Text

Sixth 10-Year Interval Inservice Testing (1ST) Program Plan
	ML22277A366
Person / Time
Site:	Point Beach
Issue date:	09/30/2022
From:	Strand D; Point Beach
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NRC 2022-0032
	Download: ML22277A366 (177)
	v • d • e

NEXTera ENERGY ~

POINT BEACH September 30, 2022 NRC 2022-0032 10 CFR 50.4 10 CFR 50.55a U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington , D. C. 20555-0001 RE : Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 Renewed Facility Operating Licenses DPR-24 and DPR-27 Sixth 10-Year Interval lnservice Testing (1ST) Program Plan Enclosed please find the NextEra Energy Point Beach , LLC Sixth 10-Year Interval lnservice Testing (1ST) Program Plan for Point Beach Nuclear Plant Units 1 and 2.

The Sixth 10-Year 1ST Interval for Unit 1 began on September 1, 2022. The Sixth 10-Year 1ST Interval for Unit 2 begins on October 1, 2022.

The Sixth 10-Year Interval 1ST Program Plan for Point Beach Nuclear Plant Units 1 and 2 is based on the requirements of the ASME OM Code, 2017 Edition.

No 1ST Relief Request for the Sixth 10-Year 1ST Interval was required.

If you have any questions, please contact Mr. Kenneth Mack, Fleet Licensing Manager, at (561) 904-3635.

Sincerely,

't:i 1

Dianne Strand

--s F~~

General Manager Regulatory Affairs Enclosure cc: USNRC Regional Administrator, Region Ill Project Manager, USNRC, Point Beach Nuclear Plant Resident Inspector, USNRC, Point Beach Nuclear Plant Public Service Commission of Wisconsin NextEra Energy Point Beach, LLC 6610 Nuclear Road, Two Rivers, WI 54241

ENCLOSURE POINT BEACH NUCLEAR PLANT, UNIT NO. 1 AND 2 INSERVICE TESTING PROGRAM DOCUMENT POINT BEACH INSERVICE TESTING PROGRAM SIXTH 10-YEAR INTERNAL (175 pages follow)

INSERVICE TESTING PROGRAM DOCUMENT PBNP Inservice Testing Program SIXTH 10-YEAR INTERVAL Point Beach Nuclear Plant, Unit No. 1 and 2 Two Rivers, WI Owned and Operated by: NEER DOCUMENT TYPE: Controlled Reference REVISION: 11 EFFECTIVE DATE: Unit 1 (09/01/2022) Commercial Date 12/21/70 Unit 2 (10/01/2022) Commercial Date 10/01/72 APPROVAL AUTHORITY: Department Manager PROCEDURE OWNER (title): Group Head OWNER GROUP: Engineering Programs

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND 2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT TABLE OF CONTENTS SECTION TITLE PAGE

1.0 INTRODUCTION

............................................................................................................... 5 2.0 CODE AND REGULATORY REQUIREMENTS ............................................................. 5 2.1 Code of Federal Regulations ................................................................................................ 5 2.2 Applicable Code ................................................................................................................... 6 3.0 GENERALIST PROGRAM GUIDELINES AND POSITIONS ....................................... 6 3.1 NUREG-0800 I NRC Technical Position BTP 5-4 .............................................................. 6 3.2 Components Required for Hot Shutdown vs. Cold Shutdown ............................................ 7 3.3 Components Required for Safe Operation vs. Safe Shutdown ............................................ 7 3.4 System Classification Criteria .............................................................................................. 7 3.5 Component Selection for Accident Mitigation .................................................................... 8 3.6 Beyond-Design-Basis Events ............................................................................................... 9 3.7 Inservice Test Frequencies ................................................................................................... 9 3.8 Inservice Examination and Test Frequency Grace ............................................................ 10 3.9 Testing Non-Code Class Components (10CFR50.55a Augmented IST Program) ............ 11 3.10 Risk-Informed Testing for ASME OM Code Components ............................................... 11 3.11 Pre-Maintenance and Post Maintenance Testing ............................................................... 11 4.0 DEFINITIONS ................................................................................................................... 12

5.0 REFERENCES

.................................................................................................................. 18 6.0 PUMP IST PROGRAM ..................................................................................................... 21 6.1 Pump Selection Criteria and Exemptions .......................................................................... 21 6.2 Group A Pumps .................................................................................................................. 21 6.2.1 Group A Pump Testing Frequency ..................................................................... 22 6.2.2 Group A Pump Test Parameters ......................................................................... 22 6.2.3 Required Instrument Accuracy ........................................................................... 23 6.2.4 Allowable Ranges for Group A Pump Hydraulic Test Parameters .................... 24 6.2.5 Allowable Ranges for Group A Pump Test Vibration Data ............................... 24 6.2.6 Group A Test Procedure ..................................................................................... 25 6.3 Group B Pumps .................................................................................................................. 25 6.3.1 Group B Pump Testing Frequency ..................................................................... 26 6.3.2 Group B Pump Test Parameters .......................................................................... 26 6.3.3 Required Instrument Accuracy ........................................................................... 27 6.3.4 Allowable Ranges for Group B Pump Hydraulic Test Parameters ..................... 27 6.3.5 Allowable Ranges for Comprehensive Pump Test Vibration Data .................... 27 6.3.6 Group B Test Procedure ...................................................................................... 28 Page 2 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT TABLE OF CONTENTS SECTION TITLE PAGE 6.4 Comprehensive Test Procedure ......................................................................................... 29 6.4.1 Allowable Ranges for Comprehensive Hydraulic Test Parameters .................... 29 6.4.2 Mandatory Appendix V Periodic Verification Test ........................................... 30 6.5 Preservice Test ................................................................................................................... 31 6.6 Tests Substitution ............................................................................................................... 30 6.7 Allowable Variance from Fixed Reference Points ............................................................ 31 6.8 Analysis and Evaluation of Test Results ........................................................................... 31 6.9 Analysis (ISTB-6200(c)) ................................................................................................... 32 6.10 Replacement, Repair, and Maintenance ............................................................................. 32 6.11 Relief Request .................................................................................................................... 33 6.12 Pump Test Table ................................................................................................................ 33 7.0 VALVE IST PROGRAM .................................................................................................. 35 7.1 Valve Selection Criteria and Exemptions .......................................................................... 35 7.2 Active/Passive Designation ............................................................................................... 36 7.3 Preservice Testing .............................................................................................................. 36 7.4 Pressure Isolation Valves (PIVs) ....................................................................................... 37 7.5 Containment Isolation Valves (CIVs) ................................................................................ 38 7.6 Valve Categories ................................................................................................................ 39 7.7 Valve Inservice Testing Frequency ................................................................................... 39 7.8 Valve Test Requirements ................................................................................................... 42 7.8.1 Valve Position Verification ................................................................................. 42 7.8.2 Active Power Operated Valve Stroke Time Testing .......................................... .44 7.8.3 Active MOVs (10CFR50.55a (b)(3)(ii) and Mandatory Appendix III) ............ .46 7.8.4 Active AOV Testing (Mandatory Appendix IV) ................................................ 47 7.8.5 Category A Valve Seat Leakage Testing ........................................................... 48 7.8.6 Fail-Safe Testing ................................................................................................. 49 7.8.7 Check Valve Testing ........................................................................................... 49 7.8.8 Check Valve Condition Monitoring (CVCM) Program Implementation ........... 50 7.8.9 Manual Valve Exercising .................................................................................... 51 7.8.10 Safety and Relief Valve Testing ......................................................................... 51 7.8.11 Vacuum Breaker Testing .................................................................................... 51 7.8.12 Rupture Disks Testing ......................................................................................... 51 7.8.13 Effects of Valve Repair, Replacement, or Maintenance ..................................... 52 7.9 Relief Requests .................................................................................................................. 53 7.10 Valve TestTable:, .............................................................................................................. 53 Page 3 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT TABLE OF CONTENTS SECTION TITLE PAGE APPENDIX A ................................................................................................. PUMP RELIEF REQUESTS APPENDIX B .............................................................................................. VALVE RELIEF REQUESTS APPENDIX C .................................................................. COLD SHUTDOWN TEST WSTIFICATIONS APPENDIX D ............................................................. REFUELING OUTAGE TEST WSTIFICATIONS APPENDIX E ......................................................................................... TECHNICAL WSTIFICATIONS APPENDIX F .................................................................................................... TECHNICAL POSITIONS APPENDIX G ............................................................................................................................ NOT USED APPENDIX H .............................................................................. SIGNIFICANT PROGRAM CHANGES Page 4 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT

1.0 INTRODUCTION

The Point Beach Nuclear Plant (PBNP) Unit 1 and 2 Inservice Testing (IST) Program Document details the technical basis and provides the overall description of activities planned to fulfill IST requirements for pumps and valves as specified in Title 10, Part 50.55a of the Code of Federal Regulations (10CFR50.55a) and the ASME Operation and Maintenance of Nuclear Power Plants (OM) Code.

In addition to the referenced ASME OM Code, this IST Program Document was prepared using the guidelines provided in Fleet Procedure ER-AA-113-1000 "Inservice Testing Procedure;"

Nonmandatory Appendix A of the OM Code, "Preparation of Test Plans;" and NUREG-1482, "Guidelines for Inservice Testing at Nuclear Power Plants." Since the initial issuance of Generic Letter (GL) 89-04, "Guidance on Developing Acceptable Inservice Testing Programs,"

NRC Staff improved guidance by revising 10CFR50.55a and NUREG-1482. These documents provide guidelines for the basis for component selection, test requirements, relief requests, and IST Program plan format; however, the owner has ultimate responsibility for format and content.

All Unit 1, Unit 2 and common unit components tested under the IST Program are identified along with relevant component information, drawings, tests, and applicable Surveillance Procedure test frequencies. Appendices to this document contain requests for relief from Code requirements ( as applicable), justifications of deferral of testing, utility technical positions, and a list of significant changes to the program (added or deleted components; added, deleted, or modified reliefrequests and deferred test justifications; changed tests or test frequencies).

Dynamic restraint (Snubber) IST Program requirements for PBNP Units 1 and 2 in accordance with Subsection ISTD of the OM Code are addressed in Fleet Procedure ER-AA-119-1000, "Snubber Program Procedure."

The Aging Management Programs and Time Limited Aging Analysis information are addressed in the appropriate ASME Section XI Programs.

2.0 CODE AND REGULATORY REQUIREMENTS 2.1 Code of Federal Regulations 10CFR50.55a Paragraph (f)(4)(ii) requires 10-year IST Programs to comply with the latest NRC approved Edition and Addenda of ASME OM Code incorporated by reference in Paragraph (a)(l)(iv) 18 months prior to the start of the 120-month inspection interval.

The sixth 10-year inspection interval for PBNP Unit 1 and Unit 2 commenced September 1, 2022 and October 1, 2022, respectively. The use of any later Edition and Addenda of the ASME OM Code is allowed if it has been incorporated in Paragraph (a)(l)(iv) of 10CFR50.55a, or if approved by the NRC as an acceptable alternative.

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POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT According to 10CFR50.55a Paragraphs (f)(l) and (f)(4), inservice testing shall be conducted in accordance with the appropriate Edition/Addenda of the Code to the extent practical within the limitations of design, geometry, and materials of construction.

Where Code requirements have been determined to be impractical, written relief has been requested pursuant to 10CFR50.55a Paragraphs (z) or (f)(5)(iii).

2.2 Applicable Code The sixth 10-year interval pump and valve IST Program was developed in accordance with the rules and requirements specified in the ASME OM Code, 2017 Edition as referenced by 10CFR50.55a Paragraph (a)(l)(iv)(C). All references to ASME OM Code, OM Code or the Code, in this document correspond to the aforementioned Edition.

The contents of this Code, as applicable to the pump and valve IST Program, are arranged as follows:

Subsection ISTA General Requirements

Subsection ISTB Inservice Testing of Pumps in Water-Cooled Reactor Nuclear Power Plants - Pre-2000 Plants

Subsection ISTC Inservice Testing of Valves in Water-Cooled Reactor Nuclear Power Plants

Mandatory Appendix I Inservice Testing of Pressure Relief Devices in Water-Cooled Reactor Nuclear Power Plants

Mandatory Appendix II Check Valve Condition Monitoring Program (Mandatory only if implemented, Owner's choice)

Mandatory Appendix III Preservice and Inservice Testing of Active Electric Motor-Operated Valve Assemblies in Water-Cooled Reactor Nuclear Power Plants

Mandatory Appendix IV Preservice and Inservice Testing of Active Pneumatically Operated Valve Assemblies in Nuclear Reactor Power Plants

Mandatory Appendix V Pump Periodic Verification Test Program 3.0 GENERALIST PROGRAM GUIDELINES AND POSITIONS The PBNP IST Program Document establishes consistent guidelines for determining IST Program scope and testing requirements. The IST Program Background Document contains evaluations of plant systems and related component functions and provides the detailed bases for including components in the IST Program or for excluding them. The following guidelines were used for evaluating pumps and valves with respect to IST Program scope and for implementation of ASME OM Code requirements.

3.1 NUREG-0800 / NRC Technical Position BTP 5-4 Page 6 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND 2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT NUREG-0800, Section 5.4.7, and NRC Technical Position BTP -4, define requirements to be capable of achieving hot shutdown, assuming the worst-case single failure. Plants with an Operating Licenses issued before April 1984 and Operating License applications docketed before April 1984 need not comply with the provisions of this item but may do so voluntarily.

3.2 Components Required for Hot Shutdown vs. Cold Shutdown The OM Code requires testing components required to bring the plant to safe shutdown.

Plants licensed for hot shutdown as the safe shutdown condition may not need to include components and systems necessary to achieve cold shutdown in the IST Program. The components necessary to achieve cold shutdown for these plants may not be safety related or subject to quality assurance requirements. PBNP is licensed for hot shutdown as the safe shutdown condition. Therefore, components required to achieve cold shutdown may not be included in the PBNP IST Program.

3.3 Components Required for Safe Operation vs. Safe Shutdown The PBNP FSAR, regulatory commitments, and related licensing basis or design basis documents (such as docketed design and testing commitments), are the primary references for determining which components perform functions within the scope of the ASME Code. Technical Specifications and several other plant source documents ( e.g.,

design guides, emergency and abnormal operating procedures, etc.) identify components that may be important to safe operation of the facility, an enhancement to system reliability, or are operated in conjunction with accident recovery. However, unless specific credit is taken for a component or system in design or licensing basis documents for achieving safe shutdown, maintaining safe shutdown, or mitigating the consequences of an accident, the component need not be included in the IST Program.

3.4 System Classification Criteria USAS B3 l .1 was the construction code for PBNP. Since PBNP was not constructed to Section III of the ASME Boiler and Pressure Vessel (BPV) Code, components were originally neither designed to ASME BPV Code Class 1, 2, and 3 requirements nor classified as such. The ASME Code classifications of systems and components at PBNP were established only to define components subject to inservice inspection and testing requirements. The NRC staff issued Regulatory Guide (RG) 1.26 to provide guidance on ASME Code classification of components for non-Section III plants. Per the NRC Standard Review Plan, NUREG-0800, Section 3.2.2, licensees may use either RG 1.26 or the ANS standards (ANSI/ANS-51.1 for PWRs) for establishing component classifications. Both documents classify components according to the safety functions that they perform. However, RG 1.26 does not cover many components and systems which may perform safety related functions such as emergency diesel support systems, HVAC systems, and instrument air/nitrogen systems. Also, RG 1.26 does not define classification requirements for primary containment penetration piping and containment isolation valves. Therefore, many components which perform safety functions may not Page 7 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT be ASME Class 1, 2, or 3 if classifications were based only on RG 1.26.

10CFR50.55a and 10CFR50, Appendix B, Criterion XI, require that all components be tested commensurate with their importance to safety regardless of Code classification.

Some non-ASME Code Class components are included in the PBNP IST Program as Augmented components. Additionally, some ASME Code Class components may have tests specified in the IST Program that are not required by the ASME OM Code but are performed based on a commitment.

3.5 Component Selection for Accident Mitigation The NRC staff provides guidance in NUREG-1482, Section 2.2.1 such that the language in ISTA-1100 for applicable components and 10 CFR 50.2, "Definitions" for safety-related components are consistent with respect to defining components for mitigating the consequences of accidents. Component selection can be based on design basis accidents in the FSAR. Design basis accidents are worst case scenarios which define bounding consequences. Less severe scenarios may exist which may still result in core damage and threaten the health and safety of the public. Therefore, the scope of the PBNP IST Program includes all components which function to prevent, or mitigate the consequences of, any accident which could result in off-site doses in excess of 10CFRl 00 or 10CFR50.67 limits.

Safety related systems are required to be capable of performing their safety function during and following design basis events given the most limiting single active component failure. However, where multiple components are capable of performing the same equivalent and redundant specified function (e.g., multiple valves closing in series) and where the components are not supplied by alternate and redundant power supplies, or are not required to meet single failure criteria, only one of the redundant components need be included in the IST Program. The component must be relied upon to perform and not simply have the capability of performance. This exemption only applies where licensing documents do not take credit for the designed redundancy. Components performing redundant functions shall be included in the testing program if, in the process of analysis or licensing justification, they are relied upon to be operable.

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POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 3.6 Beyond-Design-Basis Events Consistent with industry practice and guidance in NUREG-1482 Section 2.2.1, components required solely to mitigate the consequences of 10CFR50 Appendix R fires and station blackout events are outside the scope of the IST Program since these events are beyond the facility design basis. Beyond-design-basis events are initiated by multiple (and sometimes complete) failures of safety-related components and systems. The facility design is based on requirement that each safety system be capable of performing its safety-related functions given a failure of the most limiting active component.

Although regulations have been imposed that require the capability to cope with, or to mitigate these events, they are outside the scope of the facility accident analyses.

Components whose sole safety functions are to mitigate beyond-design-basis events are not required by regulations to be classified as safety-related. These components are non-safety-related but are classified as QA scope, augmented quality (AQ), per the PBNP Q list, and are not within the scope of the IST Program.

3.7 Inservice Test Frequencies As outlined in NUREG-1482, the intent of the ASME OM Code is that inservice tests be performed at the specified frequency with the actual time between tests being approximately equal. Based on Table ISTA-3170-1, test frequencies shall be defined as follows:

Table 3.7-1 Inservice Test Frequencies Stipulated Code or Technical Specification Specified Time Between Frequency Tests (at least once every)

Semiquarterly (for increased pump test 46 days quarterly)

Quarterly (or Every 3 Months) 92 Days Semiannually (or Every 6 Months) 184 days Annually (or Every Year) 366 days xyears x calendar years where x is a whole number?: 2 Refueling Every Refueling Outage

(:=::;18 months)

Biennially (or Every 2 Years) 24 months Page 9 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT In some cases, the Technical Specifications require a higher frequency for a given test as compared to the ASME OM Code for the same test. In such instances, the more conservative test (i.e., higher frequency) will be performed.

3.8 Inservice Examination and Test Frequency Grace As determined by TSTF-545, "TS Inservice Testing Program & Clarify SR Usage Rule Application to Section 5.5 Testing," the Inservice Testing Program (IST) has been deleted from the PBNP Technical Specifications. A new defined term, "Inservice Testing Program, 11 has been added to the definitions section of the TS and is defined as "The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10CFR50.55a paragraph (f). 11 Section IST specifies component test frequencies based either on elapsed time periods (e.g., quarterly, 2 year, etc.) or the occurrence of plant conditions or events (e.g., cold shutdown, refueling outage, upon detection of a sample failure, following maintenance, etc.). Components whose test frequencies are based on elapsed time periods shall be tested at the frequencies specified in Section IST with a specified time period between tests defined in Table ISTA-3170-1. The specified time period between tests may be extended by up to 25% for any given test period up to and including two years. For testing periods greater than 2 years, ISTA-3170 allows an extension of up to 6 months.

All periods may be reduced at the discretion of the Owner. Components whose test frequencies are based on the occurrence of plant conditions may not have their period between tests extended except as allowed by Section IST.

Period extension is to facilitate test scheduling and considers plant operating conditions that may not be suitable for performance of the required testing. Period extensions are not intended to be used repeatedly merely as an operational convenience to extend test intervals beyond those specified.

The effect of a missed inservice test on the Operability of TS equipment will be assessed under the operability determination program.

Period extensions may also be applied to accelerated test frequencies (e.g., pumps in the alert range) and other fewer than 2-year test frequencies not specified in Table ISTA-3170-1. The allowable interval for quarterly pump testing is 115 days, increasing this frequency, in accordance with ASME OM Code corrective action requirements, results in a maximum interval of 57 days for the conduct of an increase frequency pump test.

Refer to ER-AA-113-1000 Section 4.11.2 for the Fleet technical position on grace and missed surveillances.

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POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 3.9 Testing Non-Code Class Components (10CFR50.55a Augmented IST Program)

The inservice test requirements of components that are within the scope of the ASME OM Code but are not classified as ASME Code Class 1, 2, or 3 are performed under PBNP's Augmented Inservice Testing Program per 10CFR50.55a paragraph (f)(6)(ii).

Augmented inservice testing is performed in accordance with ASME OM Code requirements, unless implementing the Code provisions would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Any augmented testing OM Code deviations need not be addressed through the relief request process but should still be documented and available for NRC review per 10CFR50.55a paragraph (f)(4). The code deviation and its basis will be maintained in the applicable system IST Background Document. Refer to ER-AA-113-1000 Section 4.11.6 for the Fleet technical position on non-Code (Augmented) Components.

3.10 Risk-Informed Testing for ASME OM Code Components Active valves within the scope of Mandatory Appendices III and IV utilize risk-informed testing that incorporates risk insights to establish acceptance criteria, exercising requirements, and testing intervals. Each valve within the scope of Mandatory Appendices III and IV is categorized as either a high or low safety significant component (HSSC I LSSC) as indicated in the PBNP IST Program Background Document.

Subsection ISTE, "Risk-Informed Inservice Testing of Components in Water-Cooled Reactor Nuclear Power Plants," of the ASME OM Code is not invoked at PBNP.

3 .11 Pre-Maintenance and Post Maintenance Testing The post-maintenance testing, when performed on an IST related component, shall either reconfirm existing reference values or be used to establish new baseline data for comparison during subsequent testing.

The equipment test should be performed using the appropriate approved procedure. (Ref.

INPO SOER 98-1).

Specific post-maintenance testing requirements for various maintenance activities on plant components are called out in MA-AA-203-1000, Maintenance Functional Testing.

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POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND 2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 4.0 DEFINITIONS 4.1 Active Valves - Valves which are required to change obturator position to accomplish specific function(s) for accident mitigation or achieving/maintaining safe shutdown.

Active may also refer to a particular valve position with respect to safety function.

4.2 Administrative Controls - A valve shall be considered to be under administrative control, if; the valve is locked or de-energized in its normal position, or procedurally controlled if mispositioned. Administrative controls may also consist of stationing a dedicated operator at the controls of the valve, who is in continuous communication with the control room.

4.3 Closed System - A system or portion of a system serving an extension of the containment liner for the purpose of maintaining containment integrity. Designated closed systems shall be protected from missiles and high energy line breaks as well as designed in accordance with Class I seismic criteria.

4.4 Cold Shutdown Outage - Applies to each nonrefueling outage period in which the cold shutdown mode, as defined by plant technical specifications, is entered.

4.5 Comprehensive Pump Test Flow Rate - the flow rate established by the Owner that is effective for detecting mechanical and hydraulic degradation during subsequent testing.

4.6 Containment Isolation Valve - Valves providing a barrier between the containment environment and the outside environment which must be capable of closure to maintain containment integrity. Containment isolation valves are listed in FSAR Section 5.2. Per TS B 3.6.3, Containment Isolation Valves, no specific containment isolation time was assumed in the LOCA analysis.

4.7 Control Valve - valves used only for system control, such as pressure regulating valves.

4.8 Design Bases - That information which identifies the specific functions to be perfom1ed by a structure, system, or component of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design. These values may be (1) restraints derived from generally accepted practices for achieving functional goals, or (2) requirements derived from analysis (based on calculation and/or experiments) of the effects of a postulated accident for which a structure, system, or component must meet its functional goals.

4.9 Event V PIVs - Two check valves in series at the reactor coolant system pressure boundary interface with a low pressure system which penetrates containment. Failure of Event V check valves during a LOCA may result in leakage bypassing containment.

Further information is contained in TRM 4.16 RCS PIV Program.

4 .10 Exercising - The demonstration based on direct visual or indirect positive indications that the moving parts of a component function satisfactorily.

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POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 4.11 Fail-Safe - Characteristic of a valve and its actuator, which upon loss of actuating energy supply (typically instrument air and/or electrical control power), will cause a valve obturator member to be fully closed, fully open, or remain in the last position, as necessary to satisfy the specific function as defined in para. ISTA-1100.

4.12 Fail-Safe Test - A test of a valve with fail-safe functions, performed by observing the operation of the actuator upon loss of valve actuating power.

4.13 Full Cycle Exercise Full stroke of the valve to full open and closed position returning to its initial position.

4.14 Full Stroke Time - The time interval from initiation of the actuating signal to the indication of the end of the operating stroke (switch to light).

4.15 Group A Pumps - Pumps that are operated continuously or routinely during normal operation, cold shutdowns, or refueling operations.

4.16 Group B Pumps - Pumps in standby systems that are not operated routinely except for testing.

4.17 Instrument Loop Accuracy - The allowable inaccuracy of an instrument loop based on the square root of the sum of the square of the inaccuracies of each instrument or component in the loop when considered separately. Alternatively, the allowable inaccuracy of the instrument loop may be based on the output for a known input into the instrument loop.

4.18 Instrument Loop - Two or more instruments or components working together to provide a single output (e.g., a vibration probe and its associated signal conditioning and readout devices).

4.19 Limiting Value of Full-Stroke Time (LVFST) - The calculated or owner-specified maximum allowable valve stroke time limit established to assure that corrective action is taken on a degraded valve before it reaches the point where there is a high probability of failure to perform its safety function if called upon. If a design, Technical Specification, FSAR, or accident analysis limit exists which is more limiting, then it shall be used as the limiting value of full-stroke time in lieu of the calculated value.

4.20 Non-intrusive Testing - Testing performed on a component without disassembly or disturbing the boundary of the component.

4.21 Obturator - Valve closure member (disk, gate, plug, ball, etc.).

4.22 Operational Readiness - The ability of a pump or valve to perform its intended function.

4.23 Passive Valves - Valves which maintain obturator position and are not required to change obturator position to accomplish their safety function as defined in ISTA-1100.

Passive may also refer to a particular valve position with respect to safety function.

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POINT BEACH NUCLEAR PLANT 1ST Program Document UNITS 1 AND 2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 4.24 Performance Assessment Test - Performance testing consisting of the measurement and assessment of applicable test parameters. This test is normally performed at ambient conditions without system pressure or flow.

4.25 Plant Operation - The conditions of startup, operation at power, hot standby, and reactor cooldown, as defined by the plant Technical Specifications.

4.26 Pneumatically Operated Valve - A valve and its associated actuator, including pneumatically operated power-operated relief valves (PORVs) as defined in ISTC-2000, that uses air/gas as the motive force, including all subcomponents required for the valve assembly to perform its specific function as defined in ISTA-1100, except those exempted by para. IV-1300. For simplicity, this type of valve is referred to as an "AOV" throughout the 1ST Program Document.

4.27 Post Maintenance Testing - Required testing for equipment subsequent to performance-altering maintenance. The applicable portion of the Technical Specifications and/or ASME Codes shall be referenced for determination of appropriate post maintenance testing.

4.28 Preconditioning - The act of exercising or placing a component in service for the purpose of enhancing the results of an inservice test. The act of preconditioning could mask a degrading condition that may otherwise be detected when testing a component in the "as found" condition. The exercising of a component when required by an Operating Procedure (OP) or Instruction (01) for system reconfiguration shall not constitute or be considered as preconditioning. However, if components are routinely exercised prior to Inservice Testing, then the Inservice Test should be performed prior to exercising if practicable. Additional discussion can be found in References 5.13 and 5.14.

Refer to ER-AA-113-1000 Section 4.11.1 for the Fleet technical position on preconditioning.

NOTE: If maintenance activities are scheduled concurrent with AOV, Motor Operated Valve (MOV), or pressure relief valve periodic testing, then the periodic performance test shall be conducted in the as-found condition prior to maintenance activities, where practicable. [Ref. 1-3300, III-3300(b), IV-3410(d)]

4.29 Preservice Test - A test performed after completion of construction activities related to the component and before first electrical generation by nuclear heat, or in an operating plant, before the component is initially placed in service. The results are used to establish reference values for future tests. Typically performed after a new component is added to the plant prior to placing in service or following maintenance that could affect component performance.

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POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 4.30 Preservice Test Period - The period of time following completion of construction activities related to the component and before first electrical generation by nuclear heat, or in an operating plant, before the component is placed in service.

4.31 Pump Periodic Verification Test - a test that verifies a pump can meet the requirements (differential or discharge) pressure as applicable, at its highest design basis accident flow rate.

4.32 Pressure Isolation Valves - Two normally closed valves in series that form the reactor coolant pressure boundary and isolate reactor coolant system pressure from an attached low pressure system.

4.33 Reactor Coolant System Pressure Boundary - All those pressure retaining components of boiling and pressurized water reactors such as pressure vessels, piping, pumps, and valves which are:

4.33.1 Part of the reactor coolant system or, 4.33.2 Connected to the reactor coolant system, up to and including any and all of the following:

a. The outermost containment isolation valves in system piping which penetrates primary containment,

b. The second of two valves normally closed during normal reactor operation in system piping which does not penetrate primary containment,

c. The reactor coolant system safety and relief valves.

4.34 Reactor Coolant System Pressure Isolation - The function that prevents intersystem overpressurization between the reactor coolant system and connected low pressure systems.

4.35 Reference Point - A point of operation at which reference values are established and inservice test parameters are measured for comparison with applicable acceptance criteria.

4.36 Reference Values - One or more values oftest parameters measured or determined when the equipment is known to be operating acceptably.

4.37 Refueling Outage -Applies to the normally scheduled once-per-cycle outage period in which the refueling mode, as defined by plant technical specifications, is entered.

4.38 Routine Servicing -The performance of planned, preventive maintenance (e.g. replacing or adjusting valves in reciprocating pumps, changing oil, flushing the cooling system, adjusting packing, adding packing rings or mechanical seal maintenance or replacement).

Page 15 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 4.39 Safety-Related - Designation applied to components that are relied upon to remain functional during and following design basis events to assure:

4.39.1 The integrity of the reactor coolant pressure boundary, 4.39.2 The capability to shut down the reactor and maintain it in a safe shutdown condition, or 4.39.3 The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in 10CFRlO0 or 10CFR50.67.

4.40 Single Failure - An occurrence which results in the loss of a capability of a component to perform its intended safety functions. Multiple failures resulting from a single occurrence are considered to be a single failure. Fluid and electric systems are considered to be designed against an assumed single failure if neither (1) a single failure of any active component (assuming passive components function properly) nor (2) a single failure of a passive component (assuming active components function properly) results in a loss of the capability of the system to perform its safety functions.

4.41 Skid-Mounted Pumps and Valves - Pumps and valves which are integral to or that support operation of major components, even though these pumps and valves may not be located on the skid. In general, these pumps and valves are supplied by the manufacturer of the major component. Examples include diesel fuel oil pumps and valves, steam admission and trip throttle valves for turbines, and solenoid operated pilot valves used to control pneumatically/air operated valves (AOVs). Further discussion is provided in NUREG-1482, Sections 3.4 and4.1.10.

4.42 Stroke Test - Exercising the valve by operating one complete open and close cycle (i.e.,

"Full Cycle Exercise") and testing that includes a full-stroke time.

4.43 System Resistance - The hydraulic resistance to flow in a system.

4.44 Trending - A comparison of current data to previous data obtained under similar conditions for the same equipment.

4.45 Valve Category - The ASME Code defines test requirements by valve categories. All valves in the IST Program are assigned to one of the following categories:

4.45.1 Category A - Valves for which seat leakage is limited to a specific maximum amount in the closed position for fulfillment of their function.

4.45.2 Category B - Valves for which seat leakage in the closed position is inconsequential for fulfillment of their function.

4.45.3 Category C - Valves which are self-actuating in response to some system Page 16 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT characteristic, such as pressure or flow (check or relief valves) for fulfillment of their function.

4.45.4 Category D - Valves which are actuated by an energy source capable of only one operation such as rupture disks or explosive-actuated valves.

4.46 Vertical Line Shaft Pump - A vertically suspended pump, where the pump driver and the pumping element are connected by a line shaft within an enclosing column which contains the pump bearings, making pump bearing vibration measurements impracticable.

Page 17 of56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND 2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT

5.0 REFERENCES

5.1 Title 10, Code of Federal Regulations (10CFR), Part 50, Domestic Licensing of Production and Utilization Facilities 5.1.2 10CFR50.34a, Design objectives for equipment to control releases of radioactive material in effluents - nuclear power reactors.

5.1.3 10CFR50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants 5.1.4 10CFR50 Appendix R, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979 5.1.5 10CFR50 Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors 5.1.6 10CFR50.55a (updated 01/19/2021), Codes and Standards 5.1.7 10CFR50.67, Accident Source Term 5.2 10CFR, Part 100, Reactor Site Criteria 5.3 NRC Publications 5.3.1 NUREG-0800, Rev. 3, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants 5.3.1.1 Section 3.2.2, System Quality Group Classification 5.3.1.2 Section 3.9.6, Inservice Testing of Pumps and Valves 5.3.1.3 Section 5.4. 7, Design Requirements of the RHR System 5.3.2 NUREG-1482, Rev. 3, Guidelines for Inservice Testing at Nuclear Power Plants 5.3.3 Regulatory Guide (RG) 1.26, Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants 5.3.4 RG 1.192, Rev. 3, Operation and Maintenance Code Case Acceptability, ASME OM Code 5.3.5 Generic Letter (GL) 87-06, Periodic Verification of Leak Tight Integrity of Pressure Isolation Valves 5.3.6 GL 89-04, Guidance on Developing Acceptable Inservice Testing Programs

Point Beach Nuclear Plant Responses to GL 89-04, dated October 3, 1989, March 2, 1990, June 28, 1990, and September 11, 1990

NRC minutes of public meetings on GL 89-04, dated October 25, 1989 Page 18 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 5.3.7 GL 90-06, Resolution of Generic Issues 70, "PORVand Block Valve Reliability," and 94, "Additional Ltop Protection for PWRs" 5.3.8 NRC Technical Position RSB 5-1 5.3.9 TSTF-545, TS Inservice Testing Program & Clarify SR Usage Rule Application to Section 5.5 Testing 5.4 ASME OM Code 2017 Edition, "Operation and Maintenance ofNuclear Power Plants" 5.4.1 Code Case OMN-3 Revision 0, Requirements for Safety Significance Categorization of Components Using Risk Insights for Inservice Testing ofLWR Power Plants 5.4.2 Code Case OMN-17 Revision 0, Alternative Rules for Testing ASME Class 1 Pressure Relief/Safety Valves 5.4.3 ASME OM Code Interpretation 96-01 5.4.4 ASME OM Code Interpretation 99-03 5.4.5 ASME OM Code Interpretation 99-07 5.4.6 ASME OM Code Interpretation 01-02 5.5 Final Safety Analysis Report PBNP Units 1 & 2 5.6 Technical Specifications, PBNP Units 1 & 2 5.6.1 PBNP Units 1 and 2 -Issuance of License Amendments Re.:

Auxiliary Feedwater System Modification (TAC Nos ME1081 and ME1082) March 25, 2011 5.7 NRC Safety Evaluation Report (SER)

5. 7 .1 NRC SER dated April 7, 2017, Issuance of Amendments Regarding Technical Specifications for Inservice Testing (Amendment Nos. 259 and 263, respectively). ML l 7027A078.

5. 7 .2 NRC SER dated May 3, 2011, Point Beach Nuclear Plant (PBNP),

Units 1 And 2 - Issuance of License Amendments Regarding Extended Power Uprate (TAC Nos. MEl 044 AND MEl 045)

5. 7 .3 NRC SER, dated May 4, 2006, on the Point Beach Nuclear Plant Inservice Testing Program, Fourth 10 Year Interval (manual valve testing frequency) 5.7.4 NRC SER, dated January 30, 2003, on the Point Beach Nuclear Plant In Service Testing Program, fourth 10 year interval.

5.7.5 NRC SER, dated December 12, 1994, on the Point Beach Nuclear Plant Inservice Testing Program, Third 10 Year Interval 5.7.6 NRC SER, dated October 28, 1993, on the Point Beach Nuclear Plant Inservice Testing Program, Third 10 Year Interval Page 19 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 5.7.7 NRC Safety Evaluation Report (SER), dated April 17, 1992, on the Point Beach Nuclear Plant Inservice Testing Program, Third 10 Year Interval 5.8 NextEra Energy Fleet Procedures 5.8.1 ER-AA-113, Inservice Testing (IST) Program 5.8.2 ER-AA-113-1000, Inservice Testing Procedure 5.8.3 ER-AA-119-1000, Snubber Program Procedure 5.8.4 MA-AA-202, Work Order Execution Process 5.8.5 MA-AA-203-1000, Maintenance Functional Testing 5.8.6 PI-AA-104-1000, Condition Reporting 5.9 PBNP Program Procedures 5.9.1 CMP 2.1, Check Valves 5.9.2 CMP 2.2, Motor Operated Valves 5.9.3 CMP 2.3, Relief/ Safety Valves 5.9.4 CMP 2.5, AOV Program Document 5.9.5 CVCM PROGRAM DOCUMENT, Check Valve Condition Monitoring (CVCM) Program Document 5.9.6 CLRT Testing Program, CLRT Testing Program Basis Document 5.10 PBNP Evaluations 5.10.1 Engineering Evaluation 2001-0019 for Packing Re-consolidation of MOVs and PMT Requirements 5.10.2 CCE 2000-004 Pump Testing Increased Frequency 5.10.3 NPM 99 1313, ASME XI Valve Categorization 5.10.4 NPM 01 0736, CR 01 0647 #1, Increased Frequency Pump Testing 5.11 PBNP TRM 4.16, RCS PIV Program 5.12 NPC 2015-00144 ISTOG Position Paper, Position on IST Component Preconditioning. February 28, 201l(NRC ADAMS MLll 1930100) 5.13 CR 2008028, NRC Identified Preconditioning, initiated 11/19/2014 5.14 NPL 98-0242, Additional Reply of a Notice of Violation, 4/2/1998 5.15 INPO SOER 98-1, Safety System Status Control Page 20 of 56

POINT BEACH NUCLEAR PLANT 1ST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 6.0 PUMP IST PROGRAM 6.1 Pump Selection Criteria and Exemptions 6.1.1 The basic scope of the pump IST Program is defined in Subsections ISTA and ISTB of the ASME OM Code. Per paragraphs ISTA-1100 and ISTB-1100, 1ST requirements apply to certain centrifugal and positive displacement pumps which are provided with an emergency power source, that are required in shutting down the reactor to a safe shutdown condition, maintaining the safe shutdown condition, or mitigating the consequences of an accident.

6.1.2 Drivers are exempt from ASME OM Code testing requirements except where the pump and driver form an integral unit and the pump bearings are located in the driver.

6.1.3 Pumps which do not perform a function within the scope of the ASME OM Code but are supplied with emergency power solely for operating convenience are exempt from ASME Code testing requirements.

6.1.4 Skid-mounted pumps and component subassemblies that are tested as part of the major component and are determined by the Owner to be adequately tested are exempt. (i.e. hydraulic pumps for valve actuation, diesel fuel oil pumps, etc.)

6.2 Group A Pumps The ASME OM Code defines Group A pumps as those pumps that are operated continuously or routinely during normal operation, cold shutdown, or refueling operations. The following PBNP Unit 1 and 2 pumps meet this definition.

lP-53 and 2P Motor Driven Auxiliary Feedwater Pumps (MDAFWP). The MDAFWPs may be utilized during startup from refueling outages to fill the steam generators and to maintain steam generator level prior to initiation of normal feedwater. Note that this non-safety-related function of filling and maintaining the Steam Generator levels is normally performed by the 0P-38A/B, Standby Steam Generator Pumps. However, PBNP shall conservatively consider the MDAFWPs as Group A pumps.

lP-29 and 2P Turbine Driven Auxiliary Feedwater Pumps (TDAFWP). The TDAFWPs may be utilized during startup from refueling outages to fill the steam generators and to maintain steam generator level prior to initiation of normal feedwater. Note that this non-safety-related function of filling and maintaining the Steam Generator levels is normally performed by the 0P-38A/B, Standby Steam Generator Pumps. However, PBNP shall conservatively consider the TDAFWPs as Group A pumps.

Page 21 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 1/2P-11A and 1/2P-11B- Component Cooling Water Pumps. The CCW pumps operate continuously during normal plant operation to supply cooling water to non-essential heat loads as well as cooling water to the RCP motor bearings and thermal barrier.

1/2P-10A and 1/2P-10B - Residual Heat Removal Pumps. The RHR pumps are required to operate when maintaining the plant in a cold shutdown condition.

0P-32A through 0P-32F - Service Water Pumps. Service water pumps operate continuously during normal plant operation to supply cooling water to non-essential heat loads.

0P-12A and 0P-12B - Spent Fuel Pool Cooling Pumps. The spent fuel pool cooling pumps provide heat removal from stored fuel assemblies in the spent fuel pool by circulating fuel pool inventory through the spent fuel pool heat exchangers, allowing the residual heat to be transferred to the service water system.

1/2P-15A and 1/2P-15B - Safety Injection Pumps. The safety injection pumps are not routinely utilized during any plant operating evolution. The pumps remain in standby during all operating Modes. The pumps are required to operate only during a loss-of-coolant accident (LOCA) to provide high head safety injection and recirculation flow to the RCS, and for long term shutdown cooling during post-LOCA conditions. It should be noted that the high head safety injection pumps are periodically utilized for filling the accumulators and conservatively, based on this activity, PBNP has assigned them a Group A categorization.

6.2.1 Group A Pump Testing Frequency An inservice test shall be run on each Group A pump as specified in Table ISTB-3400-1. Table ISTB-3400-1 specifies that a Group A test be run quarterly (92 days) and a comprehensive test be performed on Group A pumps biannually (731 days). Additionally, pumps required to have a periodic verification test have this test performed biennially.

As allowed by Code, the performance of a comprehensive test may be used to satisfy Quarterly Group A test requirements (ISTB-5000). Based on the accuracy of installed pressure gauges, a number of comprehensive tests are performed on a quarterly frequency.

6.2.2 Group A Pump Test Parameters The test parameters for the Group A test, the comprehensive test, and the Periodic Verification test for Group A pumps are specified in Table ISTB-3000-1 and are reflected below in Table 6.2-1.

Table 6.2-1 Page 22 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND 2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT Group A Inservice and Comprehensive Test Parameters Periodic Group A Comprehensive Verification Test Quantity Test Test [Note l] Remarks Speed, N X X X If variable speed Differential Pressure, M X X X Centrifugal pumps, including vertical line shaft pumps Discharge Pressure, P X X X Positive displacement pumps Flow Rate, Q X X X Vibration X X Not Required Measure either Displacement (Vct) Peak-to-peak (Vct)

Velocity (Vv) Peak (Vv)

Notes:

(1) Only required for pumps identified in Div. 1 Mandatory Appendix V in the OM Code and summarized in Section 6.4.2.

6.2.3 Required Instrument Accuracy Instrument accuracy shall be within the limits specified in Table ISTB-3510-1. If a parameter is determined by analytical methods instead of measurement, then the determination shall meet the parameter accuracy requirement of Table ISTB-3510-1. For individual analog instruments, the required accuracy is percent of full scale. For digital instruments, the required accuracy is over the calibrated range. For a combination of instruments, the required accuracy is loop accuracy. Table 6.2-2 below reflects the required instrument accuracies for both the Group A test and Group B test as well comprehensive testing applicable to Group A and Group B pumps and Preservice tests.

Table 6.2-2 Required Instrument Accuracy (%)

Group A and Group B Comprehensive Quantity Tests and Preservice Tests Pressure +/-2 +/-1/2 Flow Rate +/-2 +/-2 Speed +/-2 +/-2 Vibration +/-5 +/-5 Differential Pressure +/-2 +/-1/2 Page 23 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 6.2.4 Allowable Ranges for Group A Pump Hydraulic Test Parameters The allowable ranges for Group A hydraulic test parameters are specified in Tables ISTB-5121-1 (Centrifugal Pumps), ISTB-5221-1 (Vertical Line Shaft Centrifugal Pumps) and ISTB-5321-1 (Positive Displacement Pumps) and are reflected below in Table 6.2-3. All test values shall be compared to these reference values to determine if pump performance is acceptable. It should be noted that the acceptance criteria defining Alert Range and Required Action Range are less stringent than the acceptance range imposed on the hydraulic test parameters associated with the biennially comprehensive test. When determining acceptance criteria, design basis calculations and pipe design pressures control the minimum and maximum acceptance criteria. The most limiting of the ISTB or design basis acceptance criteria are then used for testing, as appropriate.

Table 6.2-3 Group A Test Hydraulic Acceptance Criteria Test Parameter Acceptable Alert Range Range Required Action Range Low High P (Positive displacement pumps) 0.93 to 1.10 Pr 0.90 to< 0.93 Pr < 0.90 Pr > 1.l0Pr

.6.P (Vertical. line shaft pumps) 0.95 to 1.10 Af\ 0.93 to< 0.95 Mr <0.93 Mr > 1.10 Mr Q (Positive displacement and 0.95 to 1.10 Qr 0.93 to < 0.95 Qr < 0.93 Qr > 1.10 Qr vertical line shaft pumps)

.6.P (Centrifugal pumps) 0.90 to 1.10 LlPr none < 0.90 LlPr > 1.10 Mr 0 (centrifugal pumps) 0.90 to 1.10 Qr none < 0.90 Qr > 1.10 Qr Subscript r denotes reference value.

6.2.5 Allowable Ranges for Group A Pump Test Vibration Data The allowable ranges for Group A pump tests vibration data is specified in Tables ISTB-5121-1 (Centrifugal Pumps), ISTB-5221-1 (Vertical Line Shaft Centrifugal Pumps), and ISTB-5321-1 (Positive Displacement Pumps) and are reflected below in Table 6.2-4. All test values shall be compared to these reference values to determine if pump performance is acceptable. It should noted that the preceding Tables include acceptance criteria for centrifugal and vertical line shaft pumps that have a pump speed of <600 rpm. The PBNP IST Program does not contain any pumps with a rotating speed of <600 rpm.

Therefore, this acceptance criteria has been omitted from Table 6.2-4.

Page 24 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT Table 6.2-4 Group A Tests Vibration Acceptance Criteria 1 Pump Type PumpS Test Para Acceptable Alert Range Required peed meter Range Action Range Centrifugal and ve1iical ~600rpm Vv orVct :;:; 2.5Vr > 2.5Vr to 6 >6Vr or line shaft [Note (2)] Vr or >0.325 to >0.7 in/sec 0.7 in/sec Reciprocating VctorVv :;:;2.5Vr > 2.5Vr to 6 Vr >6Vr Notes:

(1) Vibration parameter is per Table ISTB-3000-1. Vr is vibration reference value in the selected unit.

(2) Including positive displacement pumps except reciprocating.

6.2.6 Group A Test Procedure Group A tests shall be conducted with the pump operating at a specified reference point. After pump conditions are as stable as the system permits, each pump shall be run at least 2 minutes. At the end of this time, at least one measurement or detem1ination of each of the parameters shall be made and recorded. The test parameters shown in Table ISTB-3000-1 (Table 6.2-1) shall be determined and recorded as required by paragraphs ISTB-5121, ISTB-5221 and ISTB-5321.

6.3 Group B Pumps The OM Code defines Group B pumps as those pumps in standby systems that are not operated routinely except for testing. The following PBNP Unit 1 and 2 pumps meet this definition.

1/2P-14A and 1/2P-14B - Containment Spray Pumps. The containment spray pumps are not utilized during any plant operating evolution. The pumps remain in standby during all operating Modes. The pumps are required to operate only during a loss-of-coolant accident (LOCA) or main steam line break (MSLB) inside containment for containment heat removal and pressure suppression. The containment spray system also serves in removing fission products released into the containment atmosphere during a LOCA by the admission of sodium hydroxide to the spray stream.

0P-206A&B and 0P-207A&B - EDG Fuel Oil Transfer Pumps. The fuel oil transfer pumps are not utilized during any plant operating evolution. The pumps remain in standby during all operating Modes. The pumps are required to operate Page 25 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT only during emergency diesel generator operation to replenish day tank inventory.

6.3.1 Group B Pump Testing Frequency An inservice test shall be run on each Group B pump at the frequency specified in Table ISTB-3400-1. Table ISTB-3400-1 specifies that a Group B test be run quarterly (92 days) and a comprehensive test be performed on Group B pumps biannually (731 days). Additionally, pumps required to have a periodic verification test have this test performed biennially.

As allowed by Code, the performance of a comprehensive test may be used to satisfy Quarterly Group B test requirements (ISTB-5000). Based on the accuracy of installed pressure gauges, a number of comprehensive tests are performed on a quarterly frequency.

6.3.2 Group B Pump Test Parameters The test parameters for the Group B test, and the comprehensive test, and the Periodic Verification test for Group B pumps is specified in Table ISTB-3000-1 and are reflected below in Table 6.3-1.

Table 6.3-1 Group B Inservice and Comprehensive Test Parameters Periodic Comprehensive Verification Test Quantity Group B Test Test [Note (2)1 Remarks Speed, N X X X If variable speed Differential X [Note (1)] X X Centrifugal pumps, Pressure,@ including vertical line shaft pumps Discharge Pressure, Not Required X X Positive displacement p pumps Flow Rate, Q X !Note (l)l X X Vibration Not Required X Not Required Measure either Displacement (Vct) Peak-to-peak (Vct)

Velocity (Vv) Peak (Vv)

Notes:

(1) For positive displacement pumps, flow rate shall be measured or determined; for all other pumps, differential pressure or flow rate shall be measured or determined.

(2) Only required for pumps identified in Div. 1 Mandatory Appendix V in the OM Page 26 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND 2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT Code and summarized in Section 6.4.2.

6.3.3 Required Instrument Accuracy Instrument accuracy shall be within the limits specified in Table ISTB-3510-1. If a parameter is determined by analytical methods instead of measurement, then the determination shall meet the parameter accuracy requirement of Table ISTB-3510-1. For individual analog instruments, the required accuracy is percent of full scale. For digital instruments, the required accuracy is over the calibrated range. For a combination of instruments, the required accuracy is loop accuracy. Table 6.2-2 in Section 6.2 reflects the required instrument accuracies for both the Group A test and Group B test as well comprehensive testing applicable to Group A and Group B pumps and Preservice tests.

6.3.4 Allowable Ranges for Group B Pump Hydraulic Test Parameters The allowable ranges for Group B hydraulic test parameters are specified in Tables ISTB-5121-1 (Centrifugal Pumps), ISTB-5221-1 (Vertical Line Shaft Centrifugal Pumps) and ISTB-5321-1 (Positive Displacement Pumps) and are reflected below in Table 6.3-2. All test values shall be compared to these reference values to determine if pump performance is acceptable. It should be noted that the acceptance criteria defining Required Action Range is less stringent than the acceptance range imposed on the hydraulic test parameters associated with the biennially comprehensive test. In addition, there is no Alert Range applicable to Group B testing. When determining acceptance criteria, design basis calculations and pipe design pressures control the minimum and maximum acceptance criteria. The most limiting of the ISTB or design basis acceptance criteria are then used for testing, as appropriate.

Table 6.3-2 Group B Test Hydraulic Acceptance Criteria Test Parameter Acceptable Ra Required Action Range nge Low Hi2h LiP (centrifugal pumps including 0.90 to 1.10 Af\ <0.90Mr > 1.10 Mr Vertical. line shaft pumps) OR 0 (All pump types) [See Note (l)] 0.90 to 1.10 Qr < 0.90 Qr > 1.10 Qr Subscript r denotes reference value.

Note:

(1) Measure Q for positive displacement pumps.

6.3.5 Allowable Ranges for Comprehensive Pump Test Vibration Data Page 27 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND 2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT The allowable ranges for the comprehensive pump test vibration data, when applied to Group pumps, is specified in Tables ISTB-5121-1 (Centrifugal Pumps), ISTB-5221-1 (Vertical Line Shaft Centrifugal Pumps) and ISTB-5321-1 (Positive Displacement Pumps) and are reflected below in Table 6.3-3. All test values shall be compared to these reference values to determine if pump performance is acceptable. It should be noted that the preceding Tables include acceptance criteria for centrifugal and vertical line shaft pumps that have a pump speed of <600 rpm. The PBNP IST Program does not contain any pumps with a rotating speed of <600 rpm. Therefore, this acceptance criteria has been omitted from Table 6.3-3. In addition, vibration data is not a required parameter for Group B tests, only when performing a comprehensive test on group B pumps.

Table 6.3-3 Comprehensive Tests Vibration Acceptance Criteria1 Pump Type PumpS Test Para Acceptable Alert Range Required peed meter Range Action Range Centrifugal and vertical :C:600rpm VvorVct s 2.5Vr > 2.5VrtO 6 >6Vr or line shaft [Note (2)] Vr or >0.325 to >0.7 in/sec 0.7 in/sec Reciprocating Vct orVv s 2.5Vr > 2.5Vr to 6 Vr >6Vr Notes:

(1) Vibration parameter is per Table ISTB-3000-1. Vr is vibration reference value in the selected unit.

(2) Including positive displacement pumps except reciprocating.

6.3.6 Group B Test Procedure Group B tests shall be conducted with the pump operating at a specified reference point. Once pump conditions are as stable as the system permits, measurement or determination of at least one test parameter shall be made and recorded. The test parameters shown in Table ISTB-3000-1 4.1-1 (Table 6.3-1) shall be determined and recorded as required by paragraphs ISTB-5122, ISTB-5222 and ISTB-5322.

Page 28 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND 2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 6.4 Comprehensive Test Procedure Comprehensive tests shall be conducted with the pump operating at a specified reference point. After pump conditions are as stable as the system permits, each pump shall be run at least 2 minutes. At the end of this time, at least one measurement or determination of each of the parameters shall be made and recorded. The test parameters shown in Table ISTB-3000-1 (Table 6.2-1 Group A and Table 6.3-1 Group B) shall be determined and recorded as required paragraphs ISTB-5123, ISTB-5223 and ISTB-5323.

Refer to Technical Position 18 for justification of comprehensive flow rates.

6.4.1 Allowable Ranges for Comprehensive Hydraulic Test Parameters The allowable ranges for comprehensive hydraulic test parameters are specified in Tables ISTB- 5121-1 (Centrifugal Pumps), ISTB-5221-1 (Vertical Line Shaft Centrifugal Pumps) and ISTB-5321-1 (Positive Displacement Pumps) and are reflected below in Table 6.4-1. All test values shall be compared to these reference values to determine if pump performance is acceptable. It should be noted that the acceptance criteria defining Alert Range and Required Action Range are more stringent than the acceptance range imposed on the hydraulic test parameters associated with the Group A and Group B quarterly test. When determining acceptance criteria, design basis calculations and pipe design pressures control the minimum and maximum acceptance criteria. The most limiting of the ISTB or design basis acceptance criteria are then used for testing, as appropriate.

Table 6.4-1 Comprehensive Test Hydraulic Acceptance Criteria Test Parameter Acceptable Alert Range Required Action Range Range Low Hi2h P (Positive displacement pumps) 0.93 to 1.06 Pr 0.90 to< 0.93 Pr < 0.90 Pr > 1.06 Pr LlP (Vertical. line shaft pumps) 0.95 to 1.06 APr 0.93 to< 0.95 Af\ < 0.93 APr > 1.06 APr Q (Positive displacement and 0.95 to 1.06 Qr 0.93 to < 0.95 Qr < 0.93 Qr > 1.06 Qr vertical line shaft pumps)

LlP (Centrifugal pumps) 0.93 to 1.06 APr 0.90 to< 0.93 ~Pr < 0.90 APr > 1.06 APr Q ( centrifugal pumps) 0.94 to 1.06 Qr 0.90 to< 0.94 APr < 0.90 Qr > 1.06 Qr Subscript r denotes reference value.

Page 29 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND 2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 6.4.2 Mandatory Appendix V Periodic Verification Test Division 1 Mandatory Appendix V in the OM Code requires a test for pumps with specific design basis accident flow rates in the Owner's credited safety analysis to verify the pumps can achieve the flow rate. The pumps included in this test category and the credited flow rates are listed in the table below. The conditions for the comprehensive test for each of these pumps currently bound the flow rates listed below. Therefore, performing the comprehensive test fulfills the code requirements for the Periodic Verification Test and no additional testing is required.

Table 6.4-2 Periodic Verification Test Flow Rates Pump Credited Flow Reference Rate (gpm)

Motor AFW Pumps 275 FSAR Ch. 14.1.10 (lP-53, 2P-53)

Turbine AFW Pumps 275 FSAR Ch. 14.1.10 (lP-29, 2P-29)

RHRPumps 1560 2006-0021 Rev. OB pg. 22 (lP-lOA, lP-l0B, 2P-10A, 2P-10B)

Safety Injection Pumps 800 2006-0021 Rev. 0 pg. 41 (1P-15A, 1P-15B, 2P-15A, 2P-15B)

Containment Spray Pumps 1200 2006-0021 Rev. 0 pg. 38 (1P-14A, 1P-14B, 2P-14A, 2P-14B)

Component Cooling Water Pumps 3500 96-0284 Rev. 3 pg. 86 (lP-1 lA, lP-1 lB, 2P-11A, 2P-11B)

Service Water Pumps 4500 I 96-0059 Rev. 12 pg. 20-21 (0P-32A, 0P-32B, 0P-32C, 0P-32D, 0P-32E, 0P-32F)

Note:

(1) The accident analysis for the SW pumps utilizes the degraded pump curve at 4500 gpm at 78 psid TDH as the minimum flow rate for the six pumps.

6.5 Preservice Test Preservice tests are required to be performed during initial start-up testing (ISTB-3100) or when specified by the owner (ISTB-3310). Accordingly, a preservice test will be required for the installation of new systems or at the discretion of the IST Engineer for significant modifications.

During the preservice test period or before implementing inservice testing, an initial set ofreference values shall be established for each pump. These tests shall be conducted Page 30 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT under conditions as near as practicable to those expected during subsequent inservice testing. Except as specified in paragraph ISTB-3310, only one preservice test is required for each pump. A set ofreference values shall be established in accordance with paragraph ISTB-3300 for each pump required to be tested. The parameters to be measured are specified in Table ISTB-3000-1.

6.6 Tests Substitution When a Group A test is required, a comprehensive test may be substituted. When a Group B test is required, a Group A or comprehensive test may be substituted. A Preservice test in accordance with ISTB-3100 may be substituted for any inservice test.

6.7 Allowable Variance from Fixed Reference Points The OM Code states that the IST tests should be performed with the pump operating as close as practical to the reference point. The following allowable variances from the reference point are established in the OM Code. Note that the variance differs based on which parameter is used as the reference value for a particular test.

Table 6.7-1 Allowable Variance from Reference Points Reference Value Parameter Code Pump Type & (centrifugal/vertical) or Q (flow rate) Reference P (positive displacement)

ISTB-5121 Centrifugal Pumps +2%/-1% +1%/-2% ISTB-5122 ISTB-5123 ISTB-5221 Vertical Line Shaft

+2% /-1% +1% /-2% ISTB-5222 Pumps ISTB-5223 ISTB-5321 Positive NIA +1%/-2% ISTB-5322 Displacement Pumps ISTB-5323 Note that the above variances are applicable to Group A, Group B, and comprehensive tests.

6.8 Analysis and Evaluation of Test Results Alert Range (ISTB-6200(a)). If the measured test parameter values fall within the alert range of Table ISTB-5121-1, Table ISTB-5221-1, Table 5321-1 or Table ISTB-5321-2, as applicable, the frequency of testing specified in paragraph ISTB-3400 shall be doubled until the cause of the deviation is determined and the condition corrected, or an analysis of the pump is performed in accordance with paragraph ISTB-6200(c).

Action Range (ISTB-6200(b )). If the measured test parameter values fall within the required action range of Table ISTB-5121-1 , Table ISTB-5221-1, Table ISTB-5321-1 or Page31 of56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND 2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT Table ISTB-5321-1, as applicable, the pump shall be declared inoperable until either the cause of the deviation has been determined and the condition corrected, or an analysis of the pump is performed in accordance with paragraph ISTB-6200(c).

Systematic Error (ISTB-6300). When a test shows measured parameter values that fall outside of the acceptable range of Table ISTB-5121-1, Table ISTB-5221-1, Table ISTB-5321-1 or Table ISTB-5321-1, as applicable, that have resulted from an identified system error, such as improper system lineup or inaccurate instrumentation, the test shall be rerun after correcting the error.

6.9 Analysis (ISTB-6200(c))

In cases where the pump's test parameters are within either the alert or required action ranges of Table ISTB-5121-1, Table ISTB-5221-1, Table ISTB-5321-1 or Table ISTB-5321-2, an analysis may be performed that supports the pump's continued use at the changed values. This analysis shall include verification of the pump's operational readiness. The analysis shall include both a pump level and a system level evaluation of operational readiness, the cause of the change in pump performance, and an evaluation of all trends indicated by available data. The analysis shall also consider whether new reference values should be established and shall justify the adequacy of the new reference values, if applicable. The results of this analysis shall be documented in the record of tests.

6 .10 Replacement, Repair, and Maintenance When a reference value or set of values may have been affected by repair or routine servicing of the pump, a new reference value or set of reference values shall be determined or the previous values reconfirmed by an inservice test and the data analyzed prior to declaring the pump operable (ISTB-3310).

The following are some examples of pump maintenance items that could affect the reference values of the pump:

Pump or motor bearing replacement.

Pump packing or seal adjustments.

Maintenance that requires full or partial disassembly of the pump or motor.

The following are some examples of pump maintenance items that have been determined to not affect pump reference values:

Pump or motor greasing.

Pump or motor oil changes.

Post-maintenance testing shall be performed in accordance with the guidance provided in MA-AA-203-1000, Maintenance Functional Testing.

Page 32 of 56

POINT BEACH NUCLEAR PLANT 1ST Program Document UNITS 1 AND 2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT Pumps in the IST program may be returned to service following maintenance activities with a vibration level between 0.325 ips (Alert Range) but less than or equal to 0.700 ips (Required Action Limit). Prior to returning the pump to service, engineering shall evaluate the condition to ensure acceptability of declaring the pump operable. Any pump returned to service with a vibration level between 0.325 ips but less than or equal to 0.700 ips shall have its testing frequency doubled as required by the Code (Ref. CCE# 2000-004).

6.11 ReliefRequest The Sixth 10-Year 1ST Interval does not have any pump-specific relief requests.

6.12 Pump Test Table The following table defines the pumps included in the PBNP 1ST Program and provides pertinent component and test information. The legend below applies to the PBNP Pump Test Table.

6.12.1 Pump

Description:

The pump name or description.

6.12.2 Pump No.: Unique component tag number.

6.12.3 Applicable Pump Group.

6.12.4 P&ID: Piping and instrumentation drawing on which the pump is depicted.

6.12.5 Coard.: Location coordinates of the pump on the P &ID.

6.12.6 Test Parameters: This column lists the applicable testing parameters that will be measured or observed. The parameters listed are those required by the Code. Any deviations from Code required measurements are described in the corresponding relief request. The following is a description of applicable parameters:

a. N = Pump speed (only required for variable speed pumps).

b. DIP = Pump differential pressure.

c. P = Pump discharge pressure.

d. Q = Pump flow rate.

e. V = Vibration velocity.

6.12.7 Code Class: ASME Code Classification of each pump.

Page 33 of 56

POINT BEACH NUCLEAR PLANT 1ST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 6.12.8 Pump Type and Driver: This column lists type and driver of each pump.

6.12.9 Pump Periodic Verification Test: Indicates whether the pump requires a biennial Pump Periodic Verification Test per ASME OM Code Mandatory Appendix V.

6.12.10 TP/ReliefRequest: This column lists the identifying numbers of any applicable pump technical positions and reliefrequests.

6.12.11 Test Procedure: This column lists the applicable pump IST Procedure.

6.12.12 Remarks: Any additional pertinent information is provided in this space.

Page 34 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 7 .0 VALVE IST PROGRAM 7.1 Valve Selection Criteria and Exemptions 7.1.1 Components within the scope of the valve IST Program are established in ISTA-1100 and ISTC-1100 of the OM Code. Sections ISTA-1100 and ISTC-1100 require preservice and inservice testing and examination of all active and passive valves (and their actuating and position indicating systems) that are required to perform a specific function in shutting down the reactor to a safe shutdown condition, in maintaining the safe shutdown condition or in mitigating the consequences of an accident. ISTA-1100 also specifies that relief devices within the scope of the OM Code are those protecting systems or portion of systems that perform a required function in shutting down the reactor to a safe shutdown condition, in maintaining the safe shutdown condition, or in mitigating the consequences of an accident. Mandatory Appendices I, II, III, and IV of the OM Code provide additional scoping discussion as well as IST requirements for pressure relief devices, check valves, active motor-operated valves (MOVs), and active AOVs, respectively.

7.1.2 Control valves are exempt from the ASME Code testing requirements as allowed by ISTC-1200 of the OM Code. However, per Section 4.2.9 of NUREG-1482, control valves that receive safety system actuation signals and/or have required fail-safe positions are required to meet all test requirements for Category B valves (exercise test, stroke time test, position indication test, and fail-safe test). Exception to this requirement includes control valves that are incapable of non-conservative positioning.

7.1.3 Per the exclusion allowed by ISTC-1200, skid-mounted valves and component subassemblies are excluded from the requirements of Subsection ISTC provided they are tested as part of the major component and are determined to be adequately tested. Further discussion pertaining to skid-mounted components is provided in Sections 3.4 and 4.1.10 ofNUREG-1482.

7.1.4 ASME non-Code Class components may still be subject to periodic testing in accordance with 10CFR50 Appendix A and Appendix B. The bases for exclusion or reason for testing non-Code components are contained in the PBNP IST Program Background Document. Non-Code class components within the scope of the OM Code are in the scope of 10CFR50.55a and are included in the Augmented IST Program.

7.1.5 Dampers are exempt from the OM Code testing requirements. Valves in safety-related ventilation systems (such as control room or primary containment ventilation butterfly valves) are within the scope of the testing requirements of the 10CFR50 IST requirements if they are ASME Code Class 2 or 3.

Page 35 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 7 .1. 6 Valves that are actuated as a result of a safety system automatic response shall be included in the IST Program to the extent that the testing shall verify valve operation required as a result of the safety system input. This applies only if valve movement is required to support those functions required as specified by the Code. This requirement extends only to testing defined by the Code and is not intended to imply the need for verifying a valve's response to automatic logic system output.

7.1.7 Thermal relief valves provide protection from overpressure due to thermal expansion of fluid for components and portions of systems when they are isolated. If a thermal relief valve protects a system or portion of systems that performs a function to achieve/maintain safe shutdown or mitigate the consequences of an accident, then it shall be included within the scope ofIST.

Thermal relief valves are defined in Mandatory Appendix I of the OM Code (I-1200). Paragraph I-1340 requires Class 1 thermal relief devices to be tested per the requirements for Class 1 pressure relief valves (I-1320). Paragraph I-1390 requires Class 2 and 3 thermal relief devices to be tested or replaced on a 10-year frequency. Test percentages and scope expansion for failures do not apply to Class 2 and 3 thermal relief devices.

7 .2 Active/Passive Designation Valves which perform safety functions are defined as being either active or passive.

Valves are stated to have an active safety function if they must actuate or change positions to perform their safety function(s) as defined in ISTA-1100. Generally, a passive safety function designation is allowed only if a valve need not actuate or change positions to perform its safety function(s); however, per NUREG-1482, Section 2.4.2, a valve need not be considered active if it is only temporarily removed from service or from its required safety position for a short period of time while maintaining administrative control over the valve. Leakage rate testing and position indication verification are the only test requirements applicable to passive valves as stated in Table ISTC-3500-1 of the OM Code. Some valves have safety functions in both the open and closed positions and, per NRC guidance, are tested to both positions. Valves with safety functions in both directions may not have to actuate or change positions to perform their required safety function in one of the two directions. For example, a normally open power operated valve in an ECCS injection line which also functions as a containment isolation valve (CIV) may not need to change position to perform its open safety function for emergency core cooling, but must be capable of closure to assure containment integrity. For cases such as these, the applicable valves are identified as having an active safety function in one direction and a passive safety function in the other direction (Ref.

OM Code Interpretation 99-03). Valves identified as performing a passive safety function shall not be exercised to their passive position as an ASME related test requirement. This has ASME concurrence via OM Code Interpretation 99-07.

7.3 Preservice Testing Page 36 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 7.3.1 The preservice test is only required once and then only before implementing inservice testing. It is typically performed after a new valve is added to the plant or following maintenance that could affect valve performance.

7.3.2 Each valve shall be tested during the preservice test period under conditions as near as practicable to those expected during subsequent inservice testing.

7.3.3 Any valve that has undergone maintenance that could affect its performance after the preservice test shall be tested in accordance with paragraphs ISTC-3310, III-3400, and N-3520.

7. 3 .4 Preservice testing of certain pressure relief devices ( safety and relief valves and nonreclosing pressure relief devices), active MOVs, and active AOVs, shall meet the preservice test requirements of Mandatory Appendices I, III, and N of this Division, respectively. Testing that meets the requirements of the applicable Mandatory Appendix but was performed prior to implementation of the Mandatory Appendix may be used.

Refer to CMP 2.3, "Relief/ Safety Valves," for additional guidance on testing before installation for Mandatory Appendix I pressure relief devices [Ref. I-3100].

Refer to CMP 2.2, "Motor Operated Valves," for additional guidance on Mandatory Appendix III MOV preservice testing [Ref. III-3200].

Refer to CMP 2.5, "AOV Program Document," for additional guidance on Mandatory Appendix N AOV preservice performance assessment testing

[Ref. N -3 300(a)]. Preservice stroke testing, fail-safe testing, leak testing, and remote position verification testing (if applicable) for Mandatory Appendix N AOV s are performed under the IST Program [Ref. N-3300(b), N-3300(c), N-3300(d), N-3300(e)].

7.4 Pressure Isolation Valves (PNs) 7.4.1 All facility pressure isolation valves are identified in this document. Pressure isolation valves (PNs) are defined as two normally closed valves in series at the reactor coolant system pressure boundary that isolate the reactor coolant system from an attached low pressure system. Event V check valves are a special sub-set of PNs, defined as two series check valves which perform PN functions and are located in piping which penetrates containment. The NRC staff guidance in NUREG-1482, Section 4.4.4 states that PN testing should be conducted in accordance with Technical Specifications and any additional commitments made in response to Generic Letter 87-06. Per NUREG-1482, any PIVs not listed in Technical Specifications should at least be tested to verify closure capability. PBNP Technical Specifications contain leak rate testing requirements for PN s; however, only Event V PN s are identified by Page 37 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT reference in Technical Specifications 5.5.16. Event V PIVs applicable to the requirements of Technical Specification 5.5.16 are monitored to ensured leakage does not exceed the limits defined in Technical Specification 5.5.16.

Those Event V PIVs applicable to the requirements of Technical Specification 5.5.16 include the following: 1(2)SI-853A, 1(2)SI-853B, 1(2)SI-853C, 1(2)SI-853D, 1(2)SI-867A, 1(2)SI-867B, 1(2)SI-845A, 1(2)SI-845B, 1(2)SI-845C, 1(2)SI-845D, 1(2)SI-845E, and 1(2)SI-845F.

7.4.2 The PBNP response to Generic Letter 87-06 identified the following nine valves in each unit which meet the definition for PIVs but were neither Event V valves nor leak rate tested per ASME Code requirements:

1(2)SI-842A, -842B, 1(2)RH-700, -701,-720, 1(2)RC-503, -541, -598, and -599. The RCS Loop Drain Valves 1(2)RC-503, -541, -598, and-599 are no longer installed. The RHR Suction Valves from the RCS Hot Leg, 1(2)RH-700 and -701, are tested consistent with the Generic Letter 87-06 commitments which is to monitor RCS boundary leakage and verify seat tightness by monitoring downstream RHR system pressure during startup. A system modification would be required to leak rate test these valves per Code requirements. Although no commitment was made in the Generic Letter 87-06 response to test the RHR Cold Leg MOV Isolation Valves, 1(2)RH-720, per ASME Code requirements, Code leak rate testing is being performed on these valves. The SI Accumulator Discharge Check Valves, 1(2)SI-842A&B, are leak rate tested by monitoring accumulator level during quarterly SI pump testing. This is consistent with the GL 87-06 commitment.

7.5 Containment Isolation Valves (CIVs)

Containment isolation valves are seat leakage tested per the PBNP Containment Leak Rate Testing (CLRT) Program as required by 10CFR50, Appendix J, Option B. All valves included in the PBNP CLRT Program shall be included in the IST Program as Category A valves (see Section 7.8.7). However, as allowed by ASME OM Code, Paragraph ISTC-3620, the OM Code acceptance criteria and corrective action requirements do not apply to CIVs which perform no leakage important safety function other than for containment isolation. For these valves, the corrective action requirements of Appendix J and the CLRT Program shall be applied.

7.5.1 Active valves which are designated as primary containment isolation valves (CIVs) or boundary valves for primary containment closed systems perform a safety-related function to close. Some CIVs and closed system boundary valves may be exempted from seat leakage testing requirements as allowed by the 10CFR50 Appendix J Program. However, all active CIVs and closed system boundary valves shall be included in the IST Program and shall be exercised in accordance with OM Code requirements, regardless of Appendix J exemptions.

7.5.2 Reactor coolant system (RCS) pressure boundary valves which are normally Page 38 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT open or routinely opened have a safety-related function to close to maintain the integrity of the reactor coolant pressure boundary. Exception to this position is allowed provided that in the event of the postulated failure of downstream components during normal operation in conjunction with failure of the RCS pressure boundary valve to close, the reactor can be shut down and cooled down in an orderly manner assuming makeup is provided by the reactor coolant makeup system (CVCS). This exception is consistent with the requirements of 10CFR50.55a(c)(2) and the guidance of ANS 51.1-1983 (formerly ANSI Nl 8.2).

7.5.3 Active ASME Code Class to non-ASME Code Class pressure boundary isolation valves generally have a safety-related function to close to maintain the integrity of the safety system pressure boundary. Exclusion of these valves from closure testing requirements is acceptable provided that failure of the pressure boundary isolation valve to close, combined with failure of downstream non-ASME Code Class components, would not impact safety system operation, including the potential effect on operability of safety-related components due to environmental concerns such as flooding or release of steam.

7.6 Valve Categories 7.6.1 All valves shall be designated as Category A, Category B, Category C, Category D, (see definitions in Section 4.43) or a combination thereof (e.g. - check valves with a leakage important safety function would be classified as Category A and C (A/C) valves).

7.6.2 Unit/train cross-connect isolation valves in closed systems outside containment which would NOT result in leakage directly to the environn1ent as a result of post-LOCA seat leakage shall be classified Category B passive valves.

7.6.3 Unit/train cross-connect isolation valves in closed systems outside containment which WOULD result in leakage directly to the environment as a result of post-LOCA seat leakage would in most cases be considered Category A valves unless off-site dose, as a result of post-LOCA seat leakage, is well within the limitations of 10CFRlO0.

7.6.4 Valves serving as cross-tie valves between redundant trains and unit cross-ties whose function are to maintain separation and to limit leakage based on total system requirements shall be classified Category B passive valves.

7.7 Valve Inservice Testing Frequency Page 39 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 7.7.1 The valve IST frequency will be as set forth in paragraphs ISTC-3510, ISTC-3630(a), ISTC-5230, ISTC-5240, ISTC-5250, I-1300, III-3310, III-3610, IV-3410, IV-3420, IV-3430 of the ASME OM Code and Mandatory Appendices. The valve IST frequency for Category A, Category Band Category C valves will be nominally every 3 months (ISTC-3510), except as provided by paras. ISTC-3520, ISTC-3540, ISTC-3550, ISTC-3570, ISTC-5221 and ISTC-5222.

Where Code required quarterly valve tests are impractical or otherwise undesirable, testing may be deferred to cold shutdown periods as permitted by paragraphs ISTC-3521(b) and 3521(c) for Category A and Category B valves, and ISTC-3522(b) for Category C Check valves, paragraph III-3721 for active MOVs, and paragraphs IV-3420(a) and IV-3430 for active AOVs.

If it is determined that testing be completed on a refueling outage frequency, testing may be deferred to refueling outages as permitted by paragraphs ISTC-3520(d) and 3520(e) for Category A and Category B valves, ISTC-3522(c) for Category C Check valves, paragraph III-3610 for active MOVs, and paragraphs IV-3420(a) and IV-3430 for active AOVs.

Valve testing which is performed during cold shutdowns and refueling outages shall be conducted in accordance with the requirements of paragraphs ISTC-3521, ISTC-3522, III-3610, IV-3420(a), and IV-3430 in their entirety along with the guidance provided in NUREG-1482, Sections 2.4.5 and 3.1.1.

Justifications for deferral of testing to cold shutdowns and refueling outages are provided in Appendices C and D of this document.

7.7.2 Valve testing shall be performed as stipulated in OM Code paragraphs ISTC-3500, unless the valve is in a system which is inoperable or not required to be operable. If the quarterly testing frequency is not followed, valve testing shall be performed within the 3 months before the system is returned to operable status as required by paragraphs ISTC-3570.

Page 40 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 7.7.3 Active MOVs (Mandatory Appendix III) 7.7.3.1 Active MOVs shall be full cycle exercised at least once per refueling cycle with the maximum time between exercises to be not greater than 24 months as set forth in paragraph III-3600. If full stroke exercising of an MOV is not practical during plant operation or cold shutdown outages, full stroke exercising shall be performed during the plant's refueling outage. Active MOVs which are categorized as HSSC that can be operated during plant operation shall be exercised quarterly, unless the potential increase in core damage frequency (CDF) and large early release frequency (LERF) associated with a longer exercise interval is small. Justifications for deferral of testing of active HSSC MOVs to cold shutdowns and refueling outages are provided in Appendices C and D of this document as needed.

7.7.3.2 The active MOV 1ST frequency will be as set forth in paragraphs III-3310, III-3 722, and III-6440. Valve testing includes a mix of static and dynamic MOV performance testing at a frequency such that the MOV functional margin does not decrease below the acceptance criteria. Refer to CMP 2.2 for additional detail.

7. 7.4 Active AOVs (Mandatory Appendix IV) 7.7.4.1 Periodic performance assessment inservice testing shall be performed as set forth in paragraph IV-3410 for HSSC AOVs per paragraph IV-3821. Periodic performance assessment testing is not required for LSSC AOVs per paragraph IV-3822. Refer to CMP 2.5, "AOV Program Document," for detailed performance assessment test guidance for AOVs within the scope of Mandatory Appendix IV.

The interval for periodic performance assessment tests shall be determined in accordance with paragraph IV-6400. If insufficient test data exists to determine the performance assessment test interval, then performance assessment testing shall be conducted every three refueling cycles or 6 years (whichever is longer) until sufficient data exist to justify a longer 1ST interval.

The periodic performance assessment test interval shall not exceed 10 years for each AOV.

7.7.4.2 A stroke test shall be performed quarterly as set forth in paragraph IV-3420 during operation at power to the position(s) required to fulfill its function(s) per ISTA-1100, if practicable.

7.7.4.3 Active AOVs with fail-safe actuators shall have a fail-safe test performed as set forth in paragraph IV-3430 in accordance with the frequency of stroke time testing described in Section 7.8.6.

Page41 of56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT

7. 7.4.4 If applicable, active Category A AOV s shall be leak tested as set forth in Section 7.8.7.

Leak test for containment isolation AOV s shall be conducted at the frequency specified in the Appendix J program.

Leak test for Category A AOV s with a leakage rate requirement based on functions other than Appendix J program shall be conducted at least once every 2 years.

Leak test for active pneumatically-operated PORVs shall be conducted at the frequency specified by Mandatory Appendix I.

7.7.4.5 Active AOVs with remote position indicators require remote position verification at least once every two years as set forth in Section 7.8.1. [Ref ISTC-3700, 10CFR50.55a (b)(3)(xi)]

7.8 Valve Test Requirements Active and passive valves in the categories defined in paragraph ISTC-1300 shall be tested in accordance with the paragraphs specified in Table ISTC-3500-1. Refer to ER-AA-113-1000 Section 4.1.8 for Fleet technical position on position indication testing.

7.8.1 Valve Position Verification 7.8.2.l Valves with remote position indicators shall be observed locally at least once every two years to verify that valve operation is accurately indicated. The Code does not; however, provide any guidance on what constitutes an acceptable demonstration of position indication verification. Consequently, good operational judgment must be exercised by personnel conducting the position indication verification as to whether or not a given valve's remote position indication is accurately reflecting the actual valve position.

The key component of a position indication verification test is that:

1) when valve travel is complete and a valve is fully open the position indication lights indicate that the valve is open, and 2) that when valve travel is complete and the valve is fully shut the position indication lights indicate that the valve is shut. The following examples are given as general guidelines:

For Motor Operated Valves (MOVs) with simple remote indicators (e.g., position indication lights) if in the best judgment of the personnel conducting the test the position indication lights show a change in valve position when the valve is greater than 15 percent from its open or shut position a maintenance work order should be initiated to correct the deviation.

Page 42 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT

For Air Operated Valves (AOVs) with simple remote indicators (e.g., position indication lights) if in the best judgment of the personnel conducting the test the position indication lights show a change in valve position when the valve is greater than 33 percent from its open or shut position a maintenance work order should be initiated to correct the deviation. Note that it is expected that these deviations will be larger on valves that have small stroke lengths because of the difficulty in setting the position indication switches.

For manually operated valves with a local position indicator (scale/pointer) ifin the best judgment of the personnel conducting the test the local position indicator does not indicate within 15 percent of the actual position of the valve when it is fully open or shut a maintenance work order should be initiated to correct the deviation.

For MOVs and AOVs with a local position indicator (scale/pointer) if in the best judgment of the personnel conducting the test the local position indicator does not indicate within 15 percent of the actual position of the valve when it is fully open or shut a maintenance work order should be initiated to correct the deviation.

The above guidance is intended to provide Operations with a consistent approach to evaluating if the position indication systems associated with valves included within the IST Program could benefit from adjustment. It is not intended to prevent plant personnel from issuing Work Orders or Action Requests if they feel that a valve's position indication system is not working up to their expectations or needs.

Personnel should note that a valve itself is not necessarily inoperable because its position indication requires correction. An example of this would be a case where the position indication lights indicate that a Motor Operated Valve is open when its open stroke is complete; however, the open light changed status when the valve was more than 15 percent from its full open position. As long as the position indication lights indicate the correct valve position at the completion of the valve stroke the position indication system does not affect the operability of the valve.

Position verification for active MOVs and active AOVs shall be verified locally during inservice testing or maintenance activities.

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POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 7.8.3.2 Local observation shall be supplemented by other indications such as the use of flowmeters or other suitable instrumentation to verify obturator position as required by 10CFR50.55a (b )(3)(xi). These observations need not be concurrent. Normal system operation is a typical indication of obturator position.

7.8.4 Active Power Operated Valve Stroke Time Testing Active power operated valve stroke time testing shall be performed in accordance with the following requirements of the ASME OM Code in its entirety:

Paragraph ISTC-5113 (PORVs),

Paragraph ISTC-5120 & Mandatory Appendix III (MOVs),

Paragraph ISTC-5130 & Mandatory Appendix IV (AOVs),

Paragraph ISTC-5140 (HOVs), and

Paragraph ISTC-5150 (SOVs).

Stroke time reference values will be determined in accordance with ISTC-3300, "Reference Values", from the results of tests performed under conditions as near as practicable to those expected during subsequent inservice testing. Reference values shall only be established for a valve when it is known to be operating acceptably.

When performing the full cycle exercise of active MOVs, the stroke time of MOVs specified in plant technical specifications shall be verified as required by 10CFR50.55a (b)(3)(ii)(D). Stroke times of active MOVs not specified in the plant technical specifications do not require verification.

Active AOVs shall have their stroke time measured during stroke testing in accordance with the requirements of paragraphs IV-3400(b), IV-3510, and IV-3420(b), IV-3530, and IV-7100. Instrumentation and test equipment accuracy shall be considered in accordance with section ISTA-4000. [Ref. IV-7100].

For new or replaced active AOVs, stroke time reference values shall be determined from the results of preservice testing; for existing AOVs stroke may be determined from the results of inservice testing (Ref. IV-3510).

Guidelines for performing Code required stroke time testing are provided in Table 7.8-1, including owner specified Limiting Value of Full Stroke Time (LVFST).

IfLVFST are established from criteria other than that specified in paragraphs ISTC-5142 (HOVs), ISTC-5152 (SOVs), or Mandatory Appendix IV (AOVs) it shall be determined in accordance with the guidance of Generic Letter 89-04, Attachment 1, Position 5. Acceptance criteria and LVFST are calculated based on the reference stroke times as shown in Table 7. 8-1. However, if a Page 44 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT design, Technical Specification, FSAR, or accident analysis limit exists which is more limiting, then it shall be used as the LVFST in lieu of the calculated value. Any valve whose reference value is less than or equal to two seconds may be (but is not required to be) designated a "rapid-acting valve" in accordance with the aforementioned paragraphs for "Stroke Time Acceptance Criteria". In such cases the maximum limiting stroke time shall be 2 sec.

Table 7.8-1 Reference Actuator Value (RV) Acceptance Criteria Type (sec) (AC) LVFST NIA NIA NOTE2 Motor NIA NIA NOTE2

S 1.5RV Other RV>lO 0.75RV - l.25RV [NOTE (including 1]

Pneumatic,

S2.0RV Solenoid, 2<RV:Sl0 0.50RV - l.50RV [NOTE Hydraulic) 1]

Any, RV:S2 :S 2 :S 2 Rapid-Acting NOTE 1: Owner specified LVST determined in accordance with the guidance of Generic Letter 89-04, Attachment 1, Position 5 as permitted by subparagraphs ISTC-5113(b),

ISTC-5141(b), and IV-3420(b)(l).

NOTE 2: For MOVs with design basis stroke times defined in Technical Specifications, this stroke time is the LVFST.

For valves with reference valve and following valve replacement, repair or other maintenance which could affect the valve's stroke time, new reference value(s) will be determined or the previous value( s) reconfirmed prior to returning the valve to service as required by ISTC-3310.

Stroke times may be impacted by changes in operating conditions (plant operation versus shutdown periods), seasonal conditions (summer versus winter), or system lineups. Therefore, it may be necessary or desirable to establish additional reference values. Additional reference values shall be established in accordance with ISTC-3320, "Establishment of Additional Set of Reference Values". Whenever additional reference values are established, the reasons for doing so shall be justified and documented in the record of tests.

Page 45 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT Corrective actions for valve stroke times which exceed the acceptance criteria specified in paragraphs ISTC-5142 (HOVs), ISTC-5152 (SOVs), Mandatory Appendix IV (AOVs), or design basis stroke times for MOVs in Technical Specifications (MOVs) or limiting values of full-stroke time specified in paragraphs ISTC-5141(b) (HOVs), ISTC-5151(b) (SOVs), Mandatory Appendix IV (AOVs), or plant Technical Specifications (MOVs) shall be in accordance with paragraphs ISTC-5143 (HOVs), ISTC-5153 (SOVs), or Mandatory Appendix IV (AOVs). Failure of a limiting value of full stroke time shall result in immediately declaring the valve inoperable. Exceeding acceptance criteria specified in the aforementioned paragraphs for stroke time acceptance criteria can result in a retest or declaring the valve inoperable. In any case, an evaluation may be performed in accordance with paragraphs ISTC-5143(c) (HOVs), ISTC-5153(c) (SOVs), or Mandatory Appendix IV (AOVs) to restore the valve to an operable status, or the valve repaired, or replaced.

Before returning a repaired or replacement valve to service, a test demonstrating satisfactory operation shall be performed per paragraphs ISTC-5143(e) (HOVs), ISTC-5153(e) (SOVs), or Mandatory Appendix IV (AOVs).

7.8.5 Active MOVs (10CFR50.55a (b)(3)(ii) and Mandatory Appendix III)

In addition to periodic exercising and stroke time testing of certain active motor-operated valves, 10CFR50.55a (b)(3)(ii) requires licensees to establish a program to ensure that motor-operated valves are capable of performing their design basis safety functions. As described in NRC GL 89-10, "Safety Related Motor-Operated Valve Testing and Surveillance," ASME OM Code testing alone is not sufficient to provide assurance ofMOV operability under design basis conditions.

Subsequent to GL 89-10, the NRC issued GL 96-05, "Periodic Verification of Design Basis Capability of Safety Related Motor-Operated Valves." This Generic Letter required licensees to establish a program to periodically confirm the continued capability of motor-operated valves to perform their safety functions, or validate the effectiveness of the current GL 89-10 Program in identifying degradation mechanisms ( such as those caused by age).

In response to the referenced Generic Letters, PBNP has established an MOV Program. This program utilizes periodic diagnostic testing, and preventative maintenance to ensure the continued operability of safety-related and augmented quality motor-operated valves. The frequency of diagnostic testing and preventative maintenance is dependent on valve classification (SR or AQ), service conditions and monitored test parameter margins.

Page 46 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT Subsequent to the establishment of the PBNP MOV Program, the Sixth Interval update of the IST Program introduced Mandatory Appendix III, "Preservice and Inservice Testing of Active Electric Motor-Operated Valve Assemblies in Water-Cooled Reactor Nuclear Power Plants". Mandatory Appendix III expands on the requirements for active MOV s established in response to GL 89-10 and 96-05. The testing requirements include a full cycle exercise and a mix of static and dynamic MOV performance testing.

Mandatory Appendix III requires a one-time design basis verification test to verify the capability of each MOV to meet its safety-related design basis requirements. The design basis verification test shall be repeated if an MOV application is changed, the MOV is physically modified, or the system is modified in a manner that invalidates its current design basis verification test results or data. Each MOV shall also have a preservice test.

As discussed in NUREG-1482 Section 4.2.13, for MOVs within the scope of the GL 89-10 and GL 96-05 programs, the testing and/or engineering analysis performed to close out those generic letters may be used to meet the design-basis verification requirement in Mandatory Appendix III. For MOV s that were outside of the scope of GL 89-10 and GL 96-05 but within the IST program, a one-time test and/or engineering analysis is required by Mandatory Appendix III.

The program also establishes requirements for the analysis and trending of the diagnostic testing and corrective action as a result of unacceptable MOV performance.

Conditions placed on Mandatory Appendix III are defined in 10CFR50.55a (b )(3)(ii)(A) through (D).

The PBNP MOV Program is subject to internal and external agency audit and inspection, and satisfies 10CFR50.55a (b)(3)(ii) requirements for ensuring the continued operability of motor-operated valves.

7.8.6 Active AOV Testing (Mandatory Appendix IV)

Mandatory Appendix IV augments the rules of Subsection ISTC to establish the requirements for preservice and inservice testing to assess the operational readiness of certain active AOVs.

The risk-informed approach for AOV inservice testing is invoked in accordance with Paragraph IV-3800 to reduce testing burden while ensuring operational readiness of active AOVs. Using the risk-informed methodology, active AOVs are evaluated and categorized into either the HSSC or LSSC Page 47 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT category. Test requirements for HSSC & LSSC categories are specified in paragraphs IV-3821 and IV-3822, respectively. Refer to PBNP Procedures CMP 2.5 and CMP 2.5.1 for additional details on risk-informed considerations and risk ranking methodology.

Grouping AOVs as described in paragraph IV-3600 is not used for periodic inservice testing at PBNP.

Requirements of Mandatory Appendix IV are summarized in the following sections.

7.8.6.1 Performance assessment testing shall be performed for HSSC AOVs and is not required for LSSC AOVs based on the Mandatory Appendix IV risk-informed considerations [Ref. IV-3821, IV-3822]. Refer to CMP 2.5, "AOV Program Document," for detailed guidance on performance assessment testing HSSC AOVs.

7.8.6.2 All active AOVs shall have inservice stroke tests performed as described in Section 7.8.2.

7.8.6.3 If applicable, active AOVs with fail-safe actuators shall have preservice and inservice fail-safe test performed as described in Section 7.8.6.

7.8.6.4 If applicable, active Category A AOVs shall have preservice and inservice leak testing performed per Section 7.8.7.

7.8.6.5 Active AOVs with remote position indicators shall have preservice and inservice remote position verification performed per Section 7.8.1. [Ref. ISTC-3700, 10CFR50.55a (b)(3)(xi)]

7.8.7 Category A Valve Seat Leakage Testing

a. Category A valves, which are containment isolation valves, shall be tested in accordance with 10CFR50, Appendix J. Additional guidance is provided in the PBNP CLRT Program.

b. Containment isolation valves that also provide a reactor coolant system pressure isolation function shall additionally be tested in accordance with paragraph ISTC-3630. Additional guidelines are provided in the PBNP CLRT Program.

c. Category A valves, which perform a function other than containment isolation, shall be seat leakage tested to verify their leak tight integrity in accordance with the requirements of paragraph ISTC-3630.

Page 48 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 7.8.8 Fail-Safe Testing Most solenoid operated valves and AOV s fail to the fully open position, fail to the fully closed position, or remain in the last position upon a loss of actuating energy supply due to the design of the actuators. However, only valves that have a fail-safe safety function shall be tested in accordance with paragraph ISTC-3560. Fail-safe testing of active AOVs with fail-safe actuators is performed in accordance with paragraph IV-3430.

Generally, doing a stroke test of a fail-safe valve, using the control switch is the functional equivalent to a loss of actuator power. Therefore, the Exercise Test is equivalent to the Fail Safe Test. Because of this functional equivalency, fail-safe tests are not typically identified in IT procedures. In cases where a specific test of fail-safe capabilities is performed, it is identified.

Refer to ER-AA-113-1000 Section 4.11.4 for Fleet technical position on fail-safe testing.

7.8.9 Check Valve Testing

a. Category C check valve exercise testing shall be performed in accordance with the requirements of paragraphs ISTC-3510, ISTC-3522, ISTC-3550 and ISTC-5221 of the OM Code.

b. As an alternative to the requirements of paragraphs ISTC-3510, ISTC-3520, ISTC-3530, ISTC-3550 and ISTC-5221 of the OM Code, PBNP has established a check valve Condition monitoring (CVCM) program in accordance with Mandatory Appendix II. Specific CVCM plans are included in the IST Program. Refer to Section 7.8.8 for additional details.

c. For series check valve configurations without provisions to verify individual reverse flow closure and the plant safety analysis assumes closure of either valve (but not both), the valve pair may be operationally tested closed as a unit. If the plant safety analysis assumes that a specific valve or both valves of the pair close to perform the safety function(s), the series valves shall be tested to demonstrate individual valve closure.

d. When exercise tests cannot practically be performed or there are no means of verifying a full-stroke open or closed, check valves may be disassembled and inspected as allowed by ASME OM Code, ISTC-5221 (c).

e. Radiography or other non-intrusive techniques may provide a means of verifying check valve full-stroke capability.

Page 49 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT

f. In accordance with the ASME OM Code check valves are subject to bi-directional testing. ISTC-3522(a) provides guidance regarding when it is impractical to conduct open and closed exercise tests during the same interval. For example, if a quarterly open exercise is performed and the closed exercise can only be performed during cold shutdowns, bi-directional exercising requirements are considered satisfied when the closed exercise is complete.

7.8.10 Check Valve Condition Monitoring (CVCM) Program Implementation (Mandatory Appendix II)

The purpose of the CVCM Program is to both (a) improve valve performance; and (b) optimize testing, examination, and preventive maintenance activities in order to maintain the continued acceptable performance of a select group of check valves. PBNP may implement this program on a valve or a group of similar valves.

Examples of candidates for (a) improved valve performance are check valves that:

have an unusually high failure rate during inservice testing or operations;

cannot be exercised under normal operating conditions or during shutdown;

exhibit unusual, abnormal, or unexpected behavior during exercising or operation; and/or

the Owner elects to monitor for improved valve performance.

Examples of candidates for (b) optimization of testing, examination, and preventive maintenance activities are check valves with documented acceptable performance that:

have had their performance improved under the Condition Monitoring Program.

cannot be exercised or are not readily exercised during normal operating conditions or during shutdowns.

can only be disassembled and examined.

the Owner elects to optimize all the associated activities of the valve or valve group in a consolidated program.

The CVCM Program shall be implemented in accordance with Mandatory Appendix II of OM Code, 2017 Edition.

Valve opening and closing must be demonstrated when flow testing or Page 50 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND 2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT examination methods (nonintrusive, or disassembly and inspection) are used.

If the Mandatory Appendix II CVCM Program for a valve or valve group is discontinued, then the requirements of Subsection ISTC shall apply.

Refer to the PBNP CVCM Program Document for additional details.

7.8.11 Manual Valve Exercising Manual valves within the IST program scope that perform an active safety function shall be exercised through a complete cycle at least once every 2 years as required by ISTC-3540. Exercise testing shall be considered acceptable if valve stem travel exhibits unrestricted movement with no abnormal resistance or binding through one complete cycle. The use of a valve persuader (cheater) for additional mechanical advantage will not invalidate the test, as it is recognized that larger valves may exhibit increased packing friction and/or increased friction associated with the disk to seat interface. In addition, a valve persuader may be used for personnel safety depending on a valve's service application (i.e. main steam). Stem length measurement may be used to supplement this testing method. Where practical, process parameters may be utilized to verify obturator movement.

However, where process parameters are utilized to verify obturator movement it is not necessary to be performed simultaneous to manual exercising. l[Ac11 7.8.12 Safety and Relief Valve Testing

a. Category C safety and relief valves shall meet the inservice test requirements of Mandatory Appendix I of the OM Code.

b. Active pneumatically operated PORVs shall meet the inservice test requirements of Mandatory Appendix IV of the OM Code.

c. Refer to ER-AA-113-1000 Section 4.11.3 for the Fleet technical position on additional relief valve testing.

7.8.13 Vacuum Breaker Testing Category C vacuum breakers shall meet the inservice test requirements of Mandatory Appendix I of the OM Code.

7.8.14 Rupture Disks Testing Category D rupture disks shall meet the inservice test requirements for non-reclosing pressure devices of Mandatory Appendix I of the OM Code.

Page 51 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND 2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT 7.8.15 Effects of Valve Repair, Replacement, or Maintenance After valve or its control system has been replaced, repaired, or has undergone maintenance that could affect the valve's performance, a new reference value (or inservice test values for active MOVs with design basis stroke times defined in Technical Specifications and active AOVs) shall be determined or the previous value reconfirmed by an inservice test run before the time it is returned to service or immediately if not removed from service.

The following are some examples of valve maintenance items that could affect the performance of the valve:

Valve packing replacement or adjustment.

Disconnection of pneumatic, hydraulic, or electrical connections to the valve operator.

Maintenance that requires full or partial disassembly of the valve or its operator.

Adjustment of the valves local and remote position indicators .

Adjustments to a valve operator's control system.

Valve back seating that may affect its performance (e.g., cause damage to the valve or bind it into its backseat).

Engineering Evaluation #2001-0019 has been prepared to address on-line packing re-consolidation ofMOVs and Post-Maintenance Testing (PMT) requirements. The Evaluation specifies the controls necessary to ensure MOV performance attributes are not altered as the result of on-line maintenance and plant conditions do not support performances of stroke time testing and/or seat leakage testing.

NOTE: It is acceptable for the plant to perform an engineering evaluation of the impact of adjusting valve stem packing or valve back seating to demonstrate that the valve's performance parameters are within acceptable limits if PMT cannot be performed under the current plant conditions. The plant must perform the PMT when the plant enters an operating Mode in which the testing is practical (Re. NUREG-1482, Rev. 3, Section 4.4.2).

If the MOV or AOV was not removed from service, inservice test values shall be immediately determined or confirmed.

As an alternative for AOVs, the activities performed may be evaluated along with the results of post-replacement/repair/modification/maintenance testing to determine if new inservice test values are warranted before the AOV is returned to service (see CMP 2.5 for additional details). See NextEra Energy Fleet Procedure MA-AA-203-1000, "Maintenance Testing," for guidance on post maintenance test activities required based on the type of work activity.

If maintenance activities are scheduled concurrent with an active MOVs or an Page 52 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT active AOVs inservice test, then the test shall be conducted in the as-found condition prior to the maintenance activity.

Safety and relief valves and nomeclosing pressure relief devices, active MOVs, and active AOVs shall be tested as required by the replacement, repair, and maintenance requirements of Mandatory Appendices I, III, and IV, respectively.

7.9 Relief Requests The Sixth 10-Year IST Interval does not have any valve-specific relief requests.

7.10 Valve Test Table The following table defines the valves included in the PBNP IST Program and provides pertinent component and test information. The legend below applies to the PBNP Valve Test Table.

7.10.1 Valve

Description:

The valve name or description.

7.10.2 Valve No.: Unique component tag number.

7.10.3 P&ID: Piping and instrumentation drawing on which the pump/valve is depicted.

7.10.4 Coord.: Location coordinates of the pump/valve on the P&ID.

7.10.5 Code Class, Cat.: ASME Code Classification of each pump/valve and the Code Valve Category.

7.10.6 Positions: The normal valve position, its safety position and its fail-safe position (if applicable). A valve that has a safety function in both the open and closed positions and is maintained in one of these positions, but is only required to move from the initial position to the other and is not required to return to the initial position is not required to be stroke timed in both directions (Ref. ASME OM Code Interpretation 01-02, dated February 1, 2000).

7.10.7 Active-Passive: Defines whether the valve performs active (A) or passive (P) safety functions(s), or no safety function(s) (N) in the open and closed positions (safety functions as defined in ISTA-1100).

7.10.8 Req. Test/Freq: The Code required tests for each valve and the frequency at which these tests are performed.

7.10.9 TP/TJ/CSJ/ROJ/RR: Listing for each valve of applicable technical positions, technical justifications, cold shutdown justifications, refueling outage Page 53 of 56

POINT BEACH NUCLEAR PLANT 1ST Program Document UNITS 1 AND 2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT justifications, and/or relief requests.

7.10.10 Test Procedure: This column lists the applicable valve test procedure.

7.10.11 Valve Type: List the valve type (i.e., gate, globe, check, relief).

7.10.12 Valve Size: Specify the valve size in inches.

7.10.13 Actuator Type: List the type of valve actuator (i.e., motor, solenoid, pneumatic, hydraulic, self).

Page 54 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND 2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT VALVE TABLE CODES VALVE TYPE ACTUATOR TYPE VALVE POSITIONS AP Angle Globe AO Pneumatic (Air) Operator 0 Open BTF Butterfly HO Hydraulic Operator C Closed CK Check SO Solenoid Operator LO Locked Open DI Diaphragm MA Manual Operator LC Locked Closed GA Gate MO Motor Operator OC Open or Closed GL Globe SA Self Actuated PO Partial Open PCV Pressure Control TH Throttled RD Rupture Disk VALVE CATEGORY IP Intermediate Position REG Regulator A Valve with specific SCK Stop Check leakage criteria SRV Safety/Relief B Valve with no specific TEST FREQUENCY VB Vacuum Breaker leakage criteria C Self-Actuating Q Quarterly CS Cold Shutdown CU Continuous Use R Refueling R* Once per refueling cycle, not to exceed 24 months between exercises 2Y Two Years SY Five Years lOY Ten Years CM Per Condition Monitoring MOV Per Mandatory Appendix III AOV Per Mandatory Appendix IV Page 55 of 56

POINT BEACH NUCLEAR PLANT IST Program Document UNITS 1 AND2 Revision 11 INSERVICE TESTING PROGRAM DOCUMENT TEST REQUIREMENTS*

INSP Check valve disassembly and inspection.

ST Power operated valve stroke time test.

CV Check valve exercise test.

ET Power operated valve exercise test.

FSM Manual valve full-stroke exercise.

FST Fail-safe test.

PIT Remote position indication verification.

IST-001 Check valve closure verification by non-intrusive testing such as radiography or ultrasonic testing.

IST-002 Flow test to verify valve is open.

III Mandatory Appendix III MOV test.

IV Mandatory Appendix IV AOV performance assessment test.

NIT Non-intrusive testing.

RVT Safety and relief valve tests.

RVT-T Thermal Relief Valve Test SP Regulator Setpoint Test SLT-1 10CFR50, Appendix J, Type C, valve seat leakage test.

SLT-2 PIV seat leakage test.

SLT-3 Pressure decay seat leakage test.

SLT-4 Leakage test of SI accumulator check valves by monitoring for accumulator level changes during SI pump testing.

SLT-5 Not Used SLT-6 Seat leakage test to identify gross leakage. Specific leakage rate will be measured and evaluated.

An "A" preceding the test requirement signifies a component requiring an Augmented test(s).

Page 56 of 56

POINT BEACH NUCLEAR PLANT APPENDIXB UNITS 1 AND 2 Revision 3 INSERVICE TESTING PROGRAM VALVE RELIEF REQUESTS APPENDIXB VALVE RELIEF REQUESTS There are NO valve relief requests submitted for the Sixth 10-Year Interval.

Page 1 of 1

POINT BEACH NUCLEAR PLANT 1ST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS APPENDIXC COLD SHUTDOWN TEST WSTIFICATIONS CSJ-01 Not Used CSJ-02 Not Used CSJ-03 Not Used CSJ-04 1/2CV-313A CSJ-05 1/2CV-371, -371A CSJ-06 Not Used CSJ-07 0CC-LW-63, -64 CSJ-08 Not Used CSJ-09 Not Used CSJ-10 1/2SI-836A, -836B CSJ-11 1/2SI-847A, -847B CSJ-12 1/2CS-466AA, -466BB, -476AA, -476BB CSJ-13 1/2CS-3124, 1/2CS-3125 CSJ-14 Not Used.

CSJ-15 1/2MS-2017, -2018 CSJ-16 Not Used CSJ-17 Not Used CSJ-18 1/2RC-570A, -570B, -575A, -575B, -580A, -580B CSJ-19 1/2RH-700, -720 CSJ-20 Not Used CSJ-21 Not Used CSJ-22 1/2SI-853A, -853B CSJ-23 1/2SI-853C, -853D CSJ-24 1/2SI-897A, -897B CSJ-25 1/2RH-702 Page 1 of26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS COLD SHUTDOWN TEST JUSTIFICATION - CSJ-01 Not Used Page 2 of26

POINT BEACH NUCLEAR PLANT 1ST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST ruSTIFICATIONS COLD SHUTDOWN TEST ruSTIFICATION - CSJ-02 Not Used Page 3 of26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST ruSTIFICA TIONS COLD SHUTDOWN TEST ruSTIFICATION - CSJ-03 Not Used Page 4 of26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS COLD SHUTDOWN TEST JUSTIFICATION - CSJ-04 System: Chemical and Volume Control Valve(s): 1(2)CV-00313A Categmy: A Code Class: 2 Function: These air operated valves are located in the CVCS seal water return line from the RCP shaft seals to the VCT. The valves have no safety function in the open position. The normal seal water return line flow path is not required for accident mitigation or to bring the plant to safe shutdown. These valves perfo1m an active safety function in the closed position. 1(2)CV-313 and -313A are designated containment isolation valves for contaimnent penetration P-11. As containment isolation valves, they must be capable of automatic closure upon receipt of a containment isolation "T signal to maintain containment integrity.

Defe1Ted Test Justification: Exercising these valves quarterly during nmmal operation would require intem1pting seal water return flow from the RCP shaft seals. The interruption of seal water return flow from the RCP shaft seals is not practical during power operation due to the potential of causing unnecessaiy accelerated wear to the seals and possible seal failure. A failed RCP shaft seal would allow unisolable leakage of reactor coolant from the RCS to the containment atmosphere possibly requiring plant shutdown per T.S. 3.4.13.

Quarterly Partial Stroke Testing: The valve control circuitry is not provided with partial stroke capability.

Alternate Test Frequency: Stroke test (including stroke time measurement) and fail-safe test to the closed position shall be perfmmed during cold shutdowns when plant conditions pe1mit the removal of the RCPs from service. Perfo1ming this testing during extended cold shutdowns when conditions pe1mit or at least each refueling is acceptable per paragraphs IV-3420(a) and IV-3430.

Page 5 of26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS COLD SHUTDOWN TEST JUSTIFICATION - CSJ-05 System: Chemical and Volume Control Valve(s): 1(2)CV-00371 1(2)CV-00371A Categmy: A Code Class: 2 Function: These air operated valves are located in the nonnal letdown line from the RCS loop "B" to the non-regenerative heat exchanger. The valves have no safety function in the open position. The process function of nmmal letdown serves to maintain a constant RCS inventmy, impurity removal, and boric acid concentration adjustment. None of these functions are required for accident mitigation or to achieve safe shutdown. These valves perform an active safety function in the closed position. 1(2)CV-371 and-371A are designated containment isolation valves for containment penetration P-10. As containment isolation valves, they must be capable of automatic closure upon receipt of a containment isolation "T" signal to maintain containment integrity.

Defened Test Justification: Exercising these valves to the closed position quarterly during power operation would require inte1111pting nmmal letdown flow. The intenuption ofnmmal letdown flow is not practical during power operation due to the potential of causing a pressurizer level control transient resulting in a reactor trip. In addition, failure of a letdown valve to reopen, subsequent to closure, while continuing to provide nmmal charging flow could result in a high RCS water level trip.

Quarterly Partial Stroke Testing: The valve control circuitry is not provided with patiial stroke capability.

Alternate Test Frequency: Stroke test (including stroke time measurement) and fail-safe test to the closed position shall be perfonned during cold shutdowns when the normal charging and letdown functions are not required as pe1mitted by paragraphs IV-3420(a) and IV-3430.

Page 6 of26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS COLD SHUTDOWN TEST JUSTIFICATION - CSJ-06 Not Used Page 7 of26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS COLD SHUTDOWN TEST JUSTIFICATION - CSJ-07 System: Component Cooling Water Valve(s): 0CC-LW-063 0CC-LW-064 Categmy: B Code Class: 3 Function: These nonnally open air operated valves are located the Unit 2 radwaste processing system CC supply and return lines. The valves perform an active safety function in the closed position and serve as ASME Class 3 to non-Code Class isolation valves. Therefore, CC-LW-63 and -64 are required to be capable of closure to isolate the non-essential CCW heat loads. These valves have no safety function in the open position. Their normally open position allows CCW flow to the non-safety-related radwaste system. The radwaste processing system is not required for accident mitigation or to achieve/maintain the plant in a safe shutdown condition. However, it is required to support normal plant operation.

Defen-ed Test Justification: The closure of these valves either by remote manual switch or upon receipt of a containment isolation signal results in the initiation of the radwaste auto shutdown circuit. This interlock is provided to prevent damage to various radwaste components which could occur as a result of a loss of CCW flow. Those components that would shutdown as a result of CC-LW-63 or -64 closing include:

the c1yogenic gas compressors, the auxiliaiy condensate return pump, and the letdown gas stripper circulating pump. Although these components are not required for accident mitigation or to achieve/maintain safe shutdown, they are required for support of nonnal process functions accomplished by the radwaste system and safe plant operation. Therefore, quarterly exercising of CC-LW-63 or

-64 with the subsequent need to manually restore operation of various radwaste system components is impractical without providing a compensating increase in the level of valve reliability.

Quarterly Partial Stroke Testing: The valve control circuitty is not provided with partial stt-oke capability.

Alternate Test Frequency: Stroke test (including stroke time measurement) and fail-safe test to the closed position shall be performed during cold shutdowns when minimal impact will occur to radwaste system operability as pe1mitted by paragraphs IV-3420(a) and IV-3430.

Page 8 of26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND 2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS COLD SHUTDOWN TEST JUSTIFICATION - CSJ-08 Not Used Page 9 of26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND 2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS COLD SHUTDOWN TEST JUSTIFICATION - CSJ-09 Not Used Page 10 of 26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS COLD SHUTDOWN TEST JUSTIFICATION - CSJ-10 System: Containment Spray Valve(s): 1(2)SI-00836A 1(2)SI-00836B Category: B Code Class: 2 Function: These n01mally closed air operated valves are located between the spray additive tanks (SAT) and the eductors. They perfonn an active safety function in the open position to provide a flow path for sodium hydroxide (NaOH) to the CS pump suction from the SAT via the spray eductors. This function allows the addition of sodium hydroxide to the spray stream for the removal of fission products released into the containment atmosphere during a LOCA. 1(2)SI-836A&B must be capable of automatically opening upon receipt of a CS actuation signal after the expiration of a 2-minute time delay. These valves perform an active safety significant function in the closed position in support of maintaining sump pH levels within design limits. However, closure capability of the valves is not an ISTA-1100 function; therefore, Mandatmy Appendix IV closure stroke testing is not required.

Deferred Test Justification: Exercising these valves requires the NaOH supply to the eductors be isolated to prevent contamination of the containment spray piping with sodium hydroxide.

NaOH is a highly corrosive fluid, requiring extensive flushing if exposed to CS system piping and components. The only means of isolation is by closure of a manual valve located in the common supply to both containment spray trains.

This action would render both trains unable to inject sodium hydroxide for containment post LOCA iodine control without manual operator action and would require declaring an entire safety system inoperable.

Quarterly Partial Stroke Testing: Partially exercising the valves would result in the same consequences as full stroke exercising.

Alternate Test Frequency: Stroke test (including stroke time measurement) and fail safe test to the open position shall be performed during cold shutdowns when the containment spray system is not required to be operable as permitted by paragraphs IV-3420(a) and IV-3430.

Page 11 of 26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST ruSTIFICA TIO NS COLD SHUTDOWN TEST ruSTIFICATION - CSJ-11 System: Containment Spray Valve(s): 1(2)SI-00847A 1(2)SI-00847B Categmy: C Code Class: 2 Function: These check valves are located in the spray additive lines from the SAT to the CS pumps' suction. They perform an active safety function in the open direction.

They must be capable of opening, subsequent to the upstream AOVs opening, to provide a flow path for NaOH to be directed to the CS pump suction. The addition ofNaOH to the spray stream is required for the removal of fission products released into the containment atmosphere following a LOCA. These valves also perform an active safety function in the closed direction. They must be capable of closure on reversal of flow to provide train separation during the event of containment spray pump operation with the SAT isolated. Additionally, the normally closed position prevents communication between the RWST supply and the NaOH piping which prevents inadve1ient dilution of the 30% weight NaOH contained in the SAT.

Defe11'ed Test Justification: Exercising these valves in the open direction requires aligning the RWST such that RWST inventmy can pass through the valves in lieu of using the highly co11'osive sodium hydroxide contained in the spray additive tank. This alignment requires the manipulation of various manual valves resulting in the inability to provide NaOH to the spray stream should a CS actuation signal occur during testing and rendering both trains of CS inoperable. In addition, contamination of the contaimnent spray piping with sodium hydroxide requires extensive flushing subsequent to exposure due to the cmTosive nature of the solution.

Alternate Test Frequency: Full stroke exercise to the open position shall be performed during cold shutdowns when the containment spray system is not required to be operable.

Closure testing will be perfmmed during refuelings with justification provided in ROJ-19.

Page 12 of26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS COLD SHUTDOWN TEST JUSTIFICATION - CSJ-12 System: Feedwater Valve(s): 1(2)CS-00466AA 1(2)CS-00466BB 1(2)CS-00476AA 1(2)CS-0047 6BB Categmy: C Code Class: 2 Function: These nonnally open check valves are in the nonnal feedwater flow path to the steam generators. The valves have no safety function in the open direction.

Nmmal feedwater flow is necessaiy for steam production during nonnal plant operation, but is not required for accident mitigation or to achieve/maintain safe shutdown of the plant. These valves perfmm an active safety function in the closed direction. The AFW injection lines tie in downstream of the main feedwater check valves. These valves must be capable of closure on reversal of flow during a loss of normal feedwater (LONF) to prevent the diversion of AFW flow from the steam generator to the non-safety-related feedwater piping. Also, subsequent to feedwater isolation during a SGTR or MSLB, automatic closure of the inboard valves on reversal of flow serve to isolate the faulted steam generator.

In the case of a MSLB the check valves close to limit the energy release to containment due to back flow from the intact steam generator to the faulted steam generator via the feedwater cross-connect. These requirements for isolation capability are redundant to that provided by the upstream (outboard) check valves for containment isolation.

Defen-ed Test Justification: Exercising the feedwater injection check valves in the reverse direction is impracticable quarterly during power operation due to the necessity of isolating nmmal feedwater flow to the associated steain generator. Isolation of feedwater flow during nmmal operation would cause a loss of steam generator level control potentially resulting in a plant trip.

Alternate Test Frequency: These check valves will be exercised in the closed direction by performing a seat leakage test during cold shutdown when feedwater is not required to be inservice.

As valves in regular use, fmward exercising shall be considered satisfactorily perfmmed by the ability to maintain steam generator level.

Page 13 of26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST ruSTIFICATIONS COLD SHUTDOWN TEST WSTIFICATION - CSJ-13 System: Main F eedwater Valve(s): 1(2)CS-03 l 24 1(2)CS-03125 Categmy: B Code Class: Non-Class Function: These nonnally open, air operated valves, are located in the Main Feedwater headers to the Steam Generators and serve as the Main Feedwater Isolation Valves (MFIVs). The MFIVs isolate main feedwater (MFW) flow to the secondaiy side of the Steam Generators following a High Energy Line Break (HELB). The safety related function of the MFRVs is to provide the second isolation of MFW flow to the secondary side of the Steam Generators following a HELB. Closure of the MFIVs, MFRVs, and MFRV bypass valves terminates flow to the Steam Generators, terminating the event for Feedwater Line Breaks (FWLBs) occmTing upstream of the MFIVs or MFRVs. The consequences of events occurring in the Main Steam lines or in the MFW lines downstream from the MFIVs will be mitigated by their closure. Closure of the MFIVs, MFRVs, and MFRV bypass valves, effectively tenninates the addition ofFeedwater to an affected Steam Generator, limiting the mass and energy release for Steam Line Breaks (SLBs) or FWLBs inside Containment, and reducing the cooldown effects for SLBs. The MFIVs have no safety function in the open direction. The MFIV s remain open during normal operation to allow F eedwater flow to the Steam Generators to support power generation. This function is not required for accident mitigation and is not a safety related function.

Deferred Test Justification: Exercising these valves during nonnal operation isolates the flow ofFeedwater to the associated Steam Generator. Isolation of a Main Feedwater header could cause a pressure transient in the associated Main Feedwater line. In addition, it would cause a mis-match between Feedwater flow into the affected Steain Generator and the steam flow out of the Steam Generator which could result in a possible plant trip. Reducing power level to pe1f01m testing without causing a transient would significantly impact plant operations and power production.

PBNP Technical Specification Basis 3.7.3, Main Feedwater Isolation, states that these valves should not be tested at power since even a part stroke exercise increases the risk of a valve closure with the unit generating power.

Quarterly Partial Stroke Testing: The valve control circuitry is not designed with partial stroke capability.

Alternate Test Frequency: The MFIVs will be stroke tested (including stroke time measurement) and fail-safe tested to the closed position during cold shutdown as permitted by paragraphs IV-3420(a) and IV-3430. The supporting control air and nitrogen system valves will also be tested during cold shutdown.

Page 14 of 26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND 2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS COLD SHUTDOWN TEST JUSTIFICATION - CSJ-14 Not Used Page 15 of 26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST ruSTIFICATIONS COLD SHUTDOWN TEST WSTIFICATION - CSJ-15 System: Main Steam Valve(s): 1(2)MS-02017 1(2)MS-02018 Category: B Code Class: 2 Function: These nmmally open, air operated check valves are located in the main steam headers from the steam generators and serve as the main steam isolation valves (MSIV). The valves perfo1m an active safety function in the closed direction to prevent the umestricted release of steam from the steam generators during a main steam line break (MSLB). This function prevents blowdown from more than one steam generator for a break upstream or downstream of an MSIV. For an MSLB upstream of the MSIV additional isolation for the adjacent steam generator is provided by the non-return check valves. Other accident conditions resulting in closure of the MSIV include a steam generator tube rupture (SGTR) and a loss of reactor coolant (LOCA). Additionally, the MSIVs are designated outboard containment isolation valves for containment penetrations P-1 and P-2. As containment isolation valves, they must also be capable of closure to maintain containment integrity. The valves have no safety function in the open direction.

The MSIVs remain open during normal operation to allow steam flow from the steam generators to the main turbine to support power generation. This function is not required for accident mitigation and is not a safety related function.

Deferred Test Justification: Exercising these valves during normal operation isolates one line of steam flow to the turbine. Isolation of a main steam header would cause a severe pressure transient in the associated main steam line possibly resulting in a plant trip.

Additionally, closure of an MSIV, at power, could potentially result in challenging the set point of the main steam relief valves causing inadvertent lifting. Reducing power level to perfmm testing without causing a transient would significantly impact plant operations and power production.

Quarterly Paiiial Stroke Testing: The valve control circuit1y is not designed with partial stroke capability. The MSIVs are check valves which open against the direction of steam flow allowing rapid closure for stream line isolation.

Alternate Test Frequency: The MSIVs will be stroke tested (including stroke time measurement) to the closed position during cold shutdown as pe1mitted by paragraphs IV-3420(a) and IV-3430. As valves in regular use, fo1ward exercising shall be considered satisfactorily perfmmed by the ability to provide steam flow to the turbine as permitted by paragraph IV-3410(e).

Page 16 of 26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS COLD SHUTDOWN TEST JUSTIFICATION - CSJ-16 Not Used Page 17 of 26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS COLD SHUTDOWN TEST JUSTIFICATION - CSJ-17 Not Used Page 18 of 26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS COLD SHUTDOWN TEST JUSTIFICATION - CSJ-18 System: Reactor Coolant Valve(s): 1(2)RC-00570A, 1(2)RC-00570B 1(2)RC-00575A, 1(2)RC-00575B 1(2)RC-00580A, 1(2)RC-00580B Categmy: B Code Class: 1 Function: These normally closed pilot operated solenoid valves are part of the RCS gas vent system and are located in the reactor vessel head vent lines. The valves perfonn an active safety function in the open position. They must be capable of opening by remote manual switch actuation to vent non-condensible gases from the reactor vessel head space during post-accident conditions.

Deferred Test Justification: Exercising these valves during power operation with subsequent failure to reclose or significant leakage following closure could result in a loss of coolant in excess of the limits imposed by TS 3.4.13 leading to a plant shutdown. Additionally, as pilot operated solenoid valves, system pressure is utilized for motive force to open the valves. The valves may not properly close if the upstream pressure is equal to or less than downstream pressure which increases the potential for seat leakage, providing further justification for exercising the valves during cold shutdown.

Quarterly Partial Stroke Testing: The control circuitiy of the valves is not provided with partial sti*oke capability.

In addition, partially exercising the valves would result in the same consequences as full sti*oke exercising.

Alternate Test Frequency: Exercise, and stroke time testing to the open position shall be perfo1med during cold shutdowns.

Page 19 of 26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS COLD SHUTDOWN TEST JUSTIFICATION - CSJ-19 System: Residual Heat Removal Valve(s): 1(2)RH-00700 1(2)RH-00720 Category: B Code Class: 1 Function: These nonnally closed motor operated valves are located in the RHR supply and return lines from the RCS. The valves perfonn no safety function in the open position. They are placed in the open position for initiation ofRHR shutdown cooling. However, the shutdown cooling mode of RHR is not required for accident mitigation or to achieve/maintain safe shutdown and is not considered safety related. It is however considered a risk significant function and components supporting this function shall be subject to augmented testing. These valves do perform an active safety function in the closed position. They must be capable of closure by remote manual switch actuation, if open for normal shutdown cooling, to allow RHR to be realigned to the ECCS mode of operation as required by TS 3.5.3 and to prevent overpressurization of the RHR system should RCS pressure rise above the RHR system design pressure. Although not identified in TRM 4.16, these valves are RCS pressure boundaiy isolation valves which perfmm a PN function.

Defetred Test Justification: Exercising these valves quarterly during power operation would require defeating an interlock and protective measures intended to protect the RHR system piping and components from overpressurization from the RCS. Full or partial-stroke exercising at power would result in overpressurizing the RHR system piping and a loss of containment integrity. Valve exercising shall be perfmmed during cold shutdown when RCS pressure is less than RHR system design pressure.

Interlocks and protective lockouts are provided to prevent inadvertent opening of the valves when RCS pressure is greater than the RHR system design pressure.

There is negligible potential increase for these active MOVs in core damage frequency (CDF) and large early release frequency (LERF) associated with maintaining the exercise interval in the Sixth IST Interval as previous IST 10-year intervals.

Quarterly Partial Stroke Testing: The control circuitry of the valves is not provided with partial stroke capability.

In addition, partially exercising the valves would result in the same consequences as full stroke exercising.

Alternate Test Frequency: Full stroke exercising and stroke timing to the closed positions shall be perfo1med during cold shutdowns when RCS pressure is less than RHR system design pressure.

Page 20 of26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS COLD SHUTDOWN TEST JUSTIFICATION - CSJ-20 Not Used Page 21 of 26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS COLD SHUTDOWN TEST JUSTIFICATION - CSJ-21 Not Used Page 22 of26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS COLD SHUTDOWN TEST JUSTIFICATION - CSJ-22 System: Residual Heat Removal Valve(s): 1(2)SI-00853A 1(2)SI-00853B Categmy: AC Code Class: 1 Function: These nonnally closed check valves are located inside containment in the low head safety injection flow path to the RCS. The valves perfonn an active safety function in the open direction. 1(2)SI-853A&B must be capable of opening subsequent to system initiation to provide a path for post-LOCA low head safety injection and recirculation flow to the RCS for emergency core cooling when RCS pressure has been reduced to below the shutoff head of the pumps (334 ft.).

These valves also perform an active safety function in the closed direction.

1(2)SI-853A&B are designated containment isolation valves for containment penetrations P-8 and P-22. As containment isolation valves, 1(2)SI-853A&B must be capable of closure on reversal of flow to maintain containment integrity.

In addition, the valves serve as ASME Code Class 1 to Class 2 bounda1y barrier valves which perform a leakage important safety function as Event V pressure isolation valves (PIV). Their nmmally closed position preserves the pressure boundaiy integrity of the RCS and isolates RCS pressure from the attached low pressure RHR piping. The valves nmmally closed position also prevents diversion of core deluge injection flow if the SI pumps are aligned to provide ECCS flow via the core deluge nozzles.

Defe1Ted Test Justification: Full stroke exercising of these valves in the forward direction quarterly during power operation is not possible due to insufficient pump discharge head to overcome reactor pressure. Exercising the valves in the forward direction during cold shutdown is not desirable unless leak testing per Technical Specification 3 .4 .14 is scheduled to ensure valve leak tight integrity is verified subsequent to closure. Reverse exercising these check valves is best accomplished during the leak tight verification testing required by T.S. 3.4.14 which is perfmmed prior to entering Mode 2 whenever the unit has been in Mode 5 for 7 days or more, if leak testing has not been perfmmed in the previous 9 months. This leak tight verification testing can not be perfmmed quarterly during power operation due to the necessity of manual realignment of the RHR system rendering both trains of low head safety injection inoperable.

Alternate Test Frequency: Valve exercising in the fo1ward and reverse directions shall be performed during cold shutdown when the testing requirements of Technical Specification 3 .4.14 are scheduled to be performed and LHSI is not required to be operable.

Page 23 of 26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST JUSTIFICATIONS COLD SHUTDOWN TEST JUSTIFICATION - CSJ-23 System: Residual Heat Removal Valve(s): 1(2)SI-00853 C 1(2)SI-00853D Categmy: AC Code Class: 1 Function: These check valves are located inside containment in the low head safety injection and SI core deluge injection lines to the reactor vessel. The valves perform an active safety function in the open direction. 1(2)SI-853C&D must be capable of opening to provide a path for post-LOCA low head safety injection and recirculation flow to the RCS for emergency core cooling. These valves also perform an active safety function in the closed direction. 1(2)SI-853C&D are one of two valves providing the ASME Code Class 1 to Class 2 boundaiy barrier and perfonn a leakage important safety function as an Event V pressure isolation valves(PIV). The normally closed position of these valves preserve the pressure boundaiy integrity of the RCS and isolates RCS pressure from the attached low pressure RHR piping.

Deferred Test Justification: Full stroke exercising of these valves in the f01ward direction quarterly during power operation is not possible due to insufficient pump discharge head to overcome reactor pressure. Exercising the valves in the fmward direction dming cold shutdown is not desirable unless leak testing per Technical Specification 3 .4.14 is scheduled to ensure valve leak tight integrity is verified subsequent to closure. Reverse exercising these check valves is best accomplished during the leak tight verification testing required by TS 3 .4 .14 which is performed prior to entering Mode 2 whenever the unit has been in Mode 5 for 7 days or more, if leak testing has not been perfmmed in the previous 9 months. This leak tight verification testing can not be performed quarterly during power operation due to the necessity of manual realignment of the RHR system rendering both trains of low head safety injection inoperable.

Alternate Test Frequency: Valve exercising in the fmward and reverse directions shall be performed during cold shutdown when the testing requirements of Technical Specification 3.4.14 are scheduled to be performed and LHSI is not required to be operable.

Page 24 of26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND 2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST ruSTIFICATIONS COLD SHUTDOWN TEST WSTIFICATION - CSJ-24 System: Safety Injection Valve(s): 1(2)SI-00897A 1(2)SI-00897B Categmy: A Code Class: 2 Function: These nonnally open air op~rated valves are located in the SI injection check valves' test return line to the RWST and are installed in series. The valves perfonn a safety function in the (PASSIVE) OPEN position. The normally open position of 1(2)SI-897A&B provides a return path for SI pump minimum flow recirculation back to the RWST. This minimum flow recirculation path is required to prevents damage to the SI pumps as a result of operating in low flow or dead-headed conditions. The open position of 1(2)SI-897 A&B also provides an overpressure protection relief path for the containment spray pumps' suction piping. A relief path is provided for the CS pumps' suction piping to prevent overpressurization as a result of RHR inleakage to the CS system during RHR system operation. These valves also perfmm an active safety function in the closed position. 1(2)SI-897A&B are required to close during the switchover from the injection mode to the recirculation mode of SI. Additionally, the SI system serves as a closed system outside containment for the purposes of contaimnent isolation. Valves 1(2)SI-897A&B are designated as containment closed system boundaiy valves. Therefore, they have a safety function in the closed position to maintain containment integrity.

Defe1Ted Test Justification: 1(2)SI-897A&B are equipped with fail-open design air actuators. At power the valves are disabled in the open position by manual isolation of their actuating air supply. Exercise testing these valves quarterly during power operation would require unisolating the actuating air supply. This activity is time consuming and could compromise minimum flow protection for the SI pumps should either valve fail to reopen subsequent to closure or should a malfunction occur with the instrnment air supply resulting in the inoperability of an entire ECCS. These valves remain open during post-accident conditions except for the high head recirculation phase of emergency core cooling. During the transitioning to recirculation, operators are dispatched to valve in the air supply and/or locally close the valves.

Quarterly Partial Stroke Testing: Partial stroke exercising will not be perfmmed quarterly for the same reasons provided for not pe1f01ming full stroke exercising.

Alternate Test Frequency: Full stroke exercising and stroke time testing to the closed position shall be performed during cold shutdown when the SI pumps ai*e not required to be operable. In addition, the valves will be manually full stroke exercised during cold shutdown due to the fail-open actuator design and the lack of a qualified actuating air supply.

Page 25 of26

POINT BEACH NUCLEAR PLANT IST APPENDIX C UNITS 1 AND2 Revision 4 INSERVICE TESTING PROGRAM COLD SHUTDOWN TEST ruSTIFICA TIONS COLD SHUTDOWN TEST WSTIFICATION - CSJ-25 System: Residual Heat Removal Valve(s): 1(2)RH-00702 Categmy: C Code Class: 2 Function: These check valves are located in a branch connection to CVCS off the RHR "A" train low head safety injection(LHSI)/shutdown cooling header inside the primaiy containment. The valves perfonn an active safety function in the open direction.

No relief valves are installed in the LHSI/RHR Train "A" piping. 1(2)RH-702 must open to provide a pressure relief flow path between the LHSI/RHR piping and the letdown orifice outlet relief valve, 1(2)CV-203. The RHR/LHSI piping has a design pressure and temperature of 700 psig and 400°F. It is connected to the reactor coolant pressure boundaiy which has a design pressure and temperature of 2580 psig and 650°F. Overpressure protection is required to prevent overpressurization of the lower pressure LHSI piping in the event of in-leakage from the high pressure RCS. There is no accident flow rate associated with the safety function of RH-702 in the forward direction. These check valves also perfmm an active safety function in the closed direction. 1(2)RH-702 are designated inboard isolation valves for contaimnent penetration P-8. The containment isolation boundaiy criteria for this penetration are remote manual isolation valves and/or valves capable of automatic closure to function as barriers inside containment and a closed system outside containment. As containment isolation valves, 1(2)RH-702 must be capable of closure to maintain containment integrity.

Deferred Test Justification: Exercising these check valves in the partially open direction quarterly during power operation would require initiating flow from RHR to the CVCS letdown flow stream. Initiating RHR flow to CVCS letdown is not possible due to insufficient discharge head of the RHR pumps to overcome CVCS system pressure.

Alternate Test Frequency: Exercising these check valves in the partially open direction will be perfmmed during cold shutdowns when CVCS charging and letdown can be removed from service. Closure verification will be perfmmed quarterly during RHR pump A testing.

Page 26 of26

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS APPENDIXD REFUELING OUTAGE TEST JUSTIFICATIONS ROJ-01 1/2AF-191, l/2AF-111 ROJ-02 1/2AF-100, -101 ROJ-03 l/2AF-106,-107,-108, lAF-102, -104,2AF-103,-105 1/2AF-148, l/2AF-193A&B ROJ-04 1/2CV-294 ROJ-05 l/2CV-370 ROJ-06 1/2CV-295 ROJ-07 1/2CV-304A, -304B ROJ-08 1/2CV-304C, -304D ROJ-09 l/2CV-383 ROJ-10 l/2CC-745 ROJ-11 1/2CC-755A, -755B, 1/2CC-767 ROJ-12 1/2RM-3200AA ROJ-13 l/2RC-528 ROJ-14 1/2RC-529 ROJ-15 0SW-112A, -l35A ROJ-16 l/2SI-834D ROJ-17 l/2SI-845A, -845B, -845C, -845D, -845E, -845F ROJ-18 1/2SI-862A, -862B ROJ-19 1/2SI-847A, -847B ROJ-20 1/2SI-889A, -889B ROJ-21 l/2SI-867B ROJ-22 l/2SI-867 A, 1/2SI-842A, -842B ROJ-23 0FP-296A, -304A ROJ-24 1/2SI-854A, -854B ROJ-25 l/2SI-875A, -875B ROJ-26 1/2SI-891A, -891B ROJ-27 1/2RH-710A, -710B ROJ-28 Not Used ROJ-29 l/2SI-858A, -858B ROJ-30 Not Used ROJ-31 1/2SW-15A, -15B, -15C, -15D ROJ-32 l/2AF-l 14, 1/2AF-196 Page 1 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIXD UNITS 1AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST msTIFICATIONS REFUELING OUTAGE TEST WSTIFICATION -ROJ-01 System: Auxiliaiy Feedwater Valve(s): 1(2)AF-001 l 1 1/2AF-00191 Catego1y: C Code Class: 3 Function: These normally closed check valves is located in AFW pumps' suction supply lines from the CSTs. The valves perform an active safety function in the closed direction. Upon depletion of CST inventory or when the CSTs are unavailable, the suction supply for the AFW pumps is provided by the service water system.

When the AFW pumps are aligned to the service water system for a suction supply source, these Class 3 to non-Code boundaty barrier check valves close to maintain pressure boundaty and to prevent the service water supply from being diverted to the CSTs. These valves perform a safety ft.mction in the open position to allow the flow of water from the CS Ts during initial pump start up to prevent pump damage until Service Water can be valved in or the pump's tripped on low suction pressure.

Deferred Test Justification: There are no test connections to enable closure verification of these check valves by leak rate testing. Additionally, closure verification by aligning the pump suction to service water is undesirable due to the necessity of injecting service water into the steam generators. Injecting service water into the steam generators would result in unnecessarily subjecting the steam generators to premature degradation due to lack of maintaining proper water chemistry. The only practical means of verifying closure capability of these check valves, with the exception of disassembly, is by performing a radiographic examination test (RT) or other qualified test method on the valve body to demonstrate the valve disk is in the closed position. Due to the labor intensive nature of non-intrnsive testing, perfmming this type of testing activity quarterly during power operation is impractical without providing a commensurate increase in the level of valve reliability. Performing this type of test activity during cold shutdown is impractical from a logistics standpoint as tests are perfmmed by an off-site contractor. During unplanned cold shutdowns the primaty concern is to safely restart the plant when the condition which required going to cold shutdown is corrected. Therefore, the coordination of outside contractor notification and the time required for equipment setup is impractical for the purpose of testing and could delay plant restart.

Page 2 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST WSTIFICATIONS Alternate Test Frequency: Closure verification of the AFW pumps' suction check valves from CSTs will be accomplished by performing a qualified test method in the CVCM Program. This method of testing and frequency is suppotied by the discussion provided in NUREG-1482, Rev.3, Section 4.1.2. Check valve exercising in the non-safety related open direction will be performed during quaiierly pump testing.

As an alternative to perf01ming a non-intrnsive testing each refueling outage, these check valves may be placed in the Check Valve Condition Monitoring (CVCM) Program. If so, the qualified test method shall be performed on a sampling basis at the frequency specified in the applicable CVCM Program Plan for the associated valve group.

Page 3 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION - ROJ-02 System: Auxiliary Feedwater Valve(s): 1(2)AF-00100 1(2)AF-00101 Category: C Code Class: 2 Function: These normally closed check valves are located inside containment in the AFW injection lines to the steam generators and serve as the first-off check from the S/Gs. The valves perfonn an active safety function in the closed direction to isolate the AFW system from the main feedwater system. They serve as a barrier at the piping class break to prevent the diversion of high temperature main feedwater into the low temperature AFW system piping. Additionally, main feedwater inleakage to the AFW system may result in voiding in the piping and could result in a loss of availability of the AFW pumps due to steam binding. The valves also perform an active safety function in the open direction to provide a path for auxiliary feedwater flow to the steam generators subsequent to a loss of normal feedwater flow and for various other postulated accidents requiring AFW actuation.

Deferred Test Justification: Serving as the first off check valves from the steam generators there are no isolation valves or test connections located downstream to enable closure verification of these check valves by leak rate testing. There are drain connections located upstream of the check valves which could be utilized to verify a pressure drop across the valves' disk. However, opening these drain valves during power operation represents substantial personnel risks and would create a condition requiring manual action to restore proper alignment should AFW receive an actuation signal. During cold shutdowns these upstream drain connections could be utilized to verify differential pressure exists across the valve seat. However, the differential pressure across the valve disk, due to static head from the steam generators, would be approximately 5 psid which may be inadequate pressure to accomplish a meaningful test.

Page 4 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS Alternate Test Frequency: These check valves will be full stroke exercised with flow during quarterly pump testing. Bi-directional exercising of these check valves (ISTC-3522(a) will be satisfied by a sample disassembly examination program to confinn obturator movement (ISTC-5221(c). One valve per Unit will be manually exercised to confirm obturator travel, open and closed, each refueling outage.

These Enertech 3 inch check valves (Model DRV-Z) are oriented horizontally (Bech.Dwg. P-140, 142,239, 242) and perform identical functions. These valves qualify for inclusion in a sampling program (ISTC-5221 (c)(1 ).

As an alternative to the disassembly and examination requirements of ISTC-5221 (c), these check valves may be placed in the Check Valve Condition Monitoring (CVCM) Program and disassembled and examined per the requirements of Appendix II of the OM Code. If so, the frequency for disassembly and examination shall be specified in the CVCM Program Plan applicable to the valve group.

Page 5 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST msTIFICATIONS REFUELING OUTAGE TEST WSTIFICATION - ROJ-03 System: Auxiliaiy Feedwater Valve(s): 1(2)AF-OO106 1(2)AF-00107 1(2)AF-00108 lAF-00102 lAF-00104 1(2)AF-00148 2AF-00103 2AF-00105 1(2)AF-193A 1(2)AF-193B Categmy: C Code Class: 2 Function: These normally closed check valves are located outside containment in the AFW injection lines to the steam generators and serve as the second-off check from the S/Gs. These valves perform an active safety function in the closed direction to isolate the AFW system from the main feedwater system. They serve as a barrier at the piping class break to prevent the diversion of high temperature main feedwater into the low temperature AFW system piping. Additionally, main feedwater inleakage to the AFW system may result in voiding in the piping and could result in a loss of availability of the AFW pumps due to steam binding.

Valve closure also provides redundant isolation capability to prevent diversion of flow from an adjacent pump thereby ensuring AFW accident flow is properly directed to S/Gs. 1(2)AF-108 and the 1(2)AF-148 pe1form no safety function in the closed direction but are verified closed to satisfy bi-directional testing and a matter of good Engineering practice to support isolation of AFW flow paths.

With the exception of lAF-102, lAF-104, 2AF-103 and 2AF-105, the valves also perfonn an active safety function in the open direction to provide a path for auxiliaiy feedwater flow to the steam generators subsequent to a loss of normal feedwater flow and for various other postulated accidents requiring AFW actuation.

Page 6 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS Defe1Ted Test Justification: The AFW second off check valves from the steam generators are not provided with downstream isolation and test connections in a configuration allowing the ability for closure verification by leak rate testing. The only practical means of verifying closure capability of these check valves, with the exception of disassembly, is by perfonning a radiographic examination test or other qualified test method on the valve body to demonstrate the valve disk is in the closed position. Due to the labor intensive nature of non-intrusive testing, perfmming this type of testing activity quarterly during power operation is impractical without providing a commensurate increase in the level of valve reliability.

Perfo1ming this type oftest activity during cold shutdown is impractical from a logistics standpoint as tests are performed by an off-site contractor. During unplanned cold shutdowns the primary concern is to safely restart the plant when the condition which required going to cold shutdown is c01Tected. Therefore, the coordination of outside contractor notification and the time required for equipment setup is impractical for the purpose of testing and could delay plant restart. The valves' nmmally closed position is continuously monitored by observing line temperature via thermocouples. Continuous monitoring of line temperature allow operators sufficient time to take appropriate action to prevent steam binding of the AFW pumps should feedwater inleakage occur.

Alternate Test Frequency: Closure verification of the AFW second-off injection check valves to the steam generators and pump discharge checks will be accomplished by performing an RT during each refueling outage. This method of testing and frequency is suppmied by the discussion provided in NUREG-1482, Rev.3, Section 4.1.2. These check valves will be full stroke exercised with flow during quarterly pump testing.

As an alternative to perf01ming a qualified test each refueling outage, these check valves may be placed in the Check Valve Condition Monitoring (CVCM)

Program. If so, the qualified test shall be perfonned on a sampling basis at the frequency specified in the applicable CVCM Program Plan for the associated valve group.

Page 7 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION - ROJ-04 System: Chemical and Volume Control Valve(s): 1(2)CV-00294 Catego1y: AC Code Class: 2 Function: These check valves are located inside containment in the bypass line around the seal return header isolation valves 1(2)CV-313A. The valves perform an active safety function in the partial open and closed directions. 1(2)CV-294 must be capable of partially opening to provide thermal overpressure protection for containment penetration P-11 when the penetration isolation valves are closed.

As containment isolation valves, 1(2)CV-294 must be capable of closure on cessation or reversal of flow to maintain containment integrity.

Defen-ed Test Justification: Exercising these valves partially open or in the reverse direction requires intenupting normal seal cooling return flow from the RCPs. To satisfactorily exercise these check valves requires the use oftempora1y test equipment inside containment to perfo1m a leak test or back flow test, in addition to passing flow through the valves to demonstrate their partial opening capability. Such testing activities, if perfmmed during power operation, could cause damage to the RCP shaft seals as a result of interrupting seal cooling water flow. Due to the considerable effort associated with these test activities, exercise testing to the partially open or closed positions during cold shutdown is considered impractical due to the necessity of utilizing tempora1y test equipment inside containment.

Exercise testing of 1(2)CV-294 to the partially open and closed positions shall be perfmmed during refueling.

Alternate Test Frequency: Closure verification of these check valves shall be perfonned during refueling outages. Demonstrating the partial opening capability shall also be performed during refueling outages or during cold shutdowns of an extended duration when both RCPs can be removed from service. To demonstrate partial opening, flow will be provided in the fmward direction by an outside pressure source. There is no accident flow rate associated with the valves' safety function in the open direction. The deferral of test frequency to refueling outages is acceptable per the discussion provided in NUREG-1482, Rev.3, Section 4.1.6.

Page 8 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST msTIFICATIONS REFUELING OUTAGE TEST mSTIFICATION - ROJ-05 System: Chemical and Volume Control Valve(s): 1(2)CV-00370 Ca,.~egory: AC Code Class: 2 F~inction: These nonnally open check valves are located inside containment in the charging header to the RCS loop A cold leg and auxiliary spray line. As containment isolation valves, they perfo1m an active safety function in the closed direction to maintain containment integrity. The valves perform no safety function in the open direction. Check valves 1(2)CV-370 open to support nmmal process functions perfo1med by the CVCS.

Defened Test Justification: The only method available to verify reverse flow closure capability of these check valves is by seat leakage testing. The test connections utilized to perform seat leakage testing are located inside containment. Therefore, it would require containment entry and the intenuption of the valves' normal process functions in order to verify their closure capability. Exercising these check valves in the reverse direction requires interrupting normal charging flow and the use of temporary test equipment inside containment. Such testing activities if performed during power operation could result in a pressurizer level transient causing a plant trip. Due to the considerable effort associated with these test activities, reverse exercise testing during cold shutdown is considered impractical due to the necessity of utilizing temporary test equipment inside containment.

Alternate Test Frequency: Closure verification of these check valves shall be perfo1med during refueling outages when performing Appendix J Type C seat leakage testing. The deferral of test frequency to refueling outages is acceptable per the discussion provided in NUREG-1482, Rev.3, Section 4.1.6. These valves remain in the open position during power operation in support of the nmmal process functions perfo1med by the CVCS.

As an alternative to reverse exercising each refueling outage, these Catego1y AC check valves may be placed in the Check Valve Condition Monitoring (CVCM)

Program and reverse exercised at the Appendix J, Option B frequency during the performance of Type C seat leakage testing. This extension of frequency for reverse exercising is per the guidelines provided in NUREG-1482, Rev.3, Section 4.4.7, "Use of Appendix J, Option B, in Conjunction with ISTC Exercising Tests".

Page 9 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION - ROJ-06 System: Chemical and Volume Control Valve(s): 1(2)CV-00295 Category: C Code Class: 1 F1mction: This check valve is located inside containment in the CVCS nmmal charging line.

CV-295 performs an active safety function in the closed direction. The valve serves as ASME Class 1 RCS boundaiy baiTier isolation valve as defmed in 10CFR50.2. As such, the valve must be capable of closure to maintain the integrity of the RCS pressure boundmy in the event of a failure of upstream components. The valve has no safety function in the open direction.

Deferred Test Justification: The nmmal charging line check valve is not provided with downstream isolation or test connections allowing the ability for closure verification by leak rate testing. The only practical means of verifying closure capability of this check valve, with the exception of disassembly, is by performing a radiographic examination test (RT) or other qualified test on the valve body to demonstrate the valve disk is in the closed position. Verifying closure capability of the nmmal charging line check valve is not possible during power operation due to the necessity of interrupting normal charging flow which could result in a pressurizer level transient causing a plant trip. Perfmming this type of test activity during cold shutdown is impractical from a logistics standpoint tests are performed by an off-site contractor. During unplanned cold shutdowns the primmy concern is to safely restart the plant when the condition which required going to cold shutdown is conected. Therefore, the coordination of outside contractor notification and the time required for equipment setup is impractical for the purpose of testing and could delay plant restart.

Alternate Test Frequency: Closure verification of the CVCS nmmal charging line check valve will be accomplished by performing an RT during each refueling outage. This method of testing and frequency is supported by the discussion provided in NUREG-1482, Rev.3, Section 4.1.2. The nmmal charging line check valves 1(2)CV-295 remain in the open position during power operation to provide a flow path for normal charging.

As an alternative to performing a qualified test each refueling outage, these check valves may be placed in the Check Valve Condition Monitoring (CVCM)

Program. If so, the qualified test shall be performed on a sampling basis at the frequency specified in the applicable CVCM Program Plan for the associated valve group.

Page 10 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION - ROJ-07 System: Chemical and Volume Control Valve(s): 1(2)CV-00304A 1(2)CV-00304B Category: C Code Class: 1 Function: These check valves are located inside containment in the CVCS seal water injection line to the RCP shaft seals. The valves perform an active safety function in the closed direction. 1(2)CV-304A&B are ASME Class 1 to Class 2 RCS pressure boundary isolation valves as defined in 10CFR50.2. Therefore, the valves must be capable of closure to maintain the integrity of the RCS pressure boundaiy in the event of a failure of upstream components. The valves perfmm a safety function in the open direction to provide a relief path during a thermally induced overpressure condition of the containment penetration piping post-LOCA. There is no design flow rate associated with the open safety function. The valves also perfmm process functions; the seal water injection flow path is one of three flow paths available to the RCS for alternate boration.

However, the SI pumps are credited with the function of providing boration if the charging pumps are unavailable. In addition, seal water injection assures the integrity of the RCP shaft seals. However, the RCPs are not relied on for accident mitigation or to achieve/maintain the plant in a safe shutdown.

Page 11 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST WSTIFICATIONS DefeITed Test Justification: These seal water injection check valves are the first-off check valves from the RCPs #1 seal with no means of isolation between the check valves and the shaft seal. The valves are provided with upstream and downstream test connections.

However, utilizing the downstream test connections to apply an outside pressure source to establish a t1P across the valve seat may not provide a meaningful reverse exercise test result. Applying sufficient pressure to prevent the backflow of reactor coolant through the seal would result in a portion of the applied pressure being injected into the RCS via the seal. In addition, verifying closure by backflow from RCS static or residual pressure is undesirable due to ALARA concerns and the potential of trapping debris in the seals causing unnecessary wear to the shaft sealing surface when the associated RCP is returned to service.

The prefe1Ted method of verifying closure capability of these check valves is by perfonning a radiographic examination test (RT) or other qualified test on the valve body to demonstrate the disk is in the closed position. Perfonning this test quarterly during power operation is not possible due to the necessity of removing an RCP for service to prevent seal damage when stopping seal water flow. TS 3.4.4 requires both reactor coolant pumps to be in service whenever the reactor is critical. Perfonning this type oftest activity during cold shutdown is impractical from a logistics standpoint tests are performed by an off-site contractor. During unplanned cold shutdowns the primary concern is to safely restart the plant when the condition which required going to cold shutdown is c01Tected. Therefore, the coordination of outside contractor notification and the time required for equipment setup is impractical for the purpose of testing and could delay plant restart.

Alternate Test Frequency: Closure verification of the CVCS seal water injection line check valves 1(2)CV-304A&B will be accomplished by performing a qualified test method during each refueling outage. This method of testing and frequency is supported by the discussion provided in NUREG-1482, Rev.3, Section 4.1.2. These valves remain in the open position during normal power operation in support of normal seal water flow to the RCPs. Sufficient flow through the valves is verified by observation of seal temperatures.

As an alternative to performing a qualified test each refueling outage, these check valves may be placed in the Check Valve Condition Monitoring (CVCM)

Program. If so, the qualified test shall be performed on a sampling basis at the frequency specified in the applicable CVCM Program Plan for the associated valve group.

Page 12 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST WSTIFICATIONS REFUELING OUTAGE TEST WSTIFICATION - ROJ-08 System: Chemical and Volume Control Valve(s): 1(2)CV-00304C 1(2)CV-00304D Category: AC Code Class: 1 Function: These check valves are located in the CVCS seal water injection line to the RCP shaft seals. The valves perfonn an active safety :function in the closed direction.

Check valves 1(2)CV-304C&D serve as inside containment isolation automatic trip valves for the CVCS seal water supply line to the RCP shaft seals. As such, 1(2)CV-304C&D must be capable of closure on reversal of flow to maintain containment integrity. Additionally, 1(2)CV-304C&D are ASME Class 1 to Class 2 RCS pressure boundary isolation valves as defined in 10CFR50.2.

Therefore, the valves must also be capable of closure to maintain the integrity of the RCS pressure boundary in the event of a failure of upstream components. The valves perfo1m an safety :function in the open direction to provide a relief path during a thermally induced overpressure condition of the containment penetration piping post-LOCA. There is no design flow rate associated with the safety

function. The valves also perform process :functions; the seal water injection flow path is one of three flow paths available to the RCS for alternate boration.

However, the SI pumps are credited with the :function of providing boration if the charging pumps are unavailable. In addition, seal water injection assures the integrity of the RCP shaft seals. However, the RCPs are not relied on for accident mitigation or to achieve/maintain the plant in a safe shutdown.

Deferred Test Justification: The prefen:ed method to verify reverse flow closure capability of these check valves is by seat leakage testing. The test connections utilized to perfmm seat leakage testing are located inside containment. Therefore, it would require containment ently and the intenuption of the valves' nonnal process :functions in order to verify their closure capability. Exercising these check valves in the reverse direction requires intenupting nmmal RCP seal water flow and the use of tempora1y test equipment inside containment. Such testing activities if perfmmed during power operation would result unnecessary wear to the seals and potential premature failure of the RCP shaft seals rendering the associated pump inoperable. An inoperable RCP would require the plant to be placed in hot shutdown within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , as TS 3.4.4 requires both RCPs to be in operation when the reactor is critical. Due to the considerable effort associated with these test activities, reverse exercise testing during cold shutdown, if the RCPs are removed from service, is considered impractical due to the necessity of utilizing temporaiy test equipment inside containment.

Page 13 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS Alternate Test Frequency: Closure verification of these check valves shall be perfo1med during refueling outages when performing Appendix J Type C seat leakage testing. The defeITal oftest frequency to refueling outages is acceptable per the discussion provided in NUREG-1482, Rev.3, Section 4.1.6. These valves remain in the open position during power operation in suppmi ofnmmal seal water flow to the RCPs.

Sufficient flow through the valves is verified by observation of seal temperatures.

As an alternative to reverse exercising each refueling outage, these Category AC check valves may be placed in the Check Valve Condition Monitming (CVCM)

Program and reverse exercised at the Appendix J, Option B frequency during the perfo1mance of Type C seat leakage testing. This extension of frequency for reverse exercising is per the guidelines provided in NUREG-1482, Rev.3, Section 4.4.7, "Use of Appendix J, Option B, in Conjunction with ISTC Exercising Tests".

Page 14 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION - ROJ-09 System: Chemical and Volume Control Valve(s): 1(2)CV-00383 Catego1y: C Code Class: 1 Function: These check valves are located inside containment in the CVCS auxilimy charging line to the RCS loop B cold leg. 1(2)CV-383 perfonn an active function in the open direction to provide a relief path during a theimally induced overpressure condition of the containment penetration piping post-LOCA. There is no design flow rate associated with the safety function of 1(2)CV-383 in the open direction. 1(2)CV-383 perfo1m an active safety function in the closed direction. The valves serve as ASME Class 1 RCS pressure boundmy isolation valves as defined in 10CFR50.2. Therefore, 1(2)CV-383 must be capable of closure to maintain the integrity of the RCS pressure bounda1y in the event of a failure of upstream components.

Deferred Test Justification: The auxilimy charging line check valves are not provided with downstream isolation or test connections allowing the ability for closure verification by leak rate testing. The only practical means of verifying closure capability of these check valves, with the exception of disassembly, is by performing a radiographic examination test (RT) or other qualified test on the valve body to demonstrate the valve disk is in the closed position. Verifying closure capability of the auxiliary charging line check valves 1(2)CV-383 quarterly during power operation is impractical without providing a commensurate increase in the level of valve reliability due to the labor intensive nature of non-intmsive testing inside containment. Performing this type of test activity during cold shutdown is impractical from a logistics standpoint tests are perfmmed by an off-site contractor. During unplanned cold shutdowns the primmy concern is to safely restart the plant when the condition which required going to cold shutdown is corrected. Therefore, the coordination of outside contractor notification and the time required for equipment setup is impractical for the purpose of testing and could delay plant restart.

Page 15 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND 2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS Flow exercising these check valves quarterly during power operation or cold shutdown would require manual manipulation of valves to facilitate alignment to auxiliary charging in lieu of nonnal charging. This activity is impractical without providing a commensurate increase in the level of valve reliability. The safety function in the open direction is for the1mal overpressure protection which has no specific accident flow rate. Exercising the valves with high pressure discharge flow from the charging pumps does not necessarily demonstrate their capability to open as a thennal overpressure relief path. F mward exercising to the partially open position will be considered as being satisfied by demonstration of the valves ability to pass flow during the change in plant operating conditions from hot shutdown to cold shutdown.

Alternate Test Frequency: Verification of partial opening capability shall be demonstrated by the valves ability to pass flow when transitioning from hot shutdown to cold shutdown. This flow path is maintained isolated by a nmmally closed manual isolation valve outside containment. Closure verification of the CVCS auxiliary charging line check valves will be accomplished by perfo1ming a qualified test. This method of testing and frequency is supported by the discussion provided in NUREG-1482, Rev.3, Section 4.1.2. The auxiliary charging line check valves 1(2)CV-383 remain in the closed position during power operation.

As an alternative to performing qualified test each refueling outage, these check valves may be placed in the Check Valve Condition Monitoring (CVCM)

Program. If so, the qualified test shall be perfmmed on a sampling basis at the

frequency specified in the applicable CVCM Program Plan for the associated valve group.

Page 16 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION - ROJ-10 System: Component Cooling Water Valve(s): 1(2)CC-00745 Categ01y: C Code Class: 2 Function: These check valves are located in the CCW return header from equipment inside containment. The valves perfonn an active safety function in the closed direction to isolate CCW main header return flow from being directed to containment in the event of a CCW line break inside containment. Closure of the CCW return header check valves 1(2)CC-745 is required to preserve the integrity of the CCW system for continued heat removal capability from essential safety-related equipment. These check valves have no safety function in the open direction.

Their normally open position provides a path for CCW return flow from the RCP motor bearings, the RCP thermal barriers, and the excess letdown heat exchanger; however, this function is not classified as safety-related.

Page 17 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND 2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS Defen-ed Test Justification: Exercising these check valves to the closed position quarterly during power operation would require isolating the main CCW supply header to the containment. Isolating the CCW supply header to containment would intenupt CCW flow to the RCP motors bearings and thermal bairiers. The inte1111ption of cooling water flow to the RCPs could result in damage to the RCP motors and thennal ban-iers rendering the associated RCP inoperable. An inoperable RCP would require the plant to be placed in hot shutdown within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , as TS 3.4.4 requires both RCPs to be in operation when the reactor is critical. Exercising these valves to the closed position during cold shutdowns would require the removal of both RCPs (per unit) from service and verifying the absence of leakage at an upstream vent/drain connection while maintaining flow in the CCW return header downstream of 1(2)CC-745. Although possible, this method of testing is undesirable due to the persmmel risks associated with verifying the absence of leakage at an upstream vent/drain connection. Component cooling water contains the cmrnsion inhibitor potassium chromate. Potassium chromate poses a moderate health hazard and is a carcinogen. Therefore, personnel contact with potassium chromate should be minimized and should not be discharged to the environment. The prefen-ed method of verifying closure capability of 1(2)CC-745 is by performing a radiographic examination test (RT) or other qualified test on the valve body to demonstrate the valve disk is in the closed position. Due to the labor intensive nature of non-intrnsive testing, perfmming this type of test activity during cold shutdown is impractical from a logistics standpoint as tests are perfotmed by an off-site contractor. During unplanned cold shutdowns the primary concern is to safely restart the plant when the condition which required going to cold shutdown is corrected. Therefore, the coordination of outside contractor notification and the time required for equipment setup is impractical for the purpose of testing and could delay plant restart.

Alternate Test Frequency: Closure verification of the CCW return header check valves from containment will be accomplished by performing a qualified test each refueling outage. This method of testing and frequency is supported by the discussion provided in NUREG-1482, Rev.3, Section 4.1.2. These valves remain in the open position during no1mal power operation in support of nmmal seal water retmn flow from the RCPs. Sufficient flow through the valves is verified by observation of seal temperatures.

As an alternative to performing a qualified test, these check valves may be placed in the Check Valve Condition Monitoring (CVCM) Program. If so, the qualified test shall be performed on a sampling basis at the frequency specified in the applicable CVCM Program Plan for the associated valve group.

Page 18 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION - ROJ-11 System: Component Cooling Water Valve(s): 1(2)CC-00755A 1(2)CC-00755B 1(2)CC-767 Categmy: C Code Class: 2 Function: These check valves are located inside containment in the individual cooling water supply lines to the RCPs. The valves perfmm an active safety function in the closed direction to provide isolation of the low pressure CCW piping outside containment subsequent to a the1mal barrier cooling coil rnpture. These valves also perfo1m an active safety function in the partially open position.

1(2)CC-755A&B must be capable of opening to provide a vent path for relief of overpressurization due to thermal expansion during post-LOCA when the penetration is in an isolated condition. 1(2)CC-767 perform no safety function in the closed direction but are verified closed to satisfy bi-directional testing. The valves also perform an active safety function in the partially open direction to provide a relief path for the1mal overpressure protection.

Deferred Test Justification: The preferred method to verify reverse flow closure capability of these check valves is seat leakage testing to confirm disk position. The test connections utilized to perfonn seat leakage testing are located inside containment. Therefore, it would require containment entiy and the interruption of the valves' normal process functions in order to verify their closure capability. Exercising 1(2)CC-75 5A&B in the reverse direction requires intem1pting normal RCP cooling water flow and the use of temporary test equipment inside containment.

Such testing activities if performed during power operation could result in damage to the RCP motor and the1mal barrier rendering the associated RCP inoperable. An inoperable RCP would require the plant to be placed in hot standby within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , as TS 3 .4.4 requires both RCPs to be in operation when the reactor is critical.. Due to the considerable effmi associated with these test activities, reverse exercise testing during cold shutdown, if the RCPs are removed from service, is considered impractical due to the necessity of utilizing temporaiy test equipment inside containment.

Alternate Test Frequency: Closure verification of these check valves shall be performed during refueling outages. These valves remain in the open position during nmmal power operation in support of nonnal cooling water flow to the RCPs and continued flow through the excess letdown heat exchangers. Satisfactmy paiiial opening of 1(2)CC-00755A&B will be demonsti*ated by monitoring CC flow and temperature indication on the return line from the RCPs. Satisfactmy partial opening of 1(2)CC-767 will be performed during seat leakage testing.

Page 19 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST WSTIFICATIONS REFUELING OUTAGE TEST WSTIFICATION - ROJ-12 System: HVAC Valve(s): 1(2)RM-03200AA Category: AC Code Class: 2 Function: These rad monitor return check valves are located inside contaimnent in the return line to containment from radiation monitors RE-211 and RE-212. These valves are designated inboard containment isolation valves and as such, must be capable of closure on reversal of flow to maintain containment integrity. The valves have no safety function in the open position. RM-3200AA opens to provide a return path to containment during normal leak detection sampling activities.

Deferred Test Justification: These check valves provide a discharge path directly to the containment atmosphere from radiation monitors RE-211 and RE-212. The only method available to verify reverse flow closure capability of these check valves is by seat leakage testing. The test connections utilized to perform seat leakage testing are located inside containment. Therefore, it would require containment entty and the interruption of the valves' normal process functions in order to verify their closure capability. Exercising these check valves in the reverse direction requires the use of tempora1y test equipment inside containment. Due to the considerable effort associated with these test activities, reverse exercise testing during cold shutdown is considered impractical due to the necessity of utilizing temporary test equipment inside containment.

Alternate Test Frequency: Closure verification of these check valves shall be perfonned during refueling outages when performing Appendix J Type C seat leakage testing. The deferral of test frequency to refueling outages is acceptable per the discussion provided in NUREG-1482, Rev.3, Section 4.1.6. These valves will be exercised to the partial open position prior to leak testing.

Page 20 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST rusTIFICATIONS REFUELING OUTAGE TEST illSTIFICATION -ROJ-13 System: Reactor Coolant Valve(s): 1(2)RC-00528 Category: AC Code Class: 2 Function: These check valves are located inside containment in the nitrogen supply line to the pressurizer relief tank and serve as Class 2 to non-Code boundary barriers.

The valves perform an active safety function in the closed direction. They serve as the inside contaimnent automatic trip valves for the PRT nitrogen supply lines.

As such, they must be capable of closure on cessation or reversal of flow to maintain containment integrity. These valves perform no safety function in the open position. The process function in the open position to supply nitrogen to the PRT is non-safety related as the PRT is classed as non-Code and is not required for accident mitigation or to achieve/maintain the plant in a safe shutdown condition.

Defen-ed Test Justification: The only method available to verify reverse flow closure capability of these check valves is by seat leakage testing. The test connections utilized to perform seat leakage testing are located inside contaimnent. Therefore, it would require containment entry and the intenuption of the valves' normal process functions in order to verify their closure capability. Due to the considerable effort associated with these test activities, reverse exercise testing quarterly or during cold shutdown is considered impractical due to the necessity of utilizing temporary test equipment inside contaimnent.

Alternate Test Frequency: Closure verification of these check valves shalt'be perfo1med during refueling outages when perfo1ming Appendix J Type C seat leakage testing. The defetTal oftest frequency to refueling outages is acceptable per the discussion provided in NUREG-1482, Rev.3, Section 4.1.6. These valves will be exercised to the partial open position prior to leak testing.

Page 21 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND 2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION - ROJ-14 System: Reactor Coolant Valve(s): 1(2)RC-00529 Catego1y: AC Code Class: 2 Function: These check valves are located inside containment in the reactor makeup water supply line to the pressurizer relief tanks and serve as Class 2 to non-Code boundaiy ban-iers. The valves perform an active safety function in the closed direction. 1(2)RC-00529 serve as inside containment automatic trip valves for the PRT fill line from RMW. As such, they must be capable of closure on cessation or reversal of flow to maintain containment integrity. These valves also perform an active safety function in the partially open position. 1(2)RC-529 must be capable of opening to provide a vent path for relief of overpressurization due to thermal expansion during post LOCA when the penetration is in an isolated condition.

Defen-ed Test Justification: The only method available to verify partially open and reverse flow closure capability of these check valves is during the performance of seat leakage testing.

The test connections utilized to perform seat leakage testing are located inside containment. Therefore, it would require containment entiy and the interruption of the valves' normal process functions in order to verify their closure capability.

As a result of the considerable effort associated with these test activities, reverse exercise testing quaiterly or during cold shutdown is considered impractical due to the necessity of utilizing temporary test equipment inside containment.

Alternate Test Frequency: Closure verification of these check valves shall be performed during refueling outages when performing Appendix J Type C seat leakage testing. The partial opening capability shall be verified by venting the LLRT volume via the check valve. The defen-al oftest frequency to refueling outages is acceptable per the discussion provided in NUREG-1482, Rev.3, Section 4.1.6.

Page 22 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION -ROJ-15 System: Service Water Valve(s): 0SW-00112A 0SW-00135A Categ01y: C Code Class: 3 Function: These check valves are located in the service water supply lines to the steam driven AFW pumps and turbines. The valves perform an active safety function in the open direction. The turbine driven AFW pump is dependent upon bearing cooling water to support long te1m operation of both the pump and turbine subsequent to a design basis accident. They must be capable of opening to provide a path for cooling water flow to the bearings whenever cooling water supply valve 1(2)MS-2090 opens. These valves perform no safety function in the closed direction. The fire water system is also capable of supplying bearing cooling water to the TDAFWP and ties-in immediately downstream of these valves. Therefore, they would be required to close to prevent diversion of TDAFWP bearing cooling water when being supplied by the fire water system.

However, the ability for the fire water system to supply bearing cooling to the TDAFWP is not a safety-related function.

Defe1Ted Test Justification: Full-stroke exercising with flow is impractical since flow indication is not provided. Additionally, crediting full stroke capability by monitoring the components' temperature parameters is impractical due to the amount of time required for component operation. Calculations have demonstrated that an extended pump rnn (~42 minutes) would be required before pump/turbine bearing temperatures would exceed the maximum allowables. Therefore, partial stroke capability will be credited during quarterly pump testing by observation of the pump/turbine bearing temperatures.

Altemate Test Frequency: Full stroke capability of the valves will be verified during refueling outages by sample disassembly in accordance with the guidelines provided in the IST Program document. This method of testing and frequency is acceptable per the guidelines provided in ISTC-5221(c). These valves are verified in the partial open position during quarterly pump testing by observation of the pump/turbine bearing temperatures.

As an altemative to the disassembly and examination requirements of ISTC-5221(c), these check valves may be placed in the Check Valve Condition Monitoring (CVCM) Program and disassembled and examined per the requirements of Appendix II of the OM Code. If so, the frequency for disassembly and examination shall be specified in the CVCM Program Plan applicable to the valve group.

Page 23 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION - ROJ-16 System: Safety Injection Valve(s): 1(2)SI-00834D Catego1y: AC Code Class: 2 Function: These check valves are located inside containment in the nitrogen supply header to the SI accumulators. These valves perfonn an active safety function in the closed direction. 1(2)SI-834D are designated inboard isolation valves for containment penetration P-14C. As such, 1(2)SI-834D must be capable of closure on reversal of flow to maintain containment integrity.

Deferred Test Justification: The only method available to verify reverse flow closure capability of these check valves is by seat leakage testing. The test connections utilized to perfotm seat leakage testing are located inside containment. Therefore, it would require containment entty and the intenuption of the valves' normal process function in order to verify their closure capability. Exercising these check valves in the reverse direction requires defeating the ability to provide nitrogen makeup to the SI accumulator tanks and the use of temporary test equipment inside containment.

As a result of the considerable effort associated with these test activities, reverse exercise testing during power operation or cold shutdown is considered impractical due to the necessity of utilizing tempormy test equipment inside containment.

Alternate Test Frequency: Closure verification of these check valves shall be performed during refueling outages when perfo1ming Appendix J Type C seat leakage testing. The deferral of test frequency to refueling outages is acceptable per the discussion provided in NUREG-1482, Rev.3, Section 4.1.6. These valves will be exercised to the partial open position prior to leak testing ..

As an alternative to reverse exercising each refueling outage, these Category AC check valves may be placed in the Check Valve Condition Monitoring (CVCM)

Program and reverse exercised at the Appendix J, Option B frequency during the performance of Type C seat leakage testing. This extension of frequency for reverse exercising is per the guidelines provided in NUREG-1482, Rev.3, Section 4.4.7, "Use of Appendix J, Option B, in Conjunction with ISTC Exercising Tests".

Page 24 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION - ROJ-17 System: Safety Injection Valve(s): 1(2)SI-00845A 1(2)SI-00845B 1(2)SI-00845C 1(2)SI-00845D 1(2)SI-00~45E 1(2)SI-00845F Categmy: AC Code Class: 1 Function: These check valves are located inside containment in the high head safety injection flow path to the RCS Loop A and B cold legs and core deluge. The valves perform an active safety function in the open direction. The valves must be capable of opening subsequent to an SI system initiation to provide a path for post-LOCA high head safety injection and recirculation flow to the RCS for emergency and long term core cooling. SI injection via the cold legs is also credited for mitigating the consequences for a steam line break (SLB). High head safety injection can occur only when RCS pressure has been reduced to below the shutoff head of the SI pumps (3500 ft.). These valves also perform an active safety function in the closed direction. The valves must be capable of closure, if open, to prevent diversion of flow from other emergency core cooling systems as the reduction in RCS pressure allows SI accumulator discharge and subsequently low head safety injection. In addition, the valves are designated containment isolation valves and must be capable of closure on reversal of flow to maintain containment integrity. The valves also serve as RCS pressure isolation valves.

As such, valve closure is required to maintain the integrity of the RCS pressure boundaty and to isolate RCS pressure from the lower pressure SI piping and components.

Defen-ed Test Justification: Full stroke and partial stroke exercising these valves in the forward direction quarterly during power operation is not possible due to insufficient SI pump discharge head to overcome reactor pressure. Full stroke exercising these valves in the fmward direction during cold shutdown is precluded by restrictions related to LTOP concerns as discussed in TS 3.4.12. The valves will be partially exercised in the fmward direction during cold shutdown whenever leak testing per TS 3.4.14 is scheduled to ensure valve leak tight integrity is verified subsequent to closure. Partial stroke exercising without subsequent leak testing creates the potential for inter-system LOCA if the valves are not verified to be properly seated.

Alternate Test Frequency: Full stroke exercising in the fo1ward and reverse directions shall be perfmmed during refueling when sufficient time is available to demonstrate proper seating of the valves per the requirements of TS 3.4.14 and sufficient expansion volume exists in the RCS to accommodate the required flow rate.

Page 25 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST WSTIFICATIONS REFUELING OUTAGE TEST WSTIFICATION - ROJ-18 System: Contaimnent Spray Valve(s): 1(2)SI-00862A 1(2)SI-00862B Category: AC Code Class: 2 Function: These check valves are located in the CS pumps' discharge lines to the containment spray nozzles. The valves perform an active safety function in the open direction to provide a path for CS pump discharge flow to the spray nozzles during post-LOCA conditions. This function serves to limit peak contaimnent pressure to less than the design pressure of 60 psig @ 286°F and removes airborne radioactive iodine from the contaimnent atmosphere minimizing the potential of exceeding the offsite dose limits specified in 10CFRl 00. These valves also perfonn an active safety function in the closed direction.

1(2)SI-862A&B are designated outboard isolation valves for contaimnent penetrations P-54 and P-55. As contaimnent isolation valves, they must be capable of automatic closure on cessation of flow to maintain contaimnent integrity.

Deferred Test Justification: The only practical means of verifying valve closure capability is by perf01ming a seat leakage test. Performing this type of test quarterly during power operation would require isolating the associated CS header to the spray nozzles and utilizing an outside pressure source. Verifying valve closure capability with flow during quarterly pump testing would require cross connecting the discharge headers with both trains isolated from the contaimnent spray nozzles by closing manual valves 1(2)SI-868A&B. This aligmnent would render both trains of CS inoperable.

Exercise testing during cold shutdown is impractical due to the necessity of utilizing an outside pressure source or diagnostic testing both of which require the use of temporary test equipment with the potential of delaying plant restart.

Alternate Test Frequency: Valve exercise testing in the closed direction shall be performed in conjunction with Type C seat leakage testing during refuelings. This deferral of testing frequency is further supported by Section 4.1.4 ofNUREG-1482. These valves are full stroke exercised in the forward direction during quarterly pump testing by utilizing a full flow test line.

As an alternative to reverse exercising each refueling outage, these Catego1y AC check valves may be placed in the Check Valve Condition Monitoring (CVCM)

Program and reverse exercised at the Appendix J, Option B frequency dilling the perfo1mance of Type C seat leakage testing. This extension of frequency for reverse exercising is per the guidelines provided in NUREG-1482, Rev.3, Section 4.4.7, "Use of Appendix J, Option B, in Conjunction with ISTC Exercising Tests".

Page 26 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST msTIFICATIONS REFUEL OUTAGE TEST msTIFICATION -ROJ-19 System: Contaimnent Spray Valve(s): 1(2)SI-00847A 1(2)SI-00847B Category: C Code Class: 2 Function: These check valves are located in the spray additive lines from the SAT to the CS pumps' suction. They perform an active safety :function in the open direction.

1(2)SI-847A&B must be capable of opening, subsequent to the upstream AOVs opening, to provide a flow path for NaOH to be directed to the CS pump suction.

The addition ofNaOH to the spray stream is required for the removal of fission products released into the contaimnent atmosphere following a LOCA. These valves also perform an active safety :function in the closed direction. They must be capable of closure on reversal of flow to provide train separation during the event of containment spray pump operation with the SAT isolated. Additionally, the normally closed position prevents communication between the RWST supply and the NaOH piping which prevents inadve1ient dilution of the 30% weight NaOH contained in the SAT.

Defe1Ted Test Justification: During quarterly pump testing the contaimnent spray pumps are tun utilizing a full flow test line which recirculates flow back to the RWST to prevent wetting the containment. This test does not verify closure of 1(2)SI-847A&B, even though there is no flow through these checks, due to the eductors creating a low pressure area downstream of the check valves. Cold shutdown testing in the open direction passes RWST water through these checks. Since the check valves are in parallel paths, open testing of one check does not verify closure of the opposite train check. The only method available to verify reverse flow closure capability of these check valves is by seat leakage testing. Due to the considerable effort associated with these test activities, reverse exercise testing during cold shutdown is considered impractical without providing a commensurate increase in the level of valve reliability.

Alternate Test Frequency: Valve closure capability shall be verified during refuel outages by performing a seat leakage test. Open testing will be performed during cold shutdowns with justification provided in CSJ-11.

Page 27 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND 2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION - ROJ-20 System: Safety Injection Valve(s): 1(2)SI-00889A 1(2)SI-00889B Categmy: C Code Class: 2 Function: These check valves are located in the discharge lines from the SI pumps to the cold leg loop A & B injection lines and the core deluge loop A & B injection lines. The valves perfonn an active safety function in the open direction.

1(2)SI-889A&B must be capable of opening subsequent to the associated pump starting to provide a flow path for SI injection and recirculation post-accident to the RCS for emergency core cooling. This function is required to mitigate the consequences of a small break LOCA and to maintain shutdown margins subsequent to a SLB. These valves also perform an active safety function in the closed direction. The SI trains are nonnally aligned to maintain 100%

redundancy without reliance on cross-tie capability. In this normal alignment configuration, failure of a SI pump discharge check valve to close, subsequent to a train failure or removal from service, would not compromise the ability of the operating train to accomplish its design safety function. However, SI train cross-tie capability is provided for operational flexibility and to satisfy single failure in the event the SI pumps are required to provide flow to the core deluge lines to prevent boron precipitation thereby maintaining shutdown margins.

Cross-connecting the discharge lines of the SI pumps would require the pump discharge check valves to be capable of closure to prevent diversion of flow from the rnnning pump through an idle pump. The conditions constituting this alignment configuration would not require closure capability of SI-889A on reversal of flow as cross tying the loops is required to align P-15A for core deluge subsequent to a loss of P-15B. However, closure capability will be verified as an augmented test requirement for good engineering judgment.

Page 28 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST msTIFICATIONS Deferred Test Justification: Exercising these valves in the reverse direction quarterly during the inservice testing of the adjacent pump would require opening the cross-tie manual isolation valves which renders both trains of SI inoperable. In addition, reverse exercising these check valves with flow by allowing the discharge of an operating pump to communicate with the non-operating pump',s discharge check valve could result in overpressurizing the pump's suction pun:rping. This testing alignment would require cross-connecting A and B trains, tnen isolating the suction and discharge of the non-operating pump. This isolated boundary is not provided with overpressure protection. If the check valve being tested had significant leakage, the suction piping could become overpressurized by the high pressure discharge from the operating pump. The prefen-ed method of reverse exercising these check valves is by utilizing an outside pressure source to establish a ~p across the valve seat. However, this method of testing has been performed resulting in the inability to positively conclude that the valves are in their shut position.

Due to the inability to positively conclude that the valves are in their shut position when utilizing temporary test equipment, verification of reverse flow closure capability of these check valves is best demonstrated by performing a radiographic examination test (RT) or other qualified test. The qualified test shall be perf01med on the valve body to demonstrate the valve disk is in the closed position. Due to the labor intensive nature of non-intrusive testing, performing this type of test activity during cold shutdown is impractical from a logistics standpoint as tests are perfonned by an off-site contractor. During unplanned cold shutdowns the primmy concern is to safely restart the plant when the condition which required going to cold shutdown is c01Tected. Therefore, the coordination of outside contractor notification and the time required for equipment setup is impractical for the purpose of testing and could delay plant restart.

Alternate Test Frequency: Closure verification shall be accomplished by performing a qualified test during extended cold shutdowns or during each refueling outage. This defe1Tal of testing frequency is further supported by NUREG-1482, Rev.3, Section 4.1.2. These valves are full stroke exercised during the perfo1mance of quarterly pump testing.

As an alternative to performing a qualified test each refueling outage, these check valves maybe placed in the Check Valve Condition Monitoring (CVCM)

Program. If so, the qualified test shall be perfo1med on a sampling basis at the frequency specified in the applicable CVCM Program Plan for the associated valve group.

Page 29 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST msTIFICATIONS REFUELING OUTAGE TEST WSTIFICATION - ROJ-21 System: Safety Injection Valve(s): 1(2)SI-00867B Category: AC Code Class: 1 Function: These normally closed check valves are located in the safety injection line to the RCS Loop B cold legs from SI accumulators 1(2)T-34B, the SI pump discharge, and the return path for RHR shutdown cooling. The valves perform an active safety function in the open direction and must be capable of opening to provide a flow path to the RCS for injection of SI accumulator contents. 1(2)SI-867B also opens to allow high head safety injection/recirculation flow from the SI pumps.

Both functions are dependent upon a reduction in RCS pressure prior to safety injection. These valves also perform an active safety function in the closed direction. The valves serve as ASME Class 1 to Class 2 pressure boundaty isolation valves. As such, they perfmm a safety function to maintain the integrity of the RCS pressure boundaiy and to isolate RCS pressure from the lower pressure SI piping and components.

Deferred Test Justification: Exercising these valves to the full open or position quarterly during power operation is not possible due to the inability of overcoming RCS pressure. The accumulators are charged with a nitrogen blanket at 700-800 psig which is insufficient to inject accumulator inventmy into the RCS during nmmal operation for full exercising. Likewise, the SI pumps have a shutoff head of ~1500 psig which is also insufficient to overcome RCS pressure at power. To exercise these valves to their full open position at cold shutdown would require the injection of approximately 1000 ft3 of highly concentrated borated water into the RCS which could cause a low temperature overpressure condition due to insufficient expansion volume to accommodate the high flow rate.

Alternate Test Frequency: Included in the Check Valve Condition Monitoring Program, 1(2)SI-00867B will be full stroke exercised in the forward direction during refueling outages by directing RHR shutdown cooling flow through the valves, or by performing an accumulator dump test while simultaneously monitoring for disk to body impacts using nonintrusive test equipment demonstrating that the disk traveled to the full open position. Verification of valve closure capability shall be demonstrated by performing seat leakage testing per TS 3.4.14 during cold shutdown and or refueling.

Page 30 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND 2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION -ROJ-22 System: Safety Injection Valve(s): 1(2)SI-00867 A 1(2)SI-00842A 1(2)SI-00842B Category: AC Code Class: 1 Function: These normally closed check valves are located in the safety injection line to the RCS cold legs from SI accumulators 1(2)T-34A/B. 1(2)SI-00867A is also in the flow path for high head safety injection via the RCS loop A cold leg. The valves perform an active safety function in the open direction and must be capable of opening to provide a flow path to the RCS for injection of SI accumulator contents. 1(2)SI-867A also opens to provide a flow path for high head safety injection/recirculation flow from the SI pumps. Both functions are dependent upon a reduction in RCS pressure prior to safety injection. These valves also perform an active safety function in the closed direction. The valves serve as ASME Class 1 to Class 2 pressure boundaiy isolation valves. As such, they perform a safety function to maintain the integrity of the RCS pressure boundaiy and to isolate RCS pressure from the lower pressure SI piping and components.

Also, upon initiation of high head safety injection 1(2)SI-842A/B must close to prevent safety injection flow from being diverted to the SI accumulator in lieu of the loop A cold leg.

DefeITed Test Justification: Exercising these valves to the full open position quarterly during power operation is not possible due to the inability of overcoming RCS pressure. The accumulators are charged with a nitrogen blanket at 700-800 psig which is insufficient to inject accumulator inventoty into the RCS during normal operation for full exercising. Likewise, the SI pumps have a shutoff head of ~1500 psig which is also insufficient to overcome RCS pressure at power. Since the check valves are in parallel paths, open testing of one check does not verify closure of the opposite train check. The only method available to verify reverse flow closure capability of 1/2SI-867A check valve is by seat leakage testing. Due to the considerable effort associated with these test activities, reverse exercise testing during cold shutdown is considered impractical without providing a commensurate increase in the level of valve reliability.

Alternate Test Frequency: As Check Valve Condition Monitoring Program components, these valves will be full stroke exercised in the forward direction during refueling outages by perfonning an accumulator dump test while simultaneously monitoring for disk to body impacts using nonintrnsive test equipment demonstrating that the disk traveled to the full open position. Verification of 1(2)-SI-867A closure capability shall be demonstrated by performing seat leakage testing per TS 3.4.14 during cold shutdown and or refueling. 1(2)SI-842A/B shall be verified closed quarterly.

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POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION - ROJ-23 System: Service Water Valve(s): 0FP-00296A 0FP-00304A Categ01y: C Code Class: 3 Function: These check valves are located in the fire water pump P-35B supply line to the steam driven AFW pumps/turbines 1(2)P-29. The valves perform an active safety function in the closed direction. The fire water supply piping to the TDAFWPs ties into the service water bearing cooling supply piping downstream of check valves SW-135A (unit 1) and SW-l 12A (unit 2). Service water is the safety-related bearing cooling water supply during post accident conditions; therefore, FP-296A and FP-304A must be capable of closure to prevent diversion of TDAFWP bearing cooling water flow from the service water system to the non-safety-related, non-Code Class, Seismic Class 3 fire water system. These valves have no safety function in the open direction. The ability for the fire water system to supply bearing cooling water to the TDAFWPs is not a safety-related function. The fire water system has the capability of providing bearing cooling water to the TDAFWPs to ensure component operability subsequent to a station blackout. The diesel-driven fire water pump P-35B can provide cooling water during a station blackout independent of AC power, DC power, and instrnment air. However, this scenario assumes multiple failures of the emergency diesel generators which is beyond the single failure design basis of the plant.

Page 32 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND 2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST ruSTIFICATIONS DefeITed Test Justification: Verification of closure capability of these check valves quarterly during power operation is not possible due to the lack of vent/drain connections upstream. In addition, a pressure regulating device is located immediately upstream which functions to maintain the line in a charged condition at a lessor pressure than service water. To properly verify reverse flow closure c~pability of these check valves would require depressurizing upstream and dowri.~tream of the valves then venting residual pressure causing the pressure regulating device to fail open.

With the pressure regulating device in the open position, vent or drain connections are available to verify check valve closure subsequent to realigning service water downstream or by utilizing an outside pressure source. Due to the necessity of depressurizing the fire protection and SW header, verification of reverse flow closure capability of these check valves is prefe1Ted to be demonstrated by performing a radiographic examination test (RT) or other qualified test. The qualified test shall be performed on the valve body to demonstrate the valve disk is in the closed position. Due to the labor intensive nature of non-intrnsive testing, performing this type of test activity during cold shutdown is impractical from a logistics standpoint as RTs are perf01med by an _

off-site contractor. During unplanned cold shutdowns the primary concern is to safely restart the plant when the condition which required going to cold shutdown is corrected. Therefore, the coordination of outside contractor notification and the time required for equipment setup is impractical for the purpose of testing and could delay plant restart.

Alternate Test Frequency: Valve closme capability will be verified each refueling outage by perfo1ming a qualified test on the valve body to demonstrate disk closure. Exercising to the non-safety related open position shall be perfmmed annually during the periodic check which demonstrates fire water can be provided to the AFW pump/turbine.

As an alternative to performing a qualified test each refueling outage, these check valves may be placed in the Check Valve Condition Monitoring (CVCM)

Program. If so, the qualified test shall be perfonned on a sampling basis at the frequency specified in the applicable CVCM Program Plan for the associated valve group.

Page 33 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION - ROJ-24 System: Residual Heat Removal Valve(s): 1(2)SI-00854A 1(2)SI-00854B Category: C Code Class: 2 Function: These check valves are located in the individual supply line from the RWST to the suction ofRHRpumps P-lOA/B. The valves perform an active safety function in the open direction. They must be capable of opening subsequent to an auto pump staii to provide a flow path for borated water from the RWSTs to the suction of the RHR pump. This function is required for initiation of low head safety injection flow for emergency core cooling following a large break LOCA.

The valves also perform an active safety function in the closed direction. They must be capable of closure to maintain containment integrity since they are designated as closed system boundaiy valves. As designated interim closed system boundaiy valves, the check valves shall be subject to system leakage testing to assure their capability to prevent a containment bypass leakage path from the containment sump to the vented RWST post-LOCA in the Containment Leakage Rate Testing Program.

DefeITed Test Justification: The only practical means of verifying valve closure capability is by performing a seat leakage test. Performing this type of test quaiterly during power operation would require isolating the associated RHR pump and utilizing an outside pressure source. Exercise testing during cold shutdown is impractical due to the necessity of utilizing an outside pressure source or diagnostic testing both of which require the use of tempora1y test equipment with the potential of delaying plant restait.

Alternate Test Frequency: Valve exercise testing in the closed direction shall be performed in conjunction with system seat leakage testing during refuelings. This deferral of testing frequency is further supp01ted by NUREG-1482, Rev.3, Section 4.1.6 .. These valves are full stroke exercised in the f01ward direction during quarterly pump testing by utilizing a full flow test line.

Page 34 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION - ROJ-25 System: Safety Injection Valve(s): 1(2)SI-00875A 1(2)SI-00875B Categ01y: C Code Class: 2 Function: These check valves are located in a 3/4" branch line connecting the high head safety injection headers to the SI check valves' test line. The valves perform an active safety function in the partially open direction. 1(2)SI-875A&B must be capable of opening to provide a relief path for overpressure protection of the safety injection headers. The SI injection piping is attached to the RCS pressure boundary and the SI piping design pressure is less than RCS operating pressure.

The relief valves for the SI injection piping, 1(2)SI-887, are located downstream of 1(2)SI-875A&B. Therefore, 1(2)SI-875A&B must open to provide overpressure protection relief path. There is no required flow rate associated with its safety function in the open direction. These valves also perform an active safety function in the closed direction. 1(2)SI-875A&B are designated containment isolation valves for penetrations P-13 and P-27. As such, 1(2)SI-875A&B must be capable of closure on reversal of flow to maintain containment integrity. A closed system serves a the containment boundaiy barrier outside containment.

Defen-ed Test Justification: These simple check valves are located inside containment. Exercising these valves in the f 01ward direction requires alternating flow in the high head safety injection headers and diverting a portion of flow through the SI test line utilized for measuring seat leakage of the Event V check valves. Establishing this alignment requires manipulation of locked closed manual containment isolation valves inside and outside containment. Performing this testing activity quarterly during power operation would result in requiring manual operator action to restore containment integrity.

Page 35 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST msTIFICATIONS Verification of reverse :flow closure capability of these check valves is best demonstrated by perfonning a radiographic examination test (RT) or other qualified test. Individually establishing a differential pressure across the valve disk by the installation of tempora1y test equipment is impractical due to the availability of test connections and the multiple branch design configuration of the piping. The qualified test shall be perf01med on the valve body to demonstrate the valve disk is in the closed position. Due to the labor intensive nature of non-intrnsive testing, performing this type of test activity during cold shutdown is impractical from a logistics standpoint as tests are perf01med by an off-site contractor. During unplanned cold shutdowns the primaiy concern is to safely restart the plant when the condition which required going to cold shutdown is con-ected. Therefore, the coordination of outside contractor notification and the time required for equipment setup is impractical for the purpose of testing and could delay plant restart.

Alternate Test Frequency: Partial stroke exercising in the forward direction shall be perf01med during cold shutdowns when containment integrity is not required. Closure verification shall be accomplished by performing a qualified test during each refueling outage.

As an alternative to perf01ming a qualified test each refueling outage, these check valves may be placed in the Check Valve Condition Monitoring (CVCM)

Program. If so, the refueling test shall be perfonned on a sampling basis at the frequency specified in the applicable CVCM Program Plan for the associated valve group.

Page 36 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST mSTIFICATIONS REFUELING OUTAGE TEST msTIFICATION - ROJ-26 System: Safety Injection Valve(s): 1(2)SI-00891A 1(2)SI-00891B Categ01y: C Code Class: 2 Function: These check valves are located in the uninstrumented minimum flowi*ecirculation line for SI pumps 1P-15A&B. The valves perform an active safety function in the open direction to provide a return minimum flow path to the RWST. Opening capability of 1/2SI-891A&B prevents damage to the pump as a result of operating in low flow or dead-headed conditions. The SI pumps mitigate the effects of relatively small break LOCAs which are indicative of slow depressurization of the RCS. Subsequent to the SI pump starting, high head injection doesn't commence until RCS pressure has been reduced to below the shutoff head of the pump (3500 ft). During the time between the auto pump start and delivery of flow to the vessel, discharge flow is recirculated back to the RWST via 1SI-891A. These check valves have no safety function in the closed direction. Flow through the minimum flow recirculation line is limited by a restricting orifice. During the injection phase of SI post-LOCA, the recirculation return path to the RWST remains unisolated (1/2SI-897A&B remain open). Minimum flow recirculation does not prevent the SI pumps from providing the required flow to the reactor. In the event of a failure of one pump to start, the flow through the minimum flow line will not increase if the idle pump's minimum flow line check valve fails to close. The suction supply from the RWST also remains unisolated during the injection phase which precludes the possibility of overpressurizing the suction piping of the idle pump. During the recirculation phase of SI, the operating SI pump receives its suction from the RHR heat exchangers. When the SI system is aligned for recirculation, the minimum flow line AOV isolation valves, SI-897A&B, and the SI pump suction isolation valves, SI-896A&B, are closed to preclude communication of sump invent01y with the RWST. Additionally, EOP-1.3 to requires closure of manual valves SI-876A&B which are located immediately downstream of minimum flow check valves SI-891A&B. Therefore, the minimum flow check valves are not required to close when the SI system is aligned for recirculation.

DefeITed Test Justification: The only means possible for verifying full stroke opening capability of these check valves is by utilizing a tempora1y flow measuring device. Performing this type of test quarterly during power operation is impractical due to the labor intensive actions involved in the installation of the tempora1y test equipment which would not provide a compensating increase in the level of valve reliability.

Exercise testing during cold shutdown is impractical due to the necessity of utilizing temporaiy test equipment with the potential of delaying plant restart.

Page 37 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST msTIFICATIONS Due to the inability to positively conclude that the valves are in their shut position when exposing the downstream side of the check valve to operating pump discharge flow, verification of reverse flow closure capability of these check valves is best demonstrated by perfonning a radiographic examination test (RT) or other qualified test. Individually establishing a differential pressure across the valve disk by the installation of temporaiy test equipment is impractical due to the availability of test connections and the multiple branch design configuration of the piping. The refueling test shall be perfonned on the valve body to demonstrate the valve disk is in the closed position. Due to the labor intensive nature of non-intrnsive testing, performing this type of test activity during cold shutdown is impractical from a logistics standpoint as tests are performed by an off-site contractor. During unplanned cold shutdowns the primaiy concern is to safely restart the plant when the condition which required going to cold shutdown is c01Tected. Therefore, the coordination of outside contractor notification and the time required for equipment setup is impractical for the purpose of testing and could delay plant restart.

Alternate Test Frequency: Full stroke opening capability of these check valves shall be demonstrated each refueling outage by utilizing temporary ultrasonic flow instrnmentation. These check valves are exercised in the non-safety related reverse direction during refueling outages by the perfomiance of a qualified test.

As an alternative to perfonning an open exercise each refueling by utilizing temporary ultrasonic flow instrnmentation and perfonning a qualified test each refueling outage for closure verification, these check valves may be placed in the Check Valve Condition Monitoring (CVCM) Program. If so, the qualified testing shall be performed on a sampling basis at the frequency specified in the applicable CVCM Program Plan for the associated valve group.

Page 38 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST msTIFICATIONS REFUELING OUTAGE TEST WSTIFICATION - ROJ-27 System: Residual Heat Removal Valve(s): 1(2)RH-00710A 1(2)RH-0071 OB Category: C Code Class: 2 Function: These normally closed check valves are located in the discharge lines from RHR pumps to the heat exchangers. The valves perform an active safety function in the open direction. They must be capable of opening subsequent to the associated pump starting to provide a path for post-LOCA low head safety injection and recirculation flow to the RCS, and long tenn shutdown cooling to mitigate the consequences of a SGTR and MSLB. These valves perform an active safety function in the closed direction. They must be capable of closure on reversal of flow, if its associated pump is secured or unavailable, to maintain separation of the RHR trains when operating in the normal shutdown cooling mode. RHR nonnal shutdown cooling operation is credited for mitigating the consequences of SGTR and MSLB accidents. Therefore, closure of these check valves prevent diversion of flow from the discharge of the opposite train to the suction side of the idle train subsequent to a loss of pump or the removal of a pump from service.

Deferred Test Justification: Exercising these valves in the reverse direction by cross connecting RHR trains "A" and "B" during cold shutdowns has been performed resulting in the inability to positively conclude that the valves are in their shut position. Justification for this test methodology during cold shutdown was based on the impracticality that cross connecting the RHR trains for the purpose of testing during power operation requires the manipulation of various manual isolation valves which would compromise the ability for the system to accomplish its design safety function as credited in the accident analysis. The starting sequence of the RHR pumps and their related emergency power equipment is designed so that delive1y of the minimum required accident flow is achieved within ~23. 7 seconds after receipt of the actuation signal. Should excessive leakage occur through one of the check valves, the system would be unable to satisfy the required response time due to the amount of time required to close the manual isolation valves to re-establish separation between the trains.

Page 39 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST msTIFICATIONS Due to the inability to positively conclude that the valves are in their shut position when exposing the downstream side of the check valve to operating pump discharge flow, verification of reverse flow closure capability of these check valves is best demonstrated by perfmming a radiographic examination test (RT) or other qualified test. Individually establishing a differential pressure across the valve disk by the installation of temporary test equipment is impractical due to the availability of test connections and the multiple branch design configuration of the piping. A qualified test shall be perfmmed on the valve body to demonstrate the valve disk is in the closed position. Due to the labor intensive nature of non-intrusive testing, perfonning this type of test activity during cold shutdown is impractical from a logistics standpoint as tests are perfmmed by an off-site contractor. During unplaimed cold shutdowns the primary concern is to safely restart the plant when the condition which required going to cold shutdown is corrected. Therefore, the coordination of outside contractor notification and the time required for equipment setup is impractical for the purpose of testing and could delay plant restart.

Alternate Test Frequency: Closure verification shall be accomplished by perfmming a qualified test during extended cold shutdowns or during each refueling outage. This deferral of testing frequency is further supported byNUREG-1482, Rev.3, Section 4.1.2. These valves are full stroke exercised during quaiierly pump testing.

As an alternative to performing a qualified test each refueling outage, these check valves may be placed in the Check Valve Condition Monitoring (CVCM)

Program. If so, the qualified test shall be performed on a sampling basis at the frequency specified in the applicable CVCM Program Plan for the associated valve group.

Page 40 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST WSTIFICATIONS REFUELING OUTAGE TEST WSTIFICATION - ROJ-28 Not Used Page 41 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION - ROJ-29 System: Containment Spray Valve(s): 1(2)SI-00858A 1(2)SI-00858B Category: C Code Class: 2 Function: This nonnally closed check valve is located in the RWST line to the Containment Spray pump suction. These valves perform an active safety function in the open direction, providing a suction source from the RWST to the pumps to support post-LOCA Spray System operation. These check valves are not credited with performance of a safety-function in the closed position. These check valves are periodically leak tested. Closure of this check valve would provide added assurance for prevention of radioactive leakage to the atmosphere via the RWST vent subsequent to placing the CS system in service during recirculation. The use of containment spray is not anticipated during the recirculation phase of SI for the purpose of containment cooling. However, intermittent operation of the Spray System may be necessary to maintain the required pH levels in the containment sump.

Deferred Test Justification: The best method of confirming the closure of these check valves is in conjunction with the Augmented seat leakage test, performed during refuelings. Confirmation of obturator position requires the installation of a hydrostatic test pump, leakage is collected upstream of the check valve. During this evolution the associated CS pump is disabled to prevent operation without a suction source. Disabling a CS pump during power operation to conduct testing to confirm disk travel to a position not credited with a safety function is detrimental to overall safety. Based on the considerable effort associated with this activity, performance on a cold shutdown frequency is also impractical.

Alternate Test Frequency: Closure verification of these check valves shall be perf01med during refuel outages in conjunction with augmented seat leakage testing.

Page 42 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND 2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION - ROJ-30 Not Used Page 43 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION -ROJ-31 System: Service Water Valve(s): 1(2)SW-15A 1(2)SW-15B 1(2)SW-15C 1(2)SW-15D Categmy: C Code Class: 3 Function: These normally open check valves are located in the service water supply lines to containment fan coolers. They perform an active safety function in the open direction to provide a path for cooling water flow to the fan coolers and fan cooler motors. The containment fan coolers provide sufficient air recirculation flow to accomplish containment heat removal following a design basis LOCA or steam line break inside containment. Proper operation of the fan coolers, as credited in the containment integrity accident analysis, is dependent upon service water as the heat sink. These valves do not perform a safety function in the closed direction.

They are located within piping associated with various containment penetrations; however, in all cases upstream manual valves are credited as providing the outboard containment boundmy barrier function with a closed system design inside containment.

Defe1red Test Justification: Reverse exercising these check valves quarterly during power operation would require the intenuption of service water flow to the containment fan coolers and ent1y into contaimnent for the purpose of downstream manual isolation to facilitate the use of a temporary pressure or valve disassembly. As part of a designated safety-feature system, the containment fan coolers are required for containment heat removal following a design basis LOCA or steam line break inside containment. This activity would result in the associated fan cooler being unable to accomplish its design safety function. In addition, when considering single failure, the entire safety system could be rendered inoperable should the redundant train fail to start subsequent to an SI actuation signal. As a result of the considerable effort associated with these test activities, reverse exercise testing during cold shutdown is considered impractical due to the necessity of utilizing tempora1y test equipment inside containment or the significant amount of time associated with valve disassembly which could delay plant restart.

Page 44 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST msTIFICATIONS Alternate Test Frequency: Reverse exercising in the non-safety related reverse direction during refueling outages by sample disassembly in accordance with the guidelines provided in paragraph ISTC-5221(c) of the OM Code and the IST Program document. These valves are full stroke exercised quarterly.

As an alternative to the disassembly and examination requirements of ISTC-5221 (c), these check valves may be placed in the Check Valve Condition Monitoring (CVCM) Program and disassembled and examined per the requirements of Appendix II of the OM Code. If so, the frequency for disassembly and examination shall be specified in the CVCM Program Plan applicable to the valve group.

Page 45 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS REFUELING OUTAGE TEST JUSTIFICATION -ROJ-32 System: Auxiliaiy Feedwater Valve(s): 1(2)AF-00196 1(2)AF-00114 Categmy: C Code Class: 3 Function: These nmmally closed check valves ai*e located in AFW pumps' minimum flow recirculation line which returns pump discharge flow back to the CSTs. The valves perfonn a safety significant function in the open direction. These check valves open with flow subsequent to the opening of recirculation line air operated control valves. The control valves automatically open when indicated flow is less than that required to provide pump protection. They are provided with a backup instrument air accumulators or nitrogen bottles to ensure opening capability for pump protection from overheating when operating in low flow or dead headed conditions. Although highly unlikely, such a condition could occur subsequent to an accident or event if operations take manual control of AFW flow to control steam generator level. A reduction or cessation of flow without a minimum flow path available could result in overheating of the pump and subsequent failure. An overall reduction in risk and core damage frequency is achieved by having the capability to establish minimum flow when instrument air is unavailable. These valves do not perform safety function in the closed direction but are verified closed to satisfy bi-directional testing.

Deferred Test Justification: There are no test connections to enable closure verification of these check valves by leak rate testing and they are not provided with position indication. The only practical means of verifying closure capability of these check valves, with the exception of disassembly, is by performing a radiographic examination test (RT) or other qualified test on the valve body to demonstrate the valve disk is in the closed position. Due to the labor intensive nature of non-intrusive testing, performing this type of testing activity quarterly during power operation is impractical without providing a commensurate increase in the level of valve reliability. Perfmming this type of test activity during cold shutdown is impractical from a logistics standpoint as tests are performed by an off-site contractor. During unplanned cold shutdowns the primaiy concern is to safely restart the plant when the condition which required going to cold shutdown is corrected. Therefore, the coordination of outside contractor notification and the time required for equipment setup is impractical for the purpose of testing and could delay plant restart.

Page 46 of 47

POINT BEACH NUCLEAR PLANT IST APPENDIX D UNITS 1 AND2 Revision 3 INSERVICE TESTING PROGRAM REFUELING OUTAGE TEST JUSTIFICATIONS Alternate Test Frequency: Verification of the AFW pumps' minimum flow check valves to travel to their non-safety related closed position will be accomplished by performing a qualified test during each refueling outage. Check valve exercising in the open direction will be perfonned during quarterly pump testing.

As an alternative to performing a qualified test each refueling outage, these check valves may be placed in the Check Valve Condition Monitoring (CVCM)

Program. If so, the qualified test shall be perfonned on a sampling basis at the frequency specified in the applicable CVCM Program Plan for the associated valve group.

Page 47 of 47

POINT BEACH NUCLEAR PLANT APPENDIXE UNITS 1 AND2 Revision 6 INSERVICE TESTING PROGRAM TECHNICAL WSTIFICATIONS APPENDIXE TECHNICAL ruSTIFICATIONS TJ-01 1/2CS-466, -476 TJ-02 1/2CS-480, -481 TJ-03 Not Used TJ-04 0IA-1335, -1338, -1605, -1606 TJ-05 OIA-1301, -1302, -1418, -1419, -1870, -1876, -1883, -1889 TJ-06 Not Used TJ-07 1/2MS-2017A, -2018A Page 1 of 8

POINT BEACH NUCLEAR PLANT APPENDIXE UNITS 1 AND2 Revision 6 INSERVICE TESTING PROGRAM TECHNICAL mSTIFICATIONS TECHNICAL mSTIFICATION - TJ-01 System: Feedwater Valve(s): 1(2)CS-00466 1(2)CS-00476 Category: B Code Class: NC Function: These normally open, quality related air operated valves are located in the main feedwater supply header to the steam generators and serves as the feedwater flow control valves. The valves perform an active safety function in the closed position to isolate feedwater flow during a MSLB as the secondary isolation device in the event of Main Feedwater Isolation Valve (MFIV) failure to isolate.

This safety related function is permitted to be performed by a non-safety related component by NUREG-0800 Chapter 6, Section 2.1.4. Isolating feedwater flow subsequent to a MSLB decreases the blowdown rate from the steam line break which reduces cooling of the primary system and reduces the post-accident containment pressure by limiting the energy mass release to containment. The valves must be capable of automatic closure upon receipt of an SI signal which is indicative of conditions requiring feedwater isolation. 1(2)CS-466 and -476 will also auto close upon receipt of a low Tave signal coincident with a reactor trip to prevent overcooling the reactor which could result in a return to criticality. Auto closure will also occur upon receipt of a high steam generator level signal to prevent steam generator flooding. The later two automatic isolation signals are not required for accident mitigation and are classified as non-QA functions.

These valves have no safety function in the open position. During normal operation, the feedwater regulator control valves modulate to control the flow of feedwater to the steam generator in response to a control air signal supplied from the steam generator water level control circuitry. The feedwater regulator valve will also auto open, if closed, upon receipt of a high Tave signal with a reactor trip to supply feedwater as quicldy as possible to reduce the reactor coolant Tave to the no-load average temperature value. The automatic opening function is not required for accident mitigation and is classified as a non-QA function.

Deferred Test Justification: Exercising the feedwater flow control valves closed quarterly during power operation would result in a loss of normal feedwater flow to the associated Steam Generator. Isolation of nmmal feedwater flow during power operation could potentially cause a severe steam generator level transient which could result in a plant trip, and would initiate an auxiliary feedwater system actuation signal unnecessarily. This closure testing is considered augmented, since the valves are not ASME Class 1, 2, or 3.

Partial Stroke Testing: Partial stroke exercising will be perfmmed through nmmal operation of the valves in their modulating capacity.

Alternate Test Frequency: Stroke test (including stroke time measurement) and fail safe test to the closed position during cold shutdowns when feedwater is removed from service.

Page 2 of 8

POINT BEACH NUCLEAR PLANT APPENDIXE UNITS 1 AND2 Revision 6 INSERVICE TESTING PROGRAM TECHNICAL JUSTIFICATIONS TECHNICAL JUSTIFICATION - TJ-02 System: Feedwater Valve(s): 1(2)CS-00480 1(2 )CS-00481 Categ01y: B Code Class: NC Function: These n01mally closed air operated valves are located in the bypass lines around the feedwater regulator control valves. The bypass valves perfonn an active safety function in the closed position to isolate feedwater flow during a MSLB as the secondary isolation device in the event of MFIV failure to isolate. Isolating feedwater flow subsequent to a MSLB decreases the blowdown rate from the steam line break which reduces cooling of the primaiy system and reduces the post-accident containment pressure by limiting the energy mass release to containment. These valves must be capable of automatic closure, if open, upon receipt of an SI signal which is indicative of conditions requiring feedwater isolation. This isolation capability is redundant to the main feedwater, condensate, and heater drain tank pump trip circuit1y which actuates on receipt of a high containment pressure signal. However, this pump trip circuit1y is non-safety related and cannot be relied on for isolation of feedwater subsequent to a MSLB. The feedwater regulator control bypass valves will also auto close upon receipt of a high steam generator level signal to prevent steam generator flooding.

This automatic isolation signal is not required for accident mitigation and is classified as a non-QA function. These valves have no safety function in the open position.

Defened Test Justification: Exercising the feedwater regulator control bypass valves to the closed position quarterly during power operation could induce perturbations in n01mal feedwater flow possibly resulting in undesirable fluctuations in steam generator level. This closure testing is considered augmented, since the valves are not ASME Class 1, 2, or 3.

Partial Stroke Testing: Partial stroke exercising will be performed during startups through normal operation of the valves in their modulating capacity.

Alternate Test Frequency: Stroke test (including stroke time measurement) and fail safe test to the closed position during cold shutdowns when feedwater is removed from service.

Page 3 of 8

POINT BEACH NUCLEAR PLANT APPENDIXE UNITS 1 AND2 Revision 6 INSERVICE TESTING PROGRAM TECHNICAL WSTIFICATIONS TECHNICAL msTIFICATION - TJ-03 NOT USED Page 4 of 8

POINT BEACH NUCLEAR PLANT APPENDIXE UNITS 1 AND2 Revision 6 INSERVICE TESTING PROGRAM TECHNICAL IDSTIFICATIONS TECHNICAL JUSTIFICATION - TJ-04 System: Instrument Air Valve(s): 0IA-01335, 0IA-01338, 0IA-01605, 0IA-01606 Category: C Code Class: NC Function: These check valves are located in the non-Code class seismic Class 1 instrument air supply lines to the pressurizer PORVs. The valves have no s,afety significant function in the open direction. The ability of the PORVs to operate is not dependent upon instrument air to accomplish their design functions. The PORVs are provided with a backup nitrogen supply source to ensure continued operability during a loss of instrument air. These valves perfonn an active augmented safety significant function in the closed direction. They must be capable of closure to prevent diversion of the backup nitrogen to the instrument air system in lieu of being directed to the PORVs. Diversion of backup nitrogen to the instrument air system could compromise the ability of the PORV to accomplish its design functions.

Deferred Test Justification: To verify reverse flow closure of the check valves requires containment entry to isolate and depressurize the instrument air system piping immediately upstream of the checks and monitoring for leakage via an opened tubing connection. Because of the time required to implement the testing, and due to the extent of the test activities, perfmming closure verification qumierly during power operation is impractical without providing a compensating increase in the level of valve reliability. This closure test is considered to be an augmented test since the valves are not ASME Class 1; 2, or 3.

Alternate Test Frequency: Exercise to the closed and open (nonsafety related) position during cold shutdown.

Page 5 of 8

POINT BEACH NUCLEAR PLANT APPENDIXE UNITS 1 AND2 Revision 6 INSERVICE TESTING PROGRAM TECHNICAL JUSTIFICATIONS TECHNICAL JUSTIFICATION - TJ-05 System: Instrument Air Valve(s): 0IA-01301, 0IA-01302, 0IA-01418, OIA-01419, OIA-01870, 0IA-01876, OIA-01883, 0IA-01889 Category: C Code Class: NC Function: Valves IA-1301, -1302, -1418, and-1419 are normally closed check valves located in the non-Code class, seismic Class 1 nitrogen supply lines from the backup nitrogen bottles to the PORVs. The valves have no safety function in the closed direction. The IA check valves perform an augmented risk significant function in the open direction. They must be capable of opening to provide an unobstructed flow path for backup nitrogen to the PORVs. The PORVs may be aligned to the backup nitrogen bottle, subsequent to a loss of normal instrument air, in order to accomplish the following augmented safety functions:

depressurization during SGTR recove1y, depressurization when the safety injection (SI) pumps are utilized as an alternate means ofborating the RCS, or to provide low temperature overpressure protection (LTOP) when the RCS is in a low temperature water solid condition.

Valves IA-1870, -1876, -1883, and-1889 are normally closed check valves located in the non-Code class, seismic Class 1 nitrogen supply lines from the SI Accumulator nitrogen inlet supply header to the PORV nitrogen backup fixed bottles. The valves have no safety function in the open direction. The IA check valves perfmm an augmented risk significant function in the closed direction.

They must be capable of closure to prevent diversion of the backup nitrogen volume to the SI Accumulator nitrogen supply header. Diversion of backup nitrogen to the SI Accumulator nitrogen supply header could compromise the ability of the PORV to accomplish its design functions.

Deferred Test Justification: Exercising these check valves requires the PORVs to be cycled, which then requires the backup fixed nitrogen bottles to be replenished. Demonstrating the associated PORV's ability to change position within the required stroke time limitations when receiving actuating air from the nitrogen bottle satisfies full stroke exercise requirements for these check valves. Due to the possibility of the PORV to stick open or fail to seal tightly when reseated the PORVs will not be exercised at power. Additionally, GL 90-06 provides guidelines to not exercise the PORVs at power. The open tests are considered to be augmented tests since the valves are not ASME Class 1, 2, or 3.

Alternate Test Frequency: Exercise to the open position during cold shutdowns by stroking the PORVs with backup nitrogen as the 'air' supply. There is no accident flow rate associated with the valves' safety function in the open direction. Satisfactmy stroke time of the PORVs will verify full open capability.

Page 6 of 8

POINT BEACH NUCLEAR PLANT APPENDIXE UNITS 1 AND2 Revision 6 INSERVICE TESTING PROGRAM TECHNICAL JUSTIFICATIONS TECHNICAL JUSTIFICATION - TJ-06 NOT USED Page 7 of 8

POINT BEACH NUCLEAR PLANT APPENDIXE UNITS 1 AND2 Revision 6 INSERVICE TESTING PROGRAM TECHNICAL JUSTIFICATIONS TECHNICAL JUSTIFICATION - TJ-07 System: Main Steam Valve(s): 1(2)MS-02017A 1(2)MS-02018A Category: C Code Class: NC Function: These MS non-return check valves are located in non-Code class piping downstream of the MSSVs and upstream of the main steam cross connection.

The valves perform an active safety function in the closed direction. A steam line mpture upstream of the non-return valves would require valve closure to prevent umestricted blowdown of the unaffected steam generator. These valves have no safety fimction in the open direction. 1(2)MS-2017A and-2018A remain open during normal operation to allow steam flow from steam generators to the main turbine in support of power generation. This function is not required for accident mitigation and is not a safety-related function.

Deferred Test Justification: Exercising these valves in the closed direction during normal operation would require isolation of one line of steam flow to the turbine. Isolation of a main steam header would cause a severe pressure transient in the associated main steam line possibly resulting in a plant trip. Additionally, isolation of a main steam header at power could potentially result in challenging the set point of the main steam relief valves causing inadve1ient lifting. Reducing power level to perform testing without causing a transient would significantly impact plant operations and power production. This closure test is considered to be an augmented test since the valves are not ASME Class 1, 2, or 3.

Alternate Test Frequency: Exercise to the closed position during cold shutdown. The valves are in continuous use during power operation which will be considered satisfactory forward exercising.

Page 8 of 8

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS APPENDIXF TECHNICAL POSITIONS TP-01 Power Operated Relief Valve (PORV) Leak Testing TP-02 Not Used TP-03 Not Used TP-04 Not Used TP-05 Not Used TP-06 Not Used TP-07 Diesel Fuel Oil Transfer Pumps and Unloader Valves TP-08 l/2AF-4000, -4001 TP-09 Detennining Differential Pressure Instrnment Accuracy Associated With Inservice Testing TP-10 Determining Category A and B Classification for Valves Within the PBNP Inservice Testing Pro gram TP-11 Rounding Convention Used to Establish Inservice Testing Acceptance Criteria TP-12 Partial Stroking of Disassembly and Inspection Tested Check Valves TP-13 Testing Service Water Pump Discharge Check Valves TP-14 Preconditioning Evaluation for Stem Lubrication of Motor Operated Valves TP-15 Service Water Pump Evaluation of Unmeasured Recirculation Flow TP-16 Containment Spray Pump Evaluation of Unmeasured Recirculation Flow TP-17 Atmospheric Dump Valve Operability Requirements TP-18 Comprehensive Pump Testing Page 1 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND 2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS TECHNICAL POSITION -TP-01 System: Main Steam (MS) and Reactor Coolant Systems (RCS)

Valve(s): 1(2)MS-02015, 1(2)MS-02016, 1(2)RC-00430, 1(2)RC-00431C Categmy: B Code Class: 2 (MS) and 1 (RCS)

Function: 1(2)MS-02015 and 1(2)MS-02016 (Atmospheric Steam Dump Control Valves)

The normally closed, fail-closed Atmospheric Steam Dump Valves, (one per steam generator) are attached to their respective Main Steam Safety Valve header B outside Containment and discharges to the atmosphere. This valve pe1forms an ACTIVE safety function in the OPEN position to provide a means of depressurizing following a steam generator tube rnpture (SGTR) coincident with a loss of AC power and cooldown of the Reactor Coolant System (RCS) to Residual Heat Removal (RHR) ently conditions. The valves must be capable of opening by remote manual actuation of its respective controller or by local manual operation within the time period required by the accident analysis.

Additionally, in the event of a small break LOCA this valve would be opened to remove heat and reduce RCS pressure until safety injection or RHR system operational limits are achieved. However, as stated in FSAR Sections 6.2.2 and 10.1, the use of these Steam Dump Valves is not required to meet core cooling objectives in the event of a small break LOCA. There is no maximum design stroke time limit associated with the valve's safety function in the open position.

By design, lMS-2015 opens automatically when a predete1n1ined pressure setpoint is reached to prevent unnecessary operation of the Code Safety Valves and to limit Steam Generator pressure during pressure excursions. In addition, instructions are provided in EOP-3 to set the controller to 1050 psig allowing the valve to lift if the set pressure is reached during a SGTR. However, the ability of these valves to open when the predete1n1ined setpoint is reached is an enhancement to safety and would not preclude recove1y from a SGTR accident.

These valves also perfo1n1 an ACTIVE safety function in the CLOSED position.

Subsequent to opening during a SGTR event, the valves must be capable of closure to minimize the release of fission products to the environment to maintain offsite dose within 10CFRl00 limits. Its failure to reclose would be equivalent to a small steam line break enhancing the severity of the SGTR event. Additionally, the valves are located on piping associated with Containment Penetrations P-1 and P-2, per FSAR Fig. 5.2-2. These penetrations meets the Class 4 containment isolation criteria, as defined in FSAR 5 .2.2; whereas, a closed system is credited inside Containment with at least one isolation valve located outside Containment.

However, upstream manual isolation valves, 1(2)MS-227 and 1(2)MS-244, are credited in FSAR Fig. 5.2-2 Containment Isolation Valves as providing the outside Containment boundaty batrier function and are manually exercised within the scope of IST.

Page 2 of36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS Since the pneumatic control circuits and air supply is not designated as safety-related, valve operability is relying on Full Stroke Manual Exercise to open and close position once every refueling outage as specified in Tech Spec SR 3.7.4.1. Stroke test (including stroke time measurement) will be considered as Augmented.

1(2)RC-00430 and 1(2)RC-00431C Pressurizer PORVs Thesr,/*nonnally closed air operated valves function as the Pressurizer Power Operated Relief Valves (PORVs) and serve as a Class 1 to non-Code boundary barrier. The valves pe1form a safety significant function in the OPEN position to provide a means for quick depressurization of the RCS during a Steam Generator tube rnpture (SGTR). The use of the PORVs for depressurization during SGTR recove1y is the qualified backup for the normal and auxiliaiy Pressurizer Spray Valves and will be used only if those means of depressurizing are unavailable.

The PORVs also serve as overpressure protection devices. During power operation the Pressurizer PORVs will automatically open if the RCS pressure increases to 2335 psig. This is a process function which precludes a high pressure reactor trip by limiting Pressurizer pressure to a value below the reactor trip setpoint (2385 psig) during all design transients up to and including a 50 percent step load decrease with steam dump actuation. The safety related overpressure protection which prevents RCS pressure from exceeding 110 percent of design is provided by the Pressurizer Safety Valves. As part of the Overpressure Mitigating System (OMS), the PORVs perform an ACTIVE safety related function of providing Low Temperature Overpressure Protection (LTOP) when the RCS is in a low temperature water solid condition. The ASME Code,Section III, Non-mandat01y Appendix G, as invoked by 10CFR50, Appendix G, establishes RCS pressure and temperature limitations to ensure protection against non-ductile fracture. During LTOP operation, either of the two PORVs opening upon receipt of an LTOP actuation signal from the RCS wide range pressure channel of less than or equal to 440 psig provides assurance these limits are not exceeded. When LTOP operation is required each PORV is provided with a compressed nitrogen bottle as a backup actuating air supply.

The PORVs also perform an ACTIVE augmented safety function in the CLOSED position. They must be capable of closure by remote manual switch actuation, if open, to maintain RCS pressure bounda1y. This function minimizes the potential for a small break LOCA condition resulting in uncontrolled RCS discharge to the PRT and a loss of Pressurizer pressure control.

The PORVs do not perform any safety function (Ref. CR 97-1207) within the scope ofISTA-1100 other than pressure retaining boundaiy. The PORVs are included within the scope of the IST Program based on PBNP response to GL 90-06, and are tested in accordance with the recommendations ofNUREG-1482 Section 4.2.10 per GL 90-06 which requires full-stroke exercising, stroke timing, and fail-safe testing during each cold shutdown or refueling cycle. Since these PORVs perfo1m an ACTIVE safety significant function in the CLOSED position, Page 3 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS Mandatmy Appendix IV requirements for perfonnance assessment testing and position verification testing apply.

Technical Position: The intent of this Technical Position is to discuss the testing requirements specified in Mandatmy Appendix IV and Section ISTC, specifically leakage testing. Paragraph IV-3300(d) identifies leak testing as a test requirement for active AOVs (including pneumatically operated PORVs) to be performed in accordance with paragraph ISTC-3600 and Section ISTC-5000, as applicable.

The Atmospheric Steam Dump Valves are categorized as Categmy B active pneumatically operated PORVs with leak test requirements provided in paragraph IST-5112. Paragraph ISTC-5112 requires leak testing in accordance with Mandatory Appendix I. It is recognized in Section ISTC and Mandatmy Appendix I that leakage criteria is Owner specified based on functionality of the valve. PBNP has detennined that leakage criteria is not applicable to the subject valves. Should leakage occur upstream isolation is provided in all configurations and the upstream isolation capability is demonstrated by valve testing within the scope of the IST Program.

In addition, the Pressurizer PORVs are provided with temperature indication, 1(2)TIA-438, in the tailpiece which alarms ifleakage occurs at which point the motor operated block valves would be closed to isolate the leakage. The Main Steam Atmospheric Dump Valves are provided with manual isolation capability at the inlet which are also credited in the FSAR as a Contaimnent boundaiy bani er and are exercised each refueling. However, PBNP will perfmm thermography once each cycle on the tailpiece of the Main Steam Atmospheric Dump Valves to check for leakage.

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POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND 2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS TECHNICAL POSITION - TP-02 TP-02 NOT CURRENTLY IN USE Page 5 of36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS TECHNICAL POSITION - TP-03 TP-03 NOT CURRENTLY IN USE Page 6 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND 2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS TECHNICAL POSITION - TP-04 TP-04 NOT CURRENTLY IN USE Page 7 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS TECHNICAL POSITION - TP-05 TP-05 NOT CURRENTLY IN USE Page 8 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND 2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS TECHNICAL POSITION - TP-06 TP-06 NOT CURRENTLY IN USE Page 9 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS TECHNICAL POSITION - TP-07 System: Diesel Fuel Oil Pump(s) Diesel Fuel Oil Transfer Pumps (FOTP) 0P-206A 0P-206B 0P-207 A 0P-207B Category: B Code Class: 3 Function: The fuel oil transfer pumps (FOTPs) perform the safety-related function of transferring fuel oil from their storage tank to their respective day tank and are classified as Code Class 3. This function ensures a continuous fuel supply in support of long term operation of the emergency diesel generators during accident conditions. These positive displacement pumps have sufficient head and capacity to transfer fuel oil at six times the maximum rate of engine fuel consumption (205 gph).

The FOTP discharge unloader valves are pressure regulating/control devices located in the recirculation line from the discharge ofFOTP back to the EDG fuel oil storage tanks T-175A/B. G0l and G02 unloader valves, FO-3982A and -3983A, perfom1 an active safety significant function in the open position to provide overpressure protection for the transfer pumps and discharge piping. This is a backup function to that performed by the internal reliefs associated with individual FOTPs. When in the standby condition, motor operated inlet isolation valves FO-3930 and FO-3931 to day tank T-3 lA/B are maintained in the closed position and auto open when a low level indication is detected in the respective day tank. Simultaneous to FO-3930/3931 opening, the respective fuel oil transfer pump automatically starts. Should FO-3930 or FO-3931 fail to auto open in conjunction with the associated pump starting, the transfer pump and discharge piping could be overpressurized ifFO-3982A or FO-3983A did not provide a relief path back to the storage tank T-l 75A. These pressure regulating devices as well as the unloader valves associated with G03 and G04 have a designated setpoint of 26 psig and discharge back to associated storage tank. However, this 26 psig setpoint is supportive in maintaining a constant flow velocity and is not based on overpressure protection as the piping system has a design pressure rating of 150 psi. The unloader valves also perfonn a passive safety function in the closed position. FO-3982A&B and FO-3983A&B communicate directly with the associated pump's discharge piping as a backpressure regulator. Failure to properly function to maintain back pressure by premature lifting could allow a portion of pump discharge flow to be diverted back to storage tank in lieu of being properly directed to the respective day tank.

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POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS Technical Position: The intent of this technical position is to address the passive safety function of the unloader valves in the closed position and their impact on pump pe1formance. In addition, to the method of testing utilized to demonstrate the valves' ability to adequately perfom1 their passive safety function in the closed position. Two separate concerns shall be addressed. Those concerns being; 1) the diversion of pump discharge flow via the unloader valve during Inservice Testing resulting in the potential of masking pump ~egradation by not measuring total flow within the test hydraulic circuit, and 2) thi: diversion of pump discharge flow via the unloader valves in an amolmt sufficient to compromise the safety related function of the pump during an accident.

In regards to the first concern, Inservice Testing of the fuel oil transfer pumps is perf01med once each quarter in accordance with the requirements ofIT-14. To allow sufficient pump run time for IST data acquisition and to satisfy the Code required 2 minute run for flow stabilization, the FOTPs are aligned to an instrumented test loop recirculating back to the EDG fuel oil storage tank. IT-14 also provides instrnction to set FOTP discharge pressure at 25 psig as the fixed parameter with flow and vibration compared to baseline reference values. As a result of setting discharge pressure at 25 psig and the unloader valves having a set point of 26 psig there is reasonable assurance that leakage past the unloader valves will not result in masking FOTP degradation. In further support of this position the following reflects unloader valve accuracy as stated by the valve manufacturer. The accuracy of a particular device is typically determined by measuring the dead band and hysteresis. These values are then added together to come up with an overall accuracy for the device. Dead band is caused primarily by "play" in the linkage, and hysteria is caused primarily by packing friction. The unloader valves (Model 42C) are a direct-acting device with no linkage and no packing. The plug is attached directly to the diaphragm which in tum is connected directly to the spring. With no linkage or packing, dead band and hysteresis are essentially zero. Pressure is sensed by the diaphragm directly, via the sensing port. The sensing port is small (1/16") relative to the volume under the diaphragm, creating an over-damped system that eliminates any over shoot.

This feature, in combination with the spring stiffness, enables the Model 42C to have excellent response time and extremely quick settling time. These back pressure regulating devices are expected to demonstrate extremely accurate set points with minimal potential for drift.

As stated in the discussion pertaining to :function, the FOTPs are required to transfer fuel oil from the storage tank to the day tank at a rate well in excess of the maximum rate of engine fuel consumption (220 gph). This safety function supp01ts long-te1m engine operation during post accident conditions.

Page 11 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS The ability for the pumps' to accomplish their design safety function is not necessarily demonstrated during performance ofinservice Testing, as IST is intended to monitor for pump degradation. IT-14 utilizes a recirculation flow path which bypasses flow restriction orifices in the flow path to the day tanks associated with G03 and G04 and l" pipe in the flow path to the day tanks associated with GO 1 and G02. When the FOTPs are providing flow to the day tanks higher discharge pressures can be expected to the point of challenging the unloader valves' set point due to increased flow resistance. Monthly EDG surveillance testing is perfonned by TS-81 (G0l), TS-82 (G02), TS-83 (G03) and TS-84 (G04). During the performance of these surveillance tests the diesel fuel oil transfer system is verified operable as required by TS 3.8.3. Operability verification consists of allowing day tank inventory depletion to the low level actuation set point for auto starting the associated pump. The associated FOTP is allowed to remain in operation for replenishment of day tank inventmy to the high-level isolation set point resulting in cessation of pump operation. Day tank Hi-Lo level indications are recorded with a percentage range applied for acceptance criteria. Pump discharge flow rates are recorded during this evolution for G03 and G04 with acceptance criteria applied to recorded flow rate values.

However, due to the lack of flow instrumentation in the pumps' discharge lines to GO 1 and G02 day tanks, flow rate is measured at a cold shutdown frequency utilizing an ultrasonic flow measuring device. Recent testing performed on G02 with pump discharge flow directed to the day tank demonstrated a flow rate of 25 gpm and a pump discharge pressure of 49 psig. Although pump discharge pressure was significantly higher than the set point of the unloader valve (26 psig), pump discharge flow rate is more than adequate for the pump to accomplish its design safety function. This determination is based on maximum rate of engine fuel consumption (220 gph). Therefore, it is concluded that the safety function of GO 1 and G02 FOTPs, P-206A and P-207 A, will not be compromised as a result of FOTP operation at a higher discharge pressure than the set point of the unloader valve. Likewise, monthly testing performed on G04 with pump discharge flow directed to the day tank demonstrated a flow rate of 40 gpm and a pump discharge pressure of 25 .8 psig. A pump discharge pressure of 25. 8 psig would result in negligible leakage past the unloader valve and the unquestionable ability for the FOTPs to accomplish their design safety function. It should be noted that EDG surveillance testing accomplished by TS-81 (G0l), TS-82 (G02),

TS-83 (G03) and TS-84 (G04) is petfonned when the system is aligned to support long term engine operation dming post-accident conditions.

Based on the above discussion, it is concluded that diversion of pump discharge flow via the unloader valves resulting in the potential masking pump degradation during IST testing is not a concern. In addition, there is reasonable assurance that any diversion of pump discharge flow via the unloader valves will not compromise the ability for the FOTP to accomplish their design safety function.

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POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND 2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS TECHNICAL POSITION - TP-08 System: Auxiliary F eedwater Valve(s): 1(2)AF-04000 1(2)AF-04001 Categ01y: B Code Class: 2 Function: These no1mally throttled-open motor operated valves are located in the TDAFWP P-29 discharge lines to the steam generators HX-lA and lB. The valves perfonn a passive safety function in the throttled open position and are procedurally maintained at a designated throttled-open position during power operation. At this position a sufficient flow path is provided to allow the required accident flow rate of 275 gpm to each of the two steam generators with no time-sensitive positioning required. The percent open throttled position is accurately set by monitoring flow when establishing percent open as left position. These valves perfo1m an active safety function in the closed position. They must be capable of closure, by remote manual switch actuation, in the event of a MSLB to limit the amount of auxiliaiy feedwater supplied to a faulted steam generator and to maximize flow to the unfaulted steam generator. This closure capability is necessaiy to minimize RCS cooldown rate and containment pressure increase during a MSLB event and to isolate the affected steam generator during a SGTR event.

Code Requirement: Remote position indication shall be verified locally during inservice testing or maintenance activities. This local observation shall be supplemented by other indications such as the use of flow meters or other suitable instrnmentation to verify obturator position. These observations need not be concurrent. Where local position is not possible, other indications shall be used for verification of valve operation. (OM-Code, ISTC-3700, III-3300(e), 10 CPR 50.55a (b)(3)(xi))

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POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS Technical Position: The intent of the Code required position indication verification is to ensure valve position is accurately reflected by the remote position indicating device during all plant conditions. Further, augmented testing shall be performed, if practicable, to provide positive evidence that the attachment between the stem and disc is intact.

AF-4000 and AF-4001 are not provided with conventional remote position indicating devices in that they are not provided with position indicating lights.

However, control room panel C03 is provided with a dial which reflects percent of valve position in addition to flow instrumentation which indicates the amount of auxiliary feedwater flow provided to each steam generator. In no instance does operations utilize the percent of valve position dial for determining valve position, setting the valves' required throttled position or for quarterly stroke timing. Due to a minimum of 5% indication uncetiainty associated with the percent of valve position dial, the throttled position of the valves is set by observing flow while manipulating the "jog" control switch. The uncetiainty associated with the accuracy of the positioning dial is primarily due to inherent voltage variation with the power supply. This condition causes drifting in the percentage value reflecting valve position resulting in frequent test failures when verifying valve position indication when in fact the valve is functioning properly. The use of flow instrumentation when verifying valve position provides positive indication of valve position as well as demonstrating the lack of stem/disc separation. As a result of the inherent drifting of percent open indication, valve position indication verification shall be augmented with monitoring of flow rate. In addition, acceptance criteria for valve position indication shall allow for a +/- 15% error in the indicated reading. The use of flow instrumentation when determining valve position in conjunction with position indication verification is much more meaningful, provides a positive indication of position, and is the method utilized by control room operators for remote position indication of AF-4000 and AF-4001.

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POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS TECHNICAL POSITION - TP-09 System: Various Pumps: Various Code Class: Various Function: Determining Differential Pressure Instrument Accuracy Associated With Inservice Testing.

The purpose of this Technical Position is to document PBNP's position on the methodology for determining the accuracy of differential pressure readings on centrifugal pumps that use separate suction and discharge pressure gauges.

Code Requirements: ASME OM Code Subsection ISTB places requirements on the accuracy of instruments used during Inservice Testing. Table ISTB-3510-1, Required Instrument Accuracy, states that for a Group A or B test pressure and differential pressure instruments must have an accuracy of+/- 2 percent. For a comprehensive and preservice test the accuracy must be+/- 0.5 percent.

Paragraph ISTB-3510, General (Data Collection) states the following in part: (a)

Accuracy. Instrument accuracy shall be within the limits of Table ISTB-3510-1. If a parameter is dete1mined by analytical methods instead of measurement, then the deteimination shall meet the parameter accuracy requirement of Table ISTB-3510-1 (e.g., :flow rate dete1mination shall be accurate to within+/- 2 % of actual). For individual analog instruments, the required accuracy is percent of full scale. For digital instruments, the required accuracy is over the calibrated range.

For a combination of the instruments, the required accuracy is loop accuracy. (b)

Range (1) The full-scale range of each analog instrument shall be not greater than three times the reference value. (b) (2) Digital instruments shall be selected such that the reference value does not exceed 90% of the calibrated range of the instrument. (ISTB-3510)

Paragraph ISTB-3520(b), Differential Pressure states:" When dete1mining differential pressure across a pump, a differential pressure gage or a differential pressure transmitter that provides direct measurement of pressure difference or the difference between the pressure at a point in the inlet and the pressure at a point in the discharge pipe shall be used."

Table ISTB-3000-1, Inservice Test Parameters states that for centrifugal pumps, including vertical inline shaft pumps, differential pressure or :flow rate shall be measured or detennined. This is in contrast to positive displacement pumps where discharge pressure is set and :flow is measured or determined.

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POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS Technical Position: PBNP considers the determination of the differential pressure reading for centrifugal pumps that use separate suction and discharge gauges to fall under the requirements ofloop accuracy requirement as stated in ISTB-3510(a).

PBNP considers the key inputs into determining differential pressure to be the readability of the gauges and the calibrated accuracy of the gauges.

PBNP considers that gauges can be read to one half the minor division of the gauge.

PBNP considers the full-scale range of the differential pressure reading to be the full- scale range of the discharge pressure instrument because it is larger than the range of the suction gauge.

PBNP dete1mines the accuracy of the differential pressure reading to be the sum of the square root of the sum of the squares of the calibrated accuracy and the readability of the suction and discharge gauges. This result must be less than or equal to the requirements listed in Table ISTB-3510-1, Required Instrument Accuracy.

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POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND 2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS TECHNICAL POSITION -TP-10 System: Various Valve(s) Various Categmy: AorB Code Class: Various Function: Determining Category A and B Classification for Valves Within the PBNP Inservice Testing Program.

Code Requirement: ASME OM Code (ISTC-1300) requires valves within the scope of the Inservice Testing Program be placed in one or more of the following categories:

Categmy A: Valves for which seat leakage is limited to a specific maximum amount in the closed position for fulfillment of their required function(s).

Categmy B: Valves for which seat leakage in the closed position is inconsequential for fulfillment of the required function(s).

Categmy C: Valves that are self-actuating in response to some system characteristics, such as pressure (relief valves) or flow direction (check valves) for fulfillment of the required function(s).

Categmy D: Valves that are actuated by an energy source capable of only one operation, such as rnpture disks or explosively actuated valves.

The purpose of this Technical Position is to document PBNP's position on the methodology for determining the categorization (A or B) of valves within the scope the PBNP Inservice Testing Program in terms of seat leakage testing requirements.

Technical Position: PBNP classifies valves within the scope of the Inservice Testing Program as Categmy A if they meet the requirements for inclusion into one or more of the following Groups:

GROUP 1 Pressure Isolation Valves.

All valves tested as Pressure Isolation Valves (PIVs) under PBNP Technical Specification 3.4.14, RCS Pressure Isolation Valve (PIV) Leakage, and listed in PBNP Technical Requirements Manual (TRM) Chapter 4.16, Reactor Coolant System (RCS)

Pressure Isolation Valve (PIV) Leakage Program, shall be designated and tested as Categmy A or A/C valves.

Page 17 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS Group 2 Containment Isolation Valves.

Based on internal guidance in Position 10 of Appendix A ofNUREG-1482, Revision 0, all containment isolation valves (CIVs) that are included in the Appendix J Program were initially categorized in the IST Program as Categ01y A or A/C valves. Per ISTC-3620, the NRC has determined that the leakage test procedures and requirements for containment isolation valves specified in 10 CFR 50, Appendix J are equivalent to the requirements of the Inservice Testing requirements. Containment isolation valves with leakage requirements based on other functions are tested in accordance with paragraph ISTC-3630. Chapter 5.2, Containment Isolation System, of the PBNP FSAR lists the valves designated as CIVs.

PBNP's Containment Leakage Rate Testing (CLRT) Program Basis Document lists which CIVs are included in PBNP's Appendix J Program. Section 2.2.2.d of the PBNP CLRT explains that as part of PBNP's operating license application the NRC required PBNP to designate at least one CIV external to containment for penetrations that have a low probability of rupture during a LOCA. PBNP designated these valves as CIVs as listed in Chapter 5 .2 of the FSAR. However, as these valves are not required by the Emergency Operating Procedures (EOPs) to go shut either during or following a LOCA, they are not required to be Type B or C leak tested because the associated closed system boundaty does not constitute a potential primaty containment atmospheric pathway. As such, these CIVs do not have specific individual seat leakage criteria associated with them and they are not included in or tested under PBNP's Appendix J valve seat leakage testing program. Based on this, CIVs that meet this criteria are classified as Category B valves rather than Categ01y A. All CIVs included in Appendix J are classified as Category A valves within the Inservice Testing Program.

Group 3 - Valves with Individual Seat Leakage Limits or Valves in Systems with System/Train Leakage Limits.

Valves with individual leakage limits and valves in systems with system leakage limits may or may not require designation as Catego1y A valves. The following two criteria are used to determine the proper categorization of valves that are included in this group:

i) In many cases an Inservice Testing Source Document (such as a program document, test procedure, or calculation) may include a limit on the amount of leakage from a system or part of a system to a specified location while not assigning a limit for leakage through the particular valve being evaluated. Some systems are configured such that there may be many potential leakage paths, each containing one or more valves. If it is determined that a total system leakage limit, rather than a valve specified leakage limit, is appropriate, Category A does not apply and the valve is classified as Category B. Reference NPM 2004-0551 for additional supp01iing documentation for this criterion.

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POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND 2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS ii) The leakage limit specified for a valve in an Inservice Testing Program Source Document is a gross leakage limit. Gross leakage is that leakage through a valve that indicates that the valve is not closed. For purposes of this technical position, gross leakage is valve leakage that exceeds the following ASME OM Code limits, as listed in ISTC-3630(e): (1) for water valve 0.5D gal/min or 5 gal/min, whichever is less, where D equals nominal valve size in inches; (2) for air valves 7.5D standardYft3/day where D equals nominal valve size in inches. Specified seat leakage limits greater then these will be considered as gross leakage limits and the valve will be classified as category B. All valves are intended to obstruct flow when fully closed. It is reasonable to utilize the ASME OM Code pennissible leakage rates as criteria above which leakage may be considered to be gross that is, indicative of a valve that is not fully closed.

Any valve that meets the requirements for inclusion as Group 3, and which is not excluded from a categ01y A classification by the two criteria listed above, shall be classified as catego1y A in the PBNP Inservice Testing Program.

Valves which do not meet the specific requirements for classification as categ01y A based on the above listed Groups, but for which there is a technical basis for classifying as categ01y A may be classified as catego1y A at PBNP's desecration. The technical basis for this classification shall be documented in the IST Program Documents.

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POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS TECHNICAL POSITION -TP-11 System: Various Valve(s)/

Pump(s) All With IST Program.

Categmy: All Code Class: All Function: Rounding Convention Used to Establish Inservice Testing Acceptance Criteria.

Technical Position: When establishing new Inservice Testing Program acceptance criteria, conservative rounding will be practiced. When establishing stroke time acceptance criteria the lower limit will be rounded up to towards the reference value and the upper limit will be rounded down towards the reference value. When establishing pump hydraulic and vibration acceptance criteria, the established values will be rounded down towards the reference values.

Example 1: When establishing the acceptance criteria for a Motor Operated Valve with an open reference valve of 67.85 seconds the following guidelines will be used.

Lower Limit: 67.85

0.85 = 57.6725 (round up to 57.68 seconds)

Upper Limit: 67.85

1.15 = 78.0275 (round down to 78.02 seconds)

The open acceptance criteria for the valve would be 57.68 to 78.02 seconds.

Example 2: When establishing vibration acceptance criteria for a Pump with a reference valve of 0.123 ips the following guidelines will be used.

Lower Alert Limit: 0.083

2.5 = 0.2075 ips (round down to 0.207 ips)

Upper Alert Limit: 0.083

6.0 = 0.498 ips (round down to 0.498 ips)

Required Action Limit: 0.083

6.0 = 0.498 ips (round down to 0.498 ips)

The alert range for this vibration point would be ">0.207 to :s;0.498 ips". The required action limit would be ">0.498 ips". The acceptance crite1ia for the vibration point would be :s;0.498 ips.

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POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND 2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS TECHNICAL POSITION -TP-12 System: Various Valves: Check Valves Within the Scope of the PBNP IST Program.

Categmy: C and AC Code Class: Various Function: Partial Stroking of Disassembly and Inspection Tested Check Valves.

Code Requirement: ASME OM Code paragraph ISTC-5221(c)(4) states "Before return to service, valves that were disassembled for examination or that received maintenance that could affect their performance, shall be exercised full- or part-stroke, if practicable, with flow in accordance with para. ISTC-3520. Those valves shall also be tested for other requirements (e.g., closure verification or leak rate testing) before returning them to service."

Technical Position: For check valves that were opened and inspected to either meet ASME OM Code testing requirements or following valve maintenance, the following requirements apply:

1) If practicable, a full-stroke exercising of the check valve shall be performed with flow.

2) If a full-stroke exercise of the check valve is not practicable, a part-stroke exercise shall be performed with flow.

3) If either a full- or part-stroke of the check valve is practicable, a final open and closed full stroke exercise of the valve obturator immediately prior to installation of the valve bonnet will be performed to satisfy post maintenance testing requirements.

These requirements are in addition to, and not in-place of any other testing that is required following an open and inspect to either meet ASME OM Code testing requirements or following valve maintenance. For example, if a check valve was credited under PBNP's Appendix J testing Program, a seat leakage test would need to be performed in addition to the above discussed open exercise testing.

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POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND 2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS TECHNICAL POSITION -TP-13 System: Service Water Components: Service Water Pump Discharge Check Valves:

SW-00032A SW-00032B SW-00032C SW-00032D SW-00032E SW-00032F Category: C Code Class: 3 Function: The Service Water System is described in the PBNP FSAR and pe1forms a credited safety-related function post accident to remove heat from various loads and to maintain Containment temperature and pressure within allowable limits. The Service Water pumps support these safety functions by creating sufficient flow and pressure to meet the system's design basis requirements. The associated discharge check valves support this function by passing required flow through operating pumps and isolating back flow to the idle pumps.

Code Requirement: ASME OM Code, ISTB-3300, Reference Values, sub-paragraph (e)(l) states that reference values shall be established within +/-20 percent of Pllmp design flow rate during the perfmmance of comprehensive pump testing.

ASME OM Code, ISTC-5221, Valve Obturator Movement, sub-paragraph (a)(l) states that for check valves that have a safety function in both the open and closed directions they shall be exercised by initiating flow and observing that the obturator has traveled to either the full open position or to the position required to perfmm it intended function, and verify that on cessation or revisal of flow, the obturator has traveled to the seat.

Sub-paragraph (c) states that if ISTC-5221 (a) is impractical for certain check valves or if sufficient flow cannot be achieved or verified, a sample disassembly examination program shall be used to verify valve obturator movement.

Page 22 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS Technical Position: Each quarter a partial open test of the Service Water Pump Discharge Checks is performed during the Inservice Test of the associated pump. The ability of the Service Water pump being test to meet its Inservice testing acceptance criteria verifies the associated discharge check valve is not stuck in the closed position. It is impractical to test the discharge check valves to their full open position using flow without introducing undesirable perturbations to the operating units. As the Service Water Pumps and check valves support the operation of both units they cannot be tied to a specific ur;iit's refueling outage. ASME OM Code paragraph ISTC-5221(c)(3) requires that all valves in a family be disassembled and examined at least once eve1y 8 years. PBNP requires that each of the Service Water Pump discharge check valves be disassembled and examined on a nominal six year frequency. During this inspection the full stroke motion of the valve disc is verified, meeting the requirements of ASME OM Code, ISTC-5221 for open stroke testing.

Closure testing of the Service Water Pump Discharge Check Valves is accomplished quarterly by verifying that the pump being tested produces its required flow rate (meets its Inservice Testing Program acceptance criteria). This proves that the check valves on the idle pumps have moved to their closed seats. In addition, on a nominal 18 month frequency, an enhanced closure test is pe1formed on each check valve by determining back leakage through the check valve. This is done by measuring the flow produced by the test pump before and after closing the manual isolation valve downstream of an idle pump's discharge check valve and determining the difference in flow rate.

Page 23 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND 2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS TECHNICAL POSITION -TP-14 System: Various Valve(s) Various Motor Operated Valves.

NOTE: This Technical Position Excludes the Following Motor Operated Valves:

ICC-00754A 2CC-00754A 1CC-00754B 2CC-00754B 1CC-00759A 2CC-00759A 1CC-00759B CC-00759B Categmy: Various Code Class: Various Function: Preconditioning Evaluation for Stem Lubrication of Motor Operated Valves.

Evaluate the affect that valve stem cleaning and lubrication has on the stroke times of Motor Operated Valves that are stroke timed under the PBNP Inservice Testing Program.

Code Requirement: The ASME OM Code does not specifically discuss the possible affects on Motor Operated Valve performance resulting from stem cleaning and lubrication. Paragraph ISTC-3310, Effects of Valve or Actuator Replacement, Repair, and Maintenance on Reference Valves, of the ASME OM Code states in part: "When a valve or its control system has been replaced, repaired, or has undergone maintenance that could affect the valve's performance, a new reference value shall be determined or the previous value reconfirmed by an inservice test run before the time it is returned to service or immediately if not removed from service." Similarly, paragraph III-6450 states in part:

"Changes to stem lubrication procedures, including the lubricant type and application schedule, might impact the engineering evaluations ... and shall be evaluated." These statements requires the owner to determine if a maintenance activity could affect a valve's perfonnance.

The 2017 Edition of the ASME OM Code does not require stroke time testing ofMOVs.

However, an OM condition (10 CPR 50.55a (b)(3)(ii)(D)) for MOV testing requires in part: "Licensees shall verify that the stroke time ofMOVs specified in plant technical specifications satisfies the assumptions in the plant's safety analysis."

The NRC has provided the indust1y with guidance on preconditioning in a number of documents. NRC INFORMATION NOTICE 97-16: PRECONDITIONING OF PLANT STRUCTURES, SYSTEMS, AND COMPONENTS BEFORE ASME CODE INSERVICE TESTING OR TECHNICAL SPECIFICATION SURVEILLANCE TESTING, documents a number of examples of what the NRC considers unacceptable preconditioning. It provides the following example related to MOVs:

Page 24 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND 2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS Motor Operated Valves: NRC IR 50-335/96-11, 50-389/96-11 (Accession No. 9609170377) for the St. Lucie reactor facility identified that the four containment spray flow control valves (two for each unit) were being unacceptably preconditioned prior to surveillance testing. Specifically, the valve stems were being lub1icated prior to perfonning stroke time testing. The failure of the licensee's administrative procedures to ensure that these stroke time tests were perfonned under suitably controlled conditions was cited as a violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control."

Section 3.5, Preconditioning of Pumps and Valves, ofNRC NUREG-1482 (Revision 3),

Guidelines for Inservice Testing at Nuclear Power Plants, states in part: "If a preventive maintenance activity (such as valve stem lubrication or pump venting) periodically occurs prior to testing, the licensee might justify the acceptability of this infrequent preconditioning of a pump or valve if the licensee evaluates the effect of the activity on the overall ability to assess the operational readiness of the pump or valve, and to trend degradation in its performance. As noted in the inspection guidance, the licensee should have evaluated and documented the activity as acceptable preconditioning before perfonning the testing." The NUREG goes on to state: "In each instance of acceptable preconditioning, the NRC staff will expect the licensee to have available a documented evaluation of the preconditioning activity and a justification for continued confidence in the capability of the IST program to assess the operational readiness of the pump or valve. Generic evaluations may be acceptable as long as the evaluation bounds the conditions of the specific activity perfmmed on the SSC."

Based on the above, the NRC requires that utilities evaluate potential preconditions concerns to ensure that they are acceptable and that the evaluations be documented.

Technical Position: PBNP conducted a review of a sample of MOVs, included in the PBNP Inservice Testing Program that have had their stems lubricated. Eleven MOVs included in the PBNP Inservice Testing Program and which had their stems cleaned and lubricated were reviewed as a part of this evaluation. The MOVs reviewed contain a number of different valves sizes from a number of different plant systems. Graphs were produced (attached to NAMS EDMS file CR01365829/CA0115018904/MOV Lub Stem.pd+/-) for each valve reviewed that provides stroke time data, in both the open and closed directions, for Inservice Tests both before and after the valves had their stems cleaned and lubricated. The change in stroke time after the MOVs had had their stem cleaned and lubricated were compared with the variation in stroke times seen between strokes when no maintenance was performed on the valve. See attached graph that compares the variation in stroke times for these two conditions.

For the eleven MOVs a total of 28 post stem clean and lubrication strokes were reviewed. The average change in measured stroke time, stroke time before stem clean and lubrication compared to the stroke time after stem clean and lubrication, was found to be -0.031 seconds. The largest decrease in stroke time between the pre and post stem clean and lubrication stroke times was -0.46 seconds and the largest increase was 0.51 seconds. The change in stroke time after stem cleaning and lubrication was found Page 25 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS to be similar for both the open and closed stroke direction for the MOVs as can be seen in the table below.

Number of Occurrences of Number of Occurrences of Negative (orno) Change in Stroke Positive Change in Stroke Time. Time.

Closed Stroke 9 5 Open Stroke 9 5 While an average change in stroke time following a stem cleaning and lubrication of -0.031 seconds appears to show that stem cleaning and lubrication has no affect on a MOVs stroke time the more conservative method to evaluate the data is to look at the average absolute change in stroke time. The average change in absolute stroke time, stroke time before stem cleaning and lubrication compared to stroke time after stem cleaning and lubrication, was found to be 0.180 seconds.

For comparison, stroke times of four of the eleven MOVs were reviewed during normal Inservice Testing when maintenance was not perfonned. For each of these four MOVs a comparison was made between six open and six closed consecutive strokes to determine the "normal" variation that can be expected during Inservice Testing. The average change in stroke time for these 48 valve strokes was -0.016 seconds with an average change in absolute stroke time of 0.280 seconds. This means that the average absolute change in stroke time was greater during Inservice Testing without stem lubrication than it was following stem lubrication.

There are three basic groups of valves, with respect to stroke speed, included and the ASME OM Code 2004 Edition through 2006 Addenda. However, the 2017 Edition of the ASME OM Code does not provide stroke time requirements for MOVs and the referenced paragraphs herein are no longer in the code. Paragraph ISTC-5122, Stroke Time Acceptance Criteria for MOVs, of the ASME Code lists them as valves that stroke in greater than 10 seconds, valves that stroke in less than or equal to 10 seconds, and valves that stroke in less than 2 seconds. PBNP has no MOVs that stroke in less than 2 seconds and ve1y few that stroke in less than 10 seconds. This leaves the great majority of MOVs stroking in greater than 10 seconds.

Paragraph ISTC-5122(a) of the ASME OM Code 2004 Edition through 2006 Addenda states that electric motor-operated valves with a reference stroke times of greater than 10 seconds shall exhibit no more than +/-15 percent in stroke time when compared to the reference value. As a result, if a MOV had a reference value of 10 seconds the acceptance criteria for the valve would 8.5 to 11.5 seconds or 10 seconds +/-1.5 seconds.

When this 1.5 second allowance is compared to the 0.180 second absolute average change in stroke time seen after a stem clean and lubrication it is clear that stem lubrication has little effect on the valve's performance.

Human Factors and Variations in Stroke Timing of Valves -The stroke timing of a valve at PBNP is typically done from control switch actuation until the desired position is indicated on the control board. This process introduces variability into the stroke Page 26 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERYICE TESTING PROGRAM TECHNICAL POSITIONS timing process. First an Operator needs to actuate the valve's control switch and the stopwatch at as close to the same time as is possible. Second, the Operator must watch for the change in position indication lights and stop the stopwatch as soon has he senses the change.

Studies have found that it takes approximately 0.2 seconds (with a range of between 0.1 to 0.4 seconds) for a human to sense and react to an external stimulus. This fact is incorporated into the starting practices used during world class track and field events(2).

In such events the time between the actuation of the starting gun to the first kick against',}

the staiiing block is measured electronically. If an athlete pushes off on his statiing block in less than 0.1 seconds he is consider to have "anticipated" the firing of the gun which is not allowed and the athlete is considered to have false started. If a highly trained athlete can not react in less than 0.1 seconds it is reasonable to assume that it would take an Operator at least 0.2 seconds (with a variation of between 0.1 to 0.4 seconds) to react to the change in position indication lights and actuate the stopwatch.

There would also be some variability introduced into the stroke timing of valves associated with the need to start the stop watch as close as possible to the actuation of the valve's control switch. Based on this, a change in stroke time of 0.5 seconds could be just as likely related to variations in the stroke timing process as any actual change in valve performance. A review of the data provided above finds that the stem cleaning and lubrication of MOY has no statically relevant effect on the time a takes an MOY to travel to the open or closed position.

Conclusions - A review of NRC documents finds that a licensee is allowed to evaluate preconditioning activities to determine if they are acceptable. These evaluations must verify that the Inservice Testing Program will continue to be able to assess the operational readiness of the pump or valve being tested and that the preconditioning activity does not mask the degraded component performance or ensure that the component's acceptance criteria are met.

PBNP has determined, through a review of actual test data and input from evaluations perfotmed at other nuclear plants that stem cleaning and lubrication ofMOYs has no significant effect on the stroke time of MOYs included in the PBNP Inservice Testing Program. PBNP has concluded that periodic stern cleaning and lubrication, petforrned at a frequency greater than the frequency applied to Inservice Testing of the MOYs represents acceptable preconditioning and that no pre- or post- maintenance testing is required for this activity.

At PBNP the practice of stern lubrication is not performed for the purpose of ensuring that the MOY will meet its testing acceptance criteria when during Inservice Testing.

Rather, it is performed to replace well-performing lubrication far in advance of any degradation of the lubrication that would adversely affect valve operation. Based on the discussion above, it is concluded that the MOYs would not have failed their tests had stem lubrication not been perf01med, nor would stem lubrication mask the as-found condition of the valve. Stem lubrication is not routinely perfonned on the MOYs before testing, but rather is perfotmed more infrequently than as specified by the MOY preventive maintenance program. This preventive maintenance is not perfotmed only Page 27 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS for scheduling convenience. It is performed in a manner to help minimize unavailability of the associated system If PBNP chooses to clean and lubricate the stem of an MOV on a frequency equal to or more often than the frequency at which the MOV is stroke time tested under the Inservice Testing Program then a pre-maintenance stroke time test will be performed under the PBNP Inservice Testing Program.

NOTE: This evaluation has found one exception to this determination. This exception relates to MOV's that have none locking worm to worm gear rations. These include the 1/2CC-754A/B and 1/2CC-759A/B MOVs. While stem cleaning and lubrication would not affect the stroke time of these valves it may have an effect on the valve's seat leakage characteristics. If the stems of these valves are lubricated their closure torque may relax over time after the valve has been shut. This could reduce seat loading and therefore allow for increased seat leakage. Based on this, both a pre- and post- stroke time and seat leakage test must be perfonned on these eight MOVs if their stems are lubricated.

Page 28 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS TECHNICAL POSITION -TP-15 System: Service Water Components: Service Water Pumps P032A P-032B P032C P032D P-032E P032F Category: Group A Code Class: 3 Function: The Service Water System performs a safety related function post-accident to remove heat from various loads and to maintain containment temperature and pressure within allowable limits. The service water pumps supp01i these safety functions by creating sufficient flow and pressure to the system's design basis requirements.

Code Requirement: ASME OM Code, paragraph ISTB-3550, Flow Rate, states: When measuring flow rate, a rate or quantity meter shall be installed in the pump test circuit. If a meter does not indicate the flow rate directly, the record shall include the method used to reduce the data. Internal recirculated flow is not required to be measured. External recirculation flow is not required to be measured if it not practical to isolate, has a fixed resistance, and has been evaluated by the Owner to not have a substantial effect on the results of the test.

Technical Position: The design of the Service Water System is such that when testing the individual pumps (P-32A/B/C/D/E/F) under the Inservice Testing Program the total flow from the pump is not measured by the installed flow meters FI-4459A/B or FI-4460A/B.

This is because the branch connections to the Circulating Water (CW) Pump coolers is upstream of the flow instruments and cannot be isolated without potentially damaging the CW pumps. After providing cooling to the CW pumps the water is recirculated back to the lake. These branch lines to the CW Pump coolers are not instrnmented for flow.

The PBNP Service Water Flow Model does not specifically account for this flow to the CW pumps when setting Service Water Pump design basis perfo1mance limits.

The basis for this is that the branch lines to the CW pumps are not isolated during accident conditions and that they act with a fixed resistance. As a result the same amount of flow is expected to be diverted from the Service Water System during an accident as is diverted during Inservice Testing of the Service Water Pumps. The Service Water Flow Model assesses the required performance of the Service Water Pumps based on the flow indicated on the FI-4459A/B or FI-4460A/B flow indicators downstream of the branch lines to the CW Pumps. Therefore, if the Service Water pumps are proven to meet their design bases perfo1mance requirements during Page 29 of36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS quarterly Inservice Testing they will also meet their design bases perf01mance requirements under accident conditions.

Dming Inservice Testing of the Service Water Pumps the North and South Headers are isolated by closing the Header Isolation Valves (SW-2890 and SW-2891). When Inservice Testing the South Service Water Header Pumps (P-32A/B/C) some of the pump's output is diverted to the Unit 1 CW Pump(s). When Inservice Testing the North Service Water Header Pumps (P-32D/E/F) some of the pump's output is diverted to the Unit 2 CW Pump(s).

PBNP periodically installs Ultra Sonic (UT) flow meters on these branch lines to verify that the quantity of water deviated to the Unit 1 and Unit 2 CW pumps does not significantly change over time. This testing is done during the Comprehensive Pump Test in accordance with the IST Program. Previous flow testing was performed under AR 01890321-02 but was superseded when pump inlet strainers were replaced by EC 280199 and EC 279378. The change in flow to the CW pumps was minor in nature.

Overall, the flow changes have no substantial effect on the SW pump performance.

The diverted flow does not exceed 110 gpm and flow fluctuations vary only 15 gpm between one pump and two pump operation.

Branch lines to the CW Pumps operate as a fixed resistance. The flow diversion to the CW pumps is small when compared to the total flow of 4500 gpm of the SW system.

The even smaller variations in this flow result in a ve1y predictable and repeatable test condition for which to validate pump performance. The purpose of the IST Program is to detect component degradation. The diversion of the SW flow and the inherent flow variations are not sufficient to impede the ability of the IST Program to trend pump performance.

Conclusion:

It is not practical to isolate Service Water flow to the Circulating Water Pumps during quarterly Inservice Testing of the Service Water Pumps without damaging the Circulating Water Pumps. Periodic testing of flow to the Circulating Water Pumps verifies that it operates as a fixed resistance system. The above analysis verifies that not measuring flow to the Circulating Water Pumps dming Inservice Testing of the Service Water Pumps does not have a substantial effect on the results of the test. PBNP's methodology used to Inservice Test the Service Water pumps is readily duplicated and provides indication of pump perf01mance that can be used to trend degradation. In addition, it verifies that the Service Water pumps can meet their design basis perf01mance requirements as documented in PBNP's Service Water Flow Model. Based on this, Service Water flow to the Circulating Water Pumps is not required to be measured during quarterly Inservice Testing of the Service Water Pumps and is in compliance with ISTB-3550.

Page 30 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS TECHNICAL POSITION -TP-16 System: Safety Injection Components: Containment Spray Pumps 1P-014A 1P-014B 2P-014A 2P-014B Categ01y: Group B Code Class: 2 Function: The containment spray system provides containment cooling and removal of elemental iodine and particulates following a loss-of-coolant accident. The containment spray pumps support these safety functions by creating sufficient flow and pressure to meet the system's design basis requirements.

Code Requirement: ASME OM Code, paragraph ISTB-3550, Flow Rate, states: When measuring flow rate, a rate or quantity meter shall be installed in the pump test circuit. If a meter does not indicate the flow rate directly, the record shall include the method used to reduce the data. Internal recirculated flow is not required to be measured. External recirculation flow is not required to be measured if it not practical to isolate, has a fixed resistance, and has been evaluated by the Owner to not have a substantial effect on the results of the test.

Technical Position: The design of the Contaimnent Spray System is such that when testing the individual pumps (1/2P-14A/B) under the Inservice Testing Program the total flow from the pump is not measured by the installed flow meters 1/2FI-661. On the six inch discharge line of each of the Containment Spray Pumps there is a two inch line that includes an eductor (1/2Z-275A/B). After passing through the eductor the two inch line tees back into the six inch suction line associated with containment spray pump.

There are no isolation valves in the eductor recirculation loop itself and it has no flow instrumentation.

There are two branches off the two inch eductor lines on each Containment Spray Pump. The first is two inch line that connects to the RWST and the second is the two inch line from the Spray Additive Tank (l/2T-38). During Inservice Testing of the Containment Spray Pumps both of these two inch lines are isolated. In the case of the line to the RWST the 1/2SI-864C/D and 1/2SI-873A/B valves are shut. The eductor line from the Spray Additive Tan1c is isolated by shutting the 1/2SI-874A/B valves.

With these lines isolated the eductor line functions as a fixed resistance recirculation loop around the Contaimnent Spray Pump. Based on this, the flow through the eductor line is consistent at the pump's fixed reference value of 1200 gpm and has no substantial effect on the results of the Inservice Test.

Page 31 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS

Conclusion:

It is not possible to isolate flow through the Containment Spray Pump eductor recirculation loop. The above analysis verifies that the eductor recirculation loop operates as a fixed resistance system and that not measming flow through the loop does not have a substantial effect on the results of the test. PBNP' s methodology used to Inservice Test the Containment Spray Pumps is readily duplicated and provides indication of pump performance that can be used to trend degradation. In addition, it verifies that the Containment Spray Pumps can meet their design basis performance requirements. Based on this, eductor recirculation loop flow is not required to be measured during Inservice Testing of the Containment Spray Pumps and is in compliance with ISTB-3550.

Page 32 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS TECHNICAL POSITION - TP-17 System: Main Stearn Components: Atmospheric Dump Valves lMS-2015 lMS-2016 2MS-2015 2MS-2016 Categmy: B Code Class: 2 Function: The Atmospheric Dump Valves are nmmally closed, fail-closed atmospheric dump valves attached to the main steam safety valve headers outside containment and discharging to the atmosphere. These valves perform an active safety function in the open position to provide a means of depressurizing following a steam generator tube rnpture (SGTR) coincident with a loss of AC power and cooldown of the reactor coolant system to RHR ent1y conditions. They must be capable of opening by remote manual actuation of the respective controller (credited in the FSAR) or by local manual operation (TS required) within the time period required by the accident analysis. Additionally, in the event of a small break LOCA this valve would be opened to remove heat and reduce reactor coolant system pressure until safety injection of RHR system operational limits are achieved. However, use of these steam dumps is not required to meet core cooling objectives in the event of a small break LOCA as stated in FSAR Section 6.2.2 and 10.1.

There is no maximum design stroke time limit associated with the valve's safety function in the open position. By design, they open automatically when a predetermined pressure setpoint is reached to prevent unnecessaiy operation of the Code safety valves and to limit steam generator pressure during pressure excursions. In addition, instrnctions are provided in EOP-3 to set the controller to 1050 psig allowing the valve to lift if set pressure is reached during a SGTR.

However, the ability of this valve to open when the predete1mined setpoint is reached is an enhancement to safety and would not preclude recove1y from a SGTR accident.

This valve also performs an ACTIVE safety function in the CLOSED position.

Subsequent to opening during a SGTR event, the ADVs must be capable of closure to minimize the release of fission products to the environment to maintain offsite dose within 10CFRI 00 limits. Its failure to reclose would be equivalent to a small steam line break enhancing the severity of the SGTR event.

Additionally, the ADVs are located within piping associated with containment penetration P-1 and P-2, per FSAR Fig. 5.2-1 and 5.2-2. This penetration meets the Class 4 containment isolation criteria, as defined in FSAR 5 .2.2, whereas a closed system is credited inside containment with at least one isolation valve located outside containment. However, upstream manual isolation valves, Page 33 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS 1/2MS-227/244, are credited in FSAR Fig. 5.2-2 as providing the outside containment boundaty banfor fonction. Due to this manual valve remaining in the open position during power operation, the normally closed, atmospheric steam dump valves maintain pressure boundaty integrity of the main steam header. The ability to travel to the fail-safe closed position on a loss of air or electrical power is required to support their safety fonction in the closed position. The valves receive control power from the emergency bus during a loss of offsite power and have no maximum design stroke time limit associated with their safety fonction in the closed position.

Code Since the AD valves' pneumatic control circuits and air supply are not designated Requirement: as safety-related, valve operability is relying on Full Stroke Manual Exercise to the open and close position once eve1y refoeling outage as specified in Tech Spec SR 3.7.4.1. Exercise testing and stroke timing are considered as Augmented, and performed quarterly, as required for pneumatically operated valves.

FSAR safety analyses rely on remote operation of the ADVs, and PBNP will continue to stroke time the valves from the control room on a quarterly basis using their power operators to maintain a high level of reliability as a matter of good engineering practice.

Technical PBNP Technical Specification Bases 3.7.4, Atmospheric Dump Valves (ADVs)

Position: Flowpaths, states that the ADVs are operable when the ADVs are capable of being locally opened and closed. The ADVs are manually exercised once per refoeling outage in IT 310 and IT 315 using their handwheel. This testing is required by Technical Specification SR 3.7.4.1.

The ability to be remotely operated is used in the PBNP FSAR safety analysis, but is not required in order to meet core cooling objectives (ref: NRC SER 2011-0004 dated May 3, 2011, FSAR 6.2.2, 10.1, 14.3.1). The air system used in the remote positioner and the air operator of the ADV are not safety related. The Atmospheric Dump Valves are stroke time tested from the control room quarterly under IT 90 Train A, IT 90 Train B, IT 95 Train A, and IT 95 Train Busing their air operator under the PBNP Inservice Testing Program as a matter of good engineering practice, under the Augmented IST program. If they fail to stroke or fail to stroke within their expected time the procedures direct a manual stroke of the valve using their handwheel to confirm their ability to perf01m their safety-related fonction and operability.

Page 34 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS TECHNICAL POSITION-TP-18 System: Various Components: Various Pumps lP-029 2P-029 lP-OlOA 2P-010A lP-053 2P-053 lP-0l0B 2P-010B lP-01 lA 2P-011A 1P-015A 2P-015A lP-0llB 2P-011B 1P-015B 2P-015B 1P-014A 2P-014A 0P-012A 0P-032A 1P-014B 2P-014B 0P-012B 0P-032B 0P-206A 0P-207A 0P-032C OP-206B 0P-207B 0P-032D 0P-032E Categmy: Group A, Group B Code Class: 2, 3 Function: The pumps are included in the PBNP IST Program on the basis that the pumps are required to perform a specific function in shutting down the reactor to the safe shutdown condition, in maintaining the safe shutdown condition, or in mitigating the consequences of an accident.

Code Requirement: ASME OM Code, paragraph ISTB-1400(c), Owner's Responsibility, states: In addition to the requirements of para. ISTA-1500, the Owner's responsibility includes (c) establishing a comprehensive pump test flow rate for each pump.

ASME OM Code, paragraph ISTB-2000, Supplemental Definitions, defines the comprehensive pump test flow rate as: the flow rate established by the Owner that is effective for detecting mechanical and hydraulic degradation during subsequent testing. The best efficiency point, system flow rates, and any other plant-specific flow rates shall be considered.

Page 35 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX F UNITS 1 AND2 Revision 13 INSERVICE TESTING PROGRAM TECHNICAL POSITIONS Technical Position: The 2012 Edition of the ASME OM Code introduced a new definition for the comprehensive pump test flow rate. In previous editions, the comprehensive flow rate was established at+/- 20% of the pump design flow rate. As discussed in NUREG-1482 Section 5.1, the new definition was developed with the knowledge that some pumps cannot be tested at the required high flow rates because of system design limitations.

PBNP does not have system design limitations which prevent from performing the comprehensive pump test at+/- 20% of the pump design flow rate. The+/-

20% of the pump design flow rate is accepted for detecting mechanical and hydraulic degradation as degradation is more readily detectable as pump hydraulic degradation typically manifests first at higher flow rates. Because PBNP is able to achieve the+/- 20% flow rates, PBNP maintains the comprehensive pump test flow rate of+/- 20% of the pump design flow rate for pumps within the IST Program with the exception of Service Water pumps.

The lower limit for Service Water pumps is truncated from 20% of the pump design flow rate. In a letter from Bechtel Corporation (PBB-W-1404) dated November 5, 1968 it was documented that the impellers for the Service Water Pumps would be trimmed to obtain a head of 175' at a flow rate of 5320 gpm.

Replacement pumps from Sulzer (EC 272153, among others) have the same design point of 5320 gpm at 175' total head. PBNP considers this to be the design flow of the Service Water Pumps. Applying the +/-20% requirement to this 5320 gpm value results in an allowable test range of 4256 to 6384 gpm (reference NPM 2003-0524).

The Inservice Testing Program sets Service Water differential pressure at a fixed reference value of 78 psid and trends pump flow. The design acceptance criterion for the service water pumps set by the IST program is at least 4500 gpm of flow per pump at 78 psid. Calculation 96-0059 establishes the SW model pump curves by degrading the nominal pump curves until they run through 4500 gpm at 78 psid. The degraded pump curves are then used in subsequent service water system design requirement calculations. This ensures that the flows provided by the SW pumps are always higher than the degraded performance that was evaluated as sufficient in the SW design basis calculations. ASME OM Code IST acceptance criteria for lower flow limits is truncated to 4500 gpm as necessmy to ensure design requirements are met.

Page 36 of 36

POINT BEACH NUCLEAR PLANT IST APPENDIX H UNITS 1 AND2 Revision 0 INSERVICE TESTING PROGRAM SIGNIFICANT PROGRAM CHANGES APPENDIXH SIGNIFICANT PROGRAM CHANGES The following significant program changes were implemented for the Sixth Interval IST Program update:

Introduction of Mandatory Appendix III, "Preservice and Inservice Testing of Active Electric Motor-Operated Valve Assemblies in Water-Cooled Reactor Nuclear Power Plai;its",':,

o Refer to valve tables for associated valves

Introduction of Mandatory Appendix IV, 11 Preservice and Inservice Testing of Active Pneumatically Operated Valve Assemblies in Nuclear Reactor Power Plants" o Refer to valve tables for associated valves

Introduction of Mandatory Appendix V, 11 Pump Periodic Verification Test Program" o Refer to pump table for valves within the Mandatory Appendix V program

Deletion of augmented components which do not either 1) perform an ISTA-1100 function, or 2) belong to a testing commitment o Refer to IST Program Background Documents for additional infmmation

Impact to Relief Requests (Appendices A and B) o Deletion: VR-02

Impact to Cold Shutdown Test Justifications (Appendix C):

o Deletion: CSJ-03, CSJ-06, CSJ-09, CSJ-26, CSJ-29 o Modification: CSJ-04, CSJ-05, CSJ-07, CSJ-10, CSJ-13, CSJ-15, CSJ-19

Impact to Refueling Outage Test Justifications (Appendix D):

o Modification: ROJ-24 o Updated NUREG-1482 to Revision 3 throughout.

Impact to Technical Justifications (Appendix E):

o Deletion: TJ-03, TJ-08 o Modification: TJ-01, TJ-02

Impact to Technical Positions (Appendix F):

o Deletion: TP-02, TP-04, TP-05, TP-06 o Modification: TP-01, TP-07, TP-08, TP-13, TP-14, TP-18 (reused)

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NRC 2022-0032, Sixth 10-Year Interval Inservice Testing (1ST) Program Plan

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1.0 INTRODUCTION

5.0 REFERENCES

1.0 INTRODUCTION

5.0 REFERENCES

Description:

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Conclusion:

Conclusion:

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NRC 2022-0032, Sixth 10-Year Interval Inservice Testing (1ST) Program Plan

Text

1.0 INTRODUCTION

5.0 REFERENCES

1.0 INTRODUCTION

5.0 REFERENCES

Description:

Description:

Conclusion:

Conclusion:

Navigation menu

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