GL MO O/C STO-1 ST-2AM STC-1 ST-2AM PIT-5 ST-41 K LKJ-6 ST-39B-X39B.

10MOV-34A E-7 2 B 14.00 GL MO O/C STO-1 ST-2AL STC-t ST-2AL PIT-5 ST-41K IOMOV-34B E-3 *2 B 14.00 GL MO O/C STO-1 ST-2AM STC-1 ST-2AM PIT-5 ST-41K IOMOV-38A E-7 2 A 4.00 GL MO O/C STO-1 ST-2AL STC-1 ST-2AL PIT-5. ST-41K LKJ-6 ST-39B-X2 11A 1OMOV-38B E-4 2 A 4.00 GL MO O/C STO-1 ST;2AM STC-1 ST-2AM ST-41K PIT-5 LKJ-6 ST-39B-X21 IB IOMOV-39A E-8 2 A 16.00 . GL MO O/C. STO-1 ST-2AL JAF-SE-96-017 STC-1 ST-2AL PIT-5 ST-41K.

10MOV-39B E-3 2 16.00 GL MO O/C STO-1 ST-2AM JAF-SE-96-017 STC-1 ST-2AM PIT-5 ST-41 K

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 50 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Residule Heat Removal DRAWING: FM-20A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REQTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 1OMOV-66A 0-8 2 B 20.00 GL MO O/C STO-1 ST-2AL STC- 1 ST-2AL PIT-5 ST-41K IOMOV-66B D-3 2 20.00 GL MO O/C STO-1 ST-2AM STC-1 ST-2AM PIT-5 ST-41K 1ORHR-262 H-3 2 C 4.00 CK SA C RFC-1 ST-2AD FFT-1 ST-2AD 10RHR-277 G-8 2 C 4.00 CK SA C RFC-1 ST-2AD FFT-1 ST-2AD 10RHR-42A C-8 .2 C 16.00 CK SA O/C FFT-1 ST-2AL RFC-1 ST-2AL IORHR-42B C-3 2 C 16.00 CK SA O/C FFT-1 ST-2AM RFC-t ST-2AM 10RHR-42C C-8 2 C 16.00 CK SA O/C FFT-1 ST-2AL RFC-1 ST-2AL 10RHR-42D C-3 2 C 16.00 CK SA O/C FFT-1 ST-2AM RFC-1 ST-2AM 10RHR-64A C-8 2 C 3.00 CK SA O/C FFT-1 ROJ-05 PFT-1 ST-2AL AT LEAST ONE VALVE PER OUTAGE RFC-1 DIS-3 MST-059.12 WITH ALL VALVES IN GROUP INSPECTED AT LEAST ONCE/8 YRS.

10RHR-64B C-3 2 C 3.00 CK SA O/C FFT-. ROJ-05 PFT-1 ST-2AM AT LEAST ONE VALVE PER OUTAGE RFC-1 DIS-3 MST-059.12 WITH ALL VALVES IN GROUP INSPECTED AT LEAST ONCE/8 YRS.

1ORHR-64C D-8 2 3.00 CK SA O/C FFT-1 ROJ-05 PFT-1 ST-2AL AT LEAST ONE VALVE PER OUTAGE RFC-1 DIS-3 MST-059.12 WITH ALL VALVES IN GROUP INSPECTED AT LEAST ONCE/8 YRS.

10RHR-64D D-3 2 C 3.00 CK SA O/C FFT-I ROJ-05 PFT-i ST-2AM AT LEAST ONE VALVE PER OUTAGE RFC-1 DIS-3 MST-059.12 WITH ALL VALVES IN GROUP INSPECTED AT LEAST ONCE/8 YRS.

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 51 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Residule Heat Reemoval DRAWING: FM-20A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE(IN) TYPE TYPE FUNCTION REOTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS IORHR-95A C-8 2 C 0.75 SK SA C RFC-1 ROJ-06 RFC-3 ST-2AB FFT-1 ISTC-3550 10RHR-95B B-5 .2 C 0.75 SK SA C RFC-1 ROJ-06 RFC-3 ST-2AB FFT-1 ISTC-3550 10RV-41A C-7 2 C 1.00 *RV SA 0 RLF-8 MP-059.07 1ORV-41B. C-4 2 C 1.00 RV SA 0 RLF-8 MP-059.07 1CRV-41C C-7 2 C 1.00 RV SA 0 RLF-8 MP-059.07 C

1ORV-41D C-4 2 1.00 RV SA 0 RLF-8 MP-059.07 10SV-35A E-8 2 C 1.00 . RV SA 0 RLF-8 MP-059.07 10SV-35B E-3 2 C 1.00 RV SA RLF-8 0 MP-059.07 10SV-40 D-5 21 1.00 RV SA 0 RLF-8 MP-059.07

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 52 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Residule Heat Removal DRAWING: FM-20B DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REQ'TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 10AOV-71 A F-6 2 B 3.00 GL AO C PIT-5 ST-41K PASSIVE 10AOV-71B F-5 2 B 3.00 GL AO C PIT-5 ST-41K PASSIVE 1OMOV-12A F-6 2 B 16.00 GA MO O PIT-5 ST-41K PASSIVE 10MOV-12B F-5 2 B 16.00 GA MO O PIT-5 ST-41K PASSIVE 1OMOV-148A E-8 3 B 16.00 GA MO C . PIT-5 ST-41K PASSIVE 1OMOV- 148B E-2 3 B 16.00 GA MO C PIT-5 ST-41K PASSIVE 10MOV-149A D-8 3 B 16.00 GA MO C PIT-5 ST-41K PASSIVE 10MOV-149B D-2 3. B 16.00 GA MO C PIT-5 ST-41K PASSIVE 10MOV-167A F-8 2 B 1.00 GL MO C PIT-5 ST-41K PASSIVE 10MOV-167.B F-3 *2 B 1.00 GL MO C PIT-5 ST-41K PASSIVE 1OMOV-65A G-6 2 B 16.00 GL MO O PIT-5 ST-41K PASSIVE 10MOV-65B G-5 2 B 16.00 GL MO O PIT-5 ST-41K PASSIVE 10MOV-89A D-6 3 B 16.00 GA MO O STO-1 ST-2XA PIT-5 ST-41D 1DMOV-89B E-5 3 B 16.00 GA MO O STO-1 ST-2XB PIT-5 ST-41D 10RHR-14A B-7 3 C . 12.00 CK SA O/C FFT- I ST-2XA RFC-1 ST-2XA I0RHR-148 B-4 3 C 12.00 CK SA O/C FFT-1 ST-2XB RFC-1 ST-2XB 10RHR-14C C-7 3 C. 12.00 CK SA O/C FFT-1 ST-2XA RFC-1 ST-2XA

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 53 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Residule Heat Removal DRAWING: FM-20B DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION RE0TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 10RHR-14D C-4 3 C 12.00 CK SA O/C FFT-1 ST-2XB RFC-I ST-2XB 10RV-43A E-7 3 C 0.75 RL SA 0 RLF-8 MP-059.07 I0RV-43B E-4 3 C 0.75 RL SA 0 RLF-8 MP-059.07 1ORV-46A F-7 2 C 0.75 RL SA 0 RLF-8 MP-059.07 10RV-46B F-3 2 C 0.75 RL SA 0 " IRLF-8 MP-059.07 10SOV-101A B-6 3 B 0.75 GL so 0 STO-1 ST-2XA FSO-1 ST-2XA

• 10SOV-101B B-5 3 B 0.75 GL so 0 STO-1 ST-2XB FSO-1 ST-2XB 10SOV-101C C-6 3 B 0.75 GL so 0 STO-I ST-2XA FSO-1 ST-2XA 16SOV-IOID C-5 3 B 0.75 GL so 0 STO-1 ST-2XB FSO-1 ST-2XB

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 54 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Residule Heat Removal DRAWING: FM-18C DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE(IN) TYPE TYPE FUNCTION RE0TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 1OSOV-203 E-7 2 B 0.50 GA SO *C PIT-5 ST-41D PASSIVE 1OSOV-204 D-7 2 B 0.50 GA SO C PIT-5 ST-41D PASSIVE

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Pa ge 55 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEMWStandby Lquid Control DRAWING' FM-21A DWG . VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION RE0-S CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 11EV-14A D-6 1 D 1.50 XP s0 0 XPT-9 MST-011.11 XVD-3 MST-011.11 11EV-14B B-6 1 D 1.50 XP s0 0 XPT-9 MST-011.11 XVD-3 MST-011.11 1ISLC-16 C-7 A/C 1.50 CK SA O/C FFT-1 ROJ-07 FFT-3 ST-6M RFC-I RFC-3 ST-39B-X42 LKJ-6 ST-39B-X42

. ST-6M 1 SLC-17 D-7 1 A/C 1.50 SK SA O/C FFT-1 ROJ-07 FFT-3 RFC-1 RFC-3 ST-39B-X42 LKJ-6 ST-39B-X42 11SLC-43A D-6 2 C. 1.50 SK SA 0/C FFT-1 ST-6HA RFC-1 ST-6HB 11SLC-43B B-6 2 C 1.50 SK SA O/C FFT-1 ST-6HB RFC-1 ST-6HA 11SV-39A D-4 2 C 1.00 RV SA C RLF-8 MP-059.07 11SV-39B C-4 2 C 1.00 RV SA C RLF-8 MP-059.07

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 56 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Reactor Water Clean Up DRAWING: FM-24A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE(IN) TYPE TYPE FUNCTION REO'TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 12MOV-15 E-8. 1 A 6.00 GA MO C STC-1 CSJ-07 ST-26M PIT-5. ST41K LKJ-6 ST-39B-X14 12MOV-18 E-7 1 A 6.00 GA MO C STC-l CSJ-07 ST-26M PIT-5 ST-41K LKJ-6 ST-39B-X14 12MOV-69 H-7 1 A 4.00 GA MO C STC-1 CSJ-07 ST-26M PIT-5 ST-41K LKJ-6 ST-39B-X9

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 57 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Reactor Core Isolation Cooling DRAWING: FM-22A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REOTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 13EFV-0IA F-7. 1 A/C 1.00 BK SA C ETC-1 ROJ-01 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 13EFV-01B . F-7 1 NC 1.00 BK SA C ETC-i ROJ-01 ETC-3. ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 13EFV-02A G-7 1 ANC 1.00 BK. SA C ETC-1 ROJ-01 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 13EFV-02B F-7 1 ANC 1.00 BK SA C ETC-I ROJ-01 ETC-3 ISP-i *VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 13MOV-15 F-7 1 A 3.00 GA MO C STC-1 ST-24J PIT-5 ST-41D LKJ-6 ST-39B-X10 13MOV-16 F-7 1 A . 3.00 GA MO C STC-1 ST-24J PIT-5 ST-41D LKJ-6 ST-39B-XIO 13MOV-21 F-5 I A 4.00 GA MO C STC-1 ST-24J PIT-5 ST-41D LKJ-6 ST-39B-X9 13MOV-27 E-5 2 B 2.00 GL MO C STC-1 ST-24J PIT-5 ST-41D 13MOV-41 D-7 2 B 6.00 GA MO C STC-1 ST-24J PIT-5 ST-41D 13MOV-130 E-6 2 B 1.50 GA MO 0 PIT-5 ST-41D PASSIVE 13RCIC-22 F-6 IA C 4.00 CK. SA 0 FFT-1 CSJ-09 MME-2 ST-4H AUGMENTED RFC-1 MME-2 ST-4H 13RCIC-37 E-6 2 C 1.50 CK SA 0 FFT-1 ROJ-24 FFT-2 ST-4T RFC-1 RFC-2. ISTC-5223 13RCIC-38 E-6 2 C 1.50 CK SA 0 FFT-1 ROJ-24 FFT-2 . ST-4T RFC-1 RFC-2 ISTC-5223 13RCIC-4 D-6 . 2 NC 8.00 LK SA C RFC-1 ROJ-08 RFC-3 ST-24J FFT-1 ST-24J LKJ-6 ST-39B-X212 13RCIC-5 C-6 2 A/C 8.00 LK SA C RFC-l ROJ-08 RFC-3 ST-24J FFT-i1 ST-24J LKJ-6 ST-39B-X212

• 13RCIC-7 C-7 2 C 2.00 SC SA, MA C RFC-1 ROJ-25 RFC-3 ST-24L FFT-1 ROJ-25 FFT-3

• ST-24L

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 58 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Core Spray DRAWING: FM-23A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS. CATEGORY SIZE (IN) TYPE TYPE FUNCTION RE0TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 14AOV-13A G-6 1 A/C 10.00 TK SA, AO O/C FFT-1 ROJ-09 FFT-3 ST-3F RFC-1 RFC-3 ST-39J PIT-5 ST-41K LKO-5 LKO-3 ST-39J PEO-2 ST-3M PEC-2 ST-3M 14AOV-13B G-5 A/C 10.00 TK SA, AO O/C FFT-i ROJ-09 FFT-3 ST-3F RFC-1 RFC-3 ST-39J PIT-5 ST-41K LKO-5 LKO-3 ST-39J PEO-2 ST-3M PEC-2 ST-3M 14CSP-10A D-8 2 C 12.00 CK SA 0 FFT-1 ROJ-22 PFT-1/FFT-3 ST-3PA/3F RFC-1 ROJ-22 NIT-4 EDP-PC-106 14CSP-10B D-3 2 C 12.00 CK SA 0 FFT-1 ROJ-22 PFT-1/FFT-3 RFC-1 ROJ-22 NIT-4 EDP-PC-106 14CSP-62A E-7 2 C 1.00 SK SA C RFC-1 ROJ-10 RFC-3 ST-3U FFT-1 ISTC-3550 14CSP-62B E-3 2 C 1.00 SK SA C RFC-1 ROJ-I0 RFC-3 ST-3U FFT-1 ISTC-3550 14CSP-76A F-7 2 C 2.00 SK SA C RFC-1 ST-30A FFT-1 ST-30A 14CSP-76B F-4 2 C 2.00 SK SA C RFC-I ST-30B FFT-1 ST-3QB 14EFV-31A E-4 1 A/C 1.00 BK SA O/C ETC-1 ROJ-01 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 14EFV-31B E-4 1 A/C 1.00 BK SA O/C ETC-i ROJ-01 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 14MOV-11A F-7 1 A 10.00 GA MO O/C STO-1 ST-3PA JAF-SE-96-017 STC-1 ST-3PA PIT-5 ST-41K 14MOV-11B F-4 I A. 10.00 GA MO O/C STO-1 ST-3PB JAF-SE-96-017 STC-1 ST-3PB PIT-5 ST-41K

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 59 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Core Spray DRAWING: FM-23A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE(IN) TYPE TYPE FUNCTION REQ'TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 14MOV-12A F-6 1 A 10.00 GA MO O/C STO-1 ST-3PA STC-1 ST-3PA PIT-5 ST-41K LKO-5 LKJ-3 TST-1!7 LKO-5 SATISFIED BY LKJ-3 LKJ-6 ST-390-X16A PER JAF-CALC-MISC-00554 14MOV-12B F-4 1 A 10.00 GA MO O/C STO-1 ST-3PB STC-1 ST-3PB PIT-5 ST-41K LKO-5 LKJ-3 TST-118 LKO-5 SATISFIED BY LKJ-3 LKJ-6 ST-39B-X16B PER JAF-CALC-MISC-00554 14MOV-26A F-7 2 B 8.00 GL MO C STC-1 ST-3PA PIT-5 ST-41 D 14MOV-26B F-3 2 B 8.00 GL MO C STC-1 ST-3P8 PIT-5 ST-41D 14MOV-5A E-7 2 B - 3.00 GA MO O/C STO-1 ST-3PA STC-1 ST-3PA PIT-5 ST-41D 14MOV-5B E-3 2 B 3.00 GA MO O/C STO-1 ST-3PB STC-1 ST-3PB PIT-5 ST-41D 14MOV-7A C-6 2 B 16.00 GA MO O/C STO-1 ST-3PA STC-1 ST-3PA PIT-5 ST-41D 14MOV-7B C-4 2 B 16.00 GA MO O/C STO-1 .ST-3P8 STC-1 ST-3PB PIT-5 ST-41D 14SV-20A E-8 2 C 1.50 . RL SA 0 RLF-8 MP-059.07 2 C 1.50 RL SA 0 RLF-8 MP-059.07 14SV-20B E-2

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 60 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Reactor Building Closed Loop Cooling Water DRAWING: FM-15B DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REOTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 15AOV-130A C-7 2 A 6.00 GL AG C STC-1 CSJ-02 STC-2 ST-IR PIT-5 ST-41K LKJ-6 ST-39B-X23 15AOV-130B D-4 2 A 4.00 GL AG C STC-1 CSJ-02 STC-2 ST-1R PIT-5 ST-41K LKJ-6 ST-39B-X24 15AOV-131A E-7 A 4.00 GL AG C STC-1 *CSJ-02 STC-2 ST-lR PIT-5 ST-41K LKJ-6 ST-39B-X66 15AOV-131B E-4 2 A 4.00 GL AO C STC-1 CSJ-02 STC-2 ST-1R PIT-5 ST-41K LKJ-6 ST-39B-X62 15AOV-132A F-4 2 A 4.00 GL AG C STC-1 CSJ-02 STC-2 ST-1R PIT-5 ST-41K LKJ-6 ST-39B-X63 15AOV-132B F-7 2 A 4.00 GL AO C STC-1 CSJ-02 STC-2 STAIR PIT-5 ST-41K LKJ-6 ST-39B-X67 15AOV-133A F-4 2 A 4.00 GL AO C STC-1 CSJ-02 STC-2 ST-AR PIT-5 ST-41K LKJ-6 ST-39B-X64 15AOV-133B F-7 2 A 4.00 GL AO C STC-1 CSJ-02 STC-2 ST-1R PIT-5 ST-41K LKJ-6 ST-39g-X68 15AOV-134A C-6 2 A 1.50 GL AO C STC-1 CSJ-02 STC-2 ST-1R PIT-5 ST-41K LKJ-6 ST-39B-X65 15RBC-61 F-7 3A C 1.00 SK SA C RFC-1 ST-41J FFT-1 Augmented exclusion

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 61 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Reactor Building Closed Loop Cooling Water DRAWING: FM-18C DWG VALVE VALVE* ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REQOTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 15RBC-214 E-7 3 C 1.00 CK SA .C RFC-1 ROJ-11 DIS-3 MST-059.45 GL-89-04 Pos. 2 FFT-1 ROJ-11 DIS-3 MST-059.46 15SOV-215 E-7 B 1.00 GL SO C PIT-5 ST-41D PASSIVE

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 62 o 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Leak Rate Aanalysis DRAWING: FM-49A DWG VALVE VALVE ACTUATOR SAFETY TEST

• RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE(IN) TYPE TYPE FUNCTION REOQTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 16-1AOV-IOIA D-7 2 A 0.375 GA AO C STC-1 ST-iC FAST ACTING VALVE FSC-1 ST-tC PIT-5 ST-41K LKJ-6 ST-39B-X45 16-lAOV-101B E-7 A 0.375 GA AO C STC-1 ST-lC FAST ACTING VALVE FSC-I ST-IC PIT-5 *ST-41K LKJ-6 ST-398-X45 16-1AOV-102A D-7 A 0.375 GA AO C STC-1 ST-lC FAST ACTING VALVE FSC-1 ST-IC PIT-5 ST-41K LKJ-6 ST-39B-X218 16-IAOV-1028 C-7 2 A 0.375 GA AO C STC-1 ST-iC FAST ACTING VALVE FSC-1 ST-iC PIT-5 ST-41K LKJ-6 ST-39B-X218

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 63 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Radwaste DRAWING: FM-17A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE(IN) TYPE TYPE FUNCTION REO'TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 20AOV-83 H-6 2 A 3.00 BL AO C STC-1 ST-lC FAST ACTING VALVE FSC-1 ST-lU PIT-5 ST-41K LKJ-6 ST-39B-X18 20AOV-95 D-6 3.00 BL AO C STC-1 ST-iC FAST ACTING VALVE FSC-1 ST-i U PIT-5 ST-41K LKJ-6 ST-39B-X19 20MOV-82 H-7 A 3.00 GA MO C STC-1 ST-1C PIT-5

• ST-41K LKJ-6 ST-39B-X18 20MOV-94 D-7 A 3.00 GA MO C STC-1 ST-iC PIT-5 ST-41K LKJ-6 ST-39B-XI9 20RD-18 H-7 D 1.00 RD SA 0 RDT-10 PM TASK

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 64 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: High Pressure Coolant Injeection DRAWING: FM-25A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE.ID CO-ORD CLASS CATEGORY SIZE,(IN) TYPE TYPE FUNCTION REOQTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 23AOV-42 G-2 2 IS 1.00 GA AO C STC-1 ST-4N FAST ACTING VALVE FSC-1 ST-4N PIT-5 ST-4ID 23EFV-01A G-6 1 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1

• 23EFV-01B G-7 1 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 23EFV-02A G-7 1 ANC 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 23EFV-02B G-7 I A/C 1.00 BK SA C " ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW ISP-1 LKO-5 LKO-3 23HOV-1 F-3 2 B 10.00 GA HO O/C STO-1 ST-4N FAST ACTING VALVE STC-I ST-4N PIT-5 ST-41D 23HPI-12 C-6 2 A/C 16.00 LK SA O/C FFT-1 ST-4N RFC-1 ROJ-12 RFC-3 ST-39B-X214 LKJ-6 ST-39B-X214

• 23HP1-13 C-7 2 C 2.00 SC . SA, MA O/C FFT-I ROJ-13 DIS-3 MST-059.11 RFC-1 ROJ-26 RFC-2 ST-4W 23HP1-130 C-5 2 C 2.00 SK SA O/C FFT-1 ROJ-17 VRR-04 DIS-3 MST-059.45 RFC-1 PFT-I ST-4N 23HP1-.131 C-5 2 C 2.00 SK SA C RFC-1 ROJ-18 VRR-04 DIS-3 MST-059.45 FFT-i DIS-3 MST-059.45 23HP1-18 F-7 I1 C. 14.00 CK SA 0 FFT-1 CSJ-03 MME-2 ST-4H RFC-I MME-2 ST-4H 23HPI-32 G-5 2 C 16.00 CK SA C RFC-1 ROJ-i4 VRR-04 DIS-3 MST-059.12 FFT-1 DIS-3 MST-059.12 23HPI-402 E-7 2A C 2.00 CK SA O/C FFT-1 ROJ-23 FFT-2 ST-4T RFC-1 RFC-2 ISTC-5223 23HPI-403 E-7 2A C 2.00 CK SA O/C FFT-1 ROJ-23 FFT-2 ST-4T RFC-1 RFC-2 ISTC-5223

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 65 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: High Pressure Coolant Injeection DRAWING: .FM-25A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REQTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 23HPI-56 C-6 2 C 2.00 SK SA 0 FFT-1 ROJ-13 DIS-3 MST-059.45 RFC-t ROJ-13 DIS-3 MST-059.45 23HPI-61 B-7 2 C 16.00 CK SA 0 FFT-1 ROJ-15 PFT-3 ST-4M RFC-1 ROJ-15 VRR-04 DIS-3 MST-059.12 23HPI-62 F-4 2 C 4.00 CK SA 0 FFT-1 ROJ-16 VRR-04 DIS-3 MST-059.12 RFC-1 ROJ-16 DIS-3 MST-059.12 23HPI-65 C-6 2 NC 20.00 LK SA O/C FFT-1 ST-4N RFC-1 ROJ-12 RFC-3 ST-39B-X214 LKJ-6 ST-39B-X214 23MOV-14 F-3 2 B 10.00 GA MO 0 STO-1 ST-4N PIT-5 ST-41D 23MOV-15 F-8 1 A 10.00 GA " MO O/C STO-1 ST-4N STC-1 ST-4N PIT-5 ST-41K LKJ-6 ST-39B-X1I1 23MOV-l16 F-7 1 A 10.00 GA MO O/C STO-1 ST-4N STC- 1 ST-4N PIT-5 ST-41K LKJ-6 ST-39B-X1 1 23MOV-17 G-5 2 B 16.00 GA MO C STC-1 ST-4N PIT-5

• ST-41D 23MOV-19 F-6 1 A 14.00 GA MO O/C STO-1 ST-4N STC-1 ST-4N PIT-5 ST-41K LKJ-6 ST-39B-X9 23MOV-20 F-6 2 B 14.00 GA MO 0 STO-1 ST-4N PIT-5 ST-41 D 23MoV-21 G-6 2 B 8.00 GL MO C STC-1 ST-4N PIT-5 ST-41D 23MOV-25 F-5 2 B 4.00 GL MO O/C STO-1

• ST-4N STC-1 ST-4N PIT-5 ST-41D 23MOV-57 F-5 2 B 16.00 GA MO 0 STO-1 ST-4N PIT-5 ST-41D

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Pa ge.66 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: High Pressure Coolant Injeection DRAWING: FM-25A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REQFS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 23MOV-58 C-7 2 . B 16.00 GA MO O/C STO-1 ST-4N STC-1 ST-4N PIT-5 ST-41D 23MOV-59 E-7 2.00 GA MO PIT-5 ST-41D 23MOV-60 F-7 1,00 GL MO C STC-1 ST-4N PIT-5 ST-41K LKJ-6 ST-398-Xl 1 23SV-34 E-5 C 1.00 RV SA O RLF-8 MP-059.07 23S V-66 D-5 C 2.00 RV SA O

" RLF-8 MP-059.07 23Z-7 F-3 16.00 RD SA O RDT-10 PM TASK 23Z-8 F-3 2A 16.00 RD SA O RDT-10 PM TASK AUGMENTED

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 67 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Containment Atmosphere Dilution / Vent & Purge DRAWING: FM-18A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REOTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 27A0V-126A G-5 2A. B 1.00 GL AO 0 STO-1 ST-25BA AUGMNTEIU FSO-1 FAST ACTING VALVE PIT-5 ST-41D 27AOV- 126B F-5 2A B 1.00 GL AO 0 STO-1 ST-25BB AUGMENTED FSO-1 ST-25BB FAST ACTING VALVE PIT-5 ST-41D 27AOV-128A G-4 2A B 1.50 GL AO O/C STO-1 ST-25BA AUGMENTED STC-1 ST-25BA FAST ACTING VALVE FSO-1 ST-25BA PIT-5 ST-41D 27AOV- 128B E-4 2A B 1.50 GL AO. O/C STO-1 ST-25BB AUGMENTED STC-1 ST-25BB FAST ACTING VALVE FSO-1 ST-258B PIT-5 ST-41 D 27AOV-129A F-4 2A B 1.00 GL AO O/C STO-1 ST-25BA AUGMENTED STC-1 ST-25BA FAST ACTING VALVE FSO-1 ST-25BA PIT-5 ST-41D 27AOV-129B F-4 2A 1.00 GL AO O/C STO-1 ST-25BB AUGMENTED STC-1 ST-25BB FAST ACTING VALVE FSO-1 ST-25BB PIT-5 ST-41D 27CAD-19A G-6 2A C 2.00 CK SA 0 FFT-1 ST-25DA AUGMENTED RFC-1 Augmented Exclusion 27CAD-19B C-6 2A C 2.00 CK SA 0 FFT-1. ST-25DB AUGMENTED RFC-1 Augmented Exclusion 27RD-1A F-7 2A D 1.00 RD SA 0 RDT-10 PM TASK AUGMENTED 27RD-1B C-7 2A D 1.00 RD SA 0 RDT-10 PM TASK AUGMENTED 27RD-2A F-6 2A D 1.00 RD SA 0 RDT-10 PM TASK AUGMENTED 27RD-2B C-6 2A D 1.00 RD SA 0 RDT-10 PM TASK AUGMENTED 27SV-1 14A G-6 2A C 1.00 RV SA 0 RLF-8 MP-059.07 AUGMENTED 27SV-i 14B D-6 2A C 1.00 RV SA 0 RLF-8 MP-059.07 AUGMENTED

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 68 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Containment Atmosphere Dilution / Vent & Purge DRAWING: FM-18A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO:ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REQ"S CSJ/ROJ. REQUEST TEST PROCEDURE REMARKS 27SV-115A. G-4 2A C 0.50 RV SA 0 RLF-8 MP-059.07 AUGMENTED 27SV-115B E-4 2A C 0.50 RV SA 0 RLF-8 MP-059.07 AUGMENTED 27SV-118A G-5 2A C 0.50 RV SA 0 RLF-8 . MP-059.07 AUGMENTED 27SV-118B C-6 2A C 0.50 RV SA 0 RLF-8 MP-059.07 AUGMENTED 27SV-1 19A F-7 2A C 0.50 RV SA 0 RLF-8 MP-059.07 AUGMENTED 27SV-1 198 C-7 2A C 0.50 RV SA 0 RLF-8 MP-059.07 AUGMENTED 27SV-201 A F-3 2A C 1.00 RV SA 0 RLF-8 MP-059.07. AUGMENTED 27SV-201 B F-3 2A C 1.00 RV SA 0 RLF-8 MP-059.07 AUGMENTED 27SV-202 H-3 2A C 1.00 RV SA 0 RLF-8 MP-059.07 AUGMENTED

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 69 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Containment Atmosphere Dilution / Vent & Purge DRAWING: FM-18B DWG VALVE VALVE ACTUATOR SAFETY. TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REOTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 27AOV-101A C-6 2 A 20.00 BF AO O/C STO-1 ST-15G STC-1 ST-15G FSC-1 ST-15G PIT-5 ST-41K LKJ-6 ST-39B-X202B/G 27AOV-101B C-6 A 20.00 BF AO O/C STO-1 ST-15G STC- 1 ST-15G FSC-1 ST-i15G PIT-5 ST-41K LKJ-6 ST-39B-X202B/G 27AOV-1 11 C-2 2 A 24.00 BF AO C STC-1 CSJ-04 STC-2 ST-68 FSC-1 FSC-2 ST-68 PIT-5 ST-4 1K LKJ-6 ST-39B-X25/71 27AOV-1 12 C-3 2 A 24.00 BF AO C STC-1 CSJ-04 STC-2 ST-68 FSC-1 FSC-2 ST-68 PIT-5 ST-41K LKJ-6 ST-39B-X25/71 27AOV-113 D-8 A 24.00 BF AO C . STC-1 CSJ-04 STC-2 ST-68 FSC-1 FSC-2 ST-68 PIT-5 ST-4 1K LKJ-6 ST-39B-X26A/B 27AOV-1 14 D-8 2 A 24.00 BF AO C STC-1 CSJ-04 STC-2 ST-68 FSC-1 FSC-2

• ST-68 PIT-5 .ST-41K LKJ-6 ST-398-X26A/B 27AOV-115 C-2 2 20.00 BF AO C STC-1 CSJ-04 STC-2 ST-68 FSC-1 FSC-2 ST-68 PIT-5 ST-41K LKJ-6 ST-39B-X220 27AOV-1 16 C-3 2 A 20.00

• BF AO C STC-1 CSJ-04 STC-2 ST-68 FSC-1 FSC-2 ST-68 PIT-5 ST-41K LKJ-6 ST-390-X220

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 70 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Containment Atmosphere Dilution / Vent & Purge DRAWING: FM-18B DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REQOTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 27AOV-117 B-8 2 A 20.00 BF AO C STC-1 ST-lC FSC-1 ST-1C PIT-5 ST-41K LKJ-6 ST-39B-X205 27AOV-118 B-8 2 A 20.00 BF AO C STC-1 ST-1C FSC-1 ST-IC PIT-5 ST-41K" LKJ-6 ST-39B-X205 27AOV-131A C-4 2 A 1.50 GL AO O/C STO-1 ST-25BA STC-1 ST-25BA FSC-1 ST-25BA PIT-5 ST-41K LKJ-6 ST-39B-X25171 27AOV-131B C-3 2 A 1.50 GL AO O/C STO-1 ST-25BB STC-1 ST-25BB FSC-t ST-25B8 PIT-5. ST-41K LKJ-6 ST-39B-X25/71 27AOV-132A C-4 2 A 1.50 GL AO O/C STO-t ST-25BA STC-1 ST-25BA FSC-1 ST-25BA PIT-5 ST-41K LKJ-6 ST-39B-X220 27AOV-132B C-3 2 A 1.50 GL AO O/C STO-1 ST-25BB STC-' ST-25BB FSC-1 ST-25BB PIT-5 ST-41K LKJ-6 ST-39B-X220 27CAD-67 C-4 2 A/C 1.50 SK SA O/C FFT-1 ST-25DA RFC-1 ST-25BA LKJ-6 ST-39B-X220 27CAD-68 C-4 2 A/C 1.50 SK SA O/C FFT-1 ST-25DA RFC-1 ST-25BA LKJ-6 ST-39B-X25/71 27CAD-69 C-3 2 A/C 1.50 SK SA O/C FFT-1 ST-25DB RFC-1 ST-25BB LKJ-6 ST-39B-X25/71

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 71 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Containment Atmosphere Dilution / Vent & Purge DRAWING: FM-18B DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE(IN) TYPE TYPE FUNCTION REO'TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 27CAD-70 C-3 2 A/C 1.50 SK SA O/C FFT-1 ST-25DB RFC-1 ST-25BB LKJ-6 ST-39B-X220 27MOV-113 C-8 2 A 3.00 BF MO O/C STO-1 ST-lC STC-l1 ST-1C PIT-5 ST-41K LKJ-6 ST-39B-X26A/B 27MOV-117 B-8 2 A .3.00 BF MO 0/C STO-1 ST-1C STC-1 ST-IC PIT-5 ST-41K LKJ-6 ST-39B-X205 27MOV-120 H-8 2 B 12.00 BF MO 0 STO-I ST-lC STC-1 ST-iC PIT-5 ST-41D 27MOV-121 H-8 2 B 6.00 BF MO 0 STO-1 ST-25BB PIT-5 ST-41D 27MOV-122 C-8 2 A 3.00 GL MO O/C STO-1 ST-iC STC-I ST-1C PIT-5 ST-41K LKJ-6 ST-39B-X26A/B ST-iC 27MOV-123 B-8 2 A 3.00 GL MO O/C STO-1 STC-1 ST-IC PIT-5 ST-41K LKJ-6 ST-39B-X205 27SOV-125A F-5 2 1.00 GL SO C STC-1 ST-iC FSC-1 ST-iC FAST ACTING VALVE PIT-5 ST-41D LKJ-6 ST-39B-X52A 27SOV-125B F-4 2 1.00 GL SO C STC-1 ST-iC FSC-1 ST-iC FAST ACTING VALVE PIT-5 ST-41D LKJ-6 ST-39B-X55B ST-1C 27SOV-125C F-5 2 A 1.00 GL SO C STC-1 FSC-1 ST-1C FAST ACTING VALVE PIT-5 ST-41D LKJ-6 ST-39B-X52A

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 72 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Containment Atmosphere Dilution / Vent & Purge DRAWING: FM-18B DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE(IN) TYPE TYPE FUNCTION REOTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 27SOV-125D F-4 2 A 1.00 GL SO C STC-1 ST-iC FSC-I ST-IC FAST ACTING VALVE PIT-5 ST-41D LKJ2 6 ST-39B-X55B 27SOV-135A E-5 2 1.00 GL SO C STC-1 ST-iC FSC-I ST-iC FAST ACTING VALVE PIT-5 ST-41D LKJ-6 ST-39B-X31AD 27SOV-135B F-5 2 1.00 GL SO C STC-1 ST-iC FSC-1 ST-1C FAST ACTING VALVE PIT-5 ST-4 1D LKJ-6 ST-396-X31 BD 27SOV- I35C E-5 A 1.00 GL SO C STC-1 ST-iC FSC-I ST-iC FAST ACTING VALVE PIT-5 ST-41D LKJ-6 ST-39B-X31AD 27SOWVi35D F-5 2 A 1.00 GL SO C STC-1 ST-iC FSC-1 ST-iC FAST ACTING VALVE PIT-5 ST-41D LKJ-6 ST-398-X31 BD 27VB-1 C-6 A/C 30.00 CK SA O/C ETO-1 MME-1 ST-15J ETC-1 MME-1 ST-15J PIT-5 ST-41D LKO-5 LKO-3 ST-39E 27VB-2 C-6 A/C 30.00 CK SA O/C ETO-I MME-1 ST-15J ETC-1 MME-1 ST-15J PIT-5 ST-4 ID LKO-5 LKO-3 ST-39E 27VB-3 C-6 A/C 30.00 CK SA O/C ETO-1 MME-I ST-15J ETC-1 MME-1 ST-15J PIT-5 ST-41D LKO-5 LKO-3 ST-39E 27VB-4 C-6 A/C 30.00 CK SA O/C ETO- I MME-1 ST-15J ETC-1 MME-1 ST-15J PIT-5 ST-41D LKO-5 LKO-3 ST-39E 1,

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 73 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Containment Atmosphere Dilution / Vent & Purge. DRAWING: FM-18B DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REO'TS CSJ/ROJ REOUEST TEST PROCEDURE REMARKS 27VB-5 C-6 2 A/C 30.00 CK SA O/C ETO-1 MME-1 ST-15J ETC-I MME-I ST-i 5J PIT-5 ST-41D LKO-5 LKO-3 ST-39E 27VB-6 C-6 A/C 20.00 CK SA O/C ETO-l MME-1 ST-15G ETC-i MME-1 ST-15G PIT-5 ST-tIS5 LKJ-6 ST-39B-X202BfG 27VB-7 C-6 2 A/C 20.00 CK SA O/C ETO-t MME-1 ST-15G ETC- 1 MME-1 ST-15G PIT-5 ST-15G LKJ-6 ST-39B-X202B/G

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 74 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Containment Atmosphere Dilution / Vent & Purge DRAWING: FM-18D DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REO'TS CSJ/ROJ REGUEST TEST PROCEDURE REMARKS 27SOV-119E1 C-7 2 A 0.375 GL SO C STC-1 ST-IC FSC-1 ST-iC FAST ACTING VALVE PIT-5 ST-41 D LKJ-6 ST-398-X203A 27SOV-119E2 C-6 2 A 0.375 GL SO C STC-1 ST-1C FSC-1 ST-iC FAST ACTING VALVE PIT-5 ST-41D LKJ-6 ST-39B-X203A 27SOV- 119Fi D-4 2 A 0.375 GL SO C STC-1 ST-iC.

FSC-1 ST-1C FAST ACTING VALVE PIT-5 ST-41D LKJ-6 ST-39B-X2116 27SOV-119F2 C-5 2 A 0.375 GL SO C* STC-1 ST-IC FSC-l. ST-iC FAST ACTING VALVE PIT-5 ST-41D LKJ-6 ST-39B-X216 27SOV-120E1 F-6 2 A 0.375 GL SO C STC-1 ST-iC FSC-1 ST-iC FAST ACTING VALVE PIT-5 ST-41D LKJ-6 ST-398-X202B/G 27SOV-120E2 F-6 2 A 0.375 GL SO C STC-1 ST-iC FSC-1 ST-1C FAST ACTING VALVE PIT-5 ST-41D LKJ-6 ST-39B-X202B/G 27SOV-120F1 F-4 2 A 0.375 GL SO C STC-i ST-iC FSC-1 ST-iC FAST ACTING VALVE PIT-5 ST-41D LKJ-6 ST-39B-X58C 27SOV-120F2 F-4 2 A 0.375 GL SO C STC-1 ST-iC FSC-1 ST-iC FAST ACTING VALVE PIT-5 ST-41D LKJ-6 ST-39E-X58C 27SOV-122E1 F-6 2 A 0.375 GL SO C STC-1 ST-iC FSC-1 ST-iC FAST ACTING VALVE PIT-5 ST-41D LKJ-6 ST-39B-X26A/B

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 75 of 121 INSERVICE TESTING PROGRAM VALVE TABLE DRAWING: FM-18D SYSTEM: Containment Atmosphere Dilution / Vent & Purge VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST DWG CATEGORY SIZE(IN) TYPE TYPE FUNCTION REQTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS VALVE ID CO-ORD CLASS 2 A 0.375 GL SO C STC-1 ST-tC 27SOV-122E2 F-6 FSC-1 ST-lC FAST ACTING VALVE PIT-5 ST-41D LKJ-6 ST-39B-X26A/B GL SO C STC-1 ST-lC 27SOV-122F1 G-4 2 A 0.375 -. ST-lC FAST ACTING VALVE FSC-l ST-41D PIT-5 ST-39B-X58B LKJ-6 ST-IC 27SOV-122F2 G-4 2 A 0.375 GL SO C STC-1 ST-lC FAST ACTING VALVE FSC-1 PIT-5 ST-41D ST-39B-X58B.

LKJ-6 ST-1C 27SOV-123E1 E-6 2 A 0.375 GL SO C STC-1 FSC-1 ST-lC FAST ACTING VALVE ST-41 D PIT-5 LKJ-6 ST-39B-X59 ST-IC 27SOV-123E2 E-6 2 A 0.375 GL SO C STC-I ST-tC FAST ACTING VALVE FSC-1 ST-41D PIT-5 ST-39B-X59 LKJ-6 ST-iC 27SOV-123FI F-4 2 A 0.375 GL SO C STC-1 FSC-1 ST-IC FAST ACTING VALVE PIT-5 ST-41 D LKJ-6 ST-39B-X58D ST-lC 27SOV-123F2 F-4 2 A 0.375 GL SO C STC-1 ST-lC FAST ACTING VALVE FSC-t ST-41D PIT-5 ST-39B-X58D LKJ-6 ST-IC 27SOV-124E1 C-4 2 A 1.00 GL SO C STC-1 FSC-1 ST-1C FAST ACTING VALVE PIT-5 ST-41D LKJ-6 ST-39B-X203B ST-1C 27SOV- 124E2 C-4 2 A 1.00 GL SO C STC-I FSC-1 ST-IC FAST ACTING VALVE ST-41 D PIT-5 LKJ-6 ST-39B-X2038

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 76 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Containment Atmosphere Dilution / Vent & Purge DRAWING: FM-18D DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST.

VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REQTS CSJIROJ REQUEST TEST PROCEDURE REMARKS 27SOV-124F1 C-4 2 A 0.375 GL SO C STC- 1

• ST-1C FSC- 1 ST-1C FAST ACTING VALVE PIT-5 ST-4 1D LKJ-6 ST-39B-X203B 27SOV-124F2 C-4 2 A 0.375 GL SO C STC-1 ST-IC FSC-1 ST-1C FAST ACTING VALVE PIT-5

• ST-41D LKJ-6 ST-39B-X203B

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 77 of 121 INSERVICE TESTING PROGRAM VALVE TABLE

. SYSTEM: Containment Atmosphere Dilution / Vent & Purge DRAWING: FM-39C DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REO'TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 27SOV-141 E-6 2 . A 1.00 GA SO O/C STO-1

• ST-1C STC- I ST-IC FAST ACTING VALVE FSO-1 ST-1C PIT-5 ST-41K LKJ-6 ST-39B-X22 27SOV-145 G-5 A 1.00 GA SO O/C STO-1 ST-lC STC-1 ST-1C FAST ACTING VALVE FSO-1 ST-iC PIT-5 ST-41K LKJ-6 ST-39B-X57C

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 78 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Main Steam DRAWING: FM-29A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE(IN) TYPE TYPE FUNCTION REQOTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 29AOV-BOA E-5 1 A 24.00 GL AO C STC-1 CSJ-08 STC-2 ST-1B FSC-1 ROJ-19 FSC-3 S-T-6BB PIT-5 ST-41K LKJ-6 ST-39B-X7A 29AOV-80B D-5 1 A 24.00. GL AO C STC-1 CSJ-08 STC-2 ST-1B FSC-1 ROJ-19 FSC-3 ST-68B PIT-5 ST-41K LKJ-6 ST-39B-X7B 29AOV-80C D-5 1 A 24.00 GL AO C STC-1 CSJ-08 STC-2 ST-AB FSC-1 ROJ-19 FSC-3 ST-68B PIT-5 ST-41K LKJ-6 ST-39B-X7C 29AOV-80D D-5 1 A 24.00 GL AO C STC-1 CSJ-08 STC-2 ST-AB FSC-1 ROJ-19 FSC-3 ST-68B PIT-5 ST-41K LKJ-6 ST-39B-X7D 29AOV-86A G-4 1 A 24.00 GL AO C STC-1 CSJ-08 STC-2 ST-1B FSC-1 CSJ-05 FSC-2 ST-68 PIT-5 ST-41K LKJ-6 ST-39B-X7A 29AOV-86B F-4 1 A 24.00 GL AO C STC.1 CSJ-08 STC-2 ST-AB FSC-1 CSJ-05 FSC-2 ST-68 PIT-5 ST-41K LKJ-6 ST-39B-X7B 29AOV-86C E-4 1 A 24.00 GL AO C STC-1 CSJ-08 STC-2 ST-lB FSC-1 CSJ-05 FSC-2 ST-68 PIT-5 ST-41K LKJ-6 ST-39B-X7C 29AOV-86D D-4 1 A 24.00 GL AO C STC-1 CSJ-08 STC-2 ST-AB FSC-1 CSJ-05 FSC-2 ST-68 PIT-5 ST-41 K LKJ-6 ST-39B-X7D

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 79 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Main Steam DRAWING: FM-29A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REQOS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 29EFV-30A F-5. 1 A/C 1.00 BK SA C ETC-I ROJ-01 VRR-03 ETC-3 ISP-1. VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 29EFV-30B F-5 A/C 1.00. BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 29EFV-30C F-5 A/C 1.00 BK SA C ETC-1 ROJ-oi VRR-03 ETC-3 ISP-I VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 29EFV-30D F-5 A/C 1.00 BK SA C ETC-1 ROJ-0i VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-i 29EFV-34A F-8 A/C 1.00 BK SA C ETC-i ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 29EFV-34B F-8 A/C 1.00 BK SA C ETC-i ROJ-0i VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 29EFV-34C F-8 A/C 1.00 BK SA C ETC-1 ROJ-o0 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 29EFV-34D F-8 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 29EFV-53A E-8 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5. LKO-3 ISP'I 29EFV-53B E-8 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 29EF V-SaC E-8 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 29EF V-53D E-8 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 29EFV-54A E-5 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 29EFV-54B E-5 A/C 1.00 BK SA C ETC-1 ROJ-Oi VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1 29EFV-54C E-5 A/C 1.00 BK SA C ETC-1 ROJ-01 VRR-03 ETC-3 ISP-1 VALVE ISOLATES ON EXCESS FLOW LKO-5 LKO-3 ISP-1

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 80 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Main Steam DRAWING:. FM-29A VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE .TEST DWG VIAI VJF Ill CLASS CATEGORY SIZE (IN) TYPE FUNCTION RE07S CSJ/ROJ REQUEST TEST PROCEDURE REMARKS VALVE ID CO-ORD 29EFV-54D E-5 1 A/C 1.UU BI\ SA' C. ECI-1. R 1 -

LKO-5 LKO-3 ISP-1 2A B 1.00 GL MO 0 STO-1 ST-1MA AUGMENTED.

29MOV-200A C-3 PIT-5 ST-41D ST-1MB *AUGMENTED 29MOV-200B B-3 2A B 1.00 GL MO 0 STO- I PIT-5 ST-41 D C-3 2A B 1.00 GL MO 0/C STO-1 ST-1MA AUGMENTED 29MoV-201 A STC-1 ST-1MA PIT-5 ST-41D 2A B 1.00 GL G MO O/C STO-1 ST-1MB AUGMENTED 29MOV-201B B-3 STC-1 ST-1MB PIT-5 ST-41D 29MOV-202A C-3 2A B 1.00 GL MO O/C STO-1 ST-1MA AUGMENTED STC-1 ST-1MA ST-41D PIT-5 29MOV-202B B-3 2A B 1.00 GL MO O/C STO-1 ST-1MB AUGMENTED STC- 1 ST- 1MB PIT-5 ST-41D 29MOV-203A H-3 2A B 1.00 GL MO 0 STO-1 CSJ-06 STO-2 ST-68 AUGMENTED PIT-5 ST-41D 29MOV-203B H-3 2A B 1.00 GL MO 0 STO-1 CSJ-06 STO-2 ST-68 AUGMENTED PIT-5 ST-41D 29MOV-204A C-3 2A B 1.00. GL MO C STC-1 ST-1MA. AUGMENTED PIT-5 ST-41D 29MOV-204B B-3 2A B 1.00 GL MO C STC-1 ST-IMB AUGMENTED PIT-5 ST-4 1D A MO STC-1 ST-1C 29MOV-74 C-6 3.00 GA C PIT-5 ST-41K LKJ-6 ST-39B-X8.

29MOV-77 C-5 A 3.00 GA MO C STC-1 ST-iC PIT-5 ST-41K LKJ-6 ST-39B-X8

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 81 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Feedwater DRAWING: FM-34A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE(IN) TYPE TYPE FUNCTION REQrOS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 34FWS-28A E-7 1 A/C 18.00 CK SA C RFC-1 ROJ-20 RFC-3 ST-39B-X9 FFT-1 ISTC-3550 LKJ-6 LKJ-3. ST-39B-X9 34FWS-28B F-7 1 A/C 18.00 CK SA C RFC-1 ROJ-20 RFC-3 ST-39E-X9 FFT-1 ISTC-3550 LKJ-6 LKJ-3 ST-39B-X9 34NRV-111A E-7 1 A/C 18.00 NK SA, AO C RFC-1 ROJ-30 RFC-2 ST-39B-X9 FFT- 1 ISTC-3550 LKJ-6 ST-39B-X9 PIT-3 ST-41K 34NRV-111B F-7 1 A/C 18.00 NK SA, AO C RFC-1 ROJ-30 RFC-2 ST-39B-X9 FFT-1 ISTC-3550 LKJ-6 ST-39B-X9 PIT-3 ST-41K'

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page.82 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Instrumenl Air DRAWING: FM-39C DWG VALVE VALVE ACTUATOR SAFTEY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REOTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 391AS-22 E-5 2A

• A/C 2.00 CK SA O/C FFT-1 ROJ-21 FFT-3 ST-68B AUGMENTED RFC-1 ST-25C LKJ-6 ST-39B-X22 391AS-29 F-3 2A A/C 1.00 CK SA O/C FFT-1 ROJ-21 FFT-3 ST-68B AUGMENTED RFC-1 ST-25C LKJ-6 ST-39B-X57C

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 83of121 INSERVICE TESTING PROGRAM VALVE TABLE.

SYSTEM: Service Water DRAWING: FB-35E DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REQ'TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 46(70)ESW-101 G-6 3 B 4.00 GA MA 0 ETO-5 ST-8S 46(70)ESW-102 C-6 3 B 4.00 GA MA 0 . ETO-5 ST-8S 46(70)ESW-103 F-6 3 B 4.00 GA MA 0 ETO-5 ST-8S 46(70)ESW-104 C-6 3 B 4.00 GA MA 0 ETO-5 ST-8S 46(70)SWS-101 H-8 3 C 6.00 CK SA C RFC-1 ST-80 FFT-1 ISTC-3550 46(70)SWS-102 H-8 3 C 6.00 CK SA C RFC-1 ST-8Q FFT-1 ISTC-3550 46(70)SWS-13 H-4 3 B 6.00 GL MA C ETC-5 ST-8VA 46(70)SWS-14 E-4 3 B 6.00 GL MA C ETC-5 ST-8VB 70TCV-120A F-7 B 2.00 3W AO 0 STO-1 VRR-05 OMN-8 FSO-1 ST-41 FA ETO-1 ST-41 FA 70TCV-120B C-6 2.00 3W AO 0 STO-1 VRR-05 OMN-8 FSO-1 ST-41FB ETO-1 ST-41FB 70TCV-121A F-6 2.00 3W AO 0 STO-1 VRR-05 OMN-8 FSO-1 ST-41 FA ETO-1 ST-41FA 70TCV-121B C-7 2.00 3W  : AO 0 STO-1 VRR-05 OMN-8 FSO-1. ST-41 FB ETO-1 ST-41FB 70WAC-12A F-6 3 B 4.00 GA MA C ETC-5 ST-8VA 70WAC-12B C-6 3 B 4.00 GA MA C ETC-5 ST-BVB 70WAC-5A F-2 3 B 4.00 GA MA C ETC-5 ST-8VA 70WAC-5B D-2 3 B 4.00 GA MA C ETC-5 ST-8VB

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 84 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Service Water DRAWING: FM-46A DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REQ'TS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 46SWS-67A B-6. 3 C 3.00 CK SA C RFC-1 ST-8Q FFT-1 ISTC-3550 46SWS-67B B-7 3 C 3.00 CK SA C RFC-1 ST-80 FFT-1 ISTC-3550 46SWS-68 B-6 3 C 3.00 CK SA. C RFC-1 ST-80 FFT-1 ISTC-3550

.46SWS-69 B-8 3 C 3.00 CK SA C RFC-1 ST-80 FFT-1 ISTC-3550 67PCV-101 D-2 3 B 2.50 GL AO .0 STO-1 VRR-05 OMN-8 FSO-1 ST-41 FA ETO-1 ST-41 FA

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 85 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Service Water DRAWING: FM-46B DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST VALVE ID CO-ORD CLASS CATEGORY SIZE(IN) TYPE TYPE FUNCTION REOTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 46ESW-1A E-7 3. C 12.00 CK SA 0 FFT-1 ST-80 RFC-1 NIT-4 46ESW-1B D-7 3 C 12.00 CK SA 0 FFT-1 ST-80 RFC-1 NIT-4 46ESW-7A E-5 3 C 6.00 . CK SA 0 FFT-1 ST-8Q RFC-1 NIT-4 46ESW-7B E-5 3 C 6.00 CK SA 0 FFT-I ST-80 RFC-1 NIT-4 46ESW-9A E-4 3 C 8.00 CK SA 0 FFT-1 ST-8O RFC-1 NIT-4 46ESW-9B D-4 3 C 8.00 CK SA 0 FFT-1 ST-8O RFC-1 NIT-4 46MOV-101A E-6 3 B 10.00 GA MO 0 STO-1 ST-8Q PIT-5 ST-41D 46MOV-101B C-6 3 B 10.00 GA MO 0 STO-1 ST-80 PIT-5 ST-41D 46MOV-102A E-6 3 B 8.00 GA MO C STC-1 ST-8O PIT-5 ST-41D 46MOV-102B D-6 3 B 8.00 GA MO C STC-1 ST-8Q PIT-5. ST-41D 46RV-112A G-7 3 C 6.00 RL SA 0 RLF-8 MP-059.07 46RV-112B F-6 3 C 6.00 RL SA 0 RLF-8 MP-059.07 46RV-1 12C F-7 3 C 6.00 RL SA 0 RLF-8 MP-059.07 46RV-112D G-6 3 C 6.00 RL SA 0 RLF-8 MP-059.07

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Page 86 of 121 INSERVICE TESTING PROGRAM VALVE TABLE SYSTEM: Service Water DRAWING: FB-10H DWG VALVE VALVE ACTUATOR SAFETY TEST RELIEF ALTERNATE TEST C *-ORD CLASS CATEGORY SIZE (IN) TYPE TYPE FUNCTION REQOTS CSJ/ROJ REQUEST TEST PROCEDURE REMARKS 46S WS-60A C-5 3 C 4.00 CK SA C RFC-1 ST-80 FFT-1 - ISTC-3550 46SWS-60B C-5 3 C 4.00 CK SA C RFC-1 ST-80 FFT-1 ISTC-3550

JAMES A. FITZPATRICK NUCLEAR POWER PLANT.

JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPENDIX B Cold Shutdown Justifications CSJ-01 SYSTEM: REACTOR WATER RECIRCULATION (RWR)

COMPONENTS: 02MOV-53A, B CATEGORY: B SAFETY FUNCTION: These valves close, on low reactor pressure to isolate the faulted loop coincident with initiation of the RHR System in the LPCI mode, to prevent diversion of LPCI flow.

JUSTIFICATION: To exercise these valves, the respective recirculation pump must be secured.

Securing either pump (single loop operation) is limited by Technical Specification requirements. Single loop operation also requires a reduction in power. This hardship is not warranted since there is no compensating increase in the level of quality and safety.

ALTERNATE TEST: These valves will be stroke time tested during cold shutdown when Reactor Water Recirculation Pumps can be secured in accordance with ISTC-3521(f) and (g).

CSJ-02 SYSTEM: REACTOR BUILDING CLOSED LOOP COOLING (RBC)

COMPONENTS: 15AOV-130A, B; 15AOV-131A, B; 15AOV-132A, B; 15AOV-133A, 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, Drywell equipment drain sump cooler, cooling water to the recirculation pump motor and seal coolers. 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, or damage to the recirculation pumps.

ALTERNATE TEST: These valves will be stroke time tested during cold shutdowns in accordance with ISTC-3521(f) and (g).

.Rev. No. 10 Page 87 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES CSJ-03 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)

COMPONENTS: 23HPI-18 CATEGORY: C SAFETY FUNCTION: This valve opens to provide a flowpath for the HPCI system injection to 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, this valve cannot be exercised during normal plant operation.

ALTERNATE TEST: This valve will be mechanical exercise tested during cold shutdown in accordance with ISTC-3522(d) and (e).

CSJ-04 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.

ALTERNATE TEST: These valves will be stroke time tested during cold shutdown in accordance with ISTC-3521(f) and (g).

Rev. No. 10 Page 88 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES CSJ-05 SYSTEM: MAIN STEAM (MSS)

COMPONENTS: 29AOV-86A, B, C, D CATEGORY: A SAFETY FUNCTION: These valves close to provide containment isolation.

JUSTIFICATION: Performance of the fail close test for the MSIVs requires entry into the Steam Tunnel. This cannot be done during normal operation.

ALTERNATE TEST: These valves will be fail safe tested during cold shutdown in accordance with ISTC-3521(f) and (g).

CSJ-06 SYSTEM:. MAIN STEAM (MSS)

COMPONENTS: 29MOV-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.

ALTERNATE TEST: These valves will be stroke time tested during cold shutdown in accordance with ISTC-3521(f) and.(g).

Rev. No. 10 Page 89 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES CSJ-07 SYSTEM: REACTOR WATER CLEANUP COMPONENTS: 12MOV-15, 12MOV-18, 12MOV-69 CATEGORY: A SAFETY FUNCTION: These valves close to provide containment isolation. The valves also close*

on, low reactor water level or high RWCU ambient temperature to protect the core in case of a break in the RWCU piping and on SLC actuation to prevent removal of boron.

JUSTIFICATION: Cycling these valves during operation has significant negative effects to reactor water chemistry that could result in power reduction or plant shutdown. Radiation exposure received during system alterations to perform the testing during operation has also resulted in excessive personnel exposure. Cycling the system during operation causes thermal transients that places undue stress on the piping and pumps. Testing of these valves during operation subjects the system to unacceptable chemical and thermal transients and excessive personnel radiation exposure. As discussed in NUREG-1482 Paragraph 2.4.5 these negative effects place impractical conditions on the system and justify cold shutdown deferral.

ALTERNATE TEST: These valves will be stroke time tested during cold shutdown in accordance with ISTC-3521(f) and (g).

CSJ-08 SYSTEM: MAIN STEAM (MSS)

COMPONENTS: 29AOV-80A, B, C, D; 29AOV-86A, B, C, D CATEGORY: A SAFETY FUNCTION: These valves close to provide containment isolation.

JUSTIFICATION: Full stroke testing of MSIV's at power places the plant in an abnormal operating condition and introduces an unnecessary challenge to plant equipment. This is in view of industry experience, both from an operational standpoint, and from the standpoint that stroking MSIV's at power is a contributor to valve seat degradation and resultant degraded containment isolation capability. (Ref: NUREG- 1482)

ALTERNATE TEST: Stroke timing during cold shutdown in accordance with ISTC-3521(f) and (g).is acceptable since valve actuator is designed to limit stroke time regardless of system dynamics present at time of testing.

Rev. No. 10 Page 90 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES CSJ-09 SYSTEM: REACTOR CORE ISOLATION COOLING (RCIC)

COMPONENTS: 13RCIC-22 CATEGORY: C SAFETY FUNCTION: This valve opens to provide a flow path for the RCIC system injection to the reactor vessel.

JUSTIFICATION: With the reactor at operating pressure, the RCIC pump can develop sufficient discharge pressure to open this valve, however RCIC 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, this valve cannot be exercised during normal plant operation.

ALTERNATE TEST: This valve will be mechanical exercise tested during cold shutdown in accordance with ISTC-3522(d) and (e).

Rev. No. 10 Page 91 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT.

JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES Refueling Outage Justifications ROJ-01 SYSTEM: VARIOUS COMPONENTS: Excess Flow Check Valves CATEGORY: A/C (Listed Below)

SAFETY FUNCTION: These valves close to isolate the respective instrument lines in the event of a pipe break downstream of the valves.

JUSTIFICATION: Exercising these valves requires isolation of their associated safety-related instrument, which could place the plant in an unsafe condition. In addition, the induced hydraulic transients resulting from establishing flow and subsequent valve closure would. most likely result in an engineered safety feature actuation. During such testing, radiation doses to test personnel would be high due to the location of these valves and reactor 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 ISTC-3522(c) and (f).

EXCESS FLOW CHECK VALVES 02-2EFV-PS-128A,B 02-3EFV-25 02-2EFV-PT-24A,B 02-3EFV-31A,B,C,D 02-2EFV-PT-25A,B 02-3EFV-31E,F,G,H 02-2EFV1-DPT- 111A,B 02-3EFV-3 1J,K,L,M 02-2EFV 1-FT-11 OA,C,E,G 02-3EFV-31N,P,R,S 02-2EFV2-DPT-1 1 1A,B 02-3EFV-33 02-2EFV2-FT- 1 10A,C,E,G 13EFV-O1A,B 02-3EFV-11 13EFV-02A,B 02-3-EFV- 13A,B 14EFV-31A,B 02-3EFV- 15A,B 23EFV-O1A,B 02-3EFV-15N 23EFV-02A,B 02-3EFV- 17A,B 29EFV-30A,B,C,D 02-3EFV- 19A,B 29EFV-34A,B,C,D 02-3EFV-21A,B,C,D 29EFV-53A,B,C,D 02-3EFV-23A,B,C,D 29EFV-54A,B,C,D 02-3EFVr23 Rev. No. 10 Page 92 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES ROJ-02 SYSTEM: REACTOR WATER RECIRCULATION (RWR)

COMPONENTS: 02-2RWR-13A, B CATEGORY: A/C SAFETY FUNCTION: These recirculation pump seal water injection valves close to provide containment isolation.

JUSTIFICATION: Exercising these valves during normal operations or cold shutdown requires securing the Recirculation pumps and entering containment to check the valves closed by using a back-leakage test. Testing during operations is therefore impossible. Testing during cold shutdown by performing back-leakage tests would require extensive time for test equipment set-up and place an undue burden on the plant staff. In addition, entry into the containment may be prohibited if the drywell remains inerted.

Back-leakage testing and leakrate testing will be performed during each refueling outage in accordance with ISTC-3522(c) and (f).

ROJ-03 SYSTEM: REACTOR WATER RECIRCULATION (RWR)

COMPONENTS: 02-2RWR-41A,B CATEGORY: A/C SAFETY FUNCTION: These recirculation pump seal purge check valves close to provide containment isolation.

JUSTIFICATION: Closing these valves any timeReactor 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.

ALTERNATE TEST: Back-leakage testing and leakrate testing will be performed during each refueling outage in accordance with ISTC-3522(c) and (f).

Rev. No. 10 Page 93 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES ROJ-04 SYSTEM: AUTOMATIC DEPRESSURIZATION (ADS)

COMPONENTS: 02RV- 1 through 02RV- I1 02VB-1 through 02VB-1 1 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..

ALTERNATE TEST: Testing will be performed during each refueling outage in accordance with ISTC-3522(c) and (f).

ROJ-05 SYSTEM: RESIDUAL HEAT REMOVAL (RHR)

COMPONENTS: 1ORHR-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.

JUSTIFICATION: These valves, are exercised open every three months by flow during pump testing. However, quantitative flow measurements as a means of verifying these valves open has been determined to be impractical.

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. 10 Page 94 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES ROJ-05 (Continued) causing the ultrasonic flow meter to go into fault. Attempts have been made at different locations and with different size transducers, and faults still occur.

This test method requires the RHR pumps to be operated repeatedly (three to four times) at minimum flow conditions for the maximum time period allowed by procedure. Running at this condition is undesirable, particularly for a test method that frequently does not yield meaningful results. NRC Information Notice 89-08 documented concerns about pump damage by operating at low flow conditions. When this test is performed with no flow measurements being taken, the time spent at minimum pump flow is short.

In addition, this testing must be performed in a radiation area, which has caused increased exposure to personnel while multiple test attempts and transducer repositioning are accomplished. It is concluded that continued efforts with this method are not practical.

Attempts were made to distinguish the check valve opening impact on the valve bonnet using a seismic vibration probe. Meaningful results could not be obtained again due to the high background noise and vibration associated with a pump start at minimum flow.

The method of using process flow and pressure instrumentation in the main line to infer the flow in the minimum flow line was investigated. 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.

ALTERNATE TEST: In accordance with Generic Letter 89-04, Position 2, during 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 eight (8) years.

Rev. No. 10 Page 95 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES ROJ-06 SYSTEM: RESIDUAL HEAT REMOVAL (RHR)

COMPONENTS: 1ORHR-95A,B CATEGORY: C SAFETY FUNCTION: These valves close to prevent reverse flow from the torus.

JUSTIFICATION: These are simple check valves with no means of determining disc position without performing a back leakage test. Performing such a test during plant operations would require setting up a test rig and performing a hydrostatic test.

As. discussed in NUREG 1482, 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.

ALTERNATE TEST: These valves will be verified to close each refueling outage during a hydrostatic leak rate test in accordance with ISTC-3522(c) and (f).

ROJ-07 SYSTEM: STANDBY LIQUID CONTROL (SLC)

COMPONENTS: 1lSLC-16 & 11SLC-17 CATEGORY: A/C SAFETY FUNCTION: These. valves prohibit backflow from the reactor vessel to the SLC System and provide for containment isolation. They open to permit SLC System flow to the reactor vessel.

JUSTIFICATION: Full or partial-stroke exercising these valves requires that flow be established through the subject check valves. The only practical means of initiating flow through these valves requires actuation of the SLC system and pumping from the SLC Tank to the reactor vessel. During normal plant operation, this would introduce boron into the reactor 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.

ALTERNATE TEST: 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. 10 Page 96 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES 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 considered to be undesirable from the aspect of potential damage to RCIC system components should the scaffold be subjected to structural failure.

Based on the foregoing discussion, testing of these valves during plant operation at power is considered to be impractical. During cold shutdowns, erection of the scaffold in addition to other activities related to test performance would place an extreme burden on the plant staff and would likely result in unwarranted extensions to all forced outages with the added negative impact on plant performance and availability.

ALTERNATE TEST: These valves will be verified to close by performing a back-leakage test at each refueling outage in accordance with ISTC-3522(c) and (f).

Rev. No. 10 Page 97 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES 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 overfilling the reactor vessel and flooding the main steam lines. This could adversely affect the performance of the main steam safety/relief valves (SRVs) since a contributing factor to the historically poor performance of the SRVs is water contamination of the operators.

ALTERNATE TEST: During cold shutdowns, each of the valves will be exercised using the installed air operators (considered a partial-stroke). This test satisfies the exercising of both safety positions.

Each. of the valves will be full-stroked exercised during each refuel outage in accordance with ISTC-3522(c) and (f) by injecting full accident flow into the reactor vessel. The closed position is leak tested every refuel outage per ISTC-3630(a). This position complies with the guidance of NUREG-1482, Section 4.1.6.

Rev. No. 10 Page 98 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES ROJ-10 SYSTEM: -CORE SPRAY (CSP)

COMPONENTS: 14CSP-62A,B CATEGORY: C SAFETY FUNCTION: These valves close to prevent reverse flow from the torus.

JUSTIFICATION: There are no position indicators or other means to verify closure of these valves.

As a result, valve closure must be verified by back-leakage testing. Performing such a test during plant operations would require setting up for and performing a hydrostatic test. As discussed in NUREG 1482, section 4.1.4, the NRC has 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.

ALTERNATE TEST: These valves will be verified close each refueling outage in accordance with ISTC-3522(c) and (f) during a hydrostatic leak rate test.

ROJ- 11 SYSTEM: REACTOR BUILDING CLOSED LOOP COOLING (RBC)

COMPONENTS: 15RBC-214 CATEGORY: C SAFETY FUNCTION: This valve closes to prevent flow diversion when the Emergency Service Water system is supplying cooling water to RBC heat loads.

JUSTIFICATION: There is no provision on this valve that provides position indication of the disc.

There are no test taps and block valves to enable a back-leakage test to verify closure.

ALTERNATE TEST: ISTC-5221(c) allows disassembly each refueling outage to verify operability as an alternative to quarterly testing.

Rev. No. 10 Page 99 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES ROJ-12 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)

COMPONENTS: 23HPI-12 and 23HPI-65 CATEGORY: A/C SAFETY FUNCTION: These valves close to provide containment isolation.

JUSTIFICATION: There is no provision on either of these valves that provides position indication of the disc. As a result, valve closure must be verified by back-leakage testing.

In order to verify valve closure by the back-leakage technique, the HPCI exhaust line must be isolated for the duration of the test causing the HPCI system to be inoperable. The potential safety impact of voluntarily placing the HPCI system in an inoperable status during plant operation at power is considered to be imprudent and unwarranted in relation to any apparent gain in system reliability derived from the closure verification. In addition, the valves are located approximately twenty (20) feet from the floor necessitating erection of a large scaffold in the vicinity of the HPCI pump. This also is considered to be undesirable from the aspect of potential damage to HPCi system components should the scaffold be subjected to structural failure.

Based on the foregoing discussion, testing of these valves during plant operation at power is considered to be impractical. During cold shutdowns, erection of the scaffold in addition to other activities related to test performance would place an extreme burden on the plant staff and would likely result in unwarranted extensions to all forced outages with the added negative impact *on plant performance and availability.

ALTERNATE TEST: These valves will be verified to close by performing a back-leakage test at each refueling outage in accordance with ISTC-3522(c) and (f).

Rev. No. 10 Page 100 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES ROJ-13 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)

COMPONENTS: 231-IPI-13 and 23HPI-56 CATEGORY:C SAFETY FUNCTION: These valves open to permit HPCI turbine condensate to drain to the Torus and close on cessation of flow.

JUSTIFICATION: There are no. means for exercising these valves to the open position where positive indication of acceptable. valve performance is verified. There is no provision that provides position indication of the disc. There are no test taps and block valves to enable a back-leakage test to verify closure.

ALTERNATE TEST: ISTC-5221(c) allows disassembly each refueling outage to verify operability as an alternative to quarterly testing.

ROJ-14 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 torus 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.

ALTERNATE TEST: ISTC-5221(c) allows disassembly each refueling outage to verify operability as an alternative to quarterly testing. Valve Relief Request VRR-04 allows this activity to be performed during on-line system outages.

Rev. No. 10 Page 101 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES 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. It closes on cessation of flow.

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. 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.

ALTERNATE TEST: ISTC-5221(c) allows disassembly each refueling outage to verify operability as an alternative to quarterly testing. Valve Relief Request VRR-04 allows this activity to be performed during on-line system outages. 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. It closes on cessation of flow.

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. 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.

ALTERNATE TEST: ISTC-5221(c) allows disassembly each refueling outage to verify operability as an alternative to quarterly testing. Valve Relief Request VRR-04 allows this activity to be performed during on-line system outages.

Rev. No. 10 Page 102 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES ROJ-17 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)

COMPONENTS: 23HPI- 130 CATEGORY: C SAFETY FUNCTION: This valve opens to provide a flowpath for cooling water circulation through the HPCI turbinelube oil cooler and closes to prevent flow diversion.

JUSTIFICATION: This valve has no means of determining disc position or flowrate and, thus there is no mechanism for verifying full accident flow. In addition, there are no test taps and block valves to enable a back-leakage test to verify closure.

ALTERNATE TEST: ISTC-5221(c) allows disassembly each refueling outage to verify operability as an alternative to quarterly testing. Valve Relief Request VRR-04 allows this activity to be performed during on-line system outages. In addition, this valve will be partial-flow tested once per operating cycle.

ROJ-18 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)

COMPONENTS: 23HPI-131 CATEGORY: C SAFETY FUNCTION: *Thisvalve 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.

ALTERNATE TEST: ISTC-5221(c) allows disassembly each refueling outage to verify operability as an alternative to quarterly testing. Valve Relief Request VRR-04 allows this activity to be performed during on-line system outages..

Rev. No. 10 Page 103 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES ROJ-19 SYSTEM: MAIN STEAM (MSS)

COMPONENTS: 29AOV-80A,B,C,D CATEGORY: A SAFETY FUNCTION: These valves are normally open to provide steam to the main turbine generator and auxiliaries, and they close to isolate steam flow and for containment isolation.

JUSTIFICATION: Fail safe exercising these valves requires local manipulation of valves located inside containment. During plant operation at power, and on occasion while in cold shutdown, the containment atmosphere is maintained in a nitrogen-inerted condition. During such periods, entry into the containment is not practical due to personnel safety concerns.

ALTERNATE TEST: These valves will be verified to fail safe close at each refueling. outage in accordance with ISTC-3521(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 feedwater 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 valves would require up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and demand significant staff resources. Also, entry into the containment at cold shutdown with the containment inerted is a personnel safety concern.

ALTERNATE TEST: Closure of these valves will be demonstrated during each refuel outage in accordance with ISTC-3522(c) and (f) by conducting a back-leakage test.

Rev. No. 10 1Page 104 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES 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 MS1V testing. During plant operation at power, this is impractical since closure of the MSIVs would cause a plant trip. Also performing such a test requires entry into the containment vessel and local manipulation of test connections located inside the drywell.

During plant operation at power and, on occasion, while in the cold shutdown mode, the containment atmosphere is maintained in a nitrogen-inerted condition.

During such periods, entry into the containment is not practical due to personnel.

safety concerns.

ALTERNATE TEST: These valves will be tested open at each refueling outage in accordance with ISTC-3522(c) and (f).

Rev. No. 10 Page 105 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES ROJ-22 SYSTEM: CORE SPRAY (CSP)

COMPONENTS: 14CSP-1OA, B CATEGORY: C SAFETY FUNCTION: The Core Spray pump discharge check valves open to allow flow to the CS spargers during injection for accident mitigation.. They rapid close upon pump stopping to prevent draining the pump. discharge piping.

DISCUSSION: Full stroke exercising as defined in Generic Letter 89-04 requires passing the maximum required accident condition flow through the valve. As defined in NEDC-31317P (JAF SAFERIGESTR), the Core Spray system maximum required flow for one pump is 5,456 gpm. Technical Specification 4.5.A.l.b requires flow rate testing of the Core Spray pumps at least 4,265 gpm against a system head corresponding toa reactor vessel pressure >; 113 psi above primary containment pressure. The TS test is performed quarterly through the Core Spray system test loop to the Torus. The test loop is currently evaluated at 4,700 gpm, therefore, testing at higher flows is not practicable.

JUSTIFICATION: Full stroke exercising these valves during power operation would require injecting 5,456 gpm into the reactor. During plant operation, the core spray pumps cannot produce sufficient discharge pressure to overcome reactor vessel pressure and provide flowinto the vessel.

• Duringcold shutdown, injecting into the reactor vessel requires a major effort to establish the requisite conditions to align 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, reverse flow testing of these valves requires installation of non-intrusive test equipment.

ALTERNATE TEST: In accordance with guidance provide in NUREG-1482 section 3.1.1 and 4.1.6, these valves will be partially stroke exercised on a quarterly basis by passing the Tech Spec flow of 4,265 gpm, and full flow exercised on a refueling basis by passing > 5,456 gpm. Additionally, these valves will be reverse flow tested on a refueling outage frequency. This meets the requirements of ISTC-3522(c) and (f).

Rev. No. 10 Page 106 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES ROJ-23 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)

COMPONENTS: 23HPI-402 and 23HPI-403 CATEGORY: C SAFETY FUNCTION: These valves open to eliminate any differential pressure that could force water from the suppression chamber into the HPCI exhaust piping when the

.suppression chamber pressure is greater than atmospheric. They close to prevent HPCI exhaust steam from entering 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, to perform this testing during plant operation at power, is considered to be impractical without a compensating level of quality and safety ALTERNATE TEST: These valves will be forward and reverse flow tested each refueling outage in accordance with ISTC-3522(c) and (f).

ROJ-24 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 impractical without a compensating level of quality and safety.

ALTERNATE TEST: These valves will be forward and reverse flow tested each refueling outage in accordance with ISTC-3522(c) and (f).

Rev. No. 10 Page 107 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES ROJ-25 SYSTEM: REACTOR CORE ISOLATION COOLING (RCIC)

COMPONENTS: 13RCIC-7 CATEGORY: C SAFETY FUNCTION: Thisvalve 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. Open exercise includes similar configuration.

ALTERNATE TEST: This valve will be reverse flow tested during refuel outages in accordance with ISTC-3522(c) and (f).

ROJ-26 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)

COMPONENTS: 23HPI-13 CATEGORY: 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 and the torus is vented to atmosphere. Placing the HPCI system and containment in this configuration during plant operation could adversely affect the plant's response in the event of an accident and is considered to be impractical without a compensating level of quality and safety.

ALTERNATE TEST: This valve will be reverse flow. tested during refuel outages in accordance with ISTC-3522(c) and (f).

Rev. No. 10 RPage 108 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES ROJ-27 SYSTEM: CONTROL ROD DRIVE HYDRAULICS (CRD)

COMPONENTS: 03HCU-115 (Typical for 137 HCUs) CATEGORY: C SAFETY FUNCTION: These valves close on initiation of a scram to prevent diversion of scram drive water into a depressurized charging header.

JUSTIFICATION: Exercising these valves during operation would require depressurization of the charging header with the potential for a loss of scram function.

ALTERNATE TEST: These valves will be reverse flow tested during refuel outages in accordance with ISTC-3522(c) and (f).

ROJ-28 SYSTEM: RESIDUAL HEAT REMOVAL (RHR)

COMPONENTS: 1OMOV-17 & 1OMOV-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. 1OMOV-17 also performs a containment

.isolation function.

JUSTIFICATION: With the reactor pressure greater than 75 psig, these valves are prevented from opening by an electrical interlock.

ALTERNATE TEST: These valves will be stroke time tested during refuel outages in accordance with ISTC-3521(e) and (h).

Rev. No. 10 Page 109 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES ROJ-29 SYSTEM: RESIDUAL HEAT REMOVAL (RHR)

COMPONENTS: 10AOV-68A, B CATEGORY: A/C SAFETY FUNCTION: These valves open to provide flow paths for LPCI injection to the reactor vessel.

They close for pressure isolation from the reactor vessel.

JUSTIFICATION: With the reactor at operating pressure, the RRHR pumps cannot develop sufficient discharge pressure to open these valves. The installed air operators are designed to open these, valves at zero differential pressure, which is* not practical with the reactor at operating pressure. Therefore, 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 due to personnel safety concerns related to the potential release of radioactive steam at high pressure.

.In accordance with recommendations of NUREG-1482 section 4.1.6, these ALTERNATE TEST:

valves will be forward and reverse flow tested during refueling outages in accordance with ISTC-3522(c) and (f).

ROJ-30 SYSTEM: FEEDWATER (FWS)

COMPONENTS: 34NRV-I11A, B CATEGORY: A/C SAFETY FUNCTION: These valves close to provide containment isolation and to prevent diversion of HPCI flow into the feedwater system.

JUSTIFICATION: Exercising these valves during operation would require isolation of feedwater flow to the reactor vessel. Such an evolution would create an adverse operating condition and potential automatic plant shutdown. To perform this testing during plant operation is considered to be impractical without a compensating level of quality and safety.

ALTERNATE TEST: In accordance with recommendations of NUREG- 1482 section 4.1.6, these valves will be forward and reverse flow tested during refueling outages in accordance with ISTC-3522(c) and (f).

Rev. No. 10 Page 110 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES ValveRelief Requests VRR-01 Withdrawn Rev. No. 10 Page 111 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES VRR-02 System:

TRAVERSING IN-CORE PROBE (TIP)

ASMIE Code Components Affected:

07SOV-104A, B, C Component/System Function:

These valves close to provide containment isolation.

Applicable Code Edition and Addenda

ASME OM Code-2001 including 2003 Addenda OM Code Category:

A Applicable Code ReQuirement:

ISTC-5151, "Valve Stroke Testing" ISTC-515 I(a), Active valves shall have their stroke times measured when exercised in accordance with ISTC-3500.

ISTC-5151(c), The stroke time of all valves shall be measured to at least the nearest second.

Reason for Request

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 opening stroke), and de-energizes the valve solenoid allowing the valve to stroke closed (< 2 seconds). The total elapsed time is specified to be </= 12 seconds.

Proposed Alternative and Basis for Use:

The overall cycle time (opened and closed) for these valves will be measured and evaluated in accordance with ISTC-5152.

Duration of Proposed Alternative:

The proposed alternative identified in this 10CFR50.55a Request shall be utilized during the Fourth Ten Year IST Interval.

Rev. No. 10 Page 112 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES Precedents:

This 10CFR50.55a Request was previously approved for the Interval 3 IST Program in NRC SER dated November 17, 1998 (TAC No. MA0096). The circumstances and basis for the previous NRC approval have not changed.

References:

None Rev. No. 10 Page 113 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES VRR-03 System:

Various Excess Flow Check Valves (Listed Below)

ASME Code Components Affected:

02-2EFV-PS-128A,B 02-2EFV-PT-24A,B 02-2EFV-PT-25A,B 02-2EFVI-DPT- 111A,B 02-2EFV 12-FT- 110A,C,E,G 02-2EFV2-DPT-111A,B 02-2EFV2-FT- 110A,C,E,G 02-3EFV-11 02-3EFV-13A,B 02-3EFV-15A,B,N 02-3EFV-17A,B 02-3EFV- 19A,B 02-3EFV-21A,B,C,D

• 02-3EFV-23A,B,C,D 02-3EFV-23 02-3EFV-25 02-3EFV-31A,B,C,D 02-3EFV-31E,F,G,H 02-3EFV-31J,K,L,M 02-3EFV-31N,P,R,S 02-3EFV-33 13EFV-01A,B 13EFV-02A,B 14EFV-3 1A,B 23EFV-01A,B 23EFV-02A,B 29EFV-30A,B,C,D 29EFV-34A,B,C,D.

29EFV-53A,B,C,D 29EFV-54A,B,C,D Component/System Function:

The reactor instrumentation lines excess flow check valves close to limit the flow in the respective instrument lines in the event of an instrument line break downstream of the EFCVs outside containment.

Applicable Code Edition and Addenda

ASME OM Code-2001 including 2003 Addenda Rev. No.. 10 Page 114 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES OM Code Category:

A/C

Applicable Code Requirement

Subsection ISTC, Inservice Testing of Valves in Light Water Reactor Power Plants, ISTC-35 10, "Exercising Test Frequency", requires these valves to be tested nominally every 3 months, except as provided by ISTC-3520, ISTC-3540, ISTC-3550, ISTC-3570, ISTC-5221 and ISTC-5222.

Reason for Request

Relax the number of EFCVs tested every refuel outage from "each" to a "representative sample" every refuel outage (nominally once every 24 months). The representative sample is based on approximately 20 percent of the valves each cycle such that each valve is tested every 10 years (nominal).

The BWROG Topical Report, B21-00658-01, dated November 1998, and associated NRC safety evaluation, dated March 14, 2000, provides the basis for this relief. The report provides justification for relaxation of the testing frequency described above. The BWROG report provides justification for relocation of the TS SR from the TS and relaxation of the testing intervals for the EFCVs. This specific request is solely for the relaxation in the testing frequency as described above.

The report demonstrates, through operating experience, a high degree of reliability with EFCVs and the low consequences of an EFCV failure. Reliability data in the report (Table 4-1) documents zero EFCV failures (failure to close) for the FitzPatrick plant. The instrument lines at FitzPatrick have a flow restricting orifice upstream of the EFCVs to limit reactor water leakage in the event of rupture. Previous evaluations contained in the James A. FitzPatrick Final Safety Analysis Report (FSAR) of such an instrument line rupture do not credit the EFCVs for isolating the rupture. Thus a failure of an EFCV, though not expected as a result of this request, is bounded by the analysis. Based on the BWROG report and the analysis contained in the FSAR, the proposed alternative to the required exercise testing frequency for EFCVs prescribed by the OM Code provides an acceptable level of quality and safety.

Proposed Alternative and Basis for Use:

Exercise test, by full-stroke to the position required to fulfill its function, a representative sample of EFCVs every refuel outage. The representative sample is based on approximately 20 percent of the valves each cycle such that each valve is tested every 10 years (nominal). EFCV failures will be documented in the FitzPatrick's Corrective Action Program as a surveillance test failure. The failure will be evaluated and corrected. An Equipment Failure Evaluation (EFE) will be required per the Corrective Action Program. The EFE will encompass common failure mode identification, industry experience evaluation, and review of similar component failure history.

Proposed Alternative Testing:

To ensure EFCV performance remains consistent with the extended test interval a minimum acceptance criteria of less than or equal to 1 failure per year on a 3 year rolling average will be required. Upon exceeding the criteria a root-cause evaluation is required to determine cause, extent of conditions, an evaluation of the testing interval to ensure reliability of the EFCVs, and a risk analysis of the effects of the failures on cumulative and instantaneous plant safety. Corrective actions and performance goals will Rev. No. 10 Page 115 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES be established based on the results of the root-cause analysis.

Duration of Proposed Alternative:

The proposed alternative identified in this 10CFR50.55a Request shall be utilized during the Fourth Ten Year IST Interval.

Precedents:

This 10CFR50.55a Request was previously approved for the Interval 3 IST Program in NRC SER dated October 10, 2000 (TAC No. MA8767). The circumstances and basis for the previous NRC approval have not changed.

(Refer to similar relief approved for Columbia Generating Station,.SER Dated March 23, 2007, TAC Nos. MD3537, MD3538, MD3539, MD3541, MD3542, MD3550 MD3551 and MD3552), RV05.)

References:

BWROG Report B21-00658-01, "Excess Flow Check Valve Testing Relaxation," dated November 1998.

Columbia Generating Station, SER Dated March 23, 2007, TAC Nos. MD3537, MD3538, MD3539, MD3541, MD3542, MD3550 MD3551 and MD3552 Rev. No. 10 Page 116 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES VRR-04 System:

HIGH PRESSURE COOLANT INJECTION (HPCI)

ASME Code Components Affected:

23HPI-130 HPCI Gland Seal Cooling Return Check Valve 23HPI-131 HPCI Condensate Pump P-141 Disch Check Valve 23HPI-32 HPCI Booster Pump P-lB Suct From CST 33TK-12A and B Check Valve 23HPI-61 HPCI Booster Pump P-lB Suct From Suppression Pool Check Valve 23HPI-62 HPCI Min Flow Line To RHR Check Valve Component/System Function:

Various

Applicable Code Edition and Addenda

ASME OM Code-2001 including 2003 Addenda OM Code Category:

C

Applicable Code Requirement

Subsection ISTC, Inservice Testingof Valves in Light Water Reactor Power Plants, ISTC-35 10, "Exercising Test Frequency", requires these valves to be tested nominally every 3 months, except as provided by ISTC-3520, ISTC-3540, ISTC-3550, ISTC-3570, ISTC-5221 and ISTC-5222.

For the listed valves, the FitzPatrick IST program exercises the provisions of ISTC-3522(c) and ISTC-5221 (c)(3) which together establish that: "As an alternative to the testing above, sample disassembly every refueling outage to verify operability of check valves may be used." Thus, a sample of these valves would be disassembled and inspected during each refueling outage.

Reason for Request

Relaxation of the "refueling outage" restriction of ISTC-3522(c) and ISTC-5221(c)(3) for testing of the listed valves to a test frequency of "a sample at least once per operating cycle."

Performance of these IST activities on a refueling outage frequency is currently acceptable in accordance with ISTC. By specifying testing activities on a frequency commensurate with each refueling outage, ISTC recognizes and establishes an acceptable time period between testing. Historically, the refueling outages have provided a convenient and defined time period in which testing activities could be safely Rev. No. 10 NPage 117 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES and efficiently performed. However, an acceptable testing frequency can be maintained separately without being tied directly to a refueling outage while still managing plant risk in accordance with 10 CFR 50.65(a)(4). IST performed on a frequency that maintains the acceptable time period between testing activities during the operating cycle is consistent with the intent of ISTC. Over time, approximately the same number of tests would be performed using the proposed operating cycle test frequency as would be performed using the current refueling outage frequency. Thus, IST activities performed during the proposed operating cycle test frequency provide an equivalent level of quality and safety as IST performed at a refueling outage frequency.

Proposed Alternative and Basis for Use:

Any on-line IST activities associated with this relief will be performed subject to the FitzPatrick program for compliance with the requirements of 10 CFR 50.65(a)(4), "Requirements for monitoring the effectiveness of maintenance at nuclear power plants."

Proposed Alternative Testing:

ISTC-5221(c)(3) allows sample disassembly each refueling outage to verify operability as an alternative to quarterly testing. This activity will be performed, with the exception that it will be done at a frequency of at least once per operating cycle in lieu of during each refueling outage.

Duration of Proposed Alternative:

The proposed alternative identified in this 10CFR50.55a Request shall be utilized during the Fourth Ten Year IST Interval.

Precedents:

This 10CFR50.55a Request was previously approved for the Interval 3 IST Program in NRC SER dated October 10, 2000 (TAC No. MA8767). The circumstances and basis for the previous NRC approval have not changed.

References:

None Rev. No. 10 Page 118 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES VRR-05 System:

ELECTRIC BAY AND TUNNEL VENTILLATION (SYSTEM 67)

CONTROL ROOM COOLING AND VENTILLATION (SYSTEM 70)

ASME Code Components Affected:

67PCV-101 70TCV- 120A, B 70TCV-121A, B Power-Operated Valves that are used for System Control and have a Safety Function currently included in the FitzPatrick Inservice Testing Program Component/System Function:

System control with an associated failsafe position feature.

Applicable Code Edition and Addenda

ASME OM Code-2001 including 2003 Addenda OM Code Category:

B

Applicable Code Requirement

ISTA-3130, "Application of Codes Cases", ISTA-3130(b) states, Code Cases shall be applicable to the edition and addenda specified in the test plan.

1. OM Subsection ISTC, Paragraph ISTC-5131, Pneumatically Operated Valves Stroke Testing

2. OM Subsection ISTC, Paragraph ISTC-5132, Stroke Test Acceptance Criteria

3. OM Subsection ISTC, Paragraph ISTC-5133(b), Stroke Test Corrective Action

Reason for Request

ISTA-3130, "Application of Codes Cases", ISTA-3130(b) states, Code Cases shall be applicable to the edition and addenda specified in the test plan. ISTA-3130(c) states, Code Cases shall be in effect at the time the test plan is filed, except as provided in ISTA-3130(d). ISTA-3130(d) states, Code Cases issued subsequent to filing the test plan may be proposed for use in amendments to the test plan. Licensees with a Code of record that is not applicable to the acceptance of this Code Case may submit a request for Rev. No. 10 Page 119 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES relief toapply the Code Case consistent with the indicated conditions to provide an acceptable level of quality and safety..

NUREG-1482, Revision 1, Section 4.2.9 states in part; Control valves that perform a safety or fail-safe function must be tested in accordance with the Code. provisions for IST to monitor the valves for degrading conditions.

The NRC staff recommends that licensees should apply ASME Code Case OMN-8, as accepted in RG 1.192, if concerns exist regarding IST of control valves with fail-safe functions.

Code Case OMN-8 states that stroke-time testing need not be performed for POVs when the only safety-related function of those valves is. to fail safe. Any abnormality or erratic action experienced during valve exercising should be recorded in the test record and an evaluation should be performed.

RG 1.192 allows licensees with an applicable Code of record to implement ASME Code Case OMN-8 in lieu of the Code provisions for Valve Stroke Testing, Stroke Time Acceptance Criteria and Stroke Test Corrective Action, without the need to submit a relief request.

The Code of record for FitzPatrick Fourth 10-Year IST Interval is OM Code-2001 Edition through 2003 Addenda. The applicable Code for OMN-8, as stated in RG 1.192, is OM Code-1998 through the 2000 Addenda.

Proposed Alternative and Basis for Use:

Pursuant to the guidelines provided in NUREG-1482, Revision 1, Section 4.2.9, FitzPatrick proposes to*

implement Code Case OMN-8 in lieu of the Code provisions for Valve Stroke Testing, Stroke Time Acceptance Criteria and Stroke Test Corrective Action specified in ISTC-5130. Code Case OMN-8 has been determined by the NRC to provide an acceptable level of quality and safety as documented in RG 1.192.

ASME Code Case OMN-8 states that stroke-time testing need not be performed for these valves when the only safety-related function of the valves is to fail safe. OM Code Committee is in the process of revising the applicability of this Code Case to the later approved OM Code editions and addenda.

Proposed Alternative Testing:

Using the provisions of this 10 CFR 50.55a request as an alternative to the AOV stroke-time testing requirements of ISTC-5130 provides an acceptable level of quality for the determination of valve operational readiness. Code Case OMN-8 should be considered acceptable for use with OM Code-2001 through 2003 Addenda as the Code of record. Therefore, pursuant to 10CFR50.55a(a)(3)(i), FitzPatrick requests relief from the specific ISTC Code requirements identified in this 10CFR 50.55a request.

These valves shall be exercised in accordance with the Subsection ISTC requirements and the failsafe position on a loss of power shall be verified. Any abnormality or erratic action experienced during valve exercising shall be evaluated per. the Corrective Action Program.

Rev. No. 10 Page 120 of 121

JAMES A. FITZPATRICK NUCLEAR POWER PLANT JAF-RPT-MULTI-03365 INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES Duration of Proposed Alternative:

The proposed alternative identified in this 10CFR50.55a Request shall be utilized during the Fourth Ten Year IST Interval.

Precedents:

None for FitzPatrick.

(Refer to similar relief approved for Columbia Generating Station, SER Dated March 23, 2007, TAC Nos. MD3537, MD3538, MD3539, MD3541, MD3542, MD3550 MD3551 and MD3552), RV05.)

(Refer to similar relief approved for Surry Power Station, Units 1 and 2 - SER Dated July 2, 2004, (TAC Nos. MC0120 through MC0146).

References:

Code Case OMN-8, "Alternative Rules for Preservice and Inservice Testing of Power-Operated Valves that are used for System Control and have a Safety Function per OM-10" Regulatory Guide 1.192, "Operation and Maintenance Code Case Acceptability, ASME OM Code",

Table 1, "Acceptable OM Code Cases" OM Code-2001 w/2003 Addenda, Paragraph ISTC-5130, "Pneumatically Operated Valves" OM Code-2001 w/ 2003 Addenda, Paragraph ISTA-3130, "Applicationof Code Cases" NUREG-1482, Revision 1, Section 4.2.9, "Control Valves with a Safety Function."

Surry Power Station, Units 1 and 2- SER Dated July 2, 2004, (TAC Nos. MC0120 through MC0146).

Columbia Generating Station, SER Dated March 23, 2007, TAC Nos. MD3537, MD3538, MD3539, MD3541, MD3542, MD3550 MD3551 and MD3552 Rev. No. 10 Page 121 of 121