TMI-14-069, Submittal of the Fifth Ten-Year Interval Inservice Testing Program

Submittal of the Fifth Ten-Year Interval Inservice Testing Program

ML14147A018
Person / Time
Site:	Three Mile Island
Issue date:	05/16/2014
From:	Jim Barstow Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TMI-14-069
Download: ML14147A018 (234)

Text

200 Exelon Way Adnow Kennett Square, PA 19348 AdENOW Exelon Generation, www.exeloncorp.com 10CFR50.55a TMI-14-069 May 16, 2014 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289

Subject:

Submittal of the Fifth Ten-Year Interval Inservice Testing Program In accordance with the ASME OM Code-2004 Edition, with Addenda through OMb-2006 (ISTA-3200(a)), attached for your information is a copy of the fifth ten-year interval Inservice Testing (IST) Program Plan for the Three Mile Island Nuclear Station, Unit 1. This copy of the program plan is being supplied for information only.

There are no regulatory commitments contained within this letter.

If you have any questions or require additional information, please contact Tom Loomis (610-765-5510).

Sincerely, James Barstow Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC

Attachment:

Three Mile Island Nuclear Station, Unit 1 Inservice Testing (IST) Program Plan Fifth Ten-Year Interval cc: Regional Administrator, Region I, USNRC USNRC Senior Resident Inspector, TMI Project Manager, USNRC

1. L47

ATTACHMENT Three Mile Island Nuclear Station, Unit 1 Inservice Testing (IST) Program Plan Fifth Ten-Year Interval

Exelon Nuclear Generation, LLC

.200 Exelon Way Kennett Square, PA 19348 Three Mile Island Unit #1 NRC Docket Number 50-289 Three Mile Island Unit #1 Route 441 South Middletown, Pennsylvania 17057-0480 Commercial Service Date:

September 2, 1974 Inservice Testing (IST) Program Program Plan Fifth Ten-Year Interval October 25, 2013 - September 22, 2024 Revision 0 October 25, 2013

Three Mile Island Unit #1 Inservice Testing Program Plan REVISION RECORD Effective Revision Description Sign & Date Date Prepared: Reviewed: Approved; Site IST Corporate Engr.

Engineer IST Programs

, A .1 Engineer Manager Revision 0:

10/25/2013 Program revised in its entirety for fifth 10-year interval to comply with ASME OM Code- Mark 2004, including OMa Code-2005 and OMb Masoero Ruff Code-2006 Addenda's. 10/2312013

'-1--f--

4. +

Revision 0 October 25, 2013 Three Mile Island Unit #1 Inservice Testing Program Plan 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 RAI Responses and SERs

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 Revision 0 October 25, 2013 Three Mile Island Unit #1 Inservice Testing Program Plan

1.0 INTRODUCTION

1.1 Purpose The purpose of this Inservice Testing (IST) Program Plan is to provide a summary description of the Three Mile Island Unit #1 IST Program in order to document its compliance with the requirements of 10 CFR 50.55a(f) for the 5th 10-year IST interval.

1.2 Scope This Inservice Testing Program Plan identifies all of the testing performed on the components included in the Three Mile Island Unit #1 Inservice Testing (IST) Program for the 5th ten-year IST interval, which will begin on October 25, 2013 and is scheduled to end on September 22, 2024. (Note: The Fourth Ten-Year Interval began on September 23, 2004, and was originally scheduled to conclude on September 22, 2014. The Fifth Ten-Year Interval scheduled end date maintains the original 10-year pattern of intervals in accordance with ISTA-3120(d).)

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 OMb-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 one 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 Three Mile Island Unit #1, 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 Revision 0 October 25, 2013 Three Mile Island Unit #1 Inservice Testing Program Plan 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 Station's IST Bases Document. In addition to a description of each component's safety function(s), the Bases Document identifies 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.

Three Mile Island Unit #1 is licensed with the Hot Standby condition 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 Standby condition, maintain the Hot Standby 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 Revision 0 October 25, 2013 Three Mile Island Unit #1 Inservice Testing Program Plan 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 P&ID's on which a depiction of the pump or valve may be located.

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. These condition monitoring plans are generated from information contained in the IST Program database - Inservice Testing Program Assistant (ISTPa-2003). Implementation and Maintenance of the Condition Monitoring Program is addressed in T&RM procedure ER-AA-321-1005, "Condition Monitoring for Inservice Testing of Check Valves.".

Revision 0 October 25, 2013 Three Mile Island Unit #1 Inservice Testing Program Plan 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.

Revision 0 October 25, 2013 Three Mile Island Unit #1 Inservice Testing Program Plan 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 OMb-2006.

1.4.3 Three Mile Island Unit #1Technical Specification 1.4.4 Exelon Corporation Administrative Procedure ER-AA-321, Administrative Requirements for Inservice Testing Revision 0 October 25, 2013 Three Mile Island Unit #1 Inservice Testing Program Plan 2.0 INSERVICE TESTING PLAN FOR PUMPS 2.1 Pump Inservice Testing Plan The Three Mile Island Unit #1 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 DescriDtion The pumps included in the Three Mile Island Unit #1 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 EPN The unique identification number for the pump, as designated on the System P&ID or Flow Diagram Test Group A or B, as defined in Reference 1.4.2 (or applicable Relief)

Safety Class The ASME Safety Class (i.e., 1, 2 or 3) of the pump. Non-ASME Safety Class pumps are designated "N/A".

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 Pump Driver The type of driver with which the pump is equipped:

A Air-motor D Diesel M Motor (electric)

T Turbine (steam)

Nominal Speed The normal rotational speed of the pump at IST reference conditions.

P&ID The Piping and Instrumentation Diagram or Flow Drawing on which the pump is shown Revision 0 October 25, 2013 Three Mile Island Unit #1 Inservice Testing Program Plan P&ID Coor. Coordinates on the P&ID or Flow Diagram where the pump is shown Test Type 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)

AP Differential Pressure P Discharge Pressure (positive displacement pumps)

Q Flow Rate, Vd Vibration (displacement)

Vv Vibration (velocity)

Test Freq. 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.

Tech. Pos. Provides the Technical Position identification number applicable to the pump or test.

Comments Any appropriate reference or explanatory information (e.g.,

technical positions, etc.)

Pump Name The descriptive name of the pump shown on the bottom line for each pump entry. [use PIMS, Passport, etc. names for consistency)

Revision 0 October 25, 2013 Three Mile Island Unit #1 Inservice Testing Program Plan 3.0 INSERVICE TESTING PLAN FOR VALVES 3.1 Valve Inservice Testing Plan The Three Mile Island Unit #1 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 Three Mile Island Unit #1 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 EPN The unique identification number for the valve, as designated on the System P&ID or Flow Diagram.

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

Non-ASME Safety Class valves are designated by

N/A".

Categroy The ASME Code category or categories of the valve as defined in Reference 1.4.2.

Size The nominal size of the valve in inches.

Valve 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 Revision 0 October 25, 2013 Three Mile Island Unit #1 Inservice Testing Program Plan Act. 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 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.

Valve Normal and Abbreviations used to identify the normal, fail, and Safety Positions safety-related positions for the valve. Abbreviations used are:

Al 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 P&ID The Piping and Instrumentation Diagram or Flow Drawing on which the valve is shown.

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

Test Type 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 Revision 0 October 25, 2013 Three Mile Island Unit #1 Inservice Testing Program Plan related closure test)

BDO Bidirectional Check Valve test (non-safety related open test) 002 Check Valve Exercise Test - Closed CO 2 Check Valve Exercise Test - Open CP 2 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 1A 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.

Revision 0 October 25, 2013 Three Mile Island Unit #1 Inservice Testing Program Plan Test 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 OP Operational Frequency Relief Request Identifies the number of the Relief Request applicable to the specified test.

Deferred Just. 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.

Tech. Pos. Provides the Technical Position identification number applicable to the pump or test.

Valve Name The descriptive name for the valve [use PIMS, Passport, etc. names for consistency].

Revision 0 October 25, 2013 Three Mile Island Unit #1 Inservice Testing Program Plan SECTION 4.0 ATTACHMENTS Revision 0 October 25, 2013 Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT I SYSTEM AND P&ID LISTING Revision 0 October 25, 2013 Al - 1

Three Mile Island Unit #1 Inservice Testing Program Plan System System Name P&ID AH Reactor Building Purge 302-831, 847 BS Reactor Building Spray 302-712 CA Reactor Plant Chemical Addition 302-671 CF Core Flood 302-711 CH Control Building Chilled Water 302-847 CM Containment Monitoring 302-721 CO Condensate 302-101 DC Decay Heat Closed Cycle Cooling Water 302-645 DF Emergency Diesel Generators Fuel Systems 302-283, 351 DH Decay Heat Removal 302-640 DR Decay Heat River Water 302-202, 645 EF/AS Emergency Feedwater/Auxiliary Steam 302-011, 082, 611 EG Emergency Diesel Generator Support Systems 302-351, 353, 354 FS Fire Service Water 302-231 FW Main Feedwater 302-081 HM Containment Hydrogen Monitoring 302-674 HP Hydrogen Purge Discharge 302-721 HR Post LOCA Hydrogen Recombiner 302-722 IA Instrument Air 302-271, 273 IC Intermediate Closed Cooling Water 302-620 MS Main Steam 302-011 MU Makeup & Purification 302-660, 661 NI Nuclear Plant Nitrogen Supply 302-720 NR Nuclear Services River Water 302-202 NS Nuclear Services Closed Cooling Water 302-610, 645 PP Penetration Pressurization 302-706 Reactor Building Emergency Cooling Water/Reactor RB/RR River 302-611 RC Reactor Coolant 302-650 SA Station Service Air 302-271 SF Spent Fuel Pool Cooling 302-630 WDG Gaseous Waste Disposal 302-694 WDL Liquid Waste Disposal 302-196, 690 Revision 0 October 25, 2013 Al -2

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 2 PUMP RELIEF REQUEST INDEX Revision 0 October 25, 2013 A2 - 1

Three Mile Island Unit #1 Inservice Testing Program Plan RELIEF RELIEF REQUEST TITLE APPROVAL DATE REQUEST NUMBER PR-01 Nuclear Services Closed Cooling Water Flow Rate 08/15/2013 Measurement During Group A Tests PR-02 Proposed Alternative to Utilize Code Case OMN- 08/15/2013 18 In Accordance with 10 CFR 50.55a(a)(3)(i).

Note: The ASME Code committee has approved Code Case OMN-18, "Alternate Testing Requirements for Pumps Tested Quarterly within

+/- 20% of Design Flow." However, this Code Case has not been approved for use in Regulatory Guide 1.192, Operation and Maintenance Code Case Acceptability, ASME OM Code, June 2003 and therefore, requires NRC relief.

Revision 0 October 25, 2013 A2 - 2

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 3 PUMP RELIEF REQUESTS Revision 0 October 25, 2013 A3- 1

Three Mile Island Unit #1 Inservice Testing Program Plan 10 CFR 50.55a Request Number: PR-01 Revision 0 Relief Request Concerning Nuclear Services Closed Cooling Water Flow Rate Measurement During Group A Tests In Accordance with 10 CFR 50.55a(f)(5)(iii)

1. ASME Code Component(s) Affected NS-P-1A, Nuclear Services Closed Cooling Water (NSCCW) Pump 1A, (Centrifugal / Group A

/ Class 3)

NS-P-1 B, Nuclear Services Closed Cooling Water Pump 1 B, (Centrifugal / Group A / Class 3)

NS-P-1 C, Nuclear Services Closed Cooling Water Pump 1C, (Centrifugal I Group A / Class 3)

Component/System Function The NSCCW system includes four 33.33 percent capacity nuclear services coolers, and three 50-percent capacity NSCCW pumps. This system, along with the intermediate cooling system, satisfies the cooling requirements of all nuclear-oriented services other than decay heat and reactor building emergency cooling. In the event of a loss-of-coolant accident, 100-percent redundancy of all nuclear services equipment may be obtained by isolating non-essential items so that flow requirements are reduced to approximately half that of normal operation.

2. Applicable Code Edition and Addenda ASME OM Code-2004 Edition, with Addenda through OMb-2006

3. Applicable Code Requirement(s)

ISTB-5121 (b) - "Group A Test Procedure" - "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..."

4. Impracticality of Compliance Pursuant to 10 CFR 50.55a(f)(5)(iii), relief is requested from the requirement of ASME OM Code ISTB-5121(b). Due to system design and plant operating requirements, it is not practical to reduce the number of pumps in service to one to allow for single-pump testing during power operation. Also, individual pump flow rates cannot be measured during the Group A test. The flow instrumentation for this system is located in the common discharge header for all three of the subject pumps. The piping configuration does not contain, nor would the system design permit the installation of accurate individual pump flow measuring devices due to the turbulence caused by the valving and elbow configuration on the discharge of the pumps. There were no provisions originally designed in the system to measure individual pump flow rate.

Revision 0 October 25, 2013 A3 - 2

Three Mile Island Unit #1 Inservice Testing Program Plan

5. Burden Caused by Compliance Individual pump flow cannot be measured during normal quarterly operations since individual flow instrumentation does not exist. Also, two pumps are normally required to be in service to provide adequate cooling for system components.

To comply with the ISTB requirement for measuring individual pump flow rates on a quarterly basis, a modification of the system would be required.

6. Proposed Alternative and Basis for Use Individual suction and discharge pressure gauges are installed at each pump allowing for measurement of differential pressure for inservice testing. A flow instrument is installed in the common discharge header.

The proposed test would test (as approved previously in PR-02 for the fourth interval)

NSCCW pumps in pump pairs. As stated previously, individual pump flow cannot be measured during quarterly operations since individual flow instrumentation does not exist.

Also, two (2) pumps are normally required to be inservice to provide adequate cooling for system components. The NSCCW pumps are centrifugal pumps (not vertical line shaft). The current quarterly inservice procedure tests all combinations of paired-pumps (A-B, B-C, A-C).

During these tests, pump dP [differential pressure] is set and combined pump flow rate is measured.

Individual pump flow rates will be calculated and compared against individual pump flow rate reference values. Corrective actions will be taken in accordance with ISTB-6200, "Corrective Action," in the event that these criteria are not met. The pumps will continue to be tested individually in accordance with ISTB-5123, "Comprehensive Test Procedure," during refueling outages.

Additionally, vibration data on the pump will be recorded and compared to the reference values. Any deviation from the reference value will be compared to the Code acceptance criteria. The current testing methodology of testing paired-combinations of pumps near two-pump design flow rate provides an adequate basis for identifying and evaluating degraded pump performance. Therefore, this testing method provides reasonable assurance of pump operational readiness.

Summary of proposed alternative testing:

a. TMI, Unit 1 will continue to perform quarterly testing using a modified Group A test procedure as described above. With two paired-pumps in service, the required group A test parameters will be measured except for individual pump flow rate. Individual pump flow rates will be calculated and compared against calculated individual pump flow rate reference values. During this test the differential pressure for each pump will be throttled to the reference value.

b. Vibration measurements will be recorded and compared to their reference values.

Deviations from the reference value will be compared with the ranges specified in Table ISTB-5121-1 for Group A tests. Corrective actions will be taken in accordance with ISTB-6200.

c. During testing of the subject pumps (quarterly and refueling), TMI, Unit 1 will perform full spectrum vibration analysis, which is above Code required vibration testing.

d. The pumps will continue to be tested individually in accordance with ISTB-5123, "Comprehensive Test Procedure," during refueling outages.

Revision 0 October 25, 2013 A3 - 3

Three Mile Island Unit #1 Inservice Testing Program Plan Using the provisions of this relief request as an alternative to the specific requirements of ISTB-5121(b) will provide reasonable assurance of pump operational readiness. Therefore, pursuant to 10 CFR 50.55a(f)(5)(iii), TMI, Unit 1 requests relief from the specific ISTB requirements identified in this request.

7. Duration of Proposed Alternative The proposed alternative identified will be utilized during the fifth IST interval which is scheduled to begin October 15, 2013'and conclude on October 14, 2023.

8. Precedents A similar Relief Request (P5) was approved for TMI, Unit 1 for the third 10-year interval as documented in the US Nuclear Regulatory Commission's Safety Evaluation Report dated July 2, 2004 (ML041670196).

A similar Relief Request (PR-02) was approved for TMI, Unit 1 for the fourth 10-year interval, as documented in the U.S. Nuclear Regulatory Commission's Safety Evaluation Report dated July 7, 2005 (ML051530406).

Revision 0 October 25, 2013 A3 - 4

Three Mile Island Unit #1 Inservice Testing Program Plan 10 CFR 50.55a Request Number: PR-02 Revision 0 Proposed Alternative to Utilize Code Case OMN-18 In Accordance with 10 CFR 50.55a(a)(3)(i)

1. ASME Code Component(s) Affected AH-P-3A&B, Control Building Chilled Water Supply Pumps (Centrifugal / Group A / Class 3)

BS-P-1A&B, Reactor Building Spray Pumps (Centrifugal / Group AB / Class 2)

DC-P-1A&B, Decay Heat Closed Cooling Water Pumps (Centrifugal / Group A / Class 3)

DR-P-1iA&B, Decay Heat River Water Pumps (Vertical Line Shaft / Group A / Class 3)

SF-P-1A&B, Spent Fuel Cooling Pumps (Centrifugal / Group A / Class 3)

Component/System Function Provide minimum flow to meet system requirements under accident conditions

2. Applicable Code Edition and Addenda ASME OM Code-2004 Edition, with Addenda through OMb-2006

3. Applicable Code Requirement(s)

" ISTB-3300, "Reference Values," states, in part, that "Reference values shall be established within +/-20 percent of pump design flow rate for the comprehensive test," and "Reference values shall be established within +/-20 percent of pump design flow for the Group A and Group B tests, if practicable."

" ISTB-3400, "Frequency of Inservice Tests", states that an inservice test shall be run on each pump as specified in Table ISTB-3400-1.

• Table ISTB-3400-1 requires Group A and Group B tests to be performed quarterly and a comprehensive test to be performed biennially.

• Table ISTB-3510-1, "Required Instrument Accuracy," specifies the instrument accuracies for Group A, Group B, comprehensive, and preservice tests.

• Table ISTB-5121-1 "Centrifugal Pump Test Acceptance Criteria" defines the required acceptance criteria for Group A, Group B, and Comprehensive tests for centrifugal pumps.

• Table ISTB-5221-1 "Vertical Line Shaft Centrifugal Pumps Test Acceptance Criteria" defines the required acceptance criteria for Group A, Group B, and Comprehensive tests for Vertical Line Shaft centrifugal pumps.

4. Reason for Request The ASME Code committees have approved Code Case OMN-18, "Alternate Testing Requirements for Pumps Tested Quarterly within +/- 20% of Design Flow." This Code Case Revision 0 October 25, 2013 A3 - 5

Three Mile Island Unit #1 Inservice Testing Program Plan has not been approved for use in Regulatory Guide 1.192, "Operation and Maintenance Code Case Acceptability, ASME OM Code," June 2003.

This Code Case allows the Owner to not perform the Comprehensive Pump Test (CPT) with the associated acceptance criteria, if the quarterly test is performed at +/- 20% of design flow and the instrumentation meets the accuracy requirements of Table ISTB-3510-1 for the comprehensive and preservice tests. The basis for the testing strategy in this Code Case is that a quarterly Group A pump test, performed at the CPT flow rate with more accurate instrumentation, is more effective in assessing a pump's operational readiness, through trending, than a standard Group A test in conjunction with a biennial CPT.

Additionally, ISTB allows the Owner to categorize the pumps in their program. As such, an Owner could categorize a pump that otherwise meets the requirements of Group B, as a Group A (or AB) pump, and test according to the provisions of Code Case OMN-18. In doing this, they are obtaining additional data (vibration and flow or differential pressure) quarterly, rather than once every two years.

As a result of the increased requirements on the parameters imposed by the proposed alternative during applicable quarterly tests, there is no added value in performing the biennial comprehensive test on the subject pumps.

5. Proposed Alternative and Basis for Use TMI, Unit 1 is proposing to utilize the provisions of Code Case OMN-18 and performing a modified Group A test in lieu of performing the Code-required CPT. The modified Group A test will be run at +/-20% of the pump's design flow rate using +/-1/2%accurate gauges to determine the pump differential pressure. Vibration tests will be performed and the vibration acceptance criteria for the proposed alternative test will remain identical to the standard Group A test. Additionally, TMI, Unit 1 will utilize an Acceptable Range High limit of 106% or lower for quarterly testing, which is also consistent with the planned Code change applicable to CPT.

The use of more accurate pressure gauges and a more limiting Acceptable Range during every modified quarterly Group A test compensates for the elimination of the CPT (with its more limiting Acceptable Range upper bound for differential pressure or flow of 103%).

Regular testing with more accurate instrumentation and tighter acceptance criteria will provide for better trending of pump performance. Instead of performing seven tests with pressure instruments with +/-2% accuracy and then performing the eighth test with pressure instruments with +/-1/2% accuracy, all eight tests will be performed with the same +/-1/2% accurate instruments. Due to the improved accuracy, consistent testing methodology, and the addition of quarterly vibration monitoring on Group AB pumps, deviations in actual pump performance indicative of impending degradation are more easily recognized during quarterly performance trending activities.

Using the provisions of this request as an alternative to the requirements of ISTB-3400 and Tables ISTB-3400-1, ISTB-5121-1, & ISTB-5221-1 provides a reasonable alternative to the Code requirements based on the determination that the proposed alternative will provide adequate indication of pump performance, permit detection of component degradation, and continue to provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), TMI, Unit 1 requests approval of this alternative to the specific ISTB requirements identified in this request.

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Three Mile Island Unit #1 Inservice Testing Program Plan

6. Duration of Proposed Alternative The proposed alternative identified will be utilized during the fifth IST interval which is scheduled to begin October 15, 2013 and conclude on October 14, 2023.

7. Precedents A similar Relief Request (PR-01) was approved for the Oyster Creek Nuclear Generating Station as discussed in the U.S. Nuclear Regulatory Commission's Safety Evaluation Report dated June 21, 2012 (ML120050329).

A similar Relief Request (PR-9) was approved for the St. Lucie, Units 1 and 2 as discussed in the U.S. Nuclear Regulatory Commission's Safety Evaluation Report dated July 1, 2011 (ML11143A077).

A similar Relief Request (PR-3) was approved for the Perry Nuclear Power Plant, Unit 1, as discussed in the U.S. Nuclear Regulatory Commission's Safety Evaluation Report dated October 8, 2009 (ML092640690).

Revision 0 October 25, 2013 A3 - 7

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 4 VALVE RELIEF REQUEST INDEX Revision 0 October 25, 2013 A4- 1

Three Mile Island Unit #1 Inservice Testing Program Plan RELIEF RELIEF REQUEST TITLE APPROVAL DATE REQUEST NUMBER VR-01 Proposed Alternative Concerning Testing of 8/28/2013 the Pressurizer Pilot operated Relief Valve (PORV) in Accordance with 10 CFR 50.55a(a)(3)(i)

VR-02 Proposed Alternative Concerning ASME OM 8/15/2013 Code Test Frequencies In Accordance with 10 CFR 50.55a(a)(3)(ii)

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 5 VALVE RELIEF REQUESTS Revision 0 October 25, 2013 A5- 1

Three Mile Island Unit #1 Inservice Testing Program Plan 10 CFR 50.55a Request Number: VR-01 Revision 0 Proposed Alternative Concerning Testing of the Pressurizer Pilot operated Relief Valve (PORV) in Accordance with 10 CFR 50.55a(a)(3)(i)

1. ASME Code Component(s) Affected 1-RC-RV-2, Pressurizer Pilot Operated Relief Valve (PORV), Class 1, Category B/C Component/System Function The PORV is used to control Reactor Coolant System (RCS) pressure as discussed in Three Mile Island Generating Station (TMI), Unit 1 Updated Final Safety Analysis Report, Section 4.2.4 ("Pressure Control and Protection").

2. Applicable Code Edition and Addenda ASME OM Code-2004 Edition, with Addenda through OMb-2006

3. Applicable Code Requirement(s)

• ISTC-3310, Effects of Valve Repair, Replacement, or Maintenance on Reference Values"

- This section requires post-maintenance valve exercising following replacement.

• ISTC-3510, "Exercising Test Frequency" - "Power operated relief valves shall be exercise tested once per fuel cycle."

4. Reason for Request Pursuant to 10 CFR 50.55a(a)(3)(i), approval of a proposed alternative is requested to the above listed requirements of the OM Code. During the current fourth IST interval, the OM Code required testing for the PORV is satisfied by manually stroking the valve once every operating cycle. This is performed during plant startup following a refueling outage. The valve must be stroke timed during this exercise test. TMI, Unit 1 is proposing an alternative to this required in-situ testing for several reasons.

There are several disadvantages to the in-situ testing of the PORV. The PORV is a 2.5 inch Dresser Electromatic, solenoid actuated, pilot operated relief valve. Operation of the pilot valve vents the chamber under the main valve disc which causes the main valve to open. The PORV requires steam pressure for the main disc to open. Stroke testing the PORV during cold shutdown conditions would not exercise the main valve disc, which would not satisfy the ASME OM Code requirements. To test the PORV in-place, the RCS must be pressurized to supply the necessary fluid (steam) pressure to open the main valve disc.

Also, since the PORV design does not provide direct obturator position indication, the valve disc position must be inferred from alternate indications (tailpipe AT, acoustic monitor, RCS pressure decrease, or quench tank pressure or level rise).

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Three Mile Island Unit #1 Inservice Testing Program Plan In-situ testing of the PORV would also result in an in-surge of cooler water from the hot leg of the RCS into the pressurizer. The resulting thermal cycle on the pressurizer surge line would be a thermal stress concern, as described in NRC Bulletin 88-08 ("Thermal Stresses in Piping Connected to Reactor Coolant Systems") and should be avoided.

Requiring that the PORV be tested in-place prevents plant personnel from verifying proper reseating of the main valve disc because its discharge is not accessible as it is during bench testing. Minor leakage would not be readily evident before it would cause damage to the main valve disc/seat. Excessive leakage from the pilot valve can lead to inadvertent opening of the main valve, and impair its ability to re-close.

The proposed alternative will allow testing of the PORV that is appropriate to demonstrate functionality without cycling the valve in place using reactor steam pressure. This is consistent with NUREG-0737, "Clarification of TMI Action Plan Requirements," Item I1.K.3.16, "Reduction of Challenges and Failures of Relief Valves," which recommended that the number of relief valve openings be reduced as much as possible and that unnecessary challenges should be avoided.

5. Proposed Alternative and Basis for Use The following alternatives to the requirements stated above are proposed.

1) Bench testing of the PORV to satisfy valve exercise and stroke time requirements is performed at the vendor test facility prior to installation. Exercising of the valve at both the normal power operation set point and the Low Temperature Overpressure Protection (LTOP) set point (as provided in Technical Specification 3.1.12, "Pressurizer Power Operated Relief Valve (PORV), Block Valve, and Low Temperature Overpressure Protection (LTOP)") will be verified during this testing. Measured stroke time will be based on the pressure response indication of main disc opening.

2) The installed valve will be removed and replaced each refueling outage, with a spare valve that has been previously bench tested.

3) The removed valve will be bench tested within one year of removal from the system.

4) In-situ exercising of the PORV will be performed only as necessary to reestablish operational readiness after maintenance on an installed valve.

Requiring in-situ testing of the PORV unnecessarily increases the number of challenges to the PORV and results in an overall reduction in plant safety. Bench testing is preferable over in-situ testing for several reasons. Bench testing is performed at normal steam inlet conditions.

Both the power operation set point and the Low Temperature Overpressure Protection (LTOP) set point (as provided in Technical Specification 3.1.12, "Pressurizer Power Operated Relief Valve (PORV), Block Valve, and Low Temperature Overpressure Protection (LTOP)") are verified. The measurement of PORV stroke time, verification of main disc movement, and seat leakage check can be verified more precisely under bench testing conditions.

Performing a bench test under controlled conditions allows the opportunity for accomplishing repairs to the valve while there is sufficient time for any needed repairs or additional testing without these activities affecting the unit's operating/outage schedule and without the additional dose that could result by having to remove or perform work on the valve while it is attached to the RCS. If a test failure during plant startup were to occur that required cold Revision 0 October 25, 2013 A5 - 3

Three Mile Island Unit #1 Inservice Testing Program Plan shutdown conditions, this would result in an additional thermal cycle on the unit and additional personnel radiation exposure that could otherwise have been avoided.

Valve testing will be performed at a steam test facility, where the entire valve assembly (i.e.,

main valve, pilot valve, and solenoid) will be installed on a steam header in the same (vertical) orientation as the plant installation. The test conditions in the test facility will be similar to those in the plant installation, including ambient temperature, valve insulation, and steam conditions. The valve will then be leak tested, and functionally exercised at the normal power operation set point and the LTOP pressure limit to ensure the valve is capable of opening and closing (including stroke timing), and leak tested a final time. Valve seat tightness will be verified by a cold bar test, and if not free of fog, leakage will be measured and verified to be below design limits. If required, based on the results of "As-found" testing, the valve will be refurbished and recertified ("As-left" tested). The valve will then be shipped to the plant. If "As-found" testing is satisfactory and refurbishment is not necessary, the valve will be returned without any disassembly or alteration of the main valve or pilot valve components. A receipt inspection will be performed in accordance with the requirements of the Exelon Quality Assurance Program. The storage requirements in effect at TMI, Unit 1 ensure the valve is protected from moisture and physical damage. Prior to installation, the valve will again be inspected for foreign material and damage. The valve will be installed, and electrically connected in accordance with a TMI, Unit 1 maintenance procedure. The procedure ensures proper reinstallation of the PORV and proper connection of controls. The post-maintenance testing then verifies operation of the solenoid-actuated pilot valve.

The proposed alternative also provides adequate assurance that the valve stroke time will be consistently measured at the test facility. The test facility replicates the control signals to the pilot solenoid allowing it to be actuated at the pressures corresponding to when it would be actuated in the plant. Stroke timing of the solenoid-actuated pilot valve will start with the initiation of the signal to the solenoid and be based on the initial main valve obturator movement as determined by the change in steam pressure when exercised at both the normal reactor coolant system pressure limit and the lower LTOP pressure. Stroke timing will be used to ensure that the valve performs acceptably compared to its baseline and design requirements. Observation of the main disc movement at the test facility is indirect, based on evidence of steam pressure response, as it is in-situ at the plant, since here is no direct indication of the main valve position. Although this difference may result in minor differences in measured stroke time compared to those measured when installed in the plant, the stroke times measured at the test facility will be under conditions identical to when the reference values were established, making the results comparable and trendable for detecting any abnormality in valve performance. The IST reference values for the PORV stroke time are less than 2 seconds, and the stroke timing will be measured to at least tenths of a second even though the OM Code only requires it to be measured to the nearest second.

The following is a review of PORV test history from the steam test facility which demonstrates that the PORV pilot valve opens well within the 2-second limiting stroke time allowed by ASME OM Code Section ISTC-5 114(c) for rapid-acting valves:

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Three Mile Island Unit #1 Inservice Testing Program Plan RC-RV-2 First and Second LTOP First and Second Serial # Test Date Pressure Stroke times Normal RCS Pressure (seconds) Stroke times (seconds)

BS03989 11/04/2011 0.182, 0.196 0.095, 0.108 BS03989 11/11/2009 0.188, 0.232 0.128, 0.130 BS03989 10/28/2007 0.126, 0.142 0.130, 0.124 BL08905 10/22/2007 0.116, 0.176 0.132, 0.172 BL08905 08/10/2006 0.144, 0.146 0.116, 0.118 BS03839 05/07/2004 0.158, 0.156 0.094, 0.110 BL08905 05/24/2002 0.166, 0.160 0.110, 0.100 BS03989 08/31/2000 0.182, 0.176 0.084, 0.076 As demonstrated above, using the provisions of this relief request as an alternative to the listed ISTC requirements will provide: 1) assurance of PORV functional capability; 2) permit detection of component degradation, and; 3) continue to provide an acceptable level of quality and safety.

6. Duration of Proposed Alternative The proposed alternative identified will be utilized during the fifth IST interval which is scheduled to begin October 15, 2013 and conclude on October 14, 2023.

7. Precedents None Revision 0 October 25, 2013 A5 - 5

Three Mile Island Unit #1 Inservice Testing Program Plan 10 CFR 50.55a Request Number: VR-02 Revision 0 Proposed Alternative Concerning ASME OM Code Test Frequencies In Accordance with 10 CFR 50.55a(a)(3)(ii)

1. ASME Code Component(s) Affected All Pumps and Valves contained within the Inservice Testing Program (IST) scope.

2. Applicable Code Edition and Addenda ASME OM Code-2004 Edition, with Addenda through OMb-2006

3. Applicable Code Requirement(s)

This request applies to the frequency specifications of the ASME OM Code. The frequencies for tests given in the ASME OM Code do not include a tolerance band (grace period).

ISTA-3120(a) - "The frequency for the inservice testing shall be in accordance with the requirements of Section IST."

ISTB-3400 - "Frequency of Inservice Tests" - "An inservice test shall be run on each pump as specified in Table ISTB-3400-1 ." Table ISTB-3400-1 lists two frequencies -

quarterly and biennially.

ISTC-3510 - "Exercising Test Frequency" - "Active Category A, Category B, and Category C check valves shall be exercised nominally every 3 months ...."

ISTC-3540 - "Manual Valves" - "Manual Valves shall be full-stroke exercised at least once every 2 years, except where adverse conditions may require the valve to be tested more frequently to ensure operational readiness."

• ISTC-3630(a) - "Frequency" - "Tests shall be conducted at least once every 2 years."

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

ISTC-5221(c)(3) - "At least one valve from each group shall be disassembled and examined at each refueling outage; all valves in a group shall be disassembled and examined at least once every 8 years."

• Appendix 1, 1-1320 - "Test Frequencies, Class 1 Pressure Relief Valves" - "Class 1 pressure relief valves shall be tested at least once every 5 years..."

• Appendix 1, 1-1330 - "Test Frequency, Class 1 Nonreclosing Pressure Relief Devices" -

"Class 1 nonreclosing pressure relief devices shall be replaced every 5 years..."

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Three Mile Island Unit #1 Inservice Testing Program Plan

• Appendix 1,1-1340 - "Test Frequency, Class 1 Pressure Relief Valves that are used for Thermal Relief Application" - This section refers to 1-1320 for test frequency.

• Appendix 1,1-1350 - "Test Frequency, Classes 2 and 3 Pressure Relief Valves" -

"Classes 2 and 3 pressure relief valves, with the exception of PWR main steam safety valves, shall be tested every 10 years,

• Appendix 1,1-1360 - "Test Frequency, Classes 2 and 3 Nonreclosing Pressure Relief Devices" - "Classes 2 and 3 nonreclosing pressure relief devices shall be replaced every 5 years, ... "

• Appendix I, 1-1370 - "Test Frequency, Classes 2 and 3 Primary Containment Vacuum Relief Valves" - "Tests shall be performed on all Classes 2 and 3 containment vacuum relief valves at each refueling outage or every 2 years, ... "

• Appendix 1,1-1380 - "Test Frequency, Classes 2 and 3 Vacuum Relief Valves Except for Primary Containment Vacuum Relief Valves" - "All Classes 2 and 3 vacuum relief valves shall be tested every 2 years,

• Appendix 1,1-1390 - "Test Frequency, Classes 2 and 3 Pressure Relief Devices that are used for Thermal Relief Application" - "Tests shall be performed on all Classes 2 and 3 relief devices used in thermal relief application every 10 years.

• Appendix II, 11-4000(a)(1 )(e) - "Performance Improvement Activities" - Subparagraph (1)(e) requires the identification of the interval for each activity.

• Appendix II, 11-4000(b)(1)(e) - "Optimization of Condition Monitoring Activities" -

Subparagraph (1)(e) requires the identification of the interval for each activity.

4. Reason for Request Pursuant to 10 CFR 50.55a(a)(3)(ii), an alternative is requested from the frequency specifications of the ASME OM Code. The basis of the relief request is that the Code requirement presents an undue hardship without a compensating increase in the level of quality or safety.

ASME OM Code Section IST establishes the inservice test frequency for all components within the scope of the Code. The frequencies (e.g., quarterly) have always been interpreted as "nominal" frequencies (generally as defined in the Table 3.2 of NUREG 1482, Revision 1) and Owners routinely applied the surveillance extension time period (i.e., grace period) contained in the plant Technical Specifications (TS) Surveillance Requirements (SRs). The TS typically allow for a less than or equal to 25% extension of the surveillance test interval to accommodate plant conditions that may not be suitable for conducting the surveillance. However, regulatory issues have been raised concerning the applicability of the TS "Grace Period" to ASME OM Code required inservice test frequencies.

The lack of a tolerance band (grace period) on the ASME OM Code inservice test frequency restricts operational flexibility. There may be a conflict where a surveillance test could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be performed until sometime after a plant condition or associated Technical Specification is applicable. Therefore, to avoid this conflict, the surveillance test should be performed when it can be and should be performed.

The NRC recognized this potential issue in the TS by allowing a frequency tolerance as described in the TMI, Unit 1 TS SR 4.0.1. The lack of a similar tolerance applied to OM Code testing places an unusual hardship on the plant to adequately schedule work tasks without operational flexibility.

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Three Mile Island Unit #1 Inservice Testing Program Plan Thus, just as with TS required surveillance testing, some tolerance is needed to allow adjusting OM Code testing intervals to suit the plant conditions and other maintenance and testing activities. This assures operational flexibility when scheduling surveillance tests that minimize the conflicts between the need to complete the surveillance and plant conditions.

5. Proposed Alternative and Basis for Use ASME OM Code establishes component test frequencies that are based either on elapsed time periods (e.g., quarterly, 2 years, etc.) or on the occurrence of plant conditions or events (e.g., cold shutdown, refueling outage, upon detection of a sample failure, following maintenance, etc.).

a. Components whose test frequencies are based on elapsed time periods shall be tested at the frequencies specified in ASME OM Code Section IST with a specified time period between tests as shown in the table below and as defined in TMI, Unit 1 TS Section 1.25.

b. The specified time period between tests may be reduced or extended as follows:

1) For periods specified as less than 2 years, the period may be extended by up to 25% for any given test. This is consistent with TMI, Unit 1 TS Section 1.25, "FREQUENCY NOTATION."

2) For periods specified as greater than or equal to 2 years, the period may be extended by up to 6 months for any given test.

3) All periods specified may be reduced at the discretion of the owner (i.e., there is no minimum period requirement).

4) Period extensions may also be applied to accelerated test frequencies (e.g.,

pumps in Alert Range).

Specified Time Period Between Tests Frequency (all values are 'not to exceed'; no minimum periods are specified)

Quarterly 92 days (or every 3 months)

Semiannually 184 days (or every 6 months)

Annually 366 days (or every year) x Years x calendar years where 'x' is a whole number of years > 2

c. Components whose test frequencies are based on the occurrence of plant conditions or events (e.g., cold shutdown, refueling outage, upon detection of a sample failure, following maintenance, etc.) may not have their period between tests extended except as allowed by the ASME OM Code.

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Three Mile Island Unit #1 Inservice Testing Program Plan Period extension is to facilitate test scheduling and considers plant operating conditions that may not be suitable for performance of the required testing (e.g., performance of the test would cause an unacceptable increase in the plant risk profile due to transient conditions or other ongoing surveillance test or maintenance activities). Period extensions are not intended to be used repeatedly merely as an operational convenience to extend test intervals beyond those specified. This request is not applicable to frequencies in Subsection ISTD.

Using the provisions of this request as an alternative to the specific frequency requirements of the OM Code identified above will provide operational flexibility and still continue to provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii) TMI, Unit 1 requests approval of the alternative to the specific OM Code frequency requirements identified in this request.

6. Duration of Proposed Alternative The proposed alternative identified will be utilized during the fifth IST interval which is scheduled to begin October 15, 2013 and conclude on October 14, 2023.

7. Precedents A similar Relief Request was submitted to the U.S. Nuclear Regulatory Commission for Quad Cities Nuclear Power Station, Units 1 and 2 for the fifth 10-year interval, as documented in Letter RS-12-026 dated February 15, 2012.

8. References TMI, Unit 1 TS Section 1.25- "FREQUENCY NOTATION" TMI, Unit 1 TS Section 4.2 - "Reactor Coolant System Inservice and Testing" Revision 0 October 25, 2013 A5 - 9

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 6 RELIEF REQUEST RAI RESPONSES AND SERs Revision 0 October 25, 2013 A6 -1

Three Mile Island Unit #1 Inservice Testing Program Plan t'I:! . ] r , . .,* i l ". ;+

AIVMW Exe.on Oeneration.

10 CFR 50.55a TMI-13-043 March 15, 2013 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289 Subject Response to Request for Additional Information - Relief Request VR-01 Associated with the Fifth Inservice Testing (IST) Interval

References:

1) Letter from M. Jesse (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Submittal of Relief Request VR-01 Associated with the Fifth Inservice Testing (IST) Interval," dated October 18, 2012

2) Letter from P. Bamford (U.S. Nuclear Regulatory Commission) to M. Pacilio (Exelon Generation Company, LLC), 'Three Mile Island Nuclear Station, Unit 1 - Request for Additional Information Regarding Relief Request VR-01, Associated with the Fifth Inservice Testing Interval (TAC No. ME9819)," dated February 25, 2013 In the Reference I letter, Exelon Generation Company, LLC (Exelon) submitted for your review and approval Relief Request VR-01 associated with the Fifth Inservice Testing (IST) interval for Three Mile Island Nuclear Station (TMI), Unit 1. In the Reference 2 letter, the U.S. Nuclear Regulatory Commission Staff requested additional information. Attached is our response.

There are no regulatory commitments in this letter.

If you have any questions concerning this letter, please contact Tom Loomis at (610) 765-5510.

Respectfully,

• lihae/ D. Jesse /+ -/

Director - Licensir(g +& uelatory Affairs Exelon Generation' bompany, LLC Revision 0 October 25, 2013 A6 - 2

Three Mile Island Unit #1 Inservice Testing Program Plan Response to Request for Additional Information -

RR VR-01 Associated with the Fifth Inservice Testing (IST) Interval March 15, 2013 Page 2

Attachment:

Response to Request for Additional Information - Relief Request VR-01 cc: Regional Administrator, Region I, USNRC USNRC Senior Resident Inspector, TMI Project Manager, [TMI] USNRC Revision 0 October 25, 2013 A6 - 3

Three Mile Island Unit #1 Inservice Testing Program Plan Attachment Response to Request for Additional Information - Relief Request VR-01 Revision 0 October 25, 2013 A6 - 4

Three Mile Island Unit #1 Inservice Testing Program Plan Response to Request for Additional Information Relief Request VR-01 Page 1 of 2 Question:

1.) The PORV, 1-RC-RV-2, is categorized as a B/C valve in the referenced request. In addition to the requirements of the OM Code, ISTC-3310 and ISTC-3510 (for which relief is being requested), are any other ASME OM Code ISTC or Mandatory Appendix I test requirements applicable to this valve? Please provide any specific ASME OM Code test requirements for this valve type, but don't provide general test requirements from the OM Code.

Response

Exelon Generation Company, LLC (Exelon) has determined that the following specific ASME OM Code requirements also apply to the Three Mile Island Nuclear Station (TMI),

Unit 1 pressurizer Pilot Operated Relief Valve (PORV) and is not requesting relief from any of these:

• ISTC-51 10, "Power-Operated Relief Valves (PORVs)." - "Power-operated relief valves shall meet the requirements of ISTC-5100 for the specific Category B valve type and ISTC-5240 for Category C valves." This paragraph specifies the applicable testing for the type of power operator and that Mandatory Appendix I has additional test requirements invoked by ISTC-5240. These requirements are outlined in the following items.

• ISTC-51 11, "Valve Testing Requirements" - This requirement specifies the sequence of the individual tests - seat tightness / leakage testing, stroke testing, and position indication testing. It also requires calibration of the pressure-sensing device in accordance with the Owner's QA program. Position indication testing has been determined to not apply to the TMI, Unit l's PORV design.

• ISTC-5112, "Leak Testing" - This requirement specifies that seat tightness of the PORV shall be verified by leak testing in accordance with the requirements of Mandatory Appendix I. This paragraph directs leak testing using the test methodology and requirements for relief valves.

• ISTC-5113, "Valve Stroke Testing" - This requirement provides the Category B exercising requirements, which include stroke time measurement, establishment of limiting stroke time, and testing at normal operating conditions.

• 1-8110, "Steam Service" - This requirement specifies the detailed testing requirements for exercising the valve. Those that apply to TMI, Unit l's PORV testing are: test media, thermal equilibrium, and ambient temperature.

• 1-8200, "Seat Tightness Testing" - This requirement specifies that seat tightness testing shall be performed in accordance with the Owner's valve test procedure.

Consideration shall be given to test media, temperature stability, and ambient temperature, as required in 1-8100. Seat tightness testing shall be performed using Revision 0 October 25, 2013 A6 - 5

Three Mile Island Unit #1 Inservice Testing Program Plan Response to Request for Additional Information Relief Request VR-01 Page 2 of 2 the same fluid used for set-pressure testing, except as provided by 1-8300. This paragraph and associated subparagraphs (1-8210, 1-8220, and 1-8230) specify seat tightness testing using relief valve requirements (specified in 1-8100). The plant's PORV is not tested with alternative media.

The PORV testing described in Exelon's relief request complies with these requirements to ensure the operational readiness of the valve.

Question:

2.) Is the Pressurizer PORV block valve included in the TMI-1 1ST program? If not, please provide the basis for not including this valve in the program.

Response

The pressurizer PORV block valve is identified as 1-RC-V-2 in the TMI, Unit 1 IST Program Plan, and this valve is included in the scope of the program with both open and closed safety functions.

Revision 0 October 25, 2013 A6 -6

Three Mile Island Unit #1 Inservice Testing Program Plan Exeon Generat ion, .

10 CFR 50.55a TMI-13-068 April 17, 2013 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289 Subject Response to Request for Additional Information - Relief Request PR-01 Associated with the Fifth Inservice Testing (IST) Interval

References:

1) Letter from M. Jesse (Exelon Generation Company, LLC) to U.S.

Nuclear Regulatory Commission, "Submittal of Relief Requests Associated with the Fifth Inservice Testing (IST) Interval," dated November 7, 2012

2) Letter from P. Bamford (U.S. Nuclear Regulatory Commission) to M. Pacilio (Exelon Generation Company, LLC), 'Three Mile Island Nuclear Station, Unit 1 - Request for Additional Information Regarding Fifth Inservice Test Interval Relief Request PR-01 Nuclear Services Closed Cooling Water Flow Measurement (TAC No. MF0046)," dated March 19, 2013 In the Reference 1 letter, Exelon Generation Company, LLC (Exelon) submitted for your review and approval Relief Request PR-01 associated with the fifth Inservice Testing (IST) interval for Three Mile Island Nuclear Station (TMI), Unit 1. In the Reference 2 letter, the U.S. Nuclear Regulatory Commission Staff requested additional information. Attached is our response.

There are no regulatory commitments in this letter.

If you have any questions concerning this letter, please contact Tom Loomis at (610) 765-5510.

Respectfully, David P. Helker Manager - Licensing Exelon Generation Company, LLC Revision 0 October 25, 2013 A6 - 7

Three Mile Island Unit #1 Inservice Testing Program Plan Response to Request for Additional Information -

RR PR-01 Associated with the Fifth Inservice Testing (IST) Interval April 17, 2013 Page 2 Attachments: 1) Response to Request for Additional Information - Relief Request PR-01

2) Relief Request PR-01, Revision 2 cc: Regional Administrator, Region I, USNRC USNRC Senior Resident Inspector, TMI Project Manager, [TMI] USNRC Revision 0 October 25, 2013 A6 - 8

Three Mile Island Unit #1 Inservice Testing Program Plan Attachment 1 Response to Request for Additional Information - Relief Request PR-01 Revision 0 October 25, 2013 A6 - 9

Three Mile Island Unit #1 Inservice Testing Program Plan Response to Request for Additional Information Relief Request PR-01 Page 1 of 2 Question:

1.) In the attachment to letter dated November 7, 2012, Relief Request PR-01, Section 6, "Proposed Alternative and Basis for Use," it is stated that, "Individual pump flow rates will be calculated and compared against individual pump flow rate reference values." Please describe the methodology used to calculate the individual pump flow rates and how the reference values were established.

Response

The quarterly Group A testing of the three Nuclear Services Closed Cooling Water pumps is done during normal plant operation. The procedure operates the pumps in the three different combinations of pump pairs. For each test, the two operating pumps are throttled, using the discharge isolation valves for the individual pumps, to a test condition of 58 psid (pounds per square inch differential) pump head. The pump head (differential pressure) is determined from the local suction and discharge pressure gauge readings for each pump. These three flow readings are designated as AB, AC, and BC in the equations below and individual pump flows in gallons per minute (gpm) are calculated using the following equations:

Method NS-P-1A flow (AB + AC - BC)/2 = - gpm NS-P-1B flow (AB + BC - AC)/2 = __ gpm NS-P-1C flow (BC + AC - AB)/2 = __ gpm The Inservice Testing (IST) reference values were established using this test methodology. The following table summarizes the raw data and results from those tests:

Test Date Pump Pairs Discharge Suction Pump Total Test De u Pi Pressure Pressure Head Flow NS-P-1A 116 psig 58 psig 58 psi 6400 Test 1 NS-P-1 C 117 psig 59 psig 58 psi gpm March 5, NS-P-1A 116 psig 58 psig 58 psi 6300 2007 Test 2 NS-P-1B 117 psig 59 psig 58 psi gpm Test 3 NS-P-11B 117 psig 59 psig 58 psi 6000 NS-P-1C 117 psig 59 psig 58 psi gpm The individual pump flow rates were determined from this data as follows:

• NS-P-1A flow = (6300 + 6400 - 6000) /2 = 3350 gpm

• NS-P-1B flow = (6300 + 6000 - 6400) /2 = 2950 gpm

• NS-P-1Cflow= (6000+6400-6300)/2 = 3050gpm These individual pump flow rates were used as the initial reference values.

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Three Mile Island Unit #1 Inservice Testing Program Plan Response to Request for Additional Information Relief Request PR-01 Page 2 of 2 Subsequently, a new reference value for the NS-P-1A pump was established using the following data:

Discharge Suction Pump Total Test Date Pump Pairs Pressure Pressure Head Flow NS-P-1A 117 psig 59 psig 58 psi 5875 Test 1 NS-P-1C 116.5 psig 58.5 psig 58 psi gpm March14, NS-P-1A 117.5 psig 59.5 psig 58 psi 5875 200814, 2 B1Test 2008 NS-P-1 B 116.5 psig 58.5 psig 58 psi gpm NS-P-1 B 116 psig 58 psig 58 psi 5750 Test 3 NS-P-1C 116.5 psig 58.5 psig 58 psi gpm The individual pump flow rates were determined from this data as follows:

• NS-P-1 A flow = (5875 + 5875 - 5750) / 2 = 3000 gpm

• NS-P-1 B flow = (5875 + 5750 - 5875) / 2 = 2875 gpm

• NS-P-1Cflow= (5750+5875-5875)/2 = 2875gpm The NR-P-1 B and NR-P-1 C flows were slightly lower than the existing reference values, but within the ASME OM Code +/-10% acceptance criteria limit. Therefore, the reference values were not revised for these pumps.

Based on the described methodology, the flow reference values for each pump are as follows (except as noted):

• NS-P-1A flow = (5875 + 5875 - 5750) / 2 = 3000 gpm NS-P-1 B flow = (6300 + 6000 - 6400) / 2 = 2950 gpm (the current value for NS-P-1 B was later revised to 2949 gpm based on additional analysis)

• NS-P-1C flow = (6000 + 6400 - 6300) / 2 = 3050 gpm These reference values provide substantial margin over the minimum design flow requirement of 1972 gpm for each of the NS-P-1A/B/C pumps.

In addition to the above discussion, Relief Request PR-01 has been revised (see Attachment 2) to reflect a proposed alternative in accordance with 10 CFR 50.55a(a)(3)(i).

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Three Mile Island Unit #1 Inservice Testing Program Plan Attachment 2 Relief Request PR-01, Revision 2 Revision 0 October 25, 2013 A6 - 12

Three Mile Island Unit #1 Inservice Testing Program Plan 10 CFR 50.55a RELIEF REQUEST: PR-01 Revision 1 (Page 1 of 3)

Relief Request Concerning Nuclear Services Closed Cooling Water Flow Rate Measurement During Group A Tests In Accordance with 10 CFR 50.55a(a)(3)(i)

1. ASME Code Component(s) Affected NS-P-1A, Nuclear Services Closed Cooling Water (NSCCW) Pump lA (Centrifugal /

Group A / Class 3)

NS-P-I B, Nuclear Services Closed Cooling Water Pump lB (Centrifugal / Group A /

Class 3)

NS-P-IC, Nuclear Services Closed Cooling Water Pump IC (Centrifugal / Group A /

Class 3)

Component/System Function The NSCCW system includes four 33.33-percent capacity nuclear services coolers, and three 50-percent capacity NSCCW pumps. This system, along with the intermediate cooling system, satisfies the cooling requirements of all nuclear-oriented services other than decay heat and reactor building emergency cooling. In the event of a loss-of-coolant accident, 100-percent redundancy of all nuclear services equipment may be obtained by isolating nonessential items so that flow requirements are reduced to approximately half that of normal operation.

2. Applicable Code Edition and Addenda ASME OM Code-2004 Edition, with Addenda through OMb-2006

3. Applicable Code Requirement(s)

ISTB-512 I(b) - "Group A Test Procedure" - "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..."

4. Reason for Request Pursuant to 10 CFR 50.55a(a)(3)(i), TMI, Unit 1 requests approval of this alternative to the requirement of ASME OM Code ISTB-5121(b). Due to system design and plant operating requirements, it is not practical to reduce the number of pumps in service to one to allow for single-pump testing during power operation. Also, individual pump flow rates cannot be measured during the Group A test. The flow instrumentation for this system is located in the common discharge header for all three of the subject pumps. The piping configuration does not contain, nor would the system design permit the installation of accurate individual pump flow measuring devices due to the turbulence caused by the Revision 0 October 25, 2013 A6 - 13

Three Mile Island Unit #1 Inservice Testing Program Plan 10 CFR 50.55a RELIEF REQUEST: PR-01 Revision 1 (Page 2 of 3) valving and elbow configuration on the discharge of the pumps. There were no provisions originally designed in the system to measure individual pump flow rate.

To comply with the ISTB requirement for measuring individual pump flow rates on a quarterly basis, a modification of the system would be required.

5. Proposed Alternative and Basis for Use Individual suction and discharge pressure gauges are installed at each pump allowing for measurement of differential pressure for inservice testing. A flow instrument is installed in the common discharge header.

The proposed test would test (as approved previously in PR-02 for the fourth interval)

NSCCW pumps in pump pairs. As stated previously, individual pump flow cannot be measured during quarterly operations since individual flow instrumentation does not exist.

Also, two (2) pumps are normally required to be in service to provide adequate cooling for system components. The NSCCW pumps are centrifugal pumps (not vertical line shaft).

The current quarterly inservice procedure tests all combinations of paired-pumps (A-B, B-C, A-C). During these tests, pump dP (differential pressure) is set, and combined pump flow rate is measured.

Individual pump flow rates will be calculated and compared against individual pump flow rate reference values. Corrective actions will be taken in accordance with ISTB-6200, "Corrective Action," in the event that these criteria are not met. The pumps will continue to be tested individually in accordance with ISTB-5123, "Comprehensive Test Procedure,"

during refueling outages.

Additionally, vibration data on the pump will be recorded and compared to the reference values. Any deviation from the reference value will be compared to.the Code acceptance criteria. The current testing methodology of testing paired-combinations of pumps near two-pump design flow rate provides an adequate basis for identifying and evaluating degraded pump performance. Therefore, this testing method provides reasonable assurance of pump operational readiness.

The following is a summary of proposed alternative testing:

a. TMI, Unit 1 will continue to perform quarterly testing using a modified Group A test procedure as described above. With two paired-pumps in service, the required Group A test parameters will be measured except for individual pump flow rate. Individual pump flow rates will be calculated and compared against calculated individual pump flow rate reference values. During this test, the differential pressure for each pump will be throttled to the reference value.

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Three Mile Island Unit #1 Inservice Testing Program Plan 10 CFR 50.55a RELIEF REQUEST: PR-01 Revision 1 (Page 3 of 3)

b. Vibration measurements will be recorded and compared to their reference values.

Deviations from the reference value will be compared with the ranges specified in Table ISTB-5121-1 for Group A tests. Corrective actions will be taken in accordance with ISTB-6200.

c. During testing of the subject pumps (quarterly and refueling), TMI, Unit I will perform full spectrum vibration analysis, which is above Code required vibration testing.

d. The pumps will continue to be tested individually in accordance with ISTB-5123, "Comprehensive Test Procedure," during refueling outages.

Using the provisions of this relief request as an alternative to the specific requirements of ISTB-5121(b) will provide reasonable assurance of pump operational readiness. Therefore, pursuant to CFR 50.55a(a)(3)(i), TMI, Unit 1 requests approval of this alternative to the specific ISTB requirements identified in this request.

I

6. Duration of Proposed Alternative The proposed alternative identified will be utilized during the fifth IST interval which is scheduled to begin October 15, 2013 and conclude on October 14, 2023.

7. Precedents A similar Relief Request (P5) was approved for TMI, Unit 1 for the third 10-year interval as documented in the U.S. Nuclear Regulatory Commission's Safety Evaluation Report dated July 2, 2004 (ML041670196).

A similar Relief Request (PR-02) was approved for TMI, Unit 1 for the fourth 10-year interval as documented in the U.S. Nuclear Regulatory Commission's Safety Evaluation Report dated July 7, 2005 (ML051530406).

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Three Mile Island Unit #1 Inservice Testing Program Plan

,f RE*(;, UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 0

August 15, 2013 Mr. Michael J. Pacilio President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

THREE MILE ISLAND NUCLEAR STATION, UNIT 1 - RELIEF REQUESTS PR-01, PR-02, AND VR-02, ASSOCIATED WITH THE FIFTH 10-YEAR INSERVICE TEST INTERVAL (TAC NOS. MF0046, MF0047 AND MF0048)

Dear Mr. Pacilio:

By letter dated November 7, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12313A344), supplemented by letter dated April 17, 2013 (ADAMS Accession No. ML13108A003), Exelon Generation Company, LLC (the licensee) submitted proposed alternative requests PR-01, PR-02, and VR-02, associated with the fifth 10-year inservice test (IST) interval, at Three Mile Island, Unit I (TMI-1). These proposed alternatives apply to certain requirements of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code). PR-01 relates to a proposed alternative method for flow measurement during testing of the Nuclear Services Closed Cooling Water Pumps, submitted pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 55a(a)(3)(i). PR-02 requests approval for the use of ASME OM Code Case OMN-18, "Alternate Testing Requirements for Pumps Tested Quarterly Within +/-20 percent of Design Flow," for various pumps in the TMI-1 IST program, also submitted pursuant to 10 CFR 50.55a(a)(3)(i). OMN-18 would allow the licensee to not perform the Comprehensive Pump Test under certain conditions, and for certain pumps, as specified in the request. VR-02 requests approval for the licensee to utilize a tolerance band (grace period) for the intervals specified in the ASME OM Code, similar to the tolerance band allowed for surveillances in the plant technical specifications. This request was submitted pursuant to 10 CFR 50.55a(a)(3)(ii).

The U.S. Nuclear Regulatory Commission (NRC) staff has completed its review of the proposed alternatives as discussed in the enclosed safety evaluation. The NRC staff review concludes that for alternative requests PR-01 and PR-02, the proposed alternatives provide an acceptable level of quality and safety. For proposed alternative VR-02, the proposed alternative provides reasonable assurance that the components are operationally ready. Therefore, the NRC staff authorizes proposed alternative requests PR-01, PR-02 and VR-02, as proposed, at TMI-1, for the fifth 10-year IST program interval, which begins on October 15, 2013, and is scheduled to end on October 14, 2023.

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Three Mile Island Unit #1 Inservice Testing Program Plan M. Pacilio If you have any questions, please contact the TMI-1 Project Manager, Mr. Peter J. Bamford, at 301-415-2833.

Sincerely, Veronica Rodrigueief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-289

Enclosure:

Safety Evaluation cc w/encl: Distribution via ListServ Revision 0 October 25, 2013 A6 - 17

Three Mile Island Unit #1 Inservice Testing Program Plan UNITED STATES 0 NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REGARDING RELIEF REQUESTS PR-01, PR-02, AND VR-02 ASSOCIATED WITH THE FIFTH 10-YEAR INSERVICE TEST INTERVAL EXELON GENERATION COMPANY, LLC THREE MILE ISLAND NUCLEAR STATION, UNIT I DOCKET NO. 50-289

1.0 INTRODUCTION

By letter dated November 7, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12313A344), as supplemented by letter dated April 17, 2013 (ADAMS Accession No. ML13108A003), Exelon Generation Company, LLC (the licensee) submitted proposed alternative requests PR-01, PR-02, and VR-02 to the U.S. Nuclear Regulatory Commission (NRC). The licensee proposed alternatives to certain inservice testing (IST) requirements of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code), for the IST program at the Three Mile Island Nuclear Station, Unit 1 (TMI-1), for the fifth 10-year IST program interval.

Specifically, PR-01 relates to a proposed alternative method for flow measurement during testing of the Nuclear Services Closed Cooling Water (NSCCW) Pumps. PR-02 requests approval for the use of ASME OM Code Case OMN-18, "Alternate Testing Requirements for Pumps Tested Quarterly within +/-20 Percent of Design Flow," for various pumps in the TMI-1 IST program.

OMN-18 would allow the licensee to not perform the Comprehensive Pump Test (CPT) under certain conditions, applicable to certain pumps, as specified in the request. VR-02 would allow the licensee to utilize a tolerance band (grace period) to the intervals specified in the ASME OM Code, similar to the tolerance band allowed for surveillances in the plant technical specifications.

Pursuant to Title 10 of the Code of FederalRegulations (10 CFR) 50.55a(a)(3)(i), the licensee proposed the use of the alternatives in PR-01 and PR-02 on the basis that the alternatives provide an acceptable level of quality and safety. Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee proposed the use of the alternative in VR-02, on the basis that the ASME OM Code requirement presents an undue hardship without a compensating increase in the level of quality or safety.

2.0 REGULATORY EVALUATION

Pursuant to 10 CFR 50.55a(f), "Inservice Testing Requirements," IST of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME OM Code and applicable addenda.

Enclosure Revision 0 October 25, 2013 A6 -18

Three Mile Island Unit #1 Inservice Testing Program Plan Pursuant to 10 CFR 50.55a(a)(3), alternatives to ASME Code requirements may be authorized by the NRC if the licensee demonstrates that: (i) the proposed alternatives provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Based on the above, and subject to the NRC's findings with respect to authorizing the proposed alternatives to the ASME OM Code given below, the NRC staff finds that regulatory authority exists for the licensee to request, and the NRC staff to authorize, the alternatives requested by the licensee.

The Code of Record for the TMI-1 fifth 10-year IST program is the ASME OM Code, 2004 Edition with Addenda through OMb-2006. The TMI-1 fifth 10-year IST interval begins on October 15, 2013, and is currently scheduled to end on October 14, 2023.

3.0 TECHNICAL EVALUATION

3.1 Alternative Request PR-01. Nuclear Service Closed Coolinq Water Flow Rate Measurement During Group A Tests In this alternative, the licensee proposed the use of a modified quarterly Group A test for the IST of the NSCCW pumps. The NSCCW pumps have plant identification designators NS-P-1A, NS-P-1B, and NS-P-1C.

3.1.1 The Licensee's Proposed Alternative The ASME OM Code, Section ISTB-5121, "Group A Test Procedure," (b), states that "The resistance of the system shall be varied until the flow rate equals the reference point. The differential pressure shall then be determined and compared to its reference value. Alternatively, the flow rate shall be varied until the differential pressure equals the reference point and the flow rate determined and compared to the reference flow rate value."

According to the licensee, it is not practical to reduce the number of pumps in service to one to allow for single-pump testing during power operation due to the system design and plant operating requirements. The licensee also states that individual pump flow rates cannot be measured during the Group A test because the flow instrumentation for this system is located in the common discharge header for all three of the subject pumps. Further, according to the licensee, the piping configuration does not contain, nor would the system design permit, the installation of accurate individual pump flow measuring devices due to the turbulence caused by the valve and elbow configuration on the discharge of the pumps.

As an alternative to meeting the ASME OM Code requirements, the licensee proposes:

a. TMI, Unit 1 will continue to perform quarterly testing using a modified Group A test procedure as described above [in the application dated November 7, 2012, supplemented on April 17, 2013]. With two paired-pumps in service, the required Group A test parameters will be measured except for individual pump flow rate. Individual pump flow rates will be calculated and compared against calculated individual pump flow rate reference values. During this test, the differential pressure will be throttled to the reference value.

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Three Mile Island Unit #1 Inservice Testing Program Plan

b. Vibration measurements will be recorded and compared to their reference values. Deviations from the reference value will be compared with the ranges specified in Table ISTB-5121-1 for Group A tests. Corrective actions will be taken in accordance with ISTB-6200.

c. During testing of the subject pumps (quarterly and refueling), TMI, Unit 1 will perform full spectrum vibration analysis, which is done above Code required vibration testing.

d. The pumps will continue to be tested individually in accordance ISTB-5123, "Comprehensive Test Procedure," during refueling outages.

3.1.2 NRC Staff Evaluation The ASME OM Code, Table ISTB-3400-1, notes that Group A and Group B tests shall be performed quarterly, and CPTs shall be performed biennially. ISTB-5121(b) requires that for Group A tests, the resistance of the system shall be varied until the flow rate equals the reference point. The differential pressure shall then be determined and compared to its reference value.

Alternatively, the flow rate shall be varied until the differential pressure equals the reference point and the flow rate determined and compared to its reference value.

The three NSCCW pumps are centrifugal pumps. The flow instrumentation for this system is located in the common discharge header for all three of the subject pumps. Due to the system design and plant operating requirements, it is not practical to reduce the number of pumps in service to one to allow for single-pump testing during power operation. The licensee states that the piping configuration does not contain, nor would the system design permit the installation of, accurate individual pump flow measuring devices due to the turbulence caused by the valve and elbow locations in the piping configuration on the discharge of the pumps. Compliance with the OM Code requirement to measure individual pump flow rate would require installation of new piping or certain system modifications. As such, the licensee requests an alternative to the quarterly requirements of individual flow measurement, and proposes to calculate the individual pump flow rate from the combined flow measurements of three paired-pump tests (A-B, B-C, and A-C). The calculated individual pump flow rates will be compared against individual pump flow rate calculated reference values. The pump will continue to be tested individually in accordance with ISTB-5123 for CPTs during refueling outages. In addition, for the quarterly Group A tests and CPTs, TMI-1 will also perform a full spectrum vibration analysis for the affected pumps.

The NRC staff finds that the method of calculating individual flow rate from three-paired pump tests (A-B, B-C, and A-C) provides similar and equally acceptable flow rates for each affected pump, as opposed to individually measured flow rates. On the basis that direct measurements of individual pump flow rate is not available, and that the modified quarterly Group A test provides similar and equally acceptable results, the NRC concludes that the proposed alternative provides an acceptable level of quality and safety for testing the NSCCW pumps.

3.2 Alternative Request PR-02, Use of Code Case OMN-18 In this alternative, the licensee proposed the use of ASME OM Code Case OMN-18, "Alternate Testing Requirements for Pumps Tested Quarterly within +/-20% [percent] of Design Flow," for the following pumps:

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Three Mile Island Unit #1 Inservice Testing Program Plan Pump Description Category AH-P-3A Control Building Chilled Water Supply Pump A Group A AH-P-3B Control Building Chilled Water Supply Pump B Group A BS-P-1A Reactor Building Spray Pump A Group AB BS-P-1B Reactor Building Spray Pump B Group AB DC-P-1A Decay Heat Closed Cooling Water Pump A Group A DC-P-1 B Decay Heat Closed Cooling Water Pump B Group A DR-P-1A Decay Heat River Water Pump A Group A DR-P-1B Decay Heat River Water Pump B Group A SF-P-1A Spent Fuel Cooling Pump A Group A SF-P-1B Spent Fuel Cooling Pump B Group A The ASME OM Code Committee has approved Code Case OMN-18, which allows licensees to perform a Group A test in lieu of the biennial CPT ifthe Group A tests are conducted at +/-20 percent of the design flow rate and pressure instruments that meet the CPT accuracy requirement of +/-1/2 percent are used. This Code Case was not reviewed for approval in Regulatory Guide (RG) 1.192, "Operation and Maintenance Code Case Acceptability, ASME OM Code," dated June 2003.

3.2.1 The Licensee's Proposed Alternative The licensee proposes that in lieu of the CPT requirements of the ASME OM Code, Table ISTB-3400-1, Group A tests will be performed quarterly within +/-20 percent of the pump design flow rate, with pressure measuring instrumentation meeting the (+/-1/2 percent) instrument accuracy requirements of Table ISTB-3510-1 specified for the biennial CPT. In addition, the licensee has elected to restrict the upper limit for acceptable range to 106 percent of the reference value (flow rate or differential pressure), which is more conservative than the 110 percent upper limit established in Tables ISTB-5121-1 and ISTB-5221-1 for Group A tests.

Vibration testing will continue to be performed under the proposed modified Group A test and the acceptance criteria for vibration will be the same as required for Group A tests as shown in Tables ISTB-5121-1 and ISTB-5221-1.

The licensee states that using the provisions of the request as an alternative to the requirements specified in ISTB-3400 and Tables ISTB-3400-1, ISTB-5121-1, and ISTB-5221-1, will provide a reasonable alternative to the ASME OM Code requirements based on the determination that the proposed alternative will provide adequate indication of pump performance, permit consistent detection of component degradation, and continue to provide an acceptable level of quality and safety.

3.2.2 NRC Staff Evaluation The licensee is proposing to perform a quarterly IST for all the pumps listed above in accordance with a modified Group A test procedure, in lieu of a quarterly Group A test and a biennial CPT.

The ASME OM Code requires that for Group A pumps, a Group A test be performed every quarter, and a CPT be performed biennially. The Group A test is performed within +/-20 percent of the pump design flow rate and the pressure instrument accuracy is +/-2 percent. The upper limit Revision 0 October 25, 2013 A6 - 21

Three Mile Island Unit #1 Inservice Testing Program Plan for the "Acceptable Range" for flow rate and differential pressure is 110 percent of the reference values, and the high value for the "Required Action Range" for flow rate and differential pressure is greater than 110 percent of the reference values. The CPT is performed within +/-20 percent of the pump design flow rate, the pressure instrument accuracy is +/-1/2 percent, and the upper limit of the "Acceptable Range" and the high value for the "Required Action Range" for flow rate and differential pressure is 103 percent of the reference values. Vibration monitoring is performed during both the Group A tests and the CPTs.

The licensee proposes that for the pumps covered by this proposed alternative, a modified Group A quarterly test will be performed using ASME OM Code Case OMN-1 8, with modified "Acceptable" and "Required Action" ranges, and the biennial CPT will not be performed. The modified Group A quarterly test would be performed within +/-20 percent of the pump design flow rate, using more accurate pressure instrumentation that is required for a CPT (+/-1/2 percent instead of +/-2 percent). The licensee will use a more limiting upper bound of 106 percent of the reference value for the "Acceptable Range" for flow rate and differential pressure, in lieu of 110 percent of the reference value that is normally required by the ASME OM Code for Group A tests.

However, this upper bound of 106 percent is greater than the upper bound value of 103 percent for the biennial CPT. Using more accurate pressure gauges and a more limiting "Acceptable Range" (compared to the Group A test "Acceptable Range") during every modified quarterly Group A test compensates for the elimination of the CPT with its more limiting "Acceptable Range" upper bound value of 103 percent.

OMN-1 8 was published in the 2009 Edition of ASME OM Code. This edition of the ASME OM Code has not been incorporated by reference into 10 CFR 50.55a, and OMN-18 has not been incorporated into RG 1.192. However, the NRC staff has reviewed OMN-18, and currently has no concerns with its use, providing that the upper end values of the Group A test "Acceptable Ranges" for flow and differential pressure are 106 percent of the respective reference values, and the high values of the "Required Action Ranges" for flow and differential pressure are greater than 106 percent of the respective reference values. The NRC staff considers the proposed alternative acceptable because all of the tests will be performed with pressure gauges with +/-1/2 percent accuracy. The elimination of the CPT, with its more limiting "Acceptable Range" upper bound of 103 percent of the reference value, is compensated for by using more accurate pressure instrumentation on every quarterly test. Regular testing with more accurate instrumentation and tighter acceptance criteria will provide for better trending of pump performance. Therefore, the NRC finds that the proposed alternative provides an acceptable level of quality and safety for testing and acceptance criteria for the pumps specified in this proposed alternative.

3.3 Alternative Request VR-02, ASME OM Code Test Frequencies This request applies to the frequency specifications of the ASME OM Code. The frequencies for tests are provided in various sections of the ASME OM Code, but do not include a tolerance band.

3.3.1 The Licensee's Proposed Alternative According to the licensee:

ASME OM Code Section IST establishes the inservice test frequency for all components within the scope of the Code. The frequencies (e.g., quarterly) have always been interpreted as "nominal" frequencies (generally as defined in the Table 3.2 of NUREG-1482, Revision 1) and Owners routinely applied the Revision 0 October 25, 2013 A6 - 22

Three Mile Island Unit #1 Inservice Testing Program Plan surveillance extension time period (i.e., grace period) contained in the plant Technical Specifications (TS) Surveillance Requirements (SRs). The TS typically allow for a less than or equal to 25% extension of the surveillance test interval to accommodate plant conditions that may not be suitable for conducting the surveillance. However, regulatory issues have been raised concerning the applicability of the TS "Grace Period" to ASME OM Code required inservice test frequencies.

The lack of a tolerance band (grace period) on the ASME OM Code inservice test frequency restricts operational flexibility. There may be a conflict where a surveillance test could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be performed until sometime after a plant condition or associated Technical Specification is applicable. Therefore, to avoid this conflict, the IST should be performed when it can be and should be performed.

The NRC recognized this potential issue in the TS by allowing a frequency tolerance as described in the TMI Unit 1 TS SR 4.0.1. The lack of a similar tolerance applied to the OM Code testing places an unusual hardship on the plant to adequately schedule work tasks without operational flexibility.

Thus, just as with TS required surveillance testing, some tolerance is needed to allow adjusting ASME OM Code testing intervals to suit the plant conditions and other maintenance and testing activities. This assures operational flexibility when scheduling surveillance tests that minimize the conflicts between the need to complete the surveillance and plant conditions.

The licensee's alternative proposes to adopt wording equivalent to the ASME Board of Nuclear Codes and Standards (BNCS)-approved OM Code Case OMN-20, for determining acceptable tolerances for pump and valve test frequencies. This Code Case was approved by the ASME OM Code Standards Committee in February 2012. The proposed alternative applies to the various frequency specifications of the ASME OM Code for all pumps and valves contained within the IST Program scope.

The licensee's proposal states:

ASME OM Code establishes component test frequencies that are based either on elapsed time periods (e.g., quarterly, 2 years, etc.) or on the occurrence of plant conditions or events (e.g., cold shutdown, refueling outage, upon detection of a sample failure, following maintenance, etc.).

a. Components whose test frequencies are based on elapsed time periods shall be tested at the frequencies specified in ASME OM Code Section IST with a specified time period between tests as shown in the table below and as defined in TMI, Unit 1 TS Section 1.25.

b. The specified time period between tests may be reduced or extended as follows:

1) For periods specified as less than 2 years, the period may be extended by up to 25% for any given test. This is consistent with TMI, Unit 1 TS Section 1.25, "FREQUENCY NOTATION."

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Three Mile Island Unit #1 Inservice Testing Program Plan

2) For periods specified as greater than or equal to 2 years, the period may be extended by up to 6 months for any given test.

3) All periods specified may be reduced at the discretion of the owner (i.e.,

there is no minimum period requirement).

4) Period extensions may also be applied to accelerated test frequencies (e.g., pumps in Alert Range).

Specified Time Period Between Tests (all Frequency values are 'not to exceed'; no minimum periods are specified)

Quarterly 92 days (or every 3 months)

Semiannually 184 days (or every 6 months)

Annually 366 days (or every year) x Years x calendar years where 'x' is a whole number of years ? 2

c. Components whose test frequencies are based on the occurrence of plant conditions or events (e.g., cold shutdown, refueling outage, upon detection of a sample failure, following maintenance, etc.) may not have their period between tests extended except as allowed by the ASME OM Code.

Period extension is to facilitate test scheduling and considers plant operating conditions that may not be suitable for performance of the required testing (e.g.,

performance of the test would cause an unacceptable increase in the plant risk profile due to transient conditions or other ongoing surveillance test or maintenance activities). Period extensions are not intended to be used repeatedly merely as an operational convenience to extend test intervals beyond those specified. This request is not applicable to frequencies in Subsection ISTD.

3.3.2 NRC Staff Evaluation Historically, licensees have applied, and the NRC staff has accepted, the standard TS definitions for IST intervals (including allowable interval extensions) to ASME OM Code required testing (Reference NUREG-1482, "Guidelines for Inservice Testing at Nuclear Power Plants," Revision 1, Section 3.1.3). Recently, the NRC staff reconsidered the allowance of the TS testing intervals and interval extensions, for IST not associated with TS SRs. As noted in Regulatory Issue Summary (RIS) 2012-10, "NRC Staff Position on Applying Surveillance Requirements 3.0.2 and 3.0.3 to Administrative Controls Program Tests," the NRC determined that programmatic test frequencies can't be extended in accordance with the Improved Standard TS SR 3.0.2. This includes all IST described in the ASME OM Code not specifically required by the TS SRs.

Following this development, the NRC staff sponsored and co-authored an ASME OM Code inquiry and Code Case to modify the ASME OM Code to include TS-tike test interval definitions and interval extension criteria. The resultant BNCS-approved Code Case OMN-20, was Revision 0 October 25, 2013 A6 - 24

Three Mile Island Unit #1 Inservice Testing Program Plan approved by the ASME Operation and Maintenance Standards Committee on February 15, 2012, with the NRC representative voting in the affirmative on this proposed Code Case. The licensee proposed to adopt language equivalent to the BNCS-approved Code Case OMN-20.

The NRC staff finds that requiring the licensee to meet the ASME OM Code requirements, without an allowance for defined frequency and frequency extensions for IST of pumps and valves, results in a hardship without a compensating increase in the level of quality and safety.

Based on the prior acceptance by the NRC staff of the similar TS test interval definitions and interval extension criteria, the staff finds that implementation of the test interval definitions and interval extension criteria contained in the ASME BNCS-approved OM Code Case OMN-20 is acceptable. Allowing usage of an alternative equivalent to Code Case OMN-20 provides reasonable assurance of operational readiness of pumps and valves subject to the ASME OM Code IST.

4.0 CONCLUSION

As set forth above, the NRC staff has determined that for alternative requests PR-01 and PR-02, the proposed alternatives provide an acceptable level of quality and safety. For alternative VR-02, the proposed alternative provides reasonable assurance that the 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(a)(3)(i) for requests PR-01 and PR-02. The NRC staff also concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii) for request VR-02. All other ASME OM Code requirements for which relief was not specifically requested and approved remain applicable. Therefore, the NRC staff authorizes alternative requests PR-01, PR-02 and VR-02, as proposed, at TMI-1, for the fifth 10-year IST program interval, which begins on October 15, 2013, and is scheduled to end on October 14, 2023.

Principle Contributors: J. Huang, NRR R. Wolfgang, NRR J. Billerbeck, NRR Date: August 15, 2013 Revision 0 October 25, 2013 A6 - 25

Three Mile Island Unit #1 Inservice Testing Program Plan M. Pacilio Ifyou have any questions, please contact the TMI-1 Project Manager, Mr. Peter J. Bamford, at 301-415-2833.

Sincerely, Ira/

Veronica Rodriguez, Acting Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-289

Enclosure:

Safety Evaluation cc w/encl: Distribution via ListServ DISTRIBUTION:

PUBLIC RidsAcrsAcnw MailCTR Resource RidsNrrLAABaxter Resource LPLI-2 R/F RidsNrrPMThreeMilelsland Resource RidsRgn1MailCenter Resource RidsNrrDeEpnb Resource RidsNrrDorlDpr Resource RidsNrrDorlLpll-2 Resource RWolfgang, NRR VCampbell, OEDO, Region I JHuang, NRR JBillerbeck, NRR ADAMS Accession Number: ML13227A024

• by email OFFICE LPLI-2/PM LPLI-2/LA EPNBIBC*- LPLI-2/BC(A)I NAME PBamford ABaxter TLupold VRodriguez DATE 08/14113 08/114113 07/25/113 08/1513 OFFICIAL RECORD COPY Revision 0 October 25, 2013 A6 - 26

Three Mile Island Unit #1 Inservice Testing Program Plan UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 August 28, 2013 Mr. Michael J. Pacilio President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

THREE MILE ISLAND NUCLEAR STATION, UNIT 1 - RELIEF REQUEST VR-01, PROPOSED ALTERNATIVE TESTING OF THE PRESSURIZER PILOT OPERATED RELIEF VALVE (TAC NO. ME9819)

Dear Mr. Pacilio:

By letter dated October 18, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12292A585), supplemented by letter dated March 15, 2013 (ADAMS Accession No. ML13074A700), Exelon Generation Company, LLC (the licensee) submitted proposed alternative request VR-01, associated with the fifth 10-year inservice test (IST) interval at Three Mile Island, Unit 1 (TMI-1). This proposed alternative applies to certain requirements of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code). VR-01 proposes an alternative method for testing of the Pressurizer Pilot Operated Relief Valve (PORV), submitted pursuant to Title 10 of the Code of FederalRegulations, Part 50, Section 55a(a)(3)(i).

The U.S. Nuclear Regulatory Commission (NRC) staff has completed its review of the proposed alternative, as discussed in the enclosed safety evaluation. The NRC staff review concludes that alternative request VR-01 provides an acceptable level of quality and safety, and that it provides reasonable assurance that the PORV is operationally ready. Therefore, the NRC staff authorizes proposed alternative request VR-01, as proposed, for the fifth 10-year IST program interval at TMI-1, which begins on October 15, 2013, and is scheduled to end on October 14, 2023.

Revision 0 October 25, 2013 A6 - 27

Three Mile Island Unit #1 Inservice Testing Program Plan M. Pacilio If you have any questions, please contact the TMI-1 Project Manager, Mr. Peter J. Bamford, at 301-415-2833.

Sincerely, Veronica Rodriguez, Actin hief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-289

Enclosure:

Safety Evaluation cc w/encl: Distribution via ListServ Revision 0 October 25, 2013 A6 - 28

Three Mile Island Unit #1 Inservice Testing Program Plan S,.UNITED STATES 0*=NUCLEAR A REGULATORY COMMISSION

, WASHINGTON, D.C. 20555-0001

• • • • • ' SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REGARDING RELIEF REQUEST VR-01 PROPOSED ALTERNATIVE TESTING OF THE PRESSURIZER PILOT OPERATED RELIEF VALVE EXELON GENERATION COMPANY, LLC THREE MILE ISLAND NUCLEAR STATION, UNIT 1 DOCKET NO. 50-289

1.0 INTRODUCTION

By letter dated October 18, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12292A585), supplemented by letter dated March 15, 2013 (ADAMS Accession No. ML13074A700), Exelon Generation Company, LLC (the licensee) submitted proposed alternative request VR-01, associated with the fifth 10-year inservice test (IST) interval, at Three Mile Island, Unit 1 (TMI-1). The proposed alternative applies to certain requirements of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code). VR-01 relates to a proposed alternative method for testing of the Pressurizer Pilot Operated Relief Valve (PORV), submitted pursuant to Title 10 of the Code of FederalRegulations (10 CFR), Part 50, Section 55a(a)(3)(i). Specifically, the proposed alternative would utilize a bench testing protocol for the PORV in lieu of certain provisions of the ASME OM Code that require in-situ testing.

2.0 REGULATORY EVALUATION

Pursuant to 10 CFR 50.55a(f), "Inservice Testing Requirements," IST of certain ASME Code Class 1, 2, and 3 components must meet the requirements of the ASME OM Code and applicable addenda.

Pursuant to 10 CFR 50.55a(a)(3), alternatives to ASME Code requirements may be authorized by the NRC if the licensee demonstrates that: (i) the proposed alternatives provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Based on the above, and subject to the NRC's findings with respect to authorizing the proposed alternative to the ASME OM Code given below, the NRC staff finds that regulatory authority exists for the licensee to request, and the NRC staff to authorize, the alternative requested by the licensee.

Enclosure Revision 0 October 25, 2013 A6 -29

Three Mile Island Unit #1 Inservice Testing Program Plan The Code of Record for the TMI-1 fifth 10-year IST program is the ASME OM Code, 2004 Edition with Addenda through OMb-2006. The TMI-1 fifth 10-year IST interval begins on October 15, 2013, and is currently scheduled to end on October 14, 2023.

3.0 TECHNICAL EVALUATION

3.1 Licensee's Alternative Request The ASME OM Code requirements that apply to the TMI-1 PORV (1-RC-RV-2), applicable to this request, include requirements to exercise the valve after replacement, and requirements to perform exercise testing of the valve once per fuel cycle.

Currently, the licensee satisfies these requirements by manually stroking the valve once every operating cycle. This is performed during plant startup following a refueling outage. The valve must also be stroke-timed during this exercise test. The licensee is proposing an alternative to this required in-situ testing for several reasons, stated as follows in the submittal dated October 18, 2012:

There are several disadvantages to the in-situ testing of the PORV. The PORV is a 2.5 inch Dresser Electromatic, solenoid actuated, pilot operated relief valve.

Operation of the pilot valve vents the chamber under the main valve disc which causes the main valve to open. The PORV requires steam pressure for the main disc to open. Stroke testing the PORV during cold shutdown conditions does not exercise the main valve disc which, therefore, does not satisfy the subject ASME OM Code requirements. To test the PORV in-place, the RCS [reactor coolant system] must be pressurized to supply the necessary fluid (steam) pressure to open the main valve disc.

Also, since the PORV design does not provide direct obturator position indication, the valve disc position must be inferred from alternate indications (tailpipe AT, acoustic monitoring, RCS pressure decrease, or quench tank pressure or level rise).

In-situ testing of the PORV also results in an in-surge of cooler water from the RCS hot leg into the pressurizer. The resulting thermal cycle on the pressurizer surge line is a thermal stress concern, as described in NRC Bulletin 88-08

("Thermal Stresses in Piping Connected to Reactor Coolant Systems") and should be avoided.

Requiring that the PORV be tested in-place prevents plant personnel from verifying proper reseating of the main valve disc because the discharge is not visible as it is during bench testing. Minor leakage would not be readily evident before it would cause damage to the main valve disc/seat. Excessive leakage from the pilot valve can lead to inadvertent opening of the main valve and impair its ability to re-close.

The proposed alternative will allow testing of the PORV that is appropriate to demonstrate functionality without cycling the valve in place using reactor steam pressure. This is consistent with NUREG-0737, "Clarification of TMI Action Plan Requirements," Item II.K.3.16, "Reduction of Challenges and Failures of Relief Revision 0 October 25, 2013 A6 - 30

Three Mile Island Unit #1 Inservice Testing Program Plan Valves," which recommended that the number of relief valve openings be reduced as much as possible and that unnecessary challenges should be avoided.

The licensee proposes the following alternatives to the requirements regarding stroking following replacement, and once per refueling cycle, for the fifth IST interval at TMI-1:

1) Bench testing of the PORV to satisfy valve exercise and stroke time requirements is performed at the vendor test facility prior to installation.

Exercising of the valve at both the normal power operation set point and the Low Temperature Overpressure Protection (LTOP) set point (as provided in Technical Specification 3.1.12, "Pressurizer Power Operated Relief Valve (PORV), Block Valve, and Low Temperature Overpressure Protection (LTOP)")

will be verified during this testing. Measured stroke time will be based on the pressure response indication of main disc opening.

2) The installed valve will be removed and replaced each refueling outage, with a spare valve that has been previously bench tested.

3) The removed valve will be bench tested within one year of removal from the system.

4) In-situ exercising of the PORV will be performed only as necessary to reestablish operational readiness after maintenance on an installed valve.

In the application, the licensee provided a detailed justification for the use of bench testing in lieu of in-situ testing. Included in this justification was a table showing bench test stroke time history between August 31, 2000, and November 4, 2011, for both the LTOP function and the normal reactor coolant system pressure function of the PORV. According to the licensee, these results.

consistently show that the valve opens well within the 2-second limiting stroke time allowed by ISTC-5114(c) for rapid acting valves.

3.2 NRC Staff Evaluation The licensee has categorized the Pressurizer PORV, 1-RC-RV-2, as OM Code category B/C and, therefore, the valve is subject to the applicable test requirements of both ASME OM Code Subsection ISTC for power-operated valves and Mandatory Appendix I for pressure relief devices.

The requirements of Mandatory Appendix I allow the valve to be removed from the system for testing and do not specifically require that the valve be exercised when it is returned to the system. However, paragraphs ISTC-3310 and ISTC-3510 could be interpreted to require in-situ exercising of the valve following replacement (ISTC-3310) or routinely at a once per fuel cycle frequency (ISTC-351 0).

The licensee has determined that exercising the PORV in-situ is undesirable for a number of reasons. The NRC staff finds that exercising the PORV in-situ at normal steam pressures does present undesirable circumstances. There is some precedent in ASME OM Code itself for not exercising relief devices following reinstallation after testing. For Boiling Water Reactor (BWR)

Class 1 Main Steam Pressure Relief Valves with Auxiliary Actuating Devices, Mandatory Appendix 1-3410(d) allows that after removal and reinstallation for testing, the electrical and Revision 0 October 25, 2013 A6 - 31

Three Mile Island Unit #1 Inservice Testing Program Plan pneumatic connections may be verified by inspection in lieu of test, and that valve main disk movement (i.e., exercise) is not required. While this ASME OM Code example is not specifically applicable to Pressurized Water Reactor PORVs, it is analogous.

There is further precedent in a number of prior licensing actions wherein the NRC staff has approved technical specification changes for various licensees to remove routine in-situ exercise requirements for BWR Class 1 safety/relief valves for the same and similar reasons as presented by this licensee.

The licensee has proposed alternatives, which include a series of verifications and controls to demonstrate the operational readiness of the valve:

Multiple bench test verifications performed in the same orientation as the plant installation and using test conditions similar to those in the plant installation, including ambient temperature, valve insulation, and steam conditions:

0 exercising the pilot and main valve disk as a unit N set point verification M obturator movement verification 0 stroke time testing 0 seat leakage verification a refurbishment, as needed The licensee indicates, and the NRC staff agrees, that performing some of these tests on the bench is actually preferred over in-situ testing because they can be performed more precisely.

Additional steps taken by the licensee to ensure operational readiness of the PORV are:

0 Receipt inspection and storage in accordance with quality procedures to ensure protection against physical damage and moisture upon return to the plant site 0 Pre-installation inspection for foreign material and damage

• Installation and connection in accordance with quality maintenance procedures 0 Operation of the solenoid-actuated pilot valve (which controls the actuation of the valve main disk) in-situ to verify electrical power and control connections The licensee has further stated that if maintenance is ever required on an installed valve, in-situ exercising will be performed to reestablish operational readiness.

The NRC staff has reviewed the licensee's proposed alternative and concludes that implementation of these alternatives will continue to meet the fundamental intent of ASME OM Code to assure the PORV operational readiness and to permit detection of PORV degradation.

The proposed alternative demonstrates proper PORV operation without the need for in-situ testing with reactor steam, and therefore provides an acceptable level of quality and safety.

Revision 0 October 25, 2013 A6 - 32

Three Mile Island Unit #1 Inservice Testing Program Plan

4.0 CONCLUSION

As set forth above, the NRC staff determines that for alternative request VR-01, the proposed alternative provides an acceptable level of quality and safety and also provides reasonable assurance that the PORV is 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(a)(3)(i). All other ASME OM Code requirements for which relief was not specifically requested and approved remain applicable. Therefore, the NRC staff authorizes alternative request VR-01, to be implemented at TMI-1 for the fifth 10-year IST program interval, which is scheduled to begin on October 15, 2013, and conclude on October 14, 2023.

Principle Contributor: John Billerbeck, NRR Date: August 28, 2013 Revision 0 October 25, 2013 A6 - 33

ML13232A051

• by email

-I O FFICE LPLI-2/PM LPLI-2/LA* EEPNBIBC* LPLI-2/1BC/A NAME PBamford ABaxter 1rLupold VRodriguez DATE 08/20113 08127/13 C18114113 08/28/13 Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 7 CODE CASE INDEX Revision 0 October 25, 2013 A7- 1

Three Mile Island Unit #1 Inservice Testing Program Plan CODE CASE TITLE NUMBER OMN-18 The ASME Code committee has approved Code Case OMN-18, "Alternate Testing Requirements for Pumps Tested Quarterly within +

20% of Design Flow".

Revision 0 October 25, 2013 A7 - 2

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 8 COLD SHUTDOWN JUSTIFICATION INDEX Revision 0 October 25, 2013 A8 -1

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 8 COLD SHUTDOWN JUSTIFICATION INDEX Designator Rev Title/Description CS-01 Intermediate Cooling Water Valves (IC-V-2,3,4,6) Exercise Close, Fail Close CS-02 Decay Heat Suction Isolation Valve (DH-V-1/2) Exercise Open/Close CS-03 Charging Line Iso Valve (MU-V-18) Exercise Close and Fail Close CS-04 RCP Seal Water Iso Valve (MU-V-20) Exercise Close and Fail Close CSJ-05 RCP Seal Water Return Iso Valves (MU-V-25/26) Exercise Close and Fail Close CSJ-06 Letdown Iso Valves (MU-V-2A/B, 3) Exercise Close and Fail Close CSJ-07 RC Motor Return Iso Valve (NS-V-1 5) Exercise Close CSJ-08 RC Motor Return Iso Valves (NS-V-35/4) Exercise Close CSJ-09 Reactor Building Cooling Water Iso Valves (RB-2A/7) Exercise Close CSJ-10 FW Isolation Valves (FW-V-16A/B, 17A/B,5A/B,92A/B) Exercise Close CSJ-11 Pressurizer Vent Valves (RC-V-28/44) Exercise Close and Fail Close CSJ-12 RC Vent Valves (RC-V-40A/B, 41A/B) Exercise Close and Fail Close CSJ-13 Reactor Vessel Vent Valves (RC-V-42/43) Exercise Close and Fail Close CSJ-14 DH Discharge Isolation Valves (DH-V-4A/B) Exercise Open/Close CSJ-15 CA-T-1 To Makeup System Check Valve (CA-V-1 77) Exercise Open/Close CSJ-16 Main Steam Isolation Valves (MS-V-1A-D) Exercise Close CSJ-17 Pressurizer Spray Line Isolation Valve (RC-V-3) Exercise Close CSJ-18 Decay Heat Pressurizer Spray Line Isolation Valve (RC-V-4)

Exercise Open/Close Revision 0 October 25, 2013 A8 - 2

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATIONS Revision 0 October 25, 2013 A9 -1

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATION CSJ-01 Valve Number System Safety Class Category IC-V-2 IC 2 A IC-V-3 IC 2 A IC-V-4 IC 2 A IC-V-6 IC 2 A Function These power operated valves close to isolate containment from the IC System on 1) receipt of any ESAS signal concurrent with a IC Surge Tank low level signal or 2) on a 30 psig Reactor Building pressure ESAS signal, regardless of IC Surge Tank level. These air-operated valves (IC-V-3, 4, and 6) also fail closed on loss of electrical power or pneumatic supply.

These valves have no safety function to open. The IC System inside the Reactor Building is non-ASME Safety Class, Seismic Category II and does not serve any components which are required for the safe shutdown of the reactor or for the mitigation of any Design Basis accidents.

Test Requirement Exercise test to the closed position in accordance with ISTC-3510.

Justification It is not practical to full stroke exercise or fail safe test these valves closed during normal power operation since closure testing at normal power imposes a significant risk of equipment damage.

Exercising these valves closed during power operation will isolate the cooling water supply to the Control Rod Drive Cooling Coils, Primary Letdown Coolers and Reactor Coolant Pump Heat Exchangers. Interruption of cooling water flow to these components would cause equipment damage and a potential trip of the reactor.

Alternative Test These valves will be exercised closed (IC-V-2, 3, 4, 6) and fail safe tested closed (IC-V-6) during cold shutdowns when cooling water supplied to the Control Rod Drive Cooling Coils, Primary Letdown Coolers and Reactor Coolant Pump Heat Exchangers is not required.

These valves will be partial stroke exercised quarterly during normal power operations.

Revision 0 October 25, 2013 A9 - 2

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATION CSJ-02 Valve Number System Safety Class Category DH-V-1 DH 1 A DH-V-2 DH 1 A Function These normally-closed motor-operated valves are required to open to allow use of the drop line to circulate water through the Core following a cold-leg LOCA in order to prevent the concentration of boron from exceeding its solubility limit. Precipitation of boron could result in clogged flow channels, limiting the ability to remove heat.

These valves are also opened in order to use the DH System as the normal and preferred method of placing and maintaining the Reactor in the Cold Shutdown condition, including those times when it is necessary to do so in response to a Technical Specification required-action statement.

These valves receive an automatic signal to close and cannot be reopened if RCS pressure exceeds 400 psig in order to prevent over-pressurization of the lower pressure-rated DH System pump suction piping.

Test Requirement Full-stroke exercise to the open and closed positions in accordance with ISTC-351 0.

Justification It is not practical to full stroke exercise or partial stroke exercise these valves open or closed during normal power operation since opening the valves during normal power operation would cause over pressurization of the lower pressure DH system. Additionally, these valves are interlocked with reactor coolant system pressure such that they can not be opened while reactor coolant system pressure is greater than 400 psig.

Alternative Test These valves will be exercised open and closed during cold shutdowns when the reactor coolant system is depressurized.

Revision 0 October 25, 2013 A9 - 3

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATION CSJ-03 Valve Number System Safety Class Category MU-V-18 MU 2 A Function This normally-open, air-operated valve is required to close on an ESAS actuation signal of 1600 psig RCS pressure or 4 psig Reactor Building pressure to 1) assure adequate, balanced HPI flow to the RCS and to isolate containment. An ESAS backup initiation signal is also provided at 500 psig RCS pressure. Closure of this valve prevents HPI flow from bypassing the cavitating venturi at the B HPI nozzle. The valve fails closed on loss of air and open on loss of power. Loss of DC power is the identified limiting single failure associated with the closing safety function of this valve.

This valve is normally in the open (i.e., throttled) position to provide makeup flow to the RCS. This open function is not required for safe shutdown or accident mitigation.

Test Requirement Exercise test to the closed position in accordance with ISTC-3510.

Justification It is not practical to full stroke exercise or fail safe test this valve closed during normal power operation since closure testing at normal power operations would lead to a potential loss of reactor coolant system inventory control and potential trip of the reactor.

Closing this valve for testing during normal power operations would interrupt makeup flow to the reactor coolant system which could cause a loss of inventory control and subsequent trip of the reactor. This valve is required to be open for throttling and maintaining RCS inventory control during power operations.

Alternative Test This valve will be exercised closed and fail safe tested during cold shutdowns when the reactor coolant system makeup control is not required. The valve will be partial stroke exercised quarterly during normal power operations.

Revision 0 October 25, 2013 A9 - 4

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATION CSJ-04 Valve Number System Safety Class Category MU-V-20 MU 2 A Function This valve is normally-open providing shaft seal injection for the reactor coolant (RC) pumps. This air-operated valve is closed by remote manual operation to isolate the Reactor Building (containment) following an accident. It does not receive an ESAS signal to close because of the importance of seal injection to the integrity of the RC pumps shaft seals. Loss of seal injection due to an erroneous actuation of ESAS or by a non-LOCA event such as RCS overcooling could cause significant damage to the associated shaft seal or pump shaft. The valve fails closed on loss of electrical power or pneumatic supply.

Test Requirement Exercise test to the closed position in accordance with ISTC-351 0.

Justification It is not practical to full stroke exercise or fail safe test this valve closed during normal power operation since closure testing at normal power operations could cause significant

.damage to the reactor coolant pumps shafts and shaft seals.

Closing this valve for testing during normal power operations would interrupt seal injection flow to the reactor coolant pumps shaft seals and risk permanent damage to the reactor coolant pump seals and cause subsequent Reactor Coolant system leakage.

Additionally, partial stroke exercising of the valve during operation would involve manipulation of the manual hand-wheel to position the "block" so that adequate RC pump seal injection is maintained. The air operator design is such that excessive use of the hand-wheel tends to wear the shaft seals, resulting in increased leakage of the air that's required to maintain the valve open upon failure of the air source. Therefore, quarterly partial stroke exercising of the valve presents a premature failure vulnerability without a significant benefit in monitoring valve performance.

Alternative Test This valve will be exercised closed and fail safe tested during cold shutdowns when the reactor coolant pumps are not in service and seal injection is not required.

Revision 0 October 25, 2013 A9 - 5

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATION CSJ-05 Valve Number System Safety Class Category MU-V-25 MU 2 A MU-V-26 MU 2 A Function These normally open valves close automatically on a 30 psig Reactor Building ESAS signal to isolate the Reactor Building makeup and purification system. The high pressure signal is used to ensure that sealing water flow remains available through the mechanical seals of the Reactor Coolant Pumps for those events in which continued operation of the pumps provides an advantage in maintaining flow through the Core. Valve MU-V-26 also fails open on loss of electrical power and closed on loss of pneumatic supply.

These valves are normally open to provide a return flow path for the seal leakoff of the Reactor Coolant Pumps. This function is not required for safe shutdown or accident mitigation.

Test Requirement Exercise test to the closed position in accordance with ISTC-3510.

Justification It is not practical to full stroke exercise or fail safe test these valves closed during normal power operation since closure testing at normal power operations could cause significant

.damage to the reactor coolant pump seals.

Closing these valves for testing during normal power operations would disrupt seal leakoff flow from the reactor coolant pumps and would present an inventory management issue since upstream relief valve, MU-V-180, will lift and pass seal return flow to the RC drain tank. This creates a significant risk of permanent damage to the reactor coolant pump

.seals, with the potential to cause subsequent Reactor Coolant system leakage.

Alternative Test These valves will be exercised closed (MU-V-25/26) and failsafe tested (MU-V-26) during cold shutdowns when the reactor coolant pumps are not in service and seal leakoff flow is not required. The valves will be partial stroke exercised quarterly during normal power operations.

Revision 0 October 25, 2013 A9 - 6

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATION CSJ-06 Valve Number System Safety Class Category MU-V-2A MU 1 A MU-V-2B MU 1 A MU-V-3 MU 2 A Function These normally open valves receive an ESAS signal to close automatically on HPI initiation at 1600 psig RCS pressure or 4 psig Reactor Building pressure in order to isolate the MU System letdown line. Valve MU-V-3 also fails closed on loss of electrical power or pneumatic supply.

The valves are open during normal Plant operation to provide letdown flow from the RCS to the MU System for chemical and radiological control. This function is not required for safe shutdown or accident mitigation.

Test Requirement Exercise test to the closed position in accordance with ISTC-351 0.

Justification Closing these valves for testing during normal power operations would interrupt letdown flow from the reactor coolant system.

It is not practical to full stroke exercise or fail safe test these valves closed during normal power operation since closure testing at normal power operations would cause: (1) a minor transient in pressurizer level control, and (2) thermal cycling of the letdown coolers. The letdown coolers are a unique "helicoil" design, and avoiding transients is necessary to prolong their life. Unnecessary cycling would lead to premature damage of the letdown coolers, whose repair would also become a significant ALARA concern.

Alternative Test These valves will be exercised closed (MU-V-2A/B and MU-V-3) and fail safe tested (MU-V-3) during cold shutdowns when the letdown coolers are not in service and RCS letdown flow may be isolated. The valves will be partial stroke exercised quarterly during normal power operations.

Revision 0 October 25, 2013 A9 - 7

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATION CSJ-07 Valve Number System Safety Class Categorv NS-V-15 NS 2 A Function This normally open motor operated valve closes on receipt of any ESAS signal coincident with a low level signal in Nuclear Services Closed Cooling Water Surge Tank (NS-T-1), or on a 30 psig Reactor Building pressure signal to isolate the NS supply header to the Reactor Coolant Pump Motor Air and Oil Coolers.

Cooling water to the components listed above is required during normal plant operation.

ESAS initiation circuitry is set up to keep this valve open during initial actuation, unless there is a low level Surge Tank signal concurrent with the ESAS signal, or until Reactor Building pressure reaches 30 psig. This open function allows for the continuous supply of cooling water to the RCP motors in the event that the pumps continue to run. This open function is not required for safe shutdown or accident mitigation.

Test Requirement Exercise test to the closed position in accordance with ISTC-3510.

Justification It is not practical to full stroke exercise this valve closed during normal power operation since closure testing at normal power operations would lead to eventual damage of the Reactor Coolant Pump Motor Air and Oil Coolers.

Closing this valve for testing during normal power operations would interrupt cooling water flow to the Reactor Coolant Pump Motor Air and Oil Coolers and would cause a brief temperature transient, and ultimately eventual damage, to the associated coolers.

Quarterly partial stroking of the valve verifies that it will perform its function to close.

Alternative Test This valve will be exercised closed during cold shutdowns when the Reactor Coolant Pump Motor Air and Oil Coolers are not in service and Nuclear Services cooling water flow may be isolated. The valve will be partial stroke exercised quarterly during normal power operations.

Revision 0 October 25, 2013 A9 - 8

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATION CSJ-08 Valve Number System Safety Class Category NS-V-35 NS .3 A NS-V-4 NS 3 A Function These normally open motor operated valves close automatically upon receipt of any ESAS signal coincident with a low level signal in Nuclear Services Closed Cooling Water Surge Tank NS-T-1, or on a 30 psig Reactor Building pressure signal to isolate the NS return header from the Reactor Coolant Pump Motor Air and Oil Coolers.

Cooling water to the components listed above is required during normal plant operation.

ESAS initiation circuitry is set up to keep this valve open during initial actuation, unless there is a low level Surge Tank signal concurrent with the ESAS signal, or until Reactor Building pressure reaches 30 psig. This open function allows for the continuous supply of cooling water to the RCP motors in the event that the pumps continue to run. This open function is not required for safe shutdown or accident mitigation.

Test Requirement Exercise test to the closed position in accordance with ISTC-3510.

Justification It is not practical to full stroke exercise these valves closed during normal power operation since closure testing at normal power operations would lead to eventual damage of the Reactor Coolant Pump Motor Air and Oil Coolers.

Closing these valves for testing during normal power operations would interrupt cooling water flow to the Reactor Coolant Pump Motor Air and Oil Coolers and would cause a brief temperature transient, and ultimately eventual damage, to the associated coolers.

Quarterly partial stroking of the valves verifies that they will perform their function to close.

Alternative Test These valves will be exercised closed during cold shutdowns when the Reactor Coolant Pump Motor Air and Oil Coolers are not in service and Nuclear Services cooling water flow may be isolated. The valves will be partial stroke exercised quarterly during normal power operations.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATION CSJ-09 Valve Number System Safety Class Category RB-V-2A RB 2 A RB-V-7 NS 2 A Function These normally open motor operated valves close on receipt of an ESAS Reactor Building isolation signal on high Reactor Building Pressure (4 psig) or low RCS pressure (1600 psig or 500 psig backup) to isolate the non-safety, non-seismically qualified normal Reactor Building Industrial Cooling Water System and allow for shifting to the Reactor Building Emergency Cooling Water System.

These valves are open to allow cooling water to be supplied to the non-safety related Reactor Building Industrial Cooling Coils during normal operation. This function is not required for safe shutdown or accident mitigation.

Test Requirement Exercise test to the closed position in accordance with ISTC-3510.

Justification It is not practical to full stroke exercise these valves closed during normal power operation since closure testing at normal power operations would isolate cooling to the Reactor Building normal cooling coils. Isolation of the RB cooling coils at any time of the year would challenge the plant's ability to maintain compliance with Reactor Building Temperature Limits (Tech Spec 3.17). Additionally, if either of these valves were to fail in the closed position during the cycling test, this would cause a loss of the system function and likely require a plant shutdown to avoid exceeding temperature limits within containment. This justifies deferral of testing to cold shutdown. Quarterly partial stroking of the valves verifies that they will perform their function to close.

Alternative Test These valves will be exercised closed during cold shutdowns when the Reactor Building Industrial Cooling Coils are not required to be in service and reactor building emergency cooling water may be isolated. The valves will be partial stroke exercised quarterly during normal power operations.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATION CSJ-10 Valve Number System Safety Class Category FW-V-5A FW N/A B FW-V-5B FW N/A B FW-V-16A FW N/A B FW-V-16B FW N/A B FW-V-17A FW N/A B FW-V-17B FW N/A B FW-V-92A FW N/A B FW-V-92B FW N/A B Function These normally open valves must close to isolate the OTSG's from the feedwater system in the event of a main steam line break. The valves are powered from an Engineered Safeguards Control Center (Class 1E) and receive an HSPS signal to close on Hi-Hi OTSG level or a Feedwater or Main Steam line break (low OTSG pressure). Downstream check valves FW-V-12A/B are relied upon for closure during a feedwater line break and for containment isolation.

The valves are open during normal power operation to provide a flow path from the feedwater system to the OTSG's. This function is not required for safe shutdown or accident mitigation.

Test Requirement Exercise test to the closed position in accordance with ISTC-3510.

Justification It is not practical to full stroke or partial stroke exercise these valves closed during normal power operation since closure of any of these valves would interrupt feedwater flow which would result in a plant transient due to a mismatch in feedwater/main steam line flow. This transient would result in a subsequent trip of the reactor and potentially injection of emergency feedwater.

Injection of emergency feedwater would result in an unacceptable thermal shock to the steam generators which could result in a degradation of the primary containment system boundary (OTSG tubes).

Alternative Test These valves will be exercised closed during cold shutdowns when the feedwater system. is not required to be in service.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATION CSJ-11 Valve Number System Safety Class Cateaory RC-V-28 RC 1 B RC-V-44 RC 1 B Function These normally closed valves are opened to vent steam and noncondensible gases from the pressurizer steam space to the reactor coolant drain tank. This function assures that the pressurizer is available for pressure and volume control. This venting capability supports core cooling for events beyond the design basis. The ability to vent post accident increases the ability of the plant to deal with large quantities of noncondensible gas which could interfere with natural circulation (i.e., core cooling). The valve is also used during an OTSG tube rupture to minimize sub cooling margin and during RCS superheat conditions.

These functions mitigate the consequences of a design basis SGTR accident.

When the required venting iscompleted, the valves are closed by remote manual operation to maintain the RCS pressure boundary and inventory. These valves do not receive any automatic actuation or isolation signals.

The individual vent path lines are sized such that an inadvertent opening of any valve will not constitute a LOCA. This design feature provides a high degree of assurance that the vents will be available when needed, and that inadvertent operation or failures will not significantly hamper the safe operation of the plant [TS 3.1.13].

Test Requirement Full-stroke exercise to the open and closed positions in accordance with ISTC-3510.

Justification It is not practical to full or partial stroke exercise or fail safe test these valves during normal power operation. Exercising either of these valves during power operations places the plant in an undesirable configuration by reducing the reactor coolant system isolation barrier to only a single valve. Exercising either valve during normal power operations leaves only the other valve as a single valve isolation barrier. During normal operation these valves are maintained closed at all times, they also fail closed.

Alternative Test These valves will be full stroke exercised closed (RC-V-28/44) and fail safe tested closed (RC-V-44) during cold shutdowns when the Pressurizer and Reactor Coolant system is depressurized below 200 psig.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATION CSJ-12 Valve Number System Safety Class Category RC-V-40A RC 1 B RC-V-40B RC 1 B RC-V-41A RC 1 B RC-V-41 B RC 1 B Function These normally closed solenoid operated valves are opened to vent steam and noncondensible gases from the reactor coolant system hot leg high points which cannot be vented through the reactor vessel or pressurizer steam space vents. These vents relieve to the reactor building atmosphere through a rupture disk. Technical Specification 3.1.13.1.c requires the RCS high point vent flow paths to be operable when the reactor is critical. This venting capability supports core cooling for events beyond the design basis. The ability to vent post accident increases the ability of the plant to deal with large quantities of noncondensible gas which could interfere with natural circulation (i.e., core cooling).

When the required venting is completed, the valves are closed by remote manual operation to maintain the RCS pressure boundary and inventory. These valves do not receive any automatic actuation or isolation signals however they fail closed on loss of electrical power.

The individual vent path lines are sized such that an inadvertent opening of any valve will not constitute a LOCA. This design feature provides a high degree of assurance that the vents will be available when needed, and that inadvertent operation or failures will not significantly hamper the safe operation of the plant [TS 3.1.13].

Test Requirement Full-stroke exercise to the open and closed positions in accordance with ISTC-3510.

Justification It is not practical to full or partial stroke exercise or fail safe test closed these valves during normal power operation. Exercising either of these valves during power operations places the plant in an undesirable configuration by reducing the reactor coolant system isolation barrier to only a single valve in the respective path. Exercising either valve during normal power operations leaves only the other valve as a single valve isolation barrier. During normal operation these valves are maintained closed at all times, they also fail closed.

Alternative Test These valves will be full stroke exercised to the open and closed positions and fail-safe tested closed during cold shutdowns when the Reactor Coolant system is depressurized below 200 psig.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATION CSJ-13 Valve Number System Safety Class Category RC-V-42 RC 1 B RC-V-43 RC 1 B Function These normally closed solenoid operated valves are opened to vent steam and noncondensible gases from the reactor vessel head and hot and cold legs to the reactor building. Technical Specification 3.1.13.1.a requires the associated reactor vessel head vent (RC-V-42/43) flow path to be operable when the reactor is critical. This venting capability supports core cooling for events beyond the design basis. The ability to vent post accident increases the ability of the plant to deal with large quantities of noncondensible gas which could interfere with natural circulation (i.e., core cooling).

When the required venting is completed, the valves are closed by remote manual operation to maintain the RCS pressure boundary and inventory. These valves do not receive any automatic actuation or isolation signals however they fail closed on loss of electrical power.

The individual vent path lines are sized such that an inadvertent opening of any valve will not constitute a LOCA. This design feature provides a high degree of assurance that the vents will be available when needed, and that inadvertent operation or failures will not significantly hamper the safe operation of the plant [TS 3.1.13].

Test Requirement Full-stroke exercise to the open and closed positions in accordance with ISTC-3510.

Justification It is not practical to full or partial stroke exercise or fail safe test closed these valves during normal power operation. Exercising either of these valves during power operations places the plant in an undesirable configuration by reducing the reactor coolant system isolation barrier to only a single valve. Exercising either valve during normal power operations leaves only the other valve as a single valve isolation barrier. During normal operation these valves are maintained closed at all times, they also fail closed.

Alternative Test These valves will be full stroke exercised to the open and closed positions and fail-safe tested closed during cold shutdowns when the Reactor Coolant system is depressurized below 200 psig.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATION CSJ-14 Valve Number System Safety Class Categorv DH-V-166A DH 2 C DH-V-166B DH 2 C Function These check valves must open to allow pressure between the seats and portions of piping of certain valves to equalize with downstream (RCS) pressure in order to prevent pressure-locking of the boundary valves allowing them to be opened when required.

The valves close to prevent back flow of RCS fluid into the area between the seats and portions of piping to prevent pressure-locking of the respective valves. However, the ability of the check valves to close is not required to achieve or maintain safe shutdown or for accident mitigation.

Test Requirement:

Exercise test to the open and closed positions in accordance with ISTC-3522.

Justification It is not practical to full flow open exercise these valves during normal power operation since the open test requires the DH system in service and also requires partially opening isolation valves DH-V-4AIB. Opening the isolation valves (DH-V-4A/B), with the DH pumps in service during normal power operation, could disturb the downstream check valves and cause over pressurization of the lower pressure DH system, subjecting the system to pressures in excess of its design pressure.

Exercising these valves closed requires a hydro pump to pressurize downstream of the check valves to verify the bi-directional closure capability of the subject valves. This test is essentially a leakage rate test which requires significant test setup. This justifies deferral of testing to cold shutdown.

Alternative Test These valves will be exercised open and verified to close during cold shutdowns when the reactor coolant system is depressurized and the DH system can be placed inservice. As permitted by ASME OM Code ISTC-3522(a) both the open and closed check valve exercise tests will be performed during cold shutdowns when it is practicable to perform both tests.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATION CSJ-15 Valve Number System Safety Class Cateciorv CA-V-1 77 CA 3 C Function This valve performs a safety function in the open position to provide a flow path for alternate emergency boration of the RCS from the Boric Acid Mix Tank, CA-T-1, via Boric Acid Injection Pump, CA-P-1A or CA-P-1B, to the Makeup (MU) System. In the open position, this valve also provides a flow path for addition of boron to the Reactor Coolant System for reactivity control during normal operation.

This valve performs no safety function in the closed position. In the closed position, this valve isolates the non-nuclear, Seismic II portion of the CA System from the ASME Class 3, Seismic I portion. However, downstream manual isolation valve MU-V-51 also performs the same function. It is normally closed isolating this flow path when not in use.

Test Requirement Exercise test to the open position in accordance with ISTC-3522.

Justification It is not practical to full stroke exercise or partial stroke exercise this valve open during normal power operation since this results in injection of concentrated boric acid into the RCS. Opening the valve during normal power operation requires running a boric acid pump to put flow through the valve. This evolution adds concentrated boric acid from Boric Acid Mix Tank CA-T-1 to Makeup Tank MU-T-1, the suction source of the running makeup pump. Changing MU tank boron concentration during normal power operation in this manner is not practical due to the negative reactivity added to the RCS and the significant effect it will have on RCS reactivity control. This justifies deferral of testing to cold shutdown.

Alternative Test This valve will be exercised open during cold shutdowns when the addition of concentrated boric acid to the reactor coolant system is not a concern.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATION CSJ-16 Valve Number System Safety Class Category MS-V-1A MS 2 B/C MS-V-1B MS 2 B/C MS-V-1C MS 2 B/C MS-V-1 D MS 2 B/C Function These normally open motor operated stop check valves receive no automatic isolation signal in order to allow for optimum RCS heat removal during most accidents and non-accident reactor trips. These valves are required to be closed from the Control Room following a large-break LOCA, small-break LOCA, Main Steam line break, Steam Generator tube rupture, or if a loss of OTSG integrity results in Containment pressure above OTSG pressure. In these cases, the valve would be closed by the motor operator, resulting in the Category B classification. These valves function as check valves in closing to prevent blowdown of both OTSG's in the event of a Main Steam line break upstream.

Therefore, they are also classified as Category C.

Although the open position is preferred under most accident and normal shutdown conditions (and is required for Plant operation), these valves do not have any safety function in the open position. These valves are considered Containment Isolation Valves:

however, they are exempt from Type C testing since they are connected to a closed loop inside containment (OTSG).

Test Requirement Exercise test to the closed position in accordance with ISTC-3510.

Justification It is not practical to full stroke exercise or partial stroke exercise these valves closed during normal power operation (see TSCR 246) since closure testing at normal power operations could cause a plant trip.

Closing these valves for testing during normal power operations would interrupt steam flow from the steam generator to the main steam/turbine systems. Exercising these valves closed would isolate the steam generator which would result in a severe power transient in the steam and reactor coolant systems which would lead to a subsequent trip of the plant.

To perform this testing quarterly would create a condition with a high potential to cause a reactor trip (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to cold shutdowns.

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Three Mile Island Unit #1 Inservice Testing Program Plan Alternative Test These valves will be exercised closed (Category B) and stroke timed during cold shutdowns (see TSCR 246) when the main steam system is not required to be in operation.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATION CSJ-17 Valve Number System Safety Class Categorv.

RC-V-3 RC 1 B .

Function This valve provides a backup means of securing flow in the event that the spray valve should stick open. This valve also provides a backup means of post LOCA boron precipitation control for the normal method using the decay heat drop line.

Test Requirement Exercise test to the closed position in accordance with ISTC-3510.

Justification It is not practical to full stroke exercise this valve during power operation because closing this valve isolates the minimum flow used to prevent thermal shock to the piping nozzle.

This justifies deferral of testing to cold shutdown.

Alternative Test This valve will be exercised closed during cold shutdowns when there is not a concern of thermal shock to the piping nozzle.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 9 COLD SHUTDOWN JUSTIFICATION CSJ-18 Valve Number System Safety Class Cateciorv RC-V-4 RC 1 A Function This valve is opened by remote manual operation to initiate auxiliary spray to provide further pressure reduction or complete depressurization of the RCS after the OTSG's and RC Pumps have been removed from service. The auxiliary spray system cannot be placed in service until the reactor coolant system is below 200 psig. Auxiliary spray is an alternate method to DH drop-line flow to prevent boron precipitation post-LOCA. It is a backup to the drop leg method, which is not single-failure proof.

In the closed position, this valve is one of two valves in series that isolate the RCS from the lower design pressure Decay Heat Removal System. The valve must close / remain closed to isolate the reactor coolant system from the DH System. The valve is a Reactor Coolant Pressure Boundary (RCPB) Pressure Isolation Valve (PIV).

Test Requirement:

Exercise test to the open and closed positions in accordance with ISTC-351 0.

Justification It is not practical to full stroke exercise this valve open and closed during normal power operation since injection into the RCS through this line is not possible with the RCS pressurized, and opening of RC-V-4 during operation would leave only a single check valve as the isolation protecting low-pressure injection piping from RCS pressure. The Decay Heat discharge pressure is approximately 200 psig while the normal RCS pressure is greater than 2100 psig. RC-V-4 is procedurally required to be closed when the RCS is

>400 psig. To perform this testing quarterly would require major plant or hardware modifications (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to cold shutdown.

Alternative Test This valve will be exercised open and closed during cold shutdowns when the reactor coolant system is depressurized.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 10 REFUELING OUTAGE JUSTIFICATION INDEX Revision 0 October 25, 2013 A10 - 1

Three Mile Island Unit #1

• Inservice Testing Program Plan ATTACHMENT 10 REFUEL OUTAGE JUSTIFICATION INDEX Designator Rev Title/Description RJ-01 DH Pump Discharge Check Valves (DH-V-16A/B) Exercise Open RJ-02 DELETED RJ-03 DELETED (Replaced by Cold Shutdown Justification CSJ-16)

RJ-04 DH to CFT Check Valves (DH-V-22A/B) Exercise Open/Closed RJ-05 EF Pump Discharge Check Valves (EF-V-1i1A/B) and EF to OTSG Check Valves (EF-V-1 2A/B) Exercise Open/Closed RJ-06 IA Check Valves (IA-V-1 628A/B and 1631A/B) Exercise Open/Closed RJ-07 EF Pump 1 Discharge Check Valve (EF-V-13) Exercise Open/Closed RJ-08 2-Hour Backup Air Valves (IA-V-1625A/B, 1626A/B) Exercise and Fail Open RJ-09 DELETED RJ-10 MS Isolation Check Valves (MS-V-1A-D) Exercise Closed RJ-11 RB Emergency Cooling Water Valves (RR-9A/B/C) Exercise Open/Closed RJ-12 MU Tank Outlet Check Valve (MU-V-1 12) Exercise Closed RJ-13 MU Pump BWST Check Valves (MU-V-14A/B) Exercise Open/Closed RJ-14 RB Emergency Cooling Water Valves (RR-8A/B) Exercise Open RJ-15 MU Pump Discharge Check Valves (MU-V-73A-C) Exercise Open/Closed RJ-16 DELETED RJ-17 DELETED RJ-18 RB Emergency Cooling Water Valves (RR-7A/B) Exercise Open/Closed RJ-19 DELETED RJ-20 DH Pump from BWST Check Valve (DH-V-14A/B) Exercise Open/Closed RJ-21 RB Sump to DH Pump Suction Valves (DH-V-6A/B) Exercise Open/Closed RJ-22 DH Equalizing Check Valves (DH-V-170/171,172) Exercise Open/Closed RJ-23 DH Pressurizer Spray Line Check Valve (RC-V-23) Exercise Open Revision 0 October 25, 2013 A10-2

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 10 REFUEL OUTAGE JUSTIFICATION INDEX Designator Rev Title/Description RJ-24 Spent Fuel Return Check Valve (DH-V-50) Exercise Open/Closed RJ-25 CST De-Icing Supply Check Valves (CO-V-25A/B) Exercise Open/Closed RJ-26 DELETED RJ-27 MU Pump Minimum Flow Valves (MU-V-193A/B/C) Exercise Open I Closed RJ-28 Reactor Vessel Internal Vent Valves (RC-V-144A thru 144H) Exercise Open Revision 0 October 25, 2013 A10 -3

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATIONS Revision 0 October 25, 2013 All -1

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-01 Valve Number System Safety Class Cateaory DH-V-16A DH 2 C DH-V-16B DH 2 C Function These check valves are required to open to permit flow from the Decay Heat Removal Pumps to the reactor coolant system or makeup pumps suction when required during accident conditions.

In the event of a Core Flood System line break, DH Pump discharge cross-connect valves DH-V-38A and DH-V-38B would need to be opened to provide balanced flow to the Core.

If the corresponding DH pump fails to start, the associated check valve must close in order to prevent bypassing flow back to the Reactor Building Sump through the idle train.

Test Requirement:

Exercise test to the open and closed positions in accordance with ISTC-3522.

Justification It is not practical to full stroke exercise these valves open during operations or during cold

.shutdown. The open exercise requires a full flow test. Performance of a full stroke forward flow test is not possible during normal operations since the decay heat removal pumps cannot overcome reactor coolant system pressure. A part stroke open test is performed as part of the pump quarterly operability test using the recirculation line to the RWST. A forward flow test during cold shutdowns is not always possible since the reactor coolant system pressure may still be at a level which prevents exercising the valves full open.

Additionally, the injection of DH system flow into the reactor coolant system during normal operations or cold shutdowns would effect boron concentration and reactivity.

It is not practical to exercise these valves closed during normal power operations. The closure test of these valves requires a reverse flow or leakage test. A reverse flow test can only be performed when the cross tie isolation valves DH-V-38A/B are both open and one pump is in operation. The corresponding idle pump discharge check valve can then be reverse flow tested. This test can only be performed during cold shutdown periods when the cross tie isolation valve may be opened. These valves are maintained closed during normal power operation to maintain train separation.

Alternative Test These check valves will be exercised open and closed during refueling outages. The valves are verified to open with full flow during refueling outages when the DH pump can deliver the required flow rates into the reactor coolant system and closed when the cross-tie isolation valves may be opened. As permitted by ASME OM Code ISTC-3522(a) both Revision 0 October 25, 2013 All -2

Three Mile Island Unit #1 Inservice Testing Program Plan the open and closed check valve tests will be performed during refueling outages when it is practicable to perform both tests.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ DELETED Valve Number System Safety Class Category CA-V-448A CA 2 C CA-V-448B CA 2 C Function The steam generator blowdown/sampling modification, ECR TM 07-00259, installed separate relief valves on the 2 penetrations, thereby eliminating the need for these check valves. Therefore, CA-V-448A/B have been removed from the plant and from the IST and 10CFR50 Appendix J programs.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-03 Valve Number System Safety Class Cate-gory MS-V-1A MS 2 B/C MS-V-1 B MS 2 B/C MS-V-l C MS 2 B/C MS-V-1 D MS 2 B/C This Refueling Outage Justification has been replaced by Cold Shutdown Justification CSJ-16.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-04 Valve Number System Safety Class Category CF-V-5A CF 1 A/C CF-V-5B CF 1 A/C DH-V-22A DH 1 A/C DH-V-22B DH 1 A/C Function These check valves are Reactor Coolant Pressure Boundary Pressure Isolation Valves (RCPB PIV). In the closed position, the valves prevent leakage from the high pressure RCS to the lower pressure rated Core Flood and Decay Heat Removal Systems. Two of the valves are considered containment isolation valves; however, they are exempt from Appendix J Type C testing since no potential atmospheric leakage path from containment exists.

The valves are required to open to admit borated water from the discharge of the Core Flood Tanks or the Decay Heat Removal Pumps to the Reactor Coolant System during the injection and recirculation phases following an intermediate to large-break LOCA.

Test Requirement:

Exercise test to the open and closed positions in accordance with ISTC-3522.

Justification It is not practical to exercise these valves during normal power operation or during cold shutdown conditions. Exercising these valves requires the performance of a leakage test to verify the closed position and forward flow test to verify the open position.

Performance of a forward flow test is not possible during normal operations since the core flood tanks and the decay heat removal pumps cannot overcome reactor coolant system pressure. This forward flow test during cold shutdowns is not always possible since the reactor coolant system pressure may still be at a level which prevents exercising the valves open. Additionally, injection into the reactor coolant system during normal operations or cold shutdowns would effect boron concentration and reactivity.

To perform a leakage test to verify closure during normal operations or cold shutdowns, requires temporary test equipment to be installed inside containment to establish a differential pressure across the valve to verify closure. This test is not practical to be performed during normal power operations or during cold shutdowns since the DH system would be required to be out of service. Additionally, containment entry to perform the necessary test setup, performance and restoration would cause an increase in personnel radiation exposure. To perform this testing quarterly would require major plant or hardware Revision 0 October 25, 2013 All -6

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-04 (Cont.)

modifications (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to refueling outages.

Alternative Test These check valves will be exercised closed during refueling outages in conjunction with their respective Pressure Isolation Valve seat leakage test. This test is performed when the CF and DH systems are not required to be in service, the reactor coolant system is depressurized, and containment entry is possible. The valves are verified to open with full flow during refueling outages when the CF tanks and the DH pumps can deliver the required flow rates into the reactor coolant system.

Revision 0 October 25, 2013 All -7

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-05 Valve Number System Safety Class Cate-gory EF-V-11A EF 3 C EF-V-11B EF 3 C EF-V-12A EF 2 C EF-V-12B EF 2 C Function These check valves open to permit the Emergency Feedwater Pump to deliver Emergency Feedwater to either or both OTSG's when Emergency Feedwater is required.

Valves EF-V-1 1A/B must close to prevent diversion of flow through an idle or faulted emergency feedwater pump. Valves EF-V-12A/B must close to prevent reverse flow thereby maintaining OTSG inventory.

Test Requirement:

Exercise test to the open and closed positions in accordance with ISTC-3522.

Justification It is not practical to exercise these valves open or closed during normal power operation or during cold shutdown conditions. Exercising these valves requires the performance of a leakage test to verify the closed position and forward flow test to verify the open position.

Performance of a forward flow test is not possible during normal operations since flow through the valves during power operation would inject cold water into a hot steam generator.. This is impractical because injecting cold water from the Condensate Storage Tank into the hot Steam Generator during operations would thermally cycle the tubes and Emergency Feedwater nozzles. Further, injection from the Condensate Storage Tank will introduce oxygenated water into the Steam Generators. The exposure of the Steam Generator tubes to oxygenated water, especially during short shutdowns, must be minimized.

To perform a leakage test to verify closure during normal operations or cold shutdowns, requires temporary test equipment to be installed to establish a differential pressure across the valve to verify closure. This closed test requires isolation of one injection line to the OTSG and the cross connecting of two of the three Emergency Feedwater Pumps. This effectively removes two pumps from service. This configuration is not desirable nor is it permitted by the Tech Specs when the plant is operating. During short duration or unplanned Cold Shutdowns, the test could extend the outage due to the amount of time necessary. for test setup, performance and restoration. To perform this testing quarterly would require. major plant or hardware modifications (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to refueling outages.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-05 (Cont.)

Alternative Test These check valves will be exercised open and closed during refueling outages when the EF pumps are able to deliver the required flow to the OTSG. The closure test performed by a differential pressure leakage test will be performed during refueling outages when the EF system and OTSG's are not required to be in service.

Revision 0 October 25, 2013 All -9

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-06

.Valve Number System Safety Class Category IA-V-1628A IA NC C IA-V-1628B IA NC C IA-V-1631A IA NC C IA-V-1631B IA NC C Function The IA-V-1628A/B check valves must close to isolate the non-safety related instrument air supply from the 2-Hour Backup Air system.

Check valve IA-V-1631A is required to open to provide air from the 2-Hour Backup Air system to be supplied to the essential equipment in the event of a loss of the Instrument Air Compressors. This valve closes to prevent backflow when the A train is out of service while the B train is required to supply PC-5 and MS-V-6.

Check valve IA-V-1 631 B is required to open to provide a flow path from the 2-Hour Backup Air system to MS-V-6 and the PC-5 Controller.

Test Requirement:

Exercise test to the open and closed positions in accordance with ISTC-3522.

Justification It is not practical to full stroke exercise these valves open or closed during normal power operation or during cold shutdowns since the closure test requires a backflow leakage test.

This testing would require rendering all of the associated essential components inoperable while performing the back flow leakage or respective forward flow tests. The required full flow and backflow leakage tests can only be performed during refueling outages when it is possible to declare the associated instrument air supplied components inoperable. To perform this testing quarterly would require major plant or hardware modifications (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to refueling outages.

Alternative Test These check valves will be exercised open and closed during refueling outages when the Instrument Air System is not required to be in service. As permitted by ASME OM Code ISTC-3522(a) both the open and closed check valve tests will be performed during refueling outages when it is practicable to perform both tests.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-07 Valve Number System Safety Class Categorv EF-V-13 EF 3 C Function This check valve opens to permit the turbine-driven Emergency Feedwater Pump EF-P1 to deliver Emergency Feedwater to either or both OTSG's when emergency feedwater is required to be in operation. The valve must open to provide a minimum of 350 gpm during design accident conditions.

This valve closes to prevent diversion of flow by recirculation through EF-P1 if idle or faulted.

Test Requirement:

Exercise test to the open and closed positions in accordance with ISTC-3522.

Justification It is not practical to exercise this valve open or closed during normal power operation or during cold shutdown conditions. Exercising the valve requires the performance of a pump full flow test to verify the open position.

Performance of a forward flow test is not possible during normal operations since flow through the valves during power operation would inject cold water into a hot steam generator. This is impractical because injecting cold water from the Condensate Storage Tank into the hot Steam Generator during operations would thermally cycle the tubes and Emergency Feedwater nozzles. Further, injection from the Condensate Storage Tank will introduce oxygenated water into the Steam Generators. The exposure of the Steam Generator tubes to oxygenated water, especially during short shutdowns, must be minimized.

The closed test requires that part of the EF system is inoperable and removes redundancy for providing Emergency Feedwater to each Steam Generator. This configuration is not permitted by the Technical Specifications when the plant is operating. To perform this testing quarterly would require major plant or hardware modifications (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to refueling outages.

Alternative Test This check valve will be exercised open during refueling outages when the EF pump is able to deliver the required flow to the OTSG. The closure test, performed by a differential pressure test, will be performed during refueling outages when the EF system and OTSG's are not required to be in service. As permitted by ASME OM Code ISTC-3522(a) both the open and closed check valve tests will be performed during refueling outages when it is practicable to perform both tests.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-08 Valve Number System Safety Class Category IA-V-1625A IA NC B IA-V-1625B IA NC B IA-V-1626A IA NC B IA-V-1 626B IA NC B Function IA-V-1625A/B These normally open air operated 3-way valves must remain open to provide a flow path from the 2-Hr Backup Supply Air Bottle Rack to the essential EF and MS air actuators and positioners. The valves are normally aligned in their safety-related position from the Train A 2-Hr Backup Supply Air Bottle Rack to the supply header to assure a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> supply of actuating air in the event of a loss of the IA Compressors. In the event of a loss of supply pressure, the valves fail to the vented position, causing the downstream valves to fail to their safe position.

IA-V-1626A/B These normally open 3-Way valves must open to provide a flow path from the 2-Hr Backup Supply Air Bottle Rack to the essential EF and MS air actuators and positioners. The valves are normally aligned from the discharge of the Instrument Air Compressors (normal supply) to the supply header for EF-V 030A, EF-V 030C, MS-V 004A, MS-V 006 and the PC-5 Controller. Upon a loss of IA System pressure, they shift to the safety-related supply source from the Train A 2-Hr Backup Supply Air Bottle Rack.

Test Requirement:

Exercise test to the open and closed positions in accordance with ISTC-351 0.

Justification It is not practical to exercise open or fail open these valves during normal power operations or during cold shutdowns. To exercise these valves requires the normal instrument air supply to be isolated to the Emergency Feedwater and Main Steam valves actuators and positioners thereby rendering them inoperable. The components supplied by the 2-Hr Backup Air Bottles are required to be operable during normal operations and during cold shutdowns when the EF and MS systems are still required to be inservice. To perform this testing quarterly would require major plant or hardware modifications (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to refueling outages.

Alternative Test These valves will be exercised and fail-safe tested open during refueling outages when the EF and MS systems are not required to be inservice and the 2-Hour backup air system is not required.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ DELETED Valve Number System Safety Class Cateaory MU-V-79 MU 2 Function This check valve has been classified as a passive valve based on upstream manual valve MU-V-78 being administratively maintained in the locked closed position and relied upon for isolation of the letdown/purification system. The valve has been determined to not perform a safety function and has been removed from the IST Program.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-10 Valve Number System Safety Class Category MS-V-1A MS 2 B/C MS-V-1 B MS 2 B/C MS-V-iC MS 2 B/C MS-V-1D MS 2 B/C Function These normally open motor operated stop check valves receive no automatic isolation signal in order to allow for optimum RCS heat removal during most accidents and non-accident reactor trips. These valves are required to be closed from the Control Room following a large-break LOCA, Small-break LOCA, Main Steam line break, or Steam Generator tube rupture, or if a loss of OTSG integrity results in Containment pressure above OTSG pressure. In these cases, the valve would be closed by the motor operator, resulting in the Category B classification. These valves function as check valves in closing to prevent blowdown of both OTSG's in the event of a Main Steam line break upstream.

Therefore, they are also classified as Category C.

Although the open position is preferred under most accident and normal shutdown conditions and is required for Plant operation, the valves do not have any safety function in the open position. These valves are considered Containment Isolation Valves; however, they are exempt from Type C testing since it is connected to a closed loop inside containment (OTSG).

Test Requirement:

Exercise test to the closed position in accordance with ISTC-3522.

Justification It is not practical to exercise these valves closed with flow (category C function) during normal power operation or during cold shutdowns.

Closing these valves for testing during normal power operations would interrupt steam flow from the steam generator to the main steam/turbine systems. Exercising these valves closed would isolate the steam generator which would result in a severe power transient in the steam and reactor coolant systems which would lead to a subsequent trip of the plant.

The closure test of the check valve function requires that one OTSG be depressurized while pressure on the other is monitored. Effectively, the closed function is verified by a differential pressure test to verify closure. This test can only be done during refueling outages when the steam generators may be isolated. To perform this testing quarterly would create a condition with a high potential to cause a reactor trip (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to refueling outages.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-10 (Cont.)

Alternative Test These valves will be exercised closed (Category C) during refueling outages when the main steam system is not required to be in operation. The open (Category C) check valve function is verified during normal operations with system flow.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-11 Valve Number. System Safety Class Category RR-V-9A RR 3 C RR-V-9B RR 3 C RR-V-9C RR 3 C Function These check valves must open to provide a return flow path to the River from the Emergency RB Cooling Coil for post-accident cooling and pressure control of the Reactor Building.

The valves close to prevent reverse flow through a faulted cooling coil, however downstream motor operated valves RR-V-4A-D are relied upon for isolation of the cooling coil. Therefore this closed function is not required for safe shutdown or accident mitigation.

Test Requirement:

Exercise test to the open position in accordance with ISTC-3522.

Justification It is not practical to exercise these check valves open or closed during normal power operation or during cold shutdowns. Exercising the valves open requires the performance of a full flow test to verify the open position. Exercising the valves closed requires a leakage or reverse flow test.

Performance of a full flow test through these valves requires the RBEC pump to introduce river water into the Reactor Building Emergency Cooling Coils. Since river water contains silt and microorganisms, the system must be drained to the Reactor Building sump and refilled with Nuclear Services Closed Cooling Water. Approximately 5,000 gallons of water must be processed through the Liquid Waste Disposal System. This is not practical for a quarterly or cold shutdown frequency. Additionally, test performance will discharge a quantity of corrosion inhibitor to the river, an environmental release that should be minimized.

To perform a leakage test or reverse flow test to verify closure during normal operations or during cold shutdown, requires temporary test equipment to be setup to establish a differential pressure across the valve to verify closure. This test is not practical to be performed during normal power operations or during cold shutdowns since the RB Emergency Cooling Water system would be required to be vented and out of service.

Additionally, the necessary test setup, performance and restoration could cause an increase in personnel radiation exposure and delay startup from a cold shutdown condition.

To perform this testing quarterly would require major plant or hardware modifications (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to refueling outages.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-11 (Cont.)

Alternative Test These check valves will be exercised open and verified to close (bi-directional test) during refueling outages when full forward and reverse flow testing can be performed and the reactor building emergency cooling coils are not required to be in service.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-12 Valve Number System Safety Class Category MU-V-112 MU 2 A/C Function This check valve is required to close to prevent pumping post-accident liquid from the RB Sump into the Makeup Tank when operating the DH and MU Systems in the "piggyback" mode. This valve is located in the outlet line of Makeup Tank MU-T-1 and forms a boundary. between the portion of the MU System which is required for safe shutdown/accident mitigation and the non-safety related makeup tank.

This valve opens to allow makeup flow from Makeup Tank MU-T-1 to the suction of the operating Makeup Pump (normally MU-P-1B) during normal operation. It also opens to allow flow from Makeup Tank MU-T-1 to the suction of the operating Makeup Pump to support alternate emergency boration requirements. Downstream motor-operated isolation valve, MU-V12, does not receive any signal to close in the event of an accident.

Test Requirement:

Exercise test to the closed position in accordance with ISTC-3522.

Justification It is not practical to exercise this valve closed during normal power operation or during cold shutdowns. Exercising the valve requires the performance of a leakage or reverse flow test to verify the closed position.

To perform a leakage test or reverse flow test to verify closure during normal operations or cold shutdowns, requires temporary test equipment to be installed to establish a differential

,pressure across the valve to verify closure. This test is not practical to be performed during normal power operations or during cold shutdowns since the MU system and DH system would be required to be vented and out of service to perform a leakage test. To perform a reverse flow test using the DH system to pressurize the valve would require removing both the MU and DH system from service. Removing both systems is not practical during normal power operations or during cold shutdown conditions. To perform this testing quarterly would require major plant or hardware modifications (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to refueling outages.

Alternative Test This valve will be exercised closed during refueling outages when the MU and DH systems are not required to be in service. The open direction is verified using normal system flow during normal power operations.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-13.

Valve Number - System Safety Class Category MU-V-14A MU 2 A/C MU-V-14B MU 2 A/C Function These motor operated stop check valves are required to open on an ESAS actuation signal of 1600 psig RCS pressure or 4 psig Reactor Building pressure to provide suction from the BWST to the MU Pumps for HPI initiation. An ESAS backup initiation signal is also provided at 500 psig RCS pressure.

The valves are required to close following the post-LOCA injection phase to isolate the BWST during the recirculation phase. In the "piggyback" mode, this is the only valve that separates the BWST from the discharge of the DH Pumps.

Test Requirement:

Exercise test to the open and closed positions in accordance with ISTC-3522.

Justification It is not practical to exercise these valves open or closed during normal power operation or during cold shutdowns. Exercising the valves requires the performance of a leakage or reverse flow test to verify the closed position and full flow test to verify the open position.

To perform a leakage test or reverse flow test to verify closure during normal operations or cold shutdowns, requires temporary test equipment to be installed to establish a differential pressure across the valve to verify closure. This test is not practical to be performed during normal power operations or during cold shutdowns since the test lineup requires isolation of the Makeup Tank and observation of level decrease over time. This procedure can significantly lengthen an unplanned Cold Shutdown. To verify the full open position, a full flow test using the BWST as a suction source is required. Injection of highly borated water from the BWST into the reactor coolant system is not practical during normal operations or during cold shutdown due the significant effect it will have on reactivity and RCS inventory control. Additionally, performance of this full flow test may lengthen the time to reach criticality. The delay would occur because the boron concentration would. need to be diluted by water from a Reactor Coolant Bleed Tank. To perform this testing quarterly would require major plant or hardware modifications (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to refueling outages.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-13 (Cont.1 Alternative Test These valves will be exercised closed during refueling outages in conjunction with the respective leakage test when the MU system is not required to be in service. The valves will be exercised open during refueling outages with flow by injecting BWST water into the reactor coolant system.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-14 Valve Number System Safety Class Catenorv RR-V-8A RR 3 C RR-V-8B RR 3 C Function These check valves must open to provide flow from the RBEC Pumps to the Reactor Building Emergency Cooling Coils for post-accident cooling and pressure control. The valves may also be required to open to provide a safety-related, seismically-qualified backup suction source to the Emergency Feedwater Pumps.

These valves are required to close to prevent diversion of flow in the event of a failure of the RBEC Pump to start or continue running or in the event of an upstream pipe break.

Test Requirement:

Exercise test to the open position in accordance with ISTC-3522.

Justification It is not practical to exercise these check valves open during normal power operation or during cold shutdowns. Exercising the valves open requires the performance of a full flow test to verify the open position.

Performance of a full flow test through these valves requires the RBEC pump to introduce river water into the Reactor Building Emergency Cooling Coils. Since river water contains silt and microorganisms, the system must be drained to the Reactor Building sump and refilled with Nuclear Services Closed Cooling Water. Approximately 5,000 gallons of water must be processed through the Liquid Waste Disposal System. This is not practical for a quarterly or cold shutdown frequency. Additionally, test performance will discharge a quantity of corrosion inhibitor to the river, an environmental release that should be minimized. To perform this testing quarterly would require major plant or hardware modifications (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to refueling outages.

Alternative Test These check valves will be exercised open during refueling outages when full flow testing can be performed and the reactor building emergency cooling coils are not required to be in service. The closure test is performed quarterly using reverse flow from the Nuclear Services Closed Cooling Water system.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-15 Valve Number System Safety Class Category MU-V-73A MU 2 C MU-V-73B MU 2 C MU-V-73C MU 2 C Function In the open position, these check valves allow required HPI flow to be supplied by the Makeup Pumps to the Reactor Coolant System (RCS) in response to an ESAS initiation at 1600 psig RCS pressure or 4 psig Reactor Building pressure, or by a backup signal at 500 psig RCS pressure.

In the closed position, the valves prevent diversion of flow by recirculation through the pump in the event that the pump fails to start or is secured by the operator during the later stages of the accident or for post-accident recirculation (i.e., "piggyback") operation.

Test Requirement:

Exercise test to the open and closed positions in accordance with ISTC-3522.

Justification It is not practical to exercise these check valves open during normal power operation or during cold shutdowns. Exercising. the valves open requires the performance of a full flow test to verify the open position.

Full flow testing of these valves requires injection into the RCS. During normal power operations or during cold shutdowns the injection flow into the RCS is limited to avoid pressure transients in the RCS. Attaining full flow requires that additional injection valves be opened. Tech Spec 3.1.12.3 will not permit opening of the injection valves (MU-V16A, B, C, D) when RCS temperature is below 275 degrees F with the Reactor Head in place.

This means testing at Cold Shutdown is not possible and that the test must be performed as the plant shuts down or restarts. To perform this testing quarterly would require major plant or hardware modifications (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to refueling outages.

Due to the design of the MU pumps (9 stage high pressure centrifugal), positive verification of closure of the discharge check valves cannot be ascertained during quarterly testing of the pumps on minimum recirculation flow.

Alternative Test These check valves will be exercised open and closed during refueling outages when injection to the reactor coolant system is possible and the system alignment allows for positive verification that the valves are exercised closed by reverse flow.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ DELETED Valve Number System Safety Class Cateaorv NR-V-20A NR 3 C NR-V-20B NR 3 C NR-V-20C NR 3 C Deleted this justification after revising the quarterly testing in 2008 to use a flow that satisfies design basis requirements.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ DELETED Valve Number System Safety Class .Categiory RC-V-2 RC 1 .B This Refueling Outage Justification has been determined to not be necessary and has been deleted.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-18 Valve Number System Safety Class Category RR-V-7A RR 3 C RR-V-7B RR 3 C Function These check valves must open to provide a flow path from the RBEC Pumps to the Reactor Building Emergency Cooling Coils for post-accident cooling and pressure control. The valves may also be required to open to provide a safety-related, seismically-qualified backup suction source to the Emergency Feedwater Pumps.

These valves close to prevent reverse flow through a faulted or idle RBEC pump, however, downstream valves RR-V-8A/B and RR-V-lA/B are relied upon for this function. Therefore the closed function is not required for safe shutdown or accident mitigation.

Test Requirement:

Exercise test to the open position in accordance with ISTC-3522.

Justification It is not practical to exercise these check valves open during normal power operation or during cold shutdowns. Exercising the valves open requires the performance of a full flow test to verify the open position.

Performance of a full flow test through these valves requires the RBEC pump to introduce river water into the Reactor Building Emergency Cooling Coils. Since river water contains silt and microorganisms, the system must be drained to the Reactor Building sump and refilled with Nuclear Services Closed Cooling Water. Approximately 5,000 gallons of water must be processed through the Liquid Waste Disposal System. This is not practical for a quarterly or cold shutdown frequency. Additionally, test performance will discharge a quantity of corrosion inhibitor to the river, an environmental release that should be minimized. To perform this testing quarterly would place undue stress on components and cause unnecessary cycling of equipment (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to refueling outages.

Alternative Test These check valves will be exercised open during refueling outages when full flow testing can be performed and the reactor building emergency cooling coils are not required to be in service. Bi-directional closure testing is performed each refueling using non-intrusive equipment (ultrasonic instrumentation) to verify water in the piping on the downstream side of the valve.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ DELETED Valve Number System Safety Class Category.

RC-RV-2 RC 1 B/C This Refueling Outage Justification has been determined to not be necessary and has been deleted.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-20 Valve Number System Safety Class Category DH-V-14A DH 2 A/C DH-V-14B DH 2 A/C Function These check valves open to permit the Decay Heat Removal Pumps to take suction from the BWST.

The valves are required to close to prevent potential flow from the Reactor Building Sump into the BWST in the event that DH-V 005A/B fail to close following an accident while the Reactor Building is under pressure.

Test Requirement:

Exercise test to the open and closed positions in accordance with ISTC-3522.

Justification It is not practical to exercise these valves during normal power operations or cold shutdown when DH system is required to be in service.

To exercise the valves open requires full flow injection from the BWST into the reactor coolant system. This test is not practical to perform during plant operation or during cold shutdown since full flow injection of borated water into the reactor coolant system would result in severe reactor power oscillations and subsequent reactor trip during plant operations. During cold shutdown, this test would result in delaying unit startup due to the amount time necessary to clean up the DH and RC systems along with refilling the BWST.

To exercise the valve closed would require a leakage test. This test would require one train of DH suction piping to be isolated and vented. The test would render the associated train of the DH system inoperable during plant operations and during cold shutdown periods. To perform this testing quarterly would require major plant or hardware modifications (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to refueling outages.

Alternative Test These valves will be exercised open and closed during refueling outages when the DH and RC systems are not required to be inservice to allow for full flow injection and leakage testing of the subject valves.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-21 Valve Number System Safety Class Category DH-V-6A DH 2 B DH-V-6B DH 2 B Function These valves must be capable of closure to provide long term Containment isolation following an accident. However, since the Decay Heat Removal (DH) System is filled with water and is a closed system outside Containment, it is very improbable that leakage would occur through these valves. Furthermore, the valves will be opened and may remain open indefinitely following an accident. The valves are exempt from Type C leakage testing since they are in penetrations which are continuously submerged in water during accident conditions (water-sealed).

The valves must be opened to shift suction of the DH Pumps from the BWST to the Reactor Building Sump when the BWST inventory is depleted following a LOCA in order to initiate the post-LOCA recirculation mode. The valves are closed by remote manual operation.

Test Requirement:

Exercise test to the open and closed positions in accordance with ISTC-3510.

Justification It is not practical to exercise these valves during normal power operations or cold shutdown.,

Exercising these valves. would require. blanking the reactor building ECCS sump suction piping and would render the associated DH and Building Spray trains inoperable. This testing would require containment entry to install blank flanges on the DH system suction piping to perform the necessary testing. This would render the DH and BS systems inoperable during normal operations and potentially delay unit startup from a cold shutdown to perform the necessary blank installation and removal.. To perform this testing quarterly would require major plant or hardware modifications (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to refueling outages.

Alternative Test These valves will be exercised open and closed during refueling outages when the DH and RC systems are not required. to be inservice and the reactor building sump level is such that radioactive fluid is not introduced into the DH system suction piping.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION R.J-22 Valve Number System Safety Class Category DH-V-170 DH 2 C DH-V-171 DH 2 C DH-V-172 DH 2 C Function These check valves must open to allow pressure between the seats and portions of piping of certain valves to equalize with downstream (RCS) pressure in order to prevent pressure-locking of the boundary valves allowing them to be opened when required.

The valves close to prevent back flow of RCS fluid into the area between the seats and portions of piping to prevent pressure-locking of the respective valves. However, the ability of the check valves to close is not required to achieve or maintain safe shutdown or for accident mitigation.

Test Requirement:

Exercise test to the open and closed positions in accordance with ISTC-3522.

Justification It is not practical to exercise these valves open or closed during normal power operations or cold shutdown.

Exercising these valves closed. requires venting of reactor coolant system fluid to perform backflow testing to verify closure capability of the subject check valves. This test is essentially a leakage rate test which would require. isolating and venting of the associated portions of the reactor coolant system.. Additionally, increased personnel radiation exposure would be realized due to the location of these valves inside the reactor building.

The full flow open test of these check valves is also not practical to perform during normal power operations due to the need to partially open the associated main process line isolation valve. Opening either of these valves (DH-V-1, DH-V-2) during normal power operation could cause over pressurization of the lower pressure DH system. To perform this testing quarterly would require major plant or hardware modifications (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to refueling outages.

Alternative Test These valves will be exercised open and verified to close during refueling outages when the DH and RC systems are not required to be inservice. As permitted by ASME OM Code ISTC-3522(a) both the open and. closed check valve exercise tests will be performed during refueling outages when it is practicable to perform both tests.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-23 Valve Number System Safety Class Category RC-V-23 RC 1 A/C Function This check valve opens on initiation of flow from the DH System to allow circulation of water through the auxiliary spray line to prevent the concentration of boron in the Core from exceeding its solubility limit. The auxiliary spray system cannot be placed in service until the reactor coolant system is below 200 psig. Auxiliary spray is an alternate method to DH drop-line flow to prevent boron precipitation post-LOCA. It is a backup to the drop leg method, which is not single-failure proof.

In the closed position, this valve is one of two valves in series that isolate the RCS from the lower design pressure Decay Heat Removal System. This valve would remain closed as long as RCS pressure is higher than DH System pressure or would close for any condition which caused flow to reverse. The valve is a Reactor Coolant Pressure Boundary (RCPB)

Pressure Isolation Valve (PIV).

Test Requirement:

Exercise test to the open and closed positions in accordance with ISTC-3522.

Justification It is not practical to full stroke exercise this valve open or closed during normal power operation since injection in to the RCS through this line is not possible with the RCS pressurized. The Decay Heat discharge pressure is approximately 200 psig while the normal RCS pressure is greater than 2100 psig. The closure test of this check valve is performed by using a back flow leakage test to verify disk closure. This test requires entry in to the reactor building to perform the necessary test setup and restoration. It is not practical to perform this leakage test during normal power operations or during cold shutdowns. Entry in to the reactor building is not possible during normal power operations due to the elevated dose. Performance of a leakage test during cold shutdowns is not practical since a significant amount of piping would be required to be vented along with an increase to personnel radiation exposure. To perform this testing quarterly would require major plant or hardware modifications (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to refueling outages.

Alternative Test This valve will be exercised open and closed during refueling outages when the reactor coolant system is depressurized and radiation levels are reduced.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-24 Valve Number System Safety Class Catenory DH-V-50 DH 2 A/C Function Closure of this valve is important to ECCS bypass leakage. This valve is a potential leakage path from the reactor building sump or associated system when contaminated water is circulated following a LOCA [C-1101-210-E270-014].

This valve closes to prevent diversion of flow from the Decay Heat Removal Pump DH-P-1A suction in the event of a loss of upstream piping, however, upstream manual isolation valve, SF-V-44 is maintained closed and is also classified as Seismic Class I. Additionally, the cleanup supply line from the spent fuel pool is maintained closed during normal power operations.

This valve opens to provide a flow path from the spent fuel pool cleanup system to the suction of the decay heat removal pump. This function is not required for safe shutdown or accident mitigation.

Test Requirement:

Exercise test to the closed position in accordance with ISTC-3522.

Justification It is not practical to full stroke exercise this valve open or closed during normal power operation or during cold shutdowns. An open exercise requires the spent fuel pool cleanup line to be opened and flow established from the spent fuel pool to the suction of the DH pumps. This evolution can only be performed during refueling outages when the DH system is in cleanup mode. The closure test of this valve requires pressurization of the DH suction piping. This test can not be performed during normal power operations without rendering the DH system inoperable. To perform this testing quarterly would require major

.plant or hardware modifications (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to refueling outages.

Alternative Test This valve will be exercised open (bi-directional test) and closed during refueling outages when the DH system can be lined up for spent fuel pool cooling. As permitted by ASME OM Code ISTC-3522(a) both the open and closed check valve tests will be performed during refueling outages when it is practicable to perform both tests.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-25 Valve Number System Safety Class Category CO-V-25A CO 3 C CO-V-25B CO 3 C Function These check valves must close to prevent backflow from the associated condensate storage tank to the non-safety related de-icing piping in the event that a pipe or alternate condensate storage tank rupture occurs. This function ensures that the contents of the respective condensate storage tank will be available to the suction of the EFW pumps.

The valve opens when the system is in the de-ice mode to provide a warm water flow path from the suction of the condensate booster pumps to the condensate storage tank. This function is not required for safe shutdown or accident mitigation since the condensate and de-icing piping is non-safety related.

Test Requirement:

Exercise test to the closed position in accordance with ISTC-3522.

Justification It is not practical to full stroke exercise these valves closed during normal power operation.

The de-icing line manual valves from the condensate booster pumps (CO-V-26A/B) are throttled during de-icing operations, and they also are not completely leak-tight when di-icing is secured. As such, they do not provide a zero leakage boundary from the condensate booster pump suction pressure. During refueling outages with the condensate system out of service, the closed safety function of CO-V-25A/B can be verified by a back leakage test. To perform this testing every quarter would require major plant or hardware modifications (NUREG-1482, revision 1, section 2.4.5). This justifies deferral of testing to refueling outages.

Alternative Test These valves will be exercised closed during refueling outages using the head of the condensate storage tanks. As permitted by ASME OM Code ISTC-3522(a) both the closed and non-safety (bi-directional) open check valve tests will be performed during refueling outages when it is practicable to perform both tests.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ DELETED Valve Number System Safety Class Cateplory NR NR-V-19 3 .B This Refueling Outage Justification has been determined to not be necessary and has been deleted.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-27 Valve Number System Safety Class Categorv MU-V-i 93A MU 2 C MU-V-i 93B MU 2 C MU-V-193C MU 2 C Function These check valves must open to allow recirculation of flow from the discharge of the makeup and purification pumps during low system flow conditions through the Seal Return Coolers and back to the Makeup (MU) Tank in order to prevent overheating and potential damage to the pumps.

The valves close to ensure full MU Pump discharge flow is directed to the RCS for accident mitigation on HPI initiation by preventing diversion of flow back through the makeup and purification pumps in the event that a pump failed to start.

Test Requirement:

Exercise test to the open and closed positions in accordance with ISTC-3522.

Justification It is not practical to full stroke exercise these check valves open during normal power operation or during cold shutdowns. The Makeup pump minimum flow lines are not instrumented to allow flow measurement during quarterly testing.

Check valve closure verification requires isolation of the pump suction, Due to the design of the MU pumps (9 stage high pressure centrifugal), positive verification of closure of the minimum flow line check valves (and discharge check valves) cannot be ascertained during quarterly testing of the pumps on minimum recirculation flow.

Alternative Test These check valves will be exercised open and closed during refueling outages when injection to the reactor coolant system is possible (full forward flow testing can be performed) and system alignment will allow the pump suctions to be isolated for positive verification that the valves are exercised closed by reverse flow. These valves are partial stroked open during pump testing quarterly.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 11 ATTACHMENT 11 REFUELING OUTAGE JUSTIFICATION RJ-28 Valve Number System Safety Class Categorv RC-V-144A RC 1 C RC-V-144B RC 1 C RC-V-144C RC 1 C RC-V-144D RC 1 C RC-V-144E RC 1 C RC-V-144F RC 1 C RC-V-144G RC 1 C RC-V-144H RC 1 C Function The RC-V-144 series internals vent valves have an open safety function to ensure adequate core cooling in the event of a reactor coolant cold leg piping LOCA. The normally closed position is passive and not a safety function.

The RC-V-144 series valves consist of eight 14 inch inside diameter vent valve assemblies installed in the cylindrical wall of the reactor internals core support shield. The internals vent valves are normally closed check valves with a safety function to open, in the event of a pipe rupture in the reactor coolant cold leg piping, to permit steam generated in the core to flow directly to the leak. This will permit the core to be rapidly recovered and adequately cooled after emergency core coolant has been supplied to the reactor vessel.

Test Requirement:

Exercise test to the open position in accordance with ISTC-3522.

Justification It is not practical to exercise these valves open during normal power operation or during cold shutdowns since these valves are located inside the reactor vessel. Exercising the valves open requires removing the reactor vessel head and the upper plenum assembly to allow access to the valves. This is only practical during refueling outages.

Alternative Test The check valves will be exercised open during refueling outages when the reactor vessel head and the upper plenum assembly can be removed to allow access to the valves.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 12 TECHNICAL POSITION INDEX Revision 0 October 25, 2013 A12- 1

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 12 TECHNICAL POSITION INDEX Technical Position Revisio Title/Description Number n CTP-IST-001 Preconditioning of IST Program Components CTP-IST-002 Quarterly Pump Testing Under Full-Flow Conditions CTP-IST-003 Quarterly Testing of Group B Pumps CTP-IST-004 Classification of Pumps: Centrifugal vs. Vertical Line Shaft CTP-IST-005 Preservice Testing of Pumps CTP-IST-006 Classification and Testing of Class 1 Safety/Relief Valves With Auxiliary Actuating Devices CTP-IST-007 Skid-Mounted Components CTP-IST-008 Position Verification Testing CTP-IST-009 ASME Class 2 & 3 Relief Valve Testing Requirements CTP-IST-010 ERV and PORV Testing Requirements CTP-IST-01 1 Extension of Exercise Testing Frequencies to Cold Shutdown or Refueling Outage CTP-IST-012 Use of ASME OM Code Cases for Inservice Testing CTP-IST-013 Exercise Testing Requirements for Valves with Fail-Safe Actuators CTP-IST-014 Bi-directional Testing of Check Valves to Their Safety and Non-Safety Related Positions Revision 0 October 25, 2013 A12 -2

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 13 TECHNICAL POSITIONS Revision 0 October 25, 2013 A13- 1

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 13 Number: CTP-IST-001, Rev. 1

Title:

Preconditioning of IST Program Components Applicability: All Exelon IST Programs. This issue also applies to other Technical Specification surveillance testing where preconditioning may affect the results of the test. This Technical Position may be adopted optionally by other Exelon organizations.

Background:

There are no specified ASME Code requirements regarding preconditioning or the necessity to perform as-found testing, with the exception of setpoint testing of relief valves and MOV testing performed in accordance with Code Case OMN-1 or Mandatory Appendix Ill. Nevertheless, there has been significant concern raised by the NRC, and documented in numerous publications, over this issue. Section 3.5 of Reference 2 provides guidance on preconditioning as it relates to IST; Section 3.6 provides additional guidance on as-found testing. It is the intent of this Technical Position to provide a unified, consistent approach to the issue of preconditioning as it applies to IST Programs throughout the Exelon fleet.

The purpose of IST is to confirm the operational readiness of pumps and valves within the scope of the IST Program to perform their intended safety functions whenever called upon. This is generally accomplished by testing using quantifiable parameters which provide an indication of degradation in the performance of the component. Preconditioning can diminish or eradicate the ability to obtain any meaningful measurement of component degradation, thus defeating the purpose of the testing.

Preconditioning is defined as the alteration, variation, manipulation, or adjustment of the physical condition of a system, structure, or component before Technical Specification surveillance or ASME Code testing. Since IST is a component-level program, this Technical Position will address preconditioning on a component-level basis. Preconditioning may be acceptable or unacceptable..

• Acceptable preconditioning is defined as preconditioning which is necessary for the protection of personnel or equipment, which has been evaluated as having insufficient impact to invalidate the results of the surveillance test, or which provides performance data or information which is equivalent or superior to that which would be provided by the surveillance test.

• Unacceptable preconditioning is preconditioning that could potentially mask degradation of a component and allow it to be returned to or remain in service in a degraded condition.

In most cases, the best means to eliminate preconditioning concerns is to perform testing in the as-found condition. When this is not practical, an evaluation must be performed to determine if the preconditioning is acceptable. Appendix 1 to this Technical Position may be used to document this evaluation.

The acceptability or unacceptability of preconditioning must be evaluated on a case-by-case basis due to the extensive variability in component design, operation, and performance requirements. Preconditioning of pumps may include filling and venting of pump casings, venting of discharge piping, speed adjustments, lubrication, adjustment of seals or packing, etc. Preconditioning of Revision 0 October 25, 2013 A13 -2

Three Mile Island Unit #1 Inservice Testing Program Plan valves may include stem lubrication, cycling of the valve prior to the "test" stroke, charging of accumulators, attachment of electrical leads or jumpers, etc.

Factors to be considered in the evaluation of preconditioning acceptability include component size and type, actuator or driver type, design requirements, required safety functions, safety significance, the nature, benefit, and consequences of the preconditioning activity, the frequencies of the test and preconditioning activities, applicable service and environmental conditions, previous performance data and trends, etc.

Lubrication of a valve stem provides an example of the variability of whether or not a preconditioning activity is acceptable. For example, lubrication of the valve stem of an AC-powered MOV during refueling outages for a valve that is exercise tested quarterly would normally be considered acceptable, unless service or environmental conditions could cause accelerated degradation of its performance.

Lubrication of a valve stem each refueling outage for an MOV that is exercise tested on a refueling outage frequency may be unacceptable if the lubrication is always performed prior to the exercise test. Lubrication of a valve stem for an AOV prior to exercise testing is likely to be unacceptable, unless it can be documented that the preconditioning (i.e., maintenance or diagnostic testing) can provide equal or better information regarding the as-found condition of the valve.

Manipulation of a check valve or a vacuum breaker that uses a mechanical exerciser to measure breakaway force prior to surveillance testing would be unacceptable preconditioning. Additional information regarding preconditioning of MOVs may be found in Reference 4.

Position:

1. Preconditioning SHALL be avoided unless an evaluation has been performed to determine that the preconditioning is acceptable. Appendix 1 to this Technical Position may be used to document this evaluation. In cases where the same information applies to more than one component, a single acceptability evaluation may be performed and documented.

2. Evaluations SHALL be prepared, reviewed and approved by persons with the appropriate level of knowledge and responsibility. For example, persons preparing an evaluation should hold a current certification in the area related to the activity. Reviewers should be certified in a related area.

3. The evaluation SHALL be approved by a Manager or designee.

4. If it is determined that an instance of preconditioning has occurred without prior evaluation, the evaluation SHALL be performed as soon as practicable following discovery. If the evaluation concludes that the preconditioning is unacceptable, an IR shall be written to evaluate the condition and identify corrective actions.

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Three Mile Island Unit #1 Inservice Testing Program Plan

References:

1. NRC Information Notice 97-16, "Preconditioning of Plant Structures, Systems, and Components before ASME Code Inservice Testing or Technical Specification Surveillance Testing".

2. NUREG-1482, Revision 1 (January, 2005), Section 3.5 "Pre-Conditioning of Pumps and Valves".

3. NRC Inspection Manual Part 9900: Technical Guidance, "Maintenance - Preconditioning of Structures, Systems and Components Before Determining Operability".

4. ER-AA-302-1006, "Generic Letter 96-05 Program Motor-Operated Valve Maintenance and Testing Guidelines"

5. ER-AA-321, "Administrative Requirements for Inservice Testing" Revision 0 October 25, 2013 A13 -4

Three Mile Island Unit #1 Inservice Testing Program Plan CTP-IST-001 APPENDIX 1 EVALUATION OF PRECONDITIONING ACCEPTABILITY Description of activity:

Secion1:

RC n~secton anual-Part.9006 Review:

Answer the following questions to determine the acceptability of the preconditioning activity based on Section D.2 of Reference 3.

Question Yes No Not Determined

1. Does the alteration, variation, manipulation or adjustment ensure that the El component will meet the surveillance test acceptance criteria?

2. Would the component have failed the surveillance without the alteration, variation, 0 E0 manipulation or adjustment?

3. Does the practice bypass or mask the as-found condition? 0 El

4. Is the alteration, variation, manipulation or adjustment routinely performed just before the testing?

5. Is the alteration, variation, manipulation or adjustment performed only for scheduling convenience?

If all the answers to Questions I thru 5 are No, the activity is acceptable; go to Section 3.

Otherwise, continue to Section 2.

Section 2::AdditionaliEValuation The following questions may be used to determine if preconditioning activities that do not meet the screening criteria of Section 1 are acceptable Question Yes No

6. Is the alteration, variation, manipulation or adjustment required to prevent personnel injury or E equipment damage? If yes, explain below.

7. Does the alteration, variation, manipulation or adjustment provide performance data or information that is equivalent or superior to that provided by the surveillance test? If yes, explain below.

8. Is the alteration, variation, manipulation or adjustment being performed to repair, replace, inspect or test an SSC that is inoperable or is otherwise Unable to meet the surveillance test acceptance criteria? If yes, explain below.

9. Is there other justification to support classification of the alteration, variation, manipulation or adjustment as acceptable preconditioning? If yes, explain below and provide references.

Explanation / Details: (attach additional sheets as necessary)

==

Conclusion:==

The preconditioning evaluated herein (is / is not) acceptable. (Circle one)

Section 3"Review i-Approve::

Prepared by: Date:

Reviewed by: Date:

Approved by: Date:

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 13 Number: CTP-IST-002, Rev. I

Title:

Quarterly Pump Testing Under Full Flow Conditions Applicability: ASME OM-1995 Code and Later, Subsection ISTB

Background:

Pumps included in the scope of the IST Program are classified as Group A or Group B. The OM Code defines a Group A pump as a pump that is operated continuously or routinely during normal operation, cold shutdown, or refueling operations. A Group B pump is defined as a pump in a standby system that is not operated routinely except for testing.

Testing of pumps in the IST Program is performed in accordance with Group A, Group B, comprehensive or preservice test procedures. In general, a Group A test procedure is intended to satisfy quarterly testing requirements for Group A pumps, a. Group. B test procedure is intended to satisfy quarterly testing requirements for Group B pumps and a comprehensive test procedure is required to be performed on a frequency of once every two years for all Group A and Group B pumps. The Code states that when a Group A test is required a comprehensive test may be substituted; when a Group B test is required a comprehensive test or a Group A test may be substituted. A preservice test may be substituted for any inservice test. The Corporate Exelon position on preservice testing requirements for pumps in the IST Program is provided in CTP-IST-005.

Subsection ISTB provides different acceptance, alert and required action ranges.

for centrifugal, vertical line: shaft, non-reciprocating positive displacement and reciprocating positive displacement pumps, for Group A, Group B and.

comprehensive pump tests. In each case, the acceptance bands for flow and differential or discharge pressure for the comprehensive test are narrower than those for the Group A and Group B tests. Since comprehensive pump test.

requirements did not exist prior to the OM-1995 Code, and since the frequency of comprehensive tests is once every two years, most stations have a limited history of comprehensive pump test performance. Thus, pumps. that have demonstrated satisfactory results during quarterly testing over a period of several years may fail a comprehensive test while continuing to operate at the same performance level.

Position: The following points summarize the Exelon position on full-flow testing of pumps:

1. Any specific pump is either Group A or Group B; it cannot be both. Any pump that is operated routinely for any purpose, except for the performance of inservice testing, is a Group A pump. A pump cannot be classified as Group A for certain modes of operation and Group B for other modes of operation (e.g.,

pumps used for shutdown cooling are Group A pumps), unless authorized by means of an NRC-approved Relief Request.

2. Under certain circumstances, similar or redundant pumps may be classified differently. For example, if a station has four identical RHR pumps with two used for shutdown cooling and two dedicated to ECCS service, the shutdown cooling pumps would be Group A, whereas the dedicated ECCS pumps would be Group B provided they were maintained in standby except when performing inservice testing.

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Three Mile Island Unit #1 Inservice Testing Program Plan

3. Quarterly testing of Group A pumps shall be performed in accordance with a Group A or comprehensive test procedure. Post-maintenance testing of Group A pumps shall be performed in accordance with a Group A, a comprehensive, or a preservice test procedure.

4. Quarterly testing of Group B pumps shall be performed in accordance with a Group B, Group A, or comprehensive test procedure. Post-maintenance testing of Group B pumps shall be performed in accordance with a Group A, a comprehensive, or a preservice test procedure.

5. Credit can only be taken for a comprehensive test if all of the OM Code requirements for a comprehensive test are met, including flow, instrument range and accuracy, and acceptance limits.

Regardless of test conditions, quarterly pump testing. is required to meet the acceptance criteria specified for Group A or Group B pumps, as applicable, in the edition/addenda of the OM Code in effect at the Plant. More restrictive acceptance criteria may be applied optionally if desired to improve trending or administrative control.

The ASME OM Code has identified quarterly and comprehensive pump testing as distinctly separate tests with separate frequency and instrumentation requirements and separate acceptance criteria. When performing a quarterly (Group A or Group B) test under full flow conditions, it may be apparent that a comprehensive test limit was exceeded. In such cases, ISSUE an IR to describe and evaluate the condition and potential compensatory measures (e.g.,

establishing new reference values) prior to the next scheduled comprehensive test. No additional corrective actions are required.

References:

1. ASME OM Code, Code for Operation and Maintenance of Nuclear Power Plants, 1995 Edition and later, Subsection ISTB.

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Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 13 Number: CTP-IST-003, Rev. 0

Title:

Quarterly Testing of Group B Pumps Applicability: ASME OM-1995 Code and Later

Background:

Pumps included in IST Programs that must comply with the 1995 Edition of the ASME OM Code and later are required to be classified as either Group A or Group B pumps. The OM Code defines a Group A pump as a pump that is operated continuously or routinely during normal operation, cold shutdown, or refueling operations. A Group B pump is defined as a pump in a standby system that is not operated routinely except for testing.

Testing of pumps is performed in accordance with Group A, Group B, comprehensive or preservice test procedures. In general, a Group A test procedure is intended to satisfy quarterly testing requirements for a Group A pump, a Group B test procedure is intended to satisfy quarterly testing requirements for a Group B pump, and a comprehensive test procedure is required to be performed on a frequency of once every two years for all Group A and Group B pumps. A Group A test procedure may be substituted for a Group B procedure and a comprehensive or preservice test procedure may be substituted for a Group A or a Group B procedure at any time.

A Group A test procedure is essentially identical to the quarterly pump test that was performed in accordance with OM-6 and earlier Code requirements. Group B testing was introduced to the nuclear industry when the NRC endorsed the OM-1995 Edition with OMa-1996 Addenda in 10 CFR 50.55a(b)(3). The intent of the Group B test was to provide assurance that safety related-pumps that sit idle essentially all of the time (e.g. ECCS pumps) would be able to start on demand and achieve a pre-established reference condition. The requirements for Group B testing were significantly relaxed when compared with the Group A (traditional) pump test requirements based on the assumption that there were no mechanisms or conditions that would result in pump degradation while the pump sat idle.

Strong differences of opinion regarding the intent and requirements for Group B testing developed and have persisted since the beginning. These differences span the industry, the NRC, and even members of the OM Code Subgroup-ISTB who created them. One opinion is that the Group B test is intended to be a "bump" test in which the pump is started, brought up to reference flow or pressure, and then stopped. The opposing opinion is that the Group B test requires the pump to be brought to the reference flow or pressure followed by recording and evaluation of both the flow and pressure readings. Both opinions can be supported by the applicable OM Code verbiage. However, NRC personnel have expressed a reluctance to accept the "bump" test interpretation.

Position: Group B pump testing should be performed as follows:

1. When performing a Group B pump test, both hydraulic test parameters (i.e.,

flow and differential pressure OR flow and discharge pressure) shall be Revision 0 October 25, 2013 A13-8

Three Mile Island Unit #1 Inservice Testing Program Plan measured and evaluated in accordance with the applicable Code requirements for the pump type.

2. Vibration measurements are not required for Group B pump tests. Vibration measurements may continue to be taken optionally. In the event that a vibration reading exceeds an alert or required action limit for the comprehensive test for the pump being tested, an IR shall be written and corrective action taken in accordance with the CAP process.

References:

1. ASME OM Code, Code for Operation and Maintenance of Nuclear Power Plants, 1995 Edition and later, Subsection ISTB Revision 0 October 25, 2013 A13 -9

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 13 Number: CTP-IST-004, Rev. 1

Title:

Classification of Pumps: Centrifugal vs. Vertical Line Shaft Applicability: All Exelon IST Programs

Background:

Early Code documents that provided requirements for inservice testing of pumps did not differentiate between pump types. Subsection IWP of the ASME Boiler and Pressure Vessel Code, Section Xl, required the measurement of flow, differential pressure and vibration and comparison of the measured data with reference values, similar to the way in which centrifugal pump testing is currently performed. Some additional measurements were required (e.g., bearing temperature, lubrication level or pressure) which were later determined to be of minimal value to IST. A major limitation in the earlier Code was that the same parameters and acceptance criteria were specified for all pumps.

With the development of the OM Standards (OM-1, OM-6, OM-10, etc.), it was recognized that pumps of different design performed differently and required different measurement criteria to determine acceptable performance. For example, discharge pressure was determined to be a more representative measurement of performance for a positive displacement pump than differential pressure. Part 6 of the OM Standards (OM-6), also introduced different criteria for inservice testing of centrifugal and vertical line shaft pumps. Unfortunately, it did not provide any definition for a vertical line shaft pump.

The definition of "vertical line shaft" pump was first incorporated into the OM-1 998 Edition of the OM Code as "a vertically suspended pump where the pump driver and pump element are connected by a line shaft within an enclosed column."

This definition failed to eliminate much of the uncertainty in determining whether certain pumps were vertically-oriented centrifugal pumps or vertical line shaft pumps. Further confusion was created by the choice of wording used in the OM Code Tables that specify the acceptance criteria for centrifugal and vertical line shaft pumps.

Position: Code requirements for vibration measurement provide the clearest indication of the difference between a centrifugal pump and a vertical line shaft pump. On centrifugal pumps, vibration measurements are required to be taken in a plane approximately perpendicular to the rotating shaft in two approximately orthogonal directions on each accessible pump-bearing housing and in the axial direction on each accessible pump thrust bearing housing. On vertical line shaft pumps, measurements are required to be taken on the upper motor-bearing housing in three approximately orthogonal directions, one of which is the axial direction.

Therefore, a pump which is connected to its driver by a vertically-oriented shaft in which vibration measurements must be taken on the pump motor due to the inaccessibility of the pump bearings will be classified as a vertical line shaft pump.

For plants using the 1998 Edition of the OM Code through the OMb-2003 addenda, Table ISTB-5100-1 applies to all horizontally and vertically-oriented centrifugal pumps; Table ISTB-5200-1 applies to vertical line shaft pumps. For plants using the 2004 Edition of the OM Code and later, Table ISTB-5121-1 Revision 0 October 25, 2013 A13 -10

Three Mile Island Unit #1 Inservice Testing Program Plan applies to all horizontally and vertically-oriented centrifugal pumps; Table ISTB-5221-1 applies to vertical line shaft pumps.

References:

1. ASME OMa-1988, ASME/ANSI Operation and Maintenance of Nuclear Power Plants, Part 6, Inservice Testing of Pumps in Light-Water Reactor Power Plants.

2. ASME OM Code, Code for Operation and Maintenance of Nuclear Power Plants, 1995 Edition and later, Subsection ISTB.

Revision 0 October 25, 2013 A13- 11

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 13 Number: CTP-IST-005, Rev. 1

Title:

Preservice Testing of Pumps Applicability: OM-1995 Code and Later

Background:

Requirements for preservice testing of pumps have been stated in ASME Code documents since the beginning. However, the 1995 Edition of the OM Code significantly expanded the scope of preservice testing by introducing the requirement that centrifugal and vertical line shaft pumps in systems where resistance can be varied establish a pump curve by measuring flow and differential pressure at a minimum of five points. These points are required to be from pump minimum flow to at least design flow, if practicable. At least one point is to be designated as the reference point for future inservice tests.

The OM Codes further state that it is the responsibility of the Owner to determine if preservice testing requirements apply when reference values may have been affected by repair, replacement, or maintenance on a pump. A new reference value or set of values is required to be determined or the previous reference value(s) reconfirmed by a comprehensive or Group A test prior-to declaring the pump operable.

Position: Whenever a pump's [hydraulic] reference values may have been affected by repair, replacement, or maintenance, a preservice test SHALL be performed in accordance with the preservice test requirements of Reference 1 of this CTP for the applicable pump design. If it is determined through evaluation that the maintenance activity did not affect the existing reference values, then the previous reference value(s) SHALL be reconfirmed. by a comprehensive or Group A test prior to declaring the pump operable. Evaluation that the maintenance activity did not affect the pump's reference values SHALL BE DOCUMENTED.

Since a preservice test may be substituted for any other required inservice test, this test could be performed in place of any quarterly or comprehensive test.

Performing it in lieu of a comprehensive test would have minimal impact on test scope or schedule and would provide valuable information for subsequent evaluations of pump performance.

For centrifugal and vertical line shaft pumps in systems with variable resistance, one of the five points on the preservice test curve (preferably one between 100%

and 120% of design flow but in no case less than 80% of design flow) SHALL be selected as the reference point for the comprehensive tests. If quarterly testing will be performed at full flow, then the same point should be selected for the quarterly pump tests. If quarterly testing cannot be performed at full flow, then another point on the preservice test curve SHALL be selected as the reference point for the quarterly tests.

References:

1. ASME OM Code, Code for Operation and Maintenance of Nuclear Power Plants, 1995 Edition and later, Subsection ISTB.

Revision 0 October 25, 2013 A13- 12

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 13 Number: CTP-IST-006, Rev. 0

Title:

Classification and Testing of Class 1 Safety/Relief Valves With Auxiliary Actuating Devices Applicability: All Exelon IST Programs

Background:

he definition for valve categories in the ASME Codes has been consistent since the beginning. Category A, B, C and D valves are basically defined the same now as they were in early editions/addenda of Section Xl of the ASME Boiler and Pressure Vessel Code. Likewise, the requirement that valves meeting the definition for more than one category be tested in accordance with all the applicable categories has been consistent over time.

Due to a lack of clear testing requirements for Class 1 Safety/Relief Valves With Auxiliary Actuating Devices in early ASME Codes, these valves were historically classified as Category B/C. As relief valves, they were required to meet the Category C testing requirements; and since the auxiliary operators essentially put them in the classification of power-operated valves, Category B requirements were imposed to address stroke-time and position indication testing considerations.

Position: The B/C categorization of these valves was initially made due to a lack of specific Code requirements. However, with the publication of ASME OM Standard OM-1 in 1981, which identified specific requirements for these valves, it became irrelevant. All applicable testing requirements for these valves were specified in OM-1, which has been superseded by Appendix I of the ASME OM Code. Efforts of the Code to exempt these valves from Category B testing requirements further demonstrate their inapplicability. Therefore, these valves should be classified as Category C.

References:

1. ASME OM-1987, ASME/ANSI Operation and Maintenance of Nuclear Power Plants, Part 1, Requirements for Inservice Performance Testing of Nuclear Power Plant Pressure Relief Devices.

2. ASME OM Code, Code for Operation and Maintenance of Nuclear Power Plants, 1995 Edition and later, Subsection ISTC and Appendix I.

Revision 0 October 25, 2013 A13- 13

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 13 Number: CTP-IST-007, Rev. I

Title:

Skid-Mounted Components Applicability: All Exelon IST Programs

Background:

The term "skid-mounted component" was coined to describe support components, such as pumps and valves for the purposes of IST, that function in the operation of a supported component in such a way that their proper functioning is confirmed by the operation of the supported component. For example, the successful operation of an emergency diesel-generator set confirms that essential support equipment, such as cooling water and lube oil pumps and valves, are functioning as required. The concept of "skid-mounted" is actually irrespective of physical location.

Position: Components 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 are required to tested in accordance with the ASME Code-in-effect for the station's IST Program. It is not the intent of the skid-mounted exemption that it be used in cases where the specific testing requirements of the Code for testing of pumps and valves can be met. For example, if adequate instrumentation is provided to measure a pump's flow and differential pressure, and if required points for vibration measurement can be accessed, then invoking the skid-mounted exemption would be inappropriate.

The "skid-mounted" exclusion as stated in references 2 and 3, below, may be applied to pumps or valves classified as "skid-mounted" in the IST Program provided that they are tested as part of the major component and are justified to be adequately tested. Such components SHALL be listed in the Program Plan document and identified as skid-mounted. Pump or Valve Data Sheets which contain the justification regarding the adequacy of their testing SHALL be provided in the IST Bases Document.

References:

1. NUREG-1482 (Rev.0 and Rev.1), Section 3.4, Skid-Mounted Components and Component Subassemblies

2. ASME OM Code, Code for Operation and Maintenance of Nuclear Power Plants, 1995 Edition OMa-1996 Addenda, ISTA 1.7, ISTC 1.2.

3. ASME OM Code, Code for Operation and Maintenance of Nuclear Power Plants, 1998 Edition and later, ISTA-2000 and ISTC-1200.

Revision 0 October 25, 2013 A13- 14

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 13 Number: CTP-IST-008, Rev. 0

Title:

Position Verification Testing Applicability: All Exelon IST Programs

Background:

Valves with remote position indicators are required to be observed locally at least once every two years to verify that valve operation is accurately indicated. This local observation should be supplemented by other indications to verify obturator position. Where local observation is not possible, other indications shall be used for verification of valve operation.

Position: All valves within the scope of the IST Program that are equipped with remote position indicators, shall be tested. The testing shall clearly demonstrate that the position indicators operate as required and are indicative of obturator position.

For example, a valve that has open and closed indication shall be cycled to demonstrate that both the open and closed indicators perform as designed, including both or neither providing indication when the valve is in mid-position.

Valves that have indication in one position only shall be cycled to ensure that the indicator is energized/de-energized when appropriate. These requirements apply to all IST valves, regardless of whether they are classified as active or passive.

References:

1. ASME OM Code, Code for Operation and Maintenance of Nuclear Power Plants, 1995 Edition with OMa-1996 Addenda, para ISTC 4.1.

2. ASME OM Code, Code for Operation and Maintenance of Nuclear Power Plants, 1998 Edition and later, para ISTC-3700.

3. NUREG-1482, Rev. 1, Section 4.2.8 Revision 0 October 25, 2013 A13 - 15

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 13 Number: CTP-IST-009, Rev. 0

Title:

ASME Class 2 & 3 Relief Valve Testing Requirements Applicability: All Exelon IST Programs

Background:

The ASME OM Code, Appendix I, provides requirements for Inservice Testing of ASME Class 1, 2, and 3 Pressure Relief Devices. The requirements for Class 1 pressure relief devices are identified separately from those for Classes 2 and 3.

The requirements for Class 2 and 3 pressure relief devices are identified together.

This Technical Position applies only to ASME Class 2 and 3 safety and relief valves. It does not include vacuum breakers or rupture discs. Class 2 PWR Main Steam Safety Valves are also not included in this Technical Position because they are required to be tested in accordance with ASME Class 1 safety valve requirements.

Position: This Technical Position applies to the classification, selection, scheduling and testing of ASME Class 2 and 3 safety and relief valves only. For the purposes of this Technical Position, the term "relief valve" will be used to apply to both types.

Classification DETERMINE whether or not the valve may be classified as a thermal relief. A thermal relief valve is one whose only over-pressure protection function is to protect isolated components, systems, or portions of systems from fluid expansion caused by changes in fluid temperature. If a relief valve is required to perform any other function in protecting a system or a portion of a system that is required..

to place the reactor in the safe shutdown condition, to maintain the safe shutdown condition, or to mitigate the consequences of an accident, it cannot be classified as a thermal relief Valve.

Class 2 and Class 3 thermal relief valves are required to be TESTED or REPLACED every 10 years unless performance data indicates the need for more frequent testing or replacement. Details regarding whether a Class 2 or Class 3 thermal relief valve is tested or replaced and the bases for the associated frequency SHALL be documented in the IST Bases Document.

Grouping, sample expansion and the requirement to test 20% of the valves within any 48-month period. do not apply to Class 2 and. Class 3 thermal relief Valves.

Class 2 and 3 thermal relief valves may be optionally tested in accordance with the more conservative requirements for non-thermal relief valves if desired.

'Non-thermal relief valves shall be grouped in accordance with the grouping criteria of Appendix I (same manufacturer, type, system application, and service media). Groups may range in size from one valve to all of the valves'meeting the grouping criteria. Grouping criteria SHALL be documented in the IST Bases Document or other document that controls Class 2 and 3 IST relief valve testing.

If two valves are manufactured at the same facility to the same specifications, dimensions, and materials of construction but under a different manufacturer's name due to a merger or acquisition, the valves may be considered to meet the requirement for same manufacturer.

Revision 0 October 25, 2013 A13 - 16

Three Mile Island Unit #1 Inservice Testing Program Plan Valves in systems containing air or nitrogen may be considered to have the same service media.

Selection Valves SHALL be selected for testing such that the valve(s) in each group with the longest duration since the previous test are chosen first. This SHALL INCLUDE any valves selected due to sample expansion..

IF an exception to this requirement is necessary due to accessibility or scheduling considerations, DOCUMENT the reason and that the valves that should have been selected will not come due prior to the next opportunity to test them (e.g.,

the next outage).

Scheduling Grace is NOT permitted for relief valve testing, unless authorized by an NRC-approved relief request.

All frequency requirements are test-to-test (i.e., they begin on the most recent date on which the valve was tested per Appendix I requirements and end. on the date of the next Appendix I test).

All Class 2 or Class 3 relief valves in any group must be tested at least once every 10 years.

Valves within each group must be tested such that a minimum of 20% of the valves are tested within any given 48-month period.

If all of the valves in a group are removed for testing and replaced with pretested valves, the removed valves shall be tested within 12 months of removal from the system.

If less than all of the valves in a group are removed for testing and replaced with pretested valves, the removed valves shall be tested within 3 months of removal from the system or before resumption of electric power generation, whichever is later.

Testing of pretested valves must have been performed such that they will meet the 10 year and 20% /48-month requirements for the entire time they are in service.

Testing of relief valves that is required to be performed during an outage SHALL BE PERFORMED as early in the outage as practicable in order to allow for contingency testing of additional valves in the event a scheduled valve fails its as-found test.

Testing Testing SHALL BE PERFORMED using the same service media wherein the valve was installed.

Testing of additional valves due to failure of a scheduled valve to meet its as-found setpoint acceptance criteria SHALL BE PERFORMED in accordance with all applicable OM Code and Technical Specification requirements.

References:

1. ASME OM Code, 1995 Edition and later, Mandatory Appendix I, Inservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants Revision 0 October 25, 2013 A13 -17

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 13 Number: CTP-IST-010, Rev. 0

Title:

ERV and PORV Testing Requirements Applicability: Exelon Stations with Electromatic Relief Valves or Power-Operated Relief Valves

Background:

Electromatic Relief Valves (ERVs) and Power-Operated Relief Valves (PORVs) are used at nuclear plants to protect the Reactor Coolant pressure boundary from overpressure under various conditions. This may include preventing excessive challenges to BWR Main Steam Safety Valves and PWR Pressurizer Safety Valves during operation at power or preventing low temperature overpressure (LTOP) conditions from exceeding brittle fracture limits when the plant is cooled down.

ERVs and PORVs come in a variety of designs, which can make their categorization and testing in accordance with OM Code requirements challenging.

Some are actual relief valves that are equipped with air operators to open the valves against spring force upon actuation by some pressure-sensing apparatus in the primary coolant system. Others may be motor-operated gate valves that open and close as a result of signals generated at predetermined pressure settings. The key to determining the proper category of the ERV or PORV is not the nomenclature of the valve (i.e., "relief valve"), but the actual physical design of the valve and its actuator.

Power-operated relief valves were not addressed by the ASME Codes until the OMa-1996 Addenda. Even then, they were only alluded to by the addition of an exclusion to paragraph ISTC 1.2 which stated: "Category A and B safety and relief valves are excluded from the requirements of ISTC 4.1, Valve Position Verification and ISTC 4.2, Inservice Exercising Test." Up to this point, Owners typically categorized these valves as Category B/C, assigned the position verification and exercise test requirements for the Category B portion, and then obtained Relief from the NRC to not perform them due to their impracticability. The Relief Requests provided a detailed description of the proposed alternative techniques, which generally matched Category C requirements for valves with auxiliary actuators.

Paragraph ISTC-5110 was introduced in the OM-1998 Edition of the OM Code which stated: "Power-operated relief valves shall meet the requirements of ISTC-5100 for the specific Category B valve type and ISTC-5240 for Category C valves." This essentially added no value, since this was already the practice.

OMb-2000 added the following definition of a power-operated relief valve to paragraph ISTC-2000, Supplemental Definitions: "a power-operated valve that can perform a pressure relieving function and is remotely actuated by either a signal from a pressure sensing device or a control switch. A power-operated relief valve is not capacity certified under ASME Section III overpressure protection requirements." In addition, OMb-2000 added the following to paragraph ISTC-3510: "Power-operated relief valves shall be exercise tested once per fuel cycle."

The addition of exclusions, definitions and test requirements to the Code for these valves has only tended to make actual testing requirements more conflicting or confusing. These valves are still being categorized as Category B, C or B/C (with Revision 0 October 25, 2013 A13 -18

Three Mile Island Unit #1 Inservice Testing Program Plan a few A's or A/C's) throughout the industry with testing requirements assigned accordingly and relief still being sought where deemed appropriate.

Position: Each Station MUST DETERMINE the proper valve category or categories for its ERVs and/or PORVs based on valve and actuator design, and IDENTIFY appropriate testing requirements and methodologies appropriate to that categorization. The following table summarizes the possible categories that can be applied to an ERV or PORV, whether or not the valve meets the definition of a PORV as defined in ISTC-2000, and the .associated test requirements:

Meets Category PORV Test Requirements Comments Def.

B C B C Valve is not a safety or relief valve; X No ISTC-3700 actuator is MO, AO or HO. Does not ISTC-5120* meet Code definition of PORV (ISTC-ISTC-5130* 2000).

00.EecsExercise test et quarterly urel per e ISTC-SC ISTC-5140" 3510, or defer to Cold Shutdown or RFO per ISTC-3521.

X Yes ISTC-3700 Valve meets Code definition of PORV ISTC-5110 (ISTC-2000). Exercise test once per fuel cycle per ISTC-3510 and ISTC-5110.

Valve is a relief valve with AO or HO STC-5240 actuator. Does not meet Code definition App. I of PORV (ISTC-2000). Exempt from Cat B testing (ISTC-3500/ISTC-3700) per ISTC-1200.

X X No Valve is a relief valve with AO or HO ISTC-5240 actuator. Does not meet Code definition App. I of PORV (ISTC-2000). Exempt from Cat B testing (ISTC-3500/ISTC-3700) per ISTC-1200.

X X Yes ISTC-3700 Should not be classified Category C.

ISTC-5110 Relief valves do not meet the Code definition of PORV (ISTC-2000).

• As applicable A Relief Request SHALL BE SUBMITTED for any ERV or PORV that does not meet the applicable test requirements specified in the above table.

A detailed description of the rationale behind the category designation, the assignment of testing requirements, and how they are satisfied SHALL BE PROVIDED on the applicable IST Bases Document Valve Data Sheets.

References:

1. ASME OM Code, 1995 Edition and later, Subsection ISTC, Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants

2. ASME OM Code, 1995 Edition and later, Mandatory Appendix I, Inservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants Revision 0 October 25, 2013 A13 - 19

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 13 Number: CTP-IST-011, Rev. 0

Title:

. Extension of Valve Exercise Test Frequencies to Cold Shutdown or Refueling Outage Applicability: All Exelon IST Programs

Background:

Requirements for exercise testing of Category A and B power-operated valves and check valves (Category C) are stipulated in the OM Code as follows:

ISTC-3510 states: "Active Category A, Category B and Category C check valves shall be exercised nominally every 3 mo, except as provided by paras. ISTC-3520, ISTC-3540, ISTC-3550, ISTC-3570, ISTC-5221 and ISTC-5222." Plant Technical Specifications for IST identify the 3 month frequency as once per 92 days with allowance for a 25% extension.

ISTC-3520 is divided into ISTC-3521 for Category A and Category B valves, and ISTC-3522 for Category C check valves. ISTC-3521 states: "Category A and B valves shall be tested as follows:

(a) full-stroke exercising of Category A and Category B valves during operation at power to the position(s) required to fulfill its function(s).

(b) if full-stroke exercising during operation at power is not practicable, it may be limited to part-stroke during operation at power and full-stroke during cold shutdowns.

(c) if exercising is not practicable during operation at power, it may be limited to full-stroke exercising during cold shutdowns.

(d) if exercising is not practicable during operation at power and full-stroke during cold shutdowns is also not practicable, it may be limited to part-stroke during cold shutdowns and full-stroke during refueling outages.

(e) if exercising is not practicable during operation at power or cold shutdowns, it may be limited to full-stroke during refueling outages.

Paragraphs (f) through (h) provide additional limitations on cold shutdown and refueling outage exercise testing.

ISTC-3522 provides essentially the same requirements for check valves except that the requirement to consider partial-stroke exercising is not included.

ISTC-3540 stipulates exercise testing frequency requirements for manual valves.

ISTC-3550 discusses valves in regular use, ISTC-3570 addresses valves in systems out-of-service, ISTC-5221 addresses special frequency considerations for check valves in a sample disassembly and inspection program, and ISTC-5222 addresses check valves in a condition monitoring program.

ISTC-3521 makes it clear that the intent of the Code is for valves to be exercised quarterly unless it is impracticable to do so. When it is impracticable, the graduated approach of ISTC-3521 through cold shutdown and refueling frequencies and partial and full-stroke exercising impose an obligation on the owner to perform at least some testing as frequently as practicable.

The determination of "practicability" is left to the owner. The industry has universally adopted the practice of writing Cold Shutdown and Refueling Outage Revision 0 October 25, 2013 A13 -20

Three Mile Island Unit #1 Inservice Testing Program Plan Justifications to document conditions that they believe to be "impracticable".

There are no Code or regulatory definitions of impracticability nor are there any Code or regulatory requirements to prepare Cold Shutdown or Refueling Outage Justifications. However, Reference 2 provides a good deal of useful guidance regarding a regulatory opinion of what constitutes it. Merriam-Webster defines "impracticable" as (1) impassable or (2) not practicable; incapable of being performed or accomplished by the means employed or at command".

Position: The following direction SHALL BE IMPLEMENTED when establishing exercise test frequencies for power-operated Category A and B valves and Category C check valves:

1. Stations SHALL DETERMINE the practicability of performing exercise testing of all valves in their IST Programs in accordance with the Code.

2. When preparing or performing a technical revision to a Cold Shutdown or Refueling Outage Justification, the Station IST Engineer SHALL OBTAIN a peer review from the Corporate IST Engineer and at least one other Site IST Program Engineer.

3. Cold Shutdown and Refueling Outage Justifications SHALL PROVIDE a strong, clear technical case for the testing deferral. References to NUREG-1482 may be made to support the justification; however, it is not to be cited as the justification itself.

References:

1. ASME OM Code, 1995 Edition and later, Subsection ISTC, Inservice Testing of*Valves in Light-Water Reactor Nuclear Power Plants

2. NUREG 1482, Revision 1, Guidelines for Inservice Testing at Nuclear Power Plants, Sections 2.4.5 and 3.1..

Revision 0 October 25, 2013 A13 - 21

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 13 Number: CTP-IST-012, Rev. 0

Title:

Use of ASME OM Code Cases for Inservice Testing Applicability: All Exelon IST Programs

Background:

Code Cases are issued to clarify the intent of existing Code requirements or to provide alternatives to those requirements. Adoption of the alternative requirements provided by Code Cases are optional; they only become mandatory when an owner commits to them. Code Cases are included as a separate section at the end of published editions/addenda of the OM Code for the user's convenience. They are not a part of any Code edition or addenda and endorsement of specific editions/ addenda of the OM Code by the NRC does not constitute endorsement of the Code Cases.

If the Code Committee desires to make the requirements of a Code Case mandatory, those requirements are incorporated into the Code at a later date. For example, Code Case OMN-1, Alternative Rules for Preservice and Inservice Testing of Active Electric Motor Operated Valve Assemblies in Light-Water Reactor Power Plants, was incorporated into the 2009 Edition of the OM Code as Mandatory Appendix III. Appendix Ill will become mandatory for IST Programs when 10 CFR 50.55a imposes the requirement that 10-year interval updates meet the requirements of the 2009 Edition of the ASME Code or later. Until such time, plants may optionally implement OMN-1 or may continue to perform stroke-time testing and position indication verification in accordance with Subsection ISTC requirements.

In order for an OM Code Case to be used in an Inservice Testing Program at a nuclear power plant, it must be authorized by ASME and approved by the NRC. A Code Case. is authorized for use by ASME as soon as. it is published, provided certain limitations included in the Code Case, such as the applicability statement, are met. OM Code Cases are published on the ASME Web site at http://

cstools.asme.org and in Mechanical Engineering magazine as they are issued.

Efforts to clarify or simplify the use of Code Cases have instead created conflicting requirements which need to be addressed in order to avoid noncompliance with the Code or CFR. These include:

The Code of Federal Regulations, paragraph 10 CFR 50.55a(b)(6) states that Licensees may apply ASME OM Code Cases listed in Regulatory Guide 1.192 without prior NRC approval subject to certain conditions. One condition states that when a licensee initially applies a listed Code case, the licensee shall apply the most recent version of the Code case "incorporated by reference in this paragraph". A second condition states that if a licensee has previously applied a Code case and a later version of the Code. case is "incorporated by reference in this paragraph", the licensee may continue to apply, to the end of the current 120-month interval, the previous Version of the Code case or may apply the later version of the Code case, including any NRC-specified conditions placed on its use. A third condition restricts the use of annulled Code cases to those that were in use prior to their annulment.

Revision 0 October 25, 2013 A13-22

Three Mile Island Unit #1 Inservice Testing Program Plan It is not clear what "incorporated by reference in this paragraph" is referring to. If "this paragraph" means 10 CFR 50.55a(b)(6), this would refer to Reg Guide 1.192. If it refers more broadly to 10 CFR 50.55a(b), this would also include 10CFR 50.55a(b)(3), which contains the endorsement of the latest edition/addenda of the OM Code approved for use by the NRC. In the first case, Reg Guide 1.192 was published in June 2003 with no revisions to date.

Versions of the Code cases referenced therein have all exceeded their expiration dates and are not applicable to current Code editions. In the latter case, since Code Cases are independent of Code editions/addenda, there is a disconnect between approval of Code versus Code Cases.

" Requirements for the use of Code Cases are stipulated in the body of the OM Code. In all cases from the OM-1995 Edition through the OMa-2011 Addenda, it is required that "Code Cases shall be applicable to the edition and addenda specified in the inservice test plan" and " Code Cases shall be in effect at the time the inservice test plan is filed". These requirements are almost never met.

" Code Cases provided as attachments up to and including the OMb-2006 Addenda contained expiration dates. These dates are usually prior to the time it is desired to use the Code Case.

" Each Code Case contains an applicability statement. Even in the latest Edition/addenda of the Code incorporated by reference in 10 CFR 50.55a, these statements usually indicate that the Code Case applies to earlier versions of the Code than what is required to be used.

Despite the inconveniences in implementing Code Cases, they often provide alternatives to the Code that are technically superior and highly desirable from a cost-efficiency perspective. Therefore, each plant should review the potential use of Code Cases with Corporate Engineering, particularly when in the process of performing 10-year updates.

Position: The following requirements SHALL BE IMPLEMENTED in order to use ASME OM Code Cases at Exelon stations:

1. All Code Cases used by a Station for their IST Program SHALL BE LISTED

.in the IST Program Plan.

2. Code Case expiration dates, applicability statements, and the Edition/

addenda of the Code-in-effect for a Station's IST Program SHALL all be compatible for Code Cases implemented in an IST Program OR a Relief Request SHALL BE SUBMITTED to use the Code Case in accordance with Reference 2 of this CTP.

References:

1. ASME OM Code, 1995 Edition and later, Subsection ISTA, General Requirements

2. ER-AA-321, Administrative Requirements for Inservice Testing.

Revision 0 October 25, 2013 A13 -23

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 13 Number: CTP-IST-013, Rev. 0

Title:

Exercise Testing Requirements for Valves with Fail-Safe Actuators Applicability: All Exelon IST Programs

Background:

Valves with fail-safe positions usually have actuators that use the fail-safe mechanism to stroke the valve to the fail-safe position during normal operation.

For example, an air-operated valve that fails closed may use air to open the valve against spring pressure. When the actuator is placed in the closed position, air is vented from the diaphragm and the spring moves the obturator to the closed position.

The fail-safe test is generally an integral part of the stroke time exercise test and is thus performed at the same frequency. Where the exercise test is performed less frequent than every 3 months, a cold shutdown justification, refueling outage justification, or relief request is required. The same justification for the stroke time exercise test would also apply to the fail-safe test.

Position: In cases where normal valve operation moves the valve to the fail-safe position by de-energizing the operator electrically, by venting air, or both (e.g., a solenoid valve in the air supply system of a valve operator moves to the vent position on loss of power), no additional fail-safe testing is required.

In cases where a fail-safe actuator does not operate as an integral part of normal actuator operation, the fail-safe feature(s) must be tested in a manner that demonstrates proper operation of each component that contributes to the fail-safe operation. The means used to meet this requirement shall be described in the IST Bases Document.

References:

1. ASME OM Code, Code for Operation and Maintenance of Nuclear Power Plants, 1995 Edition and later, Subsection ISTC.

Revision 0 October 25, 2013 A13 - 24

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 13 Number: CTP-IST-014, Rev. 0

Title:

Bi-directional Testing of Check Valves to Their Safety and Non-Safety Related Positions Applicability: All Exelon IST Programs

Background:

This CTP addresses those cases in which inservice testing of check valves is performed in accordance with the requirements of ISTC-5221. It does not address these issues for check valves that are included in a Condition Monitoring Program. References 2 and 3 of this CTP provide additional information regarding check valve testing and Condition Monitoring.

The OM Code changed the focus of inservice testing of check valves from the ability to demonstrate that a check valve was capable of being in its safety-related position to demonstrating that the obturator was capable of free, unobstructed movement in both directions. This was accomplished by introducing a bidirectional testing requirement to inservice testing of check valves. Confirmation of this change in focus is evidenced by the fact that the Code required frequency for bi-directional testing of check valves is the lesser of the frequencies that the open direction and close direction tests can be performed. In other words, if a check valve is capable of being tested in the open direction quarterly but can only be tested closed during refueling outages, the Code required frequency for the bidirectional test is every refueling outage irrespective of the valve's safety position(s).

Condition Monitoring is the preferred method for check valve testing and inspection. For check valves that are not in a Condition Monitoring Program, the OM Code provides three options: flow/flow reversal, use of an external mechanical exerciser, and sample disassembly/examination. Of these, the flow and mechanical exerciser methods are preferred; the Code limits sample disassembly/ examination to those cases where the others are impractical. In all of these non-Condition Monitoring methods, demonstration of unobstructed obturator travel in the open and closed directions is required.

Position: The following requirements SHALL BE MET when implementing this CTP:

1. When using flow to demonstrate opening of a check valve with an open safety function, OBSERVE that the obturator has traveled to EITHER the full open position OR to the position required to perform its intended safety function(s).

Travel to the position required to perform its intended safety function(s) is defined as the minimum flow required to mitigate the system's most limiting accident requirements. For example, if three different accident scenarios called for flows of 300, 600 and 1000 gpm respectively, the required test flow would be 1000 gpm.

The full open position is defined as the point at which the obturator is restricted from further travel (e.g., hits the backstop). Methods for demonstrating travel to the full open position must be qualified if less than required accident flow is used.

Revision 0 October 25, 2013 A13 - 25

Three Mile Island Unit #1 Inservice Testing Program Plan

2. When using flow to demonstrate that the obturator of a valve that does not have an open safety function has *traveled open, the test MUST DEMONSTRATE that the obturator is unimpeded.

3. Tests for check valve closure MUST DEMONSTRATE that the check valve has travelled to the closed position, not merely that it is in the closed position.

4. Whenever design requirements are used for IST acceptance criteria, instrument accuracy MUST BE CONSIDERED. This can be accomplished by determining that sufficient margin was included in the design calculation or by adding a correction to the IST acceptance criteria.

• 5. Non-intrusive methods used to credit obturator position SHALL BE QUALIFIED. Documentation of the means used to qualify the test method(s) shall be documented in the IST Bases Document.

6. The Code requirement satisfied for each check valve, identification of the method used to satisfy the Code requirement, and a description of how the method satisfies the requirement SHALL BE PROVIDED OR REFERENCED on the Valve Data Sheet in the IST Bases Document for each check valve..

References:

1. ASME OM Code, Code for Operation and Maintenance of Nuclear Power Plants, 1995 Edition and later, Subsection ISTC.

2. ER-AA-321., Administrative Requirements for Inservice Testing

3. ER-AA-321-1005, Condition Monitoring for Inservice Testing of Check Valves Revision 0 October 25, 2013 A13 - 26

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 14 INSERVICE TESTING PUMP TABLE Revision 0 October 25, 2013 A14 - 1

Three Mile Island Unit #1 Inservice Testing Program Plan Attachment 14 Inservice Testing Pump Table REACTOR BUILDING SPRAY (Page 1)

Pump EPN Test Safety Pump Pump Nominal P&ID P&ID Test Type Test Relief Tech. Comments Group Class Type Driver Speed Coor. Freq. Request Pos.

1-BS-P-1A AB 2 C M 3600 302-712 A-6 Differential Q PR-02 Pressure - Grp A AB Flow Rate - Grp A Q PR-02 AB Vibration - Grp A Q PR-02 Pump Name REACTOR BUILDING SPRAY PUMP "A" 1-BS-P-1B AB 2 C M 3600 302-712 B-6 Differential Q PR-02 Pressure - Grp A AB Flow Rate - Grp A Q PR-02 AB Vibration - Grp A Q PR-02 Pump Name REACTOR BUILDING SPRAY PUMP "B" Revision 0 October 25, 2013 A14 - 2

Three Mile Island Unit #1 Inservice Testing Program Plan

.Attachment 14 Inservice Testing Pump Table REACTOR PLANT CHEMICAL ADDITION (Page 1)

Pump EPN Test Safety Pump Pump Nominal P&ID P&ID Test Type Test Relief Tech. Comments Group Class Type Driver Speed Coor. Freq. Request Pos.

1-CA-P-1A B 3 PDR M 1725 302-670 E-6 . Discharge Y2 Pressure - Comp B Flow Rate - Comp Y2 B Flow Rate - Grp B Q B Vibration - Comp Y2 Pump Name BORIC ACID INJECTION PUMP "A" 1-CA-P-1B 3 PDR M 1725 302-670 E-5 Discharge Y2 Pressure - Comp B Flow Rate - Comp Y2 B Flow Rate - Grp B Q B Vibration - Comp Y2 Pump Name BORIC ACID INJECTION PUMP "B" Revision 0 October 25, 2013 A14 - 3

Three Mile Island Unit #1 Inservice Testing Program Plan Attachment 14 Inservice Testing Pump Table CONTROL BUILDING CHILLED WATER (Page 1)

Pump EPN Test Safety Pump Pump Nominal P&ID P&ID Test Type Test Relief Tech. Comments Group Class Type Driver Speed Coor. Freq. Request Pos.

1-AH-P-3A A 3 C M 1750 302-847 C-5 Differential Q PR-02 Pressure - Grp A A Flow Rate - Grp A Q PR-02 A Vibration - Grp A Q .PR-02 Pump Name CONTROL BUILDING CHILLED WATER SUPPLY PUMP "A" 1-AH-P-3B A 3 C M 1750 302-847 A-5 Differential Q PR-02 Pressure - Grp A A Flow Rate - Grp A 0 PR-02 A Vibration - Grp A Q PR-02 Pump Name CONTROL BUILDING CHILLED WATER SUPPLY PUMP "B" Revision 0 October 25, 2013 A14 - 4

Three Mile Island Unit #1 Inservice Testing Program Plan Attachment 14 Inservice Testing Pump Table DECAY HEAT CLOSED CYCLE COOLING WATER (Page 1)

Pump EPN Test Safety Pump Pump Nominal P&ID P&ID Test Type Test Relief Tech. Comments Group Class Type Driver Speed Coor. Freq. Request Pos.

1-DC-P-1A A 3 C M 1150 302-645 D-3 Differential Q PR-02 Pressure - Grp A A Flow Rate - Grp A Q PR-02 A Vibration - Grp A Q PR-02 Pump Name DH CLOSED COOLING WATER PUMP "A" 1-DC-P-1B A 3 C M 1150 302-645 D-2 Differential Q PR-02 Pressure - Grp A A Flow Rate - Grp A Q PR-02 A Vibration - Grp A Q PR-02 Pump Name DH CLOSED COOLING WATER PUMP "B" Revision 0 October 25, 2013 A14 - 5

Three Mile Island Unit #1 Inservice Testing Program Plan Attachment 14 Inservice Testing Pump Table EMERGENCY DIESEL GENERATORS FUEL SYSTEMS (Page 1)

Pump EPN Test Safety Pump Pump Nominal P&ID P&ID Test Type Test Relief Tech. Comments Group* Class Type Driver Speed Coor. Freq. Request Pos.

1-DF-P-1A N/A PDN M 1750 .302-351 D-7 Flow Rate 0 IST-007 Pump Name DIESEL FUEL OIL PUMP "A" 1-DF-P-1B N/A PDN M 1750 302-351 D-6 Flow Rate Q IST-007 Pump Name DIESEL FUEL OIL PUMP "B" 1-DF-P-lC N/A PDN M 1750 302-351 D-3 Flow Rate Q IST-007 Pump Name DIESEL FUEL OIL PUMP "C" 1-DF-P-1D N/A PDN M 1750 302-351 D-2 Flow Rate Q IST-007 Pump Name DIESEL FUEL OIL PUMP "D" Revision 0 October 25, 2013 A14-6

Three Mile Island Unit #1 Inservice Testing Program Plan Attachment 14 Inservice Testing Pump Table DECAY HEAT REMOVAL (Page 1)

Pump EPN Test Safety Pump Pump Nominal P&ID P&ID Test Type Test Relief Tech. Comments Group Class Type Driver Speed Coor. Freq. Request Pos.

1-DH-P-1A A 2 C M 1800 302-640 D-5 Differential Y2 Pressure - Comp A. Differential Q Pressure - Grp A A Flow Rate - Comp Y2 A Flow Rate - Grp A Q A Vibration - Comp Y2 A Vibration - Grp A Q Pump Name DECAY HEAT REMOVAL PUMP "A" 1-DH-P-1B A 2 C M 1800 302-640 D-3 Differential Y2 Pressure - Comp A Differential Q Pressure - Grp A A Flow Rate - Comp Y2 A Flow Rate - Grp A Q A Vibration - Comp Y2 A Vibration - Grp A Q Pump Name DECAY HEAT REMOVAL PUMP "B" Revision 0 October 25, 2013 A14-7

Three Mile Island Unit #1 Inservice Testing Program Plan Attachment 14 Inservice Testing Pump Table DECAY HEAT RIVER WATER (Page 1)

Pump EPN Test Safety Pump Pump Nominal P&ID

• P&ID Test Type Test Relief Tech. Comments Group Class Type Driver. Speed Coor. Freq. Request Pos.

1-DR-P-1A A 3 VLS M 1175 302-202 A-7 Differential Q PR-02 Pressure - GrpA A Flow Rate - Grp A a PR-02 A Vibration - Grp A Q PR-02 Pump Name DECAY HEAT RIVER WATER PUMP "A" 1-DR-P-1B A 3 VLS M 1175 302-202 A-6 Differential Q PR-02 Pressure - Grp A A *Flow Rate - Grp A Q. PR-02 A Vibration - Grp A Q PR-02 Pump Name DECAY HEAT RIVER WATER PUMP "B" Revision 0 October 25, 2013 A14 - 8

Three Mile Island Unit #1 Inservice Testing Program Plan Attachment 14 Inservice Testing Pump Table EMERGENCY FEEDWATER (Page 1)

Pump EPN Test Safety Pump Pump Nominal P&ID P&ID Test Type Test Relief Tech. Comments Group Class Type . Driver Speed Coor. Freq. Request Pos.

1-EF-P-1 B 3 C T 3800 302-082 D-7 Differential Y2 Pressure - Comp B Differential Q Pressure - Grp B B Flow Rate - Comp Y2 B Flow Rate - Grp B Q.

B Speed - Comp Y2 B Speed - Grp B Q B Vibration - Comp Y2 Pump Name Steam Driven Emergency Feed Water Pump 1-EF-P-2A B 3 C M 3570 302-082 F-6 Differential Y2 Pressure - Comp B Differential Q Pressure - Grp B B Flow Rate - Comp Y2 B Flow Rate - Grp B Q B Vibration - Comp Y2 Pump Name Electric Driven Emergency Feed Pump 1-EF-P-2B B 3 C M 3570 302-082 B-6 Differential Y2 Pressure - Comp B Differential Q Pressure - Grp B B Flow Rate - Comp Y2 B Flow Rate - Grp B Q B Vibration - Comp Y2 Pump Name Electric Driven Emergency Feed Pump Revision 0 October 25, 2013 A14 - 9

Three Mile Island Unit #1 Inservice Testing Program Plan Attachment 14 Inservice Testing Pump Table MAKEUP & PURIFICATION (Page 1)

Pump EPN Test Safety Pump Pump . Nominal P&ID P&ID Test Type Test Relief Tech. Comments Group Class Type Driver Speed Coor. Freq. Request Pos.

1-MU-P-1A B 2 C M 6800 302-661 C-5 Differential Y2 Pressure - Comp B Differential Q Pressure - Grp B B Flow Rate - Comp Y2 B Vibration - Comp Y2 Pump Name MAKEUP &PURIFICATIONPUMP "A" 1-MU-P-1B A 2 C M 6800 302-661 C-4 Differential Y2 Pressure - Comp A Differential Q Pressure -Grp A A Flow Rate - Comp Y2 A Flow Rate - Grp A Q A Vibration - Comp Y2 A Vibration - Grp A* Q Pump Name MAKEUP &PURIFICATION PUMP "B" 1-MU-P-lC B. 2 C M 6800 302-661 C-3 Differential Y2 Pressure - Comp B Differential Q Pressure - Grp B B Flow Rate - Comp Y2 B Vibration - Comp Y2 Pump Name MAKEUP &PURIFICATION PUMP "C" Revision 0 October 25, 2013 A14 - 10

Three Mile Island Unit #1 Inservice.Testing Program Plan Attachment 14 Inservice Testing Pump Table NUCLEAR SERVICES RIVER WATER (Page 1)

Pump EPN Test Safety Pump Pump Nominal P&ID P&ID Test Type Test Relief Tech. Comments Group Class Type Driver Speed Coor. Freq. Request Pos.

3

.1-NR-P-1A A . VLS M 1200 .302-202 A-10

• Differential Y2 Pressure - Comp A Differential a Pressure - Grp A A Flow Rate - Comp Y2 A Flow Rate - Grp A A Vibration - Comp Y2 A Vibration - Grp A a Pump Name NUCLEAR SERVICE RIVER WATER PUMP "A" 1-NR-P-1B A .3 VLS M 1200 302-202 A-9 Differential Y2 Pressure - Comp A Differential Pressure - Grp A A Flow Rate - Comp Y2 A Flow Rate - Grp A Q A Vibration - Comp Y2 A Vibration - Grp A Q Pump Name NUCLEAR SERVICE RIVER WATER PUMP "B" 1-NR-P-1C A 3 VLS M 1200 302-202 A-9 Differential Y2 Pressure - Comp A Diffeirential Q Pressure - Grp A A Flow Rate - Comp Y2 A Flow Rate - Grp A Q A Vibration - Comp Y2 A Vibration - Grp A Q Pump Name NUCLEAR SERVICE RIVER WATER PUMP 'C" Revision 0 October 25, 2013 A14- 11

Three Mile Island Unit #1 Inservice Testing Program Plan Attachment 14 Inservice Testing Pump Table NUCLEAR SERVICES CLOSED COOLING WATER (Page 1)

Pump EPN Test Safety Pump Pump Nominal P&ID P&ID Test Type Test Relief Tech. Comments Group Class Type Driver Speed Coor. Freq. Request Pos.

1-NS-P-1A A 3 C M 1770 302-610 G-3 Differential Y2 Pressure - Comp A Differential Q PR-01 Pressure -Grp A A Flow Rate - Comp Y2 A Flow Rate - Grp A Q PR-01 A Vibration - Comp Y2 A Vibration - Grp A Q PR-01 Pump Name NUCLEAR SERVICE CLOSED COOLING WATER PUMP "A" 1-NS-P-1B A 3 C M 1770 302-610 G-3 Differential Y2 Pressure - Comp A Differential Q PR-01 Pressure - Grp A A Flow Rate - Comp Y2 A Flow Rate - Grp A Q PR-01 A Vibration - Comp Y2 A Vibration - Grp A Q PR-01 Pump Name NUCLEAR SERVICE CLOSED COOLING WATER PUMP "B" 1-NS-P-1C A 3 C M 1770 302-610 G-2 Differential Y2 Pressure - Comp A Differential Q PR-01 Pressure - Grp A A Flow Rate - Comp Y2 A Flow Rate - Grp A Q PR-01 A Vibration - Comp Y2 A Vibration - Grp A Q PR-01 Pump Name NUCLEAR SERVICE CLOSED COOLING WATER PUMP "C" Revision 0 October 25, 2013 A14 - 12

Three Mile Island Unit #1 Inservice Testing Program Plan Attachment 14 Inservice Testing Pump Table REACTOR BUILDING EMERGENCY COOLING WATER (Page 1)

Pump EPN Test Safety Pump Pump Nominal P&ID P&ID Test Type Test Relief Tech. Comments Group Class Type Driver Speed Coor. Freq. Request Pos.

1-RR-P-1A B 3 VLS M 1200 302-611. D-2 Differential Y2 Pressure - Comp B Differential 0 Pressure - Grp B B Flow Rate - Comp Y2 B Vibration - Comp Y2 Pump Name RB EMERGENCY COOLING RIVER WATER PUMP "A" 1-RR-P-1B B 3 VLS M 1200 301-611 D-2 Differential Y2 Pressure - Comp B Differential Q Pressure - Grp B B Flow Rate - Comp Y2 B Vibration - Comp Y2 Pump Name RB EMERGENCY COOLING RIVER WATER PUMP "B" Revision 0 October 25, 2013 A14- 13

Three Mile Island Unit #1 Inservice Testing Program Plan Attachment 14 Inservice Testing Pump Table SPENT FUEL COOLING (Page l)

Pump EPN Test. Safety Pump . Pump Nominal P&ID P&ID Test Type Test Relief Tech. Comments Group Class Type Driver Speed

• Coor. Freq. Request Pos.

1-SF-P-1A A 3 C M 1800 302-630 F-6 Differential Q PR-02 Pressure - Grp A A Flow Rate - Grp A Q PR-02 A Vibration - Grp A Q PR-02 Pump Name SPENT FUEL COOLING PUMP "A" 1-SF-P-1B A 3 C M 1800 302-630 D-6 Differential . Q PR-02 Pressure - Grp A A Flow Rate - Grp A Q PR-02 A Vibration - Grp A Q PR-02 Pump Name SPENT FUEL COOLING PUMP "B" Revision 0 October 25, 2013 A14 - 14

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 15 INSERVICE TESTING VALVE TABLE Revision 0 October 25, 2013 A15- 1

Attachment 15 Inservice Testing Valve Table REACTOR BUILDING PURGE (Page 1)

Valve EPN Safety Category Size Valve Act. Active i Normal Safety P&&ID P&&ID Test Test Relief . Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request. Just. Pos.

I-AH-V-11A 3 B 5 3W AO A T 0 302-847 G-7 FO :M3 IST-013 STO M3 Valve Name CONTROL BLDG VENT UNIT "A"COOLING COIL DISCH VLV 1-AH-V-11B .*3 B 5 3W AO A T 0 302-847 G-4 FO *M3 IST-013 STO M3 Valve Name CONTROL BLDG VENT.UNIT "B"COOLING COIL DISCH VLV I-AH-V-1A 2 A 48 BTF. AO A C C .302-831 G-6 FC M3 IST-01 3 LTJ AJ PI Y2..

STC M3 Valve Name CONTAINMENT ISOLATION - RB PURGE OUTLET ISOL VALVE 1-AH-V-1B 2 *A .48 BTF. MO A C C 302-831 F-6 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - RB PURGE OUTLET ISOL VALVE 1-AH-V- 1C 2 A 48 BTF MO A C *C 302-831

• C-6 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - RB PURGE INLET ISOL VALVE 1-AH-V-1D 2 A 48 BTF AO A C C 302-831 C-6 FC M3 IST-013 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - RB PURGE INLET ISOL VALVE Revision 0 October 25, 2013 A15-2

Attachment 15 Inservice Testing Valve Table EMERGENCY FEEDWATER (Page 1)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-AS-V-4 3 B/C 4 SCK MO P C . C 302-011 E-5 PI Y2 Valve Name AUX STEAM STOP CHECK VALVE SUPPLY TO EF-Ul Revision 0 October 25, 2013 A15 - 3

Attachment 15 Inservice Testing Valve Table REACTOR BUILDING SPRAY (Page 1)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-BS-V-1103 2 C 0.5 XFC SA A 0 .. 0 302-712 G-8 CO M3 Valve Name RB PRESSURE INSTRUMENT EXCESS FLOW CHECK 1-BS-V-1104 2 C 0.5 XFC SA A 0 0 302-712 G-6 CO M3 Valve Name RB PRESSURE INSTRUMENT EXCESS FLOW CHECK 1-BS-V-1105 2 C 0.5 XFC SA A 0 0 302-712 F-8 CO M3 Valve Name RB PRESSURE INSTRUMENT EXCESS FLOW CHECK 1-BS-V-1106 .2 C 0.5 XFC SA A 0 0 302-712 F-6 CO M3 Valve Name RB PRESSURE INSTRUMENT EXCESS FLOW CHECK 1-BS-V-1107 2 C 0.5 XFC SA A 0 0 302-712 D-8 CO M3 Valve Name RB PRESSURE INSTRUMENT EXCESS FLOW CHECK 1-BS-V-1108 2 C 0.5 XFC SA A 0 0 302-712 D-6 CO M3 Valve Name RB PRESSURE INSTRUMENT EXCESS FLOW CHECK 1-BS-V-IA 2 B 8 GL MO A C O/C 302-712 A-7 PI Y2 STC M3 STO M3 Valve Name BS-P1A DISCHARGE ISOLATION VALVE 1-BS-V-1B 2 B 8 GL MO A C O/C 302-712 B-7 PI Y2 STC M3 STO M3 Valve Name BS-P1B DISCHARGE ISOLATION VALVE 1-BS-V-30A 2 C 8 CK SA A C O/C 302-712 A-7 CCD CM COD CM Valve Name CONTAINMENT ISOLATION - BS NOZZLE INLET CHECK VLV 1-BS-V-30B 2 C 8 CK SA A C O/C 302-712 B-7 CCD CM COD CM Valve Name CONTAINMENT ISOLATION - BS NOZZLE INLET CHECK VLV 1-BS-V-3A 2 B 10 GA MO A C O/C 302-712 A-3 PI Y2 STC M3 STO M3 Valve Name BS-P1A SUCTION VALVE 1-BS-V-3B3 2 B 10 GA MO A C O/C 302-712 B-3 PI Y2 STC M3 STO M3 Valve Name BS-P1B SUCTION VALVE 1-BS-V-45A 2 C 1 RV SA A C. O/C 302-712 B-5 RT Y10 Valve Name BS-PlA SUCTION RELIEF 1-B3S-V-45B3 2 C 1 RV SA A C O/C 302-712 C-5 RT Y10 Valve Name BS-P1B SUCTION RELIEF 1-BS-V-59 2 A 6 GA M P LC C 302-712 B-6 LTH Y2 Valve Name TEST LINE ISOLATION VALVE TO BWST Revision 0 October 25, 2013 A15-4

Attachment 15 Inservice Testing Valve Table REACTOR PLANT CHEMICAL ADDITION (Page 1)

Valve EPN Saafety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Cllass Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-CA-V-13 1 A 0.5 GL . MO A O/C C 302-671 D-7 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - RCS LETDOWN SAMPLE VALVE 1-CA-V-177 3 C 1 CK SA A SYS 0 302-670 D-3 CC CS CSJ-15 CO CS CSJ-15 Valve Name CA-T1 TO MAKEUP SYSTEM CHECK VALVE 1-CA-V-189 2 A 2 GA AO P C C 302-669 G4 LTJ AJ PI Y2 Valve Name CONTAINMENT INTEGRITY - RECLAIMED WATER TO RB VLV 1-CA-V-192 2 A/C 2 CK SA P *C C 302-669 G-3 BDO CM CCL CM LTJ AJ Valve Name CONTAINMENT ISOLATION - RECLAIMED FEED TO RB CHK 1-CA-V-2 A . 1 GA AO A O/C C 302-671 D-7 FC . M3 IST-013 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - RC SAMPLE ISOLATION VALVE 1-CA-V-443 2 A/C 0.25 RV SA A C O/C 302-669 F-3 LTJ AJ RT Y10 Valve Name RECLAIMED WATER TO RB CONTAINMENT ISOL AND RELIEF 1-CA-V-446 2 A/C 0.5 RV SA . A C O/C 302-671 D-7 *LTJ AJ RT Y10 Valve Name RC SAMPLE. CONTAINMENT ISOLATION AND RELIEF 1-CA-V-449A 2 A/C 0.25 RV SA A C O/C 302710 G-4 LTJ AJ RT Y10 Valve Name PENETRATION 211 RELIEF VALVE 1-CA-V-449B 2 A/C 0.25 RV SA A. C. O/C 302710 F4 LTJ AJ RT Y10 Valve Name PENETRATION 210 RELIEF VALVE 1-CA-V-4A 2 A 2 GL MO A C C 302-701 G-5 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - OTSG "A"FW SAMPLE VALVE 1-CA-V-4B 2 A. 2 GL MO A C C 302-701 F-5 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - OTSG "B"FW SAMPLE VALVE Revision 0 October 25, 2013 A15 -5

Attachment 15 Inservice Testing Valve Table REACTOR PLANT CHEMICAL ADDITION (Page 2)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief. Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-CA-V-5A 2 A 2 GL AO A C C 302-701 G-4 FC M3 IST-013 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - OTSG "A"FW SAMPLE VALVE 1-CA-V-5B 2 A 2 GL AO A C C 302-701 F-4 FC M3 IST-013 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - OTSG "B"FW SAMPLE VALVE Revision 0 October 25, 2013 A15 -6

Attachment 15 Inservice Testing Valve Table CORE FLOOD (Page 1)

Valve EPN Safety Category Size Valve Act. Active/ Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-CF-V-12A 2. A/C 1 CK SA A O/C C 302-711 F-7 BDO CM CCL CM LTJ AJ Valve Name CONTAINMENT ISOLATION - CF-T1A MAKEUP CHECK VLV 1-CF-V-12B 2 A/C 1 CK SA A O/C C 302-711 F-3 BDO CM CCL CM LTJ AJ Valve Name CONTAINMENT ISOLATION - CF-T1B MAKEUP CHECK VLV 1-CF-V-19A 2 A 1 GA AO A O/C C 302-711 F-8 FC M3. IST-013 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - MU TO CF-T1A 1-CF-V-19B 2 A 1 GA AO. *A O/C C 302-711 F-3 FC M3 .IST-013 LTJ. AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - MU TO CF-T1B 1-CF-V-1A 2 B 14 GA MO P 0 0 302-711 C-7 PI Y2 Valve Name CORE FLOOD TANK A"- DISCHARGE ISOLATION VALVE 1-CF-V-1B 2 B 1.4 GA MO P 0 0 302-711 C-3 PI Y2 Valve Name CORE FLOOD TANK "B"- DISCHARGE ISOLATION VALVE 1-CF-V-20A 2 *A 1 GA AO A O/C C 302-711 D-6 FC M3 IST-013 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - CF-T1A SAMPLE ISOL VLV 1-CF-V-20B 2 A 1 GA AO A O/C C 302-711 D-5 FC M3 IST-013 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - CF-T1B SAMPLE ISOL VLV 1-CF-V-21A 2 C 2 RV SA A C O/C 302-711 G-7 RT Y8 Valve Name CORE FLOOD.TANK CF-T-1A RELIEF 1-CF-V-21B 2 C 2 RV SA A C O/C 302-711 G-3 RT Y8 Valve Name CORE FLOOD TANK.CF-T-1B RELIEF Revision 0 October 25, 2013 A15 -7

Attachment 15 Inservice Testing Valve Table CORE FLOOD (Page 2)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-CF-V-2A 2 A 1 GL MO A O/C C 302-711 D-7 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - CF-T1A SAMPLE ISO VLV 1-CF-V-2B 2 A 1 GL MO A C C 302-711 D-4 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - CF-T1 B SAMPLE ISO VLV 1-CF-V-3A 2 B 1 GL MO P C C 3102-711 F-6 PI Y2 Valve Name CORE FLOOD TANK CF-T-1A REMOTE VENT 1-CF-V-3B 2 B 1 GL MO P C C 3102-711 F-4 PI Y2

• Valve Name CORE FLOOD TANK CF-T-1B REMOTE VENT 1-CF-V-46A 2 A/C 0.25 RV SA A C O/C 3102-711 D-6 LTJ AJ RT Y10 Valve Name CONTAINMENT PENETRATION 348 RELIEF 1-CF-V-46B 2 A/C 0.25 RV SA A C O/C 3'02-711 D-4 LTJ AJ RT Y10 Valve Name CONTAINMENT PENETRATION 349 RELIEF 1-CF-V-4A 1 A/C 14 CK SA A C O/C 3'02-711 B-4 CCL CM CPF CM LTH Y2 Valve Name CORE FLOOD TANK "A" OUTLET CHECK VALVE 1-CF-V-4B 1 A/C 14 CK SA A C O/C 302-711 B-7 CCL CM CPF CM LTH Y2 Valve Name CORE FLOOD TANK "B"OUTLET CHECK VALVE 1-CF-V-5A 1 A/C 14 CK SA A C O/C 302-711 B-4 CCL RR RJ-04 COF RR RJ-04 LTH Y2 Valve Name CF-T1A &DH PUMP DISCH CHECK VALVE 1-CF-V-5B 1 A/C 14 CK SA A C O/C 302-711 B-6 CCL RR RJ-04 COF RR RJ-04 LTH Y2 Valve Name CF-T1B &DH PUMP DISCH CHECK VALVE Revision 0 October 25, 2013 A15 - 8

Attachment 15 Inservice Testing Valve Table CONTROL BUILDING CHILLED WATER (Page 1)

Category Size Valve Act. Active/I Normal. Safety .P&&ID P&&ID Test Test Relief Deferred Tech.

.Valve EPN Safety Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-CH-V-22A 3 C 4 CK SA A O/C .O/C 302-847 C-4 CC M3 CO M3 Valve Name AH-3CB CHILLED WATER DISCHARGE CHECK VALVE 1-CH-V-22B 3 C 4 CK SA A 0/C o/C 302-847 B-4 CC M3 CO M3 Valve Name AH-3CB CHILLED WATER DISCHARGE CHECK VALVE 1-CH-V-24 3 C 1 .CK SA A C C 302-847 A-7 CC M3 CO M3 Valve Name NS SUPPLY TO RBCW CHECK Revision 0 October 25, 2013 A15- 9

Attachment 15 Inservice Testing Valve Table CONTAINMENT MONITORING (Page 1)

Valve EPN Safety Category Size Valve Act. Active I Normal Safety P&&lD P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-CM-V-1 2 A 1 BALL AO A 0 C 302-721 F-7 FC M3 IST-013 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - RB ATMOSPHERE SAMPLE VALVE 1-CM-V-2 2 A 1 BALL AO A 0 C 302-721 F-7 FC M3 IST-013 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - RB ATMOS SAMPLE RETURN VLV 1-CM-V-3 2 A 1 BALL AO A 0 C 302-721 H-7 FC M3 IST-013 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - RB ATMOSPHERE SAMPLE VALVE 1-CM-V-4 2 A 1 BALL AO A 0 C 302-721 H-7 FC M3 IST-013 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - RB ATMOS SAMPLE ISOL VLV Revision 0 October 25, 2013 A15 - 10

Attachment 15 Inservice Testing Valve Table CONDENSATE (Page 1)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-CO-V-IlIA 3 B 4 GA MO A 0 C 302-101 E-6 PI Y2 STC M3 Valve Name CONDENSATE STORAGE TANK "A"ISOLATION TIE VALVE 1-CO-V-111B 3 B 4 GA MO A 0 C 302-101 E-6 PI Y2 STC M3 Valve Name CONDENSATE STORAGE TANK "B"ISOLATION TIE VALVE 1-CO-V-14A 3 B 12 GA MO A 0 C 302-101 E-5 PI Y2 STC M3 Valve Name CONDENSATE STORAGE TANK "A"ISOLATION VALVE 1-CO-V-14B 3 B 12 GA MO A 0 C 302-101 F-5 PI Y2 STC M3 Valve Name CONDENSATE STORAGE TANK "B"ISOLATION VALVE 1-CO-V-16A 3 C 10 .CK SA A SYS O/C 302-082 C-8 CCD CM COD CM COF .Y2 Valve Name CONDENSATE CHECK VALVE - SUPPLY TO EFW PUMPS 1-CO-V-16B 3 C 10 CK SA A SYS O/C" 302-082 F-8 CCD CM COD CM COF Y2 Valve Name CONDENSATE CHECK VALVE - SUPPLY TO EFW PUMPS 1-CO-V-175A 3 C 2 CK SA A SYS: O/C 302-101 E-6 . CCD CM COD CM VaIve Name EFW PUMP BEARING COOLING RETURN CHECK VALVE.

1-CO-V-175B 3 C 2 CK SA.. A SYS O/C 302-101 E-6. CCD CM COD CM Valve Name EFW PUMP BEARING COOLING RETURN CHECK VALVE 1-CO-V-25A 3 C 2 CK SA A. C C 302-101 C-4 BDO RR RJ-25 IST-014 CC RR RJ-25 Valve Name CST-A DE-ICE LINE INLET CHECK 1-CO-V-25B 3 C 4. CK SA A *C C 302-101 E-5 BDO RR RJ-25 . IST-014 CC RR RJ-25 Valve Name CST-B DE-ICE LINE INLET CHECK Revision 0 October 25, 2013 A15- 11

Attachment 15 Inservice Testing Valve Table DECAY HEAT CLOSED CYCLE COOLING WATER (Page 1)

Category Size Valve Act. . Active I Normal Safety P&&ID P&&ID Valve EPN Safety. Test Test Relief Deferred Tech.

Class Type Type Passive . Position Position Coor. Type Freq. Request Just. Pos.

1-DC-V-15A 3 C 0.5 RV SA "A C O/C .302-645 A-8 RT Y10 Valve Name DH REMOVAL COOLER DH-C-1A SHELL SIDE RELIEF 1-DC-V-15B 3 C 0.5 RV SA . A C O/C 302-645 C-9 RT Y10 Valve Name DH REMOVAL COOLER DH-C-lB SHELL SIDE RELIEF 1-DC-V-16A 3 C 0.5 RV SA A C O/C 302-645 C-6 RT Y10 Valve Name BS-P-1A MOTOR AND BEARING COOLING.WATER RELIEF 1-DC-V-16B 3 C 0.5 RV SA. A C. o/C 302-645 C-7 RT Y10 Valve Name BS-P-1B MOTOR AND BEARING COOLING WATER RELIEF 1-I)C-V-17A 3 C 0.5 RV SA A C O/C 302-645 E-6 RT Y10 Valve Name DH-P-1A MOTOR AND BEARING COOLING WATER RELIEF 1-DC-V-17B 3 C 0.5 RV SA A C O/C 302-645 E-7 RT Y10 Valve Name DH-P-1B MOTOR AND BEARING COOLING WATER RELIEF 1-DC-V-18A 3 C 0.5 RV SA A C O/C 302-645. D-3 RT Y10 Valve Name DC-P-1A BEARING COOLING WATER RELIEF 1-[C-V-18B 3 C 0.5 RV SA

• A " C O/C 302-645 D-2 RT Y10 Valve Name DC-P-1B BEARING COOLING WATER RELIEF 1-OC-V-52A C 0.75 RV SA A C O/C 302-645 F-10 RT Y10 3 Valve Name DH SERVICE COOLER DC-C-2A RELIEF 1-DC-V-52B C 0.75 RV . SA A. C O/C 302-645 F-8 RT Y10 Valve Name DH SERVICE COOLER DC-C-2B RELIEF Revision 0 October 25, 2013 A15 - 12

Attachment 15 Inservice Testing Valve Table EMERGENCY DIESEL GENERATORS FUEL SYSTEMS (Page 1)

Category Size Valve Act. Active I Normal Safety P&&ID P&&ID Test Test Relief Deferred Valve EPN Safety Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-DF-V-23A N/A C 2 SCK SA A C O/C 302-283 D-7 CC M3 IST-007 CO M3 IST-007 Valve Name DF-T1 FOOT VALVE 1-DF-V-23B N/A C 2 SCK SA A C O/C 302-283 D-7 CC M3 IST-007 CO M3 IST-007 Valve Name DF-T1 FOOT VALVE 1-DF-V-7A N/A C 1 CK SA A C O/C 302-351 D-7 CO M3 IST-007 Valve Name DF-PlA DISCHARGE CHECK VALVE 1-DF-V-7B N/A C 1 CK SA A C O/C 302-351 D-6 CO M3 IST-007 Valve Name DF-P1B DISCHARGE CHECK VALVE 1-DF-V-7C N/A C 1 CK SA A C O/C 302-351 D-3 CO M3 IST-007 Valve Name DF-PlC DISCHARGE CHECK VALVE 1-DF-V-7D N/A C 1 CK SA A C O/C 302-351 D-3 CO M3 IST-007 Valve Name DF-P1D DISCHARGE CHECK VALVE Revision 0 October 25, 2013 A15-13

Attachment 15 Inservice Testing Valve Table DECAY HEAT REMOVAL (Page 1)

Valve EPN Safety Category Size Valve Act. Active / Normal. Safety P&&ID P&&ID Test Test Relief. Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-DH-VI- 1 A 12 GA MO A C ..O/C 302-640 E-7 LTH Y2 PI Y2 STC CS CSJ-02 STO CS CSJ-02 Valve Name DECAY HEAT SUCTION ISOLATION FROM "B"HOT LEG 1-DH-V-13A 2 .C 0.75 .RV SA A C O/C 302-640 C-6 RT Y10 Valve Name DH-P-1A SUCTION RELIEF 1-DH-V-13B 2 C 0.75 RV SA A C O/C 302-640 C-4 RT, Y10 Valve*Name DH-P-1B SUCTION RELIEF 1-DH-V-14A 2 A/C 14 CK SA A C. O/C 302-640 C-3 CCL RR RJ-20 COF RR RJ-20 LT Y2 Valve Name DH PUMP SUCTION FROM BWST CHECK VALVE 1-DH-V-14B 2 A/C 14 CK SA A C O/C 302-640 B-4 CCL RR RJ-20 COF RR RJ-20 LT Y2 Valve Name DH PUMP SUCTION FROM BWST CHECK VALVE 1-DH-V-166A 2 C 0.5 CK SA A C O/C 302-640 G-7 CC CS CSJ-14 CO CS CSJ-14 Valve Name DH-V 004A INLET DISC PRESSURE EQUALIZATION CHECK 1-DH-V-1666 2 C 0.5 CK SA A C O/C 302-640 H-7 CC CS CSJ-14 CO CS CSJ-14 Valve Name DH-V 004B INLET DISC PRESSURE. EQUALIZATION CHECK 1-DH-V-16A 2 C 10 CK SA A SYS O/C 302-640 D-5 CC RR RJ-01

'CO RR RJ-01 Valve Name DH-P1A DISCHARGE CHECK VALVE 1-DH-V-16B 2 C 10 CK SA A . SYS O/C 302-640 D-3 CC RR RJ-01 CO RR RJ-01 Valve Name DECAY HEAT PUMP "B"DISCHARGE CHECK VALVE 1-DH-V-170 2 C 0.5 CK SA A C 0 302-640 E-7 BDC RR RJ-22 CO RR RJ-22 Valve Name DH-V 001 INTERDISC OVERPRESSURE RELIEF CHECK 1-DH-V-171 2 C 0.5 CK SA A C 0 302-640 D-7 BDC RR RJ-22 CO RR RJ-22 Valve Name DH-V 002 INTERDISC OVERPRESSURE RELIEF CHECK 1-DH-V-172 2 C 0.5 CK SA

• A C 0 302-640 D-7 BDC RR RJ-22 CO RR RJ-22 Valve Name DH-V1NV2 DEAD LEG PRESSURE EQUALIZATION RELIEF CHK Revision 0 October 25, 2013 A15 - 14

Attachment 15 Inservice Testing Valve Table DECAY HEAT REMOVAL (Page 2)

Category Size Valve ,Act. Active I Normal Safety P&&ID P&&ID Valve EPN Safety Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-DH-V-18A 2 C 0.75 RV . SA A C .O/C 302-640 F-5 RT Y10 Valve Name DH LOOP A INJECTION HEADER RELIEF 1-DH-V-18B 2 C 0.75 RV SA A C O/C 302-640 F-4 RT Y10 Valve Name DH LOOP B INJECTION HEADER RELIEF 1-DH-V-2 1 A 12 GA MO A C o/C 302-640 E-7 LTH Y2 PI Y2 STC CS CSJ-02 STO CS CSJ-02 Valve Name CONTAINMENT ISOLATION - DH DROP LINE/PUMP SUCTION 1-DH-V-21 A 3 . GL M P LC LC 302-640 H-5 LT Y2 Valve Name DH PUMPS DISCHARGE TEST ISOLATION VALVE 1-DH-V-22A A/C

• 10 CK SA A C O/C 302-640 G-8 CCL RR RJ-04 COF RR RJ-04 LTH Y2 Valve Name CONTAINMENT ISOLATION - DH-P1A DISCHARGE CHK TO CF 1-DH-V-22B 1 A/C 10 CK SA A C O/C 302-640 G-8 CCL RR RJ-04 COF RR RJ-04 LTH Y2 Valve Name CONTAINMENT ISOLATION - DH-P1 B DISCH CHK TO CF 1-DH-V-3 2 . B 12 GA MO A C O/C 302-640 E-6 PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION - DH DROP LINE VLV 1-DH-V-37 2 C 0.75 RV SA A C O/C 302-640 E-7 RT Y10 Valve Name DECAY HEAT DROP LEG INTERVALVE RELIEF 1-DH-V-38A 2 B *6 GA M A LC O/C " 302-640 F-5 EO Y2 Valve.Name DECAY HEAT CROSSOVER VALVE 1-DH-V-38B 2 .B 6 GA M A LC o/C 302-640 F-5 EO Y2 Valve Name DECAY HEAT CROSSOVER VALVE 1-DH-V-4A 2 B 10 GA MO A C O/C 302-640 G-6 PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION - DH-P1A DISCHARGE ISOL VLV 1-DH-V-4B 2 B 10 GA MO A C O/C 302-640 G-6 PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION - DH-P1B DISCHARGE ISOL VLV 1-DH-V-50 2 A/C 4 CK SA A C C 302-640 B-5 BDO RR RJ-24 IST-014 CC RR RJ-24 LT Y2

• Valve Name SPENT FUEL RETURN CLEANUP CHECK VALVE Revision 0 October 25, 2013 A15 - 15

Attachment 15 Inservice Testing Valve Table DECAY HEAT REMOVAL (Page 3)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-DH-V-57A 2 C 0.75 RV SA A C O/C 302-640 D-3 RT Y10 Valve Name DH-P-1A SUCTION FROM BWST RELIEF 1-DH-V-57B 2 C 0.75 RV SA A C O/C 302-640 B-3 RT Y10 Valve Name DH-P-1 B SUCTION FROM BWST RELIEF 1-DH-V-5A 2 B 14 GA MO A 0 O/C 302-640 C-2 PI Y2 STC M3 Valve Name DECAY HEAT SUCTION VALVE FROM BWST 1-DH-V-5B 2 B 14 GA MO A 0 O/C 302-640 B-3 PI Y2 STC M3 Valve Name DH PUMP SUCTION FROM BWST 1-DH-V-60A NA C 0.5 RV SA A C O/C 302-640 C-5 RT Y10 Valve Name CA SYSTEM TO DH-P-1A SUCTION RELIEF 1-DH-V-60B NA C 0.5 RV SA A C O/C 302-640 C-3 RT Y10 Valve Name CA SYSTEM TO DH-P-1B SUCTION RELIEF 1-DH-V-64 2 A 2. GL M A LC O/C 302-640 F-6 EC Y2 EO Y2 LTJ AJ Valve Name CONTAINMENT ISOLATION - DH TO PZR SPR-AY.ISOL VLV 1-DH-V-67 2 C 0.75 RV SA . A C O/C 302-640 F-7 RT Y10 Valve Name PRESSURIZER SPRAY LINE RELIEF 1-DH-V-69 2 A/C 1.5 CK SA A C O/C 302-640 F-7 CCL CM COF CM LTJ AJ Valve Name CONTAINMENT ISOLATION - DH TO PZR AUX SPRAY LINE.

1-DH-V-6A 2 1B4 GA MO A . C O/C 302-640 C-7 PI Y2 STC RR RJ-21 STO RR RJ-21 Valve Name CONTAINMENT ISOLATION - RB SUMP RECIFRC SUCTION VLV 1-DH-V-6B 2 B 14 GA MO A C O/C 302-640 B-7 Pl Y2 STC RR RJ-21 STO RR RJ-21 Valve Name CONTAINMENT ISOLATION - RB SUMP RECIRC SUCTION I-DH-V-7A 2 B 4 GA MO A C 0 302-640 G-3 PI Y2 STO M3 Valve Name DH-CiA DISCHARGE VALVE TO MAKEUP SYSTEM 1-DH-V-7B 2 B 4 GA MO A C 0 302-640 G-3 PI Y2 STO :M3 Valve Name DH-ClB DISCHARGE VALVE TO MAKEUP SYSTEM Revision 0 October 25, 2013 A15 - 16

Attachment 15 Inservice Testing Valve Table DECAY HEAT RIVER WATER (Page 1)

Test Test Relief Deferred Tech.

Valve EPN Safety Category Size Valve Act. Active Normal Safety P&&ID P&&ID Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-DR-V-1A 3 B 20 BTF" MO A C 0 302-202 C-7 P1 Y2 STO M3 Valve Name DR-01A DISCHARGE.VALVE 1-DR-V-1B B 20 BTF. MO . A C 0 302-202 C-6 PI Y2 STO M3 Valve Name DR-P1B DISCHARGE VALVE 1-DR-V-24A 3 B 3 BALL MO P 0 0 302-203 A-4 PI Y2 Valve Name DR-S-1A BACKWASH VALVE 1-DR-V-24B 3 B 3 BALL MO" P 0 0 302-203 A-4 PI Y2 Valve Name DR-S-1B BACKWASH VALVE 1-DR-V-7A 3 C 2 CK SA. A C o/C 302-202 B-7 CC M3 CO M3 Valve Name DR-PlA COLUMN VACUUM BREAKER (Check Valve) 1-DR-V-7B 3 C. .2 cK SA A C. O/C 302-202 B-6 CC M3 CO M3 Valve Name DR-P1B COLUMN VACUUM BREAKER (Check Valve) 1-DR-V-8A 3 C 1.5 RV SA A C O/C 302-202 G-7 RT Yb0 Valve Name DECAY HEAT SERVICE COOLER DC-C-2A RELIEF 1-DR-V-8B 3 C 1.5 RV SA A C O/C 302-202 F-7 RT Y10 Valve Name DECAY HEAT SERVICE COOLER DC-C-2B RELIEF Revision 0 October 25, 2013 A15 - 17

Attachment 15 Inservice Testing Valve Table EMERGENCY FEEDWATER (Page 1)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

I-EF-V-1 1A 3 C 4 CK SA A SYS O/C 302-082 F-6 CC RR RJ-05 COF RR RJ-05 Valve Name EF-P2A DISCHARGE CHECK VALVE 1-EF-V-11B 3 C 4 CK SA A SYS O/C 302-082 B-6 CC RR RJ-05 COF RR RJ-05 Valve Name EF-P2B DISCHARGE CHECK VALVE 1-EF-V-12A 2 C 6 CK SA A SYS O/C 302-082 D-4 CCF RR RJ-05 COF RR RJ-05 Valve Name CONTAINMENT ISOLATION - EFW TO OTSG "A"CHECK VLV 1-EF-V-12B 2 C 6 CK SA A SYS O/C 302-082 B-3 CCF RR RJ-05 COF RR RJ-05 Valve Name CONTAINMENT ISOLATION - EFW TO OTSG "B"CHECK VLV 1-EF-V-13 3 C 6 CK SA A SYS O/C 302-082 E-6 CC RR RJ-07 CO RR RJ-07 Valve Name EF-P1 DISCHARGE CHECK VALVE 1-EF-V-19A 3 C 1.5 CK SA A SYS O/C 302-082 F-6 CC M3 CO M3 Valve Name EF-P2A RECIRC LINE CHECK VALVE 1-EF-V-19B 3 C 1.5 CK SA A SYS O/C 302-082 B-6 CC M3 CO M3 Valve Name EF-P2B RECIRC LINE CHECK VALVE 1-EF-V-21 3 C 2 CK SA A SYS O/C 302-082 D-6 CC M3 CO M3 Valve Name EF-P1 RECIRCULATION CHECK VALVE 1-EF-V-2A 3 B 6 GA MO A 0 O/C 302-082 F-5 PI Y2 STC M3 Valve Name EFW PUMP DISCHARGE HEADER CROSS CONNECT VALVE 1-EF-V-2B 3. B 6 GA MO A 0 O/C 302-082 D-5 PI Y2 STC M3 Valve Name EFW PUMP DISCHARGE HEADER CROSS CONNECT VALVE 1-EF-V-30A 2 B 3 PLG AO A C O/C 302-082 F-4 EC Y2 EO Y2 FC M3 IST-013 STC M3 STO M3 Valve Name EFW TO OTSG "A"FLOW CONTROL VALVE Revision 0 October 25, 2013 A15 - 18

Attachment 15 Inservice Testing Valve Table EMERGENCY FEEDWATER (Page 2)

Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Valve EPN Test Test Relief* Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-EF-V-30B 2 B 3 PLG AO A C O/C 302-082 B-5 EC Y2 EO Y2 FC M3 IST-013 STC M3 STO M3 Valve Name EFW TO OTSG "B"FLOW CONTROL VALVE 1-EF-V-30C 2 B 3 PLG AO A C O/C 302-082 B-5 EC Y2 EO Y2 FC M3 IST-013 STC M3 STO M3 Valve Name EFW TO OTSG "A"FLOW CONTROL VALVE 1-EF-V-30D 2 B 3 PLG AO A C O/C 302-082 G-5 EC Y2 EO Y2 FC M3 IST-013 STC M3 STO M3 Valve Name EFW TO OTSG "B"FLOW CONTROL VALVE 1-EF-V-35 3 C 2 RV SA A C O/C 302-082 F-1 RT Y4 Valve Name EF-P1 BEARING COOLING WTR RETURN RELIEF 1-EF-V-39A 3 C 1 RV SA A C O/C 302-082 F-3 RT Y8 Valve Name EF-P2A BEARING COOLING WTR RETURN RELIEF 1-EF-V-39B 3 C 1 RV SA A C O/C 302-082 F-2 RT Y8 Valve Name EF-P2B BEARING COOLING WTR RETURN RELIEF 1-EF-V-4 3 B 6. GA MO A C O/C 302-611 E-1 PI Y2 STO M3 Valve Name EMERGENCY RIVER WATER SUPPLY TO EFW PUMPS 1-EF-V-5 3 B 6 GA MO A C 0/C 302-611 F-1 PI Y2 STO M3 Valve Name EMERGENCY RIVER WATER SUPPLY TO EFW PUMPS Revision 0 October 25, 2013 A15- 19

Attachment 15 Inservice Testing Valve Table EMERGENCY DIESEL GENERATORS SUPPORT SYST (Page 1)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-EG-V-10A NA C 0.75 CK SA A C C 302-351 F-8 BDO. M3 IST-014 CC M3 IST-007 Valve Name EG-T1A-1 AIR START SYS. RECEIVER INLET CHECK 1-EG-V-10B NA C 0.75 CK SA A C C 302-351 F-1 BDO M3 IST-014 CC M3 IST-007 Valve Name EG-T1B-1 AIR START SYS. RECEIVER INLET CHECK 1-EG-V-10C NA C 0.75 CK SA A C C 302:351 D-8 BDO M3 IST-014 CC M3 IST-007 Valve Name EG-T1A-2 AIR START SYS. RECEIVER INLET CHECK 1-EG-V-10D NA C 0.75 CK SA A C C 302-351 D-1 BDO M3 IST-014 CC M3 IST-007 Valve Name EG-T1B-2 AIR START RECEIVER INLET CHECK 1-EG-V-16A N/A B 1.5 DIA SO A C O/C 302-351 F-6 STO M3 IST-007 Valve Name AIR START FOR DIESEL GENERATOR 1A 1-EG-V-16B N/A B 1.5 DIA SO A C O/C 302-351 F-3 STO M3 IST-007 Valve Name AIR START FOR DIESEL GENERATOR 1B 1-EG-V-16C N/A B 1.5 DIA SO A C O/C 302-351 E-6 STO M3 IST-007 Valve Name AIR START FOR EG-Y1A 1-EG-V-16D N/A B 1.5 DIA SO A C O/C 302-351 E-3 STO M3 IST-007 Valve Name AIR START FOR EG-Y1B 1-EG-V-2A N/A C 0.5 RV SA A C O/C 302-351 G-8 RT Y10 Valve Name STARTING AIR RECEIVER EG-T-1A-1 RELIEF 1-EG-V-2B N/A. C 0.5 RV SA A C O/C 302-351 G-1 RT Y10 Valve Name STARTING AIR RECEIVER EG-T-1B-1 RELIEF 1-EG-V-32A N/A C 4 CK . SA A SYS 0 302-354 C-6 D&I Y8 Valve Name EG-C3AJA COOL RADIATOR OUTLET CHECK VALVE 1-EG-V-32B N/A C 4 CK SA A SYS 0 302-354 C-3 D&I Y8 Valve Name EG-C3B/A COOL RADIATOR OUTLET CHECK VALVE Revision 0 October 25, 2013 A15 -20

Attachment 15 Inservice Testing Valve Table EMERGENCY DIESEL GENERATORS SUPPORT SYST (Page 2)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor.. Type Freq. Request. Just. Pos.

1-EG-V-32C N/A C 4 CK SA A SYS 0 302-354 C-5 D&I Y8 Valve Name EG-C3A/B COOL RADIATOR OUTLET CHECK VALVE l-EG-V-32D N/A C 4 CK SA A SYS 0 302-354 C-2 D&I Y8 Valve Name EG-C3B/B COOL RADIATOR OUTLET CHECK VALVE 1-EG-V-34A N/A C 5 CK SA A SYS C 302-354 E-5 D&I Y8 Valve Name JACKET COOLANT RADIATOR BYPASS CHECK VALVE 1-EG-V-34B N/A C 5 CK SA A SYS C 302-354 E-1 D&I Y8 Valve Name JACKET COOLANT RADIATOR BYPASS CHECK VALVE 1-EG-V-3A N/A C 0.5 RV SA A C O/C 302-351 E-8 RT Y10 Valve Name STARTING AIR RECEIVER EG-T-1A-2 RELIEF 1-EG-V-3B N/A *C 0.5 RV SA A C O/C 302-351 E-1 RT Y10 Valve Name STARTING AIR RECEIVER EG-T-1B-2 RELIEF 1-EG-V-48A N/A C 4 CK SA A .C O/C , 302-354 F-6 D&I Y8 Valve Name AIR COOLING PUMP "A"DISCHARGE CHECK VALVE.

1-EG-V-48B N/A C 4 CK SA A C O/C 302-354 F-3 D&I Y8 Valve Name AIR COOLING PUMP "B' DISCHARGE CHECK VALVE 1-EG-V-77A N/A C 0.5 CK. SA A SYS O/C 302-353 G-7 D&I Y8 Valve Name EG-Y-1A TURBO-CHARGER LUBE OIL SUPPLY CHECK 1-EG-V-77B N/A C 0.5 CK SA A SYS O/C 302-353 G-3 D&I Y8' Valve Name EG-Y-1B TURBO-CHARGER LUBE OIL SUPPLY CHECK 1-EG-V-7A N/A C RV SA A C O/C 302-353 .G-10 RT. Y10 Valve Name EG-Y-1A ENGINE-DRIVEN FO PUMP EG-P-9A DISCH RELIEF 1-EG-V-7B N/A . C -RV SA A C O/C 302-353 G-5 RT Y10 Valve Name EG-Y-1B ENGINE-DRIVEN FO PUMP EG-P-9B DISCH RELIEF Revision 0 October 25, 2013 A15 - 21

Attachment 15 Inservice Testing Valve Table FIRE SERVICE WATER (Page 1)

Valve EPN Safety Category Size Valve Act. Active I Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-FS-V-401 2 A 4 GL M P LC C 302-231-2 D-5 LTJ AJ Valve Name CONTAINMENT ISOLATION- FIRE SERVICE WATER Revision 0 October 25, 2013 A15 - 22

Attachment 15 Inservice Testing Valve Table FEEDWATER (Page 1)

Valve EPN Safety Category Size Valve Act. Active I Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request *Just. Pos.

1-FW-V-12A 2 C 20 CK SA A 0 C 302-081 B-6 BDO CM CCD CM Valve Name CONTAINMENT ISOLATION - OTSG "A"INLET CHECK VALVE 1-FW-V-12B 2 C 20 CK SA A 0 C 302-081 B-3 BDO CM CCD CM Valve Name CONTAINMENT ISOLATION - OTSG "B"INLET CHECK VALVE 1-FW-V-16A NA B 6 ANG AO A 0 C 302-081 B-6 PI Y2 STC CS CSJ-10 Valve Name MAIN FEEDWATER STARTUP FLOW CONTROL VALVE 1-FW-V-16B NA B 6 ANG AO A 0 C 302-081 B-3 PI Y2 STC CS CSJ-10 Valve Name MAIN FEEDWATER STARTUP FLOW CONTROL VALVE 1-FW-V-17A NA B 16 ANG AO A 0 C 302-081 C-6 PI Y2 STC CS CSJ-10 Valve Name MAIN FEEDWATER CONTROL VALVE 1-FW-V-17B NA B 16 ANG AO A 0 C 302-081 C-3 PI Y2 STC CS CSJ-10 Valve Name MAIN FEEDWATER CONTROL VALVE 1-FW-V-5A NA B 20 GA MO A 0 C 302-081 C-6 PI Y2 STC CS CSJ-10 Valve Name MAIN FEEDWATER "A"BLOCK VALVE 1-FW-V-5B NA B 20 GA MO A 0 C 302-081 C-3 PI Y2 STC CS CSJ-10 Valve Name MAIN FEEDWATER "B"BLOCK VALVE 1-FW-V-92A NA B 6 GA MO A 0 C 302-081 B-6 PI Y2 STC CS CSJ-10 Valve Name OTSG "A"STARTUP FEEDWATER BLOCK VALVE 1-FW-V-92B NA B 6 GA MO A 0 C 302-081 B-3 PI Y2 STC CS CSJ-10 Valve Name OTSG "B"STARTUP FEEDWATER BLOCK VALVE Revision 0 October 25, 2013 A15-23

Attachment 15 Inservice Testing Valve Table CONTAIMENT HYDROGEN MONITORING (Page 1).

Valve EPN Safety Category Size Valve Act. Active/ Normal Safety P&&ID P&&ID Test Test* Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-HM-V-1A 2 A 1 GL -SO A C O/C 302-674 F-5 FC M3 IST-013 LTJ AJ PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION - H2 MONITOR "A"OUTLET ISOL 1-HM-V-1B 2 A 1 GL SO A C O/C 302-674 C-5 FC M3 IST-013 LTJ AJ PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION - H2 MONITOR "B"OUTLET ISOL 1-HM-V-2A 2 A 1 GL SO A C O/C 302-674 E-5 FC M3 IST-013 LTJ AJ PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION- "A"H2 MONITOR INLET ISOL 1-HM-V-2B 2 A 1 GL SO A C O/C 302-674 C-5 FC M3 IST-013 LTJ AJ PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION - "B"H2 MONITOR INLET ISOL 1-HM-V-3A 2 A 1 GL SO A C O/C 302-674 F-5 FC M3 IST-013 LTJ AJ PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION- "A"H2 MONITOR OUTLET ISOL 1-HM-V-3B 2 A 1 GL SO A C O/C 302-674 C-5 FC M3 IST-013 LTJ AJ PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION - "B"H2 MONITOR INLET ISOL Revision 0 October 25, 2013 A15 - 24

Attachment 15 Inservice Testing Valve Table CONTAIMENT HYDROGEN MONITORING (Page 2)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-HM-V-4A A 1 GL SO A C O/C 302-674 E-5 FC M3 IST-013

.* LTJ AJ PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION - "A"H2 MONITOR INLET ISOL 1-HM-V-4B 2 A 1 GL SO A C O/C 302-674 C-5 FC M3 IST-013 LTJ AJ PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION -"B"H2 MONITOR INLET ISOL Revision 0 October 25, 2013 A15 - 25

Attachment 15 Inservice Testing Valve Table HYDROGEN PURGE DISCHARGE SYSTEM (Page 1)

Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Valve EPN Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-HP-V-1 2 A 6 GA *M P LC C 302-721 *B-7 LTJ AJ Valve Name CONTAINMENT ISOLATION - H2 PURGE ISOLATION VALVE 1-HP-V-6 2 A 6 GA M P LC *C 302-721 B-7 LTJ AJ Valve Name CONTAINMENT ISOLATION - H2 PURGE RB ISOL VALVE Revision 0 October 25, 2013 A15 - 26

Attachment 15 Inservice Testing Valve Table POST LOCA HYDROGEN RECOMBINER SYSTEM (Page 1)

Valve EPN Safety Category Size Valve Act. Active! Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-HR-V-22A 2 A 2 GL SO P C C 302-722 D-7 LTJ AJ PI Y2 Valve Name CONTAINMENT ISOLATION -"RB EXHAUST TO H2 RECOMB 1-HR-V-22B 2 A 2 GL SO P C, C 302-722 C-7 LTJ AJ PI Y2 Valve Name CONTAINMENT ISOLATION - RB EXHAUST TO H2 RECOMB 1-HR-V-23A 2 A 2 GL SO P C C 302-722 C-7 LTJ AJ PI Y2 Valve Name CONTAINMENT ISOLATION - H2 RECOMB RETURN ISOL VLV 1-HR-V-23B 2 A 2 GL SO P C C 302-722 B-7 LTJ AJ PI Y2 Valve Name CONTAINMENT ISOLATION - H2 RECOMB RETURN ISOL VLV 1-HR-V-2A. 2 A 2 GL M P LC C 302-722 E-5 LTJ AJ Valve Name CONTAINMENT ISOLATION - HR-R1NAB RB SUPPLY VALVE 1-HR-V-2B. 2 A 2 GL M P LC C 302-722 E-5 LTJ AJ Valve Name CONTAINMENT ISOLATION - HR-R1AIB RB SUPPLY 1-HR-V-4A 2 A 2 GL M P. LC C 302-722 F-5 LTJ AJ Valve Name CONTAINMENT ISOLATION - HR-RIAB RB RETURN VALVE 1-HR-V-4B 2 A 2 GL M P LC C 302-722 F-5 LTJ AJ Valve Name CONTAINMENT ISOLATION - HR-R1ANB RB RETURN VALVE Revision 0 October 25, 2013 A15 -27

Attachment 15 Inservice Testing Valve Table INSTRUMENT AIR (Page 1)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-IA-V-1621A N/A B 1 GA AO A T T 302-273 C-7 FO M3 Valve Name 2-HR BACKUP AIR SYSTEM "A"PRESSURE REDUCER 1-IA-V-1621B N/A B 1 GA AO A T T 302-273 C-3 FO M3 Valve Name 2-HR BACKUP AIR SYSTEM "B"PRESSURE REDUCER 1-IA-V-1624A N/A C 1.5 RV SA A C O/C 302-273 C-7 RT Y8 Valve Name 2-HR BACKUP AIR SYSTEM "A"RELIEF 1-IA-V-1624B N/A C 1.5 RV SA A C O/C 302-273 C-4 RT Y8 Valve Name 2-HR BACKUP AIR SYSTEM "B"RELIEF 1-IA-V-1625A N/A B 0.75 3W AO A 0 0 302-273 C-6 FO RR RJ-08 IST-013 SO RR RJ-08 Valve Name 2-HR BACKUP AIR SYSTEM "A"HEADER VENT VALVE 1-IA-V-1625B N/A B 0.75 3W AO A 0 0 302-273 C-4 FO RR RJ-08 IST-013 SO RR RJ-08 Valve Name 2-HR BACKUP AIR SYSTEM "B"HEADER VENT VALVE 1-IA-V-1626A N/A B 0.75 3W AO A 0 0 302-273 C-6 FO RR RJ-08 IST-013 SO RR RJ-08 Valve Name 2-HR BACKUP AIR SYSTEM HEADER SUPPLY VALVE 1-IA-V-1626B N/A B 0.75 3W AD A 0 0 302-273 C-4 FO RR RJ-08 IST-013 SO RR RJ-08 Valve Name 2-HR BACKUP AIR SYSTEM HEADER SUPPLY VALVE 1-IA-V-1628A NA C 1 CK SA A SYS C 302-273 C-6 BDO RR RJ-06 IST-014 CC RR RJ-06 Valve Name 2-HR BACKUP AIR SYSTEM "A"INST AIR SUPPLY CHECK 1-IA-V-1628B NA C 1 CK SA A SYS C 302-273 C-5 BDO RR RJ-06 IST-014 CC RR RJ-06 Valve Name 2-HR BACKUP AIR SYSTEM "B"INST AIR SUPPLY CHECK 1-IA-V-1631A N/A C. 1 CK SA A C O/C 302-273 D-6 CC RR RJ-06 CO RR RJ-06 Valve Name 2-HR BACKUP AIR SYSTEM "A"SUPPLY CHECK 1-IA-V-1631B N/A C 1 CK SA A C 0 302-273 E-5 BDC RR RJ-06 IST-014 CO RR RJ-06 Valve Name 2-HR BACKUP AIR SYSTEM CROSSOVER CHECK 1-IA-V-1632 N/A B 1 BALL M A C O/C 302-273 E-5 EO Y2 Valve Name 2-HR BACKUP AIR SYSTEM CROSSOVER CHECK 1-IA-V-20 2 A 2 GL M P LC LC 302-271 F-1 LTJ AJ Valve Name CONTAINMENT ISOLATION - IAINNER RB ISOL VALVE 1-IA-V-6 2 A 2 GL M P LC LC 302-271 F-1 LTJ AJ Valve Name CONTAINMENT ISOLATION - RB INST AIR CONN VALVE Revision 0 October 25, 2013 A15 -28

Attachment 15 Inservice Testing Valve Table INTERMEDIATE COOLING WATER SYSTEM (Page 1)

Valve EPN Safety Category Size Valve Act. Active/. Normal Safety .P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

l:iC-V-10 3 C 0.5 RV SA .A C '0 302-620 RT Y10 Valve Name INTER CLG - RCDT COOLER RELIEF VALVE.

1-IC-V-102 2 A/C 0.25 RV SA A C O/C 302-620 C-5 LTJ AJ RT Y10 Valve Name CONTAINMENT PENETRATION RELIEF VALVE 1-IC-V-16 2 A/C 3 CK SA A 0 O/C 302-620 G-6 CCL CM COF CM LTJ AJ Valve Name CONTAINMENT ISOLATION - CRD COOLING CHECK VALVE 1-IC-V-18 2 A/C. 6 CK SA A 0 O/C 302-620 H-6 CCL CM COF CM LTJ AJ Valve Name CONTAINMENT ISOLATION - LETDOWN SUPPLY CHECK VLV 1-IC-V-2 2 A 6 GA MO A 0 C 302-620 C-6 LTJ AJ PI Y2 SP M3 CSJ-01 STC CS CSJ-01 Valve Name CONTAINMENT ISOLATION - ICCW COOLANT RETURN VALVE 1-IC-V-3 2 A 6 PLG AO A 0 C 302-620 C-5 LTJ AJ P1 Y2 SP M3 cSJ-01 STC CS CSJ-01 Valve Name CONTAINMENT ISOLATION - ICCW COOLANT RETURN VALVE 1-IC-V-4 2 A 6 PLG AO A 0 C 302-620 H-5 LTJ AJ PI Y2 SP M3 CSJ-01 STC CS CSJ-01 Valve Name CONTAINMENT ISOLATION - ICISOL COOLANT SUPPLY 1-IC-V-6 2 A 3 GA AO A 0 C 302-620 G-5 FC CS CSJ-01 IST-013 LTJ AJ PI Y2 SP M3 CSJ-01 STC CS CSJ-01 Valve Name CONTAINMENT ISOLATION - ICCOOLANT SUPPLY TO CRDM 1-IC-V-76 2 C 0 RV SA A C O/C 302-620 D-6 RT Y10 Valve Name CRD COOLING COIL RELIEF VALVE 1-IC-V-90A 3 C 0.75 RV SA A C O/C 302-620 E-8 RT Y10 Valve Name RC PUMP 1A COOLER RELIEF VALVE 1-IC-V-90B 3 C 0.75 RV SA A C O/C 302-620 D-8 RT Y10 Valve Name RC PUMP 1B COOLER RELIEF VALVE 1-IC-V-90C 3 C 0.75 RV SA A C O/C 302-620 D-8 RT Y10 Valve Name RC PUMP 1A COOLER RELIEF VALVE 1-IC-V-90D 3 C 0.75 RV SA A C O/C 302-620 C-8 RT Y10 Valve Name RC PUMP 1A COOLER RELIEF VALVE Revision 0 October 25, 2013 A15-29

Attachment 15 Inservice Testing Valve Table MAIN STEAM (Page 1)

Valve EPN Safety Category Size Valve Act. Active I Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-MS-V-10A 3 B 6 GA MO A C C 302-011 F-3 PI Y2 STO RR Valve Name OTSG "A"TO EF-P1 JOG/THROTTLE VALVE 1-MS-V-10B 3 B 6 GA MO A C C 302-011 F-3 PI Y2 STO RR Valve Name OTSG "B"TO EF-P1 JOG/THROTTLE VALVE 1-MS-V-13A 3 B 2 GL AO A C O 302-011 F-3 FO M3 IST-013 PI Y2 STO M3 STC M3 Valve Name MAIN STEAM SUPPLY TO EF-P1 FROM OTSG "A" 1-MS-V-13B 3 B 2 GL AO A C O 302-011 F-3 FO M3 IST-013 PI Y2 STO M3 STC M3 Valve Name MAIN STEAM SUPPLY TO EF-P1 FROM OTSG "W" 1-MS-V-17A 2 C 6 RV SA A C O/C 302-011 G-8 RT Y5 Valve Name OTSG "A"MS RELIEF VjrALVE 1-MS-V-17B C 6 RV SA A C O/C 302-011 G-8 RT Y5 Valve Name OTSG "A"MS RELIEF V,4ALVE 1-MS-V-17C C 6 RV SA A C O/C 302-011 H-8 RT Y5 Valve Name OTSG "B"MS RELIEF VALVE 1-MS-V-17D C 10 RV SA A C O/C 302-011 H-8 RT Y5 Valve Name OTSG MS RELIEF VAL' FE 1-MS-V-18A C 6 RV SA A *C O/C 302-011 G-9 RT Y5 Valve Name OTSG "A"MS RELIEF VALVE 1-MS-V-18B C 6 RV SA. A C O/C 302-011 G-9 RT Y5 Valve Name OTSG "A"MS RELIEF V.ALVE 1-MS-V-18C C 6 RV SA A C O/C 302-011 H-9 RT Y5 Valve Name OTSG "B"MS RELIEF V.ALVE 1-MS-V-18D C 6 RV SA A C O/C 302-011 H-9 RT Y5 Valve Name OTSG "B"MS RELIEF VALVE 1-MS-V-19A C 6 RV SA A C O/C 302-011 G-9 RT Y5 Valve Name OTSG "A"MS RELIEF VALVE 1-MS-V-19B C .6 RV SA A .C O/C 302-011 G-9 RT Y5 Valve Name OTSG "A"MS RELIEF VALVE 1-MS-V-19C C 6 RV SA A C O/C 302-011 H-9 RT Y5 Valve Name OTSG "B"MS RELIEF VALVE 1-MS-V-19D C *6 RV SA A C

• O/C 302-011 H-9 RT Y5 Valve Name OTSG "B"MS RELIEF V Revision 0 October 25, 2013 A15-30

Attachment 15 Inservice Testing Valve Table MAIN STEAM (Page 2)

Valve EPN Safety Category Size Valve Act. Active I Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-MS-V-1A 2 B/C 24 SCK MO A 0 C 302-011 G-10 BDO M3 ISTr014 CC RR RJ-10 PI Y2 STC CS CSJ-16 Valve Name CONTAINMENT ISOLATION - OTSG "A"MS ISOL VALVE 1-MS-V-lB 2 B/C 24 SCK MO A 0 C 302-011 G-10 BDO M3 IST-014 CC RR RJ-10 PI Y2 STC CS CSJ-16 Valve*Name CONTAINMENT ISOLATION - OTSG "A"MS ISOL VALVE 1-MS-V-1C 2 B/C 24 SCK MO A 0 C 302-011 G-10 BDO M3 IST-014 CC RR RJ-10 PI Y2.

• .STC CS CSJ-16 Valve Name CONTAINMENT ISOLATION - OTSG "B"MS ISOL VALVE'.

1-MS-V-1D 2 B/C 24 SCK MO A 0 C 302-011 H-10 BDO M3 IST-014 CC RR RJ-10 PI Y2 STC CS CSJ-16 Valve Name CONTAINMENT ISOLATION - OTSG "B"MS ISOL VALVE 1-MS-V-20A 2. C 6 RV SA A C O/C 302-011Oi G-9 RT Y5 Valve Name OTSG "A"MS RELIEF VALVE 1-MS-V-20B 2 C 6 RV SA. A C O/C 302-01, G-9 RT Y5 Valve Name OTSG "A"MS RELIEF VALVE 1-MS-V-20C 2 C 6 'RV "SA A C o/C- 302-011 H-9 RT Y5 Valve Name OTSG "B"MS RELIEF VALVE I-MS-V-20D 2 C 6 RV SA A C O/C 302-011 H-9 RT Y5 Valve Name OTSG "B"MS RELIEF VALVE 1-MS-V-21A 2 C 3 RV SA A C O/C 302-011 G-9 RT Y5 Valve Name OTSG "A"MS RELIEF VALVE 1-MS-V-21B 2 C .3 RV SA A C O/C 302-011 H-9 RT Y5 Valve Name OTSG "B"MS RELIEF VALVE.

1-MS-V-22A 3 C 4 RV SA A C O/C 302-011 F-5 RT Y8 Valve Name EMERGENCY FW PUMP EF-P1 STEAM SUPPLY RELIEF 1-MS-V-22B 3 C. .4 RV SA .A C O/C. 302-011 F-5 RT Y8 l..Vave Name EMERGENCY FW PUMP EF-P1 STEAM SUPPLY RELIEF 1-MS-V-2A 2 B 12 GA MO A 0 O/C 302-011 F-4 PI  :.Y2 STC M3 Valve Name OTSG "A"MS TO EF-Pi &TURBINE BYPASS VALVES 1-MS-V-2B 2 B 12 GA MO A. 0 O/C 302-011 G-4 PI Y2 STC M3 Valve Name OTSG "B"MS TO EF-P1 &TURBINE BYPASS VALVES Revision 0 October 25, 2013 A15 - 31

Attachment 15 Inservice Testing V'alve Table MAIN STEAM (Page 3)

Valve EPN Safety Category Size Valve Act. Active / Normal S afety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Po sition Coor. Type Freq. Request Just. Pos.

1-MS-V-33A 2 B 1 GL M A 0 C 302-121 D-6 EC Y2 Valve Name MS LINE A TRAP INLET ISOLATION 1-MS-V-33B 2 B 1 GL M A 0 C 302-121 D-6 EC Y2 Valve Name MS LINE B TRAP INLET ISOLATION .

1-MS-V-33C 2 B 1 GL M A 0 C 302-121 D-6 EC Y2 Valve Name MS LINE C TRAP INLET ISOLATION 1-MS-V-33D 2 B 1 GL M A 0. C 302-121 D-6 EC Y2 Valve Name MS LINE DTRAP INLET ISOLATION 1-MS-V-42A 2 B 1 GL M A 0. C 302-121 D-5 EC Y2 Valve Name MS HEADAER ATRAP INLET ISOLATION 1-MS-V-42B 2 B 1 GL M A 0 C 302-121 D-5 EC Y2 Valve Name MS HEADAER B TRAP INLET ISOLATION 1-MS-V-4A 3 B 6 GL AO A C C 302-011 F-4 FC M3 PI Y2 STC M3 Valve Name ATMOSPHERIC DUMP VALVE FOR OTSG "A" 1-MS-V-4B 3 B 6 GL AO A C C 302-011 G-3 FC M3 PI Y2 STC M3 Valve Name ATMOSPHERIC DUMP VALVE FOR OTSG "B" 1-MS-V-6 3 B 4 GL AO A 0 O 302-011 F-5 EC Y2 EO Y2 FO M3 IST-013 STO M3 IST-007 Valve Name EF-Pi MS PRESSURE.REGULATOR CONTROLLVALVE 1-MS-V-89A 2 B 0.5 GL M A 0 C 302-012 F-4 EC Y2 Valve Name MS LINE APOST HEATING TRAP ISOLATION 1-MS-V-89B 2 B 0.5 GL M *A 0 C 302-012 D-4 EC Y2 Valve Name MS LINE B POST HEATING TRAP ISOLATION 1-MS-V-89C 2 B 0.5 GL. M A 0 C 302-012 E-4 EC Y2 Valve Name MS LINE C POST HEATING TRAP ISOLATION 1-MS-V-89D 2 B 0.5 GL M A 0 C 302-012 B-4 EC Y2 Valve Name MS LINE 0 POST HEATING TRAP ISOLATION 1-MS-V-92 2 B 2 GL M A 0 C 302-012 C-3 EC Y2 Valve Name COMMON POST HEATING DRAIN ISOLATION 1-MS-V-9A 3 C 6 CK SA A SYS O/C 302-011 F-3 CCD CM COD CM CP CM Valve Name MAIN STEAM SUPPLY CHECK VALVE TO EF-UU1 1-MS-V-9B 3 C 6 CK SA A SYS O/C 302-011 F-3 CCD CM COD CM CP .CM Valve Name MAIN STEAM SUPPLY CHECK VALVE TO EF-Ul Revision 0 October 25, 2013 A15 - 32

Attachment 15 Inservice Testing Valve Table MAKEUP & PURIFICATION (Page 1)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-MU-RV-1 3 C 2 RV SA A C O/C 302-660 A-7 RT Y4 Valve Name MAKEUP TANK MU-T-1 RELIEF 1-MU-V-107A 1 C 2.5 CK SA A C O/C 302-661 H-2 CCD CM COD CM COF CM Valve Name CONTAINMENT ISOLATION - HPI TO RC "A"SUPPLY CHECK 1-MU-V-107B 1 C 2.5 CK SA A C O/C 302-661 G-2 CCD CM COD CM COF CM Valve Name CONTAINMENT ISOLATION - HPI TO RC "B"SUPPLY CHECK 1-MU-V-107C 1 C 2.5 CK SA A C O/C 302-661 D-2 CCD CM COD CM COF CM Valve Name CONTAINMENT ISOLATION - HPI TO RC "C" SUPPLY CHECK 1-MU-V-107D 1 C 2.5 CK SA A C O/C 302-661 D-2 CCD CM COD CM COF CM Valve Name CONTAINMENT ISOLATION - HPI TO RC "D"SUPPLY CHECK 1-MU-V-112 2 A/C 4 CK SA A 0 C 302-661 B-5 CC RR RJ-12 CO M3 LT Y2 Valve Name MAKEUP TANK OUTLET CHECK VALVE 1-MU-V-116 1 A/C 1.5 CK SA A 0 C 302-661 F-3 BDO CM CCL CM LTJ AJ Valve Name CONTAINMENT ISOLATION - SEAL INJ SPRY/RC CHK 1-MU-V-14A 2 A/C 6 SCK MO A C O/C 302-661 B-4 CC RR RJ-13 CO RR RJ-13 LT Y2 PI Y2 STO M3 Valve Name MU PUMP SUCTION FROM BWST STOP CHECK VALVE 1-MU-V-14B 2 A/C 6 SCK MO A C O/C 302-661 B-2 CC RR RJ-13 CO RR RJ-13 LT Y2 PI Y2 STO M3 Valve Name MU PUMP SUCTION FROM BWST STOP CHECK VALVE Revision 0 October 25, 2013 A15 - 33

Attachment 15" Inservice Testing Valve Table MAKEUP & PURIFICATION (Page 2)

Valve EPN Safety. Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-MU-V-16A 2 B 2.5 GL MO A C .O/C 302-661 H-4 PI Y2 STC M3 STO. M3 Valve Name CONTAINMENT ISOLATION - HPI "A"CONTROL VALVE 1-MU-V-16B 2 B 2.5 .GL MO A C O/C 302-661 G-4 PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION - HPI "B"CONTROL VALVE 1-MU-V-16C 2 B 2.5 GL MO A C O/C 302-661 D-2 PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION - HPI "C" CONTROL VALVE 1-MU-V-16D 2 B 2.5 GL MO A C O/C 302-661 D-2 PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION - HPI "D"CONTROL VALVE 1-MU-V-18* 2 A 2.5 GA AO A 0 C 302-661 G-4 FC CS CSJ-03 IST-013 LTJ AJ PI Y2 SP M3 STC CS CSJ-03 Valve Name CONTAINMENT ISOLATION - CHARGE LINE ISOL VALVE 1-MU-V-193A 2 C 2 SCK SA A SYS O/C 302-661 D-5 CC RR RJ-27 CO RR RJ-27 CP .M3 Valve Name MU-P1A RECIRC.STOP CHECK VALVE 1-MU-V-193B 2 C 2 SCK SA. A SYS O/C 302-661 D-4 CC RR RJ-27 CO RR RJ-27 CP M3 Valve Name MU-P1B RECIRC STOP CHECK VALVE 1-MU-V-193C 2 C 2 SCK SA A SYS O/C 302-661 D-3 CC RR RJ-27 CO RR RJ-27 CP M3 Valve Name MU-PlC RECIRC STOP CHECK VALVE 1-MU-V-20 2 A 4 GA AO A 0 C 302-661 F-3 FC CS CSJ-04 IST-013 LTJ AJ PI Y2 STC CS CSJ-04 Valve Name CONTAINMENT ISOLATION - RCP SEAL WATER ISOL VLV 1-MU-V-219 1 A/C 2.5

• CK SA A 0 C 302-661 H-2 BDO CM CCL CM LTJ AJ Valve Name NORMAL MAKEUP LINE TO HPI LINE B CHECK

'Revision 0 October 25, 2013 A15 - 34

Attachment 15 Inservice Testing Valve Table MAKEUP & PURIFICATION (Page 3)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-MU-V-220 1 C 2.5 CK SA. A C 0 302-661 G-2 BDC CM COD CM Valve Name HPI LOOP "B"BACK FLOW FROM MAKEUP - CHECK VALVE 1-MU-V-238 2 C 0.25 RV SA A C O/C 302-660 F-5 LTJ AJ RT Y10 Valve Name RCS LETDOWN CONTAINMENT PENETRATION RELIEF 1-MU-V-25 2 A 4 GL MO A 0 C 302-660 C-3 LTJ AJ PI Y2 SP M3 CSJ-05 STC CS CSJ-05 Valve Name CONTAINMENT ISOLATION RCP SEAL RETURN ISOL VALVE 1-MU-V-26 2 A 4 GA AO A 0 C 302-660 C-4 FC CS CSJ-05 IST-013 FO CS CSJ-05 LTJ AJ PI Y2 SP M3 CSJ-05 STC CS CSJ-05 Valve Name CONTAINMENT ISOLATION RCP SEAL RETURN LETDOWN ISOL 1-MU-V-2A A 2.5 GL MO A 0 C 302-660 GA LTJ AJ PI Y2 SP M3 CSJ-06 STC CS CSJ-06 Valve Name CONTAINMENT ISOLATION - LETDOWN CLR "A"OUTLET VLV 1-MU-V-2B 1 2.5 GL MO A 0 C 302-660 F4 LTJ AJ PI Y2 SP M3 CSJ-06 STC CS CSJ-06 Valve Name CONTAINMENT ISOLATION - LETDOWN CLR "B"OUTLET VLV 1_MU-V-3 2 A .2.5 GA AO A 0 C 302-660 G-5 FC CS CSJ-06 IST-013

• LTJ AJ PI Y2 SP M3 CSJ-06 STC CS CSJ-06 Valve Name CONTAINMENT ISOLATION - LETDOWN COOLER ISOL VALVE 1-MU-V-36 2 A 2 GA MO A 0 O/C 302-661 D-6 LT Y2 PI Y2 STC M3 STO M3 Valve Name MU PUMPS RECIRC ISOLATION VALVE Revision 0 October 25, 2013 A15-35

Attachment 15 Inservice Testing Valve Table MAKEUP & PURIFICATION (Page 4)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-MU-V-37 3 A 2 GA MO A 0 O/C 302-661 D-7 LT Y2 PI Y2 STC M3 STO .M3 Valve Name MU PUMPS RECIRC ISOL VALVE 1-MU-V-51 3 B 1 DIA AO A C 0 302-660 C-10 PI Y2 STO M3 Valve Name EMERGENCY BORIC ACID ADD VALVE TO MAKEUP TANK 1-MU-V-54A 3 C 0.75 CK SA A C c .302-660 D-10 CCP CM CO CM Valve Name LITHIUM HYDROXIDE PUMP TO MU SYSTEM CHECK 1-MU-V-54B. 3 C 0.75 CK . SA A C C 302-660 D-10 CCP CM CO CM Valve Name HYDRAZINE PUMP TO MU SYSTEM CHECK 1-MU-V-73A 2 C 3 CK SA A SYS O/C 302-661 C-5 CC RR RJ-15 CO RR RJ-15 Valve Name MU-P1A DISCHARGE CHECK VALVE 1-MU-V-73B 2 C 3 CK SA A SYS O/C .302-661 C-4 CC RR RJ-15 CO RR RJ-15 Valve Name MU-P1B DISCHARGE CHECK VALVE 1-MU-V-73C 2. C 3 CK SA A SYS O/C 302-661 C-3 CC RR RJ-15 CO RR RJ-15 Valve Name MU-PlC DISCHARGE CHECK VALVE 1-MU-V-86A 1 C 2.5 CK SA. A . C 0 302-661 D-1 BDC CM COD CM Valve Name LOOP"D" RCS HPI INLET CHECK VALVE 1-MU-V-86B 1. C 2.5 CK SA A C 0 302-661 D-1 BDC CM COD CM Valve Name LOOP "C' RCS HPI INLET CHECK VALVE 1-MU-V-94 1 C 2.5 CK SA A SYS 0. 302-661. G-1 BDC CM COD CM

.Valve Name LOOP "B"RCS HPI INLET CHECK VALVE 1-MU-V-95 1 C. 2.5 CK SA A C o 302-661 H-1 BDC CM COD CM Valve Name LOOP "A"RCS HPI INLET CHECK VALVE Revision 0 October 25, 2013 A15 -36

Attachment 15 Inservice Testing Valve Table NUCLEAR PLANT NITROGEN SUPPLY (Page 1)

Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Valve EPN Class Type Type *Passive Position Position Coor. Type Freq. Request Just. Pos.

1-NI-V-26 2 A 1 GL M P LC .C 302-720 D-9 LTJ AJ Valve Name CONTAINMENT ISOLATION - 650# N2 SUPPLY TO RB 1-NI-V-27 2 A 1 GL M P LC C 302-720 D-9 LTJ AJ Valve Name CONTAINMENT ISOLATION - 650# N2 SUPPLY TO RB Revision 0 October 25, 2013 A15 -37

Attachment 15 Inservice Testing Valve Table NUCLEAR SERVICES RIVER WATER (Page 1)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-NR-V-16B 3. B 12 BTF MO P O/C/T Al 302-202 G-9 PI Y2 Valve Name NUCLEAR SERVICES HX NS-C-1B OUTLET ISOLATION 1-NR-V-16C 3 B 12 BTF MO P O/CIT Al 302-202 F-9 PI Y2 Valve Name NUCLEAR SERVICES HX NS-C-1C OUTLET ISOLATION 1-NR-V-16D 3 B 12 BTF MO P O/CIT Al 302-202 F-9 PI Y2 Valve Name NUCLEAR SERVICES HX NS-C-1D OUTLET ISOLATION 1-NR-V-19 3 B 24 BTF MO A C 0 302-202 E-5 EO Y2 PI Y2 Valve Name NR SYSTEM OUTLET TO SCREEN WASH HEADER 1-NR-V-IA 3 B 16 BTF MO A O/C 0 302-202 C-10 PI Y2 STO M3 Valve Name NUC RIVER PUMP "A"DISCH VALVE 1-NR-V-tB 3 B 16 BTF MO A O/C 0 302-202 C-9 PI Y2 STO M3 Valve Name NUC RIVER PUMP "8" DISCH VALVE 1-NR-V-IC 3 B 16 BTF MO A O/C 0 302-202 C-9 PI Y2 STO M3 Valve Name NR-PlC DISCH VALVE 1-NR-V-2 3 A 30 BTF MO P C C 302-202 C-9 LT Y2 PI Y2 Valve Name NR TO SR HEADER ISOLATION VALVE 1-NR-V-20A 3 C 16 CK SA A SYS O/C 302-202 B-10 CC M3 CO M3 Valve Name NR-P1A DISCHARGE CHECK VALVE 1-NR-V-20B 3. C 16 CK SA A SYS O/C 302-202 B-9 CC M3 CO M3 Valve Name NR-P1B DISCHARGE CHECK VALVE 1-NR-V-20C 3 C 16 CK SA A SYS O/C 302-202 B-9 CC M3 CO M3 Valve Name NR-P1C DISCHARGE CHECK VALVE 1-NR-V-21A 3 C 0.5 RV SA A C O/C 302-202 G-8 RT Y10 Valve Name INTERMEDIATE SERVICE COOLER IC-C-lA RELIEF 1-NR-V-21B 3 C 0.5 RV SA A C O/C 302-202 F-8 RT Y10 Valve Name INTERMEDIATE SERVICE COOLER IC-C-lB RELIEF Revision 0 October 25, 2013. A15 - 38

Attachment 15 Inservice Testing Valve Table NUCLEAR SERVICES RIVER WATER (Page 2)

Valve EPN Safety. Category Size Valve Act... Active/ Normal Safety . p&&ID P&&ID Test Test Relief. Deferred Tech.

Class Type Type Passive Position Position  : Coor. Type Freq. Request Just. Pos.

1-NR-V-22A 3 C 2 CK SA A C O/C 302-202 B-10 CC M3 CO .M3 Valve Name NR-P1A VACUUM BREAKER (Check Valve)

I-NR-V-22B 3 C 2 CK :SA A C O/C 302-202 B-10 CC M3 CO M3 Valve Name NR-P1 BVACUUM BREAKER (Check Valve) 1-NR-V-22C 3 C 2 CK SA A C O/C 302-202 B-9 CC M3 CO M3 Valve Name NR-P1C VACUUM BREAKER (Check Valve) 1-NR-V-25A 3 C 1.5 RV SA A C O/C 302-202 H-10 RT Y10 Valve Name NUCLEAR SERVICES HX NS-C-1A TUBE SIDE RELIEF.

1-NR-V-25B 3 C 1.5 RV SA.. A C O/C 302-202 G-10 RT Y10 Valve Name NUCLEAR SERVICES HX NS-C-1B TUBE SIDE RELIEF 1-NR-V-25C 3 C

• 1.5 RV SA A C OC O. 302-202 .G-10 RT Y10 Valve Name NUCLEAR SERVICES HX NS-C-1C TUBE SIDE RELIEF

!-NR-V-25D 3 C 1.5 RV SA A C O/C 302-202 F-10 RT Y10 Valve Name NUCLEAR SERVICES HX NS-C-1D TUBE SIDE RELIEF.

1-NR-V-4A 3 A 30 BTF MO A C C 302-202 E-10 LT Y2 PI Y2 STC M3 Valve Name DEICING MAKEUP VALVE "A" 1-NR-V-4B 3 A .30 BTF MO A C C 302-202 E-10 LT Y2 PI Y2 STC .M3 Valve Name DEICING MAKEOP VALVE "B" 1-NR-V-53A B 2 BALL MO .P 0 0 302-203 A-5 PI Y2 Valve Name NR-S-1A BACKWASH VALVE 1-NR-V-53B B *2 BALL MO P 0 0 302-203 A-5 PI Y2 Valve Name NR-S&1B BACKWASH VALVE

• PI 1-NR-V-53C B 2. BALL MO P 0 0 .302-203 A-5 Y2 Valve Name NR-S-1C BACKWASH VALVE 1-NR-V-6 A 30 BTF MO

• P C C 302-202 E-8 LT Y2 PI Y2 Valve Name HX VAULT CROSS CONNECT BETWEEN NR &SR 1-NR-V-8B B 12 BTF MO P 0 0 302-202 G-10 P1 Y2 Valve Name NUCLEAR SERVICES HX NS-C-11B INLET ISOLATION 1-NR-V-8C B 12 BTF " MO P 0 0 302-202 F-10 PI Y2 Valve Name NUCLEAR SERVICES HX NS-C-lC INLET ISOLATION 1-NR-V-8D B 12 . BTF MO P 0 0 302-202 F-10 PI. Y2 Valve Name NUCLEAR SERVICES HX NS-C-ID INLET ISOLATION Revision 0 October 25, 2013 A15 - 39

Attachment 15 Inservice Testing Valve Table NUCLEAR SERVICES CLOSED COOLING WATER (Page 1)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-NS-V-10A 3 C 12 CK SA A SYS O/C 302-610 G-3 CCF CM COD CM CPF CM Valve Name NUC SVCS PUMP "A"DISCHARGE CHECK VALVE 1-NS-V-10B 3 C 12 .CK SA A .SYS O/C 302-610 G-3 CCF CM COD CM CPF CM Valve Name N.S-P1B DISCHARGE CHECK VALVE 1-NS-V-10C 3 C 12 CK SA A SYS O/C 302-610 G-2 CCF CM COD CM CPF CM Valve Name NS-P1C DISCHARGE CHECK VALVE 1-NS-V-11 2 A/C 8 CK SA A 0 C 302-610 D-9 BDO CM CCL CM LTJ AJ Valve Name RCP MOTOR COOLER SUPPLY VALVE 1-NS-V-135 3 C 0.5 XFC SA A 0 C 302-610 E-9 BDO M3 IST-014 CC M3 Valve Name NS/RR CROSS-CONN EXCESS FLOW VALVE 1-NS-V-15 2 A 8 GA MO A 0 C 302-610 D-8 LTJ AJ PI Y2 SP M3 CSJ-07 STC CS CSJ-07 Valve Name RC MOTOR COOLER RETURN VALVE 1-NS-V-16A 3 B 8 GA MO A O/C 0 302-610 D-3 PI Y2 STO M3 Valve Name SPENT FUEL POOL COOLER SF-C-1A COOLING WATER INLET

.1-NS-V-16B 3 B 8 GA MO A 0/c 0 302-610 D-2 PI Y2 STO M3 Valve Name SPENT FUEL POOL COOLER SF-C-1B COOLING WATER INLET 1-NS-V-205 3 C 2 CK SA *A SYS C 302-610 H-6 BDO M3 IST-014 CC M3 Valve Name NSCCW SURGE TANK DEMIN WATER SUPPLY CHECK VALVE 1-NS-V-21 1 2 A/C 0.25 RV SA A C O/C 302-610 A-8 LTJ AJ RT Y10 Valve Name RB RETURN HEADER PENETRATION RELIEF Revision 0 October 25, 2013 A15-40

Attachment 15 Inservice Testing Valve Table NUCLEAR SERVICES CLOSED COOLING WATER (Page 2)

Valve EPN Safety Category Size Valve Act. Active I Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-NS-V-35 2 A 8 GA MO A 0 C 302-610 A-9 LTJ AJ PI Y2 SP M3 CSJ-08 STC CS CSJ-08 Valve Name RC MOTOR COOLER RETURN VALVE 1-NS-V-36A 2 C 0.5 RV SA A C O/C 302-610 F-8 .RT Y10 Valve Name RB FAN MOTOR COOLER 1A RELIEF 1-NS-V-36B 2 C 0.5 RV SA A C O/C 302-610 F-7 RT Y10 Valve Name RB FAN MOTOR COOLER 1B RELIEF 1-NS-V-36C 2 C 0.5 RV SA A C O/C 302-610 F-7 RT Y10 Valve Name RB FAN MOTOR COOLER 1C RELIEF 1-NS-V-37A 3 C 0.5 RV - SA A C O/C 302-610 F-5 RT Y10

• Valve Name STEAM GENERATOR SAMPLE COOLER CA-C-2A RELIEF 1-NS-V-37B 3 C 0.5 RV SA A C O/C 302-610 F-5 RT Y10 Valve Name STEAM GENERATOR SAMPLE COOLER CA-C-2B RELIEF 1-NS-V-3. 3 C 0.5 RV SA " A C O/C .302-610 F-5 RT Y10 Valve Name PRESSURIZER SAMPLE COOLER CA-C-1 RELIEF 1-NS-V-39A 3 C 0.5 RV SA A C O/C :302-610 C-2 RT Y10 Valve Name SPENT FUEL POOL COOLER SF-C-IA TUBE SIDE RELIEF 1-NS-V-39B 3 C 0.5 RV SA A *. C O/C 302-610 C-1 RT Y10 Valve Name SPENT FUEL POOL COOLER SF-C-lB TUBE SIDE RELIEF 1-NS-V-4 2 A 8 GA MO A 0 C 302-610. A-8 LTJ AJ.

PI Y2 SP M3 CSJ-08 STC CS CSJ-08.

Valve Name RCP MOTOR COOLER RETURN VALVE 1-NS-V40A 3 C 0.5. RV SA A C O/C" 302-610 B-8 RT Y10 Valve Name RC WASTE EVAP DISTILLATE COOLER WDL-C-3A RELIEF 1-NS-V-40B 3 C 0.5 RV SA A .. C O/C 3027610 B-7 RT Y10 Valve Name RC WASTE EVAP. DISTILLATE COOLER WDL-C-3B RELIEF 1-NS-V-41A 3 C 0.5 RV SA A

• C

• O/C 302-610 *C-8 RT Y10 Valve Name RC WASTE EVAPORATOR CONDENSERWDL-C-2A RELIEF 1-NSV-41B 3 C 0.5 RV SA

• A. C O/C 302-610 B-8 RT Y10 Valve Name RC WASTE EVAPORATOR CONDENSER WDL-C-2B RELIEF 1-NS-V-42A 3 C .5.0.5 RV. .SA . A C O/C 302-610 B-6 RT Y10 Valve Name WASTE GAS COMPRESSOR COOLER WDG-P-1A RELIEF Revision 0 October 25, 2013 A5 -.41

Attachment 15 Inservice Testing Valve Table NUCLEAR SERVICES CLOSED COOLING WATER (Page 3)

Valve EPN Safety Category Size Valve Act. Active Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq.. Request Just. Pos.

1-NS-V-42B 3 C 0.5 RV SA A C O/C 302-610 B-6 RT Y10.

Valve Name WASTE GAS COMPRESSOR COOLER WDG-P-1B RELIEF 1-NS-V-43A. 3 C 0.5. RV SA A C 0/C 302-610 B-6 RT Y10 Valve Name RC PUMP SEAL RETURN COOLER MU-C-2A RELIEF 1-NS-V-43B 3 C 0.5 RV SA A C O/C 302-610 B-5 RT Y10 Valve Name RC PUMP SEAL RETURN COOLER MU-C-2B RELIEF 1-NS-V-44A 3 C 0.5 RV SA .A C. O/C 302-645 B-6 RT Y10 Valve Name MU-P-1A MOTOR AND BEARING COOLING WATER RELIEF 1-NS-V-44B 3 C 0.5 RV SA A C. 0/C 302-610 D-4 RT Y10 Valve Name MU-P-1B MOTOR AND BEARING COOLING WATER RELIEF 1-NS-V-44C "3C 0.5 RV SA A C O/C 302-645 B-7 RT Y10 Valve Name MU-P-iC MOTOR AND BEARING COOLING WATER RELIEF 1-NS-V-45A 3 C 0.5 RV SA A C O/C 302-610 D-10 RT Y10 Valve Name RC-P-1A MOTOR COOLER RELIEF 1-NS-V-45B 3 C 0.5 RV SA A C O/C 302-610 D-10 RT Y10 Valve Name RC-P-1B MOTOR COOLER RELIEF 1-NS-V-45C 3 C 0.5 RV SA A C o/C 302-610 C-10 RT Y10 Valve Name RC-P-lC MOTOR COOLER RELIEF 1-NS-V-45D 3 C 0.5 RV SA A C O/C 302-610 B-10 RT Y10 Valve Name RC-P-1D MOTOR COOLER RELIEF 1-NS-V-46A 3 C 0.5 RV SA " . A C. O/C 302-610 F-6 RT Y10 Valve Name CONTROL BLDG AC HEAT EXCHANGER AH-C-4A RELIEF 1-NS-V-46B 3 C 0.5 RV SA *A C O/C 302-610 F-6 RT Y10 Valve Name CONTROL BLDG AC HEAT EXCHANGER AH-C-4B RELIEF.

1-NS-V-47 3 C 2 . RV SA A C o/C 302-610 H-7 RT Y4 Valve Name NS SURGE TANK NS-T-1 RELIEF 1-NS-V-48A 3 C 0.5 RV SA *A C 0/C 302-610 G-2 RT Y10 Valve Name NS & DC PUMP AREA COOLER AH-E-15A RELIEF 1-NS-V-48B 3 C 0.5 RV SA .. A. C O/C 302-610 G-1 RT Y10 Valve Name NS & DC PUMP AREA COOLER AH-E-15B RELIEF 1-NS-V-49A 3 C 1 .RV SA ' A . C O/C. 302-610 F-9 RT Y10 Valve Name EFW PUMP AND IACOMPR ROOM COOLER AH-E-24A RELIEF 1-NS-V-49B 3 C 0.5 RV SA A C O/C 302-610 F-8 RT Y10 Valve Name EFW PUMP AND IACOMPR ROOM COOLER AH-E-24B RELIEF Revision 0 October 25, 2013 A15 -42

Attachment 15 Inservice Testing Valve Table NUCLEAR SERVICES CLOSED COOLING WATER (Page 4)

Valve EPN. Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-NS-V-51A 3 C 0.5 RV SA A C O/C 302-610 C-3 RT Y10 Valve Name SPENT FUEL PUMP ROOM COOLER AH-E-8A RELIEF 1-NS-V-51B 3 C 1 RV SA A C O/C 302-610 C-3 RT Y10 Valve Name SPENT FUEL PUMP ROOM COOLER AH-E-8B RELIEF 1-NS-V-52A 2 B 1 GA AO P 0 0 302-610 E-8 PI Y2 Valve Name CONTAINMENT ISOLATION - AH-E1A MOTOR COOLER SUPPLY 1-NS-V-52B 2 B 1 GA AO P 0 0 302-610 E-7 PI Y2 Valve Name CONTAINMENT ISOLATION - AH-E1B MTR COOLER SUPPLY 1-NS-V-52C 2 B 1 GA AO P 0 0 302-610 E-7 PI Y2 Valve Name CONTAINMENT ISOLATION -AH-E1C MTR COOLER SUPPLY 1-NS-V-53A 2 B 1 GA AO P 0 0 302-610 F-8 PI Y2 Valve Name CONTAINMENT ISOLATION - AH-E1A MTR COOLER RETURN 1-NS-V-53B 2 B 1 GA AO P 0 0 302-610 F-7 PI Y2 Valve Name CONTAINMENT ISOLATION - AH-E1B MTR COOLER RETURN 1-NS-V-53C 2 B 1 GA AO P 0 0 302-610 F-7 PI Y2 Valve Name CONTAINMENT ISOLATION - AH-EIC MTR COOLER RETURN 1-NS-V-54A 3 B 1.5 GL AO A T 0 302-610 C-4 FO M3 IST-013 STO M3 Valve Name SPENT FUEL PUMP ROOM COOLING COIL FLOW CONTROL 1-NS-V-54B 3 B 1.5 GL AO A T 0 302-610 C-3 FO M3 IST-013 STO M3 Valve Name SPENT FUEL PUMP ROOM COOLING COIL FLOW CONTROL VLV 1-NS-V-85 3 A 0.5 GL M A 0 C 302-610 E-9 EC Y2 Valve Name NS-FI-76 ROTOMETER OUTLET ISOL VALVE 1-NS-V-96A 3 C 0.75 RV SA A C O/C 302-610 G-10 RT Y10 Valve Name NUCLEAR SERVICES HX NS-C-1A SHELL SIDE RELIEF 1-NS-V-96B 3. C 0.75 RV . SA A C O/C 302-610 G-9 RT Y10 Valve Name NUCLEAR SERVICES HX NS-C-1B SHELL SIDE RELIEF 1-NS-V-96C 3 C 0.75 RV SA A C O/C 302-610 G-9 RT Y10 Valve Name NUCLEAR SERVICES HX NS-C-1C SHELL SIDE RELIEF 1-NS-V-96D 3 C 0.75. RV SA A C O/C 302-610 G-8 RT Y10 Valve Name NUCLEAR SERVICES HX NS-C-1 DSHELL SIDE RELIEF Revision 0 October 25, 2013 A15 -43

Attachment 15 Inservice Testing Valve Table PENETRATION PRESSURIZATION SYSTEM (Page 1)

Valve EPN Safety. Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-PP-V-210 2 A 1 GL M P C *C 302-706 E-8 LTJ AJ Valve Name CONTAINMENT INTEGRITY BACKUP SUPPLY AHV1NB "

1-PP-V-211 2 A 1 GL M P C C 302-706 G-10 LTJ AJ Valve Name CONTAINMENT INTEGRITY - BACKUP SUPPLY TO AH-V1C/D 1-PP-V-212 2 A 2 GL M P C C 302-706 F-7 LTJ AJ Valve Name CONTAINMENT INTEGRITY - PP-T1B SUPPLY TO.AH-V1ANB 1-PP-V-213 .2 A 2 GL M P C C 302-706 G-9 LTJ AJ Valve Name CONTAINMENT INTEGRITY - PP-T1A SUPPLY TO AH-V1C/D Revision 0 October 25, 2013 A15 - 44

Attachment 15 Inservice Testing Valve Table REACTOR BUILDING EMERGENCY COOLING WATER (Page.1) category Size Valve Act. Active / Normal Safety* P&&ID P&&ID Valve EPN Safety Test Test Relief Deferred Tech.

Class Type Type Passive* Position Position Coor. Type Freq. Request Just. Pos.

1-RB-V-1A NA C 0.5 RV SA A C O/C 302-611 G-6 RT Y10 Valve Name RB NORMAL COOLING COIL 1A RELIEF VLV 1-RB-V-lB NA C 0.5 RV SA A C O/C 302-611 G-6 RT Y10 Valve Name RB NORMAL COOLING COIL 1B RELIEF VLV 1-RB-V-lC NA C 0.5 RV SA A C O/C 302-611 G-6 RT Y10 Valve Name RB NORMAL COOLING COIL 1C.RELIEF VLV 1-RB-V-iD NA C 0.5 RV SA

• A C O/C 302-611 G-6 RT Y10 Valve Name RX COMPARTMENT COOLING COIL AH-C-3A RELIEF VLV 1-RB-V-iE NA C 0.5

• RV SA A C O/C 302-611 G-6 RT Y10 Valve Name RX COMPARTMENT COOLING COIL AH-C-3B RELIEF VLV 1-RB-V-2A 2 A 8 GA MO A 0 C 302-611 F-8 LTJ AJ PI Y2 SP M3 CSJ-09 STC CS CSJ-09 Valve Name CONTAINMENT ISOLATION - RB NORMAL AIR COOL SUP VLV 1-RB-V-7 2 A 8 GA MO A 0 C 302-611 F-8 LTJ AJ PI Y2 SP M3 CSJ-09 STC CS CSJ-09 Valve Name CONTAINMENT ISOLATION - RB NORMAL COOL RETURN ISOL Revision 0 October 25, 2013 A15-45

Attachment 15 Inservice Testing Valve Table REACTOR COOLANT (Page 1)

Category Size Valve Act. Active I Normal Safety

.Valve EPN Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-RC-RD-1 2 D 2 RPD SA A C 0. 302-650 D-8 DT Y5 Valve Name RCS HOT LEG VENT LINE RUPTURE DISK RD-1 1-RC-RD-2 2 D 2 RPD SA A C 0 302-650 D-2 DT Y5 Valve Name RCS HOT LEG VENT LINE RUPTURE DISK RD-2 1"

1-RC-RV-1A C'" 2.5 RV SA A

• C O/C 302-650 G-6 RT Y5 Valve Name PRESSURIZER.CODE SAFETY VALVE 1-RC-RV-1B 1* C 2.5 RV SA A C O/C 302-650. G-6 RT Y5 Valve Name PRESSURIZER CODE SAFETY VALVE 1-RC-RV-2 B/C 4 RV SO A C O/C 302-650 G-6 FC RR VR-01 IST-013 RT Y5 VR-01 STC RR VR-01 STO RR VR-01 Valve Name PZR PILOT OPERATED RELIEF VALVE (PORV) 1-RC-V-144A 1 C 14 CK SA A ' C 0 CC RR RJ-28 CO  !RR RJ-28 Valve Name REACTOR VESSEL INTERNAL VENT VALVE 1-RC-V-144B -1 C 14 C.K SA A C 0 CC RR RJ-28 CO RR RJ-28 Valve Name REACTOR VESSEL INTERNAL VENT VALVE 1 C CK

.14*1-RC-V-144C SA A C -0 CC RR RJ-28 CO RR RJ-28 Valve Name REACTOR VESSEL INTERNAL VENT VALVE CC RR . RJ-28 1-RC-V-1440 1 C 14 CK SA A C 0 CO RR RJ-28 Valve Name REACTOR VESSEL INTERNAL VENT VALVE 1-RC-V-144E 1 C 14 CK SA A C 0 CC RR RJ-28 CO RR . RJ-28 Valve Name REACTOR VESSEL INTERNAL VENT VALVE 1-RC-V-144F 1 C 14 CK SA A C 0 CC RR RJ-28 CO RR RJ-28 Valve Name REACTOR VESSEL INTERNAL VENT VALVE 1-RC-V-144G 1 C 14 CK SA A C 0 CC RR RJ-28 CO RR RJ-28 Valve Name REACTOR VESSEL INTERNAL VENT VALVE 1-RC-V-144H 1 C 14 CK SA A .C 0 CC RR - RJ-28 CO RR RJ-28 R

Valve Name REACTOR VESSEL INTERNAL VENT VALVE 302-650 G-6 PI Y2 1-RC-V-2 1 CB 34 *GA MO A 0 O/C STC M3 STO M3 Valve Name PORV (RC-RV2) ISOLATION VALVE Revision 0 October 25, 2013 A15-46

Attachment 15 Inservice Testing Valve Table REACTOR COOLANT (Page 2)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-RC-V-23 .1 A/C 1.5 CK SA A. C. O/C 302-650 G-6 CC RR. RJ-23 COF RR RJ-23 LTH Y2

• Valve Name DH PRESSURIZER SPRAY LINE CHECK VALVE.

1-RC-V-28 1 8 1 GL MO A C O/C 302-650 G-8 P1 Y2 STC CS CSJ-11 STO CS .CSJm 1 Valve Name PZR VENT TO RCDT ISOLATION VALVE 1-RC-V-3 1 B 2.5 GL MO A 0 C 302-650 E-6 PI *Y2 STC CS CSJ-17 Valve Name PRESSURIZER SPRAY LINE ISOLATION VALVE 1-RC-V-4 1 A 1.5 GL MO'. A C 0(/C 302-650 G-6 LTH Y2 PI Y2 STc CS CSJ-18 STO CS CSJ-18 Valve Name DECAY HEAT PRESSURIZER SPRAY LINE ISOLATION VALVE 1-RC-V-40A 1 B 0.5 GL SO A C O/C 302-650 E-8 FC CS CSJ-12 IST-013 PI Y2 STC CS CSJ-12

• STO CS CSJ-12 Valve Name "A"HOT LEG HIGH POINT VENT TO RCDT AND ATMOSPHERE 1-RC-V-40B. 1 B 0.5 GL SO A C O/C 302-650 E-2 FC CS CSJ-12 IST-013.

PI Y2 STC CS CSJ-12 STO CS CSJ-12 Valve Name "B"HOT LEG VENT TO RCDT AND ATMOSPHERE 1-RC-V-41A 1 B 0.5 GL SO A C C 302-650 E-8 FC CS CSJ-12 IST-013 PI .Y2 STC CS CSJ-12 STO CS CSJ-12 Valve Name "A"HOT LEG VENT TO RCDT AND ATMOSPHERE 1-RC-V-41B 1 B 0.5 GL SO A C C 302-650 E-2 FC CS CSJ-12 IST-013 PI Y2 STC CS CSJ-12 STO CS CSJ-12 Valve Name "B"HOT LEG VENT TO RCDT AND ATMOSPHERE 1-RC-V-42 1 B 0.5 GL .SO. A. C O/C 302-650 G-4 FC CS CSJ-13 IST-013 P1 Y2 STC CS CSJ-13 STO CS CSJ-13 Valve Name REACTOR VESSEL VENT TO REACTOR BLDG ATMOSPHERE Revision 0 October 25, 2013 A15-47

Attachment 15 Inservice Testing Valve Table REACTOR COOLANT (Page 3)

Valve EPN Safety Category Size Valve Act. Active I Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-RC-V-43 1 B 0.5 GL SO A C C 302-650 G-4 FC CS CSJ-13 IST-013 PI Y2 STC CS CSJ-13 STO CS CSJ-13 Valve Name REACTOR VESSEL VENT TO REACTOR BLDG ATMOSPHERE 1-RC-V-44 1 B 1 GL SO A C O/C 302-650 G-7 FC CS CSJ-11 IST-013 PI Y2 STC CS CSJ-11 STO CS CSJ-11 Valve Name PRESSURIZER HIGH POINT VENT ISOLATION VALVE Revision 0 October 25, 2013 A15-48

Attachment 15 Inservice Testing Valve Table REACTOR BUILDING EMERGENCY COOLING WATER (Page 1)

Valve EPN Safety Category Size Valve Act. Active i Normal. Safety p&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

I-RR-V-1oA 3 B 2 GL AO A C o/c 302-611 E-3 FC M3 IST-013 P1 Y2 STC M3 STO M3 Valve Name RR-P1A RECIRCULATION MINIMUM FLOW BYPASS VALVE 1-RR-V-10B .3 .B 2 GL AO A C 0/C 302-611 E-2 FC M3 IST-013 PI Y2 STC M3 STO M3 Valve Name RR-P1B RECIRCULATION MINIMUM FLOW BYPASS VALVE 1-RR-V-11A 2 C 0.5

• RV SA A C O/C 302-611 F-4 RT Y10 Valve Name RBEC COIL "A"RELIEF 1-RR-V-11B 2 C 0.5 RV SA A C O/C 302-611 E-4 RT Y10 Valve Name RBEC COIL "B"RELIEF 1-RR-V-11C 2 C 0.5 RV SA A C O/C 302-611 D-4 RT Y10 Valve Name RBEC COIL "C"RELIEF 1-RR-V-12A 3 C 2 CK SA A C O/C 302-611 D-3 CC *M3 CO M3 Valve Name RIVER WATER PUMP "A"VACUUM BREAKER (Check Valve) 1-RR-V-12B 3 C 2 CK SA A C. O/C 302-611 D-2 CC M3 CO M3 Valve Name RIVER WATER PUMP "B"VACUUM BREAKER (Check Valve) 1-RR-V-1A 3 B 16 BTF MO A C 0 302-611 E-2 PI Y2 STO M3 Valve Name RR-P1A DISCHARGE VALVE 1-RR-V-1B 3 B 16 BTF MO A C 0 302-611 E-2 P1 Y2 STO M3.

Valve Name RR-P1B DISCHARGE VALVE 1-RR-V-33A 3 B 2 BALL MO P 0 0 302-203 A-6 P1 Y2 Valve Name RRwS-1A BLOWDOWN VALVE 1-RR-V-33B 3 B 2 BALL MO P 0 0 302-203 A-8 PI Y2 Valve Name RR-S-1B BLOWDOWN VALVE 1-RR-V-3A 2 B 12 GA MO A 0 O/C 302-611 H-4 PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION - RBEC COIL "A"INLET VALVE 1-RR4V-3B 2 B 12 GA MO A 0 o/C 302-611 G-4 P1 Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION - RBEC COIL "B"INLET VALVE 1-RR-V-3C 2 B 12 GA MO A 0 O/C 302-611 F-4 PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION - RBEC COIL "C"INLET VALVE Revision 0 October 25, 2013 A15 -49

Attachment 15 Inservice Testing Valve Table REACTOR BUILDING EMERGENCY COOLING WATER (Page 2)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive position Position Coor. Type Freq. Request Just. Pos.

1-RR-V-4A 2 B .12 GA MO A C O/C 302-611 E-3 PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION - RBEC COIL "IA"OUTLET VLV 1-RR-V-4B 2 B 12 GA MO *A C O/C 302-611 E-3 PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION - RBEC COIL "IB" OUTLET VLV 1-RR-V.4C 2 B 12 GA MO A C O/C 302-611 D-3 PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION - RBEC COIL "IC" OUTLET VLV 1-RR-V-4D 2 B 12 GA MO A C O/C 302-611 D-3 PI Y2 STC M3 STO M3 Valve Name CONTAINMENT ISOLATION - RBEC COIL "1C" OUTLET VLV 1-RR-V-5 3 B 10 BTF MO A C 0 302-611 B-3 PI Y2 STO M3 Valve Name RR-V6 RB COOLING COIL DISCHARGE BYPASS VALVE 1-RR-V-6 3 B 10 *BTF DIA A T 0 302-611 B-3 FO M3 IST-013 STO M3 Valve Name RB EMERG COOLING COIL BACK PRESSURE REGULATOR 1-RR-V-7A 3 C 16 CK. SA A C 0. 302-611 D-2 *BDC RR RJ-18, IST-014 COF RR RJ-18 Valve Name RB EMERG COOL PUMP "A"DISCHARGE CHECK VALVE 1-RR-V-7B 3 C .16 CK SA A C 0 .302-611 D-2 BDC RR RJ-18 IST-014 COF RR RJ-18 Valve Name RB EMERG COOL PUMP "B"DISCHARGE CHECK VALVE 1-RR-V-8A 3 C 20 CK SA A C O/C 302-611 F-2 CC M3 COF RR RJ-14 Valve Name RIVER WATER TO RB COOL UNITS CHECK VALVE 1-RR-V-8B 3 C 20 CK SA A C O/C 302-611 F-2 CC M3 COF RR RJ-14 Valve Name RIVER WATER TO RB COOL UNITS CHECK VALVE 1-RR-V-9A 3 C 12 CK SA A C 0 302-611 E-4 BDC RR RJ-11 IST-014 COF RR RJ-11 Valve Name CONTAINMENT ISOLATION - AH-E1A EMERG COOL OUTLET 1-RR-V-9B 3 C 12 CK SA *A C 0 302-611 E-4 BDC RR RJ-11 IST-014 COF RR RJ-11 Valve Name CONTAINMENT ISOLATION - AH-E1B EMERG COOL OUTLET 1-RR-V-9C 3 C 12 CK SA

• A C 0 302-611 D-4 BDC RR RJ-11 IST-014 COF RR RJ-1 1 Valve Name CONTAINMENT ISOLATION - AH-ElC EMERG COOL OUTLET Revision 0 October 25, 2013 A15 - 50

Attachment 15 Inservice Testing Valve Table STATION SERVICE AIR (Page 1)

Valve EPN Safety Category Size Valve Act. Active I Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-SA-V-2 2 A 2 GL M P LC C 302-271 E-1 LTJ AJ Valve Name CONTAINMENT ISOLATION - SERVICE AIR TURB BLDG 1-SA-V-3 2 A 2 GL M P LC C 302-271 E-1 LTJ AJ Valve Name CONTAINMENT ISOLATION - RB 1ST FL SVC AIR ISOL Revision 0 October 25, 2013 Al5 -.51

Three Mile Island Unit #1 Inservice Testing Program Plan Attachment 15 Inservice Testing Valve Table SPENT FUEL COOLING (Page 1)

Valve EPN Safety Category Size Valve Act. Active/ Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1I-SF-V-1 B 8 DIA AO A O/C 0 302-630 F-6 PI :Y2 FO M3 IST-01 3 STO M3 Valve Name SF-P-1A SUCTION FROM SF POOL "B" 1-SF-V-11 3 B 8 DIA AO A O/C 0. 302-630 E-4 PI Y2 FO M3 IST-013 STO M3 Valve Name COOLER sF-ClA RETURN TO SF POOL "B" 1-SF-V-12 3 B 8 DIA AO A O/C O 302-630 E-4 PI Y2 FO M3 IST-013 STO M3 Valve Name COOLER SF-CiA RETURN TO SF POOL "A" 1-SF-V-14 3 B 8 DIA AO A O/C O 302-630 D-4. PI Y2 FO M3 IST-013 STO M3 Valve Name COOLER SF-C1B RETURN TO SF POOL "B" 1-SF-V-15 3 B 8 DIA AO A O/C O 302-630 D-4 PI Y2 FO M3 IST-013 STO M3 Valve Name COOLER SF-C1B RETURN TO SF POOL "A" 1-SF-V-2 3 B 8 DIA A0 A O/C 0 302-630 F-6 PE Y2 FO M3 IST-013 STO M3.

Valve Name SF-P-1A SUCTION FROM SF POOL "A" 1-SF-V-22 2 A . 8 GA M .P LC C 302-630 E-7 LTJ AJ Valve Name FUEL TRANSFER CANALFILL & DRAIN LINE VALVE 1-SF-V-23 2 A 8 GA M P LC C 302-630 E-7 LTJ AJ Valve Name FUEL TRANSFER CANAL FILL &DRAIN LINE VALVE 1-SF-V-4 3 B 8 DIA AO. A O/C 0 302-630 D-6 PI Y2 FO .M3 IST-013.

STO M3 Valve Name SF-P-1B SUCTION FROM SF POOL "B" 1-SF-V-5 3' 8 DIA AO *A 0O/C 0 302-630 D-6 PI Y2 FO M3 IST-013 STO M3 Valve Name SF-P-1B SUCTION FROM SF POOL "A" 1-SF-V-50 3 C 8 CK SA A 0 0/C 302-630 H-4 CCR CM COF CM Valve Name SF POOL "A"COOLING RETURN CHECK VALVE 1-SF-V-51 .3 C 8 CK SA A 0 O/C 302-630 H-1 CCR CM COF CM Valve Name SF POOL "B"COOLING RETURN CHECK VALVE Revision 0 October 25, 2013 A15 - 52

Attachment 15 Inservice Testing Valve Table SPENT FUEL COOLING (Page 2)

Valve EPN Safety Category Size Valve Act. Active / Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-SF-V-7 3. C 8 CK SA A

• SYS O/C 302-630 F-5 CC M3 CO M3 Valve Name SF-P1A DISCHARGE CHECK VALVE 1-SF-V-8 3 C 8 CK SA A SYS O/C 302-630 D-5 CC M3 CO M3 Valve Name SF-P1B DISCHARGE CHECK VALVE Revision 0 October 25, 2013 A15 -53

Attachment 15 Inservice Testing Valve Table GASEOUS WASTE DISPOSAL (Page 1)

Valve EPN Safety. Category Size Valve Act.. Active Normal Safety P&&ID p&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-WDG-V-3 2 A 2 GL MO A 0 C 302-694 -G-6 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - RB VENT HEADER VALVE 1-WDG-V-4 2 A 2 GA SO A 0 C 302-694 G-5 FCý M3 IST-013 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - RB VENT HEADER VALVE Revision 0 October 25, 2013 A15 - 54

Attachment 15 Inservice Testing Valve Table LIQUID RADWASTE DISPOSAL (Page 1)

Valve EPN Safety Category Size Valve Act. Active I Normal Safety P&&ID P&&ID Test Test Relief Deferred Tech.

Class Type Type Passive Position Position Coor. Type Freq. Request Just. Pos.

1-WDL-V-303 2 A 3 GA MO A C C 302-690 F-8 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - RC DRAIN PUMP DISCH VALVE 1-WDL-V-304 2 A 3 GL AO A C C 302-690 E-8 FC M3 IST-013 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - RC DRAIN PUMP DISCH ISOL 1-WDL-V-534 2 A 6 GA AO A C C 302-719 B-6 FC M3 IST-013 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - RB SUMP DRAIN TO AUX BLDG 1-WDL-V-535 2 A 6 GA AO A C C 302-719 B-6 FC M3 IST-013 LTJ AJ PI Y2 STC M3 Valve Name CONTAINMENT ISOLATION - RB SUMP DRAIN TO AUX BLDG 1-WDL-V-727 2 A/C 0.25 RV SA A C O/C 302-690 F-8 LTJ AJ RT Y10 Valve Name CONTAINMENT PENETRATION NO. 331 RELIEF Revision 0 October 25, 2013 A15 - 55

Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 16 CHECK VALVE CONDITION MONITORING PLAN INDEX Revision 0 October 25, 2013 A16- 1

.Three Mile Island Unit #1 Inservice Testing Program Plan ATTACHMENT 16 Valves in Check Valve Condition.Monitoring Program Valves shown in this table are monitored in the Check Valve Condition Monitoring (CVCM)

Program. The details for each valve, including individual. valve testing information, can be found in Attachment 15 of this Program Plan.. The Test Plan and Test Analysis review for the valves are maintained in the individual Check Valve Condition Monitoring Plans.

Item CVCM Plan Valve EPNs Description Containment Isolation - RB Spray Header Check 1 CMP-BS-01 1-BS-V-30A/B Valves Containment Isolation - Reclaimed Feed To RB 2 CMP-CA-01 1-CA-V-192 Check 3 CMP-CF-01 1-CF-V-12A/B Containment Isolation - CFT Makeup Check Valves 4 CMP-CF-02 1-CF-V-4A/B Core Flood Tanks Outlet Check Valves 5 CMP-CO-01 1-CO-V-16A/B Condensate Check Valve - Supply To EFW Pumps 6 CMP-CO-02 1-CO-V-175A/B EFW Pump Bearing Cooling Return Check Valves 7 CMP-DH-01 1-DH-V-69 Containment Isolation - DH To PZR Aux Spray Line 8 CMP-FW-01 1-FW-V-122AJB Containment Isolation - OTSG Inlet Check Valves 9 CMP-IC-01 1-IC-V-16 Containment Isol - CRD Cooling Inlet Check Valve 10 CMP-IC-02 1-iC-V-18. Containment Isol - Inter CIg Inlet to RB Check Valve 11 CMP-MS-01 1-MS-V-9A/B Main Steam Supply Check Valve To EF-Ul..

1-MU-V-12 CMP-MU-01 107A/B/C/D Containment Isolation - HPI To RC Supply Checks 13 'CMP-MU-02 1-MU-V-54A/B Lithium Hydroxide Pump To MU System Checks 14 CMP-MU-03 1-MU-V-116 Containment Isolation - Seal Ini Spray/RC Check 15 CMP-MU-04 1-MU-V-219 Normal Makeup Line To HPI Line B Check HPI Loop "B" Back Flow From Makeup - Check 16 CMP-MU-05 1-MU-V-220 Valve 1-MU-V-86A/B, 17 CMP-MU-06 MU-V-95 RCS HPI Inlet Check Valves 18 CMP-MU-07 1-MU-V-94 Loop RCS HPI Inlet Check Valve 19 CMP-NS-01 1-NS-V-11 RCP Motor Cooler Supply Valve 1-NS-V-20 CMP-NS-02 1OA/B/C Nuc Svcs Pump Discharge Check Valves 21 CMP-SF-01 1-SF-V-50/51. SF Pool Cooling Return Check Valve Revision 0 October 25, 2013 A16 -2