Text

Fifth Ten-Year Interval Inservice Testing Program Plan
	ML18218A533
Person / Time
Site:	FitzPatrick
Issue date:	08/06/2018
From:	Drews W; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	JAFP-18-0078
	Download: ML18218A533 (191)
	v • d • e

Exelon Generation ~ James A FitzPatrick NPP PO Box 110 Lycoming , NY 13093 William C. Drews Regulatory Assurance Manager JAFP-18-0078 August 6, 2018 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 James A. FitzPatrick Nuclear Power Plant Renewed Facility Operating License No. DPR-59 NRC Docket No. 50-333

Subject:

Fifth Ten-Year Interval lnservice Testing Program Plan

Dear Sir or Madam:

The Fifth 10-Year lnservice Testing (IST) Interval began at James A. FitzPatrick Nuclear Power Plant on June 1, 2018. The Enclosure to this letter contains the IST program plan for your records; submitted in accordance with American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004, subsection ISTA-3200(a).

There are no new regulatory commitments contained in this letter.

If you have any questions concerning, please contact William Drews, Regulatory Assurance Manager, at (315) 349-6562.

Sincerely, William C. Drews Regulatory Assurance Manager WD/mh

Enclosure:

SEP-IST-007 lnservice Testing (IST) Program Plan cc:

NRC Regional Administrator, Region I NRC Resident Inspector NRC Project Manager

JAFP-18-0078 Enclosure SEP-IST-007 Inservice Testing (IST) Program Plan (189 Pages)

Exelon Nuclear Generation, LLC 200 Exelon Way Kennett Square, PA 19348 James A. FitzPatrick Nuclear Power Plant Unit Docket Number 50-333 PO Box 110 Lycoming, NY 13093 Commercial Service Date: October 17, 1974 SEP-IST-007 Inservice Testing (IST) Program Program Plan 5th Ten-Year Interval June 1, 2018 - September 30, 2027 Revision 10 June 1, 2018

SEP-IST-007 REVISION RECORD Ellective Revision Description Slnn&Date D*~ Prepen!d: Re~lewed: Approved; Site lST Corpotate En9r.

Engineer IST Engineer Proglllme Mana er 06/0112018

SEP-IST-007 TABLE OF CONTENTS SECTION

1.0 INTRODUCTION

1.1 Purpose 1.2 Scope 1.3 Discussion 1.4 References 2.0 INSERVICE TESTING PLAN FOR PUMPS 2.1 Pump Inservice Testing Plan 2.2 IST Plan Pump Table Description 3.0 INSERVICE TESTING PLAN FOR VALVES 3.1 Valve Inservice Testing Plan 3.2 IST Plan Valve Table Description 4.0 ATTACHMENTS

1. System and P&ID Listing

2. Pump Relief Request Index

3. Pump Relief Requests

4. Valve Relief Request Index

5. Valve Relief Requests

6. Relief Request RAIs and SER

7. Code Case Index

8. Cold Shutdown Justification Index

9. Cold Shutdown Justifications

10. Refueling Outage Justification Index

11. Refueling Outage Justifications

12. Technical Position Index

13. Technical Positions

14. Inservice Testing Pump Table

15. Inservice Testing Valve Table

16. Check Valve Condition Monitoring Plan Index ii

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant

1.0 INTRODUCTION

1.1 Purpose The purpose of this Inservice Testing (IST) Program Plan is to provide a summary description of the James A. FitzPatrick Nuclear Power Plant (JAF) IST Program in order to document its compliance with the requirements of 10 CFR 50.55a(f) for the 5th ten-year IST interval.

1.2 Scope This Inservice Testing Program Plan identifies all of the testing performed on the components included in the JAF Inservice Testing (IST) Program for the 5th ten-year IST interval, which began on June 1st, 2018 and is scheduled to end on September 30, 2027.

The Code of Federal Regulations, 10 CFR 50.55a(f)(4), requires 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 the ASME OM Code and addenda that are incorporated by reference in paragraph 10 CFR 50.55a(b)(3) for the initial and each subsequent 120-month interval.

Based on the start date identified above, the IST Program for the 5th ten-year interval is required by 10 CFR 50.55a(f)(4)(ii) to comply with the requirements of the ASME OM Code-2004, Code for Operation and Maintenance of Nuclear Power Plants, including addenda through the OM-2006, except where relief from such requirements has been granted in writing by the NRC.

The scope of the OM Code is defined in paragraph ISTA-1100 as applying to:

(a) pumps and valves that are required to perform a specific function in shutting down a reactor to the safe shutdown condition, in maintaining the safe shutdown condition, or in mitigating the consequences of an accident; (b) pressure relief devices that protect systems or portions of systems that perform on or more of the functions listed in (a), above; and (c) dynamic restraints (snubbers) used in systems that perform one or more of the functions listed in (a).

NOTE: This IST Program Plan addresses only those components included in (a) and (b) above. Dynamic restraints (snubbers) are addressed in a separate test program.

In order to determine the scope of the IST Program at JAF, an extensive scope evaluation was performed. This scope evaluation determined all of the functions required to be performed by all ASME Class 1, 2 and 3 systems in shutting down the reactor to the safe shutdown condition, in maintaining the safe shutdown condition or in mitigating the consequences of an accident. The determination of those functions was accomplished by a thorough review of licensing bases documents such as the UFSAR/FSAR, Plant Technical Specifications and Technical Specification Bases documents, etc. Next, a component-by-component review was performed to determine what function each pump and valve in the system was required to perform in order to support the safety function(s) of the system or subsystem. The results of these efforts are documented in the Stations IST Bases Document. In addition to a description of each components safety function(s), the Bases Document identifies 1-1

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant the tests and examinations that are performed on each component to provide assurance that they will be operationally ready to perform those safety function(s).

The Bases Document identifies those ASME Class 1, 2, and 3 pumps and valves that are in the scope of the IST Program, including those that do and those that do not have required testing. It also identifies those ASME Class 1, 2 and 3 pumps and valves that are outside the scope of the IST Program on the basis that they are not required to perform any specific safety function.

As stated at the beginning of this Section, the scope of this IST Program Plan is to identify all of the testing performed on those components within the scope of the IST Program. This is accomplished primarily by means of the IST Pump and IST Valve Tables contained in Attachments 14 and 15. The remaining Sections and Attachments of this document provide support information to that contained in the Tables. Components that do not require testing are not included in the IST Program Plan document.

In addition to those components that are required to perform specific safety function(s), the scope evaluation often determines that there are also ASME Safety Class 1, 2 and 3 components that are not required to perform a licensing-based safety function but which, nonetheless, may be relied upon to operate to perform a function with some significance to safety. It may also identify non-ASME Safety Class pumps or valves that have a safety function or may be relied upon to operate to perform a function with some significance to safety. None of these components are required by 10 CFR 50.55a to be included in the IST Program. However, such components may require testing in a manner which demonstrates their ability to perform their functions commensurate with their importance to safety per the applicable portions of 10 CFR 50, Appendix A or B. One option is to include pumps or valves that fit these conditions in the IST Program as augmented components.

JAF is licensed with Hot Shutdown as the safe shutdown condition. Therefore, the scope of the IST Program must include, as a minimum, all of those ASME Class 1, 2, and 3 pumps and valves which are required to shut down the Reactor to the Hot Shutdown condition, maintain the Hot Shutdown condition, or mitigate the consequences of an accident.

1.3 Discussion A summary listing of all the pumps and valves that are tested in accordance with the IST Program is provided in the IST Pump and IST Valve Tables contained in Attachments 14 and 15. The Pump and Valve Tables also identify each test that is performed on each component, the frequency at which the test is performed, and any Relief Request or Technical Position applicable to the test. For valves, the Valve Table also identifies any Cold Shutdown Justification or Refueling Outage Justification that is applicable to the required exercise tests. Additional information is provided for both pumps and valves. All of the data fields included in the IST Pump and Valve Tables are listed and described in Sections 2 and 3 of this document.

Following Sections 2 and 3 are several Attachments which provide information referenced in the Pump and Valve Tables.

Attachment 1 includes a listing of System and P&IDs on which a depiction of the pump or valve may be located.

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Attachment 2 provides an index of the Pump Relief Requests that apply to any of the pumps in the IST Program for this ten-year interval.

Attachment 3 includes a copy of each of those Relief Requests.

Attachment 4 provides an index of the Valve Relief Requests that apply to any of the valves in the IST Program for this ten-year interval.

Attachment 5 includes a copy of each of those Relief Requests.

Attachment 6 contains the Safety Evaluation Report(s) (SER) that document approval of the Relief Requests contained in Attachments 3 and 5.

Attachment 7 includes a list of the ASME OM Code Cases that are being invoked for this ten-year interval.

Attachment 8 provides an index of Cold Shutdown Justifications that apply to the exercise testing of any valves in the IST Program for this ten-year interval.

Attachment 9 includes a copy of each of those Cold Shutdown Justifications.

Attachment 10 provides an index of Refueling Outage Justifications that apply to the exercise testing of any valves in the IST Program for this ten-year interval.

Attachment 11 includes a copy of each of those Refueling Outage Justifications.

Attachment 12 provides an index of Technical Positions that apply to the IST Program for this ten-year interval. Technical Positions provide detailed information regarding how Exelon satisfies certain ASME OM Code requirements, particularly when the Code requirement may be ambiguous or when multiple options for implementation may be available. Technical Positions do not take exception to or provide alternatives to Code requirements.

Attachment 13 includes a copy of each Technical Position listed in Attachment 12.

As described previously, Attachments 14 and 15 include the IST Pump and Valve Tables.

Attachment 16 provides a listing of Check Valve Condition Monitoring (CVCM)

Program Plans. CVCM program plans are maintained in the IST program notebook located on the JAF network.

This IST Program Plan is a quality-related document and is controlled and maintained in accordance with approved Exelon Corporate Engineering and Records Management procedures.

1.4 References 1.4.1 Title 10, Code of Federal Regulations, Part 50, Section 55a (10 CFR 50.55a) 1.4.2 ASME OM Code-2004, Code for Operation and Maintenance of Nuclear Power Plant Components, including Addenda through OM 2006.

1.4.3 James A. FitzPatrick Technical Specification, 5.5.7 1.4.4 Exelon Corporation Administrative Procedure ER-AA-321, Administrative Requirements for Inservice Testing 1-3

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 2.0 INSERVICE TESTING PLAN FOR PUMPS 2.1 Pump Inservice Testing Plan The James A. FitzPatrick Inservice Testing Program for Pumps meets the requirements of Subsections ISTA and ISTB of the ASME OM Code-2004 with OMb 2006 addenda, with the exception of those specific applications identified in the Relief Requests contained in Attachment 3.

2.2 IST Plan Pump Table Description The pumps included in the James A. FitzPatrick Inservice Testing Program are listed in Attachment 14. The information contained in that table identifies those pumps required to be tested to the requirements of the ASME OM Code, the parameters measured, associated Relief Requests and comments, and other applicable information. The column headings for the Pump Table are listed below with an explanation of the content of each column.

Pump ID The unique identification number for the pump, as designated on the System P&ID or Flow Diagram Description The descriptive name for the pump Class The ASME Safety Class (i.e., 1, 2 or 3) of the pump. Non-ASME Safety Class pumps are designated N/A.

Group A or B, as defined in Reference 1.4.2.

DWG No. The Piping and Instrumentation Diagram or Flow Drawing on which the pump is shown CO-ORD The coordinates on the P&ID or Flow Diagram where the pump is shown.

Pump Type An abbreviation used to designate the type of pump:

C Centrifugal PDN Positive Displacement - Non-Reciprocating PDR Positive Displacement - Reciprocating VLS Vertical Line Shaft Driver The type of driver with which the pump is equipped.

A Air-motor D Diesel M Motor (electric)

T Turbine (steam) 2-1

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Test Lists each of the test parameters which are required to be measured for the specific pump. These include:

N Speed (for variable speed pumps, only)

P Differential Pressure P Discharge Pressure (positive displacement pumps)

Q Flow Rate Vd Vibration (displacement)

Vv Vibration (velocity)

Test Frequency An abbreviation which designates the frequency at which the associated test is performed:

Q Quarterly Y2 Once every 2 years NOTE: All tests are performed at the frequencies specified by Code unless specifically documented by a Relief Request.

Relief Request Identifies the number of the Relief Request applicable to the specified test.

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 3.0 INSERVICE TESTING PLAN FOR VALVES 3.1 Valve Inservice Testing Plan The James A. FitzPatrick Inservice Testing Program for Valves meets the requirements of Subsections ISTA and ISTC of the ASME OM Code-2004 with OMb-2006 addenda, with the exception of those specific applications identified in the Relief Requests contained in Attachment 5.

3.2 IST Plan Valve Table Description The valves included in the James A. FitzPatrick Inservice Testing Program are listed in Attachment 15. The information contained in that table identifies those valves required to be tested to the requirements of the ASME OM Code, the testing methods and frequency of testing, associated Relief Requests, comments, and other applicable information. The column headings for the Valve Table are delineated below with an explanation of the content of each column.

Valve ID The unique identification number for the valve, as designated on the System P&ID or Flow Diagram.

Description The descriptive name for the valve.

Vlv Type An abbreviation used to designate the body style of the valve:

3W 3-Way 4W 4-Way BAL Ball BTF Butterfly CK Check DIA Diaphragm GA Gate GL Globe PLG Plug RPD Rupture Disk RV Relief SCK Stop-Check SHR Shear (SQUIB)

XFC Excess Flow Check 3-1

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Actu Type An abbreviation which designates the type of actuator on the valve. Abbreviations used are:

AO Air Operator DF Dual Function (Self and Power)

EXP Explosive HO Hydraulic Operator M Manual MO Motor Operator SA Self-Actuating SO Solenoid Operator P&ID The Piping and Instrumentation Diagram or Flow Drawing on which the valve is shown.

Coord The coordinates on the P&ID or Flow Diagram where the valve is shown.

Class The ASME Safety Class (i.e., 1, 2 or 3) of the valve.

Non-ASME Safety Class valves are designated by N/A.

Positions Abbreviations used to identify the normal and safety-Norm/Safe related positions for the valve. Abbreviations used are:

AI As Is C Closed CKL Closed/Actuator Key Locked D De-energized D/E De-energized or Energized E Energized LC Locked Closed LO Locked Open LT Locked Throttled O Open O/C Open or Closed OKL Open/Actuator Key Locked SYS System Condition Dependent T Throttled Cat The ASME Code category or categories of the valve as defined in Reference 1.4.2.

Active/Passive A or P, used to designate whether the valve is active or passive in fulfillment of its safety function.

The terms active valves and passive valves are defined in Reference 1.4.2.

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Testing Requirements

Test A listing of abbreviations used to designate the types of testing which are required to be performed on the valve based on its category and functional requirements. Abbreviations used are:

BDC Bidirectional Check Valve test (non-safety related closure test)

BDO Bidirectional Check Valve test (non-safety related open test)

CC2 Check Valve Exercise Test - Closed CO2 Check Valve Exercise Test - Open CP2 Check Valve Partial Exercise Test DT Category D Test EC Exercise Test - Closed (manual valve)

EO Exercise Test - Open (manual valve)

FC Fail-Safe Exercise Test - Closed FO Fail-Safe Exercise Test - Open LT1 Leak Rate Test PI Position Indication Verification Test RT Relief Valve Test SC Exercise Closed (without stroke-timing)

SO Exercise Open (without stroke-timing)

SP Partial Exercise (Cat. A or B)

STC Exercise/Stroke-Time Closed STO Exercise/Stroke-Time Open 1 A third letter, following the LT designation for leakage rate test, may be used to differentiate between the tests. For example, Appendix J leak tests will be designated as LTJ, low pressure (non-Appendix J) leak tests as LTL, and high pressure leak tests as LTH.

2 Three letter designations should be used for check valve tests to differentiate between the various methods of exercising check valves. The letter following CC, CO or CP should be A for acoustics, D for disassembly and inspection, F for flow indication, M for magnetics, R for radiography, U for ultrasonics, or X for manual exercise.

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant

Freq An abbreviation which designates the frequency at which the associated test is performed. Abbreviations used are:

AJ Per Appendix J CM Per Check Valve Condition Monitoring Program CS Cold Shutdown M[n] Once Every n Months Q Quarterly RR Refuel Outage R[n] Once Every n Refuel Outages SA Sample Disassemble & Inspect TS Per Technical Specification Requirements Y[n] Once Every n Years

RR Identifies the number of the Relief Request applicable to the specified test.

Justification A cross-reference to the applicable Cold Shutdown Justification or Refuel Outage Justification which describes the reasons why reduced-frequency exercise testing is necessary for the applicable valve.

Comments Any appropriate reference or explanatory information (e.g., technical positions, etc.).

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant SECTION 4.0 ATTACHMENTS

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ATTACHMENT 1 SYSTEM AND P&ID LISTING

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 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-119A 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 34 Feedwater FM-34A 39 Breathing, Instrument & Service Air FM-39A 46 Service & Emergency Service Water FM-46A,B 66 Reactor Building Service Ventilation (Service Water) FB-10H 70 Control Room Service & Chilled Water FB-35E A1 - 1

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ATTACHMENT 2 PUMP RELIEF REQUEST INDEX

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant RELIEF REQUEST RELIEF REQUEST TITLE APPROVAL NUMBER DATE PRR-001 Core Spray Pump Suction Pressure Gauge Range 4/13/2018 PRR-002 RHRSW Smooth Running Pumps 4/13/2018 PRR-003 RHR Pump Vibration Alert Range 4/13/2018 PRR-004 Code Case OMN-21 4/13/2018 A2 - 1

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ATTACHMENT 3 PUMP RELIEF REQUESTS

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant PRR-01 CORE SPRAY PUMP SUCTION PRESSURE GAUGE RANGE PLANT/UNIT: James A. Fitzpatrick (JAF) Nuclear Power Plant INTERVAL: 5th Interval beginning June 1, 2018, and ending September 30, 2027 COMPONENTS 14P-1A Core Spray Pump AFFECTED: 14P-1B Core Spray Pump CODE EDITION ASME OM Code-2004 Edition with Addenda through OMb-2006 AND ADDENDA:

REQUIREMENTS: ISTB-3500, Data Collection, paragraph ISTB-3510, General, (a) Accuracy, states, in part: Instrument accuracy shall be within the limits of Table ISTB-3510-1 ISTB-3500, Data Collection, paragraph ISTB-3510, General, (b) Range, (1),

states: The full-scale range of each analog instrument shall be not greater than three times the reference value.

ISTB-3500, Data Collection, Table ISTB 3510-1, Required Instrument Accuracy, requires pressure and differential pressure accuracy to be +/- 2% for Group A and Group B tests.

REASON FOR Pursuant to 10 CFR 50.55a(z)(2), an alternative is requested from the REQUEST: ASME OM Code paragraph(s) ISTB 3500, Table ISTB 3510-1, that requires the suction and differential pressure accuracy to be within +/- 2% of full scale range and ISTB 3510(b)(1) that requires the full-scale range of an analog instrument to be less than three times the reference value. The basis of this request is that the Code requirements present an undue hardship without a compensating increase in the level of quality or safety.

PROPOSED JAF is proposing to use the existing installed plant suction pressure ALTERNATIVE gauges to determine the pump differential pressure for testing the Core AND BASIS: Spray pumps for the Group B pump tests. The comprehensive tests are performed using Measuring and Test Equipment (M&TE).

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 suction pressure gauge full-scale range of 72.28 pounds per square inch gauge (psig), [from 25 inches of mercury (12.28 psig) vacuum to 60 psig], is sufficient for a post-accident condition when the torus is at the maximum accident pressure. This, however, exceeds the range limit of 15 psig for the suction pressure under the test condition of approximately 5 psig.

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant PRR-01 CORE SPRAY PUMP SUCTION PRESSURE GAUGE RANGE PROPOSED The installed discharge pressure instrument loop is calibrated to within ALTERNATIVE +/- 2% full-scale accuracy. However, the installed suction pressure AND BASIS instrument loop is calibrated to within +/- 2.4% full-scale accuracy. The (Continued) 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 +/- 1.7 psi and in measured discharge pressure is +/- 10 psi. Therefore, the differential pressure (typical [Pdisch - Psuct]) would be (300 - 5), which is 295 +/-

11.7 psi for an inaccuracy of 4.0%. 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 maximum variation due to inaccuracy in measured suction pressure would be +/- 0.3 psi and the resulting differential pressure measurement would be 295 +/- 10.3 psi for an inaccuracy of 3.5%.

The decrease in inaccuracy of 0.5%, by using the Code compliant gauge, is insignificant and does not warrant the additional manpower and exposure required to change the suction pressure gauge quarterly for test purposes.

In addition, the Code would allow a full-scale range for the discharge pressure measurement of 900 psig for the typical 300 psig discharge pressure. This would translate into a differential pressure measurement of 295 +/- 18.3 psig or an inaccuracy of 6.2%, if the installed instrumentation met the Code requirements of 0 - 15 psig for the suction pressure gauge and 0 -

900 psig for the discharge pressure gauge. The existing measurement is significantly better than the maximum Code allowable inaccuracy.

Based on the hardship resulting from implementing the OM Code requirements without a compensating increase in the level of quality and safety, this proposed alternative is requested pursuant to 10 CFR 50.55a(z)(2).

DURATION: 5th IST Interval, beginning June 1, 2018, and ending September 30, 2027 PRECEDENT: This proposed 10 CFR 50.55a alternative (for the instrument range) was previously authorized for use as PRR-02 at JAF during the fourth ten-year IST interval via the NRCs safety evaluation, dated November 27, 2007.

(ADAMS Accession No. ML072910422)

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant PRR-002 RHRSW Smooth Running Pumps PLANT/UNIT: James A. Fitzpatrick (JAF) Nuclear Power Plant INTERVAL: 5th Interval beginning June 1, 2018, and ending September 30, 2027 COMPONENTS 10P-1A, Residual Heat Removal Service Water (RHRSW) Pump AFFECTED: 10P-1B, RHRSW Pump 10P-1C, RHRSW Pump 10P-1D, RHRSW Pump The RHRSW pumps are vertical line shaft pumps that provide cooling water to the Residual Heat Removal (RHR) Heat Exchangers during a design basis event. These smooth running pumps are in the JAF inservice testing (IST) program and are ASME OM Code Group A pumps.

CODE EDITION ASME OM Code-2004 Edition with Addenda through OMb-2006 AND ADDENDA:

REQUIREMENTS: ISTB-3300, Reference Values, states that Reference values shall be obtained as follows: 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.

REASON FOR Pursuant to 10 CFR 50.55a(z)(1), an alternative is requested from the REQUEST: ASME OM Code ISTB requirements referenced above, related to the vibration reference value (Vr), for the pumps that are identified as smooth running pumps. The basis of this request is that the alternative reference values provide an acceptable level of quality or safety.

The smooth running pumps in the JAF Nuclear Power Plant IST Program have at least one vibration reference value (Vr) that is currently less than 0.05 inches per second (ips). 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-5221-1, Vertical Line Shaft Centrifugal Pump Test Acceptance Criteria, 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, Corrective Action, (a), 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.

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant PRR-002 RHRSW Smooth Running Pumps REASON FOR For very small reference values for vibrations, a significant portion of the REQUEST reading can be from flow variations, hydraulic noise, and instrument (Continued): error, which can affect the repeatability of subsequent measurements. Also, experience gathered by the JAF Nuclear Power Plant Predictive Maintenance (PdM) group has shown that changes in vibration levels in the range of 0.05 ips do not normally indicate significant degradation in pump performance.

In order to avoid unnecessary corrective actions, a minimum value for Vr of 0.05 ips is proposed. This minimum value would be applied to individual vibration locations, for the RHRSW pumps identified in this alternative request, with reference vibration values less than 0.05 ips.

PROPOSED In lieu of applying the reference values required by ISTB-3300, for ALTERNATIVE smooth running pumps with a measured reference value below 0.05 ips, AND BASIS: JAF will apply a minimum value for Vr of 0.05 ips for the particular vibration measurement location. This minimum value would be applied to individual vibration locations for the RHRSW pumps, identified in this alternative request. The subsequent test results for that location will be compared to an Alert Range limit of 0.125 ips and a Required Action limit of 0.300 ips. These ranges, resulting from the proposed Vr of 0.05 ips and using the existing OM Code multipliers, shall be applied to vibration test results during both Group A and Comprehensive tests.

Therefore, the smallest OM Code Alert Range limit for any IST pump vibration location would be > 2.5 times Vr, or 0.125 ips, which is within the fair range of the General Machinery Vibration Severity Chart provided by IRD Mechanalysis, Inc. Likewise, the smallest OM Code Required Action limit for any IST Pump vibration location for which the pump would be inoperable would be > 6 times Vr, or 0.300 ips.

For comparison purposes, ASME Section 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 ips for an 1800 rpm pump and 0.094 ips for a 3600 rpm pump. The minimum reference value proposed (0.05 ips) for smooth-running pumps is roughly equal to the ASME Code Section XI IWP reference value for an 1800 rpm pump and more conservative than the reference value for a 3600 rpm pump.

In addition to the requirements of OM Code Subsection ISTB for IST, the pumps in the JAF Nuclear Power Plant IST Program are also included in the JAF Nuclear Power Plant PdM Program. The JAF Nuclear Power Plant PdM Program currently employs predictive monitoring techniques such as:

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

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant PRR-002 RHRSW Smooth Running Pumps PROPOSED If the measured parameters are outside the normal operating range or are ALTERNATIVE determined by analysis to be trending toward an unacceptable degraded AND BASIS state, appropriate actions are taken that may include: initiate a Condition (Continued): Report (CR), increase monitoring to establish a rate of change, review component specific information to identify the cause, and remove the affected pump from service to perform maintenance.

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

Using the provisions of this request as an alternative to the required reference value determination specified in ISTB-3300, the vibration acceptance criteria, based on the alternative reference value, 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(z)(1), JAF requests use of the proposed alternative to the specific ISTB Code requirements identified in this 10CFR50.55a Request.

DURATION: 5th IST Interval, beginning June 1, 2018, and ending September 30, 2027 PRECEDENCE: 1. This 10CFR 50.55a request was previously approved for the Interval 4 IST Program at JAF via relief request PRR-04 in NRC SER dated November 27, 2007 (ADAMS Accession No. ML072910422).

2. Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2) -

Evaluation of Inservice Testing (IST) Pump Relief Request No. 8 Revisions 1K and 2I, respectively, NRC SER dated December 27, 2004 (ADAMS Accession No. ML043430042).

REFERENCES:

1. General Machinery Vibration Severity Chart provided by IRD Mechanalysis, Inc.

2. ASME Section XI, Subsection IWP, Inservice Testing of Pumps in Nuclear Power Plants, (IWP-3200-IWP-3400),1986 Edition A3-5

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant PRR-003 RHR PUMP VIBRATION ALERT RANGE PLANT/UNIT: James A. Fitzpatrick (JAF) Nuclear Power Plant INTERVAL: 5th Interval beginning June 1, 2018, and ending September 30, 2027 COMPONENTS 10P-3A, Residual Heat Removal (RHR) Pump AFFECTED: 10P-3B, RHR Pump 10P-3C, RHR Pump 10P-3D, RHR Pump The four RHR pumps are classified as ASME Class 2 and OM Code Group A pumps.

CODE EDITION ASME OM Code-2004 Edition with Addenda through OMb-2006 AND ADDENDA:

REQUIREMENTS: ISTB Table ISTB-5121-1, Centrifugal Pump Test Acceptance Criteria provides the vibration Alert Range low-end absolute limit of 0.325 inches per second (ips) for the Group A and Comprehensive tests.

REASON FOR Pursuant to 10 CFR 50.55a(z)(2), an alternative is requested from the REQUEST: ASME OM Code, Subsection ISTB, Table ISTB-5121-1 vibration Alert Range low-end absolute limit of 0.325 ips requirement during the Group A or biennial comprehensive pump test. The basis of this request is that the Code requirements present an undue hardship without a compensating increase in the level of quality or safety.

The increased periodicity of testing, resulting from the 0.325 ips Code requirement is an additional burden to the Operations staff, plant scheduling, and adds unnecessary run time to all RHR pumps. This request is based on analysis of vibration and pump differential pressure data indicating that no pump degradation is taking place.

PROPOSED JAF is proposing to use an alternative vibration Alert Range low-end ALTERNATIVE absolute limit of 0.408 ips. This provides an alternative method that AND BASIS: continues to meet the intended function of monitoring the pump for degradation over time while keeping the required action level unchanged.

Pump Testing Methodology The RHR pumps at JAF are tested using a full flow recirculation test line back to the suppression pool for each pump surveillance test (including quarterly group A tests and biennial comprehensive pump tests). These pumps have a minimum flow line (per division) which is used only to protect the pump from overheating when pumping against a closed discharge valve. The minimum flow line isolation valve for each division is initially open when the pump is started, and flow is recirculated through A3-6

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant PRR-003 RHR PUMP VIBRATION ALERT RANGE PROPOSED the minimum flow line back to the suppression pool. Then, the full flow ALTERNATIVE test line isolation valve is throttled open to establish flow through the AND BASIS full flow recirculation test line. The minimum flow line is then (Continued): automatically isolated and all flow is directed through the full flow test line for the IST test. Based on the full flow test line configuration, this test methodology results in flow-induced, broadband vibration readings greater than 0.325 ips, but less than the required action limits.

The RHR system is operated in the same manner and under the same conditions for each IST test, regardless of whether JAF is operating or shutdown. Consequently, the pumps will experience the same potential for flow-induced, broadband vibration whenever they are tested, whether JAF is operating or shutdown. As a result, this alternative is proposed for the inservice testing of RHR pumps when vibration measurements are required.

NRC Staff Document NUREG/CP-0152 NRC Staff document NUREG/CP-0152, entitled Proceedings of the Fourth NRC/ASME Symposium on Valve and Pump Testing, dated July 15-18, 1996, included a paper prepared by the NRC staff, entitled Nuclear Power Plant Safety Related Pump Issues. That paper presented four key components that should be addressed in an alternative request of this type to streamline the review process. These four key components are as follows:

I. The licensee should have sufficient vibration history from the inservice testing which verifies that the pump has operated at the vibration level for a significant amount of time, with any spikes in the data justified.

II. The licensee should have consulted with the pump manufacturer or vibration expert about the level of vibration the pump is experiencing to determine if the pump operation is acceptable.

III. The licensee should describe attempts to lower the vibration below the defined code absolute levels through modifications to the pump.

IV. The licensee should perform a spectral analysis of the pump driver system to identify all contributors to the vibration levels.

The following is a discussion of how these four key components are addressed for this alternative request.

I. Vibration History

a. Testing Methods and Code Requirements Elevated vibration levels on the RHR pumps has been a condition that has existed since original installation. Prior to 1998, testing was measured in displacement (mils). These A3-7

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant PRR-003 RHR PUMP VIBRATION ALERT RANGE PROPOSED readings were taken in two directions, horizontal in-line with ALTERNATIVE pump flow and horizontal perpendicular to pump flow. In 1998, AND BASIS JAF entered the third 10-year interval and implemented (Continued): ASME/ANSI OM-1989, OM-6, for pump testing. During this interval, IST vibration data was taken in displacement (mils).

Additional data was also taken in velocity (ips), with an additional data point in the axial direction. At that time, the velocity data points were used as information only. Upon the fourth 10 year interval at JAF, ASME OM Code 2001 Edition with 2003 Addenda was adopted. With this adoption, it was determined that vibration measurements in velocity would be a much better indicator of pump condition and would be beneficial in terms of early identification of degradation. Therefore, data exists for two vibration points on each RHR pump from January 1986 to August 1998 in mils. Data from May of 1999 to present is in the form of ips at three vibration data points. Various analyses of this data are included as Figures 1 through10 within this alternative request.

b. Review of Vibration History Data IST trends of RHR Pump vibration (Figures 1 - 4), which include data from 1999 through present, show some readings to be at or above current IST vibration alert criteria. From Figures 9 and 10, it can be seen that, 10P-3B and 10P-3C have exceeded the current OM Code Alert criteria of 0.325 ips.

RHR pump differential pressure trends (Figures 5 - 8) illustrate the differential pressure data during the same time period as the vibration Figures 1 - 4. These graphs show a step change in flow around the 2009 time frame. This change is due to surveillance test changes in which test flow was lowered. The change was made after engineering analysis resulted in revised pump flow criterion.

These trends do not show any signs of hydraulic degradation.

A review of the maintenance history for all four pumps and motors shows very minimal maintenance has been performed beyond the normally scheduled preventive maintenance. The only maintenance deemed to have the potential to affect vibration values is that of mechanical seal replacement. RHR pumps 10P-3A and 10P-3C both had mechanical seals replaced in 2009. Post work testing of these replacements did not show any change in vibration values.

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant PRR-003 RHR PUMP VIBRATION ALERT RANGE PROPOSED Average run times for each RHR pump per cycle is ALTERNATIVE approximately 200-300 hours. Run times of this nature, AND BASIS combined with the pumps and motors being built and maintained (Continued): to the nuclear quality standards, are considered to have a low likelihood for significant wear related degradation.

c. Review of Spikes in Vibration Data Trends of recent vibration history (4th Interval) do not show any significant spikes above baseline levels. Instead, all values are seen as fairly consistent and there have been no significant degrading trends associated with vibration data for the past 16 years.

While the overall vibration trends have been steady, when compared to the OM Code, recent vibration data points have exceeded the OM Code Alert limit of 0.325 ips. RHR pump 10P-3B exceeded this value once in 2010 and once in 2014. Also, RHR pump 10P-3C exceeded this value once in 2012 and twice in 2016.

Without the relief, for the 4th IST Interval at JAF, granted by NRC in the Safety Evaluation dated September 15, 2014, these pumps would be on increased frequency testing even though there has been no pump degradation.

II. Consultation - Pump Manufacturer/Vibration Expert During the initial investigation for the cause of the failed vibration acceptance criteria, Mancini Consulting Services (an industry pump expert) and Flowserve (the pump vendor) were consulted for input.

Each RHR pump motor is vertically mounted to the pump casing, with the piping entering and exiting the pump casing horizontally. The RHR pump motors weigh approximately 6500 pounds and operate at 1800 revolutions per minute (RPM). Other motor specifications include: 1000 Horsepower (Hp), 3 Phase, 60 Hertz (Hz), 4000 Volts. The pump casing, weighing 6100 pounds, is mounted on a reinforced floor pad.

The 20-inch suction piping enters the room level with the pump centerline. An additional 20-inch line tees into the suction piping approximately 5 feet from the pump. The 16-inch discharge piping leaves the pump on the same plane as the suction piping but then elbows 90 degrees vertically 6 feet from the pump. This is then followed by a discharge check valve and an isolation valve on the vertical run. See figures 11 and 12 for the isometric layout of the suction and discharge piping near the RHR pumps.

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant PRR-003 RHR PUMP VIBRATION ALERT RANGE PROPOSED Vibration monitoring points are located on the RHR motor/pump ALTERNATIVE assembly. Points V1, V2, and V3 (Vaxial) are the specified locations for AND BASIS IST data collection. V1 is taken in the vertical direction in line with (Continued): pump flow. V2 is taken 90 degrees from V1 perpendicular to pump flow. V3 is taken 90 degrees from the V1 and V2 plane, on the underside of the motor.

Resonance testing was performed on all four RHR pump housings in the V1, V2, and V3 directions. Analysis has shown that the contributing cause of vibration in the V1 and V2 directions is from broadband peaks in the spectrum between 85.1 to 102.7 Hertz. These frequencies fall in the area of pump operation as these pumps are of a 3-vane design. In the case of the RHR pumps, vane pass frequency excitations are influencing vibration measurements.

III. Attempts to Lower Vibration Prior to completely understanding the cause of the vibration, it was thought that after the adjustment of testing flow in early 2009, vibrations would decrease. Through the recent analysis of the vibration spectrum, the structural resonance and the running speed peaks were confirmed.

This analysis indicated that the running speed spectral peaks are consistent over years of testing. With the resonant frequency being considered as a significant contributor to exceeding the alert vibration range, options to reduce resonance reside in stiffening the pump or an internal design change (i.e. modifying to a 5-vane design).

JAF initially pursued a path to add additional stiffening to the pump and piping system. With the addition of supports, a new seismic analysis would be required for each RHR pump and the associated piping. Due to the complexity and resources needed for a new analysis, combined with the industry OE (reference Fermi 2 approved submittal ML101670372) which shows that stiffening operations also lend to the potential for vibrations to be transferred to the surrounding piping, efforts were ceased.

Major modifications, such as to add stiffening to the pump/motor system or changing the pump to a 5-vane design, when the pumps are not seen as degrading, are not deemed to result in an increase in the level of safety.

IV. Spectral Analysis ASME OM Code 2004 Edition/2006 Addenda paragraph ISTB-6400 states: If the reference value of a particular parameter being measured or determined can be significantly influenced by other related conditions, then these conditions shall be analyzed1 and documented in the record of tests. (See ISTB-9000).

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant PRR-003 RHR PUMP VIBRATION ALERT RANGE PROPOSED The footnote (1) for analyzed states: Vibration measurements of ALTERNATIVE pumps may be foundation, driver, and piping dependent. Therefore, if AND BASIS initial readings are high and have no obvious relationship to the pump, (Continued): then vibration measurements should be taken at the driver, at the foundation, and on the piping and analyzed to ensure that the reference vibration measurements are representative of the pump and the measured vibration levels will not prevent the pump from fulfilling its function.

Spectral analysis was performed showing the total vibration broken down into individual frequencies over a span from 0-1500 Hz. The analysis, which included a historical compilation of the data points that coincide with the IST surveillance testing, shows the broadband vibration is seen occurring at three times the pump running speed.

These vibrations may exceed the Code alert criteria, which would trigger the corrective action process and the need to increase the testing frequency.

Spectral data indicates that the overall vibration levels are primarily made up of a spectrum from 85 to 103 Hz due to the vane pass frequency induced by a 3-vane pump at 1800 RPM. As this vibration stems from the design of the pump, all four RHR pumps are susceptible. This vibration is accentuated on 10P-3B and 10P-3C due to the similar piping configurations. Spectral data do not indicate any degradation to the bearings, pump, or motor that would lead to imbalance or misalignment.

Basis for the OM Code Alternative Alert Values By this alternative request, JAF is proposing to increase the Alert Range low-end absolute limit for vibration from 0.325 ips to 0.408 ips for all four RHR pumps in the V1 and V2 directions. The vane pass induced broadband vibration occasionally causes the overall vibration value to exceed 0.325 ips, resulting in the pumps being placed on an increased testing frequency. A new alert acceptance criterion of 0.408 ips coincides with the warning level that is already developed per the Predictive Maintenance (PdM) Program.

The basis for the 0.408 ips warning level came from the Technical Associates of Charlotte recommendations for vertical pumps. The set points recommended were 0.350 ips and 0.525 ips. The PdM program uses those set points as high and low criteria but the program also has two additional levels. These two additional levels split the difference between the suggested set points, resulting in 0.408 ips and 0.466 ips.

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant PRR-003 RHR PUMP VIBRATION ALERT RANGE PROPOSED The RHR Pump vendor (Byron Jackson, now owned by Flowserve) did not ALTERNATIVE recommend a specific value regarding the increased vibration alert limit, but AND BASIS stated that the pumps should not be adversely impacted provided that no (Continued): upward trend existed in the vibration measurement data.

Expert analyses and maintenance reviews have shown that this vibration has not resulted in degradation to the pump or motor. Data trends show that overall vibrations have remained steady since 1998.

The new Alert criterion value allows an alternative measure that still meets the intended function of monitoring the pump for degradation, while leaving the action levels as mandated by the OM Code. The proposed criterion encompasses the previous values that exceeded the Alert level, which would eliminate the unnecessary actions associated with exceeding the OM Code Alert limit when the pump is not seen as degrading. Any corrective actions triggered by vibrations between 0.408 ips to 0.7 ips will result in the same OM Code actions as previously required when exceeding the Alert limit of 0.325 ips.

The vibration specialist at JAF routinely performs a spectral analysis on all data recorded during RHR pump inservice testing. This analysis is in addition to IST total vibration values. The analysis provides additional confidence on the ability to detect degradation at an early stage.

Each RHR pump motor is also monitored through the Preventive Maintenance (PM) and PdM programs. While these actions are intended to prevent degradation, any off-normal or unexpected conditions, will act as an indicator of the early stages of degradation. PM and PdM activities include:

annual non-intrusive thermography, annual motor bearing sample, and 10-year internal visual inspection.

Conclusion Based on the hardship resulting from implementing the OM Code requirements without a compensating increase in the level of quality and safety, this proposed alternative is requested pursuant to 10 CFR 50.55a(z)(2). JAF considers the proposal to use an alternative vibration Alert Range low-end absolute limit of greater than or equal to 0.408 ips, while keeping the required action level unchanged, will provide an alternative method that continues to meet the intended function of monitoring the RHR pumps for degradation over time.

DURATION: 5th IST Interval, beginning June 1, 2018, and ending September 30, 2027 A3-12

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant PRR-003 RHR PUMP VIBRATION ALERT RANGE PRECEDENCE: 1. This 10CFR50.55a request was previously approved for the Interval 4 IST Program at JAF via Relief Request PRR-05 in the Safety Evaluation, dated September 15, 2014 (ML14241A329)

2. Cooper Nuclear Station - RE: Request for Relief (Relief Request RP-06, Revision 2) from the Requirements of ASME Code concerning Inservice Testing of Core Spray Pump CS-P-B, dated February 25, 2004 (ML040560318)

3. Evaluation of Relief Request PR-2 Associated with Second 10-Year Interval Inservice Testing Program for Pumps and Valves for Byron Station, Units 1 and 2, dated February 19, 2002 (ML020070381)

4. Fermi 2 - Evaluation of Relief Request Nos: PRR-004, PRR-005, PRR-007, and PRR-010 for the Third 10-Year Interval Inservice Program, dated July 6, 2010 (ML101670372)

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant PRR-004 CODE CASE OMN-21 PLANT/UNIT: James A. Fitzpatrick (JAF) Nuclear Power Plant INTERVAL: 5th Interval beginning June 1, 2018, and ending September 30, 2027 COMPONENTS Refer to Table PRR-04 AFFECTED:

CODE EDITION ASME OM Code-2004 Edition with Addenda through OMb-2006 AND ADDENDA:

REQUIREMENTS: ISTB-5121, "Group A Test Procedure," paragraph ISTB-5121(b) states, in part, that The resistance of the system shall be varied until the flow rate equals the reference point. Alternatively, the flow rate shall be varied until the differential pressure equals the reference pointISTB-5122, "Group B Test Procedure," paragraph ISTB-5122(c) states, System resistance may be varied as necessary to achieve the reference point.

ISTB-5123, "Comprehensive Test Procedure," paragraph ISTB-5123(b) states, in part, that the resistance of the system shall be varied until the flow rate equals the reference point. Alternatively, the flow rate shall be varied until the differential pressure equals the reference point ISTB-5221, "Group A Test Procedure," paragraph ISTB-5221(b) states, in part, that The resistance of the system shall be varied until the flow rate equals the reference point. Alternatively, the flow rate shall be varied until the differential pressure equals the reference point ISTB-5222, "Group B Test Procedure," paragraph ISTB-5222(c) states, System resistance may be varied as necessary to achieve the reference point.

ISTB-5223, "Comprehensive Test Procedure," paragraph ISTB-5123(b) states, in part, that The resistance of the system shall be varied until the flow rate equals the reference point. ...Alternatively, the flow rate shall be varied until the differential pressure equals the reference point REASON FOR Pursuant to 10 CFR 50.55a(z)(1), an alternative is proposed to the pump REQUEST: testing reference value requirements of the ASME OM Code. The basis of the request is that the proposed alternative would provide an acceptable level of quality and safety. Specifically, this alternative is requested for all inservice testing of IST Program pumps as listed in attached Table PRR-04, Pumps Affected by Alternative Request PRR-04.

For pump testing, there is difficulty adjusting system throttle valves with sufficient precision to achieve exact flow reference values during subsequent IST tests. Subsection ISTB of the ASME OM Code does not allow for variance from a fixed reference value for pump testing. However, NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plants:

Inservice Testing of Pumps and Valves and Inservice Examination and A3-14

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant PRR-004 CODE CASE OMN-21 REASON FOR Testing of Dynamic Restraints (Snubbers) at Nuclear Power Plants, REQUEST Revision 2, Section 5.3, acknowledges that certain pump system designs (Continued): do not allow for the licensee to set the flow at an exact value because of limitations in the instruments and controls for maintaining steady flow.

ASME OM Code Case OMN-21, Alternative Requirements for Adjusting Hydraulic Parameters to Specified Reference Points, provides guidance for adjusting reference flow or differential pressure (P) to within a specified tolerance during inservice testing.

The OM Code Case OMN-21 states, It is the opinion of the Committee that when it is impractical to operate a pump at a specified reference point and adjust the resistance of the system to a specified reference point for either flow rate, differential pressure or discharge pressure, the pump may be operated as close as practical to the specified reference point with the following requirements. The Owner shall adjust the system resistance to as close as practical to the specified reference point where the variance from the reference point does not exceed + 2% or - 1% of the reference point when the reference point is flow rate, or + 1% or - 2% of the reference point when the reference point is differential pressure or discharge pressure.

The NRC also discusses this ASME Code change in NUREG-1482, Revision 2, Section 5.3.

PROPOSED JAF seeks to perform future inservice pump testing in a manner consistent ALTERNATIVE with the requirements as stated in ASME OM Code Case OMN-21.

AND BASIS: Specifically, testing of all pumps identified in Table PRR-04 will be performed such that the flow rate is adjusted as close as practical to the reference value and within proceduralized limits of +2% / -1% of the reference flow rate or alternatively the differential pressure or discharge pressure is adjusted as close as practical to the reference value and within proceduralized limits of +1% / -2% of the reference pressure or differential pressure.

JAF plant operators will continue to strive to achieve the exact test reference values (flow or differential pressure) during testing. Typical test guidance will be to adjust the reference parameter to the specific reference value with additional guidance that if the reference value cannot be achieved with reasonable effort the test will be considered valid if the steady state flow rate is within the proceduralized limits of +2% / -1% of the reference value or the pressure or differential pressure is within the proceduralized limits of +1% / -2% of the reference value.

The ASME Operation and Maintenance Standards Committee approved Code Case OMN-21 on April 20, 2012, with the NRC representative voting in the affirmative. The applicability of Code Case OMN-21 is the ASME OM Code 1995 Edition through the 2011 Addenda. The language from Code Case OMN-21 has been included in the ASME OM Code, 2012 Edition.

Using the provisions of this request as an alternative to the specific A3-15

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant PRR-004 CODE CASE OMN-21 PROPOSED requirements of ISTB-5121, ISTB-5122, ISTB-5123, ISTB-5221, ISTB-5222 ALTERNATIVE and ISTB-5223, as described above, will provide adequate indication of AND BASIS: pump performance and continue to provide an acceptable level of quality (Continued) and safety.

Based on the determination that the use of controlled reference value ranges provides an acceptable level of quality and safety, this proposed alternative is requested pursuant to 10 CFR 50.55a(z)(1).

DURATION: 5th Interval beginning June 1, 2018, and ending September 30, 2027 PRECEDENTS: Callaway Plant, Unit 1 - Request for Relief PR-06, Alternative to ASME OM Code Requirements for IST for the Fourth Program Interval - Safety Evaluation dated July 15, 2014 (ML14178A769)

Wolf Creek Generating Station - Request for Relief 4PR-01 for the Fourth 10-Year Inservice Testing Program Interval - Safety Evaluation dated May 15, 2015 (ML15134A002)

REFERENCES:

ASME Code Case OMN-21, Alternate Requirements for Adjusting Hydraulic Parameters to Specified Reference Points NUREG-1482, Revision 2, Guidelines for Inservice Testing at Nuclear Power Plants: lnservice Testing of Pumps and Valves and Inservice Examination and Testing of Dynamic Restraints (Snubbers) at Nuclear Power Plants, Section 5.3, Allowable Variance from Reference Points and Fixed-Resistance Systems, dated October 2013 (ML13295A020)

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant PRR-004 CODE CASE OMN-21 Table PRR-04, Pumps Affected by Alternative Request PRR-04 Pump ID Description Pump Type Code OM Code (Units 1 & 2) Class Category 11P-2A Standby Liquid Control Positive Displacement 2 Group B Pumps 11P-2B 14P-1A Core Spray Pumps Centrifugal 2 Group B 14P-1B 10P-3A Residual Heat Removal Centrifugal 2 Group A (RHR)\Low Pressure Coolant 10P-3B Injection (LPCI) Pumps 10P-3C 10P-3D 23P-1M High Pressure Coolant Centrifugal 2 Group B Injection (HPCI) Main Pump 23P-1B High Pressure Coolant Centrifugal 2 Group B Injection (HPCI) Booster Pump 10P-1A RHR Service Water Pumps Vertical Line Shaft 3 Group A Centrifugal 10P-1B 10P-1C 10P-1D 46P-2A Emergency Service Water Vertical Line Shaft 3 Group B Pumps Centrifugal 46P-2B A3-17

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ATTACHMENT 4 VALVE RELIEF REQUEST INDEX

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant RELIEF REQUEST RELIEF REQUEST TITLE APPROVAL NUMBER DATE VRR-001 TIP CIV STROKE TIME 4/13/2018 VRR-002 ERCV SAMPLING 4/13/2018 VRR-003 CODE CASE OMN-17 4/13/2018 VRR-004 LEAK TEST AT APP J FREQUENCY 4/13/2018 A4 - 1

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ATTACHMENT 5 VALVE RELIEF REQUESTS

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-001 TIP CIV STROKE TIME PLANT/UNIT: James A. Fitzpatrick (JAF) Nuclear Power Plant INTERVAL: 5th Interval beginning June 1, 2018, and ending September 30, 2027 COMPONENTS Traversing In-Core Probe (TIP) Containment Isolation Valves, AFFECTED: 07SOV-104A, B, and C CODE EDITION ASME OM Code-2004 Edition with Addenda through OMb-2006 AND ADDENDA:

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

ISTC-5151(c), states, Stroke time shall be measured to at least the nearest second.

REASON FOR Pursuant to 10 CFR 50.55a(z)(2), an alternative is requested to the REQUEST: stroke time requirements of the ASME OM Code. The basis of this request is that the Code requirements present an undue hardship without a compensating increase in the level of quality or safety.

The Category A, containment isolation solenoid operated valves identified in this request have no safety function in the open direction as they open to allow the passage of the TIP assembly and drive cable for flux mapping operations. These valves have an active safety function in the closed direction in response to a primary containment isolation system signal to seal the TIP guide tubes. Therefore, an exercise test and subsequent stroke time test are only required in the closed direction.

However, 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.

The design of the TIP control system does not allow for measurement of the closure stroke times of valves 07SOV-104A, B, and C. Measuring the closure stroke times in accordance with the Code would be a hardship since it would require a costly computer control system modification. Closure of valves 07SOV-104A, B, and C could also be accomplished by an alternative method but this method would require manual extraction and retraction of the TIP from the shield block. This method of testing would be contrary to the principles of keeping radiation exposure as low as reasonably achievable because it would result in radiation exposure to personnel performing the test.

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-001 TIP CIV STROKE TIME REASON FOR The proposed alternative test ensures the operation of valves 07SOV-104A, REQUEST: B, and C in both directions and provides an acceptable level of quality and (Continued) safety. This method meets the desired outcome of monitoring valve stroke time for degradation since the computer controls the 10-second delay and the additional approximate 2 seconds for valve closure should indicate the actual stroke time.

PROPOSED JAF proposes to measure overall cycle time (opened and closed) for the ALTERNATIVE TIP Containment Isolation Valves, 07SOV-104A, B, and C, in accordance AND BASIS: with ISTC-5152. Specifically, JAF will time the opening (10-second delay time included) and closing cycle for valves 07SOV-104A, B, and C. The time from open initiation to receipt of the closed light for each valve will be monitored with a stop watch. JAF will apply ISTC-5152(a) which requires that each valve exhibit no more than +/- 25% change in stroke time when compared to the reference value except that the full stroke limiting time for each valve will be truncated at 12 seconds.

Based on the hardship resulting from implementing the OM Code requirements without a compensating increase in the level of quality and safety, this proposed alternative is requested pursuant to 10 CFR 50.55a(z)(2).

DURATION: 5th IST Interval beginning June 1, 2018, and ending September 30, 2027 PRECEDENT: This proposed 10 CFR 50.55a alternative was previously authorized for use at JAF under VRR-02 for the Interval 4 IST Program via the NRC safety evaluation dated November 27, 2007 (ADAMS Accession No. ML072910422).

REFERENCE:

Letter from Entergy Nuclear Operations, Inc., Pete Dietrich, to US Nuclear Regulatory Commission, dated July 31, 2007, Response to request for Additional Information Regarding Proposed Relief Requests for James A.

Fitzpatrick Nuclear Power Plant Fourth Interval In-Service Testing Program (ADAMS Accession No. ML072190608).

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-002 EFCV SAMPLING PLANT/UNIT: James A. Fitzpatrick (JAF) Nuclear Power Plant INTERVAL: 5th Interval beginning June 1, 2018, and ending September 30, 2027 COMPONENTS Excess Flow Check Valves (EFCVs) listed in the attached Table AFFECTED: VRR-02, Excess Flow Check Valve Component List.

CODE EDITION ASME OM Code-2004 Edition with Addenda through OMb-2006 AND ADDENDA:

REQUIREMENTS: ISTC-3522, Category C Check Valves, paragraph (c) states: If exercising is not practicable during operation at power and cold shutdowns, it shall be performed during refueling outages.

REASON FOR Pursuant to 10 CFR 50.55a(z)(1), an alternative is proposed to the REQUEST: requirements of ASME OM Code paragraph ISTC-3522(c). The basis of the request is that the proposed alternative would provide an acceptable level of quality and safety. The ASME OM Code requires check valves to be exercised quarterly during plant operation, or if valve exercising is not practicable during plant operation and cold shutdown, it shall be performed during refueling outages (RFOs).

These valves cannot be exercised closed during normal power operation since closing these valves would isolate instrumentation required for power operation. Isolation of any of these instruments from service may cause a spurious signal, which could result in a plant trip or an unnecessary challenge to safety systems. Testing on a cold shutdown frequency is impractical, considering the test condition for most of the EFCVs, requires that reactor pressure is available for testing. The appropriate time for performing EFCV testing is during RFOs, and for most of the EFCVs, in conjunction with the vessel in-service leakage (hydrostatic) testing. Recent improvements in RFO schedules minimized the time that is planned for refueling and testing activities during the outages. Considering the test conditions required for EFCV testing and the shortened outage durations, it is burdensome to test all these valves during RFOs.

Based on past experience, EFCV testing during in-service leakage testing can become the outage critical path and could possibly extend the outage if all EFCVs were to be tested during this time frame.

The testing described above requires isolation of the instruments associated with each EFCV and opening of a drain valve to actuate the EFCV. Process fluid will be contaminated to some degree, requiring special measures to collect flow from the drain valve and also contributes to an increase in personnel radiation exposure.

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-002 EFCV SAMPLING PROPOSED JAF proposes to exercise test, by full-stroke to the position required to ALTERNATIVE fulfill its function, a representative sample of EFCVs every refuel outage.

AND BASIS: The representative sample is based on approximately 20 percent of the valves each cycle such that each valve is tested at least once every 10 years (nominal).

Industry experience as documented in Boiling Water Reactor Owners Group (BWROG) report, NEDO-32977-A, (B21-00658-01), Excess Flow Check Valve Testing Relaxation, indicates that EFCVs have a very low failure rate.

The instrument lines at JAF have a flow restricting orifice upstream of the EFCVs to limit reactor water leakage in the event of rupture. The JAF Final Safety Analysis Report (FSAR), paragraph 7.1.6, Supplemental NSSS Supplier Information, does not credit the EFCVs, but instead credits the installed orifice for limiting the release of reactor coolant following an instrument line break. Thus, a failure of an EFCV, though not expected as a result of this request, is bounded by the FSAR analysis. The JAF test experience is consistent with the findings in the NEDO document. The NEDO document indicates similarly that many reported test failures at other plants were related to test methodologies and not actual EFCV failures.

EFCV failures will be documented in the JAFs 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.

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 be established based on the results of the root-cause analysis.

The basis of this request is that the proposed alternative, described above, would provide an acceptable level of quality and safety. Therefore, this proposed alternative is requested pursuant to 10 CFR 50.55a(z)(1).

DURATION: 5th IST Interval, beginning June 1, 2018, and ending September 30, 2027 A5 - 4

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-002 EFCV SAMPLING PRECEDENCE: 1) This 10CFR50.55a request was previously approved as relief request VRR-03 for the JAF Interval 4 IST Program in NRC Safety Evaluation dated November 27, 2007 (ML072910422).

2) Columbia Generating Station - Requests for Relief Nos. RG01, RP01, RP02, RP03, RP04, RP05, RP06, RV01, RV02, RV03 and RV04 for the Fourth 10-Year Inservice Testing Interval, NRC Safety Evaluation dated December 09, 2014, (ML14337A449).

REFERENCES 1. BWROG Topical Report, NEDO-32977-A (B21-00658-01), Excess Flow Check Valve Testing Relaxation, June 2000.

2. JAF FSAR, paragraph 7.1.6, Supplemental NSSS Supplier Information A5 - 5

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-002 EFCV SAMPLING Table VRR-02 Excess Flow Check Valve Component List ASME VALVE NUMBER SYSTEM OM CATEGORY CLASS 02-2EFV-PS-128A Reactor Water A/C 1 Recirculation 02-2EFV-PS-128B Reactor Water A/C 1 Recirculation 02-2EFV-PT-24A Reactor Water A/C 1 Recirculation 02-2EFV-PT-24B Reactor Water A/C 1 Recirculation 02-2EFV-PT-25A Reactor Water A/C 1 Recirculation 02-2EFV-PT-25B Reactor Water A/C 1 Recirculation 02-2EFV1-DPT-111A Reactor Water A/C 1 Recirculation 02-2EFV1-DPT-111B Reactor Water A/C 1 Recirculation 02-2EFV1-FT-110A Reactor Water A/C 1 Recirculation 02-2EFV1-FT-110C Reactor Water A/C 1 Recirculation 02-2EFV1-FT-110E Reactor Water A/C 1 Recirculation 02-2EFV1-FT-110G Reactor Water A/C 1 Recirculation 02-2EFV2-DPT-111A Reactor Water A/C 1 Recirculation 02-2EFV2-DPT-111B Reactor Water A/C 1 Recirculation 02-2EFV2-FT-110A Reactor Water A/C 1 Recirculation 02-2EFV2-FT-110C Reactor Water A/C 1 Recirculation 02-2EFV2-FT-110E Reactor Water A/C 1 Recirculation 02-2EFV2-FT-110G Reactor Water A/C 1 Recirculation 02-3EFV-11 Nuclear Boiler A/C 1 02-3EFV-13A Nuclear Boiler A/C 1 02-3EFV-13B Nuclear Boiler A/C 1 A5 - 6

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-002 EFCV SAMPLING Table VRR-02 Excess Flow Check Valve Component List VALVE NUMBER SYSTEM OM CATEGORY ASME CLASS 02-3EFV-15A Nuclear Boiler A/C 1 02-3EFV-15B Nuclear Boiler A/C 1 02-3EFV-15N Nuclear Boiler A/C 1 02-3EFV-17A Nuclear Boiler A/C 1 02-3EFV-17B Nuclear Boiler A/C 1 02-3EFV-19A Nuclear Boiler A/C 1 02-3EFV-19B Nuclear Boiler A/C 1 02-3EFV-21A Nuclear Boiler A/C 1 02-3EFV-21B Nuclear Boiler A/C 1 02-3EFV-21C Nuclear Boiler A/C 1 02-3EFV-21D Nuclear Boiler A/C 1 02-3EFV-23A Nuclear Boiler A/C 1 02-3EFV-23B Nuclear Boiler A/C 1 02-3EFV-23C Nuclear Boiler A/C 1 02-3EFV-23D Nuclear Boiler A/C 1 02-3EFV-23 Nuclear Boiler A/C 1 02-3EFV-25 Nuclear Boiler A/C 1 02-3EFV-31A Nuclear Boiler A/C 1 02-3EFV-31B Nuclear Boiler A/C 1 02-3EFV-31C Nuclear Boiler A/C 1 02-3EFV-31D Nuclear Boiler A/C 1 02-3EFV-31E Nuclear Boiler A/C 1 02-3EFV-31F Nuclear Boiler A/C 1 A5 - 7

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-002 EFCV SAMPLING Table VRR-02 Excess Flow Check Valve Component List ASME VALVE NUMBER SYSTEM OM CATEGORY CLASS 02-3EFV-31G Nuclear Boiler A/C 1 02-3EFV-31H Nuclear Boiler A/C 1 02-3EFV-31J Nuclear Boiler A/C 1 02-3EFV-31K Nuclear Boiler A/C 1 02-3EFV-31L Nuclear Boiler A/C 1 02-3EFV-31JM Nuclear Boiler A/C 1 02-3EFV-31N Nuclear Boiler A/C 1 02-3EFV-31P Nuclear Boiler A/C 1 02-3EFV-31R Nuclear Boiler A/C 1 02-3EFV-31S Nuclear Boiler A/C 1 02-3EFV-33 Nuclear Boiler A/C 1 13EFV-01A Nuclear Boiler A/C 1 13EFV-01B Nuclear Boiler A/C 1 13EFV-02A Nuclear Boiler A/C 1 13EFV-02B Nuclear Boiler A/C 1 14EFV-31A Core Spray A/C 1 14EFV-31B Core Spray A/C 1 23EFV-01A High Pressure A/C 1 Coolant Injection 23EFV-01B High Pressure A/C 1 Coolant Injection 23EFV-02A High Pressure A/C 1 Coolant Injection 23EFV-02B High Pressure A/C 1 Coolant Injection 29EFV-30A Main Steam A/C 1 29EFV-30B Main Steam A/C 1 A5 - 8

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-003 CODE CASE OMN-17 PLANT/UNIT: James A. Fitzpatrick (JAF) Nuclear Power Plant INTERVAL: 5th Interval beginning June 1, 2018, and ending September 30, 2027 COMPONENTS Main Steam Safety/Relief Valves (SRVs), 02RV-071A through AFFECTED: 02RV-071H and 02RV-071J through 02RV-071L CODE EDITION ASME OM Code-2004 Edition with Addenda through OMb-2006 AND ADDENDA:

REQUIREMENTS: ASME Code Mandatory Appendix I, paragraph I-1320(a), Test Frequencies, Class 1 Pressure Relief Valves, states, in part, that Class 1 pressure relief valves shall be tested at least once every 5 years REASON FOR Pursuant to 10 CFR 50.55a(z)(1), an alternative is requested to the REQUEST: frequency requirement for testing Class 1 pressure relief valves every 5 years. JAF proposes to test Class 1 pressure relief valves every 6 years in accordance with ASME Code Case OMN-17, Alternative Rules for Testing ASME Class 1 Pressure Relief/Safety Valves, which was published via the American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants (OM) Code 2009 Edition. JAF considers that testing the main steam SRVs in accordance with ASME OM Code Case OMN-17 provides an acceptable level of quality and safety.

PROPOSED ASME OM Code Mandatory Appendix I, Inservice Testing of Pressure ALTERNATIVE Relief Devices in Light-Water Reactor Nuclear Power Plants, paragraph AND BASIS: I-1320, Test Frequencies, Class 1 Pressure Relief Valves, (a) 5-Year Test Interval, requires that Class 1 pressure relief valves be tested at least once every 5 years. As an alternative to the Code required 5-year test interval per Mandatory Appendix I, paragraph I-1320(a), JAF proposes that Class 1 pressure relief valves, 02RV-071A, B, C, D, E, F, G, H, J, K, and L, be tested at least once every three (3) refueling cycles (approximately 6 years/72 months) with a minimum of 20% of the valves tested within any 24-month interval. This 20% would consist of valves (complete assemblies) that have not been tested during the current 72-month interval, if they exist. The test interval for any individual valve would not exceed 72 months except that a 6-month grace period is allowed to coincide with refueling outages to accommodate extended shutdown periods. JAF proposes to continue testing all eleven (11) installed pilot valves every refueling outage.

The main steam pressure relief system provides reactor coolant system (RCS) overpressure protection and automatic depressurization of the nuclear system by opening the SRVs. JAF updated Final Safety Analysis Report Section 4.4.4, Safety Design Bases, describes the main protection functions of the SRVs as follows:

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-003 CODE CASE OMN-17 PROPOSED 1. The pressure relief system prevents over-pressurization of the RCS in ALTERNATIVE order to prevent failure of the reactor coolant pressure boundary due to AND BASIS pressure.

(Continued):

2. The pressure relief system provides automatic depressurization for small breaks in the RCS so that the low pressure coolant injection and the core spray systems can operate to protect the fuel barrier.

3. The pressure relief system provides for manual depressurization at a remote auxiliary panel located outside the control room in the highly unlikely event the control room was to become uninhabitable.

Seven (7) of the eleven (11) SRVs are a designated part of the automatic depressurization system (ADS) emergency core cooling system (ECCS) and must open to provide automatic reactor depressurization as a result of a small break in the nuclear system for which the high pressure coolant injection system cannot maintain reactor water level (ADS function).

The relief valve testing and maintenance cycle at JAF consists of an as-found inspection, seat leakage and set pressure testing. After as-found set pressure testing, the valves shall be disassembled and inspected to verify that parts are free of defects resulting from time-related degradation or service induced wear. As-left set pressure testing shall be performed following maintenance and prior to returning the valve to service.

Prior to the return of a complement of SRVs for installation in the plant, the valves are disassembled and inspected to verify that internal surfaces and parts are free from defects or service induced wear prior to the start of the next test interval. During this process, anomalies or damage are identified and resolved. Damaged or worn parts, springs, gaskets and seals are replaced as necessary. Following reassembly, the valve's set pressure is recertified with an acceptance criterion of +/-1%. This existing process is in accordance with ASME OM Code Case OMN-17, paragraphs (d) and (e).

This Code Case has not been approved for use in US NRC Regulatory Guide 1.192, "Operation and Maintenance Code Case Acceptability, ASME OM Code, dated June 2003.

The JAF SRV pilot valves test history, as described in Attachment 1, Safety Relief Valve Pilot Valve Test Data, shows that the pilot valves are susceptible to disc-to-seat corrosion. This is a recurring industry issue that has been the subject of both NRC and Boiling Water Reactor Owners Group (BWROG) assessments. As described in Licensee Event Report (LER) 2015-002-00, dated July 30, 2015, and based on the known industry wide issues with the two-stage Target Rock SRVs, JAF has implemented the following industry recommendations:

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-003 CODE CASE OMN-17 PROPOSED

1. Installed Stellite 21 discs in the SRV pilot assemblies during ALTERNATIVE refurbishment; AND BASIS (Continued): 2. Installed the electric lift system recommended by the BWROG;

3. Installed enhanced insulation on the SRVs;

4. Redirected ventilation air flow away from the SRVs; and

5. Began a phased replacement of 2-stage SRVs with 3-stage.

As a result, JAF will continue to test and inspect 100% of the installed SRV pilot valves each refueling outage until such time that successful test performance is achieved. The inspection and testing of the SRV valve assemblies, performed to date, have not identified any relevant issues.

However, given the current 24-month operating cycle, JAF would be required to remove and test fifty (50) percent of the SRVs every refueling outage (i.e.

five or six of 11), such that all valves are removed and tested every two refueling outages. This would ensure compliance with the ASME OM Code requirements for testing Class 1 pressure relief valves within a five-year interval. Approval of extending the test interval for the valves to 6 years with a grace period of 6 months, consistent with Code Case OMN-17, would reduce the minimum number of SRVs tested at JAF over three refueling outages by a minimum of five.

Pursuant to 10 CFR 50.55a(z)(1), an alternative is proposed to the frequency requirement for testing Class 1 pressure relief valves every 5 years. JAF proposes to test Class 1 pressure relief valves every 6 years, with a grace period of six months, in accordance with ASME Code Case OMN-17, Alternative Rules for Testing ASME Class1 Pressure Relief/Safety Valves.

JAF proposes to continue testing all eleven (11) installed pilot valves every refueling outage until such time that successful test performance is achieved.

JAF considers that testing the main steam SRVs in accordance with ASME OM Code Case OMN-17 and the installed SRV pilot valves each refueling outage will provide an acceptable level of quality and safety.

DURATION: 5th IST Interval, beginning June 1, 2018, and ending September 30, 2027 PRECEDENTS: 1. Approved for JAF Interval 4 IST Program, as relief request VRR-06, in NRC Safety Evaluation dated October 1, 2009 (ML092730032).

2. Approved for Entergy Operation, Incorporated, in NRC SER dated March 23, 2015, River Bend Station, Unit 1 - Relief Request No. VRR-RBS-2014-1 Regarding the Third 10-year Inservice Test Interval (TAC No, MF4125), (ML15071A141).

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-003 CODE CASE OMN-17

REFERENCES:

1. JAF LER 2015-002-00, Safety Relief Valve Upward Setpoint Drift, dated July 30, 2015, (ML15212A272).

2. ASME OM Code Case OMN-17, Alternative Rules for Testing ASME Class 1 Pressure Relief/Safety Valves

3. JAF FSAR Update, Revision 5, Section 4.4, Pressure Relief System, subsection 4.4.4, Safety Design Basis A5 - 13

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-003 CODE CASE OMN-17 Attachment 1, Safety Relief Valve Pilot Valve Test Data Cycle 18 Installed As Found S/N Location Model Set Pressure Deviation (%) Set Pressure 1047 02RV-71C 2-stage 1184 3.29% 1145 1080 02RV-71E 2-stage 1187 3.54% 1145 1052 02RV-71F 2-stage No Lift 1145 1088 02RV-71A 2-stage 1148 0.26% 1145 1237 02RV-71B 2-stage 1192 3.94% 1145 1111 02RV-71D 2-stage 1157 1.04% 1145 1053 02RV-71G 2-stage No Lift 1145 1194 02RV-71H 2-stage 1169 2.05% 1145 1192 02RV-71K 2-stage 1156 0.95% 1145 1056 02RV-71J 2-stage 1168 1.97% 1145 1053 02RV-71L 2-stage 1177 2.72% 1145 1193 2-stage 1245 8.03% 1145 1238 1227 6.68% 1145 Cycle 19 As Found Installed Set S/N Location Model Pressure Deviation (%) Set Pressure 1045 02RV-71C 3-stage 1206 5.33% 1145 1191 02RV-71E 3-stage No Lift 1145 1236 02RV-71F 3-stage No Lift 1145 1087 02RV-71A 2-stage 1163 N/A 1145 1218 02RV-71B 2-stage 1208 5.50% 1145 1143 02RV-71D 2-stage 1145 1217 02RV-71G 2-stage 1192 4.10% 1145 1235 02RV-71H 2-stage No Lift 1145 1051 02RV-71K 2-stage No Lift 1145 1195 02RV-71J 2-stage 1246 8.82% 1145 1052 02RV-71L 2-stage 1163 1.57% 1145 1238 1189 3.84% 1145 1013 2-stage 1159 1.22% 1145 1110 2-stage 1164 1.66% 1145 A5 - 14

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-003 CODE CASE OMN-17 Attachment 1, Safety Relief Valve Pilot Valve Test Data (Continued)

Cycle 20 As Found Installed Set S/N Location Model Pressure Deviation (%) Set Pressure TR2 02RV-71C 3-stage 1132 -1.14% 1145 TR3 02RV-71E 3-stage 1155 0.87% 1145 TR4 02RV-71F 3-stage 1151 0.52% 1145 1088 02RV-71A 2-stage 1195 4.37% 1145 1193 02RV-71B 2-stage 1150 0.44% 1145 1194 02RV-71D 2-stage 1159 1.22% 1145 1047 02RV-71G 2-stage 1150 0.44% 1145 1192 02RV-71H 2-stage 1183 3.32% 1145 1111 02RV-71K 2-stage 1162 1.48% 1145 1080 02RV-71J 2-stage 1177 2.79% 1145 1056 02RV-71L 2-stage 1141 -0.35% 1145 Cycle 21 As Found Installed Set S/N Location Model Pressure Deviation (%) Set Pressure 11 02RV-71F 3-stage 1133 -1.05% 1145 58 02RV-71E 3-stage 1118 -2.36% 1145 61 02RV-71C 3-stage 1122 -2.01% 1145 1013 02RV-71A 2-stage 1140 -0.44% 1145 1045 02RV-71B 2-stage 1250 9.17% 1145 1052 02RV-71D 2-stage 1191 4.02% 1145 1110 02RV-71G 2-stage 1231 7.51% 1145 1195 02RV-71H 2-stage 1235 7.86% 1145 1217 02RV-71K 2-stage 1243 8.56% 1145 1218 02RV-71J 2-stage 1277 11.53% 1145 1235 02RV-71L 2-stage 1203 5.07% 1145 A5 - 15

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-04 Leak Test At APP J Frequency PLANT/UNIT: James A. Fitzpatrick (JAF) Nuclear Power Plant INTERVAL: 5th Interval beginning June 1, 2018, and ending September 30, 2027 COMPONENTS Pressure Isolation Valves (PIVs) listed in the attached Table AFFECTED: VRR-04-01, Pressure Isolation Valve Component List.

CODE EDITION ASME OM Code-2004 Edition with Addenda through OMb-2006 AND ADDENDA:

REQUIREMENTS: ISTC-3630, Leakage Rate for Other Than Containment Isolation Valves, paragraph ISTC-3630(a), Frequency, states: "Tests shall be conducted at least once every 2 years.

REASON FOR Pursuant to 10 CFR 50.55a(z)(1), an alternative is proposed to the REQUEST: frequency requirements of ASME OM Code ISTC-3630(a) for the subject valves. ISTC-3630(a) requires that leakage rate testing of Category A valves with a leakage requirement that is not based on 10CFR50 Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, be conducted at least every two years. At JAF, PIVs are Category A or Category A/C valves within the scope of ISTC-3630. This alternative, to allow for scheduling of leak tests, for the valves identified in Table VRR 01, to a performance-based frequency that is the same as 10 CFR 50, Appendix J, Option B, Performance-Based Requirements, testing, would provide an acceptable level of quality and safety.

PROPOSED JAF proposes an alternative test frequency in lieu of the requirements ALTERNATIVE found in ISTC-3630(a) for the 10 applicable PIVs listed in the attached AND BASIS: Table VRR-04-01. Specifically, JAF proposes to perform the valve leakage rate test at a frequency in accordance with the 10 CFR 50, Appendix J, Option B schedule. The identified valves were initially tested at the ISTC-3630(a) required interval schedule, which was every refueling outage (RFO) or 2 years, during the 3rd ten-year IST interval at JAF, and have subsequently been tested at the 10 CFR 50, Appendix J, Option B frequency, if applicable, during the 4th ten-year IST interval. Valves that demonstrated good performance for two consecutive cycles may have had their test interval extended to a maximum of 60 months. Any leakage rate test failure required the component to return to the initial interval of every RFO or 2 years until good performance could again be established.

10 CFR 50, Appendix J, was amended to improve the focus of the body of regulations by eliminating prescriptive requirements that are marginal to safety and to provide licensees greater flexibility for cost-effective implementation methods for meeting regulatory safety objectives. Consistent with this approach, 10CFR50, Appendix J, Option B is less prescriptive, utilizes risk-based insights and allows licensees to adopt A5 - 16

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-04 Leak Test At APP J Frequency PROPOSED cost effective methods, including setting test intervals, for implementing ALTERNATIVE safety objectives underlying the requirements of Appendix J.

AND BASIS (Continued): Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, as modified by the positions in Regulatory Guide 1.163, describes the risk-informed basis for extended test intervals under Option B. That justification documents valves which have demonstrated good leakage rate performance over two consecutive cycles are subject to future failures predominantly governed by the random failure rate of the component. NEI 94-01 also presents the results of a comprehensive risk analysis, including the statement that "the risk impact associated with increasing test intervals is negligible (less than 0.1 percent of total risk)".

NUREG 0933, Resolution of Generic Safety Issues, Issue 105, Interfacing Systems LOCA at LWRs, discussed the need for PIV leak rate testing based primarily on three pre-1980 historical failures of applicable valves industry-wide. These failures all involved human errors in either operations or maintenance. None of these failures involved inservice equipment degradation. The performance of PIV leak rate testing provides assurance of acceptable seat leakage with the valve in a closed condition. Typical PIV testing does not identify functional problems which may inhibit the valves ability to re-position from open to close.

For check valves, such functional testing is accomplished per ASME OM Code ISTC-3520, Exercising Requirements, and ISTC-3522, Category C Check Valves. Power-operated valves are routinely full stroke tested per ASME OM Code to ensure their functional capabilities. The functional tests for PIVs are performed only at an RFO frequency. Such testing is not performed online in order to prevent any possibility of an inadvertent Interfacing System Loss of Coolant Accident (ISLOCA) condition. The functional testing of the PIVs is adequate to identify any abnormal condition that might affect closure capability. Performance of the separate PIV leak rate testing does not contribute any additional assurance of functional capability; it verifies the seat tightness of the closed valves. JAF proposes to perform PIV testing at intervals specified in NEI 94-01. The specific interval for each valve would be a function of its performance and would be established in a manner consistent with the Containment Isolation valve (CIV) process under 10 CFR 50, Appendix J, Option B. Program guidance will be established such that if any of the valves fail either the CIV test or PIV test, the test interval for both tests will be reduced to once every 30 months until they can be re-classified as good performers per the performance evaluation requirements of Appendix J, Option B. The test intervals for the valves identified in this request will be determined in the same manner as is done for CIV testing under Option B. That is, the test interval may be extended upon completion of two consecutive periodic PIV tests with results within the A5 - 17

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-04 Leak Test At APP J Frequency PROPOSED prescribed acceptance criteria. Any PIV test failure will require a return ALTERNATIVE to the initial interval until good performance can again be established.

AND BASIS (Continued): The intent of this request is to allow for a performance-based approach to the scheduling of PIV leakage testing. It has been shown that ISLOCA represents a small risk impact to BWRs such as JAF. NUREG 0933 Issue 105, references the conclusion from NUREG/CR-5928, "Final Report of the NRC-sponsored ISLOCA Research Program," that concludes, for the units studied, ISLOCA poses little risk.

The PIVs have an excellent performance history in terms of seat leakage testing (See Attachment 1, Pressure Isolation Valve Component Leakage History). It should be noted that component tests identified in Attachment 1 as LJ-C are performed using air as the test media. Alternatively, component tests identified in Attachment 1 as LKO are performed using water as the test media.

The risks associated with extending the leakage test interval to a maximum of sixty months are extremely low. The basis for this alternative request is the historically good performance of the PIVs. This alternative will also provide significant reductions in radiation dose. Approximately 500 mRem per outage is received when a full complement of PIV tests is required. The last RFO radiation exposure for which a full complement of PIVs was tested was used to identify that PIV testing alone each RFO resulted in a total dose to personnel of approximately 500 mRem.

It should be noted that JAF submitted a license amendment request (LAR) on August 29, 2016. The LAR proposes to change the plant Technical Specification (TS) 5.5.6 by replacing the reference to Regulatory Guide 1.163 with a reference to the NEI 94-01, Revision 3-A, topical report, dated July 2, 2012. This proposed change will allow an extension from the 60-month frequency currently permitted by Option B to a 75-Month frequency for type C leakage rate testing. Upon approval of this LAR, JAF will implement the 75-month testing frequency for those PIVs that are CIVs as listed in this alternative request and which demonstrate acceptable performance over two consecutive cycles as defined by the measured leak rate being less than the leak rate acceptance criteria for PIV and CIV testing.

Based on the determination that the scheduling of leak tests, for the valves identified in this request, to a performance based frequency that is the same as 10 CFR 50, Appendix J, Option B testing, would provide an acceptable level of quality and safety, this proposed alternative is requested pursuant to 10 CFR 50.55a(z)(1).

DURATION: 5th IST Interval, beginning June 1, 2018, and ending September 30, 2027 PRECEDENT: This proposed 10 CFR 50.55a alternative was previously authorized for use at JAF under VRR-07 for the Interval 4 IST Program in NRC Safety Evaluation dated March 16, 2012 (ML12072A113).

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SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-04 Leak Test At APP J Frequency

REFERENCES:

1. NUREG 0933, Resolution of Generic Safety Issues, Issue 105, Interfacing Systems LOCA at LWRs

2. NUREG/CR-5928, ISLOCA Research Program, (ML072430731)

3. 10 CFR 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

4. NEI 94-01, Rev 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J

5. RG 1.163, Performance-Based Containment Leak-Test Program," dated September 1995 (ML003740058).

6. LAR submitted to NRC August 2016 to incorporate reference to NEI 94-01, Rev. 3-A A5 - 19

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-04 Leak Test At APP J Frequency Table VRR-04-01 Pressure Isolation Valve Component List APPENDIX J VALVE DESCRIPTION OPTION B OM CATEGORY NUMBER AIR TESTED 10AOV-68A RHR A LPCI TESTABLE No A/C CHECK VALVE 10AOV-68B RHR B LPCI TESTABLE No A/C CHECK VALVE 10MOV-25A RHR A LPCI INBD INJ Yes A VALVE 10MOV-25B RHR B LPCI INBD INJ Yes A VALVE 14AOV-13A CSP A REACTOR ISOL No A/C TESTABLE CHECK VALVE 14AOV-13B CSP B REACTOR ISOL No A/C TESTABLE CHECK VALVE 14MOV-12A CORE SPRAY LOOP A INBD Yes A ISOL VALVE 14MOV-12B CORE SPRAY LOOP B INBD Yes A ISOL VALVE 10MOV-17 RHR SHUTDOWN COOLING Yes A OUTBD ISOL VALVE 10MOV-18 RHR SHUTDOWN COOLING Yes A INBD ISOL VALVE A5 - 20

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-04 Leak Test At APP J Frequency - Pressure Isolation Valve Component Leakage History 10AOV-68A Leakage History Test Date Test Type Test Result Leakage Leakage Limit Leakage Units 10/10/2002 LKO Sat 0.5 10 GPM 10/9/2004 LKO Sat 0 10 GPM 10/29/2006 LKO Sat 0 10 GPM 9/29/2008 LKO Sat 0 10 GPM 9/19/2010 LKO Sat 0 10 GPM 9/25/2012 LKO Sat 0 10 GPM 8/30/2014 LKO Sat 0 10 GPM 10AOV-68B Leakage History Test Date Test Type Test Result Leakage Leakage Limit Leakage Units 10/17/2002 LKO Sat 0 10 GPM 9/29/2004 LKO Sat 8.5 10 GPM 10/29/2006 LKO Sat 0 10 GPM 9/21/2008 LKO Sat 0 10 GPM 9/29/2010 LKO Sat *1 10 GPM 9/21/2012 LKO Sat 0.11 10 GPM 9/12/2014 LKO Sat 0 10 GPM

1 - Documentation lost and test results not recorded A5 - 21

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-04 Leak Test At APP J Frequency - Pressure Isolation Valve Component Leakage History (Continued) 10MOV-25A Leakage History Test Date Test Type Test Result Leakage Leakage Limit Leakage Units 10/10/2004 LJ-C Sat 1.184 16 SLM 9/30/2008 LJ-C Sat 5.96 5.893 SLM 9/25/2012 LJ-C Sat 0.35 5.893 SLM 8/30/2014 LJ-C Sat 0.45 5.893 SLM 9/3/2002 LKO Sat 0 5 GPM 8/30/2006 LKO Sat 0 5 GPM 9/10/2010 LKO Sat 0.24 5 GPM 6/21/2012 LKO Sat 0 5 GPM 8/10/2016 LKO Sat 0 5 GPM 10MOV-25B Leakage History Test Date Test Type Test Result Leakage Leakage Limit Leakage Units 10/5/2004 LJ-C Sat 3.85 5.893 SLM 10/26/2006 LJ-C Sat 8.85 5.893 SLM 9/16/2008 LJ-C Sat 1.53 16 SLM 9/23/2008 LJ-C Sat 4.88 5.893 SLM 9/23/2012 LJ-C Sat 0.62 5.893 SLM 9/29/2012 LJ-C Sat 10.4 5.893 SLM 9/3/2002 LKO Sat 0 10 GPM 9/3/2006 LKO Sat 0 10 GPM 8/29/2010 LKO Sat 0 10 GPM 11/20/2016 LKO Sat 0 10 GPM A5 - 22

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant VRR-04 Leak Test At APP J Frequency - Pressure Isolation Valve Component Leakage History (Continued) 14AOV-13A Leakage History Test Date Test Type Test Result Leakage Leakage Limit Leakage Units 10/8/2002 LKO Sat 0 10 GPM 10/9/2004 LKO Sat 0 10 GPM 10/14/2006 LKO Sat 0 10 GPM 9/22/2008 LKO Sat 0 10 GPM 9/20/2010 LKO Sat 0 10 GPM 10/5/2012 LKO Sat 0 10 GPM 8/26/2014 LKO Sat 0 10 GPM A5 - 23

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ATTACHMENT 6 RELIEF REQUEST RAIs AND SERs

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 13, 2018 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT - ISSUANCE OF RELIEF REQUESTS FOR ALTERNATIVES TO CERTAIN ASME 0M CODE REQUIREMENTS (CAC NOS. MG0052, MG0053, MG0054, MG0055, MG0056, MG0057, MG0058, AND MG0061*, EPID L-2017-LLR-0067, L-2017-LLR-0068, L-2017-LLR-0069, L-2017-LLR-0070, L-2017-LLR-0071 , L-2017-LLR-0072, L-2017-LLR-0073, AND 1-2017-LLR-0074)

Dear Mr. Hanson:

By letter dated August 3, 2017, as supplemented by letter dated October 26, 2017 (Agency wide Documents Access and Management System (ADAMS) Accession Nos. ML17219A123 and ML17299A560, respectively), Exelon Generation Company, LLC (Exelon or the licensee) submitted eight relief requests to the U.S. Nuclear Regulatory Commission (NRC) for relief from certain requirements of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (0M Code), 2004 Edition with Addenda through Omb-2006, at the James A. FitzPatrick Nuclear Power Plant (FitzPatrick) associated with the fifth 10-year inservice testing interval.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1 ), the licensee requested to use the proposed alternative on the basis that the alternative provides an acceptable level of quality and safety for Relief Requests PRR-01 , PRR-02, PRR-04, VRR-02, VRR-03, and VRR-04. Pursuant to 10 CFR 50.55a(z)(2) the licensee requested to use the proposed alternative on the basis that complying with the specified requirement would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety for Relief Requests PRR-03 and VRR-01 The NRC staff has reviewed the subject requests and concludes, as set forth in the enclosed safety evaluations, that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1) and 10 CFR 50.55a(z)(2).

The NRC staff has determined that for requests PRR-01 , PRR-02, PRR-04, VRR-02, VRR-03, and VRR-04, the proposed alternatives provide an acceptable level of quality and safety.

Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1 ) for these proposed alternatives.

Therefore, the NRC staff authorizes the use of the proposed alternatives for requests PRR-01, B. Hanson A6-1

PRR-02, PRR-04, VRR-02, VRR-03, and VRR-04 for FitzPatrick for the fifth 10-year inservice testing interval at FitzPatrick, which is scheduled to begin on June 1, 2018, and end on September 30, 2027.

The NRC staff has determined that for requests PRR-03 and VRR-01, the proposed alternatives provide reasonable assurance that the affected components are operationally ready.

Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2) for these requests. Therefore, the NRC staff authorizes the use of the proposed alternatives for requests PRR-03 and VRR-01 for FitzPatrick for the fifth 10-year inservice testing interval at FitzPatrick, which is scheduled to begin on June 1, 2018, and end on September 30, 2027.

All other ASME 0M Code requirements for which relief was not specifically requested and approved in the subject requests remain applicable.

If you have any questions, please contact Tanya Hood at 301-415-1387 or Tanya.Hood@nrc.qov.

Sincerely, Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-333

Enclosures:

1. Safety Evaluation for PRR-01 and PRR-04

2. Safety Evaluation for PRR-02 and PRR-03

3. Safety Evaluation for VRR-03

4. Safety Evaluation for VRR-01, VRR-02, and VRR-04 cc: Listserv A6-2

UNITED STATES RE G NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUESTS PRR-01, REVISION 0, AND PRR-04, REVISION 0 FIFTH 10-YEAR INTERVAL INSERVICE TESTING PROGRAM EXELON GENERATING COMPANY, LLC JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 1 .0 INTRODUCTION By letter dated August 3, 2017, as supplemented by letter dated October 26, 2017 Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML17219A123 and ML17299A560, respectively), Exelon Generation Company, LLC (Exelon or the licensee) submitted Relief Requests PRR-01, Revision 0, and PRR-04, Revision 0, to the U.S. Nuclear Regulatory Commission (NRC or the Commission). The licensee requested alternative tests in lieu of certain inservice testing (IST) requirements of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (0M Code) for the IST program at the James A. FitzPatrick Nuclear Power Plant (FitzPatrick) during the fifth 10-year IST program interval, which is scheduled to begin on June 1, 2018, and end on September 30, 2027. Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee requested to use the proposed alternative in Relief Requests PRR-01 and PRR-04 on the basis that the alternative provides an acceptable level of quality and safety.

2.0 REGULATORY EVALUATION

Section 50.55a(f), "Preservice and Inservice testing requirements," of 10 CFR requires, in part, that IST of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME 0M Code and applicable addenda, except where alternatives have been authorized pursuant to paragraphs 10 CFR 50.55a(z)(1) or 10 CFR 50.55a(z)(2).

Section 50.55a(z) of 10 CFR requires that alternatives to the requirements of paragraphs (b) through (h) of 10 CFR 50.55a, or portions thereof, must be submitted and authorized by the NRC prior to implementation. The applicant or licensee must demonstrate that: (1) the proposed alternative would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. Section 50.55a of 10 CFR allows the NRC to authorize alternatives and to grant relief from ASME Code requirements upon making the necessary findings. Enclosure 1 A6-3

The guidance that the NRC staff considered in its review is NUREG-1482, Revision 2, "Guidelines for Inservice Testing at Nuclear Power Plants," October 2013 (ADAMS Accession No. ML13295A020), which provides acceptable guidance to licensees to establish a basic understanding of the regulatory basis for pump and valve IST programs.

Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request, and the Commission to authorize, the alternative requested by the licensee.

3.0 TECHNICAL EVALUATION

The applicable ASME 0M Code edition and addenda for FitzPatrick during the fifth 10-year IST program interval is the 2004 Edition through the 2006 Addenda.

3.1Licensee's Relief Request PRR-01, Revision O Applicable Code Requirements Subsection IST B-3510(a), "Accuracy," states, in part, "Instrument accuracy shall be within the limits of Table ISTB-3510-1 ."

Subsection ISTB-3510(b), "Range," (1) states, "The full-scale range of each analog instrument shall be not greater than three times the reference value."

Table IST B-3510-1 , "Required Instrument Accuracy," requires pressure and differential pressure accuracy to be +/- 2 percent for pump Group A and Group B tests.

Components for Which Relief is Requested Core Spray Pump 14P-1A Core Spray Pump 14P-1B Reason for Alternative Request The differential pressure for each core spray pump 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 suction pressure gauge full-scale range, 72.28 pounds per square inch gauge (psig), is sufficient for a post-accident condition when the torus is at the maximum accident pressure. This range exceeds the range limit of 15 psig 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 percent of full scale. However, the installed suction pressure instrument loop is calibrated to within +/- 2.4 percent full-scale accuracy. The full-scale range of the pump discharge pressure instrumentation loop is 500 psig. Pump discharge pressure during testing is typically 280 psig. The maximum variation due to inaccuracy in measured suction pressure is +/-

1.7 psi. The maximum variation due to inaccuracy in measured discharge pressure is

+/- 10 psi. Thus, the differential pressure (typical [Pdisch - Psuct]) would be 280 5, which is 275 +/- 11.7 psi, for an inaccuracy of 4.3 percent. If the full-scale range of the suction pressure gauge is within the ASME 0M Code allowable value of three times the reference value A6-4

(15 psig), the maximum variation due to inaccuracy in measured suction pressure would be

+/- 0.3 psi, and the resulting differential pressure measurement would be 275 +/- 10.3 psi, or an inaccuracy of 3.7 percent.

The decrease in inaccuracy of 0.6 percent by using the ASME 0M Code-compliant gauge is insignificant and does not warrant manpower and exposure to change the suction pressure quarterly for test purposes.

In addition, the ASME 0M Code would allow a full-scale range for the discharge pressure measurement of 840 psig for the typical 280 psig discharge pressure. This would translate into a differential pressure measurement of 275 +/- 17.1 psig, or an inaccuracy of 6.2 percent, if the installed instrument met the ASME 0M Code requirements of O - 15 psig for suction pressure gauge and O - 840 psig for the discharge pressure gauge. The existing measurement is significantly better than the maximum ASME 0M Code allowable inaccuracy.

Proposed Alternative Testinq The licensee stated that the existing installed plant suction pressure and discharge gauges will be used to determine the pump differential pressure for the core spray pumps 14P-1A and 14P-1B Group B pump tests. The comprehensive tests are performed using measuring and test equipment.

NRC Staff Evaluation

The NRC staff reviewed the information in Relief Request PRR-01, Revision O, including the supplemental information in the October 26, 2017, submittal. For Group A and Group B tests, the ASME 0M Code requires instrument accuracy to be within +/- 2 percent of full scale and the full scale range of each instrument be no greater than three times the reference value. The combination of these two requirements results in an effective accuracy requirement of approximately +/-6 percent of the reference value.

Based on its review, the NRC staff finds that the maximum inaccuracy of the installed suction and discharge pressure instruments combined is +/- 7.3 psig [(+/- 1.73 psig) + (+/- 5.6 psig)]. The accuracies of the installed core spray pump suction and discharge instruments yield differential pressure readings that are more accurate than the readings achieved from instruments that meet the ASME 0M Code requirements [(+/- 16.8 psig) + (+/- 0.3 psig), or +/- 17.1 psig] and, thus, provide an acceptable level of quality and safety.

The use of the existing pressure instrument is supported by NUREG-1482, Revision 2, paragraph 5.5.1 , which states that that the NRC staff may authorize an alternative when the combination of range and accuracy yields a reading at least equivalent to the reading achieved from the instrument that meets the ASME 0M Code requirements. Therefore, the NRC staff concludes that the proposed alternative provides an acceptable level of quality and safety.

3.2 Licensee's Relief Request PRR-04 Applicable Code Requirements Subsection IST B-5121 , "Group A Test Procedure," (b) states, in part, "The resistance of the system shall be varied until the flow rate equals the reference point. Alternatively, the flow rate shall be varied until the differential pressure equals the reference point..."

Subsection ISTB-5122, "Group B Test Procedure," (c), states, "System resistance may be varied as necessary to achieve the reference point."

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Subsection ISTB-5123, "Comprehensive Test Procedure," (b), states, in part, "the resistance of the system shall be varied until the flow rate equals the reference point. Alternatively, the flow rate shall be varied until the differential pressure equals the reference point..."

Subsection ISTB-5221, "Group A Test Procedure," (b), states, in part, "The resistance of the system shall be varied until the flow rate equals the reference point. Alternatively, the flow rate shall be varied until the differential pressure equals the reference point... "

Subsection ISTB-5222, "Group B Test Procedure," (c), states, "System resistance may be varied as necessary to achieve the reference point."

Subsection ISTB-5223, "Comprehensive Test Procedure," paragraph IST B-5123(b) states, in part, "The resistance of the system shall be varied until the flow rate equals the reference point, Alternatively, the flow rate shall be varied until the differential pressure equals the reference point..

Components for Which Relief is Requested Table 1 Pump Pump Description Pump Type Code 0M Code ID Class Category 11P-2A Positive Standby Liquid Control Pumps 2 Group B 11P-2B Displacement 14P-1A Core Spray Pumps Centrifugal 2 Group B 14P-1B 10P-3A Residual Heat Removal (RHR)/Low 10P-3B Pressure Coolant Injection (LPCI) Centrifugal 2 Group A 10P-3C Pumps 10P-3D High Pressure Coolant Injection 23P-1M Centrifugal 2 Group B (HPCI) Main Pump 23P-1B HPCI Booster Pump Centrifugal 2 Group B 10P-1A Vertical Line 10P-1B 10P-1C RHR Service Water Pump Shaft 3 Group A 10P-1D Centrifugal Vertical Line 46P-2A Emergency Service Water Pumps Shaft 3 Group B Centrifugal Reason for Alternative Request For pump testing, there is difficulty adjusting system throttle valves with sufficient precision to achieve an exact flow rate, differential pressure, or discharge pressure during subsequent IST tests. Subsection ISTB of the ASME 0M Code does not allow for variance from a fixed reference value for pump testing. However, NUREG-1482, Revision 2, Section 5.3, acknowledges that certain pump system designs do not allow for the licensee to set the flow at an exact value because of limitations in the instruments and controls for maintaining steady flow.

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The ASME 0M Code Case OMN-21, "Alternative Requirements for Adjusting Hydraulic Parameters to Specified Reference Points," provides guidance for adjusting reference flow, differential pressure (AP), or discharge pressure to within a specified tolerance during pump inservice testing. The ASME 0M Code Case OMN-21 states, in part:

It is the opinion of the Committee that when it is impractical to operate a pump at a specified reference point and adjust the resistance of the system to a specified reference point for flow rate, differential pressure or discharge pressure, the pump may be operated as close as practical to the specified reference point with the following requirements: The Owner shall adjust the system resistance to as close as practical to the specified reference point where the variance from the reference point does not exceed +2% or -1 % of the reference point when the reference point is flow rate, or +1% or -2% of the reference point when the reference point is differential pressure or discharge pressure.

Proposed Alternative Testing The licensee seeks to perform future inservice pump testing in a manner consistent with the requirements as stated in ASME 0M Code Case OMN-21. Specifically, testing of all pumps identified in Table 1 above will be performed such that the flow rate is adjusted as close as practical to the reference value and within proceduralized limits of +2 percent or -1 percent of the reference flow rate, or alternatively, the differential pressure or discharge pressure is adjusted as close as practical to the reference value and within proceduralized limits of +1 percent or -2 percent of the reference discharge pressure or differential pressure.

The FitzPatrick plant operators will continue to strive to achieve the exact test reference values (flow, differential pressure, or discharge pressure) during testing. Typical test guidance will be to adjust the reference parameter (i.e., flow, differential pressure, or discharge pressure) to the specific reference value with additional guidance that if the reference value cannot be achieved with reasonable effort, the test will be considered valid if the steady-state flow rate is within the proceduralized limits of +2 percent or -1 percent of the reference value, or the steady-state discharge pressure or differential pressure is within the proceduralized limits of +1 percent or -

2 percent of the reference value.

NRC Staff Evaluation

The NRC staff reviewed the information in Relief Request PRR-04, Revision O. The ASME 0M Code Case OMN-21 was developed to provide guidance on alternatives when it is impractical to operate a pump at a specified reference point for either flow rate, differential pressure, or discharge pressure. The pump may be operated as close as practical to the specified reference point with the following requirements.

Code Case OMN-21 was approved by the ASME Operation and Maintenance Standards Committee on April 20, 2012, with the NRC representative voting in the affirmative. The licensee proposed to adopt ASME 0M Code Case OMN-21. The applicability of Code Case OMN-21 is the ASME 0M Code, 1995 Edition through the 201 1 Addenda. The NRC staff notes that the language from ASME 0M Code Case OMN-21 has subsequently been included in the ASME 0M code, 2012 Edition, 2015 Edition, and 2017 Edition.

Based on its review, the NRC staff notes that in certain situations, it is not possible to operate a pump at a precise reference point. The NRC staff has reviewed the alternatives proposed in ASME 0M Code Case OMN-21 and found that the proposed alternatives are reasonable and appropriate when a pump cannot be operated at a specified reference point. Operation within A6-7

the tolerance bands specified in ASME 0M Code Case OMN-21 provides reasonable assurance that licensees will be able to utilize the data collected to detect degradation of the pumps. Based on the NRC staff's review of ASME 0M Code Case OMN-21, and the licensee's commitment to use the bands specified in ASME 0M Code Case OMN-21 for flow rate, the NRC staff concludes that implementation of the alternatives contained in ASME 0M Code Case OMN-21 is acceptable for the pumps listed above. Therefore, the NRC staff concludes that the licensee's proposed alternative provides an acceptable level of quality and safety.

4.0 CONCLUSION

As set forth above, the NRC staff determines that for Relief Requests PRR-01 and PRR-04, the proposed alternatives provide an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1) for Relief Requests PRR-01 and PRR-04. Therefore, the NRC staff authorizes the use of the alternative requests for FitzPatrick for the fifth 10-year IST program interval, which is scheduled to begin on June 1, 2018, and end on September 30, 2027.

All other ASME 0M Code requirements for which relief was not specifically requested and approved in the subject requests for relief remain applicable.

Principal Contributor: Gurjendra Bedi Date: April 13, 2018.

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REGue UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUESTS PRR-02, REVISION 0, AND PRR-03, REVISION 0, FIFTH 10-YEAR INTERVAL INSERVICE TESTING PROGRAM JAMES A. FITZPATRICK NUCLEAR POWER PLANT EXELON GENERATION COMPANY, LLC DOCKET NO. 50-333

1.0 INTRODUCTION

By letter dated August 3, 2017, as supplemented by letter dated October 26, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML17219A123 and ML17299A560, respectively), Exelon Generation Company, LLC (Exelon or the licensee) submitted Relief Requests PRR-02, Revision 0, and PRR-03, Revision 0, to the U.S. Nuclear Regulatory Commission (NRC or the Commission). The licensee requested alternative tests in lieu of certain inservice testing (IST) requirements of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (0M Code) for the IST program at the James A. FitzPatrick Nuclear Power Plant (FitzPatrick) during the fifth 10-year IST program interval, which is scheduled to begin on June 1, 2018, and end on September 30, 2027.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) the licensee requested to use the proposed alternative in Relief Request PRR-02, Revision 0, on the basis that the alternative provides an acceptable level of quality and safety. Pursuant to 10 CFR 50.55a(z)(2), the licensee requested to use the proposed alternative in Relief Request PRR-03, Revision 0, on the basis that the ASME 0M Code requirements present an undue hardship, without a compensating increase in the level of quality or safety.

2.0 REGULATORY EVALUATION

Section 50.55a(f), "Preservice and Inservice testing requirements," of 10 CFR requires, in part, that IST of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME 0M Code and applicable addenda, except where alternatives have been authorized pursuant to paragraphs 10 CFR50.55a(z)(1) or 10 CFR50.55a(z)(2).

Section 50.55a(z) of 10 CFR requires that alternatives to the requirements of paragraphs (b) through (h) of 10 CFR 50.55a, or portions thereof, must be submitted and authorized by the NRC prior to implementation. The applicant or licensee must demonstrate that: (1) the proposed alternative would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty, without a A 6-9

Enclosure 2 compensating increase in the level of quality and safety. Section 50.55a of 10 CFR allows the NRC to authorize alternatives and to grant relief from ASME Code requirements upon making the necessary findings.

The guidance that the NRC staff considered in its review is NUREG/CP-0152, "Proceedings of the Fourth NRC/ASME Symposium on Valve and Pump Testing," dated July 15-18, 1996 (ADAMS Accession No. ML18057B547), which provides acceptable guidance to licensees to address safety and technical issues related to valves and pumps.

Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request, and the Commission to authorize, the alternatives requested by the licensee.

3.0 TECHNICAL EVALUATION

The applicable ASME 0M Code edition and addenda for FitzPatrick during the fifth 10-year IST program interval is the 2004 Edition through the 2006 Addenda.

3.1 Licensee's Relief Request PRR-02, Revision 0 Applicable Code Requirements Subsection ISTB-3300, "Reference Values," (a) states, "Initial reference values shall be determined from the results of testing meeting the requirements of IST B-3100, Preservice Testing, or from the results of the first inservice test."

Components for Which Relief is Requested Table 1 Pump ID Pump Description ASME Code Class ASME 0M Pump Group Residual Heat Removal Service Water A 10P-1A 3 (RHRSW) Pump 10P-1B RHRSW Pump 3 A 10P-1C RHRSW Pump 3 A 10P-1D RHRSW Pump 3 A Reason for Alternative Request The smooth running pumps listed in Table 1 have at least one vibration reference value (Vr) that is currently less than 0.05 inches per second (ips). A small value for Vr produces a small acceptable range for pump operation. The ASME 0M Code acceptable range limit for pump vibrations from Table ISTB-5221-1, "Vertical Line Shaft Centrifugal Pumps Test Acceptance Criteria," 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.

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Subsection ISTB-6200, "Corrective Action," (a), "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 IST B-3400 shall be doubled until the cause of the deviation is determined and the condition is corrected.

For very small vibration reference values, a significant portion of the vibration reading can be from flow variations, hydraulic noise, and instrument error, which can affect the repeatability of subsequent measurements. Also, experience gathered by the licensee's Predictive Maintenance (PdM) group has shown that changes in vibration levels in the range of 0.05 ips do not normally indicate significant degradation in pump performance.

Proposed Alternative Testing The licensee seeks to apply a minimum value for Vr of 0.05 ips for the particular vibration measurement location. This minimum value would be applied to individual vibration locations for the residual heat removal service water (RHRSW) pumps listed in Table 1. The subsequent test results for that location will be compared to an alert range limit of 0.125 ips and a required action limit of 0.300 ips. These ranges, resulting from the proposed Vr of 0.05 ips and using the existing ASME 0M Code multipliers, shall be applied to vibration test results during both Group A tests and Comprehensive tests.

In addition to the requirements of the ASME 0M Code subsection IST B for IST, the pumps in the FitzPatrick IST program are also included in the FitzPatrick PdM program. The FitzPatrick PdM program currently uses predictive monitoring techniques such as vibration monitoring and analysis beyond that required by subsection ISTB bearing temperature trending, oil sampling and analysis, and/or thermography and 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 the following: initiate an issue report, increase monitoring to establish a rate of change, review component-specific information to identify the cause, and remove the affected pump from service to perform maintenance.

The licensee stated that the pumps in the IST program will remain in the FitzPatrick PdM program, even if certain pumps have very low vibration readings and are considered to be smooth running pumps.

NRC Staff Evaluation

The NRC staff reviewed the information in Relief Request PRR-02, Revision 0. The ASME 0M Code requires that the vibration of all pumps in a plant's IST program be measured. For centrifugal pumps, the measurements of each pump are taken in a plane approximately perpendicular to the rotating shaft in two orthogonal directions on each accessible pump-bearing housing. For vertical line shaft pumps, the vibration measurements are taken on the upper motor-bearing housing in three orthogonal directions, including the axial direction. The measurement is also taken in the axial direction on each accessible pump thrust-bearing housing. These measurements are to be compared with the ASME 0M Code vibration acceptance criteria to determine if the measured values are acceptable.

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Subsection ISTB requires that, if during an inservice test, a bearing vibration measurement exceeds 2.5 Vr, the pump is considered in the alert range. The frequency of testing is then doubled until the condition is corrected and the vibration level returns below the alert range.

Pumps whose vibration is recorded to be 6 Vr are considered in the required action range and must be declared inoperable until the cause of the deviation has been determined and the condition corrected. The vibration reference values are required to be determined when the pump is in good operating condition.

For pumps whose absolute magnitude of vibration is an order of magnitude below the absolute vibration limits established in subsection ISTB, a relatively small increase in vibration magnitude may cause the pump to enter the alert or required action range. These instances may be attributed to variation in flow, instrument accuracy, or other noise sources that would not be associated with degradation of the pump. Pumps that operate in this region are typically referred to as smooth running pumps. Based on a small acceptable range, a smooth running pump could be subject to unnecessary corrective action and additional testing.

Based on its review, the NRC staff finds that the alert and required action limits specified in the alternative request sufficiently address the previously undetected acute pump problems, and the licensee's PdM program appears to be designed to detect problems involving the mechanical condition, even well in advance of when the pump reaches its overall vibration alert limit.

Based on the experience gathered by the FitzPatrick PdM group, the licensee has proposed to establish a reference value of 0.05 ips. The use of the suggested reference value of 0.05 ips will provide an alert range of 0.125 to 0.30 ips, and the licensee's PdM program has shown that changes in vibration levels below 0.05 ips do not normally indicate significant degradation in pump performance. The reference value of 0.05 ips is consistent with previous NRC staff safety evaluations of similar issues. This alternative request js not for relief from the requirement to establish reference values, but from the method of determining the reference value. Therefore, the NRC staff concludes that the licensee's proposed alternative will provide an acceptable level of quality and safety.

3.2Licensee's Alternative Request PRR-03, Revision 0 Applicable Code Requirements Table IST B-5121-1, "Centrifugal Pump Test Acceptance Criteria," provides the vibration alert range low-end absolute limit of 0.325 ips for the Group A and Comprehensive tests.

Components for Which Relief is Requested Table 2 ASME OM Pump Pump ID Pump Description ASME Code Class Group 10P-3A Residual Heat Removal (RHR) Pump 2 A 10P-3B RHR Pump 2 A 10P-3C RHR Pump 2 A 10P-3D RHR Pump 2 A A6-12

Reason for Alternative Request The increased periodicity of testing resulting from the 0.325 ips ASME 0M Code requirement is also an increase to the licensee's operations staff, plant scheduling, and adds run time to all RHR pumps. This request is based on analysis of vibration and pump differential pressure data indicating that no pump degradation is taking place.

Proposed Alternative Testing The licensee seeks to apply an alternative vibration alert range low-end absolute limit of 0.408 ips for the RHR pumps listed in Table 2. The required action level for vibration will remain unchanged (> 6 Vr). The RHR pumps listed in Table 2 are tested using a full-flow recirculation test line back to the suppression pool for each Group A test and comprehensive pump tests.

Based on the full-flow test line configuration, this test methodology results in flow-induced, broadband vibration readings greater than 0.325 ips, but less than the required action limits.

The guidance in NUREG/CP-OI 52 presented four key components that should be addressed in an alternative request of this type to streamline the review process. These four key components are as follows:

The licensee should have sufficient vibration history from the inservice testing, which verifies that the pump has operated at the vibration level for a significant amount of time, with any "spikes" in the data justified.

ll. The licensee should have consulted with the pump manufacturer or vibration expert about the level of vibration the pump is experiencing to determine if the pump operation is acceptable.

The licensee should describe attempts to lower the vibration below the defined ASME 0M Code absolute levels through modifications to the pump.

The licensee should perform a spectral analysis of the pump driver system to identify all contributors to the vibration levels.

The licensee provided a discussion of how it addressed these four key components in its submittal. Expert analyses and maintenance reviews have shown that this vibration has not resulted in degradation to the pump or motor. Data trends show that overall vibrations have remained steady since 1998.

The new Alert criterion value allows an alternative measure that still meets the intended function of monitoring the pump for degradation, while leaving the action levels as mandated by the ASME 0M Code. The proposed criterion encompasses the previous values that exceeded the Alert level, which would eliminate the unnecessary actions associated with exceeding the ASME 0M Code Alert limit when the pump is not seen as degrading. Any corrective actions triggered by vibrations between 0.408 ips to 0.7 ips will result in the same ASME 0M Code actions as previously required when exceeding the Alert limit of 0.325 ips.

NRC Staff Evaluation

The NRC staff reviewed the information in Relief Request PRR-03, Revision 0.

Subsection ISTB-6200, "Corrective Action," (a), "Alert Range," requires that if the measured test parameter (in this case vibration) values fall within the alert range (greater than 0.325 ips through 0.7 ips) of ASME 0M Code Table ISTB-5121-1 , the frequency of testing specified in ISTB-3400 shall be doubled until the cause of the deviation is determined and the condition is corrected.

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To accept pump vibration at a higher level than the ASME 0M Code-required alert range absolute limits, NUREG/CP-0152 recommends evaluating four key elements: (1 ) vibration history to verify that pumps were operated at this level of vibration for a significant amount of time with justification of "spikes" in test data; (2) consulting with the pump manufacturer/vibration experts to verify that the vibration levels of the pumps are acceptable; (3) attempts to lower the vibration level through modifications to the pumps or the system and structures of the pumps; and (4) perform spectral analysis to identify all contributors to the vibration level. In its submittal, the licensee provided information to address each of these key elements. The licensee also included its evaluation of all of these four key elements for the RHR pumps.

The licensee stated that the pump vendor was contacted during the initial investigation of the cause for failed vibration acceptance criteria. The licensee also stated that the basis for the 0.408 inch/second (in/sec) alert limit comes from the Technical Associates of Charlotte recommendations for vertical pumps. The RHR pump vendor did not recommend a specific value regarding the increased vibration alert limit but stated that the pumps should not be adversely impacted, provided that no upward trend existed in the vibration measurement data.

Also, the data provided in the alternative request in PRR-03, Revision O, shows that the ASME 0M Code alert range value of 0.325 in/sec has been exceeded only on pumps 1 OP-3B and 1 OP-3C. When this same alternative request was submitted as PRR-05 for the fourth 10-year IST interval on February 21 , 2014 (ADAMS Accession No. ML14057A553), the NRC staff asked the licensee to provide justification on why the alternative request is necessary for RHR pumps 1 OP-3A and 10P-3D. In its response dated July 31, 2014 (ADAMS Accession No. ML14213A115), the licensee stated that RHR pumps 10P-3A and 10P-3C are common to Train A, and RHR pumps 10P-3B and 10P-3D are common to Train B. The licensee further stated that the FitzPatrick IST program implementing procedures require increased frequency testing of both pumps in each particular train if the ASME 0M Code alert range value is exceeded on one of the pumps.

Based on its review, the NRC staff found that the licensee has submitted sufficient vibration history to verify that the pumps have operated at this vibration level for a significant period of time with no adverse impacts on performance. Spike data has been justified by consultation with an independent pump expert. The licensee has described attempts to reduce vibration and has demonstrated that the cause of the vibration appears to be the vane pass frequency inherent to the pump design. Spectral analysis of the pump-driver system was performed to identify all contributors to vibration levels. Based on the evaluation of the provided historical pump vibration data, the NRC staff concluded that these are not indicative of degraded pump performance.

The licensee has proposed to raise the minimum vibration alert range for the four RHR pumps listed in Table 2 from 0.325 ips to 0.408 ips. The NRC staff reviewed the historical vibration information for the four RHR pumps and noted that the vibration parameters cited in the alternative request for RHR pumps 10-P3B and 10-P3C do occasionally exceed the ASME 0M Code, Table ISTB-5121-1 minimum alert level of 0.325 ips alert limit. The analysis and evaluation that the licensee performed provide reasonable assurance of operational readiness.

Additionally, the proposed alternative alert limit of 0.408 ips is below the required action limit of A6-14

0.700 ips, and the licensee has demonstrated that these pumps have a normal operational history at this vibration level with no adverse consequences.

Based on the NRC staffs review of the historical vibration data provided, the additional PdM activities proposed, and the identification of vane pass frequency as a primary contributor to vibration, the NRC staff finds that implementation of the proposed alternative is acceptable for the RHR pumps listed in Table 2. Therefore, the NRC staff concludes that compliance with the specified ASME 0M Code requirement would result in hardship without a compensating increase in the level of quality and safety.

4.0 CONCLUSION

As set forth above, the NRC staff determined that for Relief Request PRR-02, Revision 0, the proposed alternative provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1) for Relief Request PRR-02, Revision O. Therefore, the NRC staff authorizes the use of the alternative request for FitzPatrick for the fifth 10-year IST program interval, which is scheduled to begin on June 1, 2018, and end on September 30, 2027.

In addition, as set forth above, the NRC staff determined that for Relief Request PRR-03, Revision 0, the proposed alternative provides reasonable assurance that the affected components are operationally ready. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2) for Relief Request PRR-03, Revision 0. Therefore, the NRC staff authorizes the use of the alternative request for FitzPatrick for the fifth 10-year IST program interval, which is scheduled to begin on June 1, 2018, and end on September 30, 2027.

All other ASME 0M Code requirements for which relief was not specifically requested and approved in the subject requests remain applicable.

Principal Contributor: Robert Wolfgang Date: April 13, 2018.

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ALTERNATIVE REQUEST VRR-03, REVISION 0 FIFTH 10-YEAR INSERVICE TESTING INTERVAL EXELON GENERATION COMPANY, LLC JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 1 .0 INTRODUCTION By letter dated August 3, 2017, as supplemented by letter dated October 26, 2017 (Agencywide Documents Access and Management System Accession Nos. ML17219A123 and ML17299A560, respectively), Exelon Generation Company, LLC (Exelon or the licensee) submitted Relief Request VRR-03, Revision O, to the U.S. Nuclear Regulatory Commission (NRC or the Commission). The licensee requested alternative tests in lieu of certain inservice testing (IST) requirements of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (0M Code) for the IST program at the James A. FitzPatrick Nuclear Power Plant (FitzPatrick) during the fifth 10-year IST program interval, which is scheduled to begin on June 1, 2018, and end on September 30, 2027.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1 ),

the licensee requested to use the proposed alternatives in Relief Request VRR-03, Revision O, on the basis that the alternative provides an acceptable level of quality and safety.

2.0 REGULATORY EVALUATION

Section 50.55a(f), "Preservice and Inservice testing requirements," of 10 CFR requires, in part, that IST of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME 0M Code and applicable addenda, except where alternatives have been authorized pursuant to paragraphs 10 CFR ) or 10 CFR 50.55a(z)(2).

Section 50.55a(z) of 10 CFR states that alternatives to the requirements of paragraphs (b) through (h) of 10 CFR 50.55a, or portions thereof, must be submitted and authorized by the NRC prior to implementation. The applicant or licensee must demonstrate that: (1 ) the proposed alternative would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. Section 50.55a of 10 CFR allows the NRC to authorize alternatives and to grant relief from ASME Code requirements upon making the necessary findings.

Enclosure 3 A6-16

The guidance that the NRC staff considered in its review is Regulatory Guide 1.192, "Operation and Maintenance Code Case Acceptability, ASME 0M Code," June 2003 (ADAMS Accession No. ML030730430), which provides approved for use as voluntary alternatives to the mandatory ASME 0M Code provisions that are incorporated by reference 10 CFR Part 50.

Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request, and the Commission to authorize, the alternative requested by the licensee.

3.0 TECHNICAL EVALUATION

The applicable ASME 0M Code Edition and Addenda for FitzPatrick during the fifth 10-year IST program interval is the 2004 Edition through the 2006 Addenda.

3.1Licensee's Relief Request VRR-03, Revision O Applicable Code Requirements Mandatory Appendix l, paragraph I-1 320, "Test Frequencies, Class 1 Pressure Relief Valves," states, in part, that "Class 1 pressure relief valves shall be tested at least once every 5 years, starting with initial electric power generation."

Components for Which Relief is Requested Class 1, Target Rock pilot-operated main steam safety/relief valves (SRVs):

02RV-071A 02RV-071B 02RV-071C 02RV-071D 02RV-071E 02RV-071 F 02RV-071G 02RV-071H 02RV-071J 02RV-071K 02RV-071L Reason for Alternative Request The current 24-month operating cycle would require the removal and test of 50 percent of the SRVs every refueling outage (i.e., five or six of 1 1), such that all valves are removed and tested every two refueling outages. Approval of extending the test interval for the valves to 6 years with a grace period of 6 months, consistent with ASME 0M Code Case OMN-17, "Alternative Rules for Testing ASME Class 1 Pressure Relief/Safety Valves," would reduce the minimum number of SRVs tested at FitzPatrick over three refueling outages.

Proposed Alternative Testing As an alternative to the ASME 0M Code required 5-year test interval, the licensee proposed that Class 1 pressure relief valves, 02RV-071A, B, C, D, E, F, G, H, J, K, and L, be tested at least once every three refueling cycles with a minimum of 20 percent of the valves tested within any 24-month interval. This 20 percent would consist of valves (complete assemblies) that have not been tested during the current 72-month interval, if they exist. The test interval for any individual valve would not exceed 72 months except that a 6-month grace period is allowed to coincide with refueling outages to accommodate extended shutdown periods. The licensee proposed to continue testing all 11 installed pilot valves every refueling outage.

The relief valve testing and maintenance cycle at Fitzpatrick consists of an as-found inspection, seat leakage, and set pressure testing. After as-found set pressure testing, the valves shall be disassembled and inspected to verify that parts are free of defects resulting from time-related degradation or service-induced wear. As-left set pressure testing shall be performed following maintenance and prior to returning the valve to service.

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3.2 NRC Staff Evaluation The NRC staff reviewed the information in Relief Request VRR-03, Revision 0. The FitzPatrick SRVs are ASME Code Class 1 pressure relief valves that provide overpressure protection for the reactor coolant pressure boundary to prevent unacceptable radioactive release and exposure to plant personnel. Mandatory Appendix I of the ASME 0M Code requires that Class 1 pressure relief valves be tested at least once every 5 years.

However, Mandatory Appendix I does not require that pressure relief valves be disassembled and inspected prior to the start of the 5-year test interval. In lieu of the 5-year test interval, the licensee proposed to implement ASME 0M Code Case OMN-17, which allows a test interval of 6 years plus a 6-month grace period. The ASME Committee on 0M developed Code Case OMN-17 and published it in the 2009 Edition of the ASME 0M Code. ASME 0M Code Case OMN-17 imposes a special maintenance requirement to disassemble and inspect each pressure relief/safety valve to verify that parts are free from defects resulting from time-related degradation or service-induced wear prior to the start of the extended test interval and at each required test during the interval.

Code Case OMN-17 has not yet been added to Regulatory Guide 1.192 or included in 10 CFR 50.55a by reference. However, the NRC has allowed licensees to use ASME OM Code Case OMN-17, provided all requirements in the code case are met. This maintenance will also help to reduce the potential for setpoint drift and increase the reliability of these SRVs to perform their design requirement functions.

Furthermore, ASME 0M Code Case OMN-17 is performance-based in that it requires that the SRVs be tested more frequently if test failures occur. For example, ASME 0M Code Case OMN-17 requires that two additional valves be tested when a valve in the initial test group exceeds the set pressure acceptance criteria. All remaining valves in the group are required to be tested if one of the additional valves tested exceeds its set pressure acceptance criteria.

Based on its review, the NRC staff finds that implementation of ASME 0M Code Case OMN-17 for the testing of the FitzPatrick SRVs, in lieu of the requirements of the 2004 Edition through the 2006 Addenda, Mandatory Appendix I, paragraph 1-1320, of the ASME 0M Code, provides an acceptable level of quality and safety.

4.0 CONCLUSION

As set forth above, the NRC staff determines that for Relief Request VRR-03, Revision 0, the proposed alternative provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1) for Relief Request VRR-03, Revision 0. Therefore, the NRC staff authorizes the use of the alternative request for FitzPatrick for the fifth 10-year IST program interval, which is scheduled to begin on June 1, 2018, and end on September 30, 2027.

All other ASME 0M Code requirements for which relief was not specifically requested and approved in the subject request for relief remain applicable.

Principal Contributor: John Billerbeck Date: April 13, 2018.

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUESTS VRR-01 REVISION VRR-02 REVISION 0; AND VRR-04, REVISION 0 FIFTH 10-YEAR INTERVAL INSERVICE TESTING PROGRAM EXELON GENERATION COMPANY, LLC JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 1 .0 INTRODUCTION By letter dated August 3, 2017, as supplemented by letter dated October 26, 2017 (Agencywide Documents Access and Management System Accession Nos. ML17219A123 and ML17299A560, respectively), Exelon Generation Company, LLC (Exelon or the licensee) submitted Relief Requests VRR-01, Revision O; VRR-02, Revision O; and VRR-04, Revision O, to the U.S. Nuclear Regulatory Commission (NRC or the Commission). The licensee requested alternative tests in lieu of certain inservice testing (IST) requirements of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (0M Code) for the IST program at the James A. FitzPatrick Nuclear Power Plant (FitzPatrick) during the fifth 10-year IST program interval, which is scheduled to begin on June 1, 2018, and end on September 30, 2027.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee requested to use proposed alternatives in Relief Request VRR-02, Revision 0, and Relief Request VRR-04, Revision 0, on the basis that the alternatives provide an acceptable level of quality and safety. Pursuant to 10 CFR 50.55a(z)(2), the licensee requested to use the proposed alternative in Relief Request VRR-01 , Revision 0, on the basis that the ASME 0M Code requirements present an undue hardship, without a compensating increase in the level of quality or safety.

2.0 REGULATORY EVALUATION

Section 50.55a(f), "Preservice and Inservice testing requirements," of 10 CFR requires, in part, that IST of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME 0M Code and applicable addenda, except where alternatives have been authorized pursuant to paragraphs 10 CFR 50.55a(z)(1) or 10 CFR 50.55a(z)(2).

Section 50.55a(z) of 10 CFR states that alternatives to the requirements of paragraphs (b) through (h) of 10 CFR 50.55a, or portions thereof, must be submitted and authorized by the NRC prior to implementation. The applicant or licensee must demonstrate that: (1) the proposed alternative would provide an acceptable level of quality and safety, or (2) compliance Enclosure 4 A6-19

with the specified requirements would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety. Section 50.55a of 10 CFR allows the NRC to authorize alternatives and to grant relief from ASME Code requirements upon making the necessary findings.

The guidance that the NRC staff considered in its review include the following:

NUREG-0933, "Resolution of Generic Safety Issues," Issue 105, "Interfacing Systems LOCA [Loss-of-Coolant Accident] at LWRs [Light-Water Reactors]," December 201 1 (ADAMS Accession No. ML11353A382), which provides a single-source repository of all NRC generic safety issue reviews.

NUREG-1482, Revision 2, "Guidelines for Inservice Testing at Nuclear Power Plants,"

October 2013 (ADAMS Accession No. ML13295A020), which provides acceptable guidance to licensees to establish a basic understanding of the regulatory basis for pump and valve IST programs.

NUREG/CR-5928, "Final Report of the ISLOCA Research Program," August 2007 (ADAMS Accession No. Ml-072430731 ), which quantifies the risk associated with an interfacing system loss-of-coolant accident (ISLOCA) event.

Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request, and the Commission to authorize, the alternatives requested by the licensee.

3.0 TECHNICAL EVALUATION

The applicable ASME 0M Code Edition and Addenda for FitzPatrick during the fifth 10-year IST program interval is the 2004 Edition through the 2006 Addenda.

3.1 Licensee's Relief Request VRR-01, Revision O Applicable Code Requirements Subsection ISTC-5151 , "Valve Stroke Testing," (a), states that "Active valves shall have their stroke times measured when exercised in accordance with ISTC-3500."

Subsection ISTC-5151 , "Valve Stroke Testing," (c), states that "Stroke time shall be measured to at least the nearest second."

Components for Which Relief is Requested Table 1 Valve ID Function Category Class 07SOV-104A Traversing In-Core Probe (TIP) Containment Isolation A 2 Valve (CIV) 07SOV-104B TIP CIV A 2 07sov-104C TIP CIV A 2 A6-20

Reason for Alternative Request The Category A containment isolation solenoid operated valves identified in this request have no safety function in the open direction as they open to allow the passage of the TIP assembly and drive cable for flux mapping operations. These valves have an active safety function in the closed direction in response to a primary containment isolation system signal to seal the TIP guide tubes. Therefore, an exercise test and subsequent stroke time test are only required in the closed direction.

The design of the TIP control system does not allow for measurement of the closure stroke times of valves 07SOV-104A, B, and C. Measuring the closure stroke times in accordance with the ASME 0M Code would require a costly computer control system modification. Closure of valves 07SOV-104A, B, and C could also be accomplished by an alternative method, but this method would require manual extraction and retraction of the TIP from the shield block. This method of testing would be contrary to the principles of keeping radiation exposure as low as reasonably achievable because it would result in radiation exposure to personnel performing the test.

The proposed alternative test ensures the operation of valves 07SOV-104A, B, and C in both directions and provides an acceptable level of quality and safety. This method meets the desired outcome of monitoring valve stroke time for degradation since the computer controls the 10-second delay and the additional approximate 2 seconds for valve closure should indicate the actual stroke time.

Proposed Alternative Testing As an alternative to the ASME 0M Code required 5-year test interval, the licensee proposed to measure overall cycle time (opened and closed) for the TIP CIVs 07SOV104A, B, and C, in accordance with Subsection ISTC-5152. Exelon will time the opening (10-second delay time included) and closing cycle for valves 07SOV-104A, B, and C. The time from open initiation to receipt of the closed light for each valve will be monitored with a stop watch.

NRC Staff Evaluation

The NRC staff reviewed the information in Relief Request VRR-01 , Revision 0.

Subsection ISTC-5151 of ASME 0M Code details the requirements for valve stroke testing of solenoid operated valves. In lieu of these requirements, the licensee proposed to full stroke exercise each valve noted in Table 1 and determine proper operation by using the computer control system for the TIP solenoid valves, which includes a provision for measuring valve cycle time (opened and closed). The computer control system opens the subject valve, maintains it energized for 10 seconds, and deenergizes the valve allowing the solenoid valve to stroke closed (< 2 seconds). The total computer control system test is 12 seconds. Exelon will apply Subsection ISTC-5152(a), which requires that each valve exhibit no more than +/- 25 percent change in stroke time when compared to the reference value, except that the full-stroke limiting time for each valve will be truncated at 12 seconds.

Based on its review, the NRC staff finds that the proposed alternative is consistent with the guidance in NUREG-1482, Revision 2, paragraph 4.2.3. Therefore, the NRC staff concludes that the proposed alternative provides an acceptable level of quality and safety.

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3.2. Licensee's Relief Request VRR-02, Revision 0 Applicable Code Requirements Subsection ISTC-3510, "Exercising Test Frequency," states, in part, that "Active Category A, Category B, and Category C check valves shall be exercised nominally every 3 months, except as provided by ISTC-3520, ISTC-3540, ISTC-3550, ISTC-3570, ISTC-5221 , and ISTC-5222."

Subsection ISTC-3522, "Category C Check Valves," (c), states that "If exercising is not practicable during operation at power and cold shutdowns, it shall be performed during refueling outages."

Subsection ISTC-3700, "Position Verification Testing," states, in part, that "Valves with remote position indicators shall be observed locally at least once every 2 years to verify that valve operation is accurately indicated. "

Valve ID System Category Class 23EFV-01B HPCI EFCV A/C 1 23EFV-02A HPCI EFCV A/C 1 23EFV-02B HPCI EFCV A/C 1 29EFV-30A Main Steam (MS) EFCV A/C 1 29EFV-30B MS EFCV A/C 1 29EFV-30C MS EFCV A/C 1 29EFV-30D MS EFCV A/C 1 29EFV-34A MS EFCV A/C 1 29EFV-34B MS EFCV A/C .1 29EFV-34C MS EFCV A/C 1 29EFV-34D MS EFCV A/C 1 29EFV-53A MS EFCV A/C 1 29EFV-53B MS EFCV A/C 1 29EFV-53C MS EFCV A/C 1 29EFV-53D MS EFCV A/C 1 29EFV-54A MS EFCV A/C 1 29EFV-54B MS EFCV A/C 1 29EFV-54C MS EFCV A/C 1 29EFV-54D MS EFCV A/C 1 Reason for Alternative Request The ASME 0M Code requires check valves to be exercised quarterly during plant operation, or if valve exercising is not practicable during plant operation and cold shutdown, it shall be performed during refueling outages. Based on past experience, EFCV testing during in-service leakage testing can become the outage critical path and could possibly extend the outage if all EFCVs were to be tested during this time frame. The testing requires isolation of the instruments associated with each EFCV and opening of a drain valve to actuate the EFCV.

Process fluid will be contaminated to some degree, requiring special measures to collect flow from the drain valve and also contributes to an increase in personnel radiation exposure.

Proposed Alternative Testinq The licensee proposed to 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 at least once every 10 years (nominal). Industry experience as documented in General Electric (GE) Topical Report NEDO-32977-A/B21-00658-01 , "Excess Flow Check Valve Testing Relaxation, June 2000 (ADAMS Accession No. ML003729011 ), indicates that EFCVs have a very low failure rate.

The instrument lines at Fitzpatrick have a flow restricting orifice upstream of the EFCVs to limit reactor water leakage in the event of rupture. The Fitzpatrick Final Safety Analysis Report (FSAR), paragraph 7.1.6, "Supplemental NSSS Supplier Information," does not credit the EFCVs, but instead credits the installed orifice for limiting the release of reactor coolant A6-24

following an instrument line break. Thus, a failure of an EFCV, though not expected as a result of this request, is bounded by the FSAR analysis. The licensee's test experience is consistent with the findings in NEDO-32977-A. The NEDO-32977-A topical report indicates similarly that many reported test failures at other plants were related to test methodologies and not actual EFCV failures.

The EFCV failures will be documented in the Fitzpatrick's Corrective Action Program as an equipment and surveillance test failure. The failure will be evaluated and corrected to ensure EFCV performance remains consistent with the extended test interval, a minimum acceptance criteria of less than or equal to one failure per year on a 3-year rolling average will be required.

NRC Staff Evaluation

The EFCVs are installed on instrument lines to limit the release of fluid in the event of an instrument line break. The EFCVs are not required to close in response to a containment isolation signal and are not required to operate under post-LOCA conditions. The EFCVs are required to be tested in accordance ASME 0M Code ISTC-3510. The proposed change revises the surveillance frequency by allowing a "representative sample" of EFCVs to be tested every refueling outage.

The NRC staff reviewed NEDO-32977-A and issued its safety evaluation on March 14, 2000 (ADAMS Accession No. Ml-003691722). In its safety evaluation, the NRC staff found that the test interval could be extended up to a maximum of 10 years. In conjunction with this finding, the NRC staff noted that each licensee that adopts the relaxed test interval program for EFCVs must have a failure feedback mechanism and corrective action program to ensure EFCV performance continues to be bounded by the topical report results. Also, each licensee is required to perform a plant-specific radiological dose assessment, EFCV failure analysis, and release frequency analysis to confirm that they are bounded by the generic analyses of the topical report.

The NRC staff reviewed the licensee's current proposal and a previous NRC-approved request for the fourth IST interval dated November 27, 2007 (ADAMS Accession No. Ml-072910422),

for its applicability to GE Nuclear's topical report NEDO-32977-A, as well as conformance with the NRC staff's guidance regarding radiological dose assessment, EFCV failure rate, release frequency, and the proposed failure feedback mechanism and corrective action program.

Based on its review, the NRC staff concludes that the radiological consequences of an EFCV failure are sufficiently low and acceptable, and that the alternative testing in conjunction with the corrective action plan provides a high degree of valve reliability and operability. Therefore, the NRC staff finds that the licensee's proposed test alternative provides an acceptable level of quality and safety.

3.3. Licensee's Relief Request VRR-04, Revision O Applicable Code Requirements Subsection ISTC-3630, "Leakage Rate for Other Than Containment Isolation Valves," states, in part, that "Category A valves with a leakage requirement not based on an Owner's 10 CFR 50, Appendix J program, shall be tested to verify their seat leakages are within acceptable limits.

Subsection ISTC-3630(a), "Frequency," states that "Tests shall be conducted at least once every 2 years."

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Components for Which Relief is Requested Table 3 Valve ID Function Category Class 10AOV-68A Residual Heat Removal (RHR) A Low Pressure A/C 1 Coolant Injection (LPCI) Testable Check Valve 10AOV-68B RHR A LPCI Testable Check Valve A/C 1 10MOV-25A RHR A LPCI Inboard Injection Valve A 1 10MOV-25B RHR B LPCI Inboard Injection Valve A 1 14AOV-13A CSP A Reactor Isolation Testable Check Valve A/C 1 14AOV-13B CSP B Reactor Isolation Testable Check Valve A/C 1 14MOV-12A Core Spray Loop A Inboard Isolation Valve A 1 14MOV-12B Core Spray Loop B Inboard Isolation Valve A 1 10MOV-17 RHR Shutdown Cooling Outboard Isolation Valve A 1 10MOV-18 RHR Shutdown Cooling Inboard Isolation Valve A 1 Reason for Alternative Request At FitzPatrick, pressure isolation valves (PIVs) are Category A or Category A/C valves within the scope of Subsection ISTC-3630 of the ASME 0M Code. This alternative to allow for scheduling of leak tests for the valves identified in Table 3 to a performance-based frequency that is the same as 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," Option B, "Performance-Based Requirements,"

testing, would provide an acceptable level of quality and safety.

Proposed Alternative Testing The licensee proposed an alternative test frequency in lieu of the requirements found in Subsection ISTC-3630(a) for the ten applicable PIVs listed in Table 3. Appendix J to 10 CFR Part 50 was amended to improve the focus of the body of regulations by eliminating prescriptive requirements that are marginal to safety and to provide licensees greater flexibility for cost-effective implementation methods for meeting regulatory safety objectives.

Nuclear Energy Institute (NEI) 94-01, Revision 3A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," July 2012 (ADAMS Accession No. ML12221A202), describes the risk-informed basis for extended test intervals under Option B.

That justification documents valves that have demonstrated good leakage rate performance over two consecutive cycles are subject to future failures predominantly governed by the random failure rate of the component. The NEI 94-01, Revision 3A guidance also presents the results of a comprehensive risk analysis, including the statement that "The risk impact associated with increasing test intervals is negligible (less than 0.1 percent of total risk)."

The guidance in NUREG-0933 discussed the need for PIV leak-rate testing based primarily on three pre-1980 historical failures of applicable valves industrywide. These failures all involved human errors in either operations or maintenance. None of these failures involved inservice equipment degradation. The performance of PIV leak-rate testing provides assurance of acceptable seat leakage with the valve in a closed condition. Typical PIV leak-rate testing does not identify functional problems that may inhibit the valve's ability to reposition from open to close.

A6-26

Fitzpatrick proposed to perform PIV testing at intervals specified in NEI 94-01. Program guidance will be established such that if any of the valves fail either the CIV test or PIV test, the test interval for both tests will be reduced to once every 30 months until they can be reclassified as good performers per the performance evaluation requirements of Appendix J, Option B. The test intervals for the valves identified in this request will be determined in the same manner as is done for CIV testing under Option B. The test interval may be extended upon completion of two consecutive periodic PIV tests with results within the prescribed acceptance criteria. Any PIV test failure will require a return to the initial interval until good performance can again be established.

The risks associated with extending the leakage test interval to a maximum of 75 months are extremely low. The basis for this alternative request is the historically good performance of the PIVs. This alternative will also provide significant reductions in radiation dose.

NRC Staff Evaluation

The NRC staff reviewed the information in Relief Request VRR-04, Revision O. The licensee proposed to functionally test and verify the leakage rate of these PIVs using the 10 CFR Part 50, Appendix J, Option B performance-based schedule. Valves would initially be tested at the required interval schedule, which is currently every refueling outage, or 2 years, as specified by the ASME 0M Code, Section ISTC-3630(a). Valves that have demonstrated good performance for two consecutive cycles may have their test interval extended to 75 months. Any PIV leakage test failure would require the component to return to the initial interval of every refueling outage, or 2 years, until good performance can again be established.

Pressure isolation valves are defined as two valves in series within the reactor coolant pressure boundary, which separate the high pressure reactor coolant system from an attached lower pressure system. Failure of a PIV could result in an over-pressurization event that could lead to a system rupture and possible release of fission products to the environment. This type of failure event was analyzed under the NUREG/CR-5928 ISLOCA research program. The NUREG/CR-5928 research program analyzed boiling-water reactor (BWR) and pressurized-water reactor designs. The conclusion of the analysis resulted in ISLOCA not being a risk concern for BWR design. FitzPatrick is a BWR design.

Appendix J, Option B to 10 CFR Part 50, is a performance-based leakage test program.

Guidance for implementation of acceptable leakage rate test methods, procedures, and analyses is provided in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," September 1995 (ADAMS Accession No. Ml-003740058). Regulatory Guide 1.163 endorses NEI 94-01 , Revision O, "Industry Guideline for Implementing Performance Based Option of 10 CFR 50, Appendix J," dated July 21 , 1995 (ADAMS Accession No. ML11327A025), with the limitation that Type C components test interval cannot extend greater than 60 months. The current version of NEI 94-01 is Revision 3-A, which allows Type C containment isolation valves test intervals to be extended to 75 months, with a permissible extension for nonroutine emergent conditions of 9 months (84 months total). The NRC staff finds the guidance in NEI 94-01, Revision 3-A, acceptable (safety evaluation dated June 8, 2012, and approval letter dated December 2, 2016; available at ADAMS Accession Nos. ML121030286 and ML12226A546, respectively), with the following conditions:

1. Extended interval for Type C local leakage-rate tests (LLRTs) may be increased to 75 months, with the requirement that a licensee's post-outage report include the margin between Type B and Type C leakage rate summation and its regulatory limit.

In addition, a corrective action plan shall be developed to restore the margin to an acceptable level. Extensions of up to 9 months (total maximum interval of 84 months for Type C tests) are permissible only for nonroutine emergent conditions. This A6-27

provision (9-month extension) does not apply to valves that are restricted and/or limited to 30-month intervals in NEI 94-01 , Revision 3A, Section 10.2, "Type B and Type C Testing Frequencies" (such as BWR main steam isolation valves), or to valves held to the base interval (30 months) due to unsatisfactory LLRT performance.

2. When routinely scheduling any LLRT valve interval beyond 60 months and up to 75 months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type B and Type C total and must be included in a licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

Based on its review, the NRC staff finds that the proposed alternative was previously authorized for use at FitzPatrick for the fourth IST program interval under safety evaluation dated March 16, 2012 (ADAMS Accession No. ML12072A113). As noted in the licensee's proposed alternative, the valves have maintained a history of good performance. Extending the leakage test interval based on good performance and the low risk factor as noted in NUREG/CR-5928 is a logical progression to a performance-based program. Therefore, the NRC staff concludes that the licensee's proposed alternative provides an acceptable level of quality and safety.

4.0 CONCLUSION

As set forth above, the NRC staff determines that for Relief Request VRR-02, Revision O, and Relief Request VRR-04, Revision O, the proposed alternative provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1) for Relief Request VRR-02, Revision O, and Relief Request VRR-04, Revision O. Therefore, the NRC staff authorizes the use of the alternative request for FitzPatrick for the fifth 10-year IST program interval, which is scheduled to begin on June 1, 2018, and end on September 30, 2027.

As set forth above, the NRC staff determines that for Relief Request VRR-01 , Revision O, the proposed alternative provides reasonable assurance that the affected components are operationally ready. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2) for Relief Request VRR-01, Revision O. Therefore, the NRC staff authorizes the use of the alternative request for FitzPatrick for the fifth 10-year IST program interval, which is scheduled to begin on June 1, 2018, and end on September 30, 2027.

All other ASME 0M Code requirements for which relief was not specifically requested and approved in the subject requests remain applicable.

Principal Contributor: Michael Farnan Date: April 13, 2018 A6-28

ML18044A993 *b safet evaluation OFFICE NRR/DORL/LPLI/PM NRR/DORL/LPLI/PM NRR/DORL/LPLI/LA NAME BVenkataraman THood LRonewicz (JBurkhardt for)

DATE 02/13/2018 04/4/2018 04/04/2018 OFFICE NRR/DE/EMIB/BC* NRR/DORL/LPLI/BC NAME SBailey JDanna DATE 01/09/2018 04/13/2018 ATTACHMENT 7 CODE CASE INDEX

CODE CASE TITLE NUMBER None A7 - 1

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ATTACHMENT 8 COLD SHUTDOWN JUSTIFICATION INDEX

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant CSJ NUMBER REV # TITLE CSJ-01 Reactor Water Recirculation CSJ-02 Reactor Building Closed Loop Circulation CSJ-03 High Pressure Coolant Injection CSJ-04 Deleted CSJ-05 Main Steam CSJ-06 Main Steam CSJ-07 Reactor Water Cleanup CSJ-08 Main Steam CSJ-09 Reactor Core Isolation CSJ-10 Residual Heat Removal CSJ-11 Core Spray A8 - 1

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATIONS

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Cold Shutdown Justifications CSJ-01 SYSTEM: REACTOR WATER RECIRCULATION (RWR)

COMPONENTS: 02-2MOV-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: During normal plant operation, exercising these valves will trip the respective recirculation pump when the valve is 10% open. Securing either pump (single loop operation) is limited by Technical Specification requirements and 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).

A9-1

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant CSJ-02 SYSTEM: REACTOR BUILDING CLOSED LOOP COOLING (RBCLC)

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). 15AOV-132A/B and 15AOV-133A/B will be stroke time tested during cold shutdowns when Reactor Water Recirculation Pumps can be secured.

A9-2

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 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 because feedwater pump discharge pressure is present - 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).

A9-3

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant CSJ-04 DELETED A9-4

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant CSJ-05 SYSTEM: MAIN STEAM (MSS)

COMPONENTS: 29AOV-86A, B, C, D CATEGORY: A SAFETY FUNCTION: The Main Steam Isolation Valves (MSIVs) are normally open valves that close to isolate containment from the main steam system.

JUSTIFICATION: Exercising these valves during normal operation isolates one line of steam flow to the turbine. Isolation of a main steam header would cause a severe pressure transient in the associated main steam line possibly resulting in a plant trip. Additionally, closure of an MSIV, at power, could potentially result in challenging the setpoint of the main safety relief valves causing inadvertent lifting. Industry experience also indicates that closing the MSIVs under high steam flow conditions may be a contributing factor in observed seat degradation. Seat degradation occurring during valve exercising could result in a loss of primary containment integrity. Therefore, it is impractical to full-stroke exercise these valves to the closed position on a quarterly (nominal 92 days) frequency during plant operation.

The MSIVs have the capability and are being partial stroked during the Technical Specification MSIV scram sensor channel functional test requirements. To completely partially fail-safe exercise these valves to the closed position, the airlines to the valves must be isolated. Thus, with the loss of air, the fail-safe mechanism (springs) would be demonstrated. The resultant exercising of the Main Steam Isolation Valves (MSIV's) could place the plant in an unsafe mode of operation causing transient conditions which could result in a reactor scram.

Therefore, partial stroke exercise testing increases the risk of a valve closure when the unit is generating power. This concern was realized within the fleet and the industry and has resulted in full closure of the applicable MSIV and a reactor trip on high pressure.

NUREG-1482 "Guidelines for Inservice Testing at Nuclear Power Plants", Section 2.4.5, "Deferring Valve Testing to Cold Shutdown or Refueling Outages" identifies impractical conditions justifying test deferrals as those conditions that could result in unnecessary challenges to safety systems, place undue stress on components, cause unnecessary cycling of equipment, or unnecessarily reduce the life expectancy of the plant systems and components. As such, it is impractical to partially exercise MSIVs on a quarterly (nominal 92 days) frequency during plant operation.

ALTERNATE TEST: These valves will be full-stroke exercise tested to the closed position and fail-safe tested during cold shutdowns per ISTC-3521(c) and (f).

A9-5

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Additional Information to Support Alternative Test On November 1, 2017, Exelon Generation Company, LLC (Exelon) submitted a relief request associated with the Inservice Testing (IST) programs for Clinton Power Station, Unit 1; Dresden Power Station, Units 2 and 3; James A. FitzPatrick Nuclear Power Plant; LaSalle County Station, Units 1 and 2; Nine Mile Point Nuclear Station, Units 1 and 2; Oyster Creek Nuclear Generating Station; and Quad Cities Nuclear Power Station, Units 1 and 2.

The relief request submitted on November 1, 2017 proposed an authorization to continue to partial stroke exercise MSIVs on a limited basis with a Cold Shutdown Justification currently in place for MSIVs. Exelon proposed that the partial stroke exercise of MSIVs would be performed in accordance with the Surveillance Frequency Control Program (SFCP) and would partially stroke exercise MSIVs at variant test intervals until the final refueling outage testing interval was achieved. Exelons relief request was submitted due to the belief that ISTC-3521(b) and ISTC-3521(c) prohibited any type of exercising of MSIVs with a cold shutdown in place.

On February 26, 2018, the NRC held a public meeting to discuss the relief request. The NRC staff stated that the alternative is not needed, since the ASME OM Code does not prevent the continued exercising of MSIVs at power if a CSJ documented in the IST Program Plan document for each site demonstrates that exercising at power is not practicable. In particular, the NRC staff noted that paragraph ISTC-3521 (c) of the ASME OM Code states that exercising "may"- not "shall" - be limited to full-stroke exercising during cold shutdown.

Thus, paragraph ISTC-3521 (c) does not prevent the partial-stroke testing of the MSIVs at power. Exelon will utilize paragraph ISTC-3521 (c) for full stroke exercising of the MSIVs during cold shutdown under the IST program while continuing to partially stroke exercise the MSIVs during power operation under the SFCP at various intervals commensurate with the SFCP frequencies.

The following documents are attached to support this CSJ

NRC Summary of 022618 mtg ML18058A523

2018_0327 Fleet IST RR for MSIV Stroke Frequency Withdrawal letter

NRC withdrawal documentation ML18100A015 A9-6

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 16, 2018 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO)

Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

CLINTON POWER STATION, UNIT NO. 1 ; DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3; JAMES A. FITZPATRICK NUCLEAR POWER PLANT; LASALLE COUNTY STATION, UNITS 1 AND 2; NINE MILE POINT NUCLEAR STATION, UNITS 1 AND 2; OYSTER CREEK NUCLEAR GENERATING STATION; AND QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2WlTHDRAWAL OF PROPOSED ALTERNATIVE TO THE MAIN STEAM ISOLATION VALVE TESTING REQUIREMENTS (EPID L-2017-LLR-OI 34)

Dear Mr. Hanson:

By application dated November 1 , 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17306A014), Exelon Generation Company, LLC (the licensee) submitted a request in accordance with Paragraph 50.55a(z)(1) of Title 10 of the Code of Federal Regulations (10 CFR) for a proposed alternative to the requirements of 10 CFR 50.55a, "Codes and standards," and the American Society of Mechanical Engineers (ASME)

Code for Operation and Maintenance of Nuclear Power Plants (0M Code) at Clinton Power Station, Unit No. 1; Dresden Nuclear Power Station, Units 2 and 3; James A. FitzPatrick Nuclear Power Plant; LaSalle County Station, Units 1 and 2; Nine Mile Point Nuclear Station, Units 1 and 2; Oyster Creek Nuclear Generating Station; and Quad Cities Nuclear Power Station, Units 1 and 2. The proposed alternative was intended to allow the licensee to eliminate the quarterly partial-stroke exercise testing of the main steam isolation valves (MSIVs) for each of these facilities. By letter dated March 27, 2018 A9-7

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant (ADAMS Accession No. ML18086B221), the licensee requested to withdraw its November 1, 2017, application because the licensee has concluded that the proposed alternative is not needed.

Currently, the licensee performs partial-stroke testing of the MSlVs quarterly and full-stroke testing of the MSlVs during cold shutdown based on a cold shutdown justification (CSJ). Paragraph ISTC-3521 (b) of the ASME 0M Code states that if full-stroke exercising of a valve during operation at power is not practicable, it may be limited to partial-stroke during operation at power and full-stroke during cold shutdown. The application stated that the licensee will revise its CSJ to eliminate the quarterly partial-stroke testing of MSlVs at power, such that the only exercising of MSlVs will be the full-stroke testing during cold shutdown. Paragraph ISTC-3521 (c) of the ASME 0M Code states that if exercising of a valve during operation at power is not practicable, it "may" be limited to full-stroke exercising during cold shutdown.

B. Hanson The licensee also performs a quarterly reactor protection system (RPS) channel functional test (a surveillance requirement) that uses the partial-stroke of the MSIV to generate the MSIV position switch input into the RPS logic. The application stated that the licensee will use the surveillance frequency control program to extend the RPS channel functional test to refueling outage intervals. In order to accomplish this change, the application stated that the partial stroking of the MSlVs at power would need to continue for a number of years at longer intervals (i.e., less frequent than the current quarterly testing).

The application indicated that the licensee's proposed alternative to the ASME 0M Code was needed because continuing to perform periodic partial-stroke MSIV testing at power will contradict a CSJ that exercising at power is not practicable. However, paragraph ISTC-3521 (c) of the ASME 0M Code states that exercising "may"not "shall"be limited to full-stroke exercising during cold shutdown. Thus, paragraph ISTC-3521(c) does not prohibit the partial stroke testing of the MSlVs at power. This fact was discussed with the licensee during a public teleconference held on February 26, 2018 (ADAMS Accession No. ML18058A523). Based on this, the licensee has concluded that the proposed alternative is not needed and has requested to withdraw its November 1, 2017, application.

This letter acknowledges that the licensee has withdrawn its November 1, 2017, application. If you have any questions, please contact Blake Purnell at 301-415-1380 or via e-mail at Blake.Purnell@nrc.qov.

Sincerely, Blake Purnell, Project Manager Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-461 , 50-237, 50-249, 50-333 50-373, 50-374, 50-220, 50-410, 50-219, 50-254, and 50-265 A9-8

ML18100A015 *via email OFFICE LPL3/PM LPL3/LA DE/EMIB/BC* LPL3/BC LPL3/PM NAME BPurnell SRohrer SBaiIey DWrona BPurnell DATE 4/1 1/18 4/11/18 4/9/18 4/16/18 4/16/18 SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant CSJ-06 SYSTEM: MAIN STEAM (MSS)

COMPONENTS: 29MOV-203A, B CATEGORY: B SAFETY FUNCTION: These valves open to provide flow paths 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).

A9-10

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 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).

A9-11

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 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.

A9-12

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 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).

A9-13

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant CSJ-10 SYSTEM: RESIDUAL HEAT REMOVAL (RHR)

COMPONENTS: 10MOV-25A and B CATEGORY: A SAFETY FUNCTION: These valves open to provide a flow path for the RHR system injection to the reactor vessel. These valves provide the pressure isolation function to protect the low-pressure interconnecting RHR piping from the high pressure Reactor Coolant system.

JUSTIFICATION: These pressure isolation motor-operated valves maintain one of the two high to low pressure barriers during plant operation. The other pressure isolation barrier is a check valve. Opening a PIV with only one other PIV available, especially with it being a check valve, increases the probability of over-pressurizing the low-pressure core spray system and inter-system LOCA. These valves are not designed to open with high differential pressure across the seats. To exercise these valves during plant operation requires installation of hydrostatic test equipment and instrumentation to equalize pressure across the valve seats when the reactor is pressurized.

ALTERNATE TEST: Exercise test during cold shutdown.

A9 - 14

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant CSJ-11 SYSTEM: CORE SPRAY (CS)

COMPONENTS: 14MOV-12A and B CATEGORY: A SAFETY FUNCTION: These valves open to provide a flow path for the CS system injection to the reactor vessel. These valves provide the pressure isolation function to protect the low-pressure interconnecting CS piping from the high pressure Reactor Coolant system.

JUSTIFICATION: These pressure isolation motor-operated valves maintain one of the two high to low pressure barriers during plant operation. The other pressure isolation barrier is a check valve. Opening a PIV with only one other PIV available, especially with it being a check valve, increases the probability of over-pressurizing the low-pressure core spray system and inter-system LOCA.

The valves are not designed to open with high differential pressure across the seats and instrumentation is not installed to determine the differential pressure prior to stroking the valve from open to close. Installation of test equipment would be required to conduct the testing quarterly.

ALTERNATE TEST: Exercise test during cold shutdown.

A9 - 15

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ATTACHMENT 10 REFUELING OUTAGE JUSTIFICATION INDEX

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATIONS

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-01 DELETED. JAF has an approved relief request. A refueling outage justification is not necessary for excess flow check valve testing.

A11-1

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 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.

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

A11-2

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 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 time Reactor Water Recirculation Pumps are running subjects the pump seals to thermal transients and pressure fluctuations, thereby, shortening seal life. Pressure fluctuations and oscillations can degrade the pressure-retaining ability of either or both seal stages. Additionally, securing seal purge flow while the Reactor Water Recirculation Pumps are running introduces reactor coolant and associated corrosion products into the seal cavity, which also shortens seal life.

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

A11-3

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-04 SYSTEM: AUTOMATIC DEPRESSURIZATION (ADS)

COMPONENTS: 02RV-1 through 02RV-11 02VB-1 through 02VB-11 CATEGORY: C SAFETY FUNCTION: These valves remain closed to prevent steam from an open safety/relief valve (SRV) from entering the drywell. They open following closure of an SRV to prevent the formation of a water column within the downcomer that could cause torus damage during subsequent lifting of the same SRV.

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

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

A11-4

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-05 SYSTEM: RESIDUAL HEAT REMOVAL (RHR)

COMPONENTS: 10RHR-64A, B, C, D CATEGORY: C SAFETY FUNCTION: These valves open on forward flow to provide minimum flow protection for the RHR pumps and close on reverse flow to prevent diversion of flow through an idle parallel pump.

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 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.

A11-5

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ALTERNATE TEST: In accordance with ISTC-5221(c), a sample disassembly examination program will be implemented for this Group of valves. During each refuel outage at least one (1) valve will be disassembled, inspected, and verified operable. The acceptance criteria as stated in ISTC-5221(c)(2) is provided in the maintenance procedure used for check valve disassembly. 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. These valves will be full or part-stroke exercised, if practicable, after reassembly.

A11-6

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-06 SYSTEM: RESIDUAL HEAT REMOVAL (RHR)

COMPONENTS: 10RHR-95A,B CATEGORY: C SAFETY FUNCTION: These valves close to prevent reverse flow from the torus.

JUSTIFICATION: These are simple check valves with no means of determining disc position without performing a back leakage test. Performing such a test during plant operations would require setting up a test rig and performing a hydrostatic test. As discussed in NUREG 1482, Revision 2, 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).

A11-7

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-07 SYSTEM: STANDBY LIQUID CONTROL (SLC)

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

JUSTIFICATION: Full-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.

A11-8

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 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).

A11-9

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-09 SYSTEM: CORE SPRAY (CSP)

COMPONENTS: 14AOV-13A,B CATEGORY: A/C SAFETY FUNCTION: These valves open to provide flow paths 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 refueling outages, the valves will be exercised using a mechanical exerciser. In addition, the valves require pressure isolation valve leak rate tests once every two years.

Each of the valves will be full-stroked exercised during each refuel outage per ISTC-5221(b) and ISTC-3521(c) and (f) and leak rate tested once every two years per ISTC-3630(a) to satisfy full open and close exercising tests.

A11-10

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 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, Revision 2, section 4.1.6, 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.

A11-11

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-11 DELETED A11-12

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 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).

A11-13

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-13 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)

COMPONENTS: 23HPI-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 a sample disassembly program such that one valve is disassembled and inspected each refueling outage to verify operability as an alternative to quarterly testing. The grouping requirements of ISTC-5221(c) shall be followed. These valves will be full or part-stroke exercised, if practicable, after reassembly.

A11-14

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 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 sample disassembly program each refueling outage to verify operability as an alternative to quarterly testing. This valve will be full or part-stroke exercised, if practicable, after reassembly.

A11-15

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-15 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)

COMPONENTS: 23HPI-61 CATEGORY: C SAFETY FUNCTION: This valve opens to provide a flow path 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 a sample disassembly program each refueling outage to verify operability as an alternative to quarterly testing. This valve will be full or part-stroke exercised, if practicable, after reassembly.

A11-16

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-16 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)

COMPONENTS: 23HPI-62 CATEGORY: C SAFETY FUNCTION: This valve opens to provide a flow path 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 a sample disassembly program each refueling outage to verify operability as an alternative to quarterly testing. This valve will be full or part-stroke exercised, if practicable, after reassembly.

A11-17

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-17 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)

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

JUSTIFICATION: This valve has no means of determining disc position or flow rate 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 a sample disassembly program each refueling outage to verify operability as an alternative to quarterly testing. This valve will be full or part-stroke exercised, if practicable, after reassembly.

A11-18

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-18 SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)

COMPONENTS: 23HPI-131 CATEGORY: C SAFETY FUNCTION: This valve closes to prevent flow diversion from the HPCI booster pump.

JUSTIFICATION: There is no provision on this valve that provides position indication of the disc. There are no test taps and block valves to enable a back-leakage test to verify closure.

ALTERNATE TEST: ISTC-5221(c) allows a sample disassembly program each refueling outage to verify operability as an alternative to quarterly testing. This valve will be full or part-stroke exercised, if practicable, after reassembly.

A11-19

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 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).

A11-20

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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.

A11-21

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-21 SYSTEM: INSTRUMENT AIR (IAS)

COMPONENTS: 39IAS-22 & 39IAS-29 CATEGORY: A/C SAFETY FUNCTION: These valves open to provide nitrogen to the MSIVs and the SRV accumulators inside the containment. They close for containment isolation.

JUSTIFICATION: Exercising these valves open is performed by charging the bleed-down header following MSIV testing. During plant operation at power, this is 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).

A11-22

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-22 DELETED A11-23

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 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.

Performing this test during cold shutdown requires mobilization of test equipment and as such constitutes adequate justification to defer check valve testing to a refueling outage.

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

A11-24

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 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.

Performing this test during cold shutdown requires mobilization of test equipment and as such constitutes adequate justification to defer check valve testing to a refueling outage.

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

A11-25

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-25 SYSTEM: REACTOR CORE ISOLATION COOLING (RCIC)

COMPONENTS: 13RCIC-7 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. 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.

Performing this test during cold shutdown requires mobilization of test equipment and as such constitutes adequate justification to defer check valve testing to a refueling outage.

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

A11-26

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 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.

Performing this test during cold shutdown requires mobilization of test equipment and as such constitutes adequate justification to defer check valve testing to a refueling outage.

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

A11-27

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 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.

Performing this test during cold shutdown requires mobilization of test equipment and as such constitutes adequate justification to defer check valve testing to a refueling outage.

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

A11-28

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-28 SYSTEM: RESIDUAL HEAT REMOVAL (RHR)

COMPONENTS: 10MOV-17 & 10MOV-18 CATEGORY: A SAFETY FUNCTION: These valves remain closed to protect the RHR System piping and components from over-pressurization during plant operation and inadvertent drain down events while in cold shutdown. 10MOV-17 also performs a containment isolation function.

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

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

A11-29

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant 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 RHR pumps cannot develop sufficient discharge pressure to open these valves. The installed air operators are designed to open these valves at zero differential pressure, which is not practical with the reactor at operating pressure. Therefore, 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.

ALTERNATE TEST: In accordance with recommendations of NUREG-1482, Revision 2, Section 4.1.6, these valves will be forward and reverse flow tested during refueling outages in accordance with ISTC-3522(c) and (f).

A11-30

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-30 SYSTEM: FEEDWATER (FWS)

COMPONENTS: 34NRV-111A, 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, Revision 2, Section 4.1.6, these valves will be forward and reverse flow tested during refueling outages in accordance with ISTC-3522(c) and (f).

A11 - 31

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-31 SYSTEM: Reactor Core Isolation Cooling and High Pressure Coolant Injection COMPONENT: 13MOV-15, 23MOV-15 CATEGORY: A SAFETY FUNCTIONS: 13MOV-15 must close to provide containment isolation and also closes in the event of a RCIC steam line break. RCIC is not a credited ECCS/ESF system; therefore, the 13MOV-15 open function is not a safety function.

23MOV-15 must close to provide containment isolation and also closes in the event of a HPCI steam line break. This valve has an open safety function to supply steam to the HPCI turbine.

TEST REQUIREMENTS: Full stroke exercise and stroke time test quarterly.

JUSTIFICATION: The valves are located in the inerted containment during power operation and are inaccessible. These valves would cause a loss of system function requiring RCIC or HPCI to be inoperable if they were to fail closed during a quarterly test. A unit shutdown would be required to perform corrective maintenance.

The ASME OM Code-of-Record allows licensees to perform testing during cold shutdown if it is not practical to test such valves during power operation. Similarly, the ASME OM Code allows licenses to test valves during each refueling outage if it is impractical to test the valve during cold shutdowns. The staff has determined that it is impractical to de-inert the containment during each cold shutdown outage solely to perform such routine testing or repair activities.

ALTERNATE TEST: Full stroke exercise and stroke time test each refueling.

REFERENCES:

NUREG-1482, Revision 2, Paragraph 3.1.1 Deferring Valve Testing to Each Cold Shutdown or Refueling Outage NUREG-1482, Revision 2, 3.1.1.3 De-inerting Containment of Boiling-Water Reactors To Allow Cold Shutdown Testing A11 - 32

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-32 SYSTEM: Containment Atmosphere Dilution COMPONENT: 27CAD-67, 68, 69 and 70 CATEGORY: A/C SAFETY FUNCTIONS: These valves have a safety function in the open position to provide nitrogen dilution to primary containment to maintain hydrogen and oxygen concentrations below explosive limits.

These valves have a safety function in the closed position to maintain primary containment integrity.

TEST REQUIREMENTS: Reverse closure test quarterly ALTERNATE TEST: Closure test by Appendix J LLRT each refueling outage.

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

During normal plant operation, test equipment to conduct a pressure drop test or local leak rate test must be installed to verify disk position. Valve closure will be verified by local leak rate testing each refueling outage. When it is impracticable to verify check valve closure during plant operation or cold shutdown, it is acceptable to extend the check valve quarterly exercise test (both open and close) to the refueling outage when the closure verification may be performed in conjunction with the Type C leak rate test. The open exercise test may also be performed during the refueling outage or anytime during the fuel cycle interval.

REFERENCES:

NUREG-1482 Rev. 2, section 4.1.6.

A11 - 33

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-33 SYSTEM: Reactor Water Recirculation (RWR)

COMPONENT: 02-2AOV-39 Recirc Loop Inboard Sample Isolation Valve CATEGORY: A TEST REQUIREMENTS: Quarterly Exercise /Stroke Time Close and Fail-Safe testing SAFETY FUNCTION: The recirculation loop sample isolation valve closes to provide containment isolation. The valve has an open function to allow sampling of the water by chemistry.

Deferred Testing: ISTC-3510, 3560, & 5121: Quarterly Exercise /Stroke Time Close and Fail Safe testing JUSTIFICATION: The valve is located in the inerted containment and is inaccessible. In addition, entry into the containment during cold shutdown may be prohibited if the drywell remains inerted.

This valve would cause a loss of the open function (chemistry sampling) if it was to fail closed during a quarterly test. A unit shutdown would be required to perform corrective maintenance Although this valve is capable of being tested at power, operating experience at other Exelon and non-Exelon plants has demonstrated a potential for valves to fail or degrade because of cycling at power (Reference ICES

314926, #314808 and #305581). Failure modes have included severe packing leakage and a loss of containment isolation function. Due to the fact that this valve is located inside the drywell and is inaccessible, the inability of the valve to open or close would result in a degraded system with the potential for an unnecessary plant shutdown and challenge to safety systems.

Additionally, the containment would require de-inerting in order to perform repairs.

Thus the risks associated with testing this valve on line vs during refuel outages* outweigh the benefits of testing quarterly on line. During refuel outages, the primary containment is open and any repairs could be performed immediately without cycling the plant and challenging safety systems during the process of shutting down the plant. Thus this analysis provides the basis, in accordance with ISTC-3521(e) for deferring testing from a quarterly test frequency to refuel outages as noted in NUREG 1482 section 2.4.5 Deferring Valve Testing to Cold Shutdown or Refueling Outages.

It has been determined under this evaluation that testing during cold shutdowns (ref ISTC-3521(c) is not practicable because primary containment (drywell and suppression chambers) is not always made accessible.

REFERENCES:

NUREG-1482 Revision 2, Paragraph 3.1.1 Deferring Valve Testing to Each Cold Shutdown or Refueling Outage NUREG-1482 Revision 2, Paragraph 3.1.1.3 De-inerting Containment if Boiling-Water Reactors to Allow Cold Shutdown Testing A11 - 34

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-34 SYSTEM: Radioactive Waste COMPONENTS: 20MOV-82 Drywell Floor Drain Sump Pump Discharge Inboard Isolation Valve 20MOV-94 RDW Drywell Equipment Drain Sump Pump Discharge Inboard Isolation Valve CATEGORY: A TEST REQUIREMENTS: Quarterly Exercise & Stroke Time Close testing SAFETY FUNCTION: These Drywell Sump inboard isolation valves close to provide containment isolation. They remain open to allow pump out of the sumps for drywell leakage monitoring.

Deferred Testing: ISTC-3510, 5121: Quarterly exercise and stroke time to the closed position.

JUSTIFICATION: Although these valves are capable of being tested at power, operating experience at other Exelon and non-Exelon plants has demonstrated a potential for valves to fail or degrade as a result of cycling at power (Reference ICES #314926, #314808 and #305581). Failure modes have included severe packing leakage and a loss of containment isolation function.

Due to the fact that these valves are located inside the drywell and are inaccessible, the inability of the valves to open or close would result in a degraded system with the potential for an unnecessary plant shutdown and challenge to safety systems. Additionally, the containment would require de-inerting in order to perform repairs.

Thus the risks associated with testing these valves on line vs during refuel outages* outweigh the benefits of testing quarterly on line. During refuel outages, the primary containment is open and any repairs could be performed immediately without cycling the plant and challenging safety systems during the process of shutting down the plant. Thus this analysis provides the basis, in accordance with ISTC-3521(e) for deferring testing from a quarterly test frequency to refuel outages as noted in NUREG 1482 section 2.4.5 Deferring Valve Testing to Cold Shutdown or Refueling Outages.

It has been determined under this evaluation that testing during cold shutdowns (ref ISTC-3521(c) is not practicable because primary containment (drywell and suppression chambers) is not always made accessible.

REFERENCES:

NUREG-1482 Revision 2, Paragraph 3.1.1 Deferring Valve Testing to Each Cold Shutdown or Refueling Outage NUREG-1482 Revision 2, Paragraph 3.1.1.3 De-inerting Containment if Boiling-Water Reactors to Allow Cold Shutdown Testing A11 - 35

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ROJ-35 SYSTEM: Main Steam COMPONENT: 29MOV-74 MST Inboard Line Drain Inboard Isolation Valve CATEGORY: A TEST REQUIREMENTS: Quarterly exercise & stroke time testing.

SAFETY FUNCTION: This main steam line drain valve closes to provide containment isolation.

JUSTIFICATION: Exercising this valve during normal operations or cold shutdown would require a shutdown if maintenance was required. In addition, entry into the containment may be prohibited if the drywell remains inerted.

This valve would cause a loss of system function if it was to fail closed during a quarterly test. A unit shutdown would be required to perform corrective maintenance.

Although this valve is capable of being tested at power, operating experience at other Exelon and non-Exelon plants has demonstrated a potential for valves to fail or degrade as a result of cycling at power (Reference ICES

314926, #314808 and #305581). Failure modes have included severe packing leakage and a loss of containment isolation function. Due to the fact that this valve are located inside the drywell and are inaccessible, the inability of the valve to open or close would result in a degraded system with the potential for an unnecessary plant shutdown and challenge to safety systems. Additionally, the containment would require de-inerting in order to perform repairs.

Thus the risks associated with testing this valve on line vs during refuel outages* outweigh the benefits of testing quarterly on line. During refuel outages, the primary containment is open and any repairs could be performed immediately without cycling the plant and challenging safety systems during the process of shutting down the plant. Thus this analysis provides the basis, in accordance with ISTC-3521(e) for deferring testing from a quarterly test frequency to refuel outages as noted in NUREG 1482 section 2.4.5 Deferring Valve Testing to Cold Shutdown or Refueling Outages.

It has been determined under this evaluation that testing during cold shutdowns (ref ISTC-3521(c) is not practicable because primary containment (drywell and suppression chambers) is not always made accessible.

REFERENCES:

NUREG-1482 Revision 2, Paragraph 3.1.1 Deferring Valve Testing to Each Cold Shutdown or Refueling Outage NUREG-1482 Revision 2, Paragraph 3.1.1.3 De-inerting Containment if Boiling-Water Reactors to Allow Cold Shutdown Testing A11 - 36

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ATTACHMENT 12 TECHNICAL POSITION INDEX

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant TECHNICAL POSITION REV # TITLE NUMBER A12 - 1

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ATTACHMENT 13 TECHNICAL POSITIONS

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant None A13 - 1

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ATTACHMENT 14 INSERVICE TESTING PUMP TABLE

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant DWG No. PUMP TEST RELIEF PUMP ID Description CLASS GROUP CO-ORD TYPE DRIVER TEST FREQUENCY REQUEST 10P-1A RHR Service Water Pump A 3 A FM-20B VLS M P, Vv Quarterly PRR-04 B-6 P, Vv 2YR Comp Pump Test PRR-02 & 04 10P-1B RHR Service Water Pump B 3 A FM-20B VLS M P, Vv Quarterly PRR-04 B-5 P, Vv 2YR Comp Pump Test PRR-02 & 04 10P-1C RHR Service Water Pump C 3 A FM-20B VLS M P, Vv Quarterly PRR-04 C-6 P, Vv 2YR Comp Pump Test PRR-02 &04 10P-1D RHR Service Water Pump D 3 A FM-20B VLS M P, Vv Quarterly PRR-04 C-5 P, Vv 2YR Comp Pump Test PRR-02 & 04 10P-3A Residual Heat Removal 2 A FM-20A C M P, Vv Quarterly PRR-03 & 04 Pump A C-7 P, Vv 2YR Comp Pump Test PRR-03 & 04 10P-3B Residual Heat Removal 2 A FM-20A C M P, Vv Quarterly PRR-03 & 04 Pump B C-4 P, Vv 2YR Comp Pump Test PRR-03 & 04 10P-3C Residual Heat Removal 2 A FM-20A C M P, Vv Quarterly PRR-03 & 04 Pump C C-7 P, Vv 2YR Comp Pump Test PRR-03 & 04 10P-3D Residual Heat Removal 2 A FM-20A C M P, Vv Quarterly PRR-03 & 04 Pump D C-4 P, Vv 2YR Comp Pump Test PRR-03 & 04 11P-2A Standby Liquid Control A 2 B FM-21A PDR M Q Quarterly PRR-04 Pump D-4 Q, Vv 2YR Comp Pump Test PRR-04 11P-2B Standby Liquid Control B 2 B FM-21A PDR M Q Quarterly PRR-04 Pump B-4 Q, Vv 2YR Comp Pump Test PRR-04 14P-1A Core Spray Pump A 2 B FM-23A C M P Quarterly PRR-01 & 04 C-8 P, Vv 2YR Comp Pump Test PRR-04 14P-1B Core Spray Pump B 2 B FM-23A C M P Quarterly PRR-01 & 04 C-3 P, Vv 2YR Comp Pump Test PRR-04 23P-1B High Pressure Coolant 2 B FM-25A C T P Quarterly PRR-04 Injection Booster Pump E-5 P, Vv 2YR Comp Pump Test PRR-04 23P-1M High Pressure Coolant 2 B FM-25A C T P,,N Quarterly PRR-04 Injection Main Pump E-4 P,N, Vv 2YR Comp Pump Test PRR-04 46P-2A Emergency Service 3 B FM-46B VLS M P Quarterly PRR-04 Water Pump A D-8 P, Vv 2YR Comp Pump Test PRR-04 46P-2B Emergency Service 3 B FM-46B VLS M P Quarterly PRR-04 Water Pump B C-8 P, Vv 2YR Comp Pump Test PRR-04 A14 - 1

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ATTACHMENT 15 INSERVICE TESTING VALVE TABLE

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 01-125MOV-100A SGT A Decay Heat Cooling Inlet BTF MO FM-48A C-6 2A O/C B A STO Q Augmented Isolation valve O STC Q PI Y2 01-125MOV-100B SGT B Decay Heat Cooling Inlet BTF MO FM-48A F-6 2A 0/C B A STO Q Augmented Isolation Valve O STC Q PI Y2 01-125MOV-11 SGT RX BLDG suction above 369 BTF MO FM-48A G-8 2A O B A STO Q Augmented EL Isolation Valve C PI Y2 01-125MOV-12 SGT RX BLDG suction below 369 BTF MO FM-48A F-8 2A O B A STO Q Augmented EL Isolation Valve C PI Y2 01-125MOV-14A SGT Filter Train A Inlet Isolation BTF MO FM-48A D-6 2A O/C B A STO Q Augmented Valve C PI Y2 01-125MOV-14B SGT RX BLDG suction above 369 BTF MO FM-48A E-6 2A 0/C B A STO Q Augmented EL Isolation Valve C PI Y2 01-125MOV-15A SGT Fan A Discharge Isolation BTF MO FM-48A D-3 2A O B A STO Q Augmented Valve C PI Y2 01-125MOV-15B SGT Fan B Discharge Isolation BTF MO FM-48A F-3 2A O B A STO Q Augmented Valve C PI Y2 02AOV-17 ADS Reactor Head Vent Inboard GL AO FM-29A G-7 1 C B P PI Y2 C

02AOV-18 ADS Reactor Head Vent GL AO FM-29A G-7 1 C B P PI Y2 Outboard C 02RV-1 ADS RV-71A Discharge Vacuum CK SA FM-29A D-7 2 C C A SC/SO RR ROJ-04 Relief O 02RV-2 ADS RV-71B Discharge Vacuum CK SA FM-29A D-7 2 C C A SC/SO RR ROJ-04 Relief O 02RV-3 ADS RV-71C Discharge Vacuum CK SA FM-29A D-7 2 C C A SC/SO RR ROJ-04 Relief O 02RV-4 ADS RV-71D Discharge Vacuum CK SA FM-29A D-7 2 C C A SC/SO RR ROJ-04 Relief O 02RV-5 ADS RV-71E Discharge Vacuum CK SA FM-29A D-7 2 C C A SC/SO RR ROJ-04 Relief O 02RV-6 ADS RV-71F Discharge Vacuum CK SA FM-29A D-7 2 C C A SC/SO RR ROJ-04 Relief O 02RV-7 ADS RV-71G Discharge Vacuum CK SA FM-29A D-7 2 C C A SC/SO RR ROJ-04 Relief O 02RV-8 ADS RV-71H Discharge Vacuum CK SA FM-29A D-7 2 C C A SC/SO RR ROJ-04 Relief O 02RV-9 ADS RV-71J Discharge Vacuum CK SA FM-29A D-7 2 C C A SC/SO RR ROJ-04 Relief O 02RV-10 ADS RV-71K Discharge Vacuum CK SA FM-29A D-7 2 C C A SC/SO RR ROJ-04 Relief O 02RV-11 ADS RV-71L Discharge Vacuum CK SA FM-29A D-7 2 C C A SC/SO RR ROJ-04 Relief O A15 - 1

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 02RV-71A ADS Main Steam Line A RV SA, FM-29A G-6 1 O/C B/C A ETC RR SOV only Safety/Relief Valve AO C ETO RR SOV only RLF 6 YR VRR-03 02RV-71B ADS Main Steam Line A Safety RV SA, FM-29A G-6 1 O/C B/C A ETC RR SOV only Relief Valve AO C ETO RR SOV only RLF 6 YR VRR-03 02RV-71C ADS Main Steam Line B Safety RV SA, FM-29A G-6 1 O/C B/C A ETC RR SOV only Relief Valve AO C ETO RR SOV only RLF 6 YR VRR-03 02RV-71D ADS Main Steam Line C Safety RV SA, FM-29A F-6 1 O/C B/C A ETC RR SOV only Relief Valve AO C ETO RR SOV only RLF 6 YR VRR-03 02RV-71E ADS Main Steam Line C Safety RV SA, FM-29A F-7 1 O/C B/C A ETC RR SOV only Relief AO C ETO RR SOV only RLF 6 YR VRR-03 02RV-71F Main Steam Line C Safety Relief RV SA, FM-29A F-7 1 O/C B/C A ETC RR SOV only Valve AO C ETO RR SOV only RLF 6 YR VRR-03 02RV-71G ADS Main Steam Line D Safety RV SA, FM-29A F-7 1 O/C B/C A ETC RR SOV only Relief Valve AO C ETO RR SOV only RLF 6 YR VRR-03 02RV-71H ADS Main Steam Line D Safety RV SA, FM-29A G-7 1 O/C B/C A ETC RR SOV only Relief AO C ETO RR SOV only RLF 6 YR VRR-03 02RV-71J Main Steam Line A Safty/Relief RV SA, FM-29A G-7 1 O/C B/C A ETC RR SOV only AO C ETO RR SOV only RLF 6 YR VRR-03 02RV-71K Main Steam Line A Safety/Relief RV SA, FM-29A G-6 1 O/C B/C A ETC RR SOV only AO C ETO RR SOV only RLF 6 YR VRR-03 02RV-71L Main Steam Line D Safety Relief RV SA, FM-29A F-7 1 O/C B/C A ETC RR SOV only Valve AO C ETO RR SOV only RLF 6 YR VRR-03 02VB-1 ADS RV-71A Discharge Vacuum CK SA FM-29A C-7 2 O/C C A SC/SO RR ROJ-4 Breaker C 02VB-2 ADS RV-71B Discharge Vacuum CK SA FM-29A C-7 2 O/C C A SC/SO RR ROJ-4 Breaker C 02VB-3 ADS RV-71C Discharge Vacuum CK SA FM-29A C-7 2 O/C C A SC/SO RR ROJ-4 Breaker C 02VB-4 ADS RV-71D Discharge Vacuum CK SA FM-29A C-7 2 O/C C A SC/SO RR ROJ-4 Breaker C 02VB-5 ADS RV-71E Discharge Vacuum CK SA FM-29A C-7 2 O/C C A SC/SO RR ROJ-4 Breaker C 02VB-6 ADS RV-71F Discharge Vacuum CK SA FM-29A C-7 2 O/C C A SC/SO RR ROJ-4 Breaker C A15 - 2

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 02VB-7 ADS RV-71G Discharge Vacuum CK SA FM-29A C-7 2 O/C C A SC/SO RR ROJ-4 Breaker C 02VB-8 ADS RV-71H Discharge Vacuum CK SA FM-29A C-7 2 O/C C A SC/SO RR ROJ-4 Breaker C 02VB-9 ADS RV-71J Discharge Vacuum CK SA FM-29A C-7 2 O/C C A SC/SO RR ROJ-4 Breaker C 02VB-10 ADS RV-71K Discharge Vacuum CK SA FM-29A C-7 2 O/C C A SC/SO RR ROJ-4 Breaker C 02VB-11 ADS RV-71L Discharge Vacuum CK SA FM-29A C-7 2 O/C C A SC/SO RR ROJ-4 Breaker C 02-2AOV-39 Recirc Loop Inboard Sample GA AO FM-26A E-4 1 C A A FC RR ROJ-33 Isolation Valve O LTJ M30 PI Y2 STC RR ROJ-33 02-2AOV-40 Recirc Loop Outboard Sample GA AO FM-26A E-3 1 C A A FC Q Isolation Valve O LTJ M30 PI Y2 STC Q 02-2EFV1-DPT- Recirc Pump A 02DPT-111A Lo XFC SA FM-26A E-3 1 C A/C A CC RR VRR-02 111A Side X-32BC Excess Flow Check O LT VRR-02 Valve 02-2EFV1-DPT- Recirc Pump B 02DPT-111B Lo XFC SA FM-26A E-8 1 C A/C A CC RR VRR-02 111B Side X-32AA Excess Flow Check O LT VRR-02 Valve 02-2EFV1-FT-110A Recirc Loop B 02FT-110A&B Lo XFC SA FM-26A F-3 1 C A/C A CC RR VRR-02 Side X-31BB Excess Flow Check O LT VRR-02 Valve 02-2EFV1-FT-110C Recirc Loop A 02FT-110C&B Lo XFC SA FM-26A D-3 1 C A/C A CC RR VRR-02 Side X-31BB Excess Flow Check O LT VRR-02 Valve 02-2EFV1-FT-110E Recirc Loop B 02FT-110E&F Hi XFC SA FM-26A F-8 1 C A/C A CC RR VRR-02 Side X-32AC Excess Flow Check O LT VRR-02 Valve 02-2EFV1-FT-110G Recic Loop B 02FT-110G&H Lo XFC SA FM-26A D-8 1 C A/C A CC RR VRR-02 Side X-31AB Excess Flow Check O LT VRR-02 Valve 02-2EFV2-DPT- Recirc Pump ADS Logic 02DPT- XFC SA FM-26A E-3 1 C A/C A CC RR VRR-02 111A 111A Hi Side X-32BD Excess O LT VRR-02 Flow Check Valve 02-2EFV2-DPT- Recirc Pump B ADS Logic 02DPT- XFC SA FM-26A E-8 1 C A/C A CC RR VRR-02 111B 111B Hi Side X-32AD Excess O LT VRR-02 Flow Check Valve 02-2EFV2-FT-110A Recirc Loop A ADS Logic 02ft- XFC SA FM-26A F-3 1 C A/C A CC RR VRR-02 110A & E Hi Side X-32AE Excess O LT VRR-02 Flow Check Valve A15 - 3

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 02-2EFV2-FT-110C Recirc Loop A ADS Logic 02ft- XFC SA FM-26A D-3 1 C A/C A CC RR VRR-02 110C & D Hi Side X-32BB Excess O LT VRR-02 Flow Check Valve 02-2EFV2-FT-110E Recirc Loop B ADS Logic 02FT- XFC SA FM-26A F-8 1 C A/C A CC RR VRR-02 110E&F Lo Side X-32AB Excess O LT VRR-02 Flow Check Valve 02-2EFV2-FT-110G Recirc Loop B ADS Logic 02FT- XFC SA FM-26A D-8 1 C A/C A CC RR VRR-02 110G&H Lo Side X-28BC Excell O LT VRR-02 Flow Check Valve 02-2EFV-PS-128A RWR Loop A ADS Logic 02PS- XFC SA FM-26A B-6 1 C A/C A CC RR VRR-02 128A X-31BA Excess Flow Check O LT VRR-02 Valve 02-2EFV-PS-128B RWR Loop B ADS Logic 02PS- XFC SA FM-26A B-6 1 C A/C A CC RR VRR-02 128B X-31AA Excess Flow Check O LT VRR-02 Valve 02-2EFV-24A Recirc Pump A Seal Cavity 2 XFC SA FM-26A C-3 1 C A/C A CC RR VRR-02 Pressure Instruments X-56D O LT VRR-02 Excess Flow Check Valve 02-2EFV-24B Recirc Pump B Seal Cavity 2 XFC SA FM-26A C-8 1 C A/C A CC RR VRR-02 Pressure Instruments X-40EB O LT VRR-02 Excess Flow Check Valve 02-2EFV-25A Recirc Pump A Seal Cavity 1 XFC SA FM-26A C-3 1 C A/C A CC RR VRR-02 Pressure Instruments X-56A O LT VRR-02 Excess Flow Check Valve 02-2EFV-25B Recirc Pump B Seal Cavity 1 XFC SA FM-26A C-8 1 C A/C A CC RR VRR-02 Pressure Instruments X-40EA O LT VRR-02 Excess Flow Check Valve 02-2MOV-53A Reactor Water Recirc Pump A GA SA FM-26A C-3 1 C B A PI RR Discharge Isolation Valve O STC CS CSJ-01 02-2MOV-53B Reactor Water Recirc Pump B GA SA FM-26A C-8 1 C B A PI RR Discharge Isolation Valve O STC CS CSJ-01 02-2RWR-13A RWR Pump A Seal Purge Supply CK SA FM-26A C-3 1 C A/C A CCF Q Check Valve O COF RR ROJ-02 LTJ AJ 02-2RWR-13B RWR Pump B Seal Purge Supply CK SA FM-26A C-8 1 C A/C A CCF Q Check Valve O COF RR ROJ-02 LTJ AJ 02-2RWR-41A Recirc Pump A Seal Purge Check CK SA FM-26A D-3 1 C A/C A CCF Q Valve O COF RR ROJ-03 LTJ AJ 02-2RWR-41B Recirc Pump A Seal Purge Check CK SA FM-26A D-8 1 C A/C A CCF Q Valve O COF RR ROJ-03 LTJ AJ 02-3EFV-11 Reactor Vessel Instrumentation XFC SA FM-47A F-7 1 C A/C A CC RR VRR-02 Excess Flow O LT RR A15 - 4

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 02-3EFV-13A Reactor Vessel Instrumentation XFC SA FM-47A E-7 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-13B Reactor Vessel Instrumentation XFC SA FM-47A E-4 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-15A Reactor Vessel Instrumentation XFC SA FM-47A E-7 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-15B Reactor Vessel Instrumentation XFC SA FM-47A E-4 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-15N Reactor Vessel Instrumentation XFC SA FM-47A B-7 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-17A Reactor Vessel Instrumentation XFC SA FM-47A D-7 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-17B Reactor Vessel Instrumentation XFC SA FM-47A D-4 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-19A Reactor Vessel Instrumentation XFC SA FM-47A D-7 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-19B Reactor Vessel Instrumentation XFC SA FM-47A D-4 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-21A Jet Pump Instrumentation XFC SA FM-47A H-5 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-21B Jet Pump Instrumentation XFC SA FM-47A C-7 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-21C Jet Pump Instrumentation XFC SA FM-47A C-4 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-21D Jet Pump Instrumentation XFC SA FM-47A H-4 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-23 Reactor Vessel Instrumentation XFC SA FM-47A F-7 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-23A Jet Pump Instrumentation XFC SA FM-47A H-5 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-23B Jet Pump Instrumentation XFC SA FM-47A D-7 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-23C Jet Pump Instrumentation XFC SA FM-47A D-4 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-23D Jet Pump Instrumentation XFC SA FM-47A C-7 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-25 CRD Pressure Sensing XFC SA FM-47A C-7 1 C A/C A CC RR VRR-02 Instrumentation Excess Flow O LT RR 02-3EFV-31A Jet Pump Instrumentation XFC SA FM-47A H-5 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-31B Jet Pump Instrumentation XFC SA FM-47A H-5 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-31C Jet Pump Instrumentation XFC SA FM-47A H-5 1 C A/C A CC RR VRR-02 Excess Flow O LT RR A15 - 5

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 02-3EFV-31D Jet Pump Instrumentation XFC SA FM-47A H-5 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-31E Jet Pump Instrumentation XFC SA FM-47A D-7 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-31F Jet Pump Instrumentation XFC SA FM-47A H-5 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-31G Jet Pump Instrumentation XFC SA FM-47A G-5 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-31H Jet Pump Instrumentation XFC SA FM-47A G-5 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-31J Jet Pump Instrumentation XFC SA FM-47A H-4 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-31K Jet Pump Instrumentation XFC SA FM-47A H-4 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-31L Jet Pump Instrumentation XFC SA FM-47A H-4 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-31M Jet Pump Instrumentation XFC SA FM-47A D-4 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-31N Jet Pump Instrumentation XFC SA FM-47A H-4 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-31P Jet Pump Instrumentation XFC SA FM-47A H-4 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-31R Jet Pump Instrumentation XFC SA FM-47A G-4 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-31S Jet Pump Instrumentation XFC SA FM-47A G-4 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 02-3EFV-33 Jet Pump Instrumentation XFC SA FM-47A B-4 1 C A/C A CC RR VRR-02 Excess Flow O LT RR 03AOV-126 HCU Inlet Scram (Typical of 137) GL AO FM-27B C-4 2 O B A Exempt-Skid C

03AOV-127 HCU Outlet Scram (Typical of GL AO FM-27B D-4 2 O B A Exempt-Skid 137) C 03AOV-32 Scram Discharge Volume B Vent GL AO FM-27B H-4 2 C B A FC Q O STC Q PI Y2 03AOV-33 Scram Discharge Volume B GL AO FM-27B F-4 2 C B A FC Q Drain O STC Q PI Y2 03AOV-34 Scram Discharge Volume B Vent GL AO FM-27B H-4 2 C B A FC Q O STC Q PI Y2 03AOV-35 Scram Discharge Volume B GL AO FM-27B F-4 2 C B A FC Q Drain O STC Q PI Y2 A15 - 6

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 03AOV-36 Scram Discharge Volume A Vent GL AO FM-27B H-6 2 C B A FC Q O STC Q PI Y2 03AOV-37 Scram Discharge Volume A GL AO FM-27B F-6 2 C B A FC Q Drain O STC Q PI Y2 03AOV-38 Scram Discharge Volume A Vent GL AO FM-27B H-6 2 C B A FC Q O STC Q PI Y2 03AOV-39 Scram Discharge Volume A GL AO FM-27B F-6 2 C B A FC Q Drain O STC Q PI Y2 03HCU-114 Scram Discharge Line Check XFC SA FM-27B D-4 2 O C A Exempt Skid (Typical of 137) O 03HCU-115 Charging Water Inlet (Typical of XFC SA FM-27B C-4 2 C C A ROJ-27 Exempt Skid 137) C 03HCU-138 Cooling Water Supply (Typical of XFC SA FM-27B C-4 2 C C A Exempt Skid 137) C 03SOV-120 Withdraw Settle Solenoid GA AO FM-27B C-4 2 C B A Exempt Skid O

03SOV-121 Insert Exhaust Solenoid GA AO FM-27B C-4 2 C B A Exempt Skid O

03SOV-122 Withdraw Drive Water Solenoid GA AO FM-27B C-4 2 C B A Exempt Skid O

03SOV-123 Withdraw Settle Solenoid GA AO FM-27B C-4 2 C B A Exempt Skid O

07EV-104A TIP A Explosive Shear Valve SHR EXP FM- F-5 2A C D A DT Y2 Augmented 119A O 07EV-104B TIP B Explosive Shear Valve SHR EXP FM- F-4 2A C D A DT Y2 Augmented 119A O 07EV-104C TIP C Explosive Shear Valve SHR EXP FM- F-4 2A C D A DT Y2 Augmented 119A O 07SOV-104A TIP A Ball Valve BAL SO FM- F-5 2A C A A FC Q 119A O/C LTJ M60 PI Y2 STC Q VRR-01 07SOV-104B TIP B Ball Valve BAL SO FM- F-4 2A C A A FC Q 119A O/C LTJ M60 PI Y2 STC Q VRR-01 07SOV-104C TIP C Ball Valve BAL SO FM- F-4 2A C A A FC Q 119A O/C LTJ M60 PI Y2 STC Q VRR-01 A15 - 7

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 10AOV-68A RHR A LPCI Testable Check CK SA, FM-20A F-6 1 O/C A/C A COF RR ROJ-29 Valve AO C CCF RR ROJ-29 LTH Y2 VRR-04 10AOV-68B RHR B LPCI Testable Check CK SA, FM-20A F-5 1 O/C A/C A COF RR ROJ-29 Valve AO O CCF RR ROJ-29 LTH Y2 VRR-04 10AOV-71A RHR Heat Exchanger A Outlet to GL AO FM-20B F-6 2 C B A PI Y2 Torus or RCIC Isol Valve C 10AOV-71B RHR Heat Exchanger B Outlet to GL AO FM-20B F-5 2 C B A PI Y2 Torus or RCIC Isol Valve C 10MOV-12A RHR Heat Exchanger A Outlet GA MO FM-20B B-6 2 O B P PI Y2 Isolation Valve O 10MOV-12B RHR Heat Exchanger B Outlet GA MO FM-20B F-5 2 O B P PI Y2 Isolation Valve O 10MOV-13A RHR Pump A Suct Torus GA MO FM-20A B-6 2 O/C B A PI Y2 Isolation valve O STC Q STO Q 10MOV-13B RHR Pump B Suct Torus GA MO FM-20A C-4 2 O/C B A PI Y2 Isolation valve O STC Q STO Q 10MOV-13C RHR Pump C Suct Torus GA MO FM-20A C-6 2 O/C B A PI Y2 Isolation valve O STC Q STO Q 10MOV-13D RHR Pump D Suct Torus GA MO FM-20A C-5 2 O/C B A PI Y2 Isolation valve O STC Q STO Q 10MOV-148A RHRSW A To RHR Cross Tie GA MO FM-20B E-8 3 C B P PI Y2 Upstream Isolation Valve C 10MOV-148B RHRSW B To RHR Cross Tie GA MO FM-20B E-2 3 C B P PI Y2 Upstream Isolation Valve C 10MOV-149A RHRSW A To RHR Cross Tie GA MO FM-20B D-8 3 C B P PI Y2 Downstream Isolation Valve C 10MOV-149B RHRSW B To RHR Cross Tie GA MO FM-20B D-2 3 C B P PI Y2 Downstream Isolation Valve C 10MOV-15A RHR A Pump Shutdown Cooling GA MO FM-20A C-6 2 C B A PI Y2 Isolation Valve O/C STC Q 10MOV-15B RHR B Pump Shutdown Cooling GA MO FM-20A C-4 2 C B A PI Y2 Isolation Valve O/C STC Q 10MOV-15C RHR C Pump Shutdown Cooling GA MO FM-20A C-6 2 C B A PI Y2 Isolation Valve O/C STC Q 10MOV-15D RHR D Pump Shutdown Cooling GA MO FM-20A C-4 2 C B A PI Y2 Isolation Valve O/C STC Q 10MOV-167A RHR Heat Exch A DNSTR Vent to GL MO FM-20B F-8 2 C B P PI Y2 Torus Isolation Valve C A15 - 8

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 10MOV-167B RHR Heat Exch A DNSTR Vent to GL MO FM-20B F-3 2 C B P PI Y2 Torus Isolation Valve C 10MOV-16A RHR Loop A Min Flow Isolation GA MO FM-20A D-8 2 O/C B A PI Y2 Valve O/C STC Q STO Q 10MOV-16B RHR Loop B Min Flow Isolation GA MO FM-20A D-3 2 O/C B A PI Y2 Valve O/C STC Q STO Q 10MOV-17 RHR Shutdown Cooling GA MO FM-20A D-5 1 C A A LT VRR-04 Outboard Isolation Valve C PI Y2 ROJ-28 STC RR 10MOV-18 RHR Shutdown Cooling Inboard GA MO FM-20A E-5 1 C A A LT VRR-04 Isolation Valve C PI Y2 ROJ-28 STC RR 10MOV-25A RHR A LPCI Inboard Injection GA MO FM-20A F-8 1 O/C A A LT VRR-04 Valve C LTJ AJ PI Y2 STC CS CSJ-10 STO CS CSJ-10 10MOV-25B RHR B LPCI Inboard Injection GA MO FM-20A F-3 1 O/C A A LT VRR-04 Valve C LTJ AJ PI Y2 STC CS CSJ-10 STO CS CSJ-10 10MOV-26A RHR A Containment Spray GA MO FM-20A G-7 2 O/C A A PI Y2 Outboard Isolation Valve C STC Q STO Q 10MOV-26B RHR B Containment Spray GA MO FM-20A G-4 2 O/C A A PI Y2 Outboard Isolation Valve C STC Q STO Q 10MOV-27A RHR A LPCI Outboard Injection AN- MO FM-20A F-8 1 O/C A A PI Y2 Angle Globe Valve Valve GL O/C STC Q STO Q 10MOV-27B RHR B LPCI Outboard Injection AN- MO FM-20A F-3 1 O/C A A PI Y2 Angle Globe Valve Valve GL O/C STC Q STO Q 10MOV-31A RHR A Containment Spray GL MO FM-20A G-6 2 O/C A A LT AJ Inboard Isolation Valve C PI Y2 STC Q STO Q 10MOV-31B RHR B Containment Spray GL MO FM-20A G-5 2 O/C A A LT AJ Inboard Isolation Valve C PI Y2 STC Q STO Q A15 - 9

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 10MOV-34A RHR A Torus Cooling Supply GL MO FM-20A E-7 2 O/C B A PI Y2 Valve C STC Q STO Q 10MOV-34B RHR B Torus Cooling Supply GL MO FM-20A E-3 2 O/C B A PI Y2 Valve C STC Q STO Q 10MOV-38A RHR A To Torus Spray Isolation GL MO FM-20A E-7 2 O/C A A LTJ AJ Valve C PI Y2 STC Q STO Q 10MOV-38B RHR B To Torus Spray Isolation GL MO FM-20A E-4 2 O/C A A LTJ AJ Valve C PI Y2 STC Q STO Q 10MOV-39A RHR A Torus Cooling Isolation GL MO FM-20A E-8 2 O/C A A PI Y2 Valve C STC Q STO Q 10MOV-39B RHR Loop B Torus Cooling GL MO FM-20A E-3 2 O/C A A PI Y2 Isolation Valve C STC Q STO Q 10MOV-65A RHR Heat Exchanger A Shell GL MO FM-20B G-6 2 O B P PI Y2 Inlet Isolation Valve O 10MOV-65B RHR Heat Exchanger B Shell GL MO FM-20B G-5 2 O B P PI Y2 Inlet Isolation Valve O 10MOV-66A RHR Heat Exchanger A Bypass GL MO FM-20A D-8 2 O/C B A PI Y2 valve O STC Q STO Q 10MOV-66B RHR Heat Exchanger B Bypass GL MO FM-20A D-3 2 O/C B A PI Y2 valve O STC Q STO Q 10MOV-89A RHR Heat Exchanger A Service GA MO FM-20B D-6 3 O B A PI Y2 Water Outlet Isolation Valve C STC Q STO Q 10MOV-89B RHR Heat Exchanger B Service GA MO FM-20B E-5 3 O B A PI Y2 Water Outlet Isolation Valve C STC Q STO Q 10RHR-14A RHRSW Pump A Discharge CK SA FM-20B B-7 3 O/C C A DIS RR Check Valve C CCF Q CVCM JAF-03 COF Q 10RHR-14B RHRSW Pump B Discharge CK SA FM-20B B-4 3 O/C C A DIS RR Check Valve C CCF Q CVCM JAF-03 COF Q 10RHR-14C RHRSW Pump C Discharge CK SA FM-20B C-7 3 O/C C A DIS RR Check Valve C CCF Q CVCM JAF-03 COF Q A15 - 10

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 10RHR-14D RHRSW Pump D Discharge CK SA FM-20B C-4 3 O/C C A DIS RR Check Valve C CCF Q CVCM JAF-03 COF Q 10RHR-262 RHR Loop B Reactor Head Spray CK SA FM-20A H-3 2 C C A LT Keep Full Inner Check Valve O/C COF 10RHR-277 RHR Loop A Reactor Head Spray CK SA FM-20A G-8 2 C C A LT Keep Full Inner Check Valve O/C COF 10RHR-42A RHR Pump A Discharge Check CK SA FM-20A C-8 2 O/C C A DIS RR Valve O/C CCF Q COF Q 10RHR-42B RHR Pump B Discharge Check CK SA FM-20A C-3 2 O/C C A DIS RR Valve O/C CCF Q COF Q 10RHR-42C RHR Pump C Discharge Check CK SA FM-20A C-8 2 O/C C A DIS RR Valve O/C CCF Q COF Q 10RHR-42D RHR Pump D Discharge Check CK SA FM-20A C-3 2 O/C C A DIS RR Valve O/C CCF Q COF Q 10RHR-64A RHR Pump A Min Flow Check CK SA FM-20A C-8 2 O/C C A DIS RR ROJ-05 Valve O/C CCF Q CVCM JAF-02 COF Q ROJ-05 10RHR-64B RHR Pump B Min Flow Check CK SA FM-20A C-3 2 O/C C A DIS RR ROJ-05 Valve O/C CCF Q CVCM JAF-02 COF Q ROJ-05 10RHR-64C RHR Pump C Min Flow Check CK SA FM-20A D-8 2 O/C C A DIS RR ROJ-05 Valve O/C CCF Q CVCM JAF-02 COF Q ROJ-05 10RHR-64D RHR Pump D Min Flow Check CK SA FM-20A D-3 2 O/C C A DIS RR ROJ-05 Valve O/C CCF Q CVCM JAF-02 COF Q ROJ-05 10RHR-95A RHR Keep Full Pump A Min Flow CK SA FM-20A D-8 2 C C A CCF RR ROJ-06 Check Valve O/C COF Q 10RHR-95B RHR Keep Full Pump B Min Flow CK SA FM-20A B-5 2 C C A CCF RR ROJ-06 Check Valve O/C COF Q 10RV-41A RHR Pump A Suction Relief RV SA FM-20A C-7 2 C C A RT Y10 Thermal Valve C 10RV-41B RHR Pump B Suction Relief RV SA FM-20A C-4 2 C C A RT Y10 Thermal Valve C 10RV-41C RHR Pump C Suction Relief RV SA FM-20A C-7 2 C C A RT Y10 Thermal Valve C 10RV-41D RHR Pump D Suction Relief RV SA FM-20A C-4 2 C C A RT Y10 Thermal Valve C A15 - 11

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 10RV-43A RHR Heat Exchanger A Tube RV SA FM-20B E-7 3 O C A RT Y10 Thermal Side Relief Valve C 10RV-43B RHR Heat Exchanger B Tube RV SA FM-20B E-4 3 O C A RT Y10 Thermal Side Relief Valve C 10RV-46A RHR Heat Exchanger A Shell RV SA FM-20B F-7 2 O C A RT Y10 Thermal Side Relief Valve C 10RV-46B RHR Heat Exchanger B Shell Side RV SA FM-20B F-3 2 O C A RT Y10 Thermal Relief Valve C 10SOV-203 RHR Loop A Sample Solenoid GA SO FM-18C E-7 2 C B P PI Y2 Valve C 10SOV-204 RHR Loop B Sample Solenoid GA SO FM-18C D-7 2 C B P PI Y2 Valve C 10SV-35A RHR Loop A Safety Valve RV SA FM-20A E-8 2 C C A RT Y10 Y4 due to past C performance 10SV-35B RHR Loop B Safety Valve RV SA FM-20A E-3 2 C C A RT Y10 Y4 due to past C performance 10SV-40 RHR Shutdown Cooling Relief RV SA FM-20A D-5 2 C C A RT Y10 Y4 Group of one Valve C 11EV-14A SLC A Double Squib Activated SHR EXP FM-21A D-6 1 O D A DT Y2 Shear Explosive Valve C 11EV-14B SLC B Double Squib Activated SHR EXP FM-21A B-6 1 O D A DT Y2 Shear Explosive Valve C 11SLC-16 SLC Pumps Discharge to RX CK SA FM-21A C-7 1 O/C A/C A COF RR ROJ-07 Outboard Check Valve O/C CCF RR ROJ-07 11SLC-17 SLC Pumps Discharge to RX CK SA FM-21A D-7 1 O/C A/C A COF RR ROJ-07 Inboard Check Valve O/C CCF RR ROJ-07 11SLC-18 SLC Pumps Discharge to RX GL M FM-21A D-7 1 O B P PI Y2 Inboard Check Valve O 11SLC-43A SLC Pump A Discharge Check CK SA FM-21A D-6 2 O C A DIS CM Valve O/C CCF Q CVCM JAF-04 COF Q 11SLC-43B SLC Pump B Discharge Check CK SA FM-21A B-6 2 O C A DIS CM Valve O/C CCF Q CVCM JAF-04 COF Q 11SV-39A SLC Pump 2A Discharge Safety RV SA FM-21A D-4 2 C C A RT Y10 Y8 Group of two Valve C 11SV-39B SLC Pump 2B Discharge Safety RV SA FM-21A C-4 2 C C A RT Y10 Y8 Group of two Valve C 12MOV-15 RWCU Supply Inboard Isolation GA MO FM-24A E-8 1 C A A LTJ AJ Valve O PI Y2 STC CS CSJ-07 12MOV-18 RWCU Supply Outboard GA MO FM-24A E-7 1 C A A LTJ AJ Isolation Valve O PI Y2 STC CS CSJ-07 A15 - 12

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 12MOV-69 RWCU Return Containment GA MO FM-24A H-7 1 C A A LTJ AJ Isolation Valve O PI Y2 STC CS CSJ-07 13EFV-01A RCIC Turbine Steam Supply Leak CK SA FM-22A G-7 1 C A/C A CC RR VRR-02 Detect Instruments Excess Flow O LK RR VRR-02 Check Valve 13EFV-01B RCIC Turbine Steam Supply Leak CK SA FM-22A G-7 1 C A/C A CC RR VRR-02 Detect Instruments Excess Flow O LK RR VRR-02 Check Valve 13EFV-02A RCIC Turbine Steam Supply Leak CK SA FM-22A G-7 1 C A/C A CC RR VRR-02 Detect Instruments Excess Flow O LK RR VRR-02 Check Valve 13EFV-02B RCIC Turbine Steam Supply Leak CK SA FM-22A F-7 1 C A/C A CC RR VRR-02 Detect Instruments Excess Flow O LK RR VRR-02 Check Valve 13MOV-130 RCIC Turbine Exhaust Line GA MO FM-22A E-6 2 O B P PI Y2 Vacuum Breaker Valve O 13MOV-15 RCIC Steam Supply Inboard GA MO FM-22A F-7 1 C A A LTJ AJ Isolation Valve O PI Y2 STC RR ROJ-31 13MOV-16 RCIC Turbine Steam Supply GA MO FM-22A F-7 1 C A A LTJ AJ Outboard Isolation Valve O PI Y2 STC Q 13MOV-21 RCIC Pump Discharge to Reactor GA MO FM-22A F-5 1 C A A LTJ AJ Inboard Valve C PI Y2 STC Q 13MOV-27 RCIC Pump Min Flow Isolatio GL MO FM-22A E-5 2 C B A PI Y2 Valve C STC Q 13MOV-41 RCIC Pump Suction From GA MO FM-22A D-7 2 C B A PI Y2 Suppression Pool Inboard C STC Q Isolation Valve 13RCIC-22 RCIC Testable Check Valve CK SA FM-22A F-6 1A C C A SO CS CSJ-09 Augmented C

13RCIC-37 RCIC Turbine Exhaust Vacuum CK SA FM-22A E-6 2 O C A DIS Y2 Break Check Valve O CCF RR ROJ-24 CCO RR ROJ-24 13RCIC-38 RCIC Turbine Exhaust Vacuum CK SA FM-22A E-6 2 O C A DIS Y2 Break Check Valve O CCF RR ROJ-24 CCO RR ROJ-24 13RCIC-4 RCIC Turbine Exhaust to Tours CK SA FM-22A D-6 2 O/C A/C A LTJ AJ Downstream Check Valve O CCF RR ROJ-08 CVCM JAF-05 COF Q 13RCIC-5 RCIC Turbine Exhaust to Torus CK SA FM-22A C-6 2 O/C A/C A LTJ AJ Upstream Check Valve O CCF RR ROJ-08 CVCM JAF-05 COF Q A15 - 13

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 13RCIC-7 RCIC Vacuum Pump P-3 CK SA FM-22A C-7 2 C C A CCF RR ROJ-25 Discharge Stop Check Valve O COF RR ROJ-25 14AOV-13A CSP A Reactor Isolation Testable CK SA, FM-23A G-6 1 O/C A/C A LT RR VRR-04 Check Valve AO C PI Y2 CVCM JAF-06 SC/SO RR ROJ-09 14AOV-13B CSP B Reactor Isolation Testable CK SA, FM-23A G-5 1 O/C A/C A LT RR VRR-04 Check Valve AO C PI Y2 CVCM JAF-06 SC/SO RR ROJ-09 14CSP-10A Core Spray Pump A Discharge CK SA FM-23A D-8 2 O C A CCF Q CVCM JAF-07 Check Valve C COF Q 14CSP-10B Core Spray Pump B Discharge CK SA FM-23A D-3 2 O C A CCF Q CVCM JAF-07 Check Valve C COF Q 14CSP-62A Core Spray Hold Pump A Min CK SA FM-23A E-7 2 O/C C A CCF RR ROJ-10 Flow Check Valve C COF Q 14CSP-62B Core Spray Hold Pump B Min CK SA FM-23A E-3 2 O/C C A CCF RR ROJ-10 Flow Check Valve C COF Q 14CSP-76A Core Spray Loop A Keep Full CK SA FM-23A F-7 2 C C A CCF Q Check Valve O COF Q 14CSP-76B Core Spray Loop B Keep Full CK SA FM-23A F-4 2 C C A CCF Q Check Valve O COF Q 14EFV-31A Core Spray Loop A Spray Nozzle CK SA FM-23A E-4 1 O/C A/C A LT RR VRR-02 14DPIS-43A Excess Flow Check O SC RR VRR-02 Valve 14EFV-31B Core Spray Loop B Spray Nozzle CK SA FM-23A E-4 1 O/C A/C A LT RR VRR-02 14DPIS-43A Excess Flow Check O SC RR VRR-02 Valve 14MOV-11A Core Spray Loop A Outboard GS MO FM-23A F-7 1 O A A PI Y2 Isolation Valve O STC Q STO Q 14MOV-11B Core Spray Loop B Outboard GS MO FM-23A F-4 1 O A A PI Y2 Isolation Valve O STC Q STO Q 14MOV-12A Core Spray Loop A Inboard GA MO FM-23A F-6 1 O/C A A LTJ AJ Isolation Valve C LT VRR-04 PI Y2 STC CS CSJ-11 STO CS CSJ-11 14MOV-12B Core Spray Loop B Inboard GA MO FM-23A F-4 1 O/C A A LTJ AJ Isolation Valve C LT VRR-04 PI Y2 STC CS CSJ-11 STO CS CSJ-11 14MOV-26A CSP A Full Flow Test to GL MO FM-23A F-7 2 C B A PI Y2 Suppression Pool Isolation Valve C STC Q A15 - 14

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 14MOV-26B CSP B Full Flow Test to GL MO FM-23A F-3 2 C B A PI Y2 Suppression Pool Isolation Valve C STC Q 14MOV-5A CSP Pump A Min Flow Isolation GA MO FM-23A E-7 2 O/C B A PI Y2 Valve C STC Q STO Q 14MOV-5B CSP Pump B Min Flow Isolation GA MO FM-23A E-3 2 O/C B A PI Y2 Valve C STC Q STO Q 14MOV-7A CSP Pump A Suction From GA MO FM-23A C-6 2 O/C B A PI Y2 Suppression Pool Isolation Valve O STC Q STO Q 14MOV-7B CSP Pump B Suction From GA MO FM-23A C-4 2 O/C B A PI Y2 Suppression Pool Isolation Valve O STC Q STO Q 14SV-20A Core Spray Pump A Discharge RV SA FM-23A E-8 2 C C A RT Y10 Y8 group of two Safety Valve C 14SV-20B Core Spray Pump B Discharge RV SA FM-23A E-2 2 C C A RT Y10 Y8 group of two Safety Valve C 15AOV-130A RBCLC to Drywell Cooling GL AO FM-15B C-7 2A C A A LTJ AJ Augmented Assembly A Isolation O PI Y2 STC CS CSJ-02 15AOV-130B RBCLC to Drywell Cooling GL AO FM-15B D-4 2A C A A LTJ AJ Augmented Assembly B Isolation O PI Y2 STC CS CSJ-02 15AOV-131A RBCLC Return from Drywell GL AO FM-15B E-7 2A C A A LTJ AJ Augmented Cooling Assembly A Isolation O PI Y2 STC CS CSJ-02 15AOV-131B RBCLC Return from Drywell GL AO FM-15B E-4 2A C A A LTJ AJ Augmented Cooling Assembly B Isolation O PI Y2 STC CS CSJ-02 15AOV-132A RBCLC Supply To Recric Pump & GL AO FM-15B F-4 2A C A A LTJ AJ Augmented Motor A Coolers Isolation O PI Y2 STC CS CSJ-02 15AOV-132B RBCLC Supply To Recric Pump & GL AO FM-15B F-7 2A C A A LTJ AJ Augmented Motor B Coolers Isolation O PI Y2 STC CS CSJ-02 15AOV-133A RBCLC Return From Recric GL AO FM-15B F-4 2A C A A LTJ AJ Augmented Pump & Motor A Coolers O PI Y2 Isolation STC CS CSJ-02 15AOV-133B RBCLC Return From Recric GL AO FM-15B F-7 2A C A A LTJ AJ Augmented Pump & Motor B Coolers O PI Y2 Isolation STC CS CSJ-02 15AOV-134A RBCLC supply to Drywell GL AO FM-15B C-6 2A C A A LTJ AJ Augmented Equipment Sump Cooler O PI Y2 Isolation STC CS CSJ-02 A15 - 15

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 15RBC-214 PASS Cooler Emergency Water CK SA FM-18C E-7 3 C C A No Indication Supply Check O/C 15RBC-61 RBCLC Supply to PASS Liquid CK SA FM-15B F-7 3A C C A CCF Q Augmented Sample Cooler O 15SOV-215 RBCLC Supply to PASS Liquid GL SO FM-18C E-7 3 C B P PI Y2 Sample Cooler C 16-1AOV-101A Drywell Leak Rate Testing Inner GA AO FM-49A D-7 2A C A A FC Q Reference Isolation Valve O LTJ AJ PI Y2 STC Q 16-1AOV-101B Drywell Leak Rate Testing Outer GA AO FM-49A E-7 2A C A A FC Q Reference Isolation Valve O LTJ AJ PI Y2 STC Q 16-1AOV-102A Torus Leak Rate Testing Outer GA AO FM-49A D-7 2A C A A FC Q Reference Isolation Valve O LTJ AJ PI Y2 STC Q 16-1AOV-102B Torus Leak Rate Testing Inner GA AO FM-49A C-7 2A C A A FC Q Reference Isolation Valve O LTJ AJ PI Y2 STC Q 20AOV-83 Drywell Floor Drain Pumps A&B BAL AO FM-17A F-6 2A C A A FC Q Discharge Outboard Isolation O/C LTJ RR Valve PI Y2 STC Q 20AOV-95 Drywell Equipment Drain Pump BAL AO FM-17A C-6 2A C A A FC Q Discharge Inboard Isolation O/C LTJ RR Valve PI Y2 STC Q 20MOV-82 Drywell Floor Drain Sump Pump GA MO FM-17A F-7 2A C A A LTJ AJ Discharge Inboard Isolation O PI Y2 Valve STC RR ROJ-34 20MOV-94 RDW Drywell Equipment Drain GA MO FM-17A C-6 2A C A A LTJ AJ Sump Pump Inboard Isol Valve O PI Y2 STC RR ROJ-34 20RD-18 Penetration X-18 Overpressure RPD SA FM-17A H-7 2A C D A DT Y5 Protection Rupture Disc C 23AOV-42 SPCI Turbine Steam Supply GA AO FM-25A G-2 2 C B A FC Q Upstream Drain Isolation Valve O PI Y2 STC Q 23EFV-01A HPCI Turbine Steam Supply Leak CK SA FM-25A G-6 1 C A/C A SC RR VRR-02 Detection X-51C Excess Flow O LT SO VRR-02 Check Valve A15 - 16

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 23EFV-01B HPCI Turbine Steam Supply Leak CK SA FM-25A G-7 1 C A/C A SC RR VRR-02 Detection X-51D Excess Flow O LT SO VRR-02 Check Valve 23EFV-02A HPCI Turbine Steam Supply Leak CK SA FM-25A G-7 1 C A/C A SC RR VRR-02 Detection X-27C Excess Flow O LT SO VRR-02 Check Valve 23EFV-02B HPCI Turbine Steam Supply Leak CK SA FM-25A G-7 1 C A/C A SC RR VRR-02 Detection X-27D Excess Flow O LT SO VRR-02 Check Valve 23HPI-12 HPCI Turbine Exhaust Check CK SA FM-25A C-6 2 O/C A/C A DIS RR Valve C CCF RR ROJ-12 CVCM JAF-08 COF Q LTJ AJ 23HPI-13 HPCI Drain Pot Drain to Torus CK SA, FM-25A C-7 2 O/C C A LTJ AJ Stop Check Valve MA C CCF RR ROJ-26 COF Q ROJ-13 23HPI-130 HPCI Gland Seal Cooling Return CK SA FM-25A C-5 2 O/C C A DIS RR ROJ-17 Check Valve C 23HPI-131 HPCI Condensate Pump P-141 CK SA FM-25A C-5 2 C C A DIS RR ROJ-18 CVCM JAF-20A Discharge Check Valve C 23HPI-18 HPCI Pump Discharge to CK SA FM-25A F-7 1 O C A SC CS CSJ-03 CVCM JAF-20B Reactor Check Valve C 23HPI-32 HPCI Booster Pump P-1B CK SA FM-25A G-5 2 C C A DIS RR ROJ-14 CVCM JAF-21A Suction From CST 33TK-12A and C B Check Valve 23HPI-402 HPCI Exhaust Vacuum Breaker CK SA FM-25A E-7 2A O/C C A CCF RR ROJ-23 Check Valve C COF RR ROJ-23 23HPI-403 HPCI Exhaust Vacuum Breaker CK SA FM-25A E-7 2A O/C C A CCF RR ROJ-23 Check Valve C COF RR ROJ-23 23HPI-56 HPCI Pot Drain to Torus Check CK SA FM-25A C-6 2 O C A DIS RR ROJ-13 Valve C 23HPI-61 HPCI Booster Pump P-1B CK SA FM-25A B-7 2 O C A DIS RR ROJ-15 CVCM JAF-21B suction from Suppression Pool C Check Valve 23HPI-62 HPCI Mi Flow Line to HRH Check CK SA FM-25A F-4 2 O C A DIS RR ROJ-16 CVCM JAF-22 Valve C 23HPI-65 HPCI Turbine Exhaust To Torus CK SA FM-25A C-6 2 O/C A/C A LTJ AJ Check Valve C CCF RR ROJ-12 CVCM JAF-08 COF Q 23MOV-14 HPCI Turbine Steam Supply GA MO FM-25A F-3 2 O B A PI Y2 Isolation Valve C STO Q 23MOV-15 HPCI Steam Supply Outboard GA MO FM-25A F-8 1 O/C A A LTJ AJ Isolation Valve C PI Y2 STC RR ROJ-31 STO RR ROJ-31 A15 - 17

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 23MOV-16 HPCI Steam Supply Outboard GA MO FM-25A F-7 1 O/C A A LTJ AJ Isolation Valve O PI Y2 STC Q STO Q 23MOV-17 HPCI Booster Pump P-1B GA MO FM-25A G-5 2 C B A PI Y2 Suction From CST 33TK-12 and O STC Q Isolation Valve STO Q 23MOV-19 HPCI Pump to Reactor Inboard GA MO FM-25A F-6 1 O/C A A LTJ AJ Isolation Valve C PI Y2 STC Q STO Q 23MOV-20 HPCI Pump to Reactor GA MO FM-25A F-6 2 O B A PI Y2 Outboard Isolation Valve O STO Q 23MOV-21 HPCII full Flow Test Return To GL MO FM-25A G-6 2 C B A PI Y2 CST 33TK-12A & B UPSTR Valve C STC Q 23MOV-25 HPCI Main Pump P-1M Min GL MO FM-25A F-5 2 O/C B A PI Y2 Flow Isolation Valve C STC Q STO Q 23MOV-57 HPCI Booster Pump P-1B GA MO FM-25A F-5 2 O B A PI Y2 Suction from Suppression Pool C STO Q DNSTR ISOL Valve 23MOV-58 HPCI Booster Pump P-1B GA MO FM-25A C-7 2 O/C B A PI Y2 Suction from Suppression Pool O STC Q UPSTR ISOL Valve STO Q 23MOV-59 HPCI Turbine Exhaust Line GA MO FM-25A E-7 2 O B P PI Y2 Vacuum Breaker Valve O 23MOV-60 HPCI Steam Supply Outboard GL MO FM-25A F-7 1 C A A LTJ AJ MOV-16 Bypass Valve O PI Y2 STC Q STO Q 23SV-34 HPCI Booster Pump P-1B RV SA FM-25A E-5 2 C C A RLF Y10 Y8 Group of two Suction Safety Valve C 23SV-66 HPCI Booster Pump P-1B Recirc RV SA FM-25A D-5 2 C C A RLF Y10 Y8 Group of two Safety Valve C 23Z-7 HPCI Turbine Exhaust Rupture RPD SA FM-25A F-3 2 C C A RT Y5 Disc C 23Z-8 HPCI Turbine Exhaust Rupture RPD SA FM-25A F-3 2A C C A RT Y5 Augmented Disc C 27AOV-101A Torus Vacuum Breaker VB-6 BF AO FM-18B C-6 2A O/C A A FC Q Isolation Valve C LTJ AJ PI Y2 STC Q STO Q A15 - 18

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 27AOV-101B Torus Vacuum Breaker VB-7 BF AO FM-18B C-6 2A O/C A A FC Q Isolation Valve C LTJ AJ PI Y2 STC Q STO Q 27AOV-111 Drywell Purge and Inert Supply BF AO FM-18B C-2 2A C A A FC Q Valve C LTJ AJ PI Y2 STC Q 27AOV-112 Drywell Purge and Inert Supply BF AO FM-18B C-3 2A C A A FC Q Valve C LTJ AJ PI Y2 STC Q 27AOV-113 Drywell Exhaust Inner Isolation BF AO FM-18B D-8 2A C A A FC Q Valve C LTJ AJ PI Y2 STC Q 27AOV-114 Drywell Exhaust Outer Isolation BF AO FM-18B D-8 2A C A A FC Q Valve C LTJ AJ PI Y2 STC Q 27AOV-115 Torus Purge and Inert Supply BF AO FM-18B C-2 2A C A A FC Q Valve C LTJ AJ PI Y2 STC Q 27AOV-116 Torus Purge and inert Isolation BF AO FM-18B C-3 2A O/C A A FC Q Valve C LTJ AJ PI Y2 STC Q 27AOV-117 Torus Exhaust Inner Isolation BF AO FM-18B B-8 2A C A A FC Q Valve C LTJ AJ PI Y2 STC Q 27AOV-118 Torus Exhaust Outer Isolation BF AO FM-18B B-8 2A C A A FC Q Valve C LTJ AJ PI Y2 STC Q 27AOV-126A Ambient Vaporizer A Inlet Valve GL AO FM-18A G-5 2A O A A FC Q Augmented O LTJ AJ PI Y2 STC Q 27AOV-126B Ambient Vaporizer B Inlet Valve GL AO FM-18A F-5 2A O A A FC Q Augmented C LTJ AJ PI Y2 STC Q A15 - 19

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 27AOV-128A CAD Train A Nitrogen Make-Up GL AO FM-18A G-4 2A O/C A A FC Q Augmented Supply Valve C PI Y2 STO Q 27AOV-128B CAD Train B Nitrogen Make-Up GL AO FM-18A E-4 2A O/C A A FC Q Augmented Supply Valve C PI Y2 STO Q 27AOV-129A Drywell PCV and Instrument GL AO FM-18A F-4 2A O/C A A FC Q Augmented CAD Train A Backup Valve O PI Y2 STC Q STO Q 27AOV-129B Drywell PCV and Instrument GL AO FM-18A F-4 2A O/C A A FC Q Augmented CAD Train B Backup Valve C PI Y2 STC Q STO Q 27AOV-131A CAD Train A Nitrogen Make-Up GL AO FM-18B C-4 2 O/C A A FC Q Isolation Valve C PI Y2 STC Q STO Q 27AOV-131B CAD Train B Nitrogen Make-Up GL AO FM-18B C-3 2 O/C A A FC Q Isolation Valve C LTJ AJ PI Y2 STC Q STO Q 27AOV-132A CAD Train A Torus Nitrogen GL AO FM-18B C-4 2 O/C A A FC Q Make-Up Isolation Valve C LTJ AJ PI Y2 STC Q STO Q 27AOV-132B CAD Train A Torus Nitrogen GL AO FM-18B C-3 2 O/C A A FC Q Make-Up Isolation Valve C LTJ AJ PI Y2 STC Q STO Q 27CAD-19A Liquid Nitrogen Tank A Outlet CK SA FM-18A G-6 2A O C A CCF Q, M1 Augmented Check Valve O/C COF Q 27CAD-19B Liquid Nitrogen Tank B Outlet CK SA FM-18A C-6 2A O C A CCF Q, M1 Augmented Check Valve O/C COF Q 27CAD-67 CAD Train A Torus Nitrogen CK SA FM-18B C-6 2 O/C A/C A LTJ AJ Make-Up Check Valve O CCF RR ROJ-32 COF Q 27CAD-68 CAD Train A Drywell Nitrogen CK SA FM-18B C-4 2 O/C A/C A LTJ AJ Make-Up Check Valve O CCF RR ROJ-32 COF Q A15 - 20

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 27CAD-69 CAD Train B Drywell Nitrogen CK SA FM-18B C-4 2 O/C A/C A LTJ AJ Make-Up Check Valve O CCF RR ROJ-32 COF Q 27CAD-70 CAD Train B Torus Nitrogen CK SA FM-18B C-3 2 O/C A/C A LTJ AJ Make-Up Check Valve O CCF RR ROJ-32 COF Q 27MOV-113 Drywell Exhaust Isolation BTF MO FM-18B C-8 2 O/C A A LTJ AJ Valves 27aOV-113 & 114 Outer C PI Y2 Bypass Valve STC Q STO Q 27MOV-117 Torus Exhaust Isolation Valves BF MO FM-18B B-8 2 O/C A A LTJ AJ 27AOV-117 & 118 Inner Bypass C PI Y2 Valve STC Q STO Q 27MOV-120 Containment Exhaust to BF MO FM-18B H-8 2 O B A PI Y2 Standby Gas Treatment C STC Q Isolation Valve STO Q 27MOV-121 Containment Exhaust to BF MO FM-18B H-8 2 O B A PI Y2 Standby Gas Treatment C STC Q Isolation Valve STO Q 27MOV-122 Drywell Exhaust Isolation Valves GL MO FM-18B C-8 2 O/C A A LTJ AJ 27AOV-113 & 114 Inboard C PI Y2 Bypass Valve STC Q STO Q 27MOV-123 Drywell Exhaust Isolation Valves GL MO FM-18B B-8 2 O/C A A LTJ AJ 27AOV-113 & 114 Outboard C PI Y2 Bypass Valve STC Q STO Q 27RD-1A CAD Train A Valve Operating RPD SA FM-18A F-7 2A C D A DT Y10 Augmented Supply Line Rupture Disc C 27RD-1B CAD Train B Valve Operating RPD SA FM-18A C-7 2A C D A DT Y10 Augmented Supply Line Rupture Disc C 27RD-2A Pressure Building Coil A Outlet RPD SA FM-18A F-6 2A C D A DT Y10 Augmented Rupture Disc C 27RD-2B Pressure Building Coil B Outlet RPD SA FM-18A C-6 2A C D A DT Y10 Augmented Rupture Disc C 27SOV-119E1 Containment Analyzer A Torus GL SO FM-18D C-7 2 C A A FC Q Sample Outer Isolation Valve C/O LTJ AJ PI Y2 STC Q 27SOV-119E2 Containment Analyzer A Torus GL SO FM-18D C-6 2 C A A FC Q Sample Inner Isolation Valve C/O LTJ AJ PI Y2 STC Q A15 - 21

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 27SOV-119F1 Containment Analyzer B Torus GL SO FM-18D D-4 2 C A A FC Q Sample Inner Isolation Valve C/O LTJ AJ PI Y2 STC Q 27SOV-119F2 Containment Analyzer B Torus GL SO FM-18D C-5 2 C A A FC Q Sample Outer Isolation Valve C/O LTJ AJ PI Y2 STC Q 27SOV-120E1 Containment Analyzer A Drywell GL SO FM-18D F-6 2 C A A FC Q 310EL Sample Outer Isolation C/O LTJ AJ Valve PI Y2 STC Q 27SOV-120E2 Containment Analyzer A Drywell GL SO FM-18D F-6 2 C A A FC Q 310EL Sample Inner Isolation C/O LTJ AJ Valve PI Y2 STC Q 27SOV-120F1 Containment Analyzer B Drywell GL SO FM-18D F-4 2 C A A FC Q 310EL Sample Outer Isolation C/O LTJ AJ Valve PI Y2 STC Q 27SOV-120F2 Containment Analyzer B Drywell GL SO FM-18D F-4 2 C A A FC Q 310EL Sample Inner Isolation C/O LTJ AJ Valve PI Y2 STC Q 27SOV-122E1 Containment Analyzer A Drywell GL SO FM-18D F-6 2 C A A FC Q 343EL Sample Outer Isolation C/O LTJ AJ Valve PI Y2 STC Q 27SOV-122E2 Containment Analyzer A Drywell GL SO FM-18D F-6 2 C A A FC Q 343EL Sample Inner Isolation C/O LTJ AJ Valve PI Y2 STC Q 27SOV-122F1 Containment Analyzer B Drywell GL SO FM-18D G-4 2 C A A FC Q 343EL Sample Outer Isolation C/O LTJ AJ Valve PI Y2 STC Q 27SOV-122F2 Containment Analyzer B Drywell GL SO FM-18D G-4 2 C A A FC Q 343EL Sample Inner Isolation C/O LTJ AJ Valve PI Y2 STC Q 27SOV-123E1 Containment Analyzer A Drywell GL SO FM-18D E-6 2 C A A FC Q Sample Outer Isolation Valve C/O LTJ AJ PI Y2 STC Q A15 - 22

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 27SOV-123E2 Containment Analyzer A Drywell GL SO FM-18D E-6 2 C A A FC Q Sample Inner Isolation Valve C/O LTJ AJ PI Y2 STC Q 27SOV-123F1 Containment Analyzer B Drywell GL SO FM-18D F-4 2 C A A FC Q 276EL Sample Outer Isolation C/O LTJ AJ Valve PI Y2 STC Q 27SOV-123F2 Containment Analyzer B Drywell GL SO FM-18D F-4 2 C A A FC Q 276EL Sample Inner Isolation C/O LTJ AJ Valve PI Y2 STC Q 27SOV-124E1 Containment Analyzer A Post GL SO FM-18D C-4 2 C A A FC Q Accident Sampling Return C/O LTJ AJ Header Outer Isolation Valve PI Y2 STC Q 27SOV-124E2 Containment Analyzer A Post GL SO FM-18D C-4 2 C A A FC Q Accident Sampling Return C/O LTJ AJ Header Inner Isolation Valve PI Y2 STC Q 27SOV-124F1 Containment Analyzer B Post GL SO FM-18D C-4 2 C A A FC Q Accident Sample Return Outer C/O LTJ AJ Isolation Valve PI Y2 STC Q 27SOV-124F2 Containment Analyzer B Post GL SO FM-18D C-4 2 C A A FC Q Accident Sample Return Inner C/O LTJ AJ Isolation Valve PI Y2 STC Q 27SOV-125A Drywell Rad Monitor 17-04-1 GL SO FM-18B F-5 2 C A A FC Q Sample Return Inner Isolation O LTJ AJ Valve PI Y2 STC Q 27SOV-125B Drywell Rad Monitor 17-04-2 GL SO FM-18B F-4 2 C A A FC Q Sample Return Inner Isolation O LTJ AJ Valve PI Y2 STC Q 27SOV-125C Drywell Rad Monitor 17-04-1 GL SO FM-18B F-5 2 C A A FC Q Sample Return Outer Isolation O LTJ AJ Valve PI Y2 STC Q 27SOV-125D Drywell Rad Monitor 17-04-02 GL SO FM-18B F-4 2 C A A FC Q Sample Return Outer Isolation O LTJ AJ Valve PI Y2 STC Q A15 - 23

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 27SOV-135A Drywell Rad Monitor 17-04-1 GL SO FM-18B E-5 2 C A A FC Q Sample Second Isolation Valve O LTJ AJ PI Y2 STC Q 27SOV-135B Drywell Rad Monitor 17-04-2 GL SO FM-18B F-5 2 C A A FC Q Sample Outer Isolation Valve O LTJ AJ PI Y2 STC Q 27SOV-135C Drywell Rad Monitor 17-04-1 GL SO FM-18B E-5 2 C A A FC Q Sample First Isolation Valve O LTJ AJ PI Y2 STC Q 27SOV-135D Drywell Rad Monitor 17-04-2 GL SO FM-18B F-5 2 C A A FC Q Sample Inner Isolation Valve O LTJ AJ PI Y2 STC Q 27SOV-141 Drywell PCV and Instrument Air GL SO FM-39C E-6 2 O/C A A FC Q Or Normal N2 Cross-Tie Valve O LTJ AJ PI Y2 STC Q STO Q 27SOV-145 DW instrument Nitrogen GL SO FM-39C G-5 2 O/C A A FC Q Backup Supply Isolation Valve O LTJ AJ PI Y2 STC Q STO Q 27SV-114A CAD Train A Valve Operating RV SA FM-18A G-6 2A C C A RT Y10 Augmented Supply Safety Valve C 27SV-114B CAD Train B Valve Operating RV SA FM-18A D-6 2A C C A RT Y10 Augmented Supply Safety Valve C 27SV-115A Ambient A Vaporizer Outlet RV SA FM-18A G-4 2A C C A RT Y10 Augmented Safety Valve C 27SV-115B Ambient B Vaporizer Outlet RV SA FM-18A E-4 2A C C A RT Y10 Augmented Safety Valve C 27SV-118A Liquid Nitrogen Tank A Outlet RV SA FM-18A G-6 2A C C A RT Y10 Augmented Safety Valve C 27SV-118B Liquid Nitrogen Tank B Outlet RV SA FM-18A C-6 2A C C A RT Y10 Augmented Safety Valve C 27SV-119A Pressure Building Coil A Inlet RV SA FM-18A F-7 2A C C A RT Y10 Augmented Safety Valve C 27SV-119B Pressure Building Coil B Inlet RV SA FM-18A C-7 2A C C A RT Y10 Augmented Safety Valve C 27SV-201A Drywell Instrument NW Normal RV SA FM-18A F-3 2A C C A RT Y10 Augmented Supply Safety Valve C A15 - 24

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 27SV-201B Drywell Instrument N2 Normal RV SA FM-18A F-3 2A C C A RT Y10 Augmented Supply Safety Valve C 27SV-202 Drywell Instrument N2 Backup RV SA FM-18A H-3 2A C C A RT Y10 Augmented Supply Safety Valve C 27VB-1 Torus Downcomer Vacuum RV SA FM-18B C-6 2 O/C A/C A LTJ AJ Breaker C LTL RR PI Y2 SORT Y2 27VB-2 Torus Downcomer Vacuum RV SA FM-18B C-6 2 O/C A/C A LTJ AJ Breaker C LTL RR PI Y2 SORT Y2 27VB-3 Torus Downcomer Vacuum RV SA FM-18B C-6 2 O/C A/C A LTJ AJ Breaker C LTL RR PI Y2 SORT Y2 27VB-4 Torus Downcomer Vacuum RV SA FM-18B C-6 2 O/C A/C A LTJ AJ Breaker C LTL RR PI Y2 SORT Y2 27VB-5 Torus Downcomer Vacuum RV SA FM-18B C-6 2 O/C A/C A LTJ AJ Breaker C LTL RR PI Y2 SORT Y2 27VB-6 Reactor Building To Torus RV SA FM-18B C-6 2 O/C A/C A LTJ AJ Vacuum Breaker C PI Y2 SORT Y2 27VB-7 Reactor Building To Torus RV SA FM-18B C-6 2 O/C A/C A LTJ AJ Vacuum Breaker C PI Y2 SORT Y2 29AOV-80A Main Steam Line A Inboard GL AO FM-29A E-5 1 C A A FC RO ROJ-19 Isolation O LTJ AJ PI Y2 STC CS CSJ-08 29AOV-80B Main Steam Line B Inboard GL AO FM-29A D-5 1 C A A FC RO ROJ-19 Isolation O LTJ AJ PI Y2 STC CS CSJ-08 29AOV-80C Main Steam Line C Inboard GL AO FM-29A D-5 1 C A A FC RO ROJ-19 Isolation O LTJ AJ PI Y2 STC CS CSJ-08 A15 - 25

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 29AOV-80 Main Steam Line D Inboard GL AO FM-29A D-5 1 C A A FC RO ROJ-19 Isolation O LTJ AJ PI Y2 STC CS CSJ-08 29AOV-86A Main Steam Line A Outboard GL AO FM-29A G-4 1 C A A FC RO CSJ-05 Isolation O LTJ AJ PI Y2 STC CS CSJ-08 29AOV-86B Main Steam Line B Outboard GL AO FM-29A F-4 1 C A A FC RO CSJ-05 Isolation O LTJ AJ PI Y2 STC CS CSJ-08 29AOV-86C Main Steam Line C Outboard GL AO FM-29A E-4 1 C A A FC RO CSJ-05 Isolation O LTJ AJ PI Y2 STC CS CSJ-08 29AOV-86D Main Steam Line D Outboard GL AO FM-29A D-4 1 C A A FC RO CSJ-05 Isolation O LTJ AJ PI Y2 STC CS CSJ-08 29EFV-30A Main Steam Line B Excess Flow CK SA FM-29A E-8 1 C A A/C SC RO VRR-02 O LT RO VRR-02 29EFV-30B Main Steam Line B Excess Flow CK SA FM-29A E-8 1 C A A/C SC RO VRR-02 O LT RO VRR-02 29EFV-30C Main Steam Line A Excess Flow CK SA FM-29A E-8 1 C A A/C SC RO VRR-02 O LT RO VRR-02 29EFV-30D Main Steam Line A Excess Flow CK SA FM-29A E-8 1 C A A/C SC RO VRR-02 O LT RO VRR-02 29EFV-34A Main Steam Line D Excess Flow CK SA FM-29A E-8 1 C A A/C SC RO VRR-02 O LT RO VRR-02 29EFV-34B Main Steam Line D Excess Flow CK SA FM-29A E-8 1 C A A/C SC RO VRR-02 O LT RO VRR-02 29EFV-34C Main Steam Line C Excess Flow CK SA FM-29A E-8 1 C A A/C SC RO VRR-02 O LT RO VRR-02 29EFV-34D Main Steam Line C Excess Flow CK SA FM-29A E-8 1 C A A/C SC RO VRR-02 O LT RO VRR-02 29EFV-53A Main Steam Line C Excess Flow CK SA FM-29A E-8 1 C A A/C SC RO VRR-02 O LT RO VRR-02 29EFV-53B Main Steam Line C Excess Flow CK SA FM-29A E-8 1 C A A/C SC RO VRR-02 O LT RO VRR-02 29EFV-53C Main Steam Line D Excess Flow CK SA FM-29A E-8 1 C A A/C SC RO VRR-02 O LT RO VRR-02 29EFV-53D Main Steam Line D Excess Flow CK SA FM-29A E-8 1 C A A/C SC RO VRR-02 O LT RO VRR-02 A15 - 26

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 29EFV-54A Main Steam Line A Excess Flow CK SA FM-29A E-8 1 C A A/C SC RO VRR-02 O LT RO VRR-02 29EFV-54B Main Steam Line A Excess Flow CK SA FM-29A E-8 1 C A A/C SC RO VRR-02 O LT RO VRR-02 29EFV-54C Main Steam Line B Excess Flow CK SA FM-29A E-8 1 C A A/C SC RO VRR-02 O LT RO VRR-02 29EFV-54D Main Steam Line B Excess Flow CK SA FM-29A E-8 1 C A A/C SC RO VRR-02 O LT RO VRR-02 29MOV-200A Main Steam Leakage Collection GL MO FM-29A C-3 2A O B A PI Y2 Augmented Line A Isolation C STO Q 29MOV-200B Main Steam Leakage Collection GL MO FM-29A B-3 2A O B A PI Y2 Augmented Line B Isolation C STO Q 29MOV-201A Main Steam Leakage Collection GL MO FM-29A C-3 2A O/C B A PI Y2 Augmented Line A Upstream Isolation C STC Q STO Q 29MOV-201B Main Steam Leakage Collection GL MO FM-29A B-3 2A O/C B A PI Y2 Augmented Line B Upstream Isolation C STC Q STO Q 29MOV-202A Main Steam Leakage Collection GL MO FM-29A C-3 2A O/C B A PI Y2 Augmented Line A Downstream Isolation C STC Q STO Q 29MOV-202B Main Steam Leakage Collection GL MO FM-29A B-3 2A O/C B A PI Y2 Augmented Line B Downstream Isolation C STC Q STO Q 29MOV-203A Main Steam MSIV Stem Leakoff GL MO FM-29A C-3 2A O B A PI Y2 Augmented Line B Isolation C STO Q CSJ-06 29MOV-203B Main Steam MSIV Stem Leakoff GL MO FM-29A H-3 2A O B A PI Y2 Augmented Line B Isolation C STO CS CSJ-06 29MOV-204A Main Steam Leakage Collection GL MO FM-29A C-3 2A C B A PI Y2 Augmented Line A Bypass Isolation O STO CS 29MOV-204B Main Steam Leakage Collection GL MO FM-29A B-3 2A C B A PI Y2 Augmented Line B Bypass Isolation O STC Q 29MOV-74 Main Steam Line Drain Inboard GL MO FM-29A C-6 1 C A A LTJ AJ Isolation C PI Y2 STC RR ROJ-35 29MOV-77 Main Steam Line Drain GL MO FM-29A C-5 1 C A A LTJ AJ Outboard Isolation C PI Y2 STC RR 34FWS-28A Feedwater Supply Line A CK SA FM-34A E-7 1 C A/C A LTJ Q Inboard Check O CCF RR ROJ-20 CCO RR ROJ-20 34FWS-28B Feedwater Supply Line B CK SA FM-34A E-7 1 C A/C A LTJ Q Inboard Check O CCF RR ROJ-20 CCO RR ROJ-20 A15 - 27

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 34NRV-111A Feedwater Supply Line A CK AO, FM-34A E-7 1 C A/C A LTJ Q Outboard Check SA O PI Y2 CCF RR ROJ-30 COF RR ROJ-30 34NRV-111B Feedwater Supply Line B CK AO, FM-34A E-7 1 C A/C A LTJ Q Outboard Check SA O PI Y2 CCF RR ROJ-30 CCO RR ROJ-30 39IAS-22 DW Instrument Air/N2 Supply CK SA FM-39C E-5 2 O/C A/C A COF RO ROJ-21 Check Valve O/C LTJ AJ 39IAS-29 Drywell Instrument N2 Supply CK SA FM-39C F-3 2 O/C A/C A COF RO ROJ-21 Check Valve O/C LTJ AJ 46(70)ESW-101 CR/RR Vent 70AHU-3A & 12A GA MA FB-35E G-6 3 O B A SO Y2 ESW Supply Isolation Valve C 46(70)ESW-102 CR/RR Vent 70AHU-3B & 12B GA MA FB-35E C-6 3 O B A SO Y2 ESW Supply Isolation Valve C 46(70)ESW-103 CR/RR Vent 70AHU-3A & 12A GA MA FB-35E F-6 3 O B A SO Y2 ESW Return Isolation Valve C 46(70)ESW-104 CR/RR Vent 70AHU-3B & 12B GA MA FB-35E C-6 3 O B A SO Y2 ESW Return Isolation Valve C 46(70)SWS-101 CR/RR Vent SW to System A CK SA FB-35E H-8 3 C C A CCF Q Service Water Supply Check O COF Q CVCM JAF-02 Valve 46(70)SWS-102 CR/RR Vent SW to System B CK SA FB-35E H-8 3 C C A CCF Q Service Water Supply Check O COF Q CVCM JAF-02 Valve 46(70)SWS-13 CR/RR Chiller 2a Condenser GL MA FB-35E H-4 3 C B A SC Y2 Service Water Supply Isol Valve LO 46(70)SWS-14 CR/RR Chiller 2B Condenser GL MA FB-35E E-4 3 C B A SC Y2 Service Water Supply Isol Valve LO 46ESW-1A Emergency Service Water Pump CK SA FM-46B E-7 3 O C A CCF Q A Discharge Check Valve O COF Q 46ESW-1B Emergency Service Water Pump CK SA FM-46B D-7 3 O C A CCF Q B Discharge Check Valve O COF Q 46ESW-7A CR/RR ESW Loop A Supply CK SA FM-46B E-5 3 O C A CCF Q CVCM JAF-13A Check Valve O COF Q 46ESW-7B CR/RR ESW Loop B Supply CK SA FM-46B E-5 3 O C A CCF Q CVCM JAF-13B Check Valve O COF Q 46ESW-9A ESW Loop A Supply to RB Check CK SA FM-46B E-4 3 O C A CCF Q CVCM JAF-14A Valve O COF Q 46ESW-9B ESW Loop B Supply to RB Check CK SA FM-46B D-4 3 O C A CCF Q CVCM JAF-14B Valve O COF Q 46MOV-101A Emergency Service Water Loop GA MO FM-46B E-6 3 O B A PI Y2 A Supply Header Isolation Valve C STO Q A15 - 28

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 46MOV-101B Emergency Service Water Loop GA MO FM-46B C-6 3 O B A PI Y2 B Supply Header Isolation Valve C STO Q 46MOV-102A Emergency Service Water Pump GA MO FM-46B E-6 3 C B A PI Y2 A Test Valve O STC Q 46MOV-102B Emergency Service Water Pump GA MO FM-46B D-6 3 C B A PI Y2 B Test Valve O STC Q 46RD-112A Emergency Diesel Generator A RPD SA FM-46B G-7 3 O D A DT Y5 Jacket Water Cooler ESW Outlet C Rupture Disc 46RD-112B Emergency Diesel Generator B RPD SA FM-46B F-6 3 O D A DT Y5 Jacket Water Cooler ESW Outlet C Rupture Disc 46RD-112C Emergency Diesel Generator C RPD SA FM-46B F-7 3 O D A DT Y5 Jacket Water Cooler ESW Outlet C Rupture Disc 46RD-112D Emergency Diesel Generator D RPD SA FM-46B G-6 3 O D A DT Y5 Jacket Water Cooler ESW Outlet C Rupture Disc 46SWS-60A East Crescent Unit Coolers SWS CK SA FB-10H C-5 3 C C A CCF Q CVCM JAF-15A Loop A Supply Check Valve O COF Q 46SWS-60B East Crescent Unit Coolers SWS CK SA FB-10H C-5 3 C C A CCF Q CVCM JAF-15B Loop B Supply Check Valve O COF Q 46SWS-67A East Electric Bay 67UC-16A CK SA FM-46A B-6 3 C C A CCF Q Service Water Supply Check O CCO Q CVCM JAF-11 Valve 46SWS-67B East Electric Bay 67UC-16B CK SA FM-46A B-7 3 C C A CCF Q Service Water Supply Check O CCO Q CVCM JAF-11 Valve 46SWS-68 West Cable Tunnel 67E-11 CK SA FM-46A B-6 3 C C A CCF Q Service Water Supply Check O CCO Q CVCM JAF-12 Valve 46SWS-69 East Cable Tunnel 67E-14 CK SA FM-46A B-8 3 C C A CCF Q Service Water Supply Check O CCO Q CVCM JAF-12 Valve 67PCV-101 Elec Bay and Cable Tunnel Vent GL AO FM-46A D-2 3 O B A FO Q Service Water Return Pressure O SO Q Control Valve 70TCV-120A Relay Room Vent AHU-12A 3W AO FB-35E F-7 3 O B A FO Q Chilled Water Outlet Temp O SO Q Control Valve 70TCV-120B Relay Room Vent AHU-12B 3W AO FB-35E C-6 3 O B A FO Q Chilled Water Outlet Temp O SO Q Control Valve A15 - 29

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant Valve ID Description Valve Actu P&ID Coord Class Positions Cat Active Test Freq Relief Justification Comments Type Type Safety Passive Request Normal 70TCV-121A Control Room Vent AHU-3A 3W AO FB-35E F-6 3 O B A FO Q Chilled Water Outlet O SO Q Temperature Control Valve 70TCV-121B Control Room Vent AHU-3A 3W AO FB-35E C-7 3 O B A FO Y2 Chilled Water Outlet O SO Temperature Control Valve 70WAC-12A CR/RR Chilled Water loop A Air GA MA FB-35E F-6 3 C B A SC Y2 Separator Tank 20A Inlet O Isolation Valve 70WAC-12B Control Room/Relay Room GA MA FB-35E C-6 3 C B A SC Y2 Chiller Air Separator Tank O Outlet 70WAC-5A CR/RR Chiller A Evaporator GA MA FB-35E F-2 3 C B A SC Y2 Chilled Water Outlet Isol Valve O 70WAC-5B CR/RR Chiller B Evaporator GA MA FB-35E D-2 3 C B A SC Y2 Chilled Water Outlet Isol Valve O A15 - 30

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant ATTACHMENT 16 CHECK VALVE CONDITION MONITORING PLAN INDEX

SEP-IST-007 IST Program Plan James A. FitzPatrick Nuclear Power Plant CVCM PLAN REV # TITLE NUMBER JAF-02 CMP Draft 10RHR-64A/B/C/D JAF-03 CMP Draft 10RHR-14A/B/C/D JAF-04 CMP Draft 11SLC-43A/B JAF-05 CMP Draft 13RCIC-4/5 JAF-06 CMP Draft 14AOV-13A/B JAF-07 CMP Draft 14CSP-10A/B JAF-08 CMP Draft 23HPI-12/65 JAF-10 CMP Draft 46(70)SWS-101/102 JAF-11 CMP Draft 46SWS-67A/B JAF-12 CMP Draft 46SWS-68/69 JAF-13A CMP Draft 46ESW-7A JAF-13B CMP Draft 46ESW-7B JAF-14 CMP Draft 46ESW-9A/B JAF-15A CMP Draft 46SWS-60A JAF-15B CMP Draft 46SWS-60B JAF-20A CMP Draft 23HPI-130 JAF-20B CMP Draft 23HPI-131 JAF-21A CMP Draft 23HPI-32 JAF-21B CMP Draft 23HPI-61 JAF-22 CMP Draft 23HPI-62 A16 - 1

JAFP-18-0078, Fifth Ten-Year Interval Inservice Testing Program Plan

Text

Subject:

Dear Sir or Madam:

Enclosure:

1.0 INTRODUCTION

1.0 INTRODUCTION

REFERENCES:

REFERENCES:

REFERENCE:

REFERENCES:

REFERENCES:

SUBJECT:

Dear Mr. Hanson:

Enclosures:

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

NRC Staff Evaluation

NRC Staff Evaluation

4.0 CONCLUSION

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

NRC Staff Evaluation

NRC Staff Evaluation

4.0 CONCLUSION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 CONCLUSION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

NRC Staff Evaluation

NRC Staff Evaluation

NRC Staff Evaluation

4.0 CONCLUSION

SUBJECT:

Dear Mr. Hanson:

REFERENCES:

REFERENCES:

REFERENCES:

REFERENCES:

REFERENCES:

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