NLS2016012, Fifth Ten-Year Interval Inservice Testing Program

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Fifth Ten-Year Interval Inservice Testing Program
ML16078A128
Person / Time
Site: Cooper 
Issue date: 03/09/2016
From: Shaw J
Nebraska Public Power District (NPPD)
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2016012
Download: ML16078A128 (279)
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Text

{{#Wiki_filter:N!Nebraska Public Power District Always the're when you need us NLS2016012 March 9, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Fifth Ten-Year Interval Inservice Testing Program Cooper Nuclear Station, Docket No. 50-298, DPR-46

Dear Sir or Madam:

The purpose of this correspondence is to provide Nebraska Public Power District's Inservice Testing (IST) Program Plan for the Fifth Ten-Year Interval. Submittal of the plan is in accordance with the requirements of the American Society of Mechanical Engineers Code for Operations and Maintenance of Nuclear Power Plants, Subsection ISTA-3200(a),"Administrative Requirements." As documented within the IST Program, the relief requests included in the 1ST Program have been previously submitted to the Nuclear Regulatory Commission and approved for use. The Fifth Ten-Year Interval for Cooper Nuclear Station began on March 1, 2016 and concludes on February 28, 2026.Enclosure 1 to this letter contains the "Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves." Enclosure 2 contains the "Cooper Nuclear Station Fifth Interval Inservice Examination and Testing Program for Snubbers." There are no regualatory commitments contained in this letter.Should you have any questions regarding the information contained in this submittal, please contact me at (402) 825-2788.Sincerely, Licensin Manager-/dv COOPER0 NUCLEARSTATION P.O. Box 98 / Browrivi/Ie, NE 68321-0098 Telephone: (402) 825-3811 / Fax:-(402) 825-5211 www.nppd.comn NLS2016012 Page 2 of 2

Enclosures:

1. Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves 2. Cooper Nuclear Station Fifth Interval Inservice Examination and Testing Program for Snubbers cc: Regional Administrator w/enclosures USNRC -Region IV Cooper Project Manager w/enclosures USNRC -NRR Plant Licensing Branch IV-2 Senior Resident Inspector w/enclosures NPG Distribution wlo enclosures CNS Records w/enclosures ATTACHMENT 9.4 REGULATORY SUBMITTAL REVIEW Sheet I of 2 Letter #: NLS2016012 Response Due: 3/1/2016

Subject:

Fifth Ten-Year Interval IST Program Submittal to NRC Date Issued for Review: 2/22/2016 Correspondence Preparer I Phone #: David Van Der Kamp / x2904 Section I Letter Concurrence and Aareement to Perform Actions POSITION I NAME Action Signature/Date (concurrence, (sign, interoffice menmo, e-mail, or telecom)certification, etc.)EP&C Eng -Tom Robinson Validation LIC Spec -David Van Der Kemp Validation / ,¢EP&C Supv -Stan Domikaitis Concurrence Z./_6/ =EP&C Mgr -Troy Barker Concurrence _ /DOE -Dan Buman Concurrence ' _ -// // , COMMENTS Section II Correspondence Screening Does this letter contain commitments? If "yes," identify the commitments with due Yes LI dates in the submittal and in Section III. No [If "yes," and the due date is > 3 months, is tracking of interim milestones required? Yes LI No LI Does this letter contain any information or analyses of new safety issues performed at NRC Yes LI request or to satisfy a regulatory requirement? If "yes," reflect requirement to update the No [USAR in Section II1. __ ___Does this letter require any document changes (e.g., procedures, DBDs, USAR, TS Bases, Yes LI etc.), if approved? If "yes," indicate in Section III an action for the responsible No [department to revise the affected documents. (The Correspondence Preparer may indicate the specific documents requiring revision, if known or may initiate an action for review.)Does this letter contain information certified accurate? If "yes," identify the information Yes LI and document certification in an attachment. (Attachment 9.5 must be used.) No ____ ATTACHMENT 9.4 REGULATORY SUBMITTAL REVIEW Sheet 2 of 2 Does this letter require posting per 1OCFR1 9? If "yes," ensure posting after submittal. Yes LI No [Does this letter contain Safeguards Information? If "yes," do not scan to CNS intranet Yes LI and ensure handling per Procedure 1.2. No [Does this letter contain information to be withheld from public disclosure (e.g., Proprietary or Yes LI Non-Safeguards Security-Related Information)? If "yes," do not scan to CNS intranet No [and ensure appropriate marking and handling. ___Section III Actions and Commitments Required Actions/Tracking Numbers Due Date Responsible Dept.Note: Actions needed upon approval should be captured in the appropriate action tracking system None CommitmentslCommitment Numbers Due Date Responsible Dept.Note: Enter the commitments into the commitment management system.None Section IV Final Document Signoff for Submittal Correspondence Preparer David Van Der Kamp /v Final Submittal Review (optional) See F&F v L Responsible Department Head SeLte Director of Nuclear Safety N/A Assurance (as applicable) ATTACHMENT 9.4 REGULATORY SUBMITTAL REVIEW Sheet 1 of 2 Letter#: NLS2016012 Response Due: 3/1/2016

Subject:

Fifth Ten-Year Interval IST Program Submittal to NRC Date Issued for Review: 2/22/2016 Correspondence Preparer / Phone #: David Van Der Kamp / x2904 Section I Letter Concurrence and Aareement to Perform Actions POSITION / NAME Action Signature/Date (concurrence, (sign, interoffice memo, e-mail, or telecom)certification, etc.)EP&C Eng -Tom Robinson Validation , LIC Spec -David Van Der Kamp Validation EP&C Supv -Stan Domikaitis Concurrence z/. /EP&C Mgr -Troy Barker Concurrence DOE -Dan Buman Concurrence __ _// / , COMMENTS Section II Correspondence Screenin~Does this letter contain commitments? If "yes," identify the commitments with due Yes LI dates in the submittal and in Section III. No [If "yes," and the due date is > 3 months, is tracking of interim milestones required? Yes LI No LI Does this letter contain any information or analyses of new safety issues performed at NRC Yes LI request or to satisfy a regulatory requirement? If "yes," reflect requirement to update the No [USAR in Section IlL.Does this letter require any document changes (e.g., procedures, DBDs, USAR, TS Bases, Yes LI etc.), if approved? If "yes," indicate in Section III an action for the responsible No [department to revise the affected documents. (The Correspondence Preparer may indicate the specific documents requiring revision, if known or may initiate an action for review.)Does this letter contain information certified accurate? If "yes," identify the information Yes LI and document certification in an attachment. (Attachment 9.5 must be used.) No ____ .-.ATTACHMENT 9.4 REGULATORY SUBMITTAL REVIEW Sheet 2 of 2 Does this letter require posting per 1OCFR1 9? If "yes," ensure posting after submittal. Yes LI No [Does this letter contain Safeguards Information? If "yes," do not scan to CNS intranet Yes Ii and ensure handling per Procedure 1.2. No ____Does this letter contain information to be withheld from public disclosure (e.g., Proprietary or Yes LI Non-Safeguards Security-Related Information)? If "yes," do not scan to CNS intranet No [and ensure appropriate marking and handling. ___ __Section III Actions and Commitments Required Actions/Tracking Numbers Due Date Responsible Dept.Note: Actions needed upon approval should be captured in the appropriate action tracking system None CommitmentslCommitment Numbers Due Date Responsible Dept.Note: Enter the commitments into the commitment management system.__________ None Section IV Final Document Signoff for Submittal Correspondence Preparer David Van Der Kamp ( f Final Submittal Review (optional) See F&F .Responsible Department Head SeLte Director of Nuclear Safety N/A Assurance (as applicable) NLS2016012 Page 1 of 266 Enclosure 1 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Nebraska Public Power District Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Revision 0 Cooper Nuclear Station P.O. Box 98 Brownville, NE 68321-0098 Commercial Operation Date: July 1, 1974___________-_ d Date: _ _ _ __01#-/IST Engineer: Program Backup / Supervisor: Date: )j-________ Date: _-_/_____EP & C Manager:-I Cooper Nuclear Station Ffifh Interval Inserviee Testing Pro gram for Pumps and Valves TABLE OF CONTENTS SECTION

1.0 INTRODUCTION

1.1 Purpose 1.2 Scope 2.0 INSERVICE TESTING PLAN FOR PUMPS 2.1 Pump Inservice Testing Plan Description 2.2 Pump Plan Table Description 2.3 Measurement of Test Quantities 2.4 Allowable Ranges of Test Quantities 2.5 Instrument Accuracy 3.0 INSERVICE TESTING PLAN FOR VALVES 3.1 Valve Inservice Testing Plan Description 3.2 Valve Plan Table Description 4.0 ATTACHMENTS

1. System and P&ID Listing 2. Pump Relief Requests 3. Augmented Pump Relief Requests 4. Valve Relief Requests 5. Augmented Valve Relief Requests 6. General Relief Requests 7. Cold Shutdown Justifications
8. Refuel Outage Justifications
9. Technical Positions 10. Inservice Testing Pump Table 11. Inservice Testing Valve Table Revision 0 Pg Page 2 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves

1.0 INTRODUCTION

1.1 Purpose This document establishes the Cooper Nuclear Station (CNS) Fifth 120-Month Interval Inservice Testing (IST) Program requirements for the American Society of Mechanical Engineers (ASME)Code Class 1, 2, and 3 pumps and valves whose specific functions are required to either:* Shutdown the reactor to the safe shutdown condition,* Maintain the safe shutdown condition, and/or* To mitigate the consequences of an accident.This document will also establish the CNS Augmented Inservice Testing Program requirements for ASME non-Code Class pumps and valves whose specific functions will meet one or more of those listed above. Other pumps and valves may be included in the augmented scope at the discretion of CNS.The CNS Fifth 120-Month Interval Pump and Valve Inservice Testing Program Plan will be applicable during the following time period.Begin: 03/01/2016 End: 02/28/2026 1.2 Scope Per 10OCFR 50.55a(f)(4), pumps and valves that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the inservice test requirements set forth in the ASME OM Code and Addenda. Therefore, the regulatory IST Program scope for pumps and valves applies to the Class 1, 2, and 3 pumps and valves that meet the scope statements outlined in the ASME OM Code of record. The CNS Code of record for the Fifth 120-Month Interval is the 2004 Edition through the 2006 addenda of the ASME OM Code.This document is based on the following Subsections and Appendices of the 2004 Edition through the 2006 Addenda of the ASME OM Code:* Subsection ISTA, "General Requirements" ISTA-1 100 states that "These requirements apply to (a) pumps and valves that are required to perform a specific function in shutting down a reactor to the safe shutdown condition, in maintaining the safe shutdown condition, or in mitigating the consequences of an accident." Cooper Nuclear Station was designed and licensed to operate with the Hot Shutdown condition defined as the "safe" shutdown condition.

  • Subsection ISTB, "Inservice Testing of Pumps in Light-Water Reactor Nuclear Power Plants"* Subsection ISTC, "Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants" Revision 0 Pg Page 3 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves*Mandatory Appendix I, "Inservice Testing of Pressure Relief Devices in Light- Water Reactor Nuclear Power Plants"* Mandatory Appendix II, "Check Valve Condition Monitoring Program" Those pumps and valves that are ASME non-Code Class 1, 2, or 3, that meet the scope of the ASME OM Code, will be identified as augmented components within this document.

Other pumps and valves may be included in the augmented scope at the discretion of CNS. NRC approval for any deviation from the ASME OM Code, for the augmented components, is not required.The Cooper Nuclear Station Inservice Testing Program Basis Document includes the justification for inclusion of pumps and valves that are in the IST or Augmented IST Program scope and also many justifications for pumps and valves excluded from the IST and/or Augmented 1ST Program scope.The IST Check Valve Condition Monitoring (CVCM) Program Document contains the details on the check valves selected for this program and all the necessary requirements for implementation of this Program.The CNS IST Program Basis Document, IST CVCM Program Document, administrative procedures, surveillance testing procedures, and other records required to define and execute the Inservice Testing Program are all retained and available at Cooper Nuclear Station.Revision 0 Page 4 Revision 0 Page 4 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves 2.0 INSERVICE TESTING PLAN FOR PUMPS 2.1 Pump Inservice Testing Plan Description The CNS testing program for pumps meets the requirements of the ASME OM Code 2004 Edition through 2006 addenda, Subsection ISTB, "Inservice Testing of Pumps in Light- Water Reactor Nuclear Power Plants". Where these requirements have been determined to be impractical, specific requests for relief were written and are included in Attachment 2 (Pump Relief Requests) and Attachment 3 (Augmented Pump Relief Requests). NUREG 1482, Revision 2 has been used as guidance in the development of the 1ST Program Plan for pumps.2.2 Pump Plan Table Description The pumps included in the Cooper Nuclear Station IST Plan are listed in Attachment

10. The information contained within these tables identifies those pumps which are to be tested to the requirements of Subsection ISTB of the ASME GM Code, 2004 Edition through 2006 Addenda, the testing parameters and frequencies, and associated relief requests.

The headings for the pump tables are delineated below._System: The plant system in which the pump is located.Pump CIC: The pump component identification code.ISI Class: The ASME Inservice Inspection (ISI) classification of the component. Augmented components are classified "A".P&TD: The associated piping and instrumentation drawing number.P&ID Coor: The P&ID coordinate location of the pump.IST Group The Pump group as defined in ISTB-2000 Group A Continuous or routinely operated pumps Group B Standby pumps not operated routinely Parameters: The pump test quantities to be measured or observed. The test designators are as follows: DP Differential Pressure N Speed Pd Discharge Pressure Q Flow Rate V Vibration Frequency: The frequency of testing each pump. The following test designators are used: Q Once every 92 days (Quarterly) 6M Once every 6 months 2Y Once every 2 years Revision 0 Page 5 Rev&ion 0 Page 5 Cooper Nuclear Station Ffith Interval Inservice Testing Pro gram for Pumps and Valves 2.2 Pump Plan Table Description (continued) Notes: This column contains a brief component description, reference to any applicable relief request(s), and contains any other component-related infornation. Relief Requests are designated RP-XX for pumps or RG-XX for general Program Requirements. Augmented Relief Requests are designated ARP-XX. Station Technical Positions are designated TP-XX.2.3 Measurement of Test Quantities Speed (N) Per ASME OM Code ISTB-3530, rotational speed measurement of variable speed pumps shall be taken by a method which meets the requirements of paragraph ISTB-35 10.Pressure (DP, Pd) Differential pressure across a pump will be calculated from inlet and discharge pressure measurements or by direct differential pressure measurement. Discharge pressure will be by direct measurement of discharge pressure. Per NUREG 1482, revision 2, section 5.5.3, suction pressure may be calculated based on inlet tank or bay level.Flow Rate (0) Flow rate of the pump will be measured using a rate or quantity meter installed in the pump test circuit.Vibration (V) Pump vibration will be measured with a digital vibration meter in accordance with the applicable section of ASME CM Code ISTB -3540.2.4 Allowable Ranges of Test Quantities The applicable allowable ranges specified in ASME CM Code ISTB, Tables ISTB-5 121-1, ISTB-5221-1, and ISTB-5321 -1 will be used for differential pressure, flow and vibration measurements except where specific relief is requested and/or approved or design/licensing acceptance criteria is more restrictive than that prescribed in the tables. Should a measured test quantity fall outside the allowable range, corrective action per ASME OM Code ISTB-6200 shall be followed. Records shall be maintained in accordance with ASME OM Code ISTB-9000. 2.5 Instrument Accuracy and Range Requirements Allowable instrument (and loop, where applicable) accuracy's for pressure, flow rate, speed, vibration, and differential pressure are provided in ASME CM Code ISTB Table ISTB-35 10-1 and paragraph ISTB-35 10(a). If the accuracies of the station's instruments do not meet the requirements of this table/section, temporary instruments meeting those requirements in ASME OM Code ISTB Table ISTB-3510-1 will be used or approved relief shall be received.In determining instrument accuracy, the Code does not explicitly require the Licensee to consider physical attributes (such as orifice plate tolerances), tap locations, environmental effects (such as temperature, radiation or humidity), vibration effects (such as seismic), or process effects (such as temperature). This position is documented in NUJREG 1482, revision 2, paragraph 5.5.4.Revision 0 Page 6 Revision 0 Page 6 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Additionally, ASME OM Code ISTB-35 10(b) requires that the full-scale range of analog instruments be no more than 3 times the reference value. Digital instruments shall be selected such that the reference value does not exceed 90% of the calibrated range of the instrument. Revision 0 Page 7 Revision 0 Page 7 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves 3.0 INSERVICE TESTING PLAN FOR VALVES 3.1 Valve Inservice Testing Plan Description The CNS testing program for valves meets the requirements of the ASME OM Code 2004 Edition through 2006 addenda, Subsection ISTC "Inservice Testing of Valves in Light- Water Reactor Nuclear Power Plants"; Mandatory Appendix I "Inservice Testing of Pressure Relief Devices in Light- Water Reactor Nuclear Power Plants "; and Mandatory Appendix II "Check Valve Condition Monitoring Program. " Where these requirements are determined to be impractical, specific requests for relief have been written and are included in Attachment 4 (Valve Relief Requests) and Attachment 5 (Augmented Valve Relief Requests). 3.2 Valve Plan Table Description The table in Attachment 11 lists all ASME Class 1, 2, 3, and Augmented (A) Valves that have been scoped to be within this Program Plan, and have been assigned Valve Categories. In general, valves exempt per ASME OM Code ISTC- 1200 are not listed. The following information is included for each valve.System: The plant system in which the valve is located.Valve GIG: The valve component identification code.P&ID: The associated piping and instrumentation drawing number.P&ID Coor: The drawing coordinates location on the P&ID for the valve.ISI Class: The ASME Inservice Inspection (ISI) classification of the component. Augmented components are classified "A".IST Cat: The category(s) assigned to the valve based on the definitions per ASME GM Code ISTC-1300. The following categories are defined in the Code: Category A -Valves for which seat leakage is limited to a specific maximum amount in the closed position for fulfillment of their required function(s), as specified in ISTA-1 100.Category B -Valves for which seat leakage in the closed position is inconsequential for fulfillment of the required function(s), as specified in ISTA-1 100.Category C -Valves that are self-actuating in response to some system characteristic, such as pressure (relief valves) or flow direction (check valves) for fulfillment of the required function(s), as specified in ISTA-1100.Category D -Valves that are actuated by an energy source capable of only one operation, such as rupture disks or explosively actuated valves.Valve Size: The nominal size of the valve in inches Revision 0 Page 80 Page 8 Cooper Nuclear Station F~/lh Interval Inservice Testing Program for Pumps and Valves 3.2 Valve Plan Table Description (continued) Valve Type: The Valve Body Design as indicated by the following abbreviations: ANGLE BALL BUTTERFLY BALL CHECK DUAL DISK CHECK LIFT CHECK PISTON CHECK SWING CHECK DIAPHRAGM FLOAT VALVE GATE GLOBE PLUG PRESSURE REGULATING RUPTURE DISK RELIEF/SAFETY SOLENOID VALVE STOP VALVE STOP CHECK TILTING DISK CHECK SQUIB ANG BAL BTF CK-B CK-D CK-L CK-P CK-S DIA FOV GT GL PLG PRV RD RV SOV STOP S-CK CK-T SHR ACT Type: The type of Valve Actuator as indicated by the following abbreviations: AIR OPERATOR EXPLOSIVE CHARGE HYDRAULIC OPERATED MANUAL MOTOR OPERATION SELF ACTUATED SOLENOID OPERATOR PILOT ACTUATED AO EX HO MA MO SA SO PA NORMAL POS: The normal position of the valve during regular plant operation, specified as follows: 0 C T OPEN CLOSED THROTITLED TEST ROMT The test(s) that will be performed to fulfill the requirements of ASME OM Code ISTC. The definitions and abbreviations are identified below: Revision 0 Page 9 Revision 0 Page 9 Cooper Nuclear Station Fiflh Interval Inservice Testing Program for Pumps and Valves 3.2 Valve Plan Table Description (continued) LJ-1 Contaimnment Isolation Type C Valve Seat Leakage Test in accordance with the CNS 10CFR50 Appendix J Program (ISTC-3 620).LT- 1 Accumulator Check Valve Leakage Test LT-2 Category A Leak tests other than those already specified COD Check Valve Condition Monitoring open test per Disassembly and Examination CCD Check Valve Condition Monitoring closure test per Disassembly and Examination COF Check Valve Condition Monitoring open test per flow indication measurement CCF Check Valve Condition Monitoring backflow / closure test CCL Check Valve Condition Monitoring closure test per Leakage method of testing CCR Check Valve Condition Monitoring closure test per radiography method of testing ESO Full stroke exercise test to the open position (includes stroke time measurement except for check valves and manual valves)FSC Full stroke exercise test to the closed position (includes stroke time measurement except for check valves and manual valves)PSO Partial stroke exercise to the open position PSC Partial stroke exercise to the closed position FST Fail safe position test (includes stroke time measurement unless otherwise noted)PIT Position indication test RD Rupture Disc test RVT Test of safety/relief valves EX Explosive valve test Revision 0 Page 10 Revision 0 Page 10 Cooper Nuclear Station F/ifh Interval Inservice Testing Pro gram for Pumps and Valves 3.2 Valve Plan Table Description (continued) VBT Vacuum breaker testing SKID Component integral to or that supports operation of major component and is adequately tested as part of the major component. TEST FREO The frequency at which the applicable test will be performed. The definitions and abbreviations are identified below: Test Frequency Frequency of Testing Q At least once per 92 days CS Cold Shutdown RF Refueling Cycle 6M At least once every 6 months 2Y At least once every 2 years 5Y At least once every 5 years l0Y At least once per 10 years App I Relief Valve Frequency (OM Code Appendix I)CVCM IST CVCM Frequency (Appendix II). Refer to the 1ST CVCM Program Document for details.OPB l0CFR50 Appendix J Option B Leakage Rate Frequency PB Performance based frequency other than Option B TS In accordance with Technical Specifications SD Sample Disassembly and Examination NOTES: This column contains a brief component description and references to any applicable cold shutdown justifications(s), refueling outage justification(s), relief request(s), and any other component related information. All valves are considered "active" unless otherwise noted in this column. Cold shutdown justifications are designated CSJ-XX;Refueling Outage Justifications are designated ROJ-XX; Relief Requests are designated RV-XX for valves or RG-XX for general Program Requirements; Augmented Relief Requests are designated ARV-XX;station Technical Positions are designated TP-XX; and the Check Valve Condition Monitoring Plan Bases are designated in the 1ST CVCM Program Document.Revision 0Pae1 Page 11 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves 4.0 ATTACHMENTS: Attachment 1 Attachment 2 Attachment 3 Attachment 4 Attachment 5 Attachment 6 Attachment 7 Attachment 8 Attachment 9 Attachment 10 Attachment 11 System and P&ID Listing Pump Relief Requests Augmented Pump Relief Requests Valve Relief Requests Augmented Valve Relief Requests General Relief Requests Cold Shutdown Justifications Refuel Outage Justifications Technical Positions Inservice Testing Pump Table Inservice Testing Valve Table Revision 0Pae1 Page 12 Cooper Nuclear Station F~fih Interval Inservice Testing Pro gram for Pumps and Valves ATTACHMENT 1 SYSTEM AND P&ID LISTJING System CRD CS DW DGDO DGSA HPCI'IV IA MS NMT NBI PC RW RCIC REC RF RR RWCU RIHR SW SGT SLC SA System Name Control Rod Drive Core Spray Demineralized Water Diesel Generator Diesel Oil Diesel Generator Starting Air High Pressure Coolant Injection Heating and Ventilation Instrument Air Main Steam Neutron Monitoring Traversing Incore Probe Nuclear Boiler Instrumentation Primary Containment Radioactive Waste Reactor Core Isolation Cooling Reactor Equipment Cooling Reactor Feedwater Reactor Recirculation Reactor Water Cleanup Residual Heat Removal Service Water Standby Gas Treatment Standby Liquid Control Station Air P&LD 2039 2045 2029 2011, 2077 2077, 117.10-IC.09 2041, 2044 2019, 2020 2010, 2027, 2028 2028, 2041 2083 2026, 2027, 2028, 2041,, 2045 2022, 2027, 2028, 2084 2005, 2037, 2038 2041, 2043 2031 2043. 2044 2027 2042 2040, 2041 2006, 2036, 2077 2037 2045 2010 Revision 0 Page 13 Revision 0 Page 13 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves ATTACHMENT 2 PUMP RELEIEF REQUESTS________- PUMP RELIEF REQUEST INDEX _______Relif DecripionNRC Approval Date Request No.Decito RP-01 J[Core Spray Pumps Suction Gauge Range 2-12-16'RP-02 1[Residual Heat Removal Pumps Suction Gauge Range [[ 2-12-16'R-3 High Pressure Coolant Injection Pumps Suction Gauge 1 2-12-16~'~*~ Range __________ RP-04 J[Reactor Core IsOlation Cooling Pump Suction Gauge Range ]j 2-12-16'RP-05 I[ Pump Loop Accuracy Requirements 2-12-16' 1 RiP-06 ][Reactor Equipment Cooling Flow Gauge Range [[ 2-12-16'RP-07 j[Core Spray Pump B Vibration Alert Limits II 2-12-16'RP-08 J[Comprehensive Pump Test Upper Limit 2-12-16' 1 RP-09 1[Variance Around the Reference Values 2-12-16'(1) Approved by NRC letter, dated 2-12-16, from Meena K. Khanna, NRC, to Mr. Oscar A. Limpias, Vice President of Nuclear and CNO for CNS Revision 0 Page 14 Revision 0 Page 14 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-01 Core Spray Pump Suction Gauge Range Requirements Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected CS-P-A Core Spray Pump A CS-P-B Core Spray Pump B 2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004 Edition through 2006 Addenda 3. Applicable Code Reqiuirement ISTB-3510O(b)(1) -The full-scale range of each analog instrument shall not be greater than three times the reference value.4. Reason for Reqzuest Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(l), relief is requested from the requirement of ASME GM Code ISTB-35 10(b)(l). The proposed alternative would provide an acceptable level of quality and safety.The installed suction pressure gauge range of the core spray pumps is 30" Hg (inches Mercury) to 30.0 pounds per square inch (psig). The actual values for suction pressure during inservice testing are approximately 4.0 psig. As a result, the instrument range exceeds the requirement of ISTB-35 10(b)(1).5. Proposed Alternative and Basis for Use Pump suction pressure is used along with pump discharge pressure to determine pump differential pressure. Pump suction pressure actual values for the core spray pumps during inservice testing are approximately 4.0 psig. Based on ISTB-3510(b)(l), this would require, as a maximum, a gauge with a range of 0 to 12.0 psig (3 X 4.0 psig) to bound the actual value for suction pressure.Applying the accuracy requirement of + 2% of full scale (+- 6% of reference) for the quarterly Group B pump test, the resulting inaccuracies due to pressure effects would be +/-- 0.24 psig (0.02 X 12 psig).Pump discharge pressure actual values for the core spray pumps during inservice testing are approximately 300 psig. Based on ISTB-351 0(b)(1), this would require, as a maximum, a gauge with a range of 0 to 900 psig (3 X 300.0 psig) to bound the actual value for discharge pressure.Applying the accuracy requirement of + 2% of full scale (4- 6% of reference) for the quarterly Group B pump test, the resulting inaccuracies due to pressure effects would be +/-- 18 psig (0.02 X Revision 0Pae1 Page 15 Cooper Nuclear Station F~ifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-O1 Core Spray Pump Suction Gauge Range Requirements (Continued) 900 psig). Therefore, the maximum inaccuracies due to the suction and discharge pressure indications allowed by the code would be approximately +/- 18.24 psig.The Cooper Nuclear Station (CNS) installed suction pressure gauges (PI-36A/B), which were designed to have an accuracy of+ 0.5% of full scale, have a range of approximately 45 psig. The 45 psig gauge range is derived from the 30" Hg portion of the gauge range that is in a vacuum, which converts to approximately 15 psig, added to the 30 psig positive portion of the gauge. The+/- 0.3 psig current calibration tolerance is essentially a tolerance of approximately 0.66% of full scale (0.0066 X 45 psig = -~ +/- 0.3 psig). Currently, the installed discharge pressure indicators (PI-48A/B) are 0 to 500 psig indicators that are calibrated in a loop with corresponding pressure transmitters (PT-3 8A/B). These loops are being calibrated to +/- 10 psig, or + 2% of full scale (0.02 X 500 psig =+/- 10.0 psig).As an alternative, for the Group B quarterly test, CNS will use the installed suction pressure gauge (30" Hg to 30.0 psig), currently calibrated to within a tolerance of+/- 0.3 psig, together with the installed discharge pressure gauge (0 psig to 500 psig), currently calibrated in a loop to within a tolerance of+/- 10 psig. This results in a combined maximum inaccuracy of+ 10.3 psig due to the installed suction and discharge pressure indications, which is less than the code-allowed +/- 18.24 psig.Although the permanently installed suction pressure gauges (PI-36A/B) are above the maximum range limits of ASME OM Code ISTB-35 10(b)(1), they, in conjunction with the permanently installed discharge pressure gauges (PI-48A/B), yield a better accuracy for differential pressure than the minimum requirements dictated by the code and are, therefore, suitable for the test. The range and accuracy of the instruments used to determine differential pressure will be within +- 6%of the differential pressure reference value. Reference NUJREG 1482, Revision 2, Section 5.5.1.Although not anticipated, if any revisions to the current tolerance information provided occurs within the CNS fifth ten-year interval or actual suction and discharge pressure readings were to change significantly, this relief request will remain valid as long as the combination of range and accuracy will be less than the +/- 6% of the differential pressure reference value.Using the provisions of this relief request as an alternative to the specific requirements of ISTB-351 0(b)(1), identified above, will provide adequate indication of pump performance and continue to provide an acceptable level of quality and safety.Therefore, pursuant to 10 CFR 50.55a(z)(l), Nebraska Public Power District (NPPD) requests relief from the specific ISTB requirements identified in this request.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.Revision 0Pae1 Page 16 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves Relief Request RP-01 Core Spray Pump Suction Gauge Range Requirements (Continued)

7. Precedents This relief request was previously approved for the fourth ten-year interval at CNS as Relief Request RP-01 (TAC Nos. MC8837, MC8975, MC8976, MC8977, MC8978, MC8979, MC8980, MC8981, MC8989, MC8990, MC8991, and MC8992, June 14, 2006).Revision 0 Page 17 Revision 0 Page 17 Cooper Nuclear Station F~ith Interval Inservice Testing Program for Pumps and Valves Relief Request RP-02 Residual Heat Removal Pump Suction Gauge Range Requirements Proposed Alternative in Accordance with 10 CFR 50.55a@z)(1)

Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected RHR-P-A Residual Heat Removal (RHR) Pump A .RHR-P-B Residual Heat Removal Pump B RHR-P-C Residual Heat Removal Pump C RHR-P-D Residual Heat Removal Pump D 2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTB-35 10(b)(1) -The full-scale range of each analog instrument shall not be greater than three times the reference value.4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(l), relief is requested from the requirement of ASME OM Code ISTB-3510(b)(1). The proposed alternative would provide an acceptable level of quality and safety.The installed suction pressure gauge range of the residual heat removal pumps is 30" Hg to 150.0 psig. The actual values for suction pressure during inservice testing are approximately 5.0 psig. As a result, the instrument range exceeds the requirement of ISTB-3510(b)(1).

5. Proposed Alternative and Basis for Use Pump suction pressure is used along with pump discharge pressure to determine pump differential pressure.

Pump suction actual values for the residual heat removal pumps during inservice testing is approximately 5.0 psig. Based on ISTB-35 10(b)(1), this would require, as a maximum, a gauge with a range of 0 to 15.0 psig (3 X 5.0 psig) to bound the actual value for suction pressure. Applying the accuracy requirement of+/- 2% of full scale (+ 6% of reference) for the quarterly Group A pump test, the resulting inaccuracies due to pressure effects would be +/- 0.3 psig (0.02 X 15.0 psig).Pump discharge pressure actual values for the RHR pumps during inservice testing are approximately 170 to 195 psig. Conservatively basing it on the lowest of these discharge pressure readings, ISTB-35 10(b)(1) would require, as a maximum, a gauge with a range of 0 to 510 psig (3 X 170.0 psig) to bound the actual value for discharge pressure.Revision 0Pae1 Page 18 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-02 Residual Heat Removal Pump Suction Gauge Range Requirements (Continued) Applying the accuracy requirement of+/- 2% of full scale (+/- 6% of reference) for the quarterly Group A pump test, the resulting inaccuracies due to pressure effects would be +/- 10.2 psig (0.02 X 510 psig). Therefore, the maximum inaccuracies due to the suction and discharge pressure indications allowed by the code would be approximately +/- 10.5 psig.The CNS-installed suction pressure gauges (PI-10O6A/B/C/D), which were designed to have an accuracy of +/- 0.5% of full scale, have a range of approximately 165 psig. The 165 psig gauge range is derived from the 30" Hg portion of the gauge range that is in a vacuum, which converts to approximately 15 psig, added to the 150 psig positive portion of the gauge. The 4- 1.0 psig current calibration tolerance at the 5 psig suction pressure point is essentially a tolerance of approximately 0.6% of full scale (0.006 X 165 psig = -+ 1.0 psig). Currently, the installed discharge pressure indicators (PI-107AIB/C/D) are 0 to 400 psig indicators. The discharge indicators are being calibrated to +/- 5 psig, or+/- 1.25% of full scale (0.0125 X 400 psig = +/- 5.0 psig).As an alternative, for the Group A quarterly test, CNS will use the installed suction pressure gauge (30" Hg to 150.0 psig), currently calibrated to within a tolerance of 1 psig at the 5 psig point, together with the installed discharge pressure gauge (0 psig to 400 psig), currently calibrated to within a tolerance of+ 5 psig. This results in a combined maximum inaccuracy of+/- 6 psig due to the installed suction and discharge pressure indications, which is less than the code-allowed +/- 10.5 psig.Although the permanently installed suction pressure gauges (PI-1 06A/B/C/D) are above the maximum range limits of ASME OM Code ISTB -351 0(b)(1), they, in conjunction with the permanently installed discharge pressure gauges (PI-10O7A/B/C/D), yield a better accuracy for differential pressure than the minimum requirements dictated by the code and are, therefore, suitable for the test. The range and accuracy of the instruments used to determine differential pressure will be within + 6% of the differential pressure reference value. Reference NUJREG 1482, "Guidelines for Inservice Testing at Nuclear Power Plants," Revision 2, Section 5.5.1.Although not anticipated, if any revisions to the current tolerance information provided occurs within the CNS fifth ten-year interval or actual suction and discharge pressure readings were to change significantly, this relief request will remain valid as long as the combination of range and accuracy will be less than the +/- 6% of the differential pressure reference value.Using the provisions of this relief request as an alternative to the specific requirements of ISTB-351l0(b)(1), identified above, will provide adequate indication of pump performance and continue to provide an acceptable level of quality and safety.Therefore, pursuant to 10 CFR 50.55a(z)(l), NIPPD requests relief from the specific ISTB requirements identified in this request.Revision 0 Page 19 Revision 0 Page 19 Cooper Nuclear Station Fifthi Interval Inserviee Testing Program for Pumps and Valves Relief Request RP-02 Residual Heat Removal Pump Suction Gauge Range Requirements (Continued)

6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth 10-year interval.7. Precedents This relief request was previously approved for the fourth 10-year interval at CNS as Relief Request RP-02 (TAC Nos. MC8837, MC8975, MC8976, MC8977, MC8978, MC8979, MC8980, MC8981, MC8989, MC8990, MC8991, and MC8992, June 14, 2006).Revision 0 Page 20 Revision 0 Page 20 Cooper Nuclear Station Fifthi Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-03 High Pressure Coolant Injection Pump Suction Gauge Range Requirements Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected HTPCI-P-MP High Pressure Coolant Injection (HIPCD) Main Pump HPCI-P-BP High Pressure Coolant Injection Booster Pump 2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTB-3510 (b)(1) -The full-scale range of each analog instrument shall not be greater than three times the reference value.4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(l), relief is requested from the requirement of ASME OM Code ISTB-351 0(b)(1). The proposed alternative would provide an acceptable level of quality and safety.The installed suction pressure gauge range of the high pressure coolant injection pumps is 30" Hg to 150.0 psig. The actual value for suction pressure during inservice testing is approximately 15.0 psig. As a result, the instrument range exceeds the requirement of ISTB-35 10(b)(1).5. Proposed Alternative and Basis for Use Pump suction pressure is used along with pump discharge pressure to determine pump differential pressure. Pump suction actual values for the high pressure coolant injection pumps during inservice testing are approximately 15.0 psig. Based on ISTB-3510(b)(1) this would require, as a maximum, a gauge with a range of 0 to 45.0 psig (3 X 15.0 psig) to bound the actual value for suction pressure. Applying the accuracy requirement of+/- 2% of full scale (+/- 6% of reference) for the quarterly Group B pump test, the resulting inaccuracies due to pressure effects would be+ 0.9 psig (0.02 X 45.0 psig).The pump discharge pressure actual value for the HPCI pump during inservice testing is approximately 1200 psig. Based on ISTB-3510(b)(1), this would require, as a maximum, a gauge with a range of 0 to 3600 psig (3 X 1200.0 psig) to bound the actual value for discharge pressure.Applying the accuracy requirement of + 2% of full scale (+ 6% of reference) for the quarterly Group B pump test, the resulting inaccuracies due to pressure effects would be + 72 psig (0.02 X 3600 psig). Therefore, the maximum inaccuracies due to the suction and discharge pressure indications allowed by the code would be approximately + 72.9 psig.Revision 0Pae2 Page 21 Cooper Nuclear Station FifthI Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-03 High Pressure Coolant Injection Pump Suction Gauge Range Requirements (Continued) The CNS-installed suction pressure gauge (PI-99), which was designed to have an accuracy of-4 0.5% of full scale, has a range of approximately 165 psig. The 165 psig gauge range is derived from the 30" Hg portion of the gauge range that is in a vacuum, which converts to approximately 15 psig, added to the 150 psig positive portion of the gauge. The +i 1.0 psig current calibration tolerance is essentially a tolerance of approximately 0.6% of full scale (0.006 X 165 psig =+ 1.0 psig). Currently, the installed discharge pressure indicator (P1-81) is a 0 to 1500 psig indicator. The discharge indicator is currently being calibrated to + 7.5 psig, or + 0.5% of full scale (0.005 X 1500 psig = + 7.5 psig).As an alternative, for the Group B quarterly test, CNS will use the installed suction pressure gauge (30" Hg to 150.0 psig), currently calibrated to within a tolerance of+/-: 1 psig, together with the installed discharge pressure gauge (0 psig to 1500 psig), currently calibrated to within a tolerance of+/- 7.5 psig. This results in a combined maximum inaccuracy of+/- 8.5 psig due to the installed suction and discharge pressure indications, which is less than the code-allowed + 72.9 psig.Although the permanently installed suction pressure gauge (P1-99) is above the maximum range limits of ASME OM Code ISTB-35 10(b)(1), it, in conjunction with the permanently installed discharge pressure gauge (P1-81), yields a better accuracy for differential pressure than the minimum requirements dictated by the code and is, therefore, suitable for the test. The range and accuracy of the instruments used to determine differential pressure will be within +/-- 6% of the differential pressure reference value. Reference NUREG 1482, Revision 2, Section 5.5.1.Although not anticipated, if any revisions to the current tolerance information provided occurs within the CNS fifth ten-year interval or actual suction and discharge pressure readings were to change significantly, this relief request will remain valid as long as the combination of range and accuracy will be less than the + 6% of the differential pressure reference value.Using the provisions of this relief request as an altemnative to the specific requirements of ISTB-35 10(b)(1), identified above, will provide adequate indication of pump performance and continue to provide an acceptable level of quality and safety.Therefore, pursuant to 10 CFR 50.55a(z)(1), NPPD requests relief from the specific ISTB requirements identified in this request.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth 10-year interval.7. Precedents This relief request was previously approved for the fourth 10-year interval at CNS as Relief Request RP-03 (TAC Nos. MC8837, MC8975, MC8976, MC8977, MC8978, MC8979, MC8980, MC8981, MC8989, MC8990, MC8991, and MC8992, June 14, 2006).Revision 0Pae2 Page 22 Cooper Nuclear Station Ffith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-04 Reactor Core Isolation Cooling Pump Suction Gauge Range Requirements Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected RCIC-P-MP Reactor Core Isolation Cooling (RCIC) Main Pump 2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTB-3 510O(b)(1) -The full-scale range of each analog instrument shall not be greater than three times the reference value.4. Reason for Reqjuest Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(1), relief is requested from the requirement of ASME OM Code ISTB-35 10(b)(1). The proposed alternative would provide an acceptable level of quality and safety.The installed suction pressure gauge range of the reactor core isolation cooling pump is 30" Hg to 150.0 psig. The actual value for suction pressure during inservice testing is approximately 15.0 psig. As a result, the instrument range exceeds the requirement of ISTB-35 10(b)(1).5. Proposed Alternative and Basis for Use Pump suction pressure is used along with pump discharge pressure to determine pump differential pressure. Pump suction actual values for the reactor core isolation cooling pump during inservice testing is approximately 15.0 psig. Based on ISTB-351l0(b)(1) this would require, as a maximum, a gauge with a range of 0 to 45.0 psig (3 X 15.0 psig) to bound the lowest actual value for suction pressure. Applying the accuracy requirement of+ 2% of full scale (+ 6% of reference) for the quarterly Group B pump test, the resulting inaccuracies due to pressure effects would be +/- 0.9 psig (0.02 X 45.0 psig).The discharge pressure actual value for the RCIC pump during inservice testing is approximately 1250 psig. Based on ISTB-3510(b)(1), this would require, as a maximum, a gauge with a range of 0 to 3750 psig (3 X 1250.0 psig) to bound the actual value for discharge pressure. Applying the accuracy requirement of + 2% of full scale (+ 6% of reference) for the quarterly Group B pump test, the resulting inaccuracies due to pressure effects would be +/- 75 psig (0.02 X 3750 psig). Therefore, the maximum inaccuracies due to the suction and discharge pressure indications allowed by the code would be approximately +/- 75.9 psig.Revision 0 Page 23 Revision 0 Page 23 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-04 Reactor Core Isolation Cooling Pump Suction Gauge Range Requirements (Continued) The CNS-installed suction pressure gauge (PI-66), which was designed to have an accuracy of+/- 0.5% of full scale, has a range of approximately 165 psig. The .165 psig gauge range is derived from the 30" Hg portion of the gauge range that is in a vacuum, which converts to approximately 15 psig, added to the 150 psig positive portion of the gauge. The 4- 1.0 psig current calibration tolerance is essentially a tolerance of approximately 0.6% of full scale (0.006 X 165 psig =+ 1.0 psig). Currently, the installed discharge pressure indicator (PI-5 9) is a 0 to 1500 psig indicator. The discharge indicator is being calibrated to +/- 15 psig, or + 1.0% of full scale (0.01 X 1500 psig= + 15.0 psig).As an alternative, for the Group B quarterly test, CNS will use the installed suction pressure gauge (30" Hg to 150.0 psig), currently calibrated to within a tolerance of++/- 1 psig, together with the installed discharge pressure gauge (0 psig to 1500 psig), currently calibrated to within a tolerance of+ 15.0 psig. This results in a combined maximum inaccuracy of+ 16.0 psig due to the installed suction and discharge pressure indications, which is less than the code-allowed +/- 75.9 psig.Although the permanently installed suction pressure gauge (P1-66) is above the maximum range limits of ASME OM Code ISTB-3510(b)(l), it, in conjunction with the permanently installed discharge pressure gauge (PI-59), yields a better accuracy for differential pressure than the minimum requirements dictated by the code and is, therefore, suitable for the test. The range and accuracy of the instruments used to determine differential pressure will be within + 6% of the differential pressure reference value. Reference NUREG 1482, Revision 2, Section 5.5.1.Although not anticipated, if any revisions to the current tolerance information provided occurs within the CNS fifth ten-year interval or actual suction and discharge pressure readings were to change significantly, this relief request will remain valid as long as the combination of range and accuracy will be less than the +/- 6% of the differential pressure reference value.Using the provisions of this relief request as an alternative to the specific requirements of ISTB-351 0(b)(1), identified above, will provide adequate indication of pump performance and continue to provide an acceptable level of quality and safety.Therefore, pursuant to 10 CFR 50.55a(z)(1), NPPD requests relief from the specific ISTB requirements identified in this request.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was previously approved for the fourth ten-year interval at CNS as Relief Request RP-04 (TAC Nos. MC8837, MC8975, MC8976, MC8977, MC8978, MC8979, MC8980, MC8981, MC8989, MC8990, MC8991, and MC8992, June 14, 2006).Revision 0Pae2 Page 24 Cooper Nuclear Station F~ilh Interval Inservice Testing Pro gram for Pumps and Valves Loop Accuracy Requirements Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternate Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected CS-P-A CS-P-B HIPCI-P-MP HiPCI-P-BP SW-P-BPA SW-P-BPB SW-P-BPC*SW-P-BPD Core Spray (CS) Pump A Core Spray Pump B High Pressure Coolant Injection Main Pump High Pressure Coolant Injection Booster Pump Service Water Booster (SWB) Pump A Service Water Booster Pump B Service Water Booster Pump C Service Water Booster Pump D 2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement Table ISTB-35 10-1, "Required Instrument Accuracy" 4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(1), relief is requested from the requirement of ASME OM Code ISTB Table ISTB-3510-l for Group A and B Pump Pressure accuracy (A- 2%) and for flow rate accuracy (A- 2%). The proposed alternative would provide an acceptable level of quality and safety.The installed instrumentation for the subject pumps yield the following loop accuracies: Pump Parameter Equip. Loop Accuracy (%)Calibration Loop Accuracy (%)CS Pump Discharge Pressure CS Pump Flowrate FTPCI Pump Flowrate SWB Pump Flowrate 2.06 2.02 2.03 2.03<2 200%< 2.00%< 2.00%< 2.00%As a result, the equipment loop accuracies do not meet the A- 2% requirements of Table ISTB-35 10-1, "Required Instrument Accuracy." 5. Proposed Alternative and Basis for Use The difference between the code required and presently installed instrument loop accuracies is 0.06%, at a maximum, as presented above. This difference is insignificant when applied to the quantitative measured values for these parameters during the respective Group A or Group B Revision 0 Page 25 Cooper Nuclear Station Ffith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-05 Loop Accuracy Requirements (Continued) quarterly tests. Additionally, all calibration tolerances of the loops involved meet or exceed the code-allowed accuracies of+ 2% or better.CS pump discharge pressure loop is made up of a pressure indicator (range of 0 to 500 psig) and a pressure transmitter. The pressure indicator (PI-48A/B) has a nameplate accuracy of+ 2%, and the pressure transmitter (PT-38A/B) has a nameplate accuracy oftq 0.5%. Therefore, based on the nameplate accuracies alone, the equipment loop accuracy for discharge pressure indication is-4 2.06% (square root of the sum of the squares), which exceeds the code requirement of+/- 2%.The variation from the code of 0.06%, with a gauge range of 0 to 500 psig, would amount to a potential deviation of only 0.3 psig (0.0006 X 500). However, CNS is currently calibrating this discharge pressure loop to within +- 10 psig, which is equivalent to a +- 2% of full scale tolerance (0.02 X 500 psig =+ 10 psig), which meets the accuracy requirements of the code.CS pump flow rate loop is made up of a flow indicator (range of 0 to 6000 gallons per minute[gpm]), and a flow transmitter. The flow indicator (FI-50A/B) has a nameplate accuracy of+/- 2%, and the flow transmitter (FT-40A/B) has a nameplate accuracy of+ 0.25%. Therefore, based on the nameplate accuracies alone, the equipment loop accuracy for discharge pressure indication is+ 2.02% (square root of the sum of the squares), which exceeds the code requirement of+/- 2%.The variation from the code of 0.02%, with a gauge range of 0 to 6000 gpm, would amount to potential deviation of only 1.2 gpm (6000 X .0002). However, CNS is currently calibrating this flow loop to within +/- 50 gpm (at the Inservice Testing (1ST) reference value of 5000 gpm) or approximately +/- 0.83% of full scale (+/-- 0.0083 X 6000 = +/- 50 gpm), which is better than the+/-- 2% of full scale accuracy requirements of the code. If a preservice test were to be run, CNS would ensure that the loop was calibrated to < 2% over the full range of the test prior to performing it.HiPCI pump flow rate loop is made up of a flow indicating controller (range of 0 to 5000 gpm), a flow transmitter, and a flow square rooter. The flow indicating controller (FIC-1 08) has a nameplate accuracy of+ 0.25%, the flow transmitter (FT-82) has a nameplate accuracy of+- 0.25%, and the flow square rooter (SQRT-l118) has a nameplate accuracy of +/- 2% from approximately 0 to 1000 gpm and +/- 0.5% from approximately 1000 to 5000 gpm. Therefore, based on the nameplate accuracies alone, the equipment loop accuracy for flow indication is approximately

  • 2.03% (square root of the sum of the squares) from 0 to 1000 gpm, which does not meet the code requirement of +/- 2%, and approximately

+- 0.6 1% from 1000 to 5000 gpm, which does meet the code requirement of + 2%. The variation from the code of 0.03% in the range of 0 to 1000 gpm, with a gauge range of 0 to 5000 gpm, would amount to a potential deviation of only 1.5 gpm (5,000 X .0003). However, CNS is currently calibrating this flow loop to within + 100 gpm (at the 1ST reference of 4000 gpm and at other points from 1000 gpm to 5000 gpm) or + 2% of full scale (+ 0.02 X 5000 = ~-+ 100 gpm), which is equivalent to the + 2%of full scale accuracy requirements of the code. If a preservice test were to be run, CNS would ensure that the loop was calibrated to < 2% over the full range of the test prior to performing it.The SWB flow rate is made up of a flow indicator (range of 0 to 10,000 gpm), a flow transmitter, and a flow square rooter. The flow indicator (FI-l132A/B) has a nameplate accuracy of+/- 2%, the flow transmitter (FT-97) has a nameplate accuracy of+/- 0.25%, and the flow square rooter (SQRT-132AiB) has a nameplate accuracy of+/- 0.25%. Therefore, based on the nameplate accuracies alone, the equipment loop accuracy for flow indication is approximately + 2.03%Revision 0 Page 26 Cooper Nuclear Station Ffith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-05 Loop Accuracy Requirements (Continued)(square root of the sum of the squares), which exceeds the code requirement of + 2%. The variation from the code of 0.03%, with a gauge range of 0 to 10,000 gpm, would amount to a potential deviation of only 3 gpm (0.0003 X 10,000). However, CNS is currently calibrating this flow loop to within +/-- 100 gpm, which is equivalent to a +/-: 1% of full scale tolerance (0.01 X 10,000 gpm = +/-- 100 gpm), which is better than the +/-- 2% of full scale accuracy requirements of the code.As an alternative for the Group A or Group B quarterly test, CNS will use the installed instruments calibrated such that the loop accuracies are as indicated in the above table. No adjustments to acceptance criteria will be made as the calibrated loop accuracies will meet or exceed the code tolerances. Although the permanently installed instrument loops do not meet the accuracy requirements of ASME OM Code ISTB Table ISTB-3510-1 when looking at nameplate accuracies, the effects of these small inaccuracies are insignificant when compared to the measured values, and credit will be taken for the ability to calibrate the loop within the code-allowed tolerance. Although not anticipated, if any revisions to the current tolerance information provided occurs within the CNS fifth ten-year interval, this relief request will remain valid as long as the calibrated loop accuracies meet the code required tolerances of < 2.00% of full scale.Using the provisions of this relief request as an alternative to the specific requirements of ISTB Table 3510-1, identified above, will provide adequate indication of pump performance and continue to provide an acceptable level of quality and safety.Therefore, pursuant to 10 CFR 50.55a(z)(1), NPPD requests relief from the specific ISTB requirements identified in this request.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was previously approved for the fourth ten-year interval at CNS as Relief Request RP-05 (TAC Nos. MC8837, MC8975, MC8976, MC8977, MC8978, MC8979, MC8980, MC8981, MC8989, MC8990, MC8991, and MC8992, June 14, 2006).Revision 0 Page 27 Revision 0 Page 27 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RP-06 Reactor Equipment Cooling Pump Flow Rate Range Requirements Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected REC-P-A Reactor Equipment Cooling (REC) Pump A REC-P-B Reactor Equipment Cooling Pump B REC-P-C Reactor Equipment Cooling Pump C REC-P-D Reactor Equipment Cooling Pump D 2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTB-35 10(b)(1) -The full-scale range of each analog instrument shall not be greater than three times the reference value.4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(1), relief is requested from the requirement of ASME OM Code ISTB-35 10(b)(l). The proposed alternative would provide an acceptable level of quality and safety.The installed flow rate instrument range of the reactor equipment cooling pumps is 0 to 4000 gpm. The reference values for flow rate during inservice testing are 1100 gpm. As a result, the instrument range exceeds the requirement of ISTB-35 10(b)(1).5. Proposed Alternative and Basis for Use The permanent plant flow Instruments REC-FI-450A and REC-FI-450B are calibrated such that their accuracy is 1.25% of full scale. This yields a total inaccuracy of 50 gpm (0.0125 X 4000 gpm). Reference flow rates for the reactor equipment cooling pumps are 1100 gpm. Based on ISTB-3510(b)(1) this would require, as a maximum, a gauge with a range of 0 to 3300 gpm (3 X 1100 gpm) to bound the lowest reference value for flow.Applying the accuracy requirement of+~ 2% for the pump test, the resulting inaccuracies due to flow would be + 66 gpm (0.02 X 3300 gpm).Revision 0Pae2 Page 28 Cooper Nuclear Station Ffith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-06 Reactor Equipment Cooling Pump Flow Rate Range Requirements (Continued) As an alternative, for the reactor equipment cooling pump inservice tests, CNS will use the installed flow rate instrumentation (0 to 4000 gpm) calibrated to less than + 2% such that the inaccuracies due to flow will be less than or equal to that required by the code (+ 66 gpm). This will ensure that the installed flow rate instrumentation is equivalent to the code, or better, in terms of measuring flow rate.Although the permanently installed flow gauges are above the maximum range limits of ASME OM Code ISTB-35 1O(b)(1), they are within the accuracy requirements and are, therefore, suitable for the test. Reference NUJREG 1482, Revision 2, Section 5.5.1.Using the provisions of this relief request as an alternative to the specific requirements of ISTB-351 0(b)(1), identified above, will provide adequate indication of pump performance and continue to provide an acceptable level of quality and safety.Therefore, pursuant to 10 CFR 50.55a(z)(1), NPPD requests relief from the specific ISTB requirements identified in this request.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was previously approved for the fourth ten-year interval at CNS as Relief Request RP-06 (TAG Nos. MC8837, MC8975, MC8976, MC8977, MC8978, MC8979, MC8980, MC8981, MC8989, MC8990, MC8991, and MC8992, June 14, 2006).Revision 0 Page 29 Revision 0 Page 29 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2) Alternative Demonstrates a Hardship Without a Compensating Increase in Quality and Safety 1. ASME Code Component(s) Affected CS-P-B Core Spray Pump B 2. Applicable Code Edition and Addenda ASMEi OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTB Table ISTB-5 121-1, "Centrifugal Pump Test Acceptance Criteria" 4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(2), relief is requested from the requirement of ASME OM Code ISTB Table ISTB-5121-1 during the biennial comprehensive pump test or any other time vibrations are taken to determine pump acceptability (i.e., post-maintenance testing, other periodic testing, etc.). The proposed alternative demonstrates a hardship without a compensating increase in quality and safety.The 1ST Program has consistently required (prior to obtaining relief per RP-06 of the third interval program) that CS Pump B (CS-P-B) be tested on an increased frequency due to vibration values at Points 1H and 5H, as shown in Figure 1, periodically being in the alert range. Relief is requested from ISTB Table ISTB-5 121-1 requirements to test the pump on an increased periodicity due to vibration levels for Points 1H and/or 5H exceeding the ISTB alert range absolute limit for the comprehensive pump test. This request is based on analysis of vibration and pump differential pressure data indicating that no pump degradation is taking place. CNS is proposing to use alternative vibration alert range limits for vibration Points 1H and 5H. This provides an alternative method that continues to meet the intended function of monitoring the pump for degradation over time while keeping the required action level unchanged.

5. Proposed Alternative and Basis for Use Pump Testing Methodology CS-P-B at CNS is tested using a full flow recirculation test line back to the suppression pool each quarter. CS-P-B has a minimum flow line which is used only to protect the pump from overheating when pumping against a closed discharge valve. The minimum flow line isolation valve for CS-P-B is initially open when the pump is started, and flow is initially recirculated through the minimum flow line back to the suppression pool.Revision 0Pae3 Page 30 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued)

Then, the full-flow test line isolation valve is throttled open to establish flow through the full-flow recirculation test line. The minimum flow line is then isolated automatically, and all flow remains through the full-flow test line for the IST test.The B train of the CS system is operated in the same manner and under the same conditions for each test of CS-P-B, regardless of whether CNS is operating or shut down. Consequently, the pump will experience the same potential for flow-induced, low frequency vibration whenever it is tested, whether CNS is operating or shut down. As a result, this relief is requested for the comprehensive pump testing of CS-P-B when vibration measurements are required or any other time vibrations are recorded to determine pump acceptability (i.e., post-maintenance testing, other periodic testing, etc.).CNS considers full-flow testing to be preferable to minimum flow testing due to the ability to evaluate overall pump performance at post-accident flow design conditions. Minimum flow testing would provide only limited information about the pump.Nuclear Regulatory Commission (NRC) Staff Document NUREG/CP-O0152 NRC Staff document NUREG/CP-O0152, entitled "Proceedings of the Fourth NRC/ASME Symposium on Valve and Pump Testing," dated July 15-18, 1996, included a paper entitled Nuclear Power Plant Safety Related Pump Issues, by Joseph Colaccino of the NRC staff. That paper presented four key components that should be addressed in a relief request of this type to streamline the review process. These four key components are as follows: I. The licensee should have sufficient vibration history from inservice testing which verifies that the pump has operated at this vibration level for a significant amount of time, with any "spikes" in the data justified. II. The licensee should have consulted with the pump manufacturer or vibration expert about the level of vibration the pump is experiencing t~o determine if pump operation is acceptable. III. The licensee should describe attempts to lower the vibration below the defined code absolute levels through modifications to the pump.IV. The licensee should perform a spectral analysis of the pump-driver system to identify all contributors to the vibration levels.The following is a discussion of how these four key components are addressed for this relief request.Revision 0Pae3 Page 31 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) I. Vibration History (Key Component No. 1)A. Testing Methods and Code Reqiuirements Inconsistent higher vibrations on CS-P-B have been a condition that has existed since original installation of this pump in 1973. During the construction and preoperational testing, vibrations were measured in "mils" at the top and side of the motor outboard (farthest from the pump), the side of the motor inboard (nearest the pump), and pump inboard (nearest the motor). The vibration signals were tape recorded along with the dynamic pressure pulsations in the suction and discharge of the pump as thc flow was varied. The intention was to see if hydraulic disturbances were responsible for the observed phenomena. Observation of the vibration signals on the oscilloscope showed conclusively that the motor was vibrating with randomly distributed bursts of energy at the natural frequency of the total system. Therefore, it was determined that the hydraulic disturbances found in the piping was the source of the energy. Pipe restraints were added that reduced the piping system vibrations. The monitoring of multiple vibration points over the years had not been a requirement of Section XI of the ASME Code until the adoption of the OM Standards/Codes. Therefore, at CNS, the first and second ten-year interval IST code requirements did not include the monitoring of multiple vibration points. The CNS second interval 1ST Program was committed to the 1980 Edition, Winter 1981 Addenda of Section XI. Paragraph IWP-4510 of this code required that "at least one displacement vibration amplitude shall be read during each inservice test." This code was in effect at CNS until the start of the third ten-year interval, which began on March 1, 1996. The CNS third interval IST Program was committed to the 1989 Edition of Section XI, which required multiple vibration points to be recorded during IST pump testing in accordance with the ANSI/ASME Operations and Maintenance Standard, Part 6, 1987 Edition with the 1988 Addenda.However, CNS proactively began monitoring vibration on pumps in the IST Program in velocity units (inches per second) at multiple vibration points in 1990 in accordance with an approved relief request. Therefore, data exists for vibration Points 11H and 5H from April 1,990 to the present. This data is included in the figures provided in this relief request. In April 1990, an analog velocity meter was utilized to begin measuring five different points in units of velocity. These are the same points measured today. Further technological advances resulted in the utilization of more reliable vibration meters beginning in late 1996. For the fourth interval, which began on March 1, 2006, the 2001 Edition through 2003 Addenda of the ASME OM Code was the code of record.Vibration measurements were required to be taken only during the comprehensive test since the CS-P-B pump is considered a Group B pump. The same will be true for the fifth interval, beginning on March 1, 2016, in which the 2004 Edition through the 2006 Addenda of the ASME OM Code will be the code of record.Revision 0 Page 32 Revision 0 Page 32 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) B. Review of Vibration History Data Beginning in April 1990, five vibration points (lV, 1H, 2H, 3H, 5H) were recorded for CS-P-B. However, the pump was tested at 4720 gpm from April 1990 to April 1992, then at 4800 gpm from April 1992 through December 1994, and finally at 5000 gpm from January 1995 to the present. The January 1995 test was also a post-maintenance test following the work that replaced the restricting orifice in the test return line. The last re-baseline occurred on November 6, 1996, due to the implementation of a new vibration meter with new instrument settings. Therefore, it would be appropriate to review the data from this date forward to track for degradation. This would be over eighteen years of data at the same reference points.CS-P-B IST vibration trend graphs for vibration points 5H, 1V, 2H, and 3H (Figures 3a, 4a, 5a, and 6a in this relief request), which include data from November 6, 1996, to the present, show flat or slightly downward trends. Vibration point 1H shows an essentially flat trend from -~2002 to the present (Figure 2a) and when including the data since 1990 (Figure 2b). These observations indicate that CS-P-B vibrations are not increasing in magnitude. These trends also show that Points 1H and 5H occasionally exceed the alert range criteria (Figures 2a and 3a). Figure 12 illustrates the trend for CS-P-B differential pressure (D/P) readings from January 1995 (re-baselined pump at 5000 gpm) to the present. This represents approximately twenty years of data for pump D/P with the testing at 5000 gpm. As can be seen from Figure 12, no degradation in pump D/P has occurred.Trend Graphs 2b, 3b, 4b, 5b, and 6b illustrate vibration data dating back to April 1990 for all vibration points. The data prior to 1996 represents data taken with analog, less reliable vibration instruments and, as discussed previously, at differing flows. However, it does clearly indicate that the piping-induced vibrations for vibration Points 1 H and 5H were present in the early 1990s. This condition was also documented in the 1980s. In July 1985, CNS work item #85-2497 documented high vibration readings on the horizontal motor position. A pipe resonance problem was suspected at that time.Vibrational readings varied between 0.3 and 0.5 in/sec with spikes to 0.7 ir/sec every few seconds. This 1985 documentation, available vibration data since 1990, along with the testing performed during the preoperational time period, substantiates that the piping-induced vibrations have been in existence since the pump was installed. These graphs indicate that the vibration point trends since April 1990 are essentially flat or slightly downward. Therefore, based on the available data at CNS, this pump has experienced essentially no degradation in vibration levels for ~-24.5 years or in DiP for -20 years.Revision 0 Page 33 Revision 0 Page 33 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) C. Review of "Spikes" in Vibration Data In reviewing the trend data for vibration points 1 H (Figures 2a and 2b) and 5H (Figures 3a and 3b), which includes the code-required frequency ranges (one-third pump running speed to 1000 Hertz [Hz].), random spikes were observed throughout the data that resulted in values above the alert range. These spikes are best described in a 2001 report by Machinery Solutions, Inc., an industry expert on vibrations, as follows: Most of the vibration that is measured on the motor casing is due to excitation of the structural resonances of the motor/pump by turbulent flow. These structural resonances are poorly damped and can be easily excited. Most vertical pumps have similar types of behavior, and it is not necessarily problematic by itself. A problem occurs when a pump has a continuous forcing function whose frequency coincides with a resonance (i.e., running speed). The forcing function in this case is flow turbulence caused in large part by the S-curve in the piping just off the pump discharge. The flow through this area generates lateral broadband forces, due to elbow effects, that excite the resonances in a non-continuous fashion.This is why the amplitude swings so dramatically on the motor case (the location of vibration points 1 H and 5H). The system goes from brief periods of excitation to brief periods of no excitation. The discharge riser is also moving side to side from the same forces. Although the discharge piping configuration is both non-standard and less than optimum for this application, it poses no threat to the long-term reliability of either the pump or the motor. The only negative impact is on vibration levels relative to a generic standard.As illustrated previously, there have been no degrading trends associated with vibration data points 1H and 5H for -24.5 years (Figures 2b and 3b). Since June 2002, filtered data (removal of one-third pump running speed to one-half pump running speed frequencies) has been recorded in addition to the current code-required values for vibration points 111 and 5H (reference Figures 2c and 3c for data since 2010). In reviewing this data, the trends are lower in value, steady, and without the spikes that the code-required data contains. This further supports the fact that the spikes in the original code data are due to the piping-induced, non-detrimental vibration occurring at the one-third to one-half pump running speed.Revision 0Pae3 Page 34 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) II. Consultation -Pump Manufacturer/Vibration Expert (Key Component No. 2)A. Pump Manufacturer Evaluation of CS-P-B Vibrations Byron Jackson is the pump manufacturer for CS-P-B. The pump is an 8 x 14 x 30 DVSS, vertical mount, single stage centrifugal pump. The pump impeller is mounted on the pump motor's extended shaft. As outlined in the Core Spray System Summary of Preoperational Test, the data obtained for the B Core Spray Pump indicated high vibration. The high yibration had been recognized early in the construction testing phase, and Byron Jackson sent a representative to the site to investigate. In a letter dated February 16, 1973, the Byron Jackson representative indicated the following:

1. Tests indicated that the natural frequency of the pump was 940 revolutions per minute (rpm) (approximately one-half pump speed) in the direction of the piping and 720 rpm (between one-third and one-half of pump speed) in the direction perpendicular to the piping.2. Observation of the test signals on the oscilloscope showed very conclusively that the motor was vibrating with randomly distributed bursts of energy, the frequency of which matched the natural frequency of the total system. This can only mean that the energy is coming from the hydraulic disturbances found in the piping.3. Whenever large flows are carried in piping, there is usually considerable turbulence associated with the elbows, tees, etc., of the piping configuration, all of which results in piping reactions and motion. Apparently, the vibrating piping was, in turn, vibrating the pump.4. When jacks were installed between the top of the pump and the bottom of the motor flange in an effort to stiffen the motor pump system, the motor vibrations went up due to more energy being transmitted from the pipe-pump system into the motor.5. Testing was performed to determine any weaknesses in the pump-motor mechanical system. The vibration amplitude using the IRD instrument, with the filter set at operating speed, sampled many points vertically along the pump-motor structure.

Plots of the data (along with phase angle determined by means of the strobe light) showed very clearly that the total structure was vibrating as a rigid assembly from the floor mounting. Examination of the high amplitude vibration signals showed them to be at the extremely low system natural frequencies as determined earlier.6. Such low acceleration levels, along with the system acting as a rigid structure (between motor and pump), means that the motor and pump can operate Revision 0Pae3 Page 35 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) with these levels of vibration with absolutely no impairment of operating life.This is the picture that seems very clearly described by the data obtained during these tests. There is absolutely no reason to restrict the operation of these pumps in any way.Although the vibration was found to be acceptable, CNS took actions to install new pipe supports as an attempt to reduce these piping-induced vibrations. This action was successful as will be discussed in a later section of this relief request.B. CNS Expert Analysis of CS-P-B Vibrations As the Vibration Monitoring Program expanded in the early 1 990s, it became evident that the low frequency, piping-induced vibrations still remained in CS-P-B. Design Change (DC) 94-046 resulted in the replacement of the orifices in the test return line. A March 16, 1995, memo to the CNS 1ST Engineer from the CNS Lead Civil/Structural Engineer discussed the CS-P-B vibration measurements obtained during DC 94-046 acceptance testing.The vibration data was collected using peak velocity measuring instrumentation as required for the performance of the IST test and with instrumentation that provides displacement and velocity versus frequency data. It was observed that the significant vibrations in the 1H direction were occurring around 700 cycles per minute (cpm), while the pump speed is at 1780 cpm (i.e., rpm). Given the piping movement of the system, and the knowledge that piping vibrations can commonly occur in the 700 cpm (12 Hz)range, CNS concluded that the pump vibrations were piping dependent. The CNS Lead Civil/Structural Engineer concluded that the significant pump vibrations are occurring at less than one-half of the pump operating speed. The pumps are rigidly mounted at their bases, and any impeller-induced vibrations would occur at the pump running speed or at the vane passing frequency. Therefore, the sub-synchronous pump vibrations are clearly piping induced, non-detrimental to pump/motor service or reliability, and should not be used as a basis for pump degradation. This is because the purpose of pump in-service testing is to diagnose and trend internal pump degradation. The memo further states that the vibration data collection requirement specified in the 1ST procedure consists of peak velocity recordings, which may be masked by piping-induced vibrations, negating internal pump degradation diagnosis and trending. Based on the historical trending data for both CS pumps, the vibration has remained at a consistent amplitude, trending neither upward nor downward, indicating that the induced vibrations are not impairing pump operability, nor capable of preventing the pump from fulfilling its safety function. The piping vibration is present when flow is present through the test return line. It was visually observed during DC 94-046 acceptance testing that piping vibrations were minimal when flow was directed through the minimum flow line.Revision 0Pae3 Page 36 Cooper Nuclear Station Fifth Interval lnservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Following the DC 94-046 testing, CNS noted that the deflections observed in the discharge piping were significantly reduced. Based on these results, it was determined by the Nuclear Engineering Department, Civil/Structural Group, that the CS Loop B piping vibration stresses are less than the endurance limit of the piping.On October 17, 2002, a Plant Engineering Supervisor at CNS, knowledgeable in the area of pump vibration analysis, issued a memo to the CNS Risk & Regulatory Affairs Manager discussing the low frequency vibration issue with the CS-P-B.In the memo, it is stated that the pipe is vibrating as a reaction to flow turbulence, which in turn is causing the pump to vibrate. The memo documents the basis for why the low frequency vibration (less than one-half pump running speed) experienced during CS-P-B operation is not indicative of degrading pump performance and is not expected to adversely impact pump operability. To summarize, in the area of pump performance, aside from the randomness of the low frequency peaks, the spectral data shows no degrading trend in performance over several years of data. The low frequency piping-induced vibrations are not expected to adversely impact pump operability. C. Independent Industry Vibration Expert Evaluation of CS-P-B In 2001, Machinery Solutions, Inc. was retained to perform an independent study of the CS-P-B vibrations. The following discussion was obtained from their report, issued in September of 2001. Machinery Solutions, Inc. utilized seven transducers and acquired data from CS-P-B continuously while it was operating, and data was stored every 3 seconds. Orbit plots, spectrum plots, bode and polar plots, cascade/waterfall plots, overall amplitude plots, trend plots, XY graph plots, and tabular lists were utilized to analyze the data. The data obtained by Machinery Solutions, Inc., indicated that the vibration amplitudes during the run were much higher at the top of the motor than they were at the bottom of the motor. The amplitudes decreased even further on the pump.The spectrum plots showed that most of the vibration was occurring below running speed. They also showed that the low frequency vibration is a different frequency in each direction. The predominant peaks occur at approximately 870 cpm (less than one-half pump running speed) in line with discharge and at approximately 630 cpm (less than one-half pump running speed) perpendicular to discharge. The amplitude of each of these peaks varied significantly from second to second. The natural frequency of the pump-motor-piping structure was determined via impact testing prior to starting the pump. The natural frequencies were determined to be approximately 830 cpm in line with discharge and 670 cpm perpendicular to discharge. Such a vibration response is typical for vertical pumps.Revision 0 Page 37 Revision 0 Page 37 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Machinery Solutions, Inc. concluded the following:

1. Most of the vibration that is measured on the motor casing is due to excitation of the structural resonances of the motor/pump by turbulent flow. These structural resonances are poorly damped and can be easily excited. Most vertical pumps have similar types of behavior, and it is not necessarily problematic by itself. A problem occurs when a pump has a continuous forcing function whose frequency coincides with a resonance (i.e., running speed). The forcing function in this case is flow turbulence caused in large part by the S-curve in the piping just off the pump discharge.

The flow through this area generates lateral broadband forces, due to elbow effects, that excite the resonances in a non-continuous fashion. This is why the amplitude swings so dramatically on the motor case (the location of vibration points 1 H and 5H). The system goes from brief periods of excitation to brief periods of no excitation. The discharge riser is also moving side to side from the same forces. Although the discharge piping configuration is both non-standard and less than optimum for this application, it poses no threat to the long-term reliability of either the pump or the motor. The only negative impact is on vibration levels relative to a generic standard.2. The balance condition of the motor and pump are acceptable with no corrective action required at this time.3. The shaft alignment between the motor and the pump is acceptable for long-term operation.

4. There is no evidence of motor bearing wear.Machinery Solutions, Inc. recommended the following actions: 1. Create a new IST vibration data point configuration within the data collector database to use an overall level that is generated from spectral data above 950 cpm. This will eliminate the energy from the resonances from the data set and still allow for protection from bearing degradation, impeller degradation, and motor malfunctions.

The only potential failure mode that could occur within this excluded frequency range would be a fundamental train pass frequency generated by a rolling element bearing. This frequency only occurs with increased bearing clearance. On vertical machines, this increased bearing clearance causes increased bearing compliance and the 1X component will become larger. The 1X change will be evident in the monitored data set.Revision 0Pae3 Page 38 Cooper Nuclear Station Fifth Interval lnservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued)

2. Continue to acquire the old data points with the low-frequency data "for information only" to verify that the system response does not change.III. Attempts to Lower Vibration (Key Component No. 3)CNS installed additional pipe restraints during the preoperational period in order to reduce piping-induced vibrations.

Testing on October 26 and 27, 1973, following the installation of these new supports, demonstrated significantly reduced vibrations. Low-frequency piping-induced vibrations continued, but with reduced amplitude following the installation of the pipe restraints. However, the issue resurfaced in the early 1990s when additional vibration points were recorded, more strict acceptance criteria were adopted for vibrations, and new technology was incorporated into the CNS vibration program.These new points were more influenced by the low-frequency piping-induced vibrations than the one or two points recorded in the 1 980s. It was evident that the piping-induced vibrations were still pievalent with thc CS-P-B pump.In 1993, a deficiency report was written to address increased frequency IST testing of CS-P-B due to vibration. It was suspected that the pump vibrations were piping induced.Preliminary investigation of the vibration issue concluded that cavitation at the CS test return line throttle valve and/or restriction orifices was likely causing the elevated piping vibration in both CS System loops. Vibration testing of the CS piping confirmed this conclusion. To reduce these flow-induced vibrations, DC 94-046 was developed to replace the existing simple, single-stage orifices on both CS subsystem test return lines with multi-stage orifices. Post-installation testing with these multi-stage orifices demonstrated lower vibration levels on CS-P-A, but higher vibration levels on CS-P-B. A multi-hole single-stage orifice was fabricated and installed in the CS-P-B test return line (and later in the CS-P-A test return line) with significantly improved results. Visual observation and vibration data collected during acceptance testing determined that CS-P-B pump vibrations had been reduced, but one direction (location 1H in Figure 1) still demonstrated peak velocity reading in the alert range. The pump vibrations in the 1H direction were occurring at frequencies much lower than the pump operating speed.The major vibration peaks were occurring at approximately 700 (cpm), while the pump speed is at 1780 cpm, indicating that the vibration was piping induced. It was also observed during acceptance testing that vibrations were minimal during operation in the minimum flow condition. IV. Spectral Analysis (Key Component No. 4)Figures 7 through 11 in this relief request show spectrum plots for CS-P-B, as well as spectrum trends. These plots show that the peak energy spikes for points 1H and 5H Revision 0 Page 39 Revision 0 Page 39 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) remain below one-half pump running speed and that the pump vibration signature remains fairly uniform. Figure 12 shows that pump differential pressure is consistently acceptable. This data validates the analysis performed by Machinery Solutions, Inc., and the earlier conclusions that the elevated vibrations are piping induced, and not indicative of degraded pump performance. No pump or motor faults and/or degradation are evident in the spectral analysis for this pump. This test data also shows that the vibrations experienced remain in the region of the CS-P-B pump-motor-piping system natural frequency, at less than half the pump's operating speed.Vibrations occurring at these low frequencies are not expected to be detrimental to the long-term reliability of either the pump or the motor. Typical pump faults, i.e., impeller wear, bearing problems, alignment problems, shaft bow, etc., would result in measurable vibration response in frequencies equal to or greater than one-half of the pump's running speed. Such faults would also be evident in pump trends. However, the vibrations are being expe ienced below one-half pump operating speed, have existed since initial operation, and are not trending higher. Visual inspection by Machinery Solutions, Inc., in 2001 of the pump base plate, soleplate, and grout, identified no visible cracks or degradation. Further, they concluded that the balance condition and shaft alignment of the pump and motor were acceptable, and detected no evidence of motor bearing wear.D. Maintenance History The maintenance history for CS-P-B reflects that there have been no significant work items applicable to CS-P-B due to the low-frequency vibrations that have been experienced since the construction phase of the plant. A review of maintenance history for the CS-P-B pump and motor was performed. The search consisted of a historical review of CS-P-B pump and motor maintenance in addition to a more general search of CS System vibrational issues. This search identified that the pump and motor installed in the plant today is the same combination that was installed during the construction phase of the plant. Some of the key items reviewed arc summarized below: 1. 1973: Additional supports installed on "B" CS System during pre-operational stage. As discussed previously, this resulted in lowering CS-P-B vibrations.

2. January 1977: Vibration eliminator on "B" CS test line, CS-VE7, required tightening of wall plate bolts per Maintenance Work Request (MWR) 77-1-10.Bolts in pipe clamp were replaced and clamp was realigned.

Design was determined to be adequate, but lock washers should be used to prevent recurrence of the problem. MWR 77-1-262 completed this action.Revision 0 Page 40 Revision 0 Page 40 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued)

3. April 1989 (Work Item [WI] 89-0269);

November 1991 (WI 91-1507), February 1993 (MWR #92-2876): CS-P-B stator end turn bracing brackets inspected for stress corrosion cracking or unusual conditions such as loose bolts or bending.No cracks, loose bolts, or other unusual conditions were observed.4. March 1993: A magnetic particle examination of CS-P-B support attachment weld revealed an indication at Lug #5 of the pump support. The indication was ground out, repaired, and retested satisfactorily. The indication was very small and would not have affected the overall stiffness of the pump. In 2003, no recurrence of this indication was identified.

5. April 1993: Work Order #93-1631 was initiated due to mechanical seal leakage.A complete inspection of the pump/motor was also completed.

The pump was found with the keyway not properly aligned with the mechanical seal, causing the leakagc. Thc impeller was found to have minor pitting at the base of the wear ring area. The pump casing and cover had minor erosion and pitting. No significant problems with the pump or motor were noted.6. July 1994: Bolt torque checked for lower end bell and lower bearing housing on CS-P-B motor due to a loose bolt found on the "A" RHIR pump motor. No movement on lower bearing housing bolts. Movement of lower end bell bolts were as follows: 1/16 flat on #1, 3, 4, and 5 and no movement on #2, 6, 7, and 8.These were very minor adjustments.

7. Late 1994: DC 94-046 installs new orifices in CS-P-B test line. As previously discussed, this reduced piping deflections in the test line.8. Oil Samples (Dates: 09-22-95, 10-22-95, 11-24-95, 02-28-97, 03-26-98, 04-05-99, 01-24-00, 12-26-00, 10-28-02, 08-30-04, 01-05-05, 08-14-06, 02-28-07, 08-14-07, 02-11-08, 08-14-08, 02-19-09, 08-12-09, 02-09-10, 08-25-10, 03-11-11, 09-02-11, 12-13-11, 03-02-12, 08-24-12, 02-12-13, 08-13-13, 02-11-14, 08-13-14): Periodic Oil Sample Analysis of the upper and lower motor bearings in accordance with Preventive Maintenance Program. Results of CS-P-B Motor oil analysis were satisfactory with no corrective actions required.9. Numerous Visual Motor Inspections completed satisfactory (i.e., January of 2002): Visual motor inspection satisfactory per Work Order #4199724.10. February 2003: Notification
  1. 10225272 identified an indication approximately 3/8" on a CS-P-B integral attachment (CS-PB-Al).

The indication is at the top of one of the small gusset supports where the gusset is welded to the cast pump Revision 0Pae4 Page 41 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) bowl extension (different spot than the 1993 indication). Within Engineering Evaluation 03-030, the indication was determined to be on the gusset side of the weld and appears to be an incomplete fusion of the weld and not a service load-induced flaw. Poor accessibility was the most likely cause. Engineering Calculation 03-007 demonstrated that, even if the five minor gusset plates were ignored, the pump support is still qualified under the most severe design loads.This search of the maintenance history, covering a time period of approximately forty years, identified no significant maintenance or corrective actions that had to be implemented for the "B" CS pump and motor due to the piping-induced vibrations. Only minor indications were noted on the pump impeller and casing during the last significant motor/pump disassembly in 1993.No other documentation of pump/motor disassembly inspection results was found during this review. Oil analyses of the CS-P-B lower and upper motor bearing housings were found to be satisfactory for all the results documented since 1995 to the present. Wear metals, contaminants, additives, etc., were all at acceptable levels. The addition of pipe supports in 1973 and new orifices in the test lines were necessary modifications and were previously discussed. Other than these modifications, only minor corrections have been made with pipe and/or pump supports (tightening bolts, minor indication, etc.), none of which were found to be significant. Therefore, the maintenance history supports the basis of this relief request in that the piping-induced vibrations occurring on CS-P-B have not degraded the pump or motor in any way.E. Basis for Code Alternative Alert Values for Points 1H and 5H By this relief request, NPPD is proposing to increase the absolute alert limit for vibration points 1H and 5H from 0.325 in/s to 0.400 in/s. The piping-induced vibration, which occurs at low frequencies, occasionally causes the overall vibration value for these two points to exceed 0.325 mI/s, resulting in CS-P-B being on an increased test frequency. However, several expert analyses and maintenance history reviews have shown that this piping-induced vibration has not resulted in degradation to the pump. Additionally, the overall vibration levels have remained steady over the past ~-24.5 years. Therefore, it has been demonstrated that doubling the test frequency under the current conditions does not provide additional assurance as to the condition of the pump and its ability to perform its safety function.These new values are reasonable as they represent an alternative method that still meets the intended function of monitoring the pump for degradation over time while keeping the required action level unchanged. The proposed values encompass the majority of the historical values, but not all of them (reference Figures 2a, 2b, 3a, 3b). With these new values, a reading above 0.400 in/s would require NPPD to place the pump on an increased testing frequency and to evaluate the pump performance to determine the cause Revision 0Pae4 Page 42 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps" and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) of the reading. It is expected that a small amount of degradation occurring in the pump or a slight increase in the piping-induced vibration would be quickly identified with these new parameters. The new alert limits will still allow for early detection of pump degradation or piping-induced vibration increases prior to component failure, while the required action absolute limit will remain at the code value of 0.700 in/s. Therefore, the intent of the code will be maintained. Conclusions Several expert evaluations have documented that no internal pump or motor degradation is occurring due to the piping-induced vibration, which has been present since the pre-operational testing time period. The available vibration data over the past years and differential pressure data over nearly the past -~20 years supports this fact as essentially no degradation has been indicated. A maintenance history review and review of oil analyses results further supports these conclusions. Based on this information, CNS concludes that doubling the test frequency for CS-P-B does not provide additional information nor does it provide additional assurance as to the condition of the pump and its ability to perform its safety function. Testing of this pump on an increased frequency places an unnecessary burden on CNS resources. All four key components discussed in NUREG/CP-01 52 have been addressed in detail, supporting the alternative testing recommended in this relief request.CNS concludes that CS-P-B is operating acceptably and will perform its safety function as required during normal and accident conditions. The increased alert limits proposed for vibration points 111 and 511 in this relief request will continue to assure long-term reliability of CS-P-B.During the performance of CS-P-B inservice comprehensive pump testing, or any other time vibrations are recorded to determine pump acceptability (i.e., post-maintenance testing, other periodic testing, etc.), pump vibration shall be monitored in accordance with ISTB-3510(e) and ISTB-3540(a). The acceptance criteria for vibration points 2H, 3H, and 1V will follow the criteria specified in ISTB Table ISTB-5121-1. The acceptance criteria of vibration points 1 H and 5H1 will have increased absolute alert limit values of 0.400 in/s. The absolute required action limits for all points will continue to be 0.700 in/s in accordance with ISTB Table ISTB-5121-1. The absolute alert and required action limits for all vibration points associated with CS-P-B are summarized in the table below.Revision 0Pae4 Page 43 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Absolute Vibration Acceptance Criteria for CS-P-B: Vibration Acceptable Range Alert Range Required Action Parameter Range 111 < 0.400 in./sec. > 0.400 in./sec. > 0.700 in./sec.5H1 < 0.400 in./sec. > 0.400 in./sec. > 0.700 in./sec.1V < 0.325 in./sec. > 0.325 in./sec. > 0.700 in./sec.2H1 < 0.325 in./sec. > 0.325 in./sec. > 0.700 in./sec.3H < 0.325 in./sec. > 0.325 in./sec. > 0.700 in./sec.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was previously approved for the fourth ten-year interval at CNS as Relief Request RP-07 (TAC Nos. MC8837, MC8975, MC8976, MC8977, MC8978, MC8979, MC8980, MC8981, MC8989, MC8990, MC8991, and MC8992, June 14, 2006).Revision 0 Page 44 Revision 0 Page 44 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) CS-P-B Figures Figure Description Attach. 2 Number Page Number 1 CS-P-B Vibration Monitoring Points 46 2a CS-P-B Vibration Point lH from November 1996 to the Present 47 2b CS-P-B Vibration Point 11H from April 1990 to the Present 48 2c Trend of Vibration Point 1 H with Data Below One-Half Pump 49 Running Speed Filtered from May 2010 to the Present 3a CS-P-B Vibration Point 5H from November 1996 to the Present 50 3b CS-P-B Vibration Point 5H from April 1990 to the Present 51 3c Trend of Vibration Point 5H with Data Below One-Half Pump 52 Running Speed Filtered from February 2010 to Present 4a CS-P-B Vibration Point 1V from November 1996 to the Present 53 4b CS-P-B Vibration Point 1V from April 1990 to the Present 54 5a CS-P-B Vibration Point 2H from November 1996 to the Present 55 5b CS-P-B Vibration Point 2H from April 1990 to the Present 56 6a CS-P-B Vibration Point 3H from November 1996 to the Present 57 6b CS-P-B Vibration Point 3H from April 1990 to the Present 58 7 Spectral Trend for Vibration Point 111 59 8 Spectral Trend for Vibration Point 5H 60 9 Spectral Trend for Vibration Point 1 V 61 10 Spectral Trend for Vibration Point 2H 62 11 Spectral Trend for Vibration Point 3H 63 12 CS-P-B Differential Pressure since January 1995 to the Present 64 Revision 0Pae4 Page 45 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) 1V 5H-~3 H--I DISCHARGE,----SUCTION 62CS101A Figure 1 CS-P-B Vibration Monitoring Points Revision 0 Page 46 Revision 0 Page 46 Cooper Nuclear Station F~fih Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) ~56 9 V~(1H1~*~t ~~~14~2O0O 01 4~6Q#II i Date Figure 2a CS-P-B Vibration Point 1H1 from November 1996 to the Present Revision 0 Page 47 Revision 0 Page 47 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Figure 2b CS-P-B Vibration Point 11H from April 1990 to the Present Revision 0 Page 48 Revision 0 Page 48 Cooper Nuclear Station JFih Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) CORE SPRAY PUMP MOTOR B I 1H- -MOTOR UPPR HORIZO'NTAL SOUTIfH(H0) 14 83-1K]Hz 1ST-Baseline-Value: 0.1 35 5118/1198* AM r- 2f12.12015 15 i 20:11PMd Amn0: 1r 20 O02e-1112412012 5:32:33 PM 0392 V -DG Pk = 0 394 LOA=D -100.00 RPN-= 1780.0 (2967liz)021 -I ,,"0.14-0.07 -L I 0 -~ _____________________________________ Frm7V113 I 0 1o00o 20000 30000 Frequency (CPM)00000 o00000 rct32.10 Amp: 0.00019 List of Trend Points Station: REACTOR BUILDING Machine: CS-MOT-B --> CORE SPRAY PUMP MOTOR B Meas Point: 1H --> MOTOR UPPR HORIZONTAL SOUTH (H01)Parameter: 14.83-1KHZ (PK Velocity in In/Sec)Date Time 10-May-10 14:21 16-Nov-10 13:09 21-Apr-11 15:06 24-May-11 13:43 23-Nov-11 13:32 23-May-12 10:51 24-Nov-12 17:32 Value.103.109.135.110.122.125.113 Date Time Value 16-May-13 11:28 .109 13-Nov-13 12:35 .108 13-May-14 22:51 .124 14-Aug-14 10:55 .112 12-Oct-14 02:08 .130 10-Nov-14 23:21 .120 12-Feb-15 23:20 .120 Figure 2c Early Warning Limits--Alert Limit Values--Fault Limit Values--.169.300.700 Trend of Vibration Point 1H1 with Data Below One-Half Pump Running Speed Filtered from May 2010 to the Present0 Page 49 Cooper Nuclear Station F/ith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Figure 3a CS-P-B Vibration Point 511 from November 1996 to the Present Revision 0 Page 50 Revision 0 Page 50 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RIP-07 Core Spray Pump B Vibration Alert Limits (Continued) Figure 3b CS-P-B Vibration Point 5H from April 1990 to the Present Revision 0 Page 51 Revision 0 Page 51 Cooper Nuclear Station Ffi~h Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) CORE SPRAY PUMP MOTOR B81 5H -MOTOR UPPR HORIZONTAl. W.'EST (H05)40 C ED 0~14.83-1KI-4z 1ST-Baseline -Value: 0.269 5/11811 998 1 2:00 AM 211 212 015 1 1:21:36 Ph Amp. 0.266 021-211212015 1121;36 PM 0.374 V -ID Pk 0 0374 LOAO- 100 00 RPM = 1780.J0 (29.67 lIZ).0-14-.1 01 I .~10000 20000 30000 Frequency (CPM)80000-I F~eci31~4 80000 04d.1 ~'9Amp 0 0O~504 List of Trend Points Station: REACTOR BUILDING Machine: CS-MOT-B --> CORE SPRAY PUMP MOTOR B3 Meas Point: 5H --> MMOTOR UPPR HORIZONTAL WEST (H05)Parameter: 14.83-1KHZ (PK Velocity in in/Sec)Date Time 09-Feb-10 14:48 10-Feb-10 10:04 10-May-10 14:21 16-Nov-10 13:09 21-Apr-11 15:06 24-May-11 13:43 23-Nov-11 13:33 23-May-12 10:52 Value.252.252.238.252.238.201.215.232 Date Time 24-Nov-12 17:32 16-May-13 11:30 13-Nov-13 12:35 13-May-14 22:52 14-Aug-14 10:56 12-Oct-14 02:09 10-Nov-14 23:22 12-Feb-15 23:21 Value.229.223.243.262.224.229.250.266 Early Warning Limits--Alert Limit Values--Fault Limit Values--.309.300.700 Figure 3c Trend of Vibration Point 5H with Data Below One-Half Pump Running Speed Filtered from February 2010 to the Present Revision 0 Page 52 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Figure 4a CS-P-B Vibration Point 1V from November 1996 to the Present Revision 0Pae3 Page 53 Cooper Nuclear Station F~th Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Figure 4b CS-P-B Vibration Point 1V from April 1990 to the Present Revision 0 Page 540 Page 54 Cooper Nuclear Station F~fih Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Figure 5a CS-P-B Vibration Point 2H from November 1996 to the Present Revision 0 Page 55 Revision 0 Page 55 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Figure 5b CS-P-B Vibration Point 2H from April 1990 to the Present Revision 0 Page 56 Revision 0 Page 56 Cooper Nuclear Station F~ith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Figure 6a CS-P-B Vibration Point 3H from November 1996 to the Present Revision 0 Page 57 Revision 0 Page 57 Cooper Nuclear Station Ffifh Interval Inser-vice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued)

  • A Vb~3N)A*P I U v'~ ~O(~sc)0 ror.miniq c~-p ~ I95~' ~o~.I 041 S 11~U1S96 I6fl4~OOO 0tiZ~O04 04~~I201I t1fl~Z0t4Figure 6b CS-P-B Vibration Point 3H from April 1990 to the Present Revision 0 Page 58 Revision 0 Page 58 Cooper Nuclear Station Ffi~h Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued)

CORE SPRAY PUMP MOTOR B / 1H -MOTOR UPPR HORIZONTAL SOUTH(H01) r-) 0 2112/2015 11:20:11 PM SRoute 0.302 V-DG PR= 0.302 LOAD= 100.00 RPMl= 1700.0.. .. (29.67 Hz)-i Freq:1 26060 60000 Ord.0 708 mp: 0.00644 30000 Frequency (CPM)CORE SPRAY PUMP MOTOR B / 1H -MOTOR UPPR HORIZONTAL SOUTH(H01) a)CC, C C C a)0~l-0.28;I-I k!0.28', 2/12/2016~11/10/2014 _/1/o,,20?14 6//13/2014-I 1/13/2013 616/2013/11/24/2012 6/¢523/2012 lLi , ,11/23/'2011 ,/24/2011 42/2011/11/16/2010 6110/20"10 60000l -- m¸ = I m I It-LL ..t 16000 30000 Frequency (CPM)46000 Figure 7 Spectral Trend for Vibration Point 1H Revision 0 Page 59 Revision 0 Page 59 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) CORE SPRAY PUMP MOTOR B / 5H -MOTOR UPPR HORIZONTAL WEST (H05)028.021 C_).'0.14 0.07.2/12/2015 11:21:36 PM Route 0.374 v-Do Pk= 0.374 LOAD = 100.00 RPM=1 780.0..... (29.67 Hz)L...Ut. ., ..a 10000 20000 306oo Frequency (CPM)40000 506o0 i~~ Freq:1 250.0 80000 0'rd.0.08 Amp; 0 00770 CORE SPRAY PUMP MOTOR B 1 5H -MOTOR UPPR HORIZONTAL WEST (HO5)t-0.28 U)C,)C C 0 U)V~1L~1 I to *t'U)-~I~1=~ I.. I*Illi .1 I.1i~C.11 t 1 .1 III I I I LI:1 t1.~,.0.28 -II b i J A l.1m I I-I 211212015/ 11/1012014 S10112120 14.. 5/1312014.'I 11/13.r2013 ,5/11/2013 .11124/20125,2312012 111231"2011 j"5,2412011 S412112011 11/1612010 S5/101'2010 L L II ,J h 1= .11 S II A '.i~ -Ph KI..J.. I I, I 2/1212015 11:21 PM RPM: 1780.0 Freq: 59810.6 Ord:33.60 Amp: 0.0001 II [ ............ v r I 0 100003 20000 ao00o Frequency (CPM)400(o 6o6000( 60000 Figure 8 Spectral Trend for Vibration Point 5H Revision 0 Page 60 Revision 0 Page 60 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) CORE SPRAY PUMP MOTOR B /1V -MOTOR UPPR AXL, .(V01)0.2.2/12/2015 11:19.19 PM Route 0.191 V-DO Pk= 0.191 LOAD =100 00 RPM= 1-I780.0 (29.67 Hz)0.15-5, C U 0 15 0~cl.1 0.05 LL t.J~-. -.10000 3000 0~requency (CPM)40600 so6oo Freq5e6~,8 60000 Ord.32.96 Aii~p 0000~CORE SPRAY' PUMP MOTOR B / lV -MOTOR UPPR AXLAL. (V0l)-0.2-0O.1 0.1-I'F-~r I---- -&L .1 Ii I I*1 I 2/12f2Qi~11/10/2014 10/12/2014 8/14'2014 9/12/2014 1111 ~'2O 1 Z~9/19/2013 11/2412012 6t22t2012 11/2312011 ~/24t2O1 i 4/21/2011 11119/2010 0/10/2010 I.1 1 11 i II~---~- I ViY..--.L, I I I1'-~L.2t12.,201S 11.19 PM RPM: 0 Freq:5901 0 600001-- -I i ii J i [ i f 0 ICMXX0 20000 30X000 Frequency (CPMI 40000 50000 500000 Figure 9 Spectral Trend for Vibration Point 1V Revision 0 Page 61 Revision 0 Page 61 Cooper Nuclear Station F~fih Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) CORE SPRAY PUMP MOTOR 0 / 2H -MOTOR LWR HORIZONTAL SOUTH(H02) 2/12/2015 11:22:21 PM 0.08--0.67-S0.04-S00027-0.013.]Route 0.104 V-DG Pk=0.105 LOAD = 10000 RPM = 1780.0 (29.67 Hz)L L4~L I[II ; ---_+. ...... ., ,L , + , ,+,.,.0 ioo 20000 30000 40000 5000 60000 Frequency (CPM)CORE SPRAY PUMP MOTOR B ! 2H -MOTOR LWdR HORIZONTAL SOUTH(H02) ._)O0¢.#'0 0 U)2/i41 2/1 2015 11/10/2014 7012/2014 812014 2/1 2/2015 11:22 PM RPM: 1780.0 Freq:59968.1 Ord:33.69 Amp: 0 0005 3 30000 40000 50000 50000 Frequency (CPM)Figure 10 Spectral Trend for Vibration Point 2H Revision 0 Page 62 Revision 0 Page 62 Cooper Nuclear Station Ffi~h Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) RX -CORE SPRAY PUMP MOTOR B CS-MOT-B -3H MOTOR LWR HORIZONTAL WEST (H03)0.15-017-2112(2015 11:23:07 PM... Route 0.183 V -DG Pk = 0.183.LOAD = 100 00 RPM = 1780.0 29,67 Hz)o0.06-(]3 0.03-Frecj £9F5:3 0 D~d 33 62 Amp C ['6039 Frequency (C PM)CORE SPRAY PUMP MOTOR B I 3H -MOTOR LWR HORIZONTAL WEST (H03)U)C,, C C 0 U)2112/2015/, t 810/1212014 611412014 5/132014 11/13/'201 -0,1 S1112412012 512312012/1112312011 4J/2/24/2011I 4212011"11/1182010 211 2"2015 11:23 PM RPM: 1780.0 Freq: 59968 .1 0rd:33 69 Amp: 00004 Frequency (CPM)Figure 11 Spectral Trend for Vibration Point 311 Revision 0 Page 63 Revision 0 Page 63 Cooper Nuclear Station Fifth lnterval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Cooper Nuclear Station F~ih Interval Inservice Testing Program for Pumps and Valves Relief Request RLP-07 Core Spray Pump B Vibration Alert Limits (Continued) Figure 12 CS-P-B Differential Pressure Since January 1995 to the Present Revision 0 Page 64 Revision 0 Page 64 Cooper Nuclear Station Ffifh Interval Inser-vice Testing Pro gram for Pumps and Valves Relief Request RP-08 Comprehensive Pump Test Upper Limit Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected REC-P-A/B/C/D Reactor Equipment Cooling Pumps RHR-P-A/B/C/D Residual Heat Removal Pumps SW-P-A/B/C/D Service Water Pumps 2. Applicable Code Edition and Addenda ASME CM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTB-5123 "Comprehensive Test Procedure," (e), refers to Table ISTB-5121-1, which utilizes a multiplier of 1.03 times the reference value for the comprehensive pump test's upper "Acceptahie, Range" and "Required Action Range High" criteria.ISTB-5223 "Comprehensive Test Procedure," (e), refers to Table ISTB-5221-1, which utilizes a multiplier of 1.03 times the reference value for the comprehensive pump test's upper "Acceptable Range" and "Required Action Range High" criteria.ISTB-5323 "Comprehensive Test Procedure," (e), refers to Tables ISTB-5321-1 and ISTB-5321-2, both of which utilize a multiplier of 1.03 times the reference value for the comprehensive pump test's upper "Acceptable Range" and "Required Action Range High" criteria.4. Reason for Request Occasionally, NPPD has had some difficulty with implementing the high required action range limit of 1.03% above the established hydraulic parameter reference value due to normal data scatter. NPPD has had to address an inoperability of a pump on at least two occasions during the fourth ten-year interval in which a pump was declared inoperable during a comprehensive pump test due to exceeding this upper limit. The result was that the plant had to enter (or remain in) an applicable Technical Specification Limiting Condition for Operation (LCO) for reasons other than a pump degradation issue.Based on the similar difficulties experienced by other Owners, ASME OM Code Case OMN-1 9 was developed and has been published in the 2011 Addenda of the ASME CM Code. The white paper for this code case, Standards Committee Ballot 09-6 10, record 09-657, discussed the impact of instrument inaccuracies, human factors involved with setting and measuring test parameters, readability of gauges and other miscellaneous factors on the ability to meet the 1.03% acceptance criteria. Industry operating experience is also discussed in the white paper.Revision 0 Page 65 Revision 0 Page 65 Cooper Nuclear Station Ffith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-08 Comprehensive Pump Test Upper Limit (Continued) Code Case OMN-1 9 has not yct bcen approved for use in RG 1.192, "Operations and Maintenance Code Case Acceptability, ASME OM Code." Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(1), relief is requested from the multiplier of 1.03 times the reference value for the comprehensive pump test's upper "Acceptable Range" and "Required Action Range High" criteria, referenced in Tables ISTB-5 121-1, ISTB-5221-1, ISTB-5321-1, and ISTB-5321-2. The proposed alternative would provide an acceptable level of quality and safety.5. Proposed Alternative and Basis for Use CNS proposes to use the ASME OM Code Case OMN-19 as published in the 2011 Addenda of the ASME OM Code for the fifth ten year interval IST Program. The ASME OMN-19 Code Case allows for the use of a multiplier of 1.06 times the reference value in lieu of the 1.03 multiplier for the comprehensive pump test's upper "Acceptable Range" criteria and "Required Action Range, High" criteria referenced in the applicable ISTB test acceptance criteria tables ISTB-5121-1, ISTB-5221-1, ISTB-5321I-1 and ISTB-5321-2. The bases for the approval of OMN-1 9, as discussed in the Standards Committee Ballot white paper, are summarized below: 1) Instrument inaccuracies of measured hydraulic value.2) Instrument inaccuracies of set value and its effect on measured value.3) Instrument inaccuracies and allowed tolerance for speed.4) Human factors involved with setting and measuring flow, DiP, and speed.5) Readability of Gauges based on the smallest gauge increment.

6) Miscellaneous Factors.These inaccuracies may cause the measured value to exceed the existing code allowed comprehensive pump test's upper "Acceptable Range" criteria and the "Required Action Range, High" criteria of 3%. The new upper limit of 6%, as approved in Code Case OMN-19, will eliminate declaring the pump inoperable and entering an unplanned Technical Specification LCO or will eliminate the extension of an existing LCO.As a condition for using OMN- 19, CNS will implement a pump periodic verification (PPV) test program to verifyr that a pump can meet the required differential (or discharge) pressure, as applicable, at its highest design basis accident flow rate, as discussed in Mandatory Appendix V, which was published in the 2012 Edition of the ASME OM Code. CNS will not be required to perform a PPV test if the design basis accident flow rate in the licensee's safety analysis is bounded by the comprehensive pump test or Group A test. Also, if a pump does not have a design basis accident flow rate, then a PPV test is not required.

Therefore, any IST pump that is utilizing the 1.06 multiplier for the comprehensive pump test will meet this condition. On June 30, 2015, in a response to the Nuclear Regulatory Commission's Request for Additional Information, CNS provided the following list of pumps that RP-08 is applicable to along with the requested information. The last column of the table also indicates which pumps will have a pump periodic verification (PPV) test based on the current design basis accident flow rate and the Revision 0 Page 66 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RP-08 Comprehensive Pump Test Upper Limit (Continued) current comprehensive pump test flow rate. As is required by the 2012 ASME GM Code, Mandatory Appendix V, Section V-3000(e), the basis for the PPV test parameters will be documented by the owner.Instrument inaccuracies associated with the PPV test parameters will be accounted for within the safety analyses and/or within the test acceptance criteria. CNS considers this a clarification to Mandatory Appendix V, Section V-3000(f), which states that the owner shall account for the pump periodic verification test instrument accuracies in the test acceptance criteria. Although not expected, any flow rate changes associated with this table would be available for NRC inspection, upon request.RP-08 IST Class 1, 2, and 3 Applicable Pumps Pump Pump Pump Type ASME ASME Design Basis IST PPV Test Name Number .Code CM Code Accident Flow Comprehensive Required Class Category Rate (gallons Pump Test (Yes/No)per minute) Flow Rate (gallons per________minute) Reactor Horizontal Equipment REC-P- centrifugal 3 GruA41 Cooling A/B/C/D pump GopA461100 No Pumps Residual Vertical Heat RHR-P- centrifugal 2 Group A 7700 7800 No Removal A/BIC/D pump Pumps ____Service Vertical line Water 5500-Ye Pumps A/B/C/D shaft pump 3 Group A 5846 50 e Using the upper limit of 1.06 times the reference value in lieu of the 1.03 multiplier for the comprehensive pump test's upper "Acceptable Range" criteria and "Required Action Range, High" criteria referenced in the applicable ISTB test acceptance criteria tables will provide an acceptable level of quality and safety.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was approved for the fourth ten-year interval at Columbia Generating Station as RP-06 (TAC Nos. MF3847, MF3848, MF3849, MF3851, MF3852, MF3853, MF3854, MF3855, MF3856, MF3857, and MF3858, December 9 and February 9, 2015).Revision 0 Page 67 Cooper Nuclear Station Fifth Interval Inserviee Testing Program for Pumps and Valves Relief Request RP-09 Variance Around the Reference Values Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected CS-P-A/B HPCI-P-MP &HPCI-P-BP RCIC-P-MP REC-P-A/B/C/D RHR-P-A/B/C/D SW-P-AIB/C/D SW-P-BPA/B/C/D Core Spray Pumps High Pressure Coolant Injection Main & Booster Pumps Reactor Core Isolation Cooling Main Pump Reactor Equipment Cooling Pumps Residual Heat Removal Pumps Service Water Pumps Residual Heat Removal Service Water Booster Pumps 2. Applicable Code Edition and Addenda ASME GM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTB-5 121 Group A Test Procedure ISTB-5 122 Group B Test Procedure ISTB-5 123 Comprehensive Test Procedure ISTB-5221 Group A Test Procedure ISTB-5222 Group B Test Procedure ISTB-5223 Comprehensive Test Procedure ISTB-5321 Group A Test Procedure ISTB-5322 Group B Test Procedure ISTB-5323 Comprehensive Test Procedure 4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(l), an alternative is proposed to the pump testing reference value requirements of the ASME OM Code. The basis of the request is that the proposed alternative would provide an acceptable level of quality and safety.For pump testing, there is difficulty adjusting system throttle valves with sufficient precision to achieve exact flow reference values during subsequent IST exams. Section ISTB of the ASME GM Code does not allow for variance from a fixed reference value for pump testing. However, Revision 0 Page 68 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-09 Variance Around the Reference Values (Continued) NUJREG-1482, Revision 2, Section 5.3, acknowledges that certain pump system designs do not allow for the licensee to set the flow at an exact value because of limitations in the instruments and controls for maintaining steady flow.ASME OM Code Case OMN-21 provides guidance for adjusting reference flow/DP to within a specified tolerance during Inservice Testing. The Code Case states "It is the opinion of the Colmmittee that when it is impractical to operate a pump at a specified reference point and adjust the resistance of the system to a specified reference point for either flow rate, differential pressure or discharge pressure, the pump may be operated as close as practical to the specified reference point with the following requirements. The Owner shall adjust the system resistance to as close as practical to the specified reference point where the variance from the reference point does not exceed +2% or -1% of the reference point when the reference point is flow rate, or +1% or -2% of the reference point when the reference point is differential pressure or discharge pressure." 5. Proposed Alternative and Basis for Use CNS seeks to perform Inservice Pump testing in a manner consistent with the requirements as stated in ASME OM Code Case OMN-2 1. Specifically, testing will be performed such that flow rate is adjusted as close as practical to the reference value and within proceduralized limits not to exceed +2%/-l1% of the reference value. Or, if differential pressure or discharge pressure is set, then it will be set as close as practical to the reference value and within proceduralized limits not to exceed +l%/-2% of the reference value.CNS plant operators will still strive to achieve the exact test flow reference values during testing.Typical test guidance will be to adjust flow to the specific reference value. If necessary, additional guidance will be provided such that if the reference value cannot be achieved with reasonable effort, the test will be considered valid if the steady state reference value is within the procedural limits. The procedural limits will be carefully determined on a case by case basis, and will not exceed the limits provided in Code Case OMN-2 1. The test will be considered valid if the steady state reference value is within the proceduralized limits of the procedure. On June 30, 2015, in response to the Nuclear Regulatory Commission's Request for Additional Information, CNS provided the following list of pumps that RP-09 is applicable to along with the requested information. Revision 0 Page 69 Revision 0 Page 69 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-09 Variance Around the Reference Values (Continued) RP-09 IST Class 1, 2, and 3 Applicable Pumps Pump Name Pump Number Pump Type ASME ASME OM Code Code Category____ ___ ___ ___ ___ _ __ ___ ___ ___ ___Class Core Spray Pumps CS-P-A/B Vertical centrifugal pump 2 Group B High Pressure Coolant HCIPM Tubndrvnhizta2GopB Injection Main & HPCI-P-BP cerbnt rivega pump ota Booster Pumps HC--P cnrfglpm ru Rctorolin M soatinoum Turbine driven horizontal2GruB ReCtororen IsoationCPCmpM centrifugal pump Reactor Equipment REC-P-AIB/C/D Horizontal centrifugal pump 3GruA Cooling Pumps3GruA Residual Heat Removal RIR.-ABCD Vertical centrifugal pump2GruA Pumps ______Service Water Pumps SW-P-.A/B/CiD Vertical line shaft pump 3 Group A Residual Heat Removal Service Water Booster SW-P-BPA/B/C/D Horizontal centrifugal pump 3 Group A Pumps Using the provisions of this request as an alternative to the specific requirements of ISTB-5 121, ISTB-5122, ISTB-5123, ISTB-5221, ISTB-5222, ISTB-5223, ISTB-5321, ISTB-5322, and ISTB-5323 as described above will provide adequate indication of pump performance and continue to provide an acceptable level of quality and safety.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was approved for Callaway for their fourth ten-year interval as PR-06 (TAC Nos. MF2784, MF2785, MF2786, MF2787, MF2788, and MF2789, July 15, 2014).Revision 0 Page 70 Revision 0 Page 70 Cooper Nuclear Station F~fih Interval Inservice Testing Pro gram for Pumps and Valves ATTACHMENT 3 AUGMENTED PUMP RELIEF REQUESTS AUGMENTED PUMP RELIEF REQUEST INDEX Relief 1Description 1CNS Approval Request No. JJDate ARP-01 ]jElevated Release Point Sump Pump Testing ]I 3-1-2016 ALRP -02 ][Standby Liquid Control Pump Vibration Accuracy ][ 3-1-2016 ARP-03 ]jStandby Liquid Control Pump Testing ]j 3-1-2016 ARP-04 ][Diesel Generator Fuel Oil Transfer Pump Testing ]j 3-1-2016 Revision 0Pae1 Page 71 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request ARP-O1 Elevated Release Point Sump Pump Testing Alternative Provides Acceptable Level of Quality and Safety 1. Augmented Code Component(s) Affected RW-P-Zl Elevated Release Point Sump Pump A RW-P-Z2 Elevated Release Point Sump Pump B 2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004 Edition through 2006 Addenda 3. Applicable Code Reouirement ISTB-3000, "General Testing Requirements." This includes Preservice and Inservice testing of pumps.ISTB-5000, "Specific Testing Requirements." This includes the requirements of a Group A tesit, Comprehensive Pump Test, and Preservice Test.ISTB-6000. "Monitoring, Analysis, and Evaluation. This includes actions to take based on alert and required action levels.4. Reason for Req~uest Augmented Relief is requested from the requirements of ISTB-3000, IST-5000, and ISTB-6000. The proposed alternative provides an acceptable level of quality and safety. This Augmented Relief Request does not require NRC approval.5. Proposed Alternative and Basis for Use Elevated release point sump pumps (Z-Sumps) remove water from the sump supporting the Standby Gas Treatment system drains. Pump failure could cause SGT drain lines to backup and interfere with SGT operation. These pumps operate intermittently depending on sump level. Testing requires manually providing sufficient sump inventory to facilitate pump operation. The pumps receive automatic actuation signals from sump level switches. Pump testing with insufficient inventory would result in damage to the pumps. Additionally, the pumps are submerged and water cooled.Suction pressure, differential pressure, flowrate, and vibration measurements are not feasible due to inaccessibility, the short time the pump runs, lack of available test instrumentation, and the change in suction pressure throughout the test.Testing shall be performed quarterly, utilizing pump start and stop level switches and measuring the time (pump run time) it takes to pump a specified quantity of fluid from the sump. Since the quantity of fluid is constant, pump run time, TM, shall be the test parameter that is measured each test and compared with the corresponding acceptable, alert and action limits. The acceptable Revision 0 Page 72 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request ARP-O1 Elevated Release Point Sump Pump Testing (Continued) range shall be 0.80 to 1.20'1TMr (T1Mrreference run time); the alert range low shall be <0.80 TMr; the alert range high shall be >1.20 TMr to 1.50 TMr; and the required action range shall be>1.50 TMr, not to exceed the maximum pump run time value documented in the Cooper Nuclear Station Inservice Testing Program Basis Document. If run time falls within the alert range, the test frequency shall be doubled until the cause of the deviation is determined and the condition corrected by repair, replacement or an evaluation which resolves the condition. If run time falls into the required action range, the pump shall be declared Non-Functional until the cause of the deviation has been determined and the condition corrected by repair, replacement or an evaluation which resolves the condition. These pumps are ASME non-code class pumps outside the scope of the IST Program. This method of monitoring these pumps provides a level of testing that is commensurate to the level of safety for these components.

6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents A similar relief request was previously approved by CNS for the third and fourth ten-year intervals as ARP-0 1.Revision 0 Page 73 Revision 0 Page 73 Cooper Nuclear Station Ffith Interval Inser-vice Testing Pro gram for Pumps and Valves Relief Request ARP-02 Standby Liquid Control Pump Vibration Accuracy.Alternative Provides Acceptable Level of Quality and Safety 1. Aug~mented Code Component(s)

Affected SLC-P-A Standby Liquid Control Pump A SLC-P-B Standby Liquid Control Pump B 2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004 Edition through 2006 Addenda 3. Applicable Code Requirement Table ISTB-35 10-1, "Required Instrument Accuracy" and ISTB-35 10(a), "Accuracy." Vibration instruments require an accuracy of+/- 5% over the calibrated range.4. Reason for Request Augmented Relief is requested from the requirements of table ISTB-35 10-1 and paragraph ISTB-3510(a) for vibration measurement accuracy. The proposed alternative provides an acceptable level of quality and safety. This Augmented Relief Request does not require NRC approval.5. Proposed Alternative and Basis for Use The SLC pumps function to pump a boron neutron absorber solution into the reactor if the reactor cannot be shut down or kept shut down with control rods. The SLC System also has a safety-related accident mitigation function based on the implementation of the Alternate Source Term (AST) methodology as the radiological source term design basis accident analysis (in accordance with RG 1.183). SLC is credited for controlling the pH of the water in the Suppression Pool, Reactor Vessel (Rx), and Core Cooling systems following a Design Basis LOCA.The Code requires vibration equipment to meet the accuracy of +/- 5% across the frequency response range, which includes the minimum frequency response of 1/3 pump shaft speed. For the SLC pumps, this is 173.3 rpm or 2.8 Hz. The vibration meters used at CNS are calibrated to meet the code down to and including 5 Hz. Currently, there are no calibration points being taken below 5 Hz. Therefore, the accuracy below 5 Hz. may not meet the code tolerance. The average velocity for an IST test is a single, average energy reading. The effect of this potential change in accuracy below 5 Hz., when averaged into the overall reading, is quite small.It would only be a concern if a single frequency in the spectrum were being evaluated between 2 -5 Hz. Furthermore, detection of pump degradation via vibration data is based on changes in vibration measurement from one test to another. Thus, if the calibration accuracy is consistent, then the change in vibration measurement from one test to another is appropriate information for trending purposes. Therefore, existing vibration equipment will provide adequate trending information and may be used for SLC pump vibration data collection. These pumps are ASME non-code class pumps outside the scope of the 1ST Program.Revision 0 Page 74 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Relief Request ARP-02 Standby Liquid Control Pump Vibration Accuracy (Continued) Vibration data for the SLC pumps will be taken with equipment calibrated fr'om 5 Hz. to at least 1000 Hz. at +/- 5% or better, and will not be calibrated to +/- 5% or better below 5 Hz.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents A similar relief request was previously approved by CNS for the third and fourth ten-year intervals as ARIP-02.Revision 0Pae7 Page 75 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request ARP-03 Standby Liquid Control Pump Testing Alternative Provides Acceptable Level of Quality and Safety 1. Augmented Code Component(s) Affected SLC-P-A Standby Liquid Control Pump A SLC-P-B Standby Liquid Control Pump B 2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTB-3400, "Frequency of Inservice Tests" JSTB-53 10, "Preservice Testing" ISTB 5323, "Comprehensive Test Proccdurc" 4. Reason for Request Augmented Relief is requested from the requirements of ISTB-3400, ISTB-53 10, and ISTB-5323. The proposed alternative provides an acceptable level of quality and safety. This Augmented Relief Request does not require NRC approval.5. Proposed Alternative and Basis for Use The SLC pumps function to pump a boron neutron absorber solution into the reactor if the reactor cannot be shut down or kept shut down with control rods. The SLC System also has a safety-related accident mitigation function based on the implementation of the Alternate Source Term (AST) methodology as the radiological source term design basis accident analysis (in accordance with RG 1.183). SLC is credited for controlling the pH of the water in the Suppression Pool, Reactor Vessel (Rx), and Core Cooling systems following a Design Basis LOCA.Each SLC system pump shall be capable of delivering no less than 38.2 gal/min against a system head of 1300 psig to be considered operable. This flow rate is based on the original system design requirement that a single standby liquid control pump be capable of shutting down the reactor from the most reactive condition at any time in core life and maintaining it subcritical during cooldown with all control rods withdrawn in the rated power pattern.The Standby Liquid Control pumps are categorized as Group B pumps since they are standby emergency pumps and are only operated for testing. They are horizontally-mounted reciprocating positive displacement pumps.As an alternative to the code requirement for performing a comprehensive pump test, each of these pumps will have a modified Group A test performed each quarter in place of the Group B quarterly test and the 2-yr Comprehensive pump test. The pumps will be operated at a reference Revision 0 Page 76 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Relief Request ARP-03 Standby Liquid Control Pump Testing (Continued) discharge pressure (Pr) of 1300 psig with pump flow rate (Q) measured and compared to the required action and alert range requirements of Table ISTB-5321 -2 for the group A test, which is more stringent than the Group B testing that would normally be applied each quarter. In addition, vibration measurements will be recorded every 6 months (every other quarter) rather than only once every 2 years during the comprehensive pump test. Vibration measurements will be compared to the range requirements of Table ISTB-532 1-2 for the Group A test. Corrective actions will be taken in accordance with ISTB-6200. Permanently installed plant instrumentation will be used to determine discharge pressures and flow rates. Portable vibration instruments will be used to determine vibration measurements. All instrumentation will meet the accuracy requirements of a Group A test unless specific relief is requested. This level of accuracy is sufficient for these augmented pumps, especially based on the large margin to the minimum flow of 38.2 gpm per pump. However, the current discharge pressure gauge is calibrated to +/-1/2%, which meets the accuracy requirements of the Group A test (+/-2%) and the Comprehensive pump test (+/-1/2%), so no variabilities between the Group A and Comprehensive Pump Test instrumentation currently exist.One of thie requiirements of the comprehensive test is to perform the test at oubotoantial flow (+/-20% of design flow). CNS will meet this requirement each quarter by performing the test at the design flow discharge pressure: Design Point: 1300 psig with a minimum of 38.2 gpm Test Point: 1300 psig with a minimum of 38.2 gpm Although these are Group B pumps, the OM Code allows the substitution of a Group A or comprehensive test. CNS will perform a modified Group A test as stated above such that the acceptance criteria for hydraulic performance will meet the code requirements for a Group A test.Additionally, CNS will perform vibration monitoring on these Group B pumps on a frequency of once every 6 months.The Standby Liquid Control pumps are tested at a set discharge pressure of 1300 psig. Per Table ISTB-5321 -2, the required action range for the Group A flow measurement would be <0.93Qr and > 1.lOQr, with an alert range of 0.93Qr to < 0.95Qr. This is the same as the comprehensive test requirement with the exception that the upper range is >1l.O3Qr. With reference values of approximately only 53-54 gpm, this upper limit for the comprehensive pump test may not encompass the normal data scatter associated with acceptable SLC pump operation. Therefore, CNS will monitor the flow of these pumps at the Group A test criteria each quarter as follows: Acceptable Range 0.95 to 1.10 Qr Alert Range 0.93 to <0.95 Qr Required Action <0.93 Qr or >1.10 Qr CNS will evaluate all ranges against the design conditions to ensure that all procedure lower limits bound the more conservative of the design or ASME GM Code ranges delineated above.Revision 0Pae7 Page 77 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request ARP-03 Standby Liquid Control Pump Testing (Continued) Performance of a substantial flow test each quarter would result in eight sets of data over a two-year period instead of the required one comprehensive test. Monitoring of vibration on these pumps every six months will result in four sets of mechanical data versus the required one every two years. CNS believes this testing regime provides an overall better assessment of pump mechanical and hydraulic health and will determine operational readiness on a quarterly frequency. Additionally, this modified group A positive displacement pump test performed with vibrations will verify that the pump is operating acceptably and may be utilized as the post-maintenance test following significant maintenance. Multi-point preservice testing for positive displacement pumps is not required by the OM Code per ISTB-5323. These pumps are ASME non-code class pumps outside the scope of the IST Program. This method of monitoring these pumps provides a level of testing that is commensurate to the level of safety for these components.

6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was previously approved by CNS for the fourth ten-year interval as ARP-03.Revision 0 Page 78 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request ARP-04 Diesel Generator Fuel Oil Transfer Pump Testing Alternative Provides Acceptable Level of Quality and Safety 1. Augmented Code Component(s)

Affected DGDO-P-DOTA Diesel Generator Fuel Oil Transfer Pump A DGDO-P-DOTB Diesel Generator Fuel Oil Transfer Pump B 2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004 Edition through 2006 Addenda 3. Applicable Code Req~uirement ISTB-3 400, "Frequency of Inservice Tests" ISTB-51 10, "Preservice Testing" ISTB-5121, 'Group A Test Procedure" ISTB-5 123, "Comprehensive Test Procedure" 4. Reason for Request Augmented Relief is requested from the requirements of ISTB-3400, ISTB-51 10, ISTB-5 121, and and ISTB-5 123. The proposed alternative provides an acceptable level of quality and safety.This Augmented Relief Request does not require NRC approval.5. Proposed Alternative and Basis for Use Thc diesel fuel oil transfer pumps 'have an active safety function to transfer fuel oil to the respective diesel fuel oil day tank during normal diesel generator operation. The pumps will automatically start and stop under certain low and high levels in their respective day tanks.Each DGDO system pump must provide sufficient fuel flow to one diesel engine to meet consumption requirements. The design flow rate of the diesel fuel oil transfer pumps is 15 gpm.To support continuous Diesel Generator operation at full load, the transfer pump must be capable of delivering 4.64 gpm. Therefore, significant margin exists for these pumps.The DGDO pumps are conservatively categorized as group A pumps since they are operated routinely in support of diesel generator operation in addition to testing for IST purposes. The pumps are considered vertically mounted centrifugal pumps. Some plants consider these pumps skid-mounted in support of the diesel generators. CNS has decided to individually test these pumps in order to better monitor their performance. As an alternative to the code requirements for performing a Preservice test, Comprehensive test, and a Group A test, a modified Group A test will be performed. Revision 0Pae7 Page 79 Cooper Nuclear Station F/ifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request ARP-04 Diesel Generator Fuel Oil Transfer Pump Testing (Continued) A Comprehensive pump test will not be performed, which would be essentially the same test as the Group A test. During the modified Group A test, the pumps will be operated at a reference flow rate (Qr) of 6 gpm (+/-0.3 gpm) with pump differential pressure (P) measured and compared to its reference value. Deviations from the reference value will be compared to the required action and alert range requirements of Table ISTB-5121-1 for the Group A test. In addition, vibration measurements will be recorded once every 6 months (every other quarter). This 6 month frequency for vibrations is more than adequate based on the steady vibration trends observed over the past several years. The vibration measurements will be compared to the range requirements of Table ISTB-5 121-1 for the Group A test. Corrective actions will be taken in accordance with ISTB-6200. Permanently installed plant instrumentation will be used to determine differential pressures and flow rates. Portable vibration instruments will be used to determine vibration measurements. All instrumentation will meet the accuracy requirements of a Group A test unless specific relief is requested. This level of accuracy is sufficient for these augmented pumps, especially based on the large margin to the minimum flow of 4.64 gpm per pump. The current discharge pressure gauge is calibrated to < +1%, which is lower than the Group A accuracy requirement (+/-2%) and nearly mccts thc Comprchcnsivc pump tcst accuracy requirement (+1/2%), su nijujhnal variabilities between the Group A and Comprehensive Pump Test instrumentation currently exist.One of the requirements of the comprehensive test is to perform the test at substantial flow (4-20% of design flow). Since the accident design flow is such a low value, CNS will continue to test the pump at a value slightly higher than 20% above the design value, which allows for plenty of margin for inaccuracies in instrumentation. Design Point: 4.64 gpm Test Point: 6 gpm CNS will perform a modified Group A test as stated above such that the acceptance criteria for hydraulic performance will meet the code requirements for a Group A test. Additionally, CNS will perform vibration monitoring once every six months (every other quarter).The DGDO pumps are tested at a set flow of 6 gpm (+/-0.3 gpm) and differential pressure is measured. Per Table ISTB-5 121-1, the required action range for the Group A differential measurement would be <0.90 A Pr and > 1.10 A Pr. There is no alert range. This is the same as the comprehensive test requirement with the exception that the upper range is >1.03 A Pr and a lower alert range is in place for the measured differential pressure. With reference values in the low twenties for differential pressure, this upper limit for the comprehensive pump test may not encompass the normal data scatter associated with acceptable DGDO pump operation. Also, since the measured values are low and there is significant design margin built into the pump design, alert ranges for differential pressure are not necessary. Therefore, CNS will monitor the differential pressure of these pumps utilizing the Group A test criteria each quarter as follows: Acceptable Range 0.90 to 1.10 APr Required Action <0.9OA Pr or >l.10A Pr Revision 0 Page 80 Cooper Nuclear Station F~ifh Interval Inservice Testing Program for Pumps and Valves Relief Request ARP-04 Standby Liquid Control Pump Testing (Continued) CNS will evaluate all ranges against the design conditions to ensure that all procedure lower limits bound the more conservative of the design or ASME OM Code ranges delineated above.Performance of a substantial flow test each quarter would result in eight sets of data over a two-year period instead of the required one comprehensive test. CNS believes that there would be no benefits added to implementing a 2-year comprehensive test, which would essentially be identical to the quarterly test with vibrations. Therefore, CNS believes that the proposed testing regime establishes an acceptable assessment of pump mechanical and hydraulic health and will determine operational readiness on a quarterly frequency. Additionally, this modified group A centrifugal pump test performed with vibrations will verify that the pump is operating acceptably and may be utilized as the post-maintenance test (in place of a preservice test) following significant maintenance. This one point test is adequate to verify acceptable pump operation due to the simplification of the function the DGDO pumps perform. The fuel oil is being delivered from one large tank to a smaller tank and the minimum amount of oil is verified to be met. No other modes of function occur. As long as this function may be met, and pump performance is trended within the IST Program, this testing will adequately verify pump operability. Thcsc pumps arc ASME non-codc class pumps outsidc thc scope of the IST Program. This method of monitoring these pumps provides a level of testing that is commensurate to the level of safety for these components.

6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fourth ten-year interval.7. Precedents A similar relief request was previously approved by CNS for the fourth ten-year interval as ARP-04.Revision 0 Page 810 Page 81 Cooper Nuclear Station F~/iIi Interval Inservice Testing Pro gram for Pumps and Valves ATTACHMENT 4 VALVE RELIEF REQUESTS VALVE RELIEF REQUEST INDEX Relief Description 1NRC Approval Date Request No. _____________________

RV-0 1 HPCI Solenoid Operated Drain Valve Testing 21 2-2161 RV-02 ftMain Steam Safety Valve Testing per Code Case OMN- 17 ] 2412416'RV-03___ [Main Steam Safety Relief Valve Testing ] 2-12-16'RV-04 [Control Rod Drive (CRD) Technical Specification Testing 2-12-16'RV-05 ftPerformance-Based Scheduling of PIV Leakage Tests ] 2-12-16'(1) Approved by NRC letter, dated 2-12-16, from Meena K. Khanna, NRC, to Mr. Oscar A. Limpias, Vice President of Nuclear and CNO for CNS Revision 0Pae2 Page 82 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RY-Ol HiPCI Solenoid Operated Drain Valve Testing Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected Valve Class Category System HPCI-SOV-SSV-64 2 B HPCI HPCI-SOV-SSV-87 2 B 2. Applicable Code Edition and Addenda ASME CM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTC-3300 Reference Values -Reference values shall be determined from the results of preservice testing or from the results of inservice testing.ISTC-33 10 Effects of Valve Repair, Replacement, or Maintenance on Reference Values -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... ISTC-3 500 Valve Testing Requirements -Active and passive valves in the categories defined in ISTC-l1300 shall be tested in accordance with the paragraphs specified in Table ISTC-3500-1 and the applicable requirements of ISTC-5 100 and ISTC-5200. ISTC-35 10 Exercising Test Frequency -Active Category A, Category B, and Category C check valves shall be exercised nominally every 3 months except as provided by ISTC-3520, ISTC-3540, ISTC-3550, ISTC-3570, ISTC-5221, and ISTC-5222. ISTC-3560 Fail-Safe Valves -Valves with fail-safe actuators shall be tested by observing the operation of the actuator upon loss of valve actuating power in accordance with the exercising frequency of ISTC-35 10.ISTC-5 151 Valve Stroke Testing -(a) Active valves shall have their stroke times measured when exercised in accordance with ISTC-3500.(b) The limiting value(s) of full-stroke time of each valve shall be specified by the Owner.(c) Stroke time shall be measured to at least the nearest second.ISTC-5 152 Stroke Test Acceptance Criteria -Test results shall be compared to reference values established in accordance with ISTC-3300, ISTC-33 10, or IsTc-3320. ISTC-5 153 Stroke Test Corrective Action.Revision 0 Page 83 Revision 0 Page 83 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-01 IIPCI Solenoid Operated Drain Valve Testing (Continued)

4. Reason for Requnest Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(1), relief is requested from the listed requirements of the ASME OM Code. The proposed alternative would provide an acceptable level of quality and safety.The HPCI turbine and exhaust steam drip leg drain to gland condenser (HiPCI-SO V-S SV-64) and HPCI turbine and exhaust steam drip leg drain to equipment drain isolation valve (HPCI-SOV-SSV-87) have an active safety function in the closed position to maintain pressure boundary integrity of the IIPCI turbine exhaust line. These valves serve as a Class 2 to non-code boundary barrier.These valves are rapid acting, encapsulated, solenoid-operated valves. Their control circuitry is provided with a remote manual switch for valve actuation to the open position and an auto function which allows the valves to actuate from signals received from the associated level switches I{PCI-LS-98 and HPCI-LS-680.

Both valves receive a signal to change disc position during testing of drain pot level switches. However, remote position indication is not provided for positive verification of disc position. Additionally, their encapsulated design prohibits the ability to visually verify the physical position of the operator, stem, or internal components. Modification of the system to verify valve closure capability and stroke timing is not practicable nor cost beneficial since no commensurate increase in safety would be derived.5. Proposed Alternative and Basis for Use CNS has been performing a robust exercise test for these two valves that verifies obturator movement since 1998 on a quarterly basis. In 2001, this test identified some leakage past IJPCI-SOV-SSV64 and the valve was removed and refurbished. For the past -44 years, the exercise test has been completed without any issues. This test is accomplished through the performance of surveillance procedure, 6.H!PCI.204, HPCI-SOV-SSV64 and HPCI-SOV-SSV87 1ST Closure Test. With HPCI not in operation, a demineralized water source is utilized to verify that HPCI-SOV-SSV64 opens when level switch HPCI-LS-680 (turbine exhaust drain pot high level) trips, allowing level in the gland seal condenser to start to rise due to water flow through HPCL-SOV-SSV64. After HPCI-LS-680 resets and I-IPCI-SOV-SSV64 closes, the gland seal condenser level is verified to be steady.Similarly, CNS verifies that HPCI-SOV-SSV87 opens when level switch I-IPCI-LS-98 (turbine exhaust drip leg high) trips, allowing the observation of water flow to a floor drain from a drain pipe downstream of HPCI-SOV-SSV87. After HPCI-LS-98 resets and HIPCI-SOV-SSV87 closes, CNS observes the drain pipe downstream of HIPCI-SOV-SSV87 for gross leakage past the valve.Therefore, CNS verifies valve obturator movement for both valves open and closed while simultaneously verifying the calibration of two level switches.Typically, tests that involve hooking up pressure sources and various amounts of test tubing are not performed on a quarterly basis due to their complexities (i.e. local leak rate tests). In addition, each time this "quarterly" test has been performed, IJPCI unavailability time ('-.5 Revision 0 Page 84 Rev&ion 0 Page 84 Cooper Nuclear Station JFifh Interval Inser-vice Testing Program for Pumps and Valves Relief Request RY-01 Hi-PCI Solenoid Operated Drain Valve Testing (Continued) hours) is consumed in addition to some minor radiological dose. Finally, this exercise test is actually a much better method of determining the valve's operational readiness than a quarterly fast acting stroke time test would have been. Therefore, based on the complexities of the test, consuming unnecessary HIPCI unavailability time and personnel radiation exposure, the exceptional test history dating back to 2001, and the fact that this is a robust test that verifies obturator movement, CNS proposes to exercise each valve to the full closed position, as described, on a 6 month basis.In addition to performing this robust exercise test every 6 months, each solenoid valve will be disassembled and examined for degradation on a periodic basis per the Preventative Maintenance Program. The valve body, insert, piston, plunger/stem assembly, and stem spring will all be examined per criteria outlined in surveillance procedure 6.IHPCI.404. In addition, continuity and the physical condition of the coil will also be checked. The valve and/or valve parts will be refurbished and/or replaced, as necessary, based on this examination. This maintenance shall be performed at an optimized frequency, not to exceed 48 months (2 cycles). The purpose of this enhanced preventative maintenance is to ensure the long term reliability of the components and to monitor for internal degradation. This is consistent with NUREG 1482, Section 4.2.3. The 6 month exercise tests will ensure that the valves are operational and will fulfill thcir safcty function when called upon.On June 9, 2015, in response to the Nuclear Regulatory Commission's Request for Additional Information, CNS provided an explanation as to how the frequency of the preventative maintenance task of disassembly, inspection, and refurbishment is developed, maintained, and optimized. The following several paragraphs and Table 1 were submitted to the NRC.The frequency of the preventative maintenance (PM) task was developed after reviewing the*maintenance and test histories for these two solenoid valves and after reviewing the Electric Power Research Institute (EPRI) recommendations for PMs on solenoid valves. The maintenance history for these valves since 2005 is documented in Table 1 located at the end of this response.A revicw of this data demonstrates that each valve has had two examinations that resulted in minor issues resulting in the replacement of parts (March of 2011 and April of 2014 for HPCI-SOV-SSV64; April of 2012 and April of 2014 for HPCI-SOV-SSV87). For these cases, the PM was doing its job by identifying parts that had minor issues and replacing them prior to them becoming a major issue and impacting the safety function of the valve. The exercise testing performed prior to and after these examinations was completed with acceptable results. As long as the exercise testing of the valves continues to demonstrate acceptable performance and the examination PMs do not identify any major issues that could have impacted the closure safety function, then the maximum frequency of 48 months may be utilized for these HPCI PMs.The maximum frequency of 48 months (2 cycles) is conservative when comparing this frequency to the EPRI PM recommendations; The EPRI recommended task for elastomer replacement and internal inspection of a solenoid valve is 5 years for a severe environment and up to ten years for a mild environment. Cooper Nuclear Station (CNS) considers the location of these valves to be a severe environment, so the maximum frequency allowed by CNS would be one year less than what is recommended by EPRI.Revision 0 Page 85 Revision 0 Page 85 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RV-O1 HLPCI Solenoid Operated Drain Valve Testing (Continued) The frequency will be maintaincd through the CNS work management system and the PM process. A maintenance plan has been established with the necessary tasks required to satisfy the PM. A PM work order with these required tasks is automatically created well ahead of the scheduled due date and is scheduled based on the CNS work schedule process. Any frequency changes must be approved by the IST Engineer.The monitoring of these valves will be done by tracking the proposed six month exercise testing and the results of the internal examinations. As was described in the relief request, CNS has had excellent results with the exercise tests. Based on internal valve degradation, October of 2001 was the last time one of these valves (HPCI-SOV-SSV64) failed its closure acceptance criteria.For clarification purposes, however, there was a system issue in June of 2002 in which foreign material was causing HPCI-SOV-SSV64 to leak. An internal examination identified that there was foreign material found under the valve disc of HPCI-SOV-SSV64, but the valve itself, was examined and found to be in an acceptable condition. The CNS corrective action program addressed the issue and no other foreign material issues have impacted the closure function of these valves since then. Therefore, no internal valve degradation issue has impacted the closure function of these components since October of 2001 and no system issue has impacted the closure function of these components since June of 2002.The frequency of the PMs is optimized by balancing the component reliability with the correct PM frequency. The goal is to ensure that the solenoid valves continue to perform their closure function in a reliable manner without performing the internal examination PMs too frequently. As long as the PM ensures that any minor issue is taken care of prior to it becoming an issue with the closure function of the valve meeting its acceptance criteria, then the frequency is set at an acceptable duration. This, in ,conjunction with acceptable exercise tests, justifies the acceptability of the frequency. If the exercise testing results in a failure of the closure acceptance criteria of one of the solenoid valves, or the examination PM of one of the solenoid valves identifies a significant component issue that may have resulted in the respective valve not bcing able to perform its closure function, then the examination frequency of both solenoid valves shall be moved from 48 month frequencies to 24 month frequencies. From this point, two periodic examinations would have to be performed and completed satisfactorily at the 24 month frequency prior to returning the frequency to the 48 month frequency. In conclusion, the PM was developed based on a review of the maintenance and test history results, and review of EPRI recommendations. The existing frequency will be monitored as acceptable as long as the exercise testing is completed satisfactory and the internal examinations are either satisfactory or identify parts for replacement prior to when the parts issue would have caused a failure with the closure exercise testing. The frequency of internal examinations will be reduced from 48 months to 24 months for both valves if one valve were to fail its acceptance criteria for the closure exercise testing or if the findings of an internal examination of one of the valves results in the determination that it would not have met its closure function. Two successful examinations at the 24 month frequency would be required in order to return the PM(s)to a 48 month frequency. This is how the frequency of the preventive maintenance task of disassembly, inspection, and refurbishment was developed, and how it will be maintained, monitored, and optimized, if approved.Revision 0 Page 86 Cooper Nuclear Station Ffifh Interval Inser-vice Testing Program for Pumps and Valvesi Relief Request RV-01 HPCI Solenoid Operated Drain Valve Testing (Continued) Table 1 : Maintenance historics for HPCI-SOV-SSV64 and HPCI-SOV-SSV87 HPCI-SOV-SSV64 HPCI-SOV-SSV87 02-10-05: Visual exam satisfactory 02-10-05: Visual exam satisfactory (PM work order #4363336) (PM work order #4363336)N/A 06-21-05: Replaced valve at same time as non-essential valve, HPCI-SOV-SSV88, was replaced. Valves are in close proximity. New valve allows parts to be procured.(Corrective Maintenance [CM] work order#4211944)11-7-06: Visual exam satisfactory 11-14-06: Visual exam satisfactory (PM work order #4446767) (PM work order #4446767). 03-18-08: Visual exam satisfactory 03-18-08: Visual exam satisfactory (PM work order #4569097) (PM work order #4569097)08-19-09: Visual exam satisfactory 08-19-09: Visual exam satisfctory (PM work order #4626047) ____ (PM work order #462 6047)03-21-l- 1: Valve replaced for parts reasons 03-22-11: Visual exam satisfactory with a valve upgrade to match that of HPCI- (PM work order #4750715)SOV-SSV87 (CM work order #479 1033) ____________________ 04-24-12: Visual exam satisfactory 04-25-12: Seat plug on the bottom was found (PM work order #4803 767) curled around the edges and was replaced.This issue did not impact the valve's closure function as the previous closure testing was performed successfully (PM work order#4803 767)4-23-13: Visual exam satisfactory 04-23-13: Visual exam satisfactory (PM work order #4895831). (PM work order #4895831). 4-22-14: Plunger found slightly corroded and 4-22-14: Insert showed minor stem assembly was scored in the seating area. erosion/corrosion and plunger/stem has a small Both parts were replaced. Did not impact the groove around seating area. Both parts were valve's closure function as the previous closure replaced. Did not impact the valve's closure testing was performed successfully. function as the previous closure testing was (PM work order #493 8492) performed successfully. ________________________________(PM work order #493 8492)2-10-15: Visual exam satisfactory 2-10-15: Visual exam satisfactory (PM work order 5003464). (PM work order 5003464).Revision 0Pae8 Page 87 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RV-01 ILPCI Solenoid Operated Drain Valve Testing (Continued) The robust 6 month exercise testing and the enhanced preventative maintenance will provide an adequate indication of valve performance and will continue to provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(z)(1), NPPD requests relief from the specific ISTC requirements identified in this request.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents A version of this relief request was previously approved for the fourth ten-year interval at CNS as Relief Request RV-01, Revision 1 (TAG NO. ME7021, August 28, 2012) and Revision 0 (TAG Nos. MC8837, MC8975, MC8976, MC8977, MC8978, MC8979, MC8980, MC8981, MC8989, MC8990, MC8991, and MC8992, June 14, 2006).A version of this relief request was previously approved for the fifth ten-year interval at Dresden Nuclear Power Station as Relief Request RV-23H (TAG Nos. ME9865, ME9866, ME9869, ME9870, ME9871, and ME9872, October 31, 2013).Revision 0 Page 88 Revision 0 Page 88 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RV-02 Main Steam Safety Valve Testing per Code Case OMN-17 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected Valve Class Category System MS-RV-70ARV 1 C Main Steam (MS)MS-RV-70BRV 1 C MS MS-RV-70CRV 1 C MS 2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Codk Requirement ISTC-5240 -Safety and Relief Valves. Safety and relief valves shall meet the inservice test requirements of Mandatory Appendix I.ASME OM Code Mandatory Appendix I, "Inservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants," Section I-1320, "Test Frequencies, Class 1 Pressure Relief Valves," paragraph (a), "5-Year Test Interval," states that Class 1 pressure relief valves shall be tested at least once every 5 years.4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(l), relief is requested from the requirements of ASME GM Code Appendix I, 1-1320(a). The proposed alternative would provide an acceptable level of quality and safety.Section ISTC-3200, "Inservice Testing," states that inservice testing shall commence when the valves are required to be operable to fulfill their required function(s). Section ISTC-5240, "Safety and Relief Valves," directs that safety and relief valves meet the inservice testing requirements set forth in Appendix I of the ASME GM C ode. Appendix I, Section 1-1320(a), of the ASME GM Code states that Class 1 pressure relief valves shall be tested at least once every 5 years, starting with initial electric power generation. This section also states a minimum of 20 percent of the pressure relief valves are tested within any 24-month interval and that the test interval for any individual valve shall not exceed 5 years. Prior to Cycle 28, CNS had refueling cycles of 18 months. With three safety valves, CNS has been meeting the ASME GM Code by removing, Revision 0Pae8 Page 89 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-02 Main Steam Safety Valve Testing per Code Case OMN-17 (Continued) testing, rebuilding, and re-installing one valve per refiueling outage. All three of these safety valves have an acceptable test history since 1997 as will be described in section 5.However, after Refueling Outage (RE) 27 (Fall/20 12), CNS began the current 24-month refueling cycle. The five year frequency was met for the safety valve due in RE28 (Fall/20 14), but a relief request, requesting the use of Code Case OMN-l17, will be necessary in order to continue with the process of testing only one valve each refueling outage for the fifth ten-year interval, beginning March 1, 2016. Without this relief request, CNS would be required to remove and test all three valves within a two cycle frequency (two one outage and one the next) in order to ensure that all three valves are removed and tested in accordance with the ASME OM Code requirements. This testing pattern would ensure compliance with the ASME GM Code requirements for testing Class 1 pressure relief valves within a 5 year interval.Extending the test interval to 6 years, as described in Code Case OMN- 17, would allow CNS to continue with the current method of removing, testing, rebuilding, and re-installing one safety valve per outage so that all three safety valves would be replaced over three refuel cycles (i.e., years).Without Code relief, thc incremenital outage work due to the inclusion of an additional safety valve every other outage would be contrary to the principle of maintaining radiation dose As Low As Reasonably Achievable (ALARA). The removal and replacement of the additional safety valve every other outage results in an additional exposure of approximately 450 millirem (mrem) to 726 mrem. This estimate is based on the actual radiation received to remove and re-install a safety valve each of the last three refueling outages.In accordance with 10 CFR 50.55a(z)(1), NPPD requests approval of an alternative to the 5 year test interval requirement of the ASME GM Code, Appendix I, Section 1-1320(a) for the safety valves at CNS.5. Proposed Alternative and Basis for Use NPPD requests that the test interval be increased from 5 years to 6 years in accordance with Code Case OMN-17. All aspects of Code Case OMN-17 will be followed for the MS safety valves.As an alternative to the Code required 5-year test interval per Appendix I, paragraph I-1320(a), NPPD proposes that the subject Class 1 safety valves be tested at least once every three refueling cycles (approximately 6 years/72 months) with a minimum of 20% of the valves tested within any 24-month interval. This 20% would consist of valves that have not been tested during the current 72-month interval, if they exist. The test interval for any individual valve would not exceed 72 months except that a 6-month grace period is allowed to coincide with refueling outages to accommodate extended shutdown periods and certification of the valve prior to installation. This is all in accordance with OMN-1 7, paragraph (a).After as-found set-pressure testing, the valves shall be disassembled and inspected to verify that parts are free of defects resulting from time-related degradation or service induced wear. As-left Revision 0Pae9 Page 90 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RV-02 Main Steam Safety Valve Testing per Code Case OMN-17 (Continued) set-pressure testing shall be performed following maintenance and prior to returning thle valve to service. Each valve shall have been disassembled and inspected prior to the start of the 72-month interval. Disassembly and inspection performed prior to the implementation of Code Case OMN-17 may be used.Each refueling outage, CNS will remove one safety valve to be sent off-site to a test facility.Upon receipt at the off-site facility, the valves are subject to an as-found inspection, as-found seat leakage test, and as-found set pressure test in accordance with Appendix I of the ASME OM Code. Prior to the returning the valve to the plant for re-installation, the safety valve is disassembled and inspected to verify that internal surfaces and parts are free 'from defects or service induced wear. During this process, anomalies or damage are identified for resolution. Damaged or worn parts (i.e. springs, gaskets and seals) are replaced or repaired, as necessary. Following reassembly, the valve's set pressure is recertified. This existing process is in accordance with ASME OM Code Case OMN-1 7, paragraphs (d) and (e). Alternatively, CNS may elect to replace the removed valve with a spare valve that has previously already been through the process just described. Up to three spare valves may be used in accordance with paragraph (b) of OMN-17.NPPD has reviewed the as-found set point tcst results for all three safety valves tested since 1997 as detailed in Table 1. Since 1997, all as found lift tests have been within a +/-3% tolerance (maximum of +2.02%). The current Technical Specification requirements are that the as found test results fall within a +/-3% tolerance. Technical Specifications require the as left certification of the valves to meet a +/-1% tolerance. If an as found test is found to be outside of the +/-3 %tolerance, the other 2 safety valves will be removed and tested in accordance with Code Case OMN-17, paragraph (c).Accordingly, the proposed alternative of implementing all aspects of OMN-1 7, which will increase the test interval for the subject Class 1 safety valves from 5 years to 3 fuel cycles (approximately 6 years/72 months), will provide an acceptable level of quality and safety. This will also restore the operational and maintenance flexibility that was lost when the 24-month fuel cycle created the unintended consequences of more frequent testing. This proposed alternative will continue to provide assurance of the valves' operational readiness and provides an acceptable level of quality and safety pursuant to 10 CFR 50.55a(z)(1).

6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents A similar relief was previously approved at Peach Bottom for the fourth ten-year interval as Relief Request 01A-VRR-3 (TAC Nos. MF2509 and MF25 10, April 30, 2014).Revision 0Pae9 Page 91 Cooper Nuclear Station FJith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-02 Main Steam Safety Valve Testing per Code Case OMN-17 (Continued)

Monticello Nuclear Generating Plant Relief Request VR-04 was approved in a NRC Safety Evaluation Report dated September 26, 2012 (ML 12244A272). Quad Cites Nuclear Power Station, Units 1 and 2 Relief Request RV-05 was approved in a NRC Safety Evaluation Report dated February 14, 2013 (ML13042A348). Table 1: Cooper Nuclear Station Safety Valve Test History Safety Valve AF Test Date Set Pressure AsFudSt Dvainfo Pressure Set Pressure 4/9/1997 1240 1217 -1.85%10/9/1998 1240 1252 +0.97%MS-RV-70ARV 3/8/2003 1240 1226 -1.13%4/19/2008 1240 1232 -0.65%10/21/2012 1240 1255 +1.21%4/10/1997 1240126-.3 3/12/2000 1240 13 07 MS-RV-7OBRV 1/25/2005 1240121+.8 10/3/2009 1240 1260 +1.61%10/8/2014 1240 1253 +1.05%4/10/1997 1240 1262 + 1.77%11/12/2001 1240 1237 -0.24%MS-RV-70CRV 10/26/2006 1240 1265 +2.02%3/21/2011 1240 1262 +1.77%Revision 0Pae9 Page 92 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RV-03 Main Steam Safety Relief Valve Testing Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected Valve Class Category System MS-RV-71ARV 1 B/C MS MS-RV-71BRV 1 B/C MS MS-RV-71CRV 1 B/C MS MS-RV-7lDRV 1 B/C MS MS-RV-71ERV 1 B/C MS MS-RV-71FRV 1 B/C MS MS-RV-71GRV 1 B/C MS M S-RV-71HIRV 1 B/C M S 2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTC-5 240 -Safety and Relief Valves. Safety and relief valves shall meet the inservice test requirements of Mandatory Appendix I.ASME OM Code Mandatory Appendix I, "lnservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants," Section I-1320, "Test Frequencies, Class 1 Pressure Relief Valves," paragraph (a), "5-Year Test Interval," states that Class 1 pressure relief valves shall be tested at least once every 5 years.ASME OM Code Mandatory Appendix I, 1-3310 Class 1 Main Steam Pressure Relief Valves with Auxiliary Actuation Devices -Tests before maintenance or set-pressure adjustment, or both, shall be performed for 1-3310(a), (b) and (c) in sequence. The remaining shall be performed after maintenance or set-pressure adjustments:

a. visual examination; Revision 0Pae3 Page 93 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RV-t03 Main Steam Safety Relief Valve Testing (Continued)
b. seat tightness determination, if practicable;
c. set-pressure determination;
d. determination of electrical characteristics and pressure integrity of solenoid valve(s);e. determination of pressure integrity and stroke capability of air actuator;f. determination of operation and electrical characteristics of position indicators;
g. determination of operation and electrical characteristics of bellows arm switch;h. determination of actuating pressure of auxiliary actuating device sensing element, where applicable, and electrical continuity;
i. determination of compliance with the Owner's seat tightness criteria.4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(l), relief is requested from the requirements of ASME GM Code Appendix I, sections 1-1320(a) and 1-3310. The proposed alternative would provide an acceptable level of quality and safety.Section ISTC-5240, "Safety and Relief Valves," directs that safety and relief valves meet the inservice testing requirements set forth in Appendix I of the ASME OM Code.Appendix I, Section 1-1320(a), of the ASME GM Code states that Class 1 pressure relief valves shall be tested at least once every 5 years, starting with initial electric power generation.

This section also states a minimum of 20 percent of the pressure relief valves are tested within any 24-month interval and that the test interval for any individual valve shall not exceed 5 years.CNS has eight MS safety relief valves (SRV). The approach for the past several years has been to remove either 2 or 3 of the entire valves (i.e. main body and pilot assembly) every refueling outage and send them off for as found testing, refurbishment, rebuilding, and re-certification in preparation for the next time they are re-installed into the plant. Those 2 or 3 entire valves have been replaced with refurbished valves that were recertified just prior to the outage. The schedule is planned so that all eight entire valves get sent off, as found tested, refurbished, and re-certified within a three cycle frequency. In addition, CNS has replaced the remainder of the pilot assemblies (5 or 6 per outage) and sent them off for testing, refurbishment, and re-certification in preparation for the next time they are re-installed into the plant. These 5 or 6 additional pilot assemblies are replaced with refurbished and recertified pilot assemblies that were recertified just prior to the outage. Therefore, the pilot assemblies for the full complement of 8 valves have been set pressure tested every outage for several years.Revision 0Pae9 Page 94 Cooper Nuclear Station Ftifh Interval Inservice Testing Program for Pumps and Valves Relief Request RV-03 Main Steam Safety Relief Valve Testing (Continued) CNS plans to continue this approach into the fifth ten-year interval. However, refucling outage 27 (Fall/20 12) was the last refueling outage under an 18-month cycle. CNS is now operating with 24-month cycles. With this in mind, the refurbishment of the entire valves will eventually align with a six year frequency, which is consistent with Code Case OMN-1 7. However, all eight of the pilot assemblies are being removed, tested and replaced with refurbished/recertified spare pilot assemblies every refueling outage, which means a full complement of the set pressure portion of the valves are being tested every refueling outage. Therefore, although this approach is very conservative, documenting acceptability of this approach is being pursued per this relief request.Additionally, since 5-6 pilot assemblies, alone, are being replaced every outage (versus the entire valve), documenting acceptability of how portions of Appendix 1-3310 are being satisfied is also being pursued per this relief request.5. Proposed Alternative and Basis for Use These eight SRVs are considered Class 1 main steam pressure relief valves with auxiliary actuating devices. They are located on the main steam lines. In addition to their automatic function of opening to prevent over pressurization of the reactor vessel, six of these valves are associated with the Automatic Depressurization System and two are associated with the Low Low Set logic. The valves are two-stage Target Rock valves, each equipped with a main body, a pilot assembly for set pressure control, a solenoid valve, and an air operator assembly.CNS proposes to follow the Code Case OMN-1 7, paragraph (d), recommendations for Maintenance on these eight valves. Therefore, on a three cycle (up to 6 year) frequency, CNS proposes to remove the entire valve unit (i.e. main body and pilot assembly) for each one of these valves and ship it off for as found testing, refurbishment, and re-certification. CNS will replace these entire valve units with spare refurbished and re-certified entire valve units.As mentioned earlier, each valve is equipped with a pilot valve assembly that controls the set pressure. The remainder of the pilot valve assemblies (5 or 6 per refueling outage) will be removed from the main body and sent off site for examination, as found testing, refurbishment, and re-qualification testing (set point, reseat, and pilot stage seat tightness). The test facility has a main body slave for this purpose. The removed pilot valve assemblies are replaced with previously refurbished and re-qualified pilot valve assemblies. By testing all of the pilot valve assemblies every outage, the potential need to expand to test additional valves due to set pressure failures is alleviated and the future valve reliability is improved. Test results are being monitored by serial numbers. Any as found set pressure failure will be addressed via the CNS Corrective Action Program.ASME OM Code Interpretation, 98-8, clarifies that a pilot operated relief valve with an auxiliary actuating device is not required to be tested as a unit. Furthermore, it clarifies that set pressure determination on the pilot operator may be perforned after the pilot operator is removed from the valve body.Revision 0Pae9 Page 95 Cooper Nuclear Station Fith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-03 Main Steam Safety Relief Valve Testing (Continued) Appendix I, 1-3310(a) visual examination is completed at the test facility for those main bodies and pilot assemblies being sent there for examination, testing and refurbishment. With the removal of the pilot assemblies from the main bodies at the plant, the accessible portions of the main bodies will be examined in place without further disassembly as permitted by I-1310(c). Appendix I, 1-3310(b) seat tightness, and 1-3310(c) set pressure, is satisfied through as found seat leakage and set pressure testing at the offsite test facility for those main valves and pilot valve assemblies being sent there for inspection, testing and refurbishment. Paragraph 1-3310(i) is satisfied through as left seat leakage testing at the facility. Seat leakage of installed main valves is continuously monitored and also satisfies 1-3310(i). Pressure switches in the SRV discharge lines annunciate in the control room and indicate when the main valve seat is open. In addition, there are temperature elements on the valve discharge lines which provide leakage indication. During startup, the main valve and Auxiliary Actuation Devices are verified to function properly by being full stroke exercised open and closed. Successfully exercising these valves open and closed verifies the electrical characteristics and pressure integrity of the solenoid valve and air actuator (satisfying Appendix I, paragraphs (d) and (e)). During this exercise, Appendix I, paragraph 1-3310(f), is also satisfied through the use of the valve indicating lights, discharge pressure switches, and temperature elements.Finally, Appendix I, paragraphs 1-3310(g) and 1-3310(h), are not applicable to the CNS MS safety relief valves.This proposed alternative is conservative in nature and will continue to provide an acceptable level of quality and safety pursuant to 10 CFR 50.55a(z)(1).

6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents A version of this relief request was previously approved for the fourth ten-year interval at CNS as Relief Request RV-04 (TAC Nos. MC8 837, MC8975, MC8976, MC8977, MC8978, MC8979, MC8980, MC8981, MC8989, MC8990, MC8991, and MC8992, June 14, 2006).Revision 0 Page 96 Revision 0 Page96 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves Relief Request RV-04 Control Rod Drive (CRD) Technical Specification Testing Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected Valve Class Category System CRD-SOV-SO120* 2 B CRD CRD-SOV-SO121* 2 B CRD CRD-SOV-SO122* 2 B CRD CRD-SOV-SO123* 2 B CRD CRD-AOV-CV126* 2 B CRD CRD-AOV-CV127* 2 B CRD CRD-CV-114CV* 2 C CRD CRD-CV-138CV* 2 C CRD SOV=Solenoid Operated Valve AOV=Air Operated Valve CV==Check Valve*Typical of 137 Hydraulic Control Units (HCU)2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ASME OM Code ISTC-3500 Valve Testing Requirements -Active and passive valves in the categories defined in ISTC-1 300 shall be tested in accordance with the paragraphs specified in Table ISTC-3500-1 and the applicable requirements of ISTC-5 100 and ISTC-5200. ISTC-35 10 Exercising Test Frequency -Active Category A, Category B, and Category C check valves shall be exercised nominally every three (3) months, except as provided by ISTC-3520, ISTC-3540, ISTC-3550, ISTC-3570, ISTC-5221, and ISTC-5222. ISTC-3560 Fail-Safe Valves -Valves with fail-safe actuators shall be tested by observing the operation of the actuator upon loss of valve actuating power in accordance with the exercising frequency of ISTC-35 10.ISTC-5 131 (a) Valve Stroke Testing -Active valves shall have their stroke times measured when exercised in accordance with ISTC-3500. ISTC-5 151 (a) Valve Stroke Testing -Active valves shall have their stroke times measured when exercised in accordance with ISTC-3500. Revision 0 Page 97 Revision 0 Page 97 Cooper Nuclear Station Fdith Interval Inservice Testing Program for Pumps and Valves Relief Request RV-04 Control Rod Drive (CRD) Technical Specification Testing (continued) ISTC-5221 (a) Valve Obturator Movement -The necessary valve obturator movement during exercise testing shall be demonstrated by performing both an open and a close test.4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(l), relief is requested from the requirements of ASME GM Code ISTC-3500, ISTC-3510, ISTC-3560, ISTC-5131 (a), ISTC-5151 (a), and ISTC-5221 (a). The proposed alternative would provide an acceptable level of quality and safety.This relief is needed to make the fifth ten-year inservice test program consistent with NTJREG 1482, Revision 2.5. Proposed Alternative and Basis for Use Background Information It is typical for Boiling Water Reactors (BWR) to perform the subject CRD) testing per their respective plant Technical Specifications. This originated from Generic Letter (GL) 89-04, Position 7. Per section 1.3 of NUIREG 1482, Revision 2, specific relief is required to implement the guidance derived from GL 89-04, which is why this testing is being documented under a relief request. The proposed alternatives and the basis for use are discussed in further detail below.CRD-CV-13 8CV: CRD-SOV-SO1 20, SO121. SO 122. SO 123: The CRD cooling water header check valve, CRD-CV-138CV (typical of 137 HCUs), has a safety function to close in the event of a scram to prevent diversion of pressurized HCU accumulator water to the cooling water header. The exhaust water withdrawal/settle (CRD-SOV-S0120), exhaust water insert (CRD-SOV-SO121), drive water withdrawal (CRD-SOV-SO122), and drive water insert (CRD-SO V-SO 123) solenoid valves (typical of 137), have a safety function to close in order to provide a boundary to non-code class piping.Normal control rod motion will verify that the associated cooling water check valve has moved to its safety function position of closed. Industry experience has shown that rod motion may not occur if this check valve were to fail in the open position.The solenoid valves listed above have a safety function to close in order to provide a class 2 to non-code class boundary isolation. During normal operation, these solenoid valves are used for control rod insertion and withdrawal. They are exercised open and closed during normal operation of the associated CRD. They are not equipped with position indication or control switches. They automatically change position to affect control rod movement.Therefore, control rod exercising in accordance with the CNS Technical Specifications, Surveillance Requirement (SR) 3.1.3.3, will provide an acceptable level of quality and safety for these valves. This testing method is consistent with GL 89-04, Position 7, and NUREG 1482, Revision 2, Section 4.4.6.Revision 0 Page 98 Cooper Nuclear Station F~ith Interval Inservice Testing Program for Pumps and Valves Relief Request RV-04 Control Rod Drive (CRD) Technical Specification Testing (continued) CRD-AOV-CV126, CRD-AOV-CV127. and CRD-CV-1 14CV: These valves operate as an integral part of their respective HCU to rapidly insert the control rods in support of a scram. The CRD scram inlet valve, CRD-AOV-CV 126 (typical of 137), opens with a scram signal to pressurize the lower side of the Control Rod Drive Mechanism (CRDM)pistons from the accumulator or from the charging water header. The CRD outlet isolation valve, CRD-AOV-CV 127 (typical of 137), opens with scram signal to vent the top of the CRDM piston to the scram discharge header. The CRD scram outlet check valve, CRD-CV-1 14CV (typical of 137), opens to allow flow from the top of the CRDM piston to the scram discharge header.Individual stroke time measurements of air-operated valves CRD-AOV-CV1 26 and CRD-AOV-CV127 are impractical due to their rapid acting operation and they are not equipped with position indication. '[herefore, valve stroke times will not be measured. Additionally, the air-operated valves fail-open on a loss of air or power. Normal opening removes power to the pilot solenoid valve, simulating a loss of power. On loss of power, the solenoid vents the air operator and CRD-AOV-CV126 and CRD-AOV-CV 127 are spring-driven open. Thus, each time a scram signal is given, the valves "experience" a loss of air/power to verity' each valve's fail-safc open feature.Testing these valves simultaneously would result in a full reactor scram. An excess number of scrams performed routinely could cause thermal and reactivity transients, which could lead to fuel, vessel, CRD, or piping damage. The CRDs cannot be tested during cold shutdown because the control rods are inserted and must remain inserted.Therefore, control rod scram time testing in accordance with the CNS Technical Specifications, SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4, will provide an acceptable level of quality and safety for these valves. This testing method for these valves is consistent with GL 89-04, Position 7, and NUJREG 1482, Revision 2, Section 4.4.6.'6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was previously approved for the fourth ten-year interval at CNS as relief request RV-06 (TAC No. ME1 521, April 26, 2010). A similar alternative was approved at Perry-1 for relief request VR-1, revision 1 (TAC No. ME7380, February 22, 2012).Revision 0 Page 99 Revision 0 Page 99 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-O5 Performance-Based Scheduling of Pressure Isolation Valve Leakage Tests Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternative Provides Acceptable Level of Quality and Safety 1' ASME Code Component(s) Affected Valve Class Category System RHR-MOV-MO25A 1 A RHR RHR-MOV-MO25B 1 A RHR RHR-MOV-MO274A 1 A RHR RHIR-MOV-MO274B 1 A RHR RHR-CV-26CV 1 A/C RHIR RHR-CV-27CV 1 A/C RHR RHR-MOV-MO017 1 A RHR RHR-MOV-MO018 1 A RHR CS-MOV-MO012A 1 A CS CS-MO V-MO 12B 1 A CS CS-C V-18CV 1 A/C CS CS-C V-19CV 1 A/C CS MOV=Motor Operated Valve 2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTC-3630 -Leakage Rate for Other Than Containment Isolation Valves.ISTC-3630(a) -Frequency. Tests shall be conducted at least once every two years.4. Reason for Request Pursuant to 10 CER 50.55a, "Codes and standards," paragraph (z)(1), relief is requested from the requirement of ASME OM Code ISTC-3630(a). ISTC-3630(a) requires that leakage rate testing (water) for pressure isolation valves (PIV) be performed at least once every two years. Data from RE25 and RE26 was used to identify that PIV testing alone each refueling outage incurs a total dose of at least 600 mRem. The reason for this relief request is to reduce outage dose. The basis of this relief request is that the proposed alternative would provide an acceptable level of quality and safety.5. Proposed Alternative and Basis for Use The RHR and CS systems at CNS contain valves that function as PIVs. PI~s are defined as two normally closed valves in series at the reactor coolant system boundary that isolate the reactor Revision 0 Pg 0 Page 100 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RV-05 Performance-Based Scheduling of Pressure Isolation Valve Leakage Tests (continued) coolant system from an attached low pressure system. These affected valves, listed in Section 1, are located on the 'A' and 'B' CS and RHR injection lines and the RHR shutdown cooling line.PI~s are not specifically included in the scope for performance-based testing as provided for in 10 CFR 50 Appendix J, Option B. The concept behind the Option B alternative for containment isolation valves is that licensees should be allowed to adopt cost effective methods for complying with regulatory requirements. Additionally, NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," describes the risk-informed basis for the extended test intervals under Option B. That justification shows that for valves which have demonstrated good performance by passing their leak rate tests (air) for two consecutive cycles, further failures appear to be governed by the random failure rate of the component. NEI 94-01 also presents the results of a comprehensive risk analysis, including the statement that "the risk impact associated with increasing [leakrate] test intervals is negligible (less than 0.1 percent of total risk)." The valves identified in this relief request are in water applications. The PIV testing is performed with water pressurized to normal plant operating pressures. This relief request is intended to provide for a performance-based scheduling of PIV tests at CNS.As stated in the previous section, the reason for requesting this relief is dose reduction. Data reviewed from RE25 and RE26 identified that PIV testing alone incurred a total dose of approximately 600 mrem in RE26, which benefited from the chemical decontamination that was performed, and approximately 1600 mrem in RE25. Therefore, assuming the PIVs remain classified as good performers, extended test intervals of three refueling outages would provide a savings of at least 1200 mrem over a three-cycle period.NUREG 0933, "Resolution of Generic Safety Issues," Issue 105, discusses the need for PIV leak rate testing based primarily on three pre-1980 historical failures of applicable valves industry-wide. These failures involved human errors in either operations or maintenance. None of these failures involved inservice equipment degradation. The performance of PlY leak rate testing provides assurance of acceptable seat leakage with the valve in a closed condition. Typical PIV testing does not identify functional problems which may inhibit the valves ability to re-position from open to closed. For check valves, such functional testing is accomplished per ASME OM Code ISTC-3522 and ISTC-3520. Power-operated valves are routinely full stroke tested per ASME OM Code to ensure their functional capabilities. The periodic functional testing of the PI~s is adequate to identify abnormal conditions that might affect closure capability. Performance of the separate 24-month PIV leak rate testing does not contribute any additional assurance of functional capability; it only determines the seat tightness of the closed valves.Revision 0 Pg 0 Page 101 Cooper Nuclear Station F~fhz Inteival Inserviee Testing Program for Pumps and Valves Relief Request RV-05 Performance-Based Scheduling of Pressure Isolation Valve Leakage Tests (continued) The functional test and position indication test (PIT) frequencies are as follows: Valve Functional Test PIT RITR-MOV-MO25A Quarterly 2 years RHR-MOV-MO25B Quarterly 2 years RITR-MOV-M0274A Normally De-energized Closed Refueling Outage (exercised during PIT test)RIIR-MOV-MO274B Normally De-energized Closed Refueling Outage (exercised during PIT test)RIIR-CV-26CV Refueling Outage Refueling Outage RHR-CV-27CV Refueling Outage Refueling Outage RETR-MOV-MO017 Cold S/D Refueling Outage RHR-MO V-MO 18 Cold SiD Refueling Outage CS-MO V-MO 12A Cold S/D Refueling Outage CS-MOV-MO12B Cold S/D Refueling Outage CS-C V-i18CV Refueling Outage Refueling Outage CS-C V-19CV Refueling Outage Refueling Outage CNS proposes to perform PlV testing at intervals ranging from every refueling outage to every third refueling outage. The specific interval for each valve would be a function of its performance and would be established in a manner consistent with the containment isolation valve (CIV) process under 10 CFR 50 Appendix J, Option B. Five of the 12 valves listed in Section 1 (RHR-MOV-MO25A, RHIR-MOV-M025B, CS-MOV-MO12A, CS-MOV-MO12B, RHR-MO V-MO 17) are also classified as CIVs and are leak rate tested with air at intervals determined by 10 CFR 50 Appendix J, Option B. Appendix J and inservice leak testing program guidance will be established such that if any of those five valves fail either their as found CIV test or their PIV test, the test interval for both tests will be reduced to every refueling outage until they can be re-classified as good performers per Appendix J, Option B requirements. The test intervals for the seven remaining valves with a PIV-only function will be determined in the same manner as is done under Option B. That is, the test interval may be extended to every three refueling outages (not to exceed a nominal six year period) upon completion of two consecutive, periodic PIV tests with results within prescribed acceptance criteria. Any test failure will require a return to the initial interval (every refueling outage) until good performance can again be established. The primary basis for this relief request is the historically good performance of the PI~s. There have been no PIV seat leakage failures since PlV testing began at CNS in 1995 through the present. Leakages recorded have been a very small percentage of the overall allowed leakage.Revision 0 Pg 0 Page 102 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RV-05 Performance-Based Scheduling of Pressure Isolation Valve Leakage Tests (continued) The test results for the PI~s listed in Section 1 have been exceptional. For example, a plot of the RHR-MO V-MO 17 test results is shown below: This graph is typical of the affected PI~s listed in Section 1; however, there have been cases where the CIV air testing has indicated a failure with components identified in this relief request.There is a general industry-wide consensus that CIV air testing is a more challenging and accurate measurement of seat condition, and more likely to identify any seat condition degradation. PIV testing has also been utilized at CNS as a post-maintenance test following packing replacements on the CS and RHR injection check valves to ensure the packing is adjusted adequately at normal system pressure. Therefore, PIV testing will continue to be utilized as post-maintenance testing, as necessary. On June 8, 2012, the NRC staff reviewed and endorsed NEI 94-01, Revision 3 (see the safety evaluation at Accession No. ML12 1030286), which allows for up to a 75-month frequency for"Type C tests." Per the NRC safety evaluation report (SER) for the fourth interval IST Program (TAC No. ME7021, dated August 28, 2012), to obtain a frequency extension beyond 60 months (up to 75 months), licensees should provide additional information, such as maintenance history, acceptance tests criteria, condition monitoring programs, etc., to justify the acceptability of the extension. In order to further justify the proposed maximum frequency of 3 cycles (72 months)with a standard grace period of 6 months, additional information is being provided.Table 1 of this relief request contains the maintenance history and Local Leak Rate Test (LLRT)/PIV test history for all 12 of these pressure isolation valves for the past 10 years (since 01/01/2005). The table includes the as found and as left LLRT and PIV test results with the Revision 0 Page 103 Revision 0 Page 103 Cooper Nuclear Station Ffifh Interval lnservice Testing Pro gram for Pumps and Valves Relief Request RV-05 Performance-Based Scheduling of Pressure Isolation Valve Leakage Tests (continued) associated operability limits. Note that corrective and preventative maintenance has been performed over the past 10 years (and beyond) in order to maintain the acceptable performance of the components. For instance, the MOV Program requires regular inspections and diagnostic tests of the motor operators to ensure that they continue to be relied upon throughout the life of the plant and the check valves have preventative maintenance plans to replace the valve packing on a periodic basis to ensure the packing material is properly maintained. Note that not all of the maintenance performed impacts the seating ability of the components or the test boundary of the associated LLRT/PIV tests, so pre- or post-LLRT/P1V testing may not have been required to be performed. Exercise testing, stroke time testing, and position indication testing was not listed in Table 1.As can be observed from Table 1, the As Found LLRT test results have been excellent with no failures associated with these valves over the past 10 years and a significant amount of margin has been maintained to the administrative component operability limit. Even more so, a very large margin exists between the PIV test results and the operability limit for each PTV test. With a limit of 5 gpm, the highest recorded PIV leakage in the last 10 years was 0.43 5 gpm, which is only 8.7% of thc allowcd leakage. Historically, since 1995, all of the PIV valves have maintained this much or more of a margin to the 5 gpm acceptance criteria as shown below.Test # Components Maximum PIV Percent of Percent of leakage recorded allowed leakage margin to 5 gpm since 1995 (gpm) limit 1 RHiR-MOV-MO25A 0.299 5.98% 94.02%2 RHIR-MOV-MO25B 0.272 5.44% 94.56%3 RHR-CV-26CV /RH-O-07A0.1224 2.45% 97.55%4 RHR-CV-27CV /RH-O-07B0.326 6.52% 93.48%5 RHiR-MOV-MO017 0.0272 0.54% 99.46%6 RI-R-MOV-M0 18 0.218 4.36% 95.64%7 CS-MO V-MO 12A 0.435 8.70% 91.30%8 C S-MOV-MO 12B 0.082 1.64% 98.36%9 CS-CV-18CV 0.3264 6.53% 93.47%10 CS-CV-19CV 0.082 1.64% 98.36%The NRC SER for NEI TR 94-01, Revision 3, resulted in a condition that the licensee report the margin between the Type B and Type C leakage rate summation and its regulatory limit and maintain an acceptable margin to the regulatory limit. A second condition requires the licensee to include considerable extra margin in order to extend the LLRT intervals beyond 5 years to a 75-month interval. In comparison, for these PIV tests, CNS will establish an administrative limit of<1 gpm for each of the PlY tests in order to maintain each test on an extended frequency. This administrative limit is only 20% of the allowed leakage and will provide considerable extra margin to the limit of 5 gpm when looking at the historical test results.Revision 0 Pg 0 Page 104 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-05 Performance-Based Scheduling of Pressure Isolation Valve Leakage Tests (continued) NUREG/CR-5928, "ISLOCA Research Program Final Report," evaluated the likelihood and potential severity of inter-system loss-of-coolant accident (ISLO CA) events in BWR and pressurized water reactors. The BWR design used as a reference for this analysis was a BWR/4 with a Mark 1 containment. CNS was listed in Section 4.1 of NUREG/CR-5928 as one of the applicable plants. The applicable BWR systems were individually analyzed and in each case, this report concluded that the system was "...judged to not be a concern with respect to ISLOCA risk." Section 4.3 concluded the BWR portion of the analysis by saying "ISLOCA is not a risk concern for the BWR plant examined here." Summary of bases / rationale for this relief request:* Performance-based PIV testing would yield a dose reduction of up to 1200 mrem over a three-cycle period.* Performance of separate functional testing of PIVs per ASME Code.a Excellent historical performance results from PlV testing for the applicable valves.* Low likelihood of valve mispositioning during power operations (procedures, interlocks).

  • Air testing versus water testing -degrading seat conditions are identified much sooner with air testing.* Relief valves in the low pressure piping -these relief valves may not provide ISLOCA mitigation for inadvertent PIV mispositioning (gross leakage), but their relief capacity can easily accommodate conservative PIV seat leakage rates.* Alarms that identify high pressure to low pressure leakage -Operators are highly trained to recognize symptoms of a present or incipient ISLOCA and to take appropriate actions.The intent of this relief request is simply to allow for a performance-based approach to the scheduling of PTV leakage testing. It has been shown that ISLOCA represents a small risk impact to BWRs such as CNS. CNS PIVs have an excellent performance history in terms of seat leakage testing. The risks associated with extending the leakage test interval to a maximum of three refueling outages (nominal 24 months) are extremely low. The performance-based interval shall not exceed 72 months. Standard scheduling practice may extend the program interval by 25%, not to exceed six months. This relief will provide significant reductions in radiation dose.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents A version of this relief request was previously approved for the fourth ten-year interval at CNS as relief request RV-07 (TAC No ME7021, dated 8-28-20 12). Fermi 2 received a Safety Evaluation by the NRC, dated September 28, 2010, on a similar relief request for the performance-based testing of PI~s (TAC No. ME2558, ME2557, and ME2556).Revision 0 Page 105 Revision 0 page lO5 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005

_____Component(s) [ Date hOutage jWork Order Work Order Description f AF Tests [ AL Tests Comments RIHR-MOV-MO25A (Test #1)Sprnn/2005 RE22 N/A N/A LLRT: 2.02 scfhi (< 30 scfh)PIV: 0.299 gpm (< 5 nom)N/A LLRT and PIV test due.10/02/2005 Online CM 4464719 Remove insulation and N/A N/A No impact on validate leak; tightened LLRT or PIV cap screws on pressure testing.seal; slowed leakage. _______02/10/2006 Online CM 4465302 Efforts were made to stop N/A N/A No impact on bonnet seal leak. LLRT or PIV testing.05/30/2006 Online PM 4390882 Clean & Lubricate Stem N/A N/A No impact on LLRT or PIV testing.Fall/2006 RE23 CM 4464912 Repaired pressure seal LLRT: LLRT (Final AL): Major CM 4534360 leak, refurbed motor 0.83 scfha (< 30 scflh) 7.95 scfh ( < 30 maintenance operator, disassembled scfh)) resets LLRT and examined valve, and PIV: Freq. to every diagnostically tested. 0.08 gpm (AL) refueling outage.04/03/2007 Online PM 4542913 Perform Motor Pinion N/A N/A No impact on Inspection LLRT or PIV testing.10/04/2007 Online PM 4498618 Examine MO-Mech N/A N/A No impact on PM 4498668 Examine MO-Elect LLRT or PIV testing.Spring/2008 RiE24 N/A N/A LLRT: N/A 1 st periodic 1.25 scfh (< 50 scflh) test for LLRT PIV: (and PIV)0.109 gpm (< 5 test.gpm)_________________ 03/30/2009 Online PM 4625205 PM 4625262 PM 4625267 Clean & Lubricate Stem Examine MO-Mech Examine MO-Elec N/A N/A No impact on LLRT or PIV testing.Revision 0 Pg 0 Page 106 Cooper Nuclear Station Fifth Interval lnservice Testing Program for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 Component(s) '-Date Outage Work Order Work Order Description AF Tests AL Tests C~mments FalL'2009 RE25 CM 4641890 MOV Program Diagnostic LLRT: PIV: No impact to test/motor replacement. 1.75 scfha (< 50 scfha) 0.109 gpm (< 5 gpm) LLRT or PIV No AL LLRT required testing. 2nd due to minimal seat thrust periodic test change, for LLRT (and PIV)test.Spring/2011I RE26 N/A N/A LLRT: N/A 3rd (extra)2.1 scfha (< 50 scfha) periodic test PIV: for LLRT 0.136 gpm (< 5 (and PIV)gpm) test.06/05/2012 Online PM 4802964 Clean and Lubricate Stem N/A N/A *No impact on PM 4803040 Examine MO-Mech LLRT or PIV PM 4803052 Examine MO-Elec testing.Fall/2012 RE27 N/A N/A N/A N/A No tests due to Option B /approved PIV relief request.Fall/2014 RE28 N/A N/A LLRT: N/A No PIV test 3.82 scfha (< 50 scfha) due to approved PIV_____________relief request.RHR-MOV-M025B Spring/2005 RE22 CM 4335229 Votes diagnostic test LLRT: LLRT: MOV (Test #2) 24 scfli (< 30 scfth) 23.8 scflh (< 30 sofia) periodic test.PIV: 0.0544 gpm (< 5 gpm)04/11/2005 Online PM 4381354 Clean and Lubricate Stem N/A N/A No impact on LLRT or PIV testing.10/17/2006 Online CM 4531030 Examine Torque Switch N/A N/A No impact on LLRT or PIV testing.10/18/2006 Online CM 4531090 Replace Motor Pinion N/A N/A No impact on Gear LLRT or PIV testing.Revision 0 Pg 0 Page 107 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 _____Component(s) [ Date [Outage ]Work Order ]Work Order Description AF Tests AL Tests [Comments Fall/2 006 RE23 CM 4537229: Packing Leak; No AL LLRT required due to minimal packing/seat load change.LLRT: 17.5 scfl ( 50 scf)PIV: 0 gpm (< 5 gpm)No impact on LLRT or PIV testing.Monitoring LLRT;assume 1st periodic test for PIV.04/18/2007 Online PM 4485530 Examine MO-Elec N/A N/A No impact on PM 4498698 Examine MO LLRT or PIV testing.Spring/2008 RE24 CM 4632406 Repack valve LLRT: N/A No impact on CM 4631163 Adjust Packing/Viper Test 32.14 scfih (< 50 LLRT or PIV CM 4531210 Refurbed AO; scfh) testing.No AL LLRT/PIV PIV: Monitoring required due to minimal 0.0544 gpm (< 5 LLRT; 2nd packing/seat load change. gpm) periodic test__________for PIV.10/14/2008 N/A PM 4600595 Clean and Lubricate Stem N/A N/A No impact on LLRT or PIV testing.Fall/2009 RE25 N/A N/A LLRT: N/A Monitoring 12.74 sceti (< 50 LLRT; 3rd scfh) periodic test PIV: for PIV.0.054 gpm (< 5gpm)07/13/2010 Online PM 4664227 Examine MO-Elec N/A N/A No impact on PM 4664250 Examine MO-Mech LLRT or PIV____________testing. Spring/201 1 RE26 N/A N/A LLRT: N/A Monitoring 23.16 scifh (< 50 LLRT (1st scflh) periodic test);PIV: 4th periodic 0.136 gpm (< 5 test for PIV.gpm) _ _ _07/19/2011 Online PM 4749837. Clean and Lubricate Stem N/A N/A No impact on LLRT or PIV testing.Revision 0 Pg 0 Page 108 Cooper Nuclear Station Ffifh Interv'al Inservice Testing Pro gram for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 Component(s) '-Date Outage Work Order Work Order Description AF Tests AL Tests Comments Fall/2012 RE27 CM 4842207 Viper diagnostic test; No LLRT: N/A No impact on AL LLRT/PIV required 0.17 scfha (< 50 sctfh) LLRT or PIV due to minimal seat load testing.change. Mc~nitorinlg LLRT (2nd periodic test);No PIV test due to approved PIV relief request.01/14/20 13 Online PM 4864090 Examine MO-Mech N/A N/A No impact on PM 4864089 Examine MO-Elec LLRT or PIV__________ ________________ ________________ testing.01/16/2014 Online CM 4996016 Reterminate Motor Wiring N/A N/A No impact on LLRT or PIV testing.Fall/2014 RE28 N/A N/A N/A N/A No tests due to Option B /approved PIV relief request.01/12/2015 Online PM 4953672 Clean and Lubricate stem N/A N/A No impact on LLRT or PIV_______________________testing. RHIR-MOV-Spring/2005 RE22 RiHR-MO- Examine Motor Operator LLRT: N/A No impact on MO274A MO274A 7.5 sctlh (< 35 scfh) LLRT or PIV& PM 4363586 PIV: testing.RHR-CV-26CV 0.1224 gpm (< 5 (Test #3) ______gpm) Fall/2006 RE23 RHR-MOV- Evaluate Packing -Adjust LLRT: LLRT: Routine PM MO274A: or Repack -Repacked 0.75 scfh (< 35 scfh) 4.7 scth (< 35 scflh)PM 4446728 valve PIV: 0.041 gpm (< 5 gpm)RHR-MO- Examine Motor Operator MO274A______________________ ________PM 4446878 ____________________ Revision 0 Page 109 Revision 0 Page 109 Cooper Nuclear Station Ffith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 Component(s) ~ -Date Outage Work Order Work Order Description -AF Tests AL Tests Comments Spring/2008 RE24 N/A N/A -LLRT: N/A Assume 1 st 4.27 scfha (< 35 scfh) periodic test PIV: for LLRT 0.054 gpm (< 5 (and PIV)______gpm) test.Fall/2009 RE25 RHR-MO- Examine MO-Mech LLRT: N/A No impact on MO274A 8.31 scfha(< 35 scth) LLRT or PIV PM 4645290 PIV: testing. 2nd 0.082 gpm (< 5 periodic test RHR-CV- Adjust Reed Switches gpm) for LLRT 26CV (and PIV)CM 4723494 test.Spring/201 1 RE26 RHR-MOV- Evaluate Packing -Adjust _N/A PIV: No impact on MO274A: or Repack -Tightened 0.082 gpm (< 5 gpm) PIV test.PM 4744619: Packing LLRT no longer required due to closed ioop analysis. 3rd periodic test_______________________________________ ____________________________________________or__PforPIVsest Revision 0 Page 110 Revision 0 Page 110 Cooper Nuclear Station Fifth interval Inservice Testing Pro gram for Pumps and V'alves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 Component(s) -Date Outage Work Order Work Order Description AF Tests AL Tests Comments Fall/2012 RE27 RHR-CV- Repack Valve; performed N/A N/A No PIV test 26CV: exercise test and no due to PM 4848060 extemnal leakage at approved PIV pressure as PMT; PlV test relief request.not required as had no impact to seating ability and located inside Drywell.RIIR-CV- Adjust Limit Switch 26CV: CM 4918074 RHR-MO- Examine MO-Mech MO274A: PM 4848151 __________ Fai1/2014 RE28 N/A N/A N/A N/A No PIV test due to approved PIV___________________relief request.RIIR-MOV-Spring/2005 RE22 RiHR-MO- Refurbish MO LLRT: LLRT: AF ~AL MO274B MO274B Examine MO 9.6 scflh (< 35 scflh) 9.4 scflh (< 35 scfha) LLRT.& CM 4299766 PIV: RHR-CV-27CV PM 4363585 0.136 gpm (< 5 gpm)(Test #4) Fall/2006 RF23 RHR-MOV- Evaluate Packing -Adjust LLRT: PIV: No impact on MO274B or Repack -No packing 9.8 scfha (< 35 scflh) 0.109 gpm (< 5 gpm) LLRT or PIV PM 4446729 adjustment required, testing.Examine MO PM 4446875 Spring/2008 RE24 RHR-CV- Repacked Valve LLRT: LLRT: Assume 27CV: 14.5 scfha (< 35 scfih) 28.59 scfha (< 35 scfh) resets LLRT PM 4541360 PIV: frequency. _____________ ________ _______ ________ ________________ _____________0.163 gpm (< 5 gpm) ______Revision 0 Page 111 Revision 0 Page 111 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 Component(s) '-Date Outage Work Order Work Order Description AF Tests AL Tests " Comments Fall/2009 RE25 RHR-MO- Examine MO-Mech LLRT: N/A No impact on 274B: 15.7 scfha (< 35 scth) LLRT or PIV PM 4645289 PIV: tests.0.218 gpm ( <5 Assume 1st gpm) periodic test for LLRT and PIV test.Spring/201 1 RE26 RHR-MOV- Evaluate Packing -Adjust N/A PIV: LLRT no MO274B: or Repack -Tightened 0.326 gpm ( < 5 gpm) longer PM 4744620 packing required due to closed loop analysis. No impact on PLV test; 2nd periodic PIV_____________test. Fall/2012 RE27 PM 4848150 Examine MO-Mech N/A N/A No PIV test due to approved PIV relief request.Fall/2014 RE28 N/A N/A N/A N/A No PIV test due to approved PIV relief request.RHIR-MOV-M017 Spring/2005 RE22 PM 4363507 Examine MO and Verify LLRT: No impact on (Test #5) Indication 2.95 scfha (< 30 scfha) LLRT or PIV PIV: testing.PM 4363526 Examine MO Ci.027 gpm (< 5 Assume 1st gpm) periodic test for LLRT and________PIV. Fall/2006 RE23 PM 4446718 Clean and Lube Stem LLRT: No impact on CM 4535994 Perform Motor Pinion 1.75 scfh (< 30 scflh) LLRT or PIV Inspection PIV: testing. 2nd 0 gpm (< 5 gpm) periodic test for LLRT and____________________________________________PIV. Revision 0 Page ll2 Cooper Nuclear Station Ffifh Inter'al Inservice Testing Program for Pumps and VZ~lves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 Component(s) -'Date Outage Work Order Work Order Description AF Tests AL Tests Comments Spring/2008 RE24 CM 4546759 Replace MO and LLRT: PIV: No impact on diagnostic test; AL LLRT 0.68 scfha (< 30 scflh) 0 gpm (< 5 gpm) LLRT or PIV not required due to testing. 3rd minimal change in seat periodic test load, for LLRT and PlV.Fall/2009 RE25 PM 4645142 Clean and Lubricate PIV: LLRT not 0' gpm (< 5 gpm) required due to Option B;4th periodic PIV test.Spring/201 1 RE26 PM 4744696 Examine MO -Mech PIV: LLRT not CM 4740307 Motor Pinion Inspection C.027 gpm (< 5 required due PM 4744691 Examine MO -Elect gpm) to Option B;5th periodic PIV test.Fall/2012 RE27 PM 4848600 Examine MO LLRT: No PIV test 5.32 scfih (< 30 scfha) due to approved PIV relief request.Fa11/2014 RE28 PM 4983676 Examine MO N/A No tests due to Option B /PIV relief reqtuest.RHR-MOV-MO18 Spring/2005 RE22 PM 4363506 Examine MO and Verify LLRT: LLRT: Assume (Test #6) Indication 1.96 scflh (< 30 scflh) 2.06 scfha (< 30 scfih) resets LLRT PIV: frequency. CM 4212544 Refurb MO and diagnostic 0.0408 gpm (< 5 test gpm)PM 4363568 Examine MO Revision 0 Page 113 Revision 0 Page 113 Cooper Nuclear Station Ffifh Interv'a[Inservice Testing Pro gram for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 _____Component(s) Outage Work Order Work Order Description AF Tests AL Tests Comments Fall/2006 RE23 PM 4446727 Evaluate Packing Adjust or Repack -(tightened 2 flats); AL LLRT not required due to minimal change in packing and seating forces.Examine Motor Operator Examine MO and Verify Indication LLRT: 2.6scth(<30scfh) PIV: 0.109 gpm (<5 5gpm)No impact on LLRT or PIV testing. 1st periodic test for LLRT and PIV.PM 4446867 PM 4446860 Spring/2008 RE24 PM 4549525 Examine Motor Operator LLRT: No impact on 0.7 scfha (< 30 scflh) LLRT or PIV CM 4531750 Motor Pinion Gear PIV: testing. 2nid Inspection '. 109 gpm (< 5 periodic test gpm) for LLRT and PlV.Fall/2009 RE25 CM 4640553 Motor Pinion Gear ILLRT: PIV: LLRT no Inspection N'/A 0.218 gpm (< 5 gpm) longer required due to closed loop analysis. No impact on PIV test. 3rd periodic PIV test.Spring/201 1 RE26 PM 4744618 Evaluate Packing Adjust LLRT: PIV: No impact on or Repack -(1 flat): N/A 0.027 gpm (< 5 gpm) PIV test. 4th periodic PIV PM 4746148 Examine MO-Mech test._________ ________PM 4744690 Examine MO-Elec___________________ ______Pall/LULL KILL/I'M 48480U1 tExammne MU-Mechi SN!A N/A No impact on PIV test. No PIV test due to approved PIV relief request..1 _____________ 1 ____________ C ______________ C ________________________________________________ i ______________________ .1 Revision 0 Page 114 Revision 0 Page 114 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRTr Test History' Since 01/01/2005 Component(s) -Date Outage Work Order Work Order Description AF Tests AL Tests Comments Fali12014 RE28 CM 4945389 Viper Test -no torque N/A N/A No PIV test switch or packing due to adjustments required approved PIV (AF=AL); PIV not relief request.required PM 4949440 Evaluate Packing Adjust or Repack -Not needed PM 4950106 Examine MO (Mech/Elec) CS-MOV-MO12A Spring/2005 RE22 N/A N/A LLRT: N/A Periodic (Test #7) 0.004 scfha (< 10 LLRT andPIV test.PIV: Assume 1st (1.299 gpm (< 5 periodic PIV________gpm) test.08/02/2005 Online PM 4387217 Clean, Lubricate, Partial N/A N/A No impact on Stroke LLRT or PIV testing.08/02/2006 Online CM 4447691 Adjust Packing (tightened N/A N/A No impact on 2 flats); no AL LLRT/PIV LLRT or PIV required. testing.Fall/2006 RE23 CM 4531453 Motor Pinion Gear N/A PIV: .No impact on Inspection 0.435 gpm (< 5 gpm) LLRT or PIV testing.LLRT not required due to option B.2nd periodic PIV test.02/05/2008 Online PM 4532685 Examine MO -Mech N/A N/A No impact on LLRT or PIV__________________ ___________ _________ ___________ _______________________________________ _____________ testitesigg Revision 0 Pg 1 Page 115 Cooper Nuclear Station Fifth Interva!Inservice Testing Program for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 _____Component(s) ~ -Date jjOutage 1Work Order Work Order Description AF Tests AL Tests TComments Spring/2008 RE24 CM 4547083 CM 4561 197 Refurb and test MO Install ETT/QSS LLRT: 0.86 scfth (< 10 scflh)LLRT: 1.05 sofhi ( < 10 scfhi)PIV: 0.19 gpm (< 5 gpm)AF-AL LLRT. No impact on LLRT or PIV testing. 3rd periodic PlV test.Fall/2009 RE25 CM 4723418 Adjust close limit switch PIV: N/A No impact on 0 gpm (< 5 gpm) LLRT or PIV testing.LLRT not required due to option B.4th periodic PIV test.Spring/201 1 RE26 N/A N/A PIV: N/A LLRT not C, gpm (< 5 gpm) required due to option B.5th periodic PIV test.08/10/2011 Online PM 4749833 Clean, Lubricate, and N/A N/A No impact on Partial Stroke LLRT or PIV testing.Fall/2012 RE27 PM 4848626 Examine MO (Mech & LLRT: N/A No impact on Elec); 0.1528 scfh (< 10 LLRT or PlV sefh) testing. 6th PIV: periodic PIV_______ ______ _______0 gpm (< 5 gpmn) __________test. Fall/2014 RE28 CM 4945454 PM 4950123 Viper Test; AL LLRT/PIV tests not required due to minimal change in packing and seating forces.Examine MO (Clean/Lube Stem)LLRT: 1.1 scth (< 10 scfhi)No impact on LLRT or PIV testing. PIV test not required due to approved relief request.Revision 0 Pg 1 Page 116 Cooper Nuclear Station Ffifh Interval lnservice Testing Pro gram for Pumps and Vailves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 _____Component(s) ~ -Date jOutage Work Order LWork Order Description ] AF Tests ] AL Tests Comments CS-MOV-MO12B (Test #8)Spring/2005 RE22 N/A N/A LLRT: 1 .23 scfih (< 10 scflh)PIV: 0 gpm (< 5 gpm)N/A Periodic LLRT and assume 1st periodic PIV test.08/14/2006 Online PM 4465037 Clean/Lube/Partial Stroke -N/A N/A No impact on LLRT or PIV CM 4334765 Periodic Diagnostic Test; testing.no AL LLRT/PIV_______ Fall/2006 RE23 CM 4534089 Motor Pinion Gear PIV: N/A No impact on Inspection .C gpm (< 5 gpm) LLRT or PIV testing.LLRT not required due to option B.2nd periodic PIV test.Spring/2008 RE24 CM 4561198 Install ETT/QSS PIV: N/A No impact on 0.082 gpm (< 5 LLRT or PIV gpm) testing.LLRT not required due to option B.3rd periodic PIV test.Fa11/2009 RE25 PM 4658094 Examine & Clean LLRT: N/A No impact on Operator 1.67 scfha (< 10 scfha) LLRT or PIV PIV: testing. 4th 0 gpm (< 5 gpm) periodic PIV test.11/12/2009 Online PM 4625209 Clean/Lube/Partial Stroke N/A N/A No impact on LLRT or PIV testing.Revision 0 Pg 1 Page ll7 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 Component(s) -'Date Outage Work Order Work Order Description AF Tests AL Tests Comments Spring/201 1 RE26 PM 4767601 Examine Motor Operator PIV: N/A No impact on 0Ogpm (< 5gpm) LLRT or PIV testing.LLRT not required due to option B.4th periodic___________PIV test.Fall/2012 RE27 CM 4840074 Viper Test LLRT: LLRT: No impact on 1.82 scfih (< 10 scfh) 2.02 scfh LLRT or PIV PM 4848541 Examine MO (Mech/Elec) PIV: testing. 5th 0.0 136 gpm (< 5 periodic PIV________gpm) test.Fall/2014 RE28 PM 4950054 Examine MO (Clean/Lube N/A N/A No impact on Stem) LLRT or PIV testing. No LLRT test performed due to option B and no PIV test performed due to an approved relief request.CS-CV-18CV Spring/2005 RE22 N/A N/A PIV: N/A LLRT not (Test #9) 0 3264 gpm (< 5 required due gpem) to Option B;periodic PIV______________test. Fa11/2006 RE23 N/A N/A PIV: N/A LLRT not 0.326 gpm (< 5 required due gpm) to Option B;periodic PIV___________ __________ ___________ _______________________________________ _______________ testest Revision 0 Page 118 Revision 0 Page 118 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 Component(s) '-Date Outage Work Order Work Order Description AF Tests AL Tests Comments Spring/2008 RE24 N/A N/A LLRT: N/A Periodic 1.19 scflh (< 15 scifi) LLRT and PIV: PIV tests.0.136 gpm (< 5____________ ~~~gpm) __________ ______Fall/2009 RE25 PM 4645121 Repack Valve LLRT: LLRT (Final AL): Significant CM 4724012 Disassemble and Repair 0.13 1 scfih (< 15 0.79 scfha (< 15 scfla) Maint. Resets following issues during scfh) PIV: LLRT/PIV repack 0 gpm (< 5 gpm) frequency. Spring/2011 RE26 N/A N/A PIV: N/A LLRT no 0 gpm (< 5 gpm) longer required due to closed loop analysis.First periodic PIV test.Fall/2012 RE27 N/A N/A PIV: N/A Second o gpm (< 5 gpm) periodic PIV test.Fall/2014 RE28 N/A N/A N/A PIV test not required due to approved relief request.CS-CV-19CV Spring/2005 RE22 N/A N/A LLRT: N/A Periodic (Test #10) 0.95 scfh (< 15 scflh) LLRT and PIV: PIV tests.0 gpm (< 5 gpm)Fall/2006 RE23 N/A N/A PIV: N/A LLRT not 0 gpm (< 5 gpm) required due to Option B;periodic PIV____________test. Spring/2008 RE24 CM 4631924 Adjust/add packing LLRT: LLRT (Final AL): Elected to PM 4541346 Repack valve 0.65 scfha (< 15 sctha) 1.4 scfha Reset PIV (Final AL): LLRT/PIV______________________________ ________0.05 gpm (< 5 gpm) frequency. Revision 0 Pg 1 Page 119 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 _____Component(s) '-Date Outage Work Order Work Order Description AF Tests AL Tests Comments Fall/2009 RE25 N/A N/A LLRT: N/A First periodic 1.37 scih (< 15 setha) LLRT and PIV: PIV test.o gpm (< 5 gpm)Spring/2011 RE26 N/A N/A PIV: N/A LLRT no o gpm (< 5 gpm) longer required due to closed ioop analysis.Second periodic PIV test.Fall/2012 RE27 N/A N/A N/A N/A No PIV test required due to approved relief request.Fall!2014 RE28 N/A N/A N/A N/A No IPIV test required due to approved__________________ ___________ _________ ____________ _____________________ _________________ _________re___ie__ rlie ureqest AF =As Found AL = As Left CM = Corrective Maintenance PM = Preventative Maintenance Revision 0 Pg 2 Page 120 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves ATTACHMENT 5 AUGMENTED VALVE RELIEF REQUESTS AUGMENTED VALVE RELIEF REQUEST INDEX Relief [Description 1CNS Approval Date Request No._____ARV-01 DGDO Day Tank Valve Test Method ] 3-1-20 16 ARV-02 Diesel Fuel Oil Relief Valve Test Media ] 3-1-20 16 ARV-03 IfDGDO Check Valve Closure Tests j 3-1-20 16 Revision 0 Pg 2 Page 121 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request ARV-01 DGDO Day Tank Valve Test Method Alternative Provides Acceptable Level of Quality and Safety 1. Augmented Code Component(s) Affected Valve Class Category System DGDO-SOV-SSV5028 A B DGDO DGDO-SOV-SSV5029 A B DGDO 2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTC-5 151 (a), Active valves shall have their stroke times measured when exercised in accordance with ISTC-3500. ISTC-3560, "Fail-Safe Valves." Valves with fail-safe actuators shall be tested by observing the operation of the actuator upon loss of valve actuating power in accordance with the exercising frequency of ISTC-35 10.4. Reason for Request Augmented Relief is requested from the requirements of table ISTC-5 151l(a) and ISTC-3560, as applicable, for the valves listed above. The proposed alternative provides an acceptable level of quality and safety. This Augmented Relief Request does not require NRC approval.5. Proposed Alternative and Basis for Use The DG Day Tank Fuel Safety Solenoid Valves (DGDO-SOV-SSV5028, SSV5029) have a passive safety function in the open position to permit filling of the associated Day Tank during fuel oil transfer operations. These solenoid operated valves also have an active defense-in-depth (non-safety) function in the closed position to prevent overfilling of the associated Diesel Fuel Tank. However, this is only a back-up function to the safety-related day tank high level switch function to stop the fuel oil transfer pump in the same division on a high level in the day tank.DGDO-SOV-SSV-5028, 5029 are encapsulated solenoid valves and are not provided with remote position indication or remote manual switches. The design of these valves prohibits visual verification of the physical position of the valve operator, stem, or internal components. If the safety-related high level switch fails, the solenoid valve will close at the high-high level.Revision 0 Pg 2 Page 122 Cooper Nuclear Station Fifth Interval lnservice Testing Program for Pumps and Valves Relief Request ARV-01 DGDO Day Tank Valve Test Method (Continued) Modification of the system to verify individual valve exercising capability is not practicable nor cost beneficial since no commensurate increase in safety would be derived. These valves are ASME non-code class valves and are not within the scope of the IST Program.Since the Diesel Fuel Oil transfer system is an Augmented IST System, the components within this system are not required to follow the requirements of class 1, 2, or 3 components. However, components are required to be tested in a manner that is commensurate with the level of safety that they provide. Although these valves are considered to be passive open valves, CNS will conservatively test these valves periodically in the open position. At least quarterly, the diesel fuel oil day tank level alarms and transfer pump control level switches are functionally tested during diesel runs. The solenoid valves shall be verified open by observing restoration of day tank levels and measurement of sufficient flows during the IST transfer pump test surveillances. The testing being performed will ensure that these valves may be relied upon to fulfill their open safety function.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents A version of this relief request was previously approved by CNS for the fourth ten-year interval as ARV-O 1.Revision 0 Page 123 Revision 0 Page 123 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Relief Request ARV-02 Diesel Fuel Oil Relief Valve Test Media Alternative Provides Acceptable Level of Quality and Safety 1. Augmented Code Component(s) Affected Valve Class Category System DGDO-RV-l10RV A C DGDO DGDO-RV-l11RV A C DGDO 2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004 Edition through 2006 Addenda 3. Applicable Code Requirement Mandatory Appendix I, paragraph 1-4130(a), "Test Media." Valves shall be tested with the normal system operating fluid and temperature for which they are designed. Alternative liquids and different temperatures may be used, provided the requirements of 1-43 00 are met.4. Reason for Request Augmented Relief is requested from the requirements of Mandatory Appendix I, paragraph I-4130(a), concerning the liquid test media. The proposed alternative provides an acceptable level of quality and safety. This Augmented Relief Request does not require NRC approval.5. Proposed Alternative and Basis for Use These augmented relief valves provide over pressure protection of the engine driven fuel oil pump suction due to thermal expansion. These relief valves are periodically tested onsite at CNS. However, performing this relief valve set pressure testing with diesel fuel oil would be an unsafe practice due to the dangerous characteristics associated with this media. The fuel oil is flammable, and could be damaging to the eyes and/or skin. Utilizing the fuel oil as the test media would also contaminate the test equipment, making it difficult to clean and re-use this test equipment for other relief valve testing.In order to alleviate the concerns with utilizing diesel fuel oil as the test media for these relief valves, water may be used with negligible effects on the results of the relief valve testing performed to satisfy Mandatory Appendix I. The diesel fuel oil is slightly more viscous than water, but this should not affect the 1ST set pressure or seat leakage tests for the following reasons. The set pressure is determined by applying a static force against the upstream portion of the valve disc and recording the initial gush or first sign of continuous flow through the valve.Revision 0 Pg 2 Page 124 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request ARV-02 Diesel Fuel Oil Relief Valve Test Media (Continiued) Utilizing fuel oil or water should have a negligible effect on this test. In support of this statement, the Anderson Greenwood Crosby Test Report #5595, which compared opening pressures with water, lube oil and fuel oil, concluded that a one-to-one correlation exists between Fuel oil and water. The test results indicated no significant differences in the opening pressures. The seat leakage testing criteria is set at zero leakage at 90% of set pressure. If water passes this test, then diesel fuel oil should also pass the test since water is less viscous than diesel fuel.These valves are Seismic Class IS and are ASME non-Code Class. Therefore, they are outside the scope of the 1ST requirements of 10CFR50.55a. The valves are included in the 1ST Augmented Program.Relief Valve testing will be performed per Mandatory Appendix I, with the exception that water will be utilized as the test media rather than diesel fuel oil.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was previously approved by CNS for the fourth ten-year interval as ARV-02.Revision 0 Page 125 Revision 0 Page 12 5 Cooper Nuclear Station Fifth Interval Inser-vice Testing Pro gram for Pumps and Valves Relief Request ARV-03 DGDO Check Valve Closure Tests Alternative Provides Acceptable Level of Quality and Safety 1. Augmented Code Component(s) Affected Valve Class Category System DGDO-CV-l0CV A C DGDO DGDO-CV-1lCV A C DGDO DGDO-CV-12CV A C DGDO DGDO-CV-13CV A C DGDO 2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTC-5221 (a), "The necessary valve obturator movement during exercise testing shall be demonstrated by performing both an open and a close test." 4. Reason for Request Augmented Relief is requested from the requirements of ISTC-522 1(a), concerning the closure testing of these check valves. The proposed alternative provides an acceptable level of quality and safety. This Augmented Relief Request does not require NRC approval.5. Proposed Alternative and Basis for Use DGDO-CV- 10CV and DGDO-C V- 11CV are the fuel oil transfer pump discharge check valves and DGDO-CV-12CV and DGDO-CV-13CV are the day tank inlet check valves. These augmented check valves have a safety function in the open position to allow transfer of fuel oil to their respective diesel fuel oil day tank during normal diesel generator operation. These check valves have no safety function to close, but the code requires bi-directional testing of check valves.Some plants may have elected to classify the Diesel Fuel Oil Transfer pumps and their associated check valves as being skid-mounted in accordance with the ISTA-2000 definition for skid-mounted pumps/valves and the ISTC-l1200 exemptions. This position would allow these components to be considered adequately tested with an acceptable diesel generator run.However, CNS has elected to include the pumps within the augmented 1ST Program to better track the performance of these pumps. As a part of this pump testing, the check valves associated with each transfer pump are tested to the open position by verification that the valves can deliver Revision 0 Pg 2 Page 126 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request ARV-03 DGDO Check Valve Closure Tests (Continued) > 4.64 gpm, which is the flow required to support continuous operation of a diesel generator at full load. The closure function, however, cannot be tested within this surveillance. The fuel oil transfer pump headers are isolated downstream of the check valves via normally closed cross-tie manual valve, DGDO-V-l19. With this manual valve in the passive closed position, the potential of a diversion of fuel oil from one division to the other is eliminated. Satisfactory surveillance testing of the diesel generators and fuel oil transfer pumps ensure that sufficient fuel flow is transferred to the day tanks. Finally, the diesel fuel oil discharge header cross-tie valve, DGDO-V-1 6, is a normally closed passive manual valve that is never opened during normal operations or during transient or accident conditions, which alleviates the conccrn of backflow through this line.Since the Diesel Generator Diesel Fuel Oil transfer System is an Augmented IST System, the components within this system are not required to follow the requirements of class 1, 2, and 3 components. However, the components are required to be tested in a manner that is commensurate with the level of safety that they provide. Testing DGDO-CV-10CV, DGDO-CV 11CV, DGDO-CV-l12CV, and DGDO-CV-l13CV in the open position clearly demonstrates that the check valves within the diesel fuel oil transfer system are adequately tested and may be relied upon to fulfill their safety functions. This is consistent with CNS Engineering Evaluation, EE 08-026, which reconfigured DGDO-V-l19 from open to closed. For these reasons, the closure testing of these check valves are not required.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was previously approved by CNS for the fourth ten-year interval as ARV-03.Revision 0 Page 127 Revision 0 Page 127 Cooper Nuclear Station F~fih Interval Inservice Testing Program for Pumps and Valves ATTACHMENT 5 GENERAL RELIEF REQUESTS VALVE RELIEF REQUEST INDEX Relief Request Tecito R prvlDt No.DeritoNRAprvlDe RG-O1 [ASME OM Code Test Frequencies Pending Revision 0 Pg 2 Page 12 8 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves ASME OM Code Test Frequencies Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2) Hardship or Unusual Difficulty without a Compensating Increase in Level of Quality and Safety 1. ASME Code Component(s) Affected All Pumps and Valves contained within the Inservice Testing Program (IST) scope.2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement This request for relief applies to the frequency specification of the ASME OM Code for all pump and valve testing contained within the IST Program scope. The applicable ASME OM Code (2004 Edition through the 2006 Addenda) sections include the following: Code Paragraph Description ISTA-320(a)The frequency for inservice testing shall be in accordance with the ISTA3 12(a)requirements of Section 1ST ISTB-3400 Frequency of I~nservice Tests ISTB-6200 Corrective Action ISTC-35 10 Exercising Test Frequency ISTC-3540 Manual Valves ISTC-3560 Fail-Safe Valves ISTC-3630(a) Frequency ISTC-3700 Position Verification Testing At least one valve from each group shall be disassembled and examined ISTC-522 1 (c)(3) at each refueling outage; all valves in a group shall be disassembled and examined at least once every 8 years.ISTC-5222 Condition-Monitoring Program ISTC-5230 Vacuum Breaker Valves ISTC-5 240 Safety and Relief Valves ISTC-5260 Explosively Actuated Valves Appendix 1*, 1-1320 Test Frequencies, Class 1 Pressure Relief Valves Appendix I, 1-1330 Test Frequency, Class 1 Nonreclosing Pressure Relief Devices Test Frequency, Class 1 Pressure Relief Valves That Are Used for Appedix , 1-340 Thermal Relief Applications Appendix I, 1-1350 Test Frequency, Classes 2 and 3 Pressure Relief Valves Appendix I, 1-1360 Test Frequency, Classes 2 and 3 Nonreclosing Pressure Relief Devices Test Frequency, Classes 2 and 3 Primary Containment Vacuum Relief Appendix I, 1-1370 Vle Test Frequency, Classes 2 and 3 Vacuum Relief Valves, Except for Appedix , 1-380 Primary Containment Vacuum Relief Valves Revision 0 Pg 2 Page 129 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RG-01 ASME OM Code Test Frequencies (continued) Code Paragraph Description Test Frequency, Classes 2 and 3 Pressure Relief Devices that are Used Appedix , 1-390 for Thermal Relief Application. Appendix II**, II- Performance Improvement Activities 4000(a)Appendix II, II- Optimization of Condition-Monitoring Activities 4000(b) ______________________________

  • Appendix I is for Pressure Relief Devices** Appendix II is for the Check Valve Condition Monitoring Program (CVCM)4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(2), relief is requested from the frequency specification of the ASME OM Code. The basis of the Relief Request is that the Code requirement presents an undue hardship without a compensating increase in the level of quality and safety.The ASME GM Code, 2004 Edition through the 2006 Addenda, establishes the inservice test frequency for all components within the scope of the Code. The frequencies (e.g., quarterly) have always been interpreted as "nominal" frequencies (generally as defined in Table 3.2 of NUREG 1482, Revision 2) and if necessary, owners applied the surveillance extension time period (i.e.grace period) contained in the plant Technical Specifications (TS) SRs. The CNS TS SR 3.0.2 states that the specified frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency.

This would allow an extension of up to 25% of the surveillance test interval to accommodate plant conditions that may not be suitable for conducting the surveillance. However, regulatory issues have been raised concerning the applicability of the TS grace period to ASME GM Code required inservice test frequencies. The lack of a tolerance band (grace period) on the ASME GM Code IST frequency restricts operational flexibility. There may be a conflict where an IST test could be required (i.e., its frequency could expire), but it is not possible or not desired that it be performed until sometime after a plant condition or associated TS is applicable. Therefore, to avoid this conflict, the IST test intervals should be allowed to be extended by up to 25%.Thus, just as with TS required surveillance testing, some tolerance is needed to allow adjusting GM Code testing intervals to suit the plant conditions and other maintenance and testing activities. This assures operational flexibility when scheduling 1ST tests that minimize the conflicts between the need to complete the test and plant conditions.

5. Proposed Alternative and Basis for Use Code Case OMN-20 is included in the ASME GM Code, 2012 Edition, and will be used as an alternative to the frequencies of the ASME GM Code. The requirements of Code Case OMN-20 are described below.Revision 0 Pg 3 Page 13 0 Cooper Nuclear Station Fifthi Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RG-01 ASME OM Code Test Frequencies (continued)

ASME OM, Division 1, Section 1ST and all earlier editions and addenda specify 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.).(a) 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 as shown in Table 1. The specified time period between tests may be reduced or extended as follows: (1) For periods specified as fewer than 2 years, the period may be extended by up to 25% for any given test.(2) For periods specified as greater than or equal to 2 years, the period may be extended by up to 6 months for any given test.(3) All periods specified may be reduced at the discretion of the owner (i.e., there is no minimum period requirement). Period extension is to facilitate test scheduling and considers plant operating conditions that may not be suitable for performance of the required testing (e.g., performance of the test would cause an unacceptable increase in the plant risk profile due to transient conditions or other ongoing surveillance, test, or maintenance activities). Period extensions are not intended to be used repeatedly merely as an operational convenience to extend test intervals beyond those specified. Period extensions may also be applied to accelerated test frequencies (e.g., pumps in alert range)and other fewer than 2 year test frequencies not specified in Table 1.Period extensions may not be applied to the test frequency requirements specified in Subsection ISTD, Preservice and Inservice Examination and Testing of Dynamic Restraints (Snubbers) in Light-Water Reactor Nuclear Power Plants, as Subsection ISTD contains its own rules for period extensions.(b) Components whose test frequencies are based on the occurrence of plant conditions or events may not have their period between t ests extended except as allowed by ASME OM, Division 1, Section IST, 2009 Edition through OMa-20 11 Addenda and all earlier editions and addenda.Revision 0 Page 131 Revision 0 Page 131 Cooper Nuclear Station F~fih Interval Inservice Testing Program for Pumps and Valves Relief Request RG-01 ASME OM Code Test Frequencies (continued) Table 1 Specified Test Freauencies Frequency Specified Time Period Between Tests Quarterly92dy (or every 3 months)92dy Semiannually 14dy (or every 6 months) 14dy Annually36 (or every year) 36days x yearsx calendar years where x is a whole number of x_____years_____________years > 2 6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was previously approved for the Fermi-2 third ten-year interval as Relief Request PVRR-001 (TAG No. MF2967, dated July 16, 2014).Three Mile Island Nuclear Station, Unit 1 -Relief Requests PR-0 1, PR-02, and VR-02, Associated With The Fifth 10-Year Inservice Test Interval (TAG Nos. MF0046, MF0047 and MF0048, dated August 15, 2013).Revision 0 Pg 3 Page 132 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves ATTACHMENT 7 COLD SHUTDOWN JUSTIFICATIONS COLD SHUTDOWN JUSTIFICATION INDEX Cold Shutdown Derito Justification Derito No.[CSJ-02 ]jLPCI-AOV-PCV50 Exercising K CSJ-03 ]jMS-AO V-AO80AiB/C/D and MS-AGOV-AO86A/B/C/D Exercising -CSJ-04 ]~RR-MOV-MO53AIB Exercising [CSJ-05 ]~RHR-MOV-MO17/18 Exercising [CSJ-06 ][RHR-MOV.-920MV/921MV Exercising CSJ-07 ][SW-AOV-TCV451lA/B Exercising CSJ-08 tt IV-MOV-262/264/266/268MV Exercising (Augmented) CSJ-09 ]~CS-MOV-MO 12AIB Exercising-CSJ-1 0 ][Exercising of Backseated Valves Revision 0 Pg 3 Page 133 Cooper Nuclear Station F~ifth Interval Inservice Testing Pro gram for Pumps and Valves Cold Shutdown Justification CSJ-O1 (Augmented) Valve Number System Class Categ~ory, SGT-CV-14CV SGT A C SGT-CV-15CV SGT A C Function These check valves open to provide flow paths from their respective SGT filter trains and close to prevent back flow from the operating SGT train discharge. Justification The required full flowrate for these valves is > 1602 cfmn. During plant power operations, system conditions exist that prevent the SGT system from achieving a flowrate of > 1602 cfm. Reactor building differential pressure and back pressure from the Off Gas Dilution Fans act against SGT system pressure restricting SGT system flowrate. Therefore, it is impracticable to perform a fall flowrate test of these valves at power operations. These valves are ASME non-code class and are not within the scope of the IST Program.Alternative Test A full flow open test will be performed on these valves during cold shutdown periods. The full flow test shall also satisfy the closure exercise requirements of the check valve in the idle train.Revision 0 Page 134 Revision 0 Page 134 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Cold Shutdown Justification CSJ-02 Valve Number System Class Category, IIPCI-AO V-PC V50 HIPCI 2 B Function Air operated, pressure regulating valve for the cooling water supply line to the HPCI lube oil cooler. The valve perforns an active safety function in the open/throttled position to allow cooling water flow to the lube oil cooler.Justification Thifs valve functions to control pressure in the cooling water supply line to the H{PCI turbine lube oil cooler. Cooling water is supplied from the HPCI booster pump discharge. The valve is normally maintained in the closed position as a result of the HPCI pump being idle and pressure maintenance, supplied by the auxiliary condensate system, maintaining pressure above the control valve's set point.The valve travels to a throttled position when the HPCI pump starts to automatically maintain pressure im the cooling water line at 50 psig. The valve is designed to fail to the open position on a loss of instrument air to ensure continuity of cooling water flow to the lube oil cooler. Fail open travel is limited by a travel stop in order to prevent the downstream relief valve from lifting. The travel stop is set such that the required flow rate to the lube oil cooler is met under all operating conditions. Stroke timing and fail-safe testing to the open position online would require HIPCI to be declared inoperable and the LIPCI pressure controller, HTPCI-PC-50, to be taken out of service. Then, with instrument air isolated, a test rig would be utilized to apply air to the HPCI-AO V-PC V50 actuator in order to close the valve. Upon completion of valve closure, air to the actuator would then be removed, allowing the valve to be timed as it travels to the open position. This complex process to time this valve open is impracticable to perform every quarter independently during power operation and is also impracticable to perform in conjunction with the HLPCI run. HIPCI testing duration is limited by the resulting suppression pool heat up and Tech Spec temperature limitations. Alternative Test This pressure control valve shall be stroke timed and fail-safe tested open during cold shutdowns by manipulating the valve position by controlling air to the actuator, as discussed above. A partial stroke exercise test will be considered to be performed quarterly with the HPCI pump test.Revision 0 Page 135 Revision 0 Page 13 5 Cooper Nuclear Station F/ifth Interval Inservice Testing Pro gram for Pumps and Valves Cold Shutdown Justification CSJ-03 Valve Number System Class Category.MS-AGOV-AO8OA MS 1 A MS-AO V-AO80B MS 1 A MS-AGOV-AO80C MS 1 A MS-AOV-AO80D MS 1 A MS-AOV-AO86A MS 1 A MS-AO V-A086B MS 1 A MS-AO V-A086C MS 1 A MS-AOV-AO86D MS. 1 A Function The inboard and outboard main steam isolation valves (MSIVs) must be capable of automatic closure to limit the release of radioactivity during a reactor transient or accident condition and to prevent damage to the fuel barrier by limiting the loss of reactor coolant. water in case of a major leak from the steam piping outside of primary containment. Juotification Quarterly full closure testing of the MSJVs during 100% power operation is impracticable due to the potential for reactor transients and scrams. Also, full MSIV closure could create the potential of lifting the main steam safety relief valves (SRVs) due to an increase in steam line pressure. Failure of an SRV to re-close could result in reactor vessel depressurization. These affects may be minimized by performance at a reduced power level of < 70%, but then this activity contributes to the financial burden of reducing reactor power levels to facilitate valve testing. More importantly, the full stroke testing of MSIVs, even at reduced power, places the plant in an abnormal operating condition and introduces an unnecessary challenge to plant equipment. For example, the MSIVs are challenged to close and then re-open with steam in the lines, the plant must stabilize following the isolation and un-isolation of a Main Steam Line. Also, the testing has the potential to cause the plant to remain at a reduced power level and/or cause the initiation of a shutdown in order to make repairs.This would introduce additional equipment cycling and plant thermal transients. Therefore, the testing conditions associated with full-stroke testing the MSJVs online meets the deferral criteria outlined in NUJREG-1 482 Revision 2, section 2.4.5.Finally, per Technical Specification surveillance requirement, SR 3.6.1.3.6, which verifies that the isolation time of each MSIV is between 3 and 5 seconds, the frequency of the test is to be determined by the IST Program (via this cold shutdown). Technical Specification surveillance requirement, SR 3.3.1.1.9, requires a channel functional test to be performed every 92 days and may be satisfied by a partial stroke test closed.Alternative Test The MSJVs shall be partially exercised closed from the full open position, at least once per quarter, to satisfy Technical Specification requirement, SR 3.3.1.1.9, and in accordance with ASME GM Code ISTC-3520. Stroke timing to the closed position and fail-safe testing closed shall be performed on a cold shutdown basis in accordance with ASME GM Code ISTC-3 520.Revision 0 Page 136 Cooper Nuclear Station F~ith Interval Inservice Testing Program for Pumps and Valves Cold Shutdown Justification CSJ-04 Valve Number System Class Category RR-MOV-MO53A RR 1 B RR-MOV-M053B RR 1 B Function These valves are the Reactor Recirculation Pump 1 A and 1 B Discharge Isolation valves and have an active safety function to close in order to prevent diversion of LPCI injection flow following a LOCA.Justification Closure of either of the RR pump discharge valves at power would reduce recirculation flow and result in reactor water temperature transients and reactivity transients. These transients would reduce control of power distribution and fuel usage, and increase the risk of other plant transients. This could lead to decreased fuel reliability and increase the possibility of a fuel element failure. In addition, failure of these valves during operation would require reactor shutdown due to inaccessibility. Alternative Test These valves will be exercised (and stroke timed) to the closed position during cold shutdowns when the reactor recirculation system is not required to be mn service.Revision 0 Pg 3 Page 137 Cooper Nuclear Station F~ifh Interval Inser-vice Testing Program for Pumps and Valves Cold Shutdown Justification CSJ-05 Valve Number System Class Category.RHR-MOV-MO017 RI-ll 1 A RHR-MOV-MO018 RNR 1 A Function These valves are the reactor vessel return to the RHIR pump suction and containment isolation valves during reactor operations. These valves are only opened for low pressure shutdown cooling.Justification Valves RHR-MO V-MO 17 and RHIR-MO V-MO 18 are interlocked for pressure isolation during plant operation. Opening these valves during normal operation could possibly allow high pressure reactor coolant water into the low pressure suction lines of the RHR system. Therefore, it is essential that these valves remain closed during plant operations. Alternative Test These valves will be exercised (and stroke timed) to the closed position during cold shutdowns when reactor pressure isolation is not required.Revision 0 Pg 3 Page 13 8 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Cold Shutdown Justification CSJ-06 Valve Number System Class Category.RHR-MOV-920MV RER 2 B RHR-MOV-921MV RHR A B Function These valves provide isolation of main steam to the Augmented Off Gas (AOG) system.Justification The steam supply cannot be isolated during normal plant operation without causing significant Augmented Off Gas (AOG) system transients. Transients could include a fast or uncontrolled burn of hydrogen gas in the AOG piping buried underground and leading outside the plant. Also, routine quarterly testing of either of these valves could cause a release of radioactive material several orders of magnitude above normal release activities. Alternative Test These valves will be exercised (and stroke timed) to the closed position during cold shutdowns when the steam supply to the augmented off gas system may be isolated.Revision 0 Pg 3 Page 139 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Cold Shutdown Justification CSJ-07 Valve Number System Class Category.SW-AOV-TCV451A SW 3 B SW-AOV-TCV451B SW 3 B Function These valves open to provide a flow path for cooling water to the REC heat exchangers. Justification One temperature control valve is normally open to control flow to the associated REC heat exchanger. During the hot summer months both heat exchangers are in service. Placing either valve in the closed position for an exercise test during this period would interrupt the flow to the associated heat exchanger. The REC heat exchangers provide cooling water for a variety of essential and non-essential components. Therefore, it is essential that both of these valves remain open during plant operations. During cold shutdowns, when the heat load is reduced, one REC heat exchanger can be removed from service. The associated temperature control valve can then be closed and exercised to the full open position.Alternative Test Valve exercising (and stroke timing) to the open position will be performed quarterly except when both heat exchangers are in service. When both heat exchangers are in service, valve exercising (and stroke timing) to the open position will be performed during cold shutdowns, when the heat load is reduced.When a refueling outage falls within the seasonal period in which these valves may be tested quarterly, then an additional test during the refueling outage will not be required (i.e. valves may be tested online just prior to and after the refueling outage, being maintained on a quarterly frequency). Revision 0 Pg 4 Page 140 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Cold Shutdown Justification CSJ-08 (Augmented) Valve Number System Class Category.HV-MOV-262MV HV A B HV-MOV-264MV I-IV A B HV-MOV-266MV H-V A B HV-MOV-268MV LIV A B Function These valves open to provide flow paths for ventilation to be supplied through the casing of both MG sets for cooling purposes, and close upon receipt of a PCIS Group VI signal to provide secondary containment isolation. Justification These valves are required to remain in the open position during power operation to support reactor recirculation pump operation. Closure of these valves during reactor operation could result in overheating the MG set which would compromise reactor recirculation pump operation causing plant shutdown.Also, the valves' control circuitry does not provide for partial stroke capability. These valves are ASME non-code class valves and are not within the scope of the 1ST Program.Alternative Test These valves will be exercised (and stroke timed) to the closed position during cold shutdowns when the reactor recirculation pumps are in an idle state. During unscheduled cold shutdowns when the recirculation pumps are required to remain in operation, valve exercising will be deferred until the next available opportunity when the recirculation pumps can be removed from service.Revision 0 Pg 4 Page141 Cooper Nuclear Station F~ith Interval Inservice Testing Program for Pumps and Valves Cold Shutdown Justification CSJ-09 Valve Number System ClassCaeoy C S-MOV-MO 12A C s 1 A CS-MOV-MO12B CS 1 A Function Open to admit Core Spray water to the reactor vessel to mitigate the consequences of a LOCA and close for primary containment / pressure isolation valve functions. Justification These valves are normally closed for primary containment isolation and to isolate the Core Spray System from the reactor vessel pressure. Operating these valves at RCS pressures above 450 psig (quarterly) would require them to be manually cracked off their seat in the open direction. This action serves to equalize pressure across the valves since downstream of the valves may be pressurized by minor leak-by of the inboard check valves, CS-C V-i 8CV and CS-CV-19CV. Pressure equalization greatly decreases the forces required to pull the valve disk out of the seat and makes it possible to electrically stroke time the valve.Design calculation, NEDC 95-003, lists the MOV limiting components for the CS valves. The stem to disc T-head is the most limiting sub-component for the opening direction at normal operating conditions. The forces affecting the valve stem T-head are equivalent whether the valve is manually opened or electrically opened. Pressure forces acting on the disks must still be overcome by tension on the T-head.Manipulation of the equations in NEDC 9 5-003 that calculate opening forces indicate that the valve stem T-head limits could be exceeded under maximum reactor pressure and worse-case seat friction coefficients. In order to eliminate the possibility of overstressing the valve T-head during the quarterly surveillances, the stroke time test should be limited to periods during cold shutdowns when the reactor pressure is below 450 psig.Alternative Test Valve exercising (and strke timing) to the open and closed position for CS-MO V-MO 12A/B will be performed during cold shutdowns when the reactor pressure is below 450 psig.Revision 0 Pg 4 Page 142 Cooper Nuclear Station Ffith Interval Inser-vice Testing Program for Pumps and Valves Cold Shutdown Justification CSJ-IO Valve Number System Class Category.H{PCI-MOV-MO015 HPCI 1 A MS-MOV-MO74 MS 1 A RCIC-MOV-MO015 RCIC 1 A Function HPCI-MO V-MO 15: Normally open to provide steam to the HPCI turbine and closes automatically on high flow, low pressure, or high temperature signals.MS-MOV-M074: Normally open valve provided to prevent condensed steam from accumulating in the main steam lines and closes to isolate the main steam line inboard drain penetration. RCIC -MO V-MO 15: Normally open to provide steam to the RCIC turbine and closes automatically on high flow, low pressure, or high temperature signals.Justification This cold shutdown justification only applies if one or more of the motor operated valves listed have been power backseated in an effort to reduce Drywell leakage. This method of backseating is performed outside of the Drywell at the motor control centers for the valves. The valves, themselves, are located inside the Drywell. However, the backseated valves will be verified to meet any analyzed closed stroke time limits from the power backseated position prior to declaring the valve operable. The valve would then be placed back into the same backseated configuration and returned to service. If one or more of these valves have been backseated in this manner, it would be impracticable to perform subsequent stroke time testing each quarter based on the following discussion. Repeat backseating each quarter introduces the potential for causing component damage each quarter in addition to potentially re-initiating a Drywell packing leak that had previously been stopped. Power backseating of motor operated valves entails bypassing the normal valve open limit control and using the actuator motor to stall the valve into its backseat. The end result is that the valve stem seals off leakage past the packing by interference with a beveled backseat area in the valve bonnet. Valve vendors have supplied the maximum allowable backseat forces that can be applied. Although there are some risks involved, power backseating has been successfully performed on a limited basis in the Entergy fleet after engineering analysis.Once the motor has been stalled, it is desirable to not repeat this process more than necessary. The backseating process intentionally takes the motor out of the normal control circuit and causes motor heating that with repeated events would be severely damaging to the motor and motor cabling. In addition, if the backseating was previously successful in stopping a Drywell packing leak, the potential exists to re-initiate the packing leak with each backseating evolution. Finally, quarterly testing should not be pursued on backseated valves due to the significant evolution that would be required every quarter. If quarterly testing did continue, the valve would first be stroke timed.Then, an LCO would have to be entered and the backseating test equipment would have to be set up. In order to backseat the valve, maintenance must access the motor control center and auxiliary power. A variable transformer is connected to an alternate power supply, then motor cables are lifted at the motor control center. Outputs of the leads of the variable transformer are connected to the motor power cables, and this current is monitored with meters and the diagnostic test system. The circuit is closed and the Revision 0 Pg 4 Page 143 Cooper Nuclear Station Ffith Interval Inserviee Testing Program for Pumps and Valves Cold Shutdown Justification CSJ-1O (Augmented)(Continued) valve is first stroked from thc normally open position to partially closed. Opening the valve from the mid range position allows for the technicians to start motor movement at full voltage and then dial the voltage down to a predetermined reduced voltage level that controls the stall forces that impact the valve backseat. Therefore, it would be a significant evolution every quarter if the valve were to be stroke timed and re-backseated at this frequency. In conclusion, any one of the three valves listed may be backseated at some time in the future as a mitigating strategy associated with reducing Drywell leakage. Once one or more of these valves are backseated, the stroke time frequency will be changed from quarterly to a cold shutdown frequency. Quarterly testing is impracticable due to the increased potential for causing component damage, the potential of re-initiating a Drywell leak, and the significant evolution that would need to be undertaken every quarter to stroke time the valve and re-backseat it. There would be little to gain from this process, especially since MOVs are characteristically very steady with their stroke times over long periods of time.Therefore, the stroke time testing for these valves would require a re-classification from testing quarterly to a cold shutdown frequency for an interim period until the presence of a packing leak can be ruled out or the packing repaired.Alternative Test If backseated, valve stroke time testing for IHPCI-MO V-MO 15, MS-MO V-MO74, or RCIC -MO V-MO 15 will be performed during cold shutdowns. Revision 0 Pg 4 Page 144 Cooper Nuclear Station F~fih Interval Inservice Testing Program for Pumps and Valves REFUELING OUTAGE JUSTIFICATIONSIDE REFUELING OUTAGE JUSTIFICATION Th~DEX Refueling Outage Description Justification No.ROJ-O1 IINBI-CV-49B/50B/5 1B/52BCV and NBI-SOV-SSV73 8/739 Exercising ROJ-02 NBI-CV-55/56CV Exercising ROJ-03 RF-CV-1 3/14/15/1 6CV Exercising ROJ-04 [j RWCU-CV-15CV Exercising ROJ-05 IRWCU-MO-1 5/18 Exercising ROJ-06 CRD-CV-CV1 15 (Typical of 137) Exercising ROJ-07 IA-C V-17/1 8/1 9/20/21/22/36/37CV Exercising (Augmented) ROJ-08 J] IA-CV-28/29/30/3 1/32/33/34/35CV Exercising (Augmented) ROJ-09 ][ HIPCI-CV-29CV and RCIC-CV-26CV Exercising ROJ-10 ]~SW-MOV-MO89A/B Exercising ROJ-1 1 IICore Spray and RHRE Injection Check Valve Exercising,.. ROJ- 12 ]1 CRD-CV-25/26CV Exercising (Augmented) ROJ-13 ]j IA-CV-57/58/59/60CV Exercising (augmented) Revision 0 Pg 4 Page 145 Cooper Nuclear Station Fifthi Interval Inservice Testing Program for Pumps and Valves Refueling Outage Justification ROJ-O1 Valve Number System Class Category.NBI-CV-49BCV NBI 3 C NBI-CV-J0BCV NBI 3 C NBI-CV-51BCV NBI 3 C NBI-CV-52BCV NBI 3 C NBI-SOV-SSV738 NBI 3 B NBI-SOV-SSV739 NBI 3 B Function The reference leg injection check valves and solenoid operated valves have an active safety function in the open position to inject Core Spray water to the reactor vessel level instrumentation lines in case the reference leg water has flashed or boiled off due to accident conditions in the drywell.Justification This system provides the capability for the Core Spray System to supply a backfill of water for maintaining inventory of the Nuclear Boiler Instrumentation System cold reference legs (condensing chambers 3A and 3B) during accident conditions in the drywell where the reference leg inventory could be compromised. Exercising these valves to the open position, full or partial, would require manually isolating and venting the Cold Reference Leg Backfill System. This is not practicable during power operation or cold shutdown, other than refueling, due to the possible introduction of air into the system.This could cause a spurious reactor vessel level indication which could cause a reactor trip during power operation. During cold shutdown spurious level indications could interrupt the operation of systems required for decay heat removal, thereby placing the reactor in an unsafe condition. During refueling outages, sufficient time exists for decay heat to be reduced to a level which minimizes the impact of momentary interruption in the operation of systems required for decay heat removal such that testing can be performed. Alternative Test Exercising these check valves to the full open and closed positions, and full exercising with stroke timing to the open position of the solenoid operated valves, shall be performed during refueling outages when the cold reference leg backfill system may be isolated and vented to allow testing.Revision 0 Pg 4 Page 146 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Refueling Outage Justification ROJ-02 Valve Number System Class g r NBI-CV-55CV NBI 3 C NBI-CV-56CV NBI 3 C Function These Cold Reference Leg Continuous Backfill System check valves have an active safety function in the closed position to isolate the Class 3 instrumentation piping from the Seismic uIS non-class CRD piping.Justification This system provides for a continuous flow of water from the CRD drive water pumps to prevent noncondensible gases from building up in the Nuclear Boiler Instrumentation System cold reference legs (condensing chambers 3A and 3B). Exercising these valves to the closed position would require manually isolating and venting of the Cold Reference Leg Continuous Backfill System upstream of the check valves. This is not practicable during power operation or cold shutdown, other than refueling, due to the possibility of causing a spurious reactor vessel level indication from entrained air in the system.False level indications resulting from entrained air in the system may either cause a reactor trip during power operation or interrupt the operation of systems required during cold shutdown for decay heat removal, thereby placing the reactor in an unsafe condition. During refueling outages, sufficient time exists for decay heat to be reduced to a level which minimizes the impact of momentary interruption in the operation of systems required for decay heat removal such that testing can be performed. Alternative Test Exercising these check valves to the closed position shall be performed during refueling outages when the cold reference leg backfill system may be isolated and vented to allow testing. Exercise testing shall be accomplished by performing a seat leakage test. The open test will also be credited during each reactor refueling as is allowed per ISTC-3522(a). Revision 0 Pg 4 Page 147 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Refueling Outage Justification ROJ-03 Valve Number System Class Category.R+/--CV-13CV RF 1 A/C RF-CV-14CV RF 1 A/C RF-CV-15CV RF 1 A/C RF-CV- 16CV RF 1 A/C Function These valves are the main feedwater check valves which open to allow normal feedwater flow.Additionally, RF-CV-14CV and RF-CV-l6CV open to allow HPCI and RCIC flow to the vessel, respectively. All four of these valves must be capable of closure to provide containment isolation. Additionally, RF-C V-i13CV and RE-C V-l5CV must be capable of closure to prevent diversion of HPCI and RCIC flow, respectively. Justification These valves are normally open and must remain open during reactor operations to ensure adequate feedwater flow. Exercising these valves closed during plant operation could cause a transient in reactor water level resulting in a reactor scram. The observation of specified leakage during local leak-rate testing provides the only means for verification to the closed position.With the installation of highly accurate ultrasonic flow meters on the feedwater injection lines in RE24, per CED 6023681, individual flow rates for each feedwater injection line may now be determined at any time during normal operation. Therefore, the safety-related required flows through RF-CV-14CV and RE-C V-l6CV (in addition to the non-safety related flows through RE-C V-13CV and RE-C V-15C V), may be verified during normal operations. If the flow meters were to become unavailable, the open test for all the valves may be satisfied via an IST valve disassembly and examination (which requires manual exercising of the valve and visual inspection), or a torque test, both of which would be completed during a refueling outage due to their location and the allowance provided by ISTC-3522(a). In addition, the non-safety open test for RE-C V-i13CV and RE-CV-1 5CV may be satisfied through Operation logs, in which verification of being at full power would verify that adequate feedwater flows exist in supporting this power.Alternative Test These valves will be exercised to the closed position during the Type C leak rate test performed each refueling outage in accordance with the requirements of ASME OM Code ISTC-3 522 for category C check valves and ISTC-3620 for Containment Isolation valves. The open direction test will be credited at least once each reactor refueling as is allowed per ISTC-3522(a). Normally, the open flow test will be satisfied through the recording of acceptable flows through each injection line during normal operation per ISTC-3550. If the flow meters become unavailable, disassembly and examinations (all), torque testing (all), or verification of full power (1 3CV, 1 5CV only) may be utilized.Revision 0 Page 148 Revision 0 Page 148 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves Refueling Outage Justification ROJ-04 Valve Number System Class Category.RWCU-CV- 15 CV RW 1 A/C Function This check valve is normally open to allow Reactor Water Cleanup (RWCU) return and must close to provide containment isolation. Justification This valve cannot be verified as being closed upon reversal or stopping of flow without opening and venting the line upstream of the check valve. Opening or venting the RWCU line during operations could cause a leak of high pressure reactor coolant and potentially lead to the release of radioactive material.Stopping RWCU flow during normal operations or cold shutdown for an extended period would lead to a degradation of reactor water purity. This would add to the radioactive contamination in the reactor coolant system and could lead to additional exposure of site personnel. It is essential that RWCU remain in operation as much as possible and RWCU-C V-i15CV be exercised to the closed position only during refueling outages.Alternative Test The open capability of RWCU-C V-i15CV is verified during normal operations. The closure capability will be verified during refueling outages by performing a Type C local leak rate test per the requirements of ASME OM Code JSTC-3 522 and the CNS Primary Containment Leakage Rate Testing Program. The open direction test will be credited at least once each reactor refueling as is allowed per ISTC-3522(a). Revision 0 Pg 4 Page 149 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Refueling Outage Justification ROJ-05 Valve Number System Class Category.RWCU-MO- 15 RW 1 A RWCU-MO -18 RW 1 A Function These normally open Reactor Water Cleanup (RWCU) valves must close for containment isolation. The open function is not required for safe shutdown of the plant and is considered an operational function only.Justification These normally open containment isolation valves have a function during normal operation to provide a path for reactor coolant to and from the reactor water cleanup system to maintain high reactor water purity. In order to exercise these valves during normal plant operation, the RWCU system would need to be shutdown. Shutdown of the RWCU system induces chemistry transients in the reactor and should be minimized in order to maintain consistent reactor water chemistry. Additionally, shutdown of the RWCU system can lead to hydraulic transients and crud bursts that will result in increases in radiation levels and higher worker dose.Failure of one of these RWCU valves in the closed position would result in the complete loss of the RWCU system. This, in turn, could result in a plant shutdown to repair the valve due to the loss of reactor coolant chemistry parameters. Additionally, shutting down the RWCU system every quarter cycles the equipment without a compensating increase in safety. Shutdown of the RWCU system in a forced outage will also inhibit the ability to cleanup the vessel and result in an increase in radiation levels and personnel dose. Additionally, RWCU-MO-1 5 is located inside the primary contaimnment and is inaccessible during power operation due to high radiation levels and the inerted atmosphere. It is also impractical to de-inert containment for repair of this valve if it fails during cold shutdown testing.Refueling outages have sufficient duration to allow the RWCU System to adequately cleanup the primary coolant prior to being shutdown for testing. Additionally, the refueling outage schedules include periods in which RWCU must be shutdown while maintenance is performed on its support systems. If a tested RWCU valve does fail in the closed position during a refueling outage, adequate time is available to correct the condition without impacting unit availability and without adverse ALARA effects.Alternative Test The closure capability of RWCU-MO-15 and RWCU-MO-l18 will be verified during refueling outages by performing a stroke time test in the closed direction. Revision 0 Pg 5 Page 150 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Refueling Outage Justification ROJ-06 Valve Number System Class Category.CRD-CV-CV 115 CRD 2 C (Typical of 137)Function These valves have a safety function to close to prevent the loss of water pressure in the event that charging supply pressure is lost which prevents bypassing scram water (from the accumulator) to the charging water header (if depressurized). This valve has a function to open to allow charging water to pass from the control rod drive pumps to the hydraulic control units. Flow to the accumulators is required only during scram reset or system startup. This valve has no safety function to open.Justification Exercising these valves requires the depressurization of the charging water header. The header is depressurized by either stopping the CRD pumps or by valving out and depressurizing the charging water header. Stopping the CRD pumps could result in seal damage to the control rod drive mechanisms (CRDM) from a loss of seal cooling water. Additionally, stopping the pumps would interrupt seal cooling water flow to the reactor recirculation pumps resulting in shaft seal damage. This is impracticable during normal plant operation since valving out and depressurizing the charging water header would render the CRD accumulators inoperable and stopping the CRD pumps could cause pressure variations in the CRLD System during the test evolution. Exercising these valves during cold shutdown is not possible due to the interruption of shaft seal cooling water flow as previously discussed. If the recirculation pump became idle stopping the CRD pumps or manually isolating the charging water header for reverse exercise testing could delay plant startup due to the necessity of depressurizing upstream of each individual valve (137 each) in order to accomplish an adequate test. This additional test activity during cold shutdown represents an unusual burden without a compensating increase in the level of quality and safety.Alternative Test These valves will be tested during each reactor refueling outage. Proper closure shall be verified by isolating each of the CRD scram accumulators and venting pressure on the upstream side of the check valve. Accumulator pressure decay would be observed should the respective valve fail to close properly.The open test will also be credited during each reactor refueling as is allowed per ISTC-3522(a). Revision 0 Pg 5 Page 151 Cooper Nuclear Station F/ifh Interval Inservice Testing Program for Pumps and Valves Refueling Outage Justification ROJ-07 (Augmented) Valve Number IA-CV- 1 7CV IA-C V-i18CV IA-C V-i 19CV IA-C V-20C V IA-C V-21 CV IA-C V-22C V IA-C V-36C V IA-C V-37C V System IA IA IA IA IA IA IA IA Class A A A A A A A A Cateaorv A/C A/C A/C A/C A/C A/C A/C A/C Function These valves are the Instrument Air/Nitrogen supply inlet check valves for Main Steam Relief Valve (SRV) accumulators. These check valves must be capable of closure to maintain accumulator integrity in the event of a loss of normal actuating air supply.The check valves must open to allow flow to their respective accumulators. Justification These valves are located inside the drywell and are inaccessible during normal operations or cold shutdowns. They cannot be exercised during each cold shutdown because the drywell is not routinely de-inerted each cold shutdown. Valve exercising during cold shutdown when the drywell is de-inerted could delay plant restart due to the necessity of using portable test equipment inside the drywell. The additional test activity during cold shutdown represents an unusual burden without a compensating increase in the level of quality and safety. Testing these valves during refueling outages is consistent with NUR.EG 1482, revision 2, section 3.1.1.3.These valves are ASME non-code class valves that are not within the scope of the IST Program.Alternative Test An extended time/pressure decay procedure will be used to verify each valve's closure. This will be done by venting the upstream side of the check valve and monitoring accumulator pressure to ensure each check valve functions properly. The above valves will be tested each refueling outage to verify valve closure. The open test will also be credited during each reactor refueling as is allowed per ISTC-3522(a). Revision 0 Page 152 Revision 0 Page 152 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Refueling Outage Justification ROJ-08 (Augmented) Valve Number System Class Ctgr IA-C V-28CV IA A A/C IA-CV-29CV IA A A/C IA-C V-30CV IA A A/C IA-CV-31CV IA A A/C IA-C V-32CV IA A A/C IA-CV-33CV IA A A/C IA-C V-34CV IA A A/C IA-C V-35CV IA A A/C Function These check valves must close to isolate individual Main Steam Isolation Valve accumulators for emergency gas supply.The check valves must open to allow flow to their respective accumulators. Justification These check valves do not have position indication devices. The only practicable method to verify valve closure is a pressure decay test. The valves are located in the steam tunnel and the drywell. They are inaccessible during operation and normal cold shutdowns. The complexity of the pressure decay test could delay plant startup after a cold shutdown when the drywell is de-inerted. Since these emergency air supply accumulators are a backup to the normal pneumatic supply, performing the test at refueling outages is adequate to assess valve operational readiness. Testing these valves during refueling outages is consistent with NUREG 1482, revision 2, section 3.1.1.3.These valves are ASME non-code class valves and are not within the scope of the IST Program.Alternative Test A pressure decay test will be performed each refueling outage to verify valve closure. The open test will also be credited during each reactor refueling as is allowed per ISTC-3522(a). Revision 0 Page 153 Revision 0 Page 153 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Refueling Outage Justification ROJ-09 Valve Number System Class Catgory, HIPCI-CV-29CV HPCI 1 A/C -RCIC-CV-26CV RCIC 1 A/C Function HTPCI-CV-29CV -Opens to provide a flow path from the HIPCI pump to the reactor vessel via the feedwater system; closes for primary containment isolation. RCIC-CV-26CV -Opens to provide a flow path from the RCIC pump to the reactor vessel via the feedwater system; closes for primary containment isolation. Justification These valves are normally closed to isolate the reactor coolant system and the HPCI and RCIC systems.Exercising these check valves to the open position during normal plant operation would require HPCI or RCIC injection to the reactor vessel. This would result in a perturbation of normal feedwater flow and unnecessary thermal cycling of the feedwater nozzles. It would also cause severe power fluctuations due to the relatively cold water from the Emergency Condensate Storage Tanks. Furthermore, these valves are located in the Steam Tunnel. During power operations, this area experiences temperatures of approximately 130 -1 40°F, and high radioactivity. In general, plant personnel are prohibited from entering this area during power operation due to these conditions. Testing during Cold Shutdowns is impractical. This testing would be performed from the Steam Tunnel and would require support from Radiological Protection, Operations, Mechanical Maintenance, and possibly Engineering. During a typical forced outage, entry into the Steam Tunnel would not be made.Therefore, for the sole purpose of performing this test, multiple departments would need to support the evolution. Radiological support would be required to obtain the necessary survey information in the steam tunnels and/or be in attendance with the personnel performing the testing. To exercise these testable check valves, the system is required to be removed from service such that pressure is equalized across the valves prior to exercise testing. In order to accomplish this, this test requires operations personnel to enter the steam tunnel in order to remove pipe caps on vent and test connections on either side of the check valves, connect hoses, and open these valves to equalize pressure.In comparison, set up of this type is required with the performance of an Appendix J leak test, which is specifically allowed to be extended to a refueling outage per section 4.1.6 of NUREG 1482, revision 2.Then, Mechanical Maintenance personnel would perform the required mechanical exercising of the check valves, using appropriate torque wrenches. Following the testing, Operations would be required to restore the system. If any questions arose concerning the testing, which could occur, as described in section 4.1.7 of NUJREG 1482, revision 2, Engineering Personnel could be required to get involved to analyze the data, possibly prolonging the cold shutdown, unnecessarily. A refueling outage would be of a significant duration that this would not be a concern. In addition, personnel safety concerns are heightened due to the lack of lighting in the steam tunnel during a forced outage. In conclusion, because of the required test setup and complexities described, this testing is impracticable to perform during cold shutdowns and may delay unit startup.Revision 0 Page 154 Revision 0 Page 154 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Alternative Test These valves will be mechanically exercised, verifying open and closure capability, during refueling outages when the IIPCI and RCIC systems are not required to be inservice. Revision 0 Pg 5 Page 155 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Refueling Outage Justification ROJ-10 Valve Number System Class Category.SW-MOV-MO89A SW 3 B SW-MO V-MO89B SW 3 B Function These are the Loop A and Loop B outlet isolation for the Service Water booster pump cooling water to the RHiR heat exchangers. These normally closed valves have an active safety function in the throttled position to provide a flow path for cooling water flow through the RI-R heat exchangers during transient and accident conditions. Justification These valves are exercised during quarterly Service Water Booster Pump flow testing to a throttled position required to satisfy Technical Specification flow requirements. Valve stroke timing to the fully opened position is impracticable to perform at power due to the potential to cause RHR Service Water Booster Pump run out. Each Service Water Booster Pump is electrically interlocked with its respective RHIR heat exchanger outlet valve (SW-MOV-MO89A/B). When the pump control switch is taken to START, its respective RHR heat exchanger outlet valve receives an open signal. When the valve reaches a position that ensures pump minimum flow requirements can be met, the pump receives a start signal.The RHIR heat exchanger outlet valve is throttled to obtain the desired flow. Each RHIR heat exchanger outlet valve is electrically interlocked to close when both associated RIHR Service Water Booster pumps are shutdown. These valves are also utilized to maintain RHR-SW shell pressure higher than the RHR system pressure to prevent the potential release of radioactivity into the river.It would also be impracticable to perform full stroke testing at a cold shutdown frequency since Service Water and Service Water Booster pumps are essential for providing cooling water to the RHR heat exchangers during this time period. Defeating the Service Water Booster pump interlock and windmilling the Service Water Booster pumps would not be a desirable activity. It would be more appropriate to perform this testing once the decay heat has lowered during a refueling outage.For the reasons and since this valve is verified to be capable of performing its safety function each quarter, it would be impractical to defeat the interlocks associated with this valve on a quarterly or cold shutdown frequency to obtain a full open stroke time test. It would be practical, however, on a once per refueling outage basis, to full stroke open these valves for trending either through the use of control room hand switches or MOV diagnostic equipment once decay heat has lowered. If hand switches are utilized, flow through the subsystem being tested should be manually isolated and interlock defeated to allow the full stroking of the valve.Alternative Test These valves will be exercised to their safety-related throttled position quarterly, but stroke times will not be measured. During refueling outages, these valves will be stroke time tested to the full open position through the use of the SW-MO V-MO89A/B hand switches with flow isolated and interlocks defeated. If desired, the stroke time may be obtained during MOV diagnostic testing.Revision 0 Pg 5 Page 156 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Refueling Outage Justification ROJ-11 Valve Number System Class. Category, CS-CV-l18CV CS 1 A/C CS-CV-19CV Cs 1 A/C RHR-CV-26CV RIHR 1 A/C RHR-CV-27CV RHR 1 A/C Function These valves open for Core Spray or LPCI injection and close for primary containment isolation. Justification These valves are normally closed for primary containment isolation. They are also closed to isolate the related low pressure systems from the Reactor Recirculation system and the reactor vessel. Opening these valves during power operation is not possible due to the downstream side being exposed to reactor pressure. A drywell entry during cold shutdown would be necessary to facilitate testing, thereby requiring de-inerting, which could potentially delay restart. Also, cold shutdown conditions in the Drywell have previously resulted in high radiation fields near the RHR lines and other areas of the Drywell. Finally, personnel safety concerns due to the Core Spray lines being located in the upper drywell elevations with little to no lighting may also be an issue. The reasons presented are acceptable reasons for deferral to a refueling outage basis per NUREG 1482, rev. 2, sections 2.4.5 and 3.1.1.3.Alternative Test These valves will be mechanically exercised, verifying open and closure capability during refueling outages in accordance with ISTC-3522 and ISTC-5221. Revision 0 Pg 5 Page 15 7 Cooper Nuclear Station Ffifhi Interval Inservice Testing Pro gram for Pumps and Valves Refueling Outage Justification ROJ-12 (Augmented) Valve Number System Class Category.CRD-CV-25CV CRD A A/C CRD-CV-26CV CRD A A/C Function These valves close to prevent possible CRD bypass leakage from exiting secondary containment. They open to supply drive water and charging water to the Hydraulic Control Units (HCUs) and seal water to the Reactor Recirculation pumps.Justification It is impracticable to perform a closure test on these valves during power operations or cold shutdowns. These valves are located in the line from the CRD pumps supplying drive water and charging water to the control rod's HCUs.During power operations, these valves are open since drive water is constantly supplied to the HCUs.Closure testing would require the CRD pumps to be secured and the portion of the system containing these valves to be isolated. Isolating the valves or securing the CRD pumps will terminate the constant drive water supply to the HCUs, causing all control rods to be inoperable. Also, without a continuous charging water supply, HCU accumulators would eventually depressurize and administrative controls require a scram initiation upon depressurization of two accumulators, which is a highly undesirable situation. The interruption of the drive water flow would also interrupt CRLD seal water cooling to the Reactor Recirculation pumps. The stopping of flow would impose a severe thermal transient on the RR pump seals, which could possibly lead to premature seal failure.The method of testing these valves for closure would be through a local leak rate test, which involves the draining of the system, establishment of test boundaries, equipment setup, etc., which is more suitable during a refueling outage. This test method, alone, could delay a plant startup from a cold shutdown.In conclusion, testing these valves during refueling outages is consistent with NUJREG 1482, rev. 2, sections 2.4.5, and 3.1.1.4.Alternative Test The closure capability will be verified during refueling outages by performing a local leak rate test. The open direction test will be credited at least once each reactor refueling as is allowed per ISTC-3522(a). Revision 0 Pg 5 Page 158 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Cold Shutdown Justification ROJ-13 (Augmented) Valve Number System Class Category.IA-C V-57CV IA A C IA-C V-58CV IA A C IA-C V-5 9CV IA A C IA-CV -6OCV IA A C Function These valves are required to close to maintain pressure in the associated air operated valve's accumulator in the event of a loss of the Instrument Air (IA) supply. These IA supplies are for various safety related I-&V system AOV's.Justification It is impracticable to perform a closure test on these valves during power operations or cold shutdowns. IA-CV-57CV, -58CV, -59CV, AND 6OCV are in the IA supply lines to the H&V supply and exhaust isolation dampers for the RRMG Set IA and lB. Isolation and testing of these valves potentially affects the operation of the RRMG Sets. This would be undesirable during power operations and during cold shutdowns when the recirculation pumps are required to be operational. Also, the closure test for these check valves requires that the IA supply piping upstream of the check valves to the associated accumulators be isolated and depressurized. The accumulator pressure is then monitored for one hour via installed test equipment to verify that the check valve will hold.Depressurization of the accumulator below 70 psig with the respective H&V valve open will place the component in an inoperable condition. Due to the complexities of this test and the potential to delay a cold shutdown, it would be more appropriate to perform this testing during refueling outages.These valves are ASME non-code class valves and are not within the scope of the IST Program.Alternative Test A reverse flow leakage test of these check valves will be performed during refueling outages by performing a pressure decay test. The open test will also be credited during refueling outages shutdown as is allowed per ISTC -3522 (a).Revision 0 Page 1590 Page 159 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves ATTACHMENT 9 TECHNICAL POSITIONS TECIHNICAL POSITION INDEX Technical Decito Position No. Decito TP-O01 JfBi-directional Testing of Check Valves TP-02 ]I Sample Disassembly of HPCI Vacuum Breaker Check Valves TP-03 ][Passive Valves without Test Requirements TP-04 ]IFail Safe Testing of Valves TP-05 ]~Classification of Skid-Mounted Components TP-06 ]jCheck Valves in Regular Use TP-07 ]ICategorization of IST Pumps (Group A or B)TP-08 ]jVacuum Breaker Testing Revision 0 Page 160 Revision 0 Page 160 Cooper Nuclear Station Fifthi Interval Inservice Testing Pro gram for Pumps and Valves Technical Position TP-O1 (Page 1 of 3)Bi-directional Testing of Check Valves with Non-Safety Positions Purpose The purpose of this Technical Position is to establish the station position for the verification of the non-safety direction exercise testing of check valves by normal plant operations. Applicability. This Technical Position is applicable to those valves which are included in the Inservice Testing Program that are required to be exercise tested in their non-safety related direction of flow. This position applies to those check valves required to be tested in accordance with Subsection ISTC (ASME OM Code 2004 Edition through 2006 Addenda) and Appendix II. This Technical Position does not apply to testing of the safety function (direction) of check valves included in the Inservice Testing Program.Backg~round The ASME OM Code 2004 Edition through 2006 Addenda section ISTC-3550, "Valves in Regular Use", states: "Valves that operate in the course of plant operation at a frequency that would satisfy the exercising requirements of this Subsection need not be additionally exercised, provided that the observations otherwise required for testing are made and analyzed during such operation and recorded in the plant record at intervals no greater than specified in ISTC-3510." Section ISTC-35 10 requires that check valves shall be exercised nominally every 3 months with exceptions (for extended periods) referenced. Section ISTC-522 1 (a)(2) states: "Check valves that have a safety function in only the open direction 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 perform its intended function(s) (see ISTC-1 100), and verify closure." Section ISTC-5221 (a)(3) states: "Check valves that have a safety function in only the close direction shall be exercised by initiating flow and observing that the obturator has traveled [to] at least the partially open position, 3 and verify that on cessation or reversal of flow, the obturator has traveled to the seat."" 3 The partially open position should correspond to the normal or expected system flow." Normal and/or expected system flow may vary with plant configuration and alignment. Cooper Nuclear Station Operations staff is trained in recognizing normal plant conditions. For check Revision 0 Page 161 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves Technical Position TP-O1.(Page 2 of 3)valves that have a non-safety function in the open position, Operator judgment has been deemed acceptable in determining whether or not the normal or expected flow rates for plant operation has been obtained. For check valves that have a non-safety related function in the closed position, Operator judgment is also deemed acceptable in determining whether or not flow has occurred at a normal or expected flow rate, in order to cause obturator travel.Position Typically, Cooper Nuclear Station will verify the non-safety position of check valves included in the Jnservice Testing Program using a periodic activity within the plant surveillance or preventative maintenance program. In lieu of a dedicated surveillance or preventative maintcnance activity to perform the non-safety direction testing, the following alternate verifications may be performed as follows: 1. An appropriate means shall be determined which establishes the method for determining the open/closed non-safety function of the check valve during normal operations. The position determination may be by direct indicator, or by other positive means such as changes in system pressure, flow rate, level, temperature, seat leakage, etc. This determination shall be documented in the respective Condition Monitoring Plan for the specific check valve group. For check valves included in the Inservice Testing Program and not included in the Condition Monitoring Plan, this determination shall be documented in the 1ST Bases Document for the specific check valve group.2. Observation and analysis of plant processes that a check valve is satisfying its non-safety direction function may be used. For an example, consider a check valve that has a safety function only in the closed direction and normally provides a flow path to maintain plant operations. If this check valve does not open to pass flow when required, an alarm or indication would identify a problem to the operator. The operator would respond by takcing the appropriate actions. A Condition Report would then be generated for the abnormal plant condition which would identify the check valve failure.3. Observation and analysis of plant logs (i.e. Operations or Chemistry) and other records may be an acceptable method for verifying a check valve's non-safety direction function verification during normal plant operations. The open/closed non-safety function shall be recorded at a frequency required by ISTC-35 10, nominally every 3 months, (with exceptions as allowed), in plant records such as Cooper Nuclear Station Operating or Chemistry Logs, Electronic Rounds, chart recorders, automated data loggers, etc. The safety function direction testing requires a periodic Quality Record typically documented within the surveillance and/or preventative maintenance program.Records as indicated above in 1 through 3 are satisfactory for the non-safety direction testing. A condition report shall be generated for any issues regarding check valve operability. Justification This Technical Position establishes the acceptability of the methods used in determining the ability of a check valve to satisfy its non-safety function. Typically, Cooper Nuclear Station will verify the non-safety position of check valves included in the Inservice Testing Program using a Ravision 0 Page 162 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Technical Position TP-O1 (Page 3 of 3)periodic activity within the plant surveillance or preventative maintenance program.Alternatively, through normal plant system operation and operator actions, a valve's non-safety function is verified through either observation or analysis of plant records and logs. Additionally, the recording of parameters which demonstrate valve position is satisfied at a frequency in accordance with ISTC-35 10. These actions collectively demonstrate the non-safety position of Jnservice Testing Program check valves in regular use as required by ISTC-3550. Revision 0 Pg 6 Page 163 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Technical Position TP-02 (Page 1 of 2)Sample Disassembly of HIPCI Vacuum Breaker Check Valves The purpose of this Technical Position is to establish the station position for testing the HPCI Vacuum Breaker Check Valves.Applicabilit, This Technical Position is applicable to the IIPCI Vacuum Breaker Check Valves, HPCI-CV-24CV, 25CV, 26CV, and 27CV.Position These valves are normally closed checkc valves with two in series cross connected to two in parallel (H pattern). The four HPCI valves are swing check valves. The valves are located in the suppression pool fre~e space. In the closed position, they prevent steam from the exhaust line from entering the free space of the suppression chamber. Either two inboard valves or two outboard valves must be closed to perform this function. Two valves in series provide added assurance that steam will not enter the suppression chamber.The valves open to prevent siphoning suppression pooi water into the exhaust line due to steam condensing when the associated HiPCI system is isolated. Each pair of valves is cross connected to the parallel pair of valves (H pattern) so that a single failure will not prevent the vacuum relief function. These valves are not required to be leak tight and are not equipped with position indication or pressure sensing devices. During power operation, the suppression chamber is inerted and inaccessible. Due to the location and configuration of these valves (located on the open ended turbine exhaust lines) a typical closure or open test cannot be performed. These vacuum breaker check valves are not capacity certified. Therefore, they may be tested in accordance with ISTC-5220 (as a check valve). In accordance with the requirements of ISTC-5221 (c), a sample disassembly examination program is being utilized to satisfy 1ST testing for the HPCI vacuum breaker check valves.These valves are grouped together because they perform the same function, have the same manufacturer, design, service conditions and have the same size, materials of construction and orientation. A different valve from this group will be disassembled, inspected and manually exercised during each refueling outage until the entire group has been tested. If the disassembled valve selected is not capable of being full-stroke exercised or there is binding or failure of valve internals, the remaining valves in the group will be disassembled, inspected, and manually exercised during the same outage.Procedural requirements ensure the valve is re-installed correctly, so no open/closed testing is required following reassembly. Revision 0 Page 164 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Technical Position TP-D2 (Page 1 of 2)Justification All requirements of ISTC-5221(c) and ISTC-9200(c) are met.Revision 0 Page 1650 Page 165 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Technical Position TP-03 (Page 1 of 1)Passive Valves Without Test Requirements Purpose The purpose of this Technical Position is to establish the station position for valves which perform a passive safety function. However, no testing is required in accordance with LSTC.Applicability This Technical Position is applicable to valves that perform a passive function in accordance with ISTA-2000 and do not have inservice testing requirements per Table ISTA-3500-1. This position is typical of Category B, passive valves that do not have position indication. 'An example is a manual valve which must remain in its normal position during an accident, to perform its intended function.' Typically, manual valves that perform a safety function are locked in their safety position and administratively controlled by Cooper Nuclear Station procedures. These valves would be considered passive. If they do not have remote position indicating systems and categorized as B, they would not be subjected to any test requirements in accordance with Table ISTC-3500-l. Position The Cooper Nuclear Station Inservice Testing Program, Valve Tables -Attachment 11, will not list valves that meet the following criteria.* The valve is categorized B (seat leakage in the closed position is inconsequential for fulfillment of the valves' required function(s)) in accordance with ISTC- 1300.* The valve is considered passive (valve maintains obturator position and is not required to change obturator position to accomplish the required function(s)) in accordance with ISTA-2000.

  • The valve does not have a remote position indicating system which detects and indicates valve position.Justification Valves that meet this position will not be listed in the Cooper Nuclear Station Inservice Testing Program, Valve Tables -Attachment 11, however, the basis for categorization and consideration of active/passive functions should be documented in the 1ST Program Basis Document.Revision 0 Pg 6 Page 166 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Technical Position TP-04 (Page 1 of 1)Fail Safe Testing of Valves Purpose The purpose of this Technical Position is to establish the station position for fail safe testing of valves in conjunction with stroke time exercising or position indication testing.Applicabilit.

This Technical Position is applicable to valves with fail-safe actuators required to be tested in accordance with ISTC-3560.

Background

The ASME OM Code 2004 through 2006 Addenda section ISTC-3560 requires;"Valves with fail-safe actuators shall be tested by observing the operation of the actuator upon loss of valve actuating power in accordance with the exercising frequency of ISTC-35 10." Section ISTC-3510 states;"Active Category A, Category B, and Category C check valves shall be exercised nominally every 3 months..." Position In cases where the valve operator moves the valve to the open or closed position following de-energizing the operator electrically, by venting air, or both, the resultant valve exercise will satisfy the fail-safe test requirements and an additional test specific for fail safe testing will not be performed. Cooper Nuclear Station will also use remote position indication as applicable to verify proper fail-safe operation, provided that the indication system for the valve is periodically verified in accordance with ISTC-3700. Justification Fail-Safe Testing tests the ability of the fail-safe mechanism of the valves to go to its fail safe condition. Whether or not the actuation of this fail-safe mechanism is due to Operator Action of failure of either the valve's air or electric power source, the resultant action of the valve will be the same. Therefore, the verification of a valve's fail safe ability can be taken credit for with the performance of either a stroke time exercising or position indication test.'Revision 0 Pg 6 Page 167 Cooper Nuclear Station Ffifh Interval Inserviee Testing Program for Pumps and Valves Technical Position TP-05 (Page 1 of 2)Classification of Skid-Mounted Components Purpose The purpose of this technical position is to clarify requirements for classification of various skid-mounted components, and to clarify the testing requirements of these components.

Background

The ASME Code allows classification of some components as skid-mounted when their satisfactory operation is demonstrated by the satisfactory performance of the associated major components. Testing of the major component is sufficient to satisfy Inservice Testing requirements for skid-mounted components. In section 3.4 of NUREG 1482 Rev l, the NRC supports the designation of components as skid mounted: "The staff has determined that testing the major component is an acceptable means to verify the operational readiness of the skid-mounted components and component subassemblies if the licensee discusses this approach in the IST program document. Licensees should consider and document the specific measurements and attributes of major component testing which relate to the assessment of skid-mounted component condition. In addition, various continuous and periodic observations of the major components (such as System Monitoring Walkdowns or Operator Logs) may also support assurance of skid-mounted component readinesss. This is acceptable for both Code class components and non-Code class components that are tested and tracked by the IST Program." The 2004 Edition through the 2006 Addenda of the ASME Cm Code, Subsection ISTA-2000 provides the following definition. Skid mounted pumps and valves -pumps and valves integral to or that support operation of major components, even though these pumps and valves may not be located directly on the skid. In general, these pumps and valves are supplied by the manufacturer of the major component. Examples include: (a) diesel fuel oil pumps and valves;(b) steam admission and trip throttle valves for high-pressure coolant injection turbine-driven pumps;(c) steam admission and trip throttle valves for auxiliary feedwater turbine driven pumps;(d) solenoid-operated valves provided to control an air-operated valve.Revision 0 Pg 6 Page 16 8 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Technical Position TP-05 (Page 2 of 2)Additionally the Subsections pertaining to pumps (ISTB) and valves (ISTC) include exclusions/exemptions for skid-mounted components; ISTB-1 200(c) Exclusions Skid-mounted pumps that are tested as part of the major component and are justified by the Owner to be adequately tested.ISTC-1 200 Exemptions Skid-mounted valves are excluded from this Subsection, provided they are tested as part of the major component and are justified by the Owner to be adequately tested.Position The 2004/2006a ASME OM Code definition of skid-mounted will be used for classification of components in the Cooper Nuclear Station Inservice Testing Program. In addition, for a component to be considered skid-mounted:

  • The major component associated with the skid-mounted component must be surveillance tested at a frequency sufficient to meet ASME Code test frequency for the skid mounted component.
  • Satisfactory operation of the skid-mounted component must be demonstrated by satisfactory operation of the major component.

If the skid mounted component is a check valve, it does not have to be exercised in both directions if both direction testing is not required to indicate satisfactory operation.

  • The IST Bases Document should describe the bases for classifying a component as skid-mounted, and the IST Program Plan should reference this technical position for the component, if listed.Justification Recognition and classification of components as skid-mounted eliminates the need for the redundant testing of the sub component(s) as the testing of major (parent) component satisfactory demonstrates operation of the "skid mounted" component(s).

Revision 0 Page 169 Revision 0 Page 169 Cooper Nuclear Station F~ifh lnterval Inser-vice Testing Program for Pumps and Valves Technical Position TP-06 (Page 1 of 2)Check Valves in Regular Use Purpose The purpose of this Technical Position is to establish the station position for check valves that are in regular use during normal plant operations. Applicability. This Technical Position is applicable to check valves that are capable of being demonstrated to be open during routine operations.

Background

The ASME OM Code 2004 through 2006 Addenda section ISTC-3550, "Valves in Regular Use", states: "Valves that operate in the course of plant operation at a frequency that would satisfy the exercising requirements of this Subsection need not be additionally exercised, provided that the observations otherwise required for testing are made and analyzed during such operation and recorded in the plant record at intervals no greater than specified in ISTC-3510." Section ISTC-35 10 requires that check valves shall be exercised nominally every 3 months with exceptions (for extended periods) referenced. Check valves that are a part of the 1ST Check Valve Condition Monitoring Program shall be tested per the frequency requirements of that program.Normal and/or expected system flow may vary with plant configuration and alignment. The open"safety function" of a check valve typically requires a specified design accident flow rate. For these subject valves, the normal system flow is above the design accident flow rates. Since the Cooper Nuclear Station Operations staff is trained so as to be able to recognize normal plant conditions, Operator judgment has been deemed acceptable for the purpose of determining check valve open demonstration by observing either normal or expected flow rates for the plant operating condition. Position Cooper Nuclear Station will verify the open position of these subject check valves by observing plant logs, computer systems, strip chart recorders, etc., during normal plant operations. The open/closed safety function shall be recorded at a frequency required by ISTC -3510 [or ISTC-5222, Condition Monitoring Program, if applicable], nominally every 3 months, (with exceptions as provided), in plant records such as Cooper Nuclear Station Operating Logs, Electronic Rounds, chart recorders, automated data loggers, etc.Revision 0 Page 170 Revision 0 Page 170 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Technical Position TP-06 (Page 2 of 2)Justification Normal plant systems operation and operator actions provide for the observations and analysis that these subject valves are capable of satisfying their open safety function. Additionally, the recording of parameters which demonstrate valve position is satisfied at a frequency in accordance with JSTC-35 10 or ISTC-5222. These actions collectively demonstrate the open safety function of Inservice Testing Program check valves in regular use as required by ISTC-3550.Revision 0 Pg 7 Page 171 Cooper Nuclear Station F~fh~ Interval Inservice Testing Program for Pumps and Valves Technical Position TP-07 (Page 1 of 3)Categorization of IST Pumps (Group A or B)Position Cooper Nuclear Station has categorized the pumps required to be included in the Inservice Testing Program or Augmented (class "~A") Inservice Testing Program as either Group A or B in accordance with the requirements of ISTB-2004/2006a. Group A pumps are pumps that are operated continuously or routinely during normal operation, cold shutdown, or refueling operations. The following pumps are categorized as Group A at Cooper Nuclear Station: Pump CIC Class Group Type Function DGDO-P-DOTA A A Centrifugal Diesel Fuel Oil Transfer DGDO-P-DOTA A A Centrifugal Diesel Fuel Oil Transfer RW-P-Zl1 A A Centrifugal Elevated Release Point Sump RW-P-Z2 A A Centrifugal Elevated Release Point Sump REC-P-A 3 A Centrifugal 'Reactor Equipment Cooling REC-P-B 3 A Centrifugal Reactor Equipment Cooling REC-P-C 3 A Centrifugal Reactor Equipment Cooling REC-P-D 3 A Centrifugal Reactor Equipment Cooling RIIR-P-A 2 A Centrifugal Residual Heat Removal RHR-P-B 2 A Centrifugal Residual Heat Removal RHIR-P-C 2 A Centrifugal Residual Heat Removal RHIR-P-D 2 A Centrifugal Residual Heat Removal SW-P-A 3 A Vertical Service Water SW-P-B 3 A Vertical Service Water SW-P-C 3 A Vertical Service Water SW-P-D 3 A Vertical Service Water SW-P-BPA 3 A Centrifugal Service Water SW-P-BPB 3 A Centrifugal Service Water SW-P-BPC 3 A Centrifugal Service Water SW-P-BPD 3 A Centrifugal Service Water Revision 0 Pg 7 Page 172 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Technical Position TP-07 (Page 2 of 3)Group B pumps are those pumps in standby systems that are not operated routinely except for testing. The following pumps are categorized as Group B at Cooper Nuclear Station: Pump Number Class Group Type Function CS-P-A 2 B Centrifugal Core Spray CS-P-B 2 B Centrifugal Core Spray HTPCI-P-MP 2 B Centrifugal High Pressure Coolant min HiPCI-P-BP 2 B Centrifugal High Pressure Coolant mIn RCIC-P-MP 2 B Centrifugal Reactor Core Iso Cooling SLC-P-A A B Positive Disp Standby Liquid SLC-P-B A B Positive Disp Standby Liquid The following summarizes the Group A, B, and Comprehensive Pump Test requirements as specified by the ASME OM Code Subsection JSTB. This testing must be performed unless relief for alternate testing has been approved. The design flow rate is defined as the maximum accident flow rate for the pump.Group A Pump Tests -Group A tests are performed quarterly for each pump categorized as A.Reference values are established within +/--20% of pump design flow rate, if practicable. If not practicable, the reference point flow rate shall be established at the highest practical flow rate.For centrifugal pumps, the pump is operated at a nominal motor speed for constant speed drives or at a speed adjusted to the reference point (+/--1%) for variable speed drives. The resistance of the system is varied until the flow rate equals the reference point. Then, differential pressure and vibration measurements are determined and compared to their reference values.For positive displacement pumps, the pump is operated at a nominal motor speed for constant speed drives or at a speed adjusted to the reference point (+/-1%) for variable ,speed drives. The resistance of the system is varied until the discharge pressure equals the reference point. Then, flow and vibration measurements are determined and compared to their reference values.Group B Pump Tests -Group B tests are performed quarterly for each pump categorized as B.Reference values are established within +20% of pump design flow rate, if practicable. If not practicable, the reference point flow rate shall be established at the highest practical flow rate.For centrifugal pumps, the pump is operated at a nominal motor speed for constant speed drives or at a speed adjusted to the reference point (+ 1%) for variable speed drives.Then, the differential pressure or flow rate is determined and compared to its reference value.For positive displacement pumps, the pump is operated at a nominal motor speed for constant speed drives or at a speed adjusted to the reference point (+/--1%) for variable speed drives. Then, the flow rate is determined and compared to its reference value.Revision 0 Page 17 3 Cooper Nuclear Station F~fih Interval Inservice Testing Program for Pumps and Valves Technical Position TP-07 (Page 2 of 3)Comprehensive Pump Tests -Comprehensive pump tests are performed biennially for all pumps in the Inservice Testing Program. Reference values are established within +/-20% of pump design flow rate. The procedure to perform a comprehensive pump test is similar to the Group A test.The following instrument accuracy requirements apply to each test type of test: Parameter Group A Group B Comprehensive Pressure +/- 2.0% +/- 2.0% +/- 0.5%Flow Rate +/- 2.0% +/-/- 2.0% +/- 2.0%Speed +/- 2.0% +/- 2.0% +/- 2.0%Vibration +/- 5.0% +/- 5.0% +/- 5.0%Differential Pressure +/- 2.0% +/- 2.0% +/- 0.5%Revision 0 Pg 7 Page 174 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Technical Position TP-8 (Page 1 of 2)Vacuum Breaker Testing Purpose The purpose of this Technical Position is to establish the station position for testing vacuum breakers.Applicabilit. This Technical Position is applicable to vacuum breakers that are included in the Inservice Testing Program.Position In accordance with the requirements of ISTC-5230 for vacuum breaker valves, vacuum breakers shall meet the applicable inservice test requirements of ISTC-5220 and Mandatory Appendix I.ISTC-5220 is the requirement for the valve obturator movement of check valves and Mandatory Appendix I is the requirement for the Inservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants. The testing performed at Cooper will meet both requirements. Check valves that are capacity certified and are functioning as a vacuum breaker, will be tested in accordance with Mandatory Appendix I. Per 1-3370, Class 2 and 3 Vacuum Relief Valves shall be actuated to verify open and close capability, set-pressure, and performance of any pressure and position-sensing accessories. Seat tightness shall be in compliance with the owner's seat tightness criteria. The disposition after testing or maintenance shall be in accordance with 1-3470 for class 2 and 3 vacuum relief valves.Per 1-1370, the frequency of testing Class 2 and 3 containment vacuum relief valves shall be at each refueling outage or every 2 years, whichever is sooner, unless historical data requires more frequent testing. This code required frequency recognizes that this testing may be able to be performed online by having the "or every 2 years" criteria and "unless historical data requires more frequent testing." Leak test frequencies are designated by the owner in accordance with Table ISTC-3500-l. Per 1-13 80, those Class 2 and 3 vacuum relief valves, other than Primary Containment Vacuum Relief Valves, shall be tested every 2 years, unless performance data suggest the need for a more appropriate test interval.Mandatory Appendix I requirements contain all the necessary requirements to adequately test 1ST vacuum relief valves. However, in order to satisfy ISTC-5220, as required by JSTC-5230, exercise testing in the open (FSO) and closed (FSC) positions have also been indicated within the Attachment 11 Inservice Testing Valve Table in addition to the VBT designation for the vacuum breaker test. The exercise testing requirements of ISTC-5220 (FSO/FSC) are being met within the testing performed per Mandatory Appendix I. No further testing is required.Revision 0 Pg 7 Page 175 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Technical Position TP-8 (Page 2 of 2)Justification This position was only established to clearly define the basis for how the vacuum breakers are being tested at CNS and at what frequency. All code requirements are being met. As stated above, the vacuum breaker testing fully encompasses the exercise testing requirements of IST-5220. The required code frequencies for testing are clearly being met.Revision 0 Page 176 Revision 0 Page 176 Cooper Nuclear Station Fifthi Interval Inservice Testing Pro gram for Pumps and Valves ATTACHMENT 10 INSERVICE TESTING PUMP TABLE Revision 0 Page 177 Revision 0 Page 177 Cooper Nuclear Station F/ifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: CORE SPRAY (CS), P&ID No. 2045 Sheet 1 Pump P&IID ISI Pump __Parameters CIC Coor Class Group Q dP V N Notes CS-P-A F-3 2 B Q Q 2Y (1) cs Pump A________ RP-01, RP-05, RP-09, RG-01, TP-07 CS-P-B D-3 2 B Q Q 2Y (1) CS Pump B RP-01, RP-05, RP-07, RP-09, RG-01,____ ___ _ __ ___ ___ __ _ ___ ___ ___ ___TP-07 NOTES: (1) Pump is directly coupled to a constant speed synchronous or induction type driver.SYSTEM: DIESEL GENERATOR FUEL OIL TRANSFER (DGDO), P&1ID No. 2077 and 2011 Sheet 1 Pump P&ID ISI Pump Parameters CIC Coor Class Group Q dP V N Notes DGDO-P- G-3 A NA NA NA NA NA Skid-Mounted, RG-01, TP-05 EDF 1 DGDO-P- G-4 A NA NA NA NA NA Skid-Mounted, RG-Ol, TP-05 EDF 1 DGDO-P- A-8 A A Q Q 6M (1) DGDO Pump A, ARP-04, RG-01, DOTA __TP-07 DGDO-P- A-10 A A Q Q 6M (1) DGDO Pump B, ARP-04, RG-01, DOTB TP -07 NOTES: (1) Pump is directly coupled to a constant speed synchronous or induction type driver.SYSTEM: HIGH PRESSURE COOLANT INJECTION (IIPCI), P&llD No. 2044 Pump P&ID ISI Pump Parameters__ CIC Coor Class -Group Q dP V N Notes HPCI-P-MP E-4 2 B Q Q 2Y -Q -HiPCI Pump Main (1)_________ ____ ____ _________ ___ ___ ___RP-03, RP-05, RP-09, RG-01, TP-07 HPCI-P-BP E-3 2 B Q Q -2Y -Q HPCI Pump Booster(l) _________ ____ ____ _________ ___ ___ ___RP-03, RP-05, RP-09, RG-01, TP-07 NOTES: (1) HPCI main and booster pumps will be tested simultaneously. Revision 0 Page 178 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: RADIOACTIVE WASTE (RW), P&LD No. 2005 Sheet 2 Pump dCI P&ID ISI Coor Class Pump Group Q Parameters JdP [V N INotes RW-P-Z1 G-10 A A Q NA NA (1) Elevated Release Point Sump Pump, RWPZ2 G1O A A (2) N NAARP-01, RG-01, TP-07 RWPZ2 G10 A A Q N NA()Elevated Release Point Sump Pump,___________ ____ ____(2) _____ ARIP-01, RG-O1, TP-07 NOTES: (1) Pump is directly coupled to a constant speed synchronous or induction type driver.(2) The time (TM) to pump a specified quantity of water from the sump will be measured and trended.SYSTEM: REACTOR CORE ISOLATION COOLING (RCIC), P&ID No. 2043 Pump P&ID ISI Pump Parameters CIC Coor Class Group Q dP V N Notes RCIC.-P-MP G-3 2 B Q Q 2Y Q RCJC Pump 1A RP-04, RP-09, RG-01, TP-07 SYSTEM: REACTOR EQUIPMENT COOLING (REC), P&ID No. 2031 Sheet 2 Pump P&LD ISI Pump Parameters CIC Coor Class Group Q dP V N Notes REC-P-A G-l 3 A Q Q Q (1) REC Pump lA, Loop A RP-06. RP-08, RP-09, RG-01, TP-07 REC-P-B G-2 3 A Q Q Q (1) REC Pump 1B, Loop A______________________________RP-06, RP-08, RP-09, RG-01, TP-07 REC-P-C G-3 3 A Q Q Q (1) REC Pump 1C, Loop B RP-06, RP-08, RP-09, RG-01, TP-07 REC-P-D G-3 3 A Q Q Q (1) REC Pump 1D, Loop B_____________________________ __________RP-06, RP-08, RP-09, RG-O1, TP-07 NOTES: (1) Pump is directly coupled to a constant speed synchronous or induction type driver.Revision 0 Page 179 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RHR), P&ll) No. 2040 Pump P&LD ISI Pump Parameters CIC Coor Class Group Q dP V N Notes RHR.-P-A G-4 2 A Q Q Q (1) RHR Pump 1A, Loop A RP-02, RP-08, RP-09, RG-01, TP-07 RHiR-P-B G-9 2 A Q Q Q (1) RHR Pump iB, Loop B RP-02, RP-08, RP-09, RG-Ol, TP-07 RHR-P-C H-4 2 A Q Q Q (1) RHR Pumpl1C, Loop A_____ ______RP-02, RP-08, RP-09, RG-O1, TP-07 RHR-P-D H-9 2 A Q Q Q (1) RHR Pump 1D, Loop B_______________RP-02, RP-08, RP-09, RG-Ol, TP-07 NOTES: (1) Pump is directly coupled to a constant speed synchronous or induction type driver.SYSTEM: SERVICE WATER (SW), P&1D No. 2006 Sheets 1 and 4'Pump P&ID 1S1 Pump Parameters CIC Coor Class Group Q dP V N Notes SW-P-A B3-10 3 A Q Q Q (1) sw Pump lA, RP-08, RP-09, RG-01, TP-07 SW-P-B B3-9 3 A Q Q Q (1) sw Pump 11B, RP-08, RP-09, RG-01,______TP-07 SW-P-C B3-8 3 A Q Q Q (1) SW Pump 1C, RP-08, RP-09, RG-01, TP-07 SW-P-D 13-7 3 A Q Q Q (1) sw Pump 1D, RP-08, RP-09, RG-01, TP-07 SW-P-BPA F-7 3 A Q Q Q (1) SW Booster Pump 1A RP-05, RP-09, RG-01, TP-07 SW-P-BPB C-7 3 A Q Q Q (1) sw Booster Pump lB RP-05, RP-09, RG-0l, TP-07 SW-P-BPC E-7 3 A Q Q Q (1) SW Booster Pump IC RP-05, RP-09, RG-01, TP-07 SW-P-BPD A-7 3 A Q Q Q (1) SW Booster Pump 1D____________RP-05, RP-09, RG-01, TP-07 NOTES: (1) Pump is directly coupled to a constant speed synchronous or induction type driver.Revision 0 Page 180 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: STANDBY LIQUID CONTROL (SLC), P&IID No. 2045 Sheet 2 Pump P&ID ISI Pump ___Parameters __CIC Coor Class Group Q Pd V N Notes SLC-P-A E-10 A B Q Q 6M (1) SLC Pump lA_________ ____ _____ _____ ___ ___ ARP-02, ARP-03, RG-01, TP-07 SLC-P-B F-10 A B Q -Q -6M (1) SLC Pump 1A_________ ____ _____ _____ ___ ___ ___ ARP-02, ARP-03, RG-01, TP-07 NOTES: (1) Positive displacement pump Revision (9 Page 181 Revision 0 Page 181 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves ATTACHMENT 11 INSERVICE TESTING VALVE TABLE Revision 0 Page 182 Revision 0 Page 182 Cooper Nuclear Station Fifth Interval lnservice Testing Pro gram for Pumps and Valves SYSTEM: CONTROL ROD DRIVE (CR]))VALVE CIC P&IDP&ID ISI IST 1VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR _CLASS CAT SIZE TYPE TYPE POS RQMT FREQ CRD-AOV-CV32A 2039 J2 2 B 1 GB AO 0 FSC Q SOUTH SDIV INBOARD VENT FST Q ISOLATION VALVE, RG-01, TP-04_________ ____ ___ __ _____PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _CRD-AOV-CV32B 2039 J4 2 B 1 GB AO 0 FSC Q NORTH SDIV INBOARD VENT FST Q ISOLATION VALVE, RG-01, TP-04_________ ____ ___ __ _____PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _CRD-AOV-CV33' 2039 H3 2 B 2 GB AO 0 FSC Q SOUTH SDIV INBOARD DRAIN FST Q ISOLATION VALVE, RG-01, TP-04______________ ______PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _CRD-AOV-CV34 2039 H4 2 B 2 GB AO 0 FSC Q NORTH SDIV INBOARD DRAIN FST Q ISOLATION VALVE, RG-01, TP-04_________ ____ ___ __ _____PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _CRD-AOV-CV35 2039 H3 2 B 2 GB AO 0 FSC Q SOUTH SDIV OUTBOARD DRAIN FST Q ISOLATION VALVE, RG-01, TP-04_________ ____ ____ _ ______PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _CRD-AO V-C V36 -2039 H4 2 B 2 GB AO 0 FSC Q NORTH SDIV OUTBOARD DRAIN FST Q ISOLATION VALVE, RG-01, TP-04"_________ ____ ___ __ _____PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _CRD-AOV-CV38A 2039 J3 2 B 1 GB AO 0 FSC Q SOUTH SD1V OUTBOARD VENT FST Q ISOLATION VALVE, RG-01, TP-04_________ ____ ____ _ ______PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _FSC Q CRD-AOV-CV38B 2039 J4 2 B 1 GB AO 0 FST Q NORTH SDIV OUTBOARD VENT______________ ___ ____ ___ ____ ____ ________ PIT 2Y ISOLATION VALVE, RG-01, TP-04 Revision 0 Page 183 Revision 0 Page 183 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: CONTROL ROD DRIVE (CRD)VALVE GIG P&ID P&ID IS I5 ST 1VALVE VALVE ACT 1NORM TEST 1TEST NOTES/DESCRIPTION COOR _CLASS CATj SIZE TYPE TYPE jPOS RQMT FREQ CRD-AOV-CV126 2039 C10 B 1 DIA AO C FSO TS SCRAM INLET, (TYP. OF 137) ________FST TS RV-04, RG-01*CRD-AOV-CV127 2039 B10 B 3/4 DIA AO C FSO TS SCRAM OUTLET, (TYP. OF 137) ____FST TS RV-04, RG-01 CRD-CV-13CV 2039 A9 1 A/C 3/4 CK-P SA 0 LJ-1 OPB CRD WATER TO REACTOR CCL CVCM RECIRCULATION PUMP A, COF CVCM RG-01, TP-01, TP-06 CRD-CV-14CV 2039 B9 1 A/C 3/4 CK-P SA 0 LJ-1 OPB CRD WATER TO REACTOR CCL CVCM RECIRCULATION PUMP A,_________ ____ ___COP CVCM RG-0i, TP-01, TP-06 CRD-CV-15CV 2039 A8 1 A/C 3/4 CK-P SA 0 LJ-1 OPB CRD WATER TO REACTOR*CCL CVCM RECIRCULATION PUMP B,_________ ____ ___COF CVCM RG-01, TP-01, TP-06 CRD-CV-16CV 2039 B8 1 A/C 3/4 CK-P SA 0 LJ-1 OPB CRD WATER TO REACTOR CCL CVCM RECIRCULATION PUMP B,__________COF CVCM RG-01, TP-01, TP-06 CRD-CV-25CV 2039 B4 A A/C 1 1/2A CK-S SA 0 LT-2 RF CRD SYSTEM ISOLATION CHECK FSC RF VALVE, RG-01, TP-01, ROJ-12___ ___ __ ___ __ __ ___ _ ___ _ _FSO5 RF CRD-CV-26CV 2039 B4 A A/C 1 1/22 CK-S SA 0 LT-2 RF CRD SYSTEM ISOLATION CHECK FSC RE VALVE, RG-01, TP-01, ROJ-12_____ ____ ___ ____ ____ ___ __ _ ___ ____ ___ __ _ ___ ____ FSO RF _ _ _ _ _ _ _ _ _ _ _Revision 0 Page 184 Cooper Nuclear Station Ffith Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: CONTROL ROD DRIVE (CRD)VALVE GIGC P&ID {P&ID ISI{ 1ST VALVE VALVE 1ACT[ NORM ITEST[ TEST NOTES/DESCRIPTION CORCASCAT SIZE TYETYPE POSRQMT[FE CRD-CV-28CV 2039 H2 A C 1/2A CK-P SA C SKID Q SOUTH SDIV DRAIN VALVE AIR__________SUPPLY BYPASS CV, RG-01, TP-05 CRD-CV-29CV 2039 H5 A C 1/2A CK-P SA C SKID Q NORTH SDIV DRAIN VALVE AIR____ _____ _____SUPPLY BYPASS CV, RG-01, TP-05 CRD-CV-30CV 2039 H3 A C 1/2A CK-P SA C SKID Q SOUTH SDIV VENT V AIR____ ____ _________ ______SUPPLY BYPASS CV, RG-01, TP-05 CRD-CV-31CV 2039 H4 A C 1/2A CK-P SA C SKID Q NORTH SDV VENT VAIR SUPPLY____ ____ _____BYPASS CV, RG-01, TP-05 CRD-CV-32CV 2039 H3 A C 1/2A CK-P SA C SKID Q SOUTH SDIV DRAIN VALVE AIR____ ____ ______SUPPLY BYPASS CV, RG-01, TP-05 CRD-CV-33CV 2039 H5 A C 1/2A CK-P SA C SKID Q NORTH SDIV DRAIN VALVE AIR____ ____ ______SUPPLY BYPASS CV, RG-01, TP-05 CRD-CV-34CV 2039 J3 A C 1/2A CK-P SA C SKID Q SOUTH SDV VENT V AIR SUPPLY_____ _____BYPASS CV, RG-01, TP-05 CRD-CV-35CV 2039 J4 A C 1/2A CK-P SA C SKID Q NORTH SDV VENT V AIR SUPPLY____ ____ ____ ______BYPASS CV, RG-01, TP-05 CRD-CV-CV114 2039 Bl 21 C 3/4 CK-B SA 0 FSO TS SCRAM OUTLET CHECK VALVE, (TYP. OF 137) ____ ___RV-04, RG-01, TP-05 CRD-CV-CV1 15 2039 D9 2 C 1/2A CK-B SA 0 FSC RF SCRAM INLET CHECK VALVE, (TYP. OF 137) ____________FSO RjF RG-01, TP-01, ROJ-06 Revision 0 Page 185 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: CONTROL ROD DRIVE (CRD)VALVE GIG P&1ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ CRD-CV-138 2039 C9 2 C 1/22 CK-B SA 0 FSC TS COOLING/SCRAM HEADER (TYP. OF 137) CHECK VALVE, RV-04, RG-01, TP-_________________ ______ _____ ______ __ ___ ___ 05 CRD-SOV-SO120 2039 C9 2 B 114 SOV SO C FSC TS CRD WITHDRAWAL EXHAUST (TYP. OF 137) F___ __ _______ ST TS VALVE, RV-04, RG-01, TP-05 CRD-SOV-SO121 2039 B10 B 1/4 SOV SO C FSC TS CRD INSERT EXHAUST (TYP. OF 137) _______ ___ __ ___ ___FST TS VALVE, RV-.04, RG-01, TP-05 CRD-SOV-SO 122 2039 B10 2 B 1/4 SOV SO C FSC TS CRD WITHDRAWAL VALVE, (TYP. OF 137) ____ ___ _______ ____FST TS RV-04, RG-01, TP-05 CRD-SOV-SO0123 2039 B10 2 B 1/4 SOV SO C FSC TS CRD INSERT VALVE, (TYP. OF 137) ____ _______ _______ ___ _______ FST TS RV-04, RG-01, TP-05 Revision 0 Page 186 Revision 0 Page 186 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: CORE SPRAY (CS){P&ID ISI IST VALVE VALVE fACT NORM TEST TEST VALVE CIC P& COO CLASS CAT SIZE TYPE [TYPE POS -RQMT FREQ NOTES/DESCRIPTION CS-CV-10CV 2045 F5 2 C 12 CK-S SA C FSO Q CS PUMP A DISCHARGE CHECK SH 1 ______FSC Q R.G-01, TP-01 CS-CV-11CV 2045 C5 C 12 CK-S SA C FSO Q CS PUMP B DISCHARGE CHECK SH 1 ___ __ __ FSC Q RG-01, TP-01 CS-CV-12CV 2045 G7 A C 2 CK-P SA C COF CVCM CS LOOP A OUTBOARD PRESSURE SH 1 CCF CVCM MAINTENANCE SUPPLY, RG-01, TP-_______01 CS-CV-13CV 2045 G7 2 C 2 CK-P SA C COF CVCM CS LOOP A INBOARD PRESSURE SH 1 CCF CVCM MAINTENANCE SUPPLY, RG-01, TP-CCD/ CVCM 01______ _____ ~~~~CCR _ _ _ _ _ _ _ _ _ _ _ _ _CS-CV-14CV 2045 D7 A C 2 CK-P SA C COF CVCM CS LOOP B OUTBOARD PRESSURE SHi1 CCF CVCM MAINTENANCE SUPPLY, RG-01, TP-______ _______01 CS-CV-15CV 2045 D7 2 C 2 CK-P SA C COF CVCM CS LOOP B INBOARD PRESSURE SH 1 CCF CVCM MAINTENANCE SUPPLY, RG-0 1, TP-CCD/ CVCM 01 CCR__ _ _ _ _ _ _ _ _ _ _ _CS-CV-18CV 205 D0 1 AC 10 CK-S SA C LT-2 OPB CS SYSTEM A TESTABLE CHECK SH1FSC RF ROJ-11, RV-05, RG-01 FSO RF PIT RE Revision 0 Page 187 Cooper Nuclear Station Ffifti Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: CORE SPRAY (CS)1P&ID ISI IST VALVE VALVE TACT NORM TEST TEST VALVE CIC P&ID COOR CLASS [CAT SIZE TYPE _TYPE P0S RQMT FREQ NOTES/DESCRIPTION CS-CV- 19CV 2045 Bl 10 A/C 10 CK-S SA C LT-2 OPB CS SYSTEM B TESTABLE CHECK SHi1 FSC RF ROJ-11, RV-05, RG-01 FSO RE_____PIT RF CS-MOV-MO5A 2045 E5 2 B 3 GT MO 0 FSO Q CS PUMP A MINIMUM FLOW SH 1 FSC Q RECIRCULATION ISOLATION, RG-____PIT 2Y 01 CS-MOV-MO5B 2045 B5 2 B 3 GT MO 0 FSO Q CS PUMP BMINIMUM FLOW SHi1 FSC Q RECIRCULATION ISOLATION, RG-____PIT 2Y 01 CS-MOV-MO7A 2045 F2 2 B 14 GT MO 0 FSO Q CS PUMP A SUCTION, RG-01 SHi1 FSC Q_____ _____ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _CS-MOV-MO7B 2045 C2 2 B 14 GT MO 0 FSO Q cs PUMP B SUCTION, RG-01 SHi1 FSC Q______________ _____ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _CS-MOV-MO11lA 2045 F8 2 B 10 GT MO 0 PIT 2Y LOOP A INJECTION THROTITLE, RG-__ _ _ _ _ _ _ SH 1_ __ 01 CS-MOV-MO11B 2045 C8 2 B 10 GT MO 0 PIT 2Y LOOP B INJECTION THROTTPLE, RG-__ _ _ _ _ _ _ SH 1 __ _ _ _ _ _ _ _ _ _ _ _ _01 CS-MOV-MO 12A 24 FS 1 A 10 GT MO C LJ-l1 OPB LOOP A INJECTION BLOCK SiLT-2 OPB CSJ-09, RV-05, RG-01 FSO CS Rev&ion 0 Page 188 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: CORE SPRAY (CS)P&ID ISI IST VALVE VALVE ACT 1NORM TEST TTEST VALVE CIC P&ID COOR [C LASS CAT SIZE TYPE TYPE POS RQMT jFREQ NOTES/DESCRIPTION FSC CS_____________ _____ _____ PIT 2Y_ _ _ _ _ _ _ _ _ _ _ _ _CS-MO V-MO 12B 2045 C8 1 A 10 GT MO C LJ-1 OPB LOOP B INJECTION BLOCK Sill LT-2 OPB CSJ-09, RV-05, RG-01 FS0 CS FSC CS___ __ __ __ _ __ _ __ __PIT 2Y CS-MOV-MO26A 2045 F7 2 B 10 GB MO C PIT 2Y CS PUMP A TEST LINE PASSIVE__ _ _ _ _ _ _ Sil 1 _ _ _ _ __ _ __ _ _ ISOLATION, RG-01 CS-MOV-MO26B 2045 C7 2 B 10 GB MO C PIT 2Y CS PUMP B TEST LINE PASSIVE_ _ _ _ _ _ _ SI-I 1 __ _ ___ISOLATION, RG-01 CS-RV-10RV 2045 H3 2 C 3/4 RV SA C RVT APP I CS PUMP A SUCTION RELIEF, RG-__ _ _ _ _ _ _ Sil 1 _ _ _ _ __ _ _ 01 CS-RV-11RV 2045 F6 2 C 2 RV SA C RVT APP I CS PUMP A DISCHARGE RELIEF,_ _ _ _ _ _ Sil 1__ RG-01 CS-RV-12RV 2045 E3 2 C 3/4 RV SA C RVT APP I CS PUMP B SUCTION RELIEF, RG-01___________ Sil 1 _ _ _ _ __ _ _ _ _ _ _ _ _ _CS-RV-13RV 2045 C6 2 C 2 RV SA C RVT APP I CS PUMP B DISCHARGE RELIEF, Sill RG-01 Revision 0 Pg 8 Page 189 Cooper Nuclear Station F~ith Interval lnservice Testing Program for Pumps and Valves SYSTEM: DEMINERALIZED WATER SYSTEM (DW)VALVE CIC P&ID 1P&ID ISI IST 1VALVE IVALVE ACT INORM TEST JTEST f NOTES/DESCRIPTION COOR CLASS CAT jSIZEJTYPE TYPE POS RQMT FREQ DW-V-133 2029 G8 2 A 4 GT MA C LJ-1 OPB PASSIVE DRYWELL OUTBOARD____ ___ ___ __ __ ___ __ ___ ___ ___ _ _ ___ ___ _ _ ___ ___ _ _ ___ SUPPLY VALVE DW-V-219 2029 G8 2 A 4 GT MA C L -i OPB PASSIVE DRYWELL INBOARD SUPPLY VALVE Revision 0 Page 190 Revision 0 Page 190 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: DIESEL GENERATOR DIESEL OIL (DGDO)VALVE CIC P&ID P&ID ISI IST VALVE VALVE IACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ_____________ DGDO-CV-°10CV 2011 A8 A C 2 CK-P SA C FSO Q DGDO TRANSFER PUMP A SH 1______ __ __ DISCHARGE, ARV-03, RG-01 DGDO-CV-1 1CV 2011 A10 A C 2 CK-P SA C FSO Q DGDO TRANSFER PUMP B SHi 1______ __ __ DISCHARGE, ARV-03, RG-01 DGDO-CV-12CV 2011 A4 A C- 2 CK-P SA C FSO Q DGDO DAY TANK 1 INLET, ARV-03,_ _ _ _ _ _ SHi 1_ _ _ _ _ _ _ _ _ _ RG-01 DGDO-CV-13CV 2011 B4 A C 2 CK-P SA C FSO Q DGDO DAY TANK 2 INLET, ARV-03, SHi 1 _ _ _ _ _ _ _ _ _ _ __ _ RG-01 DGDO-CV-14CV 2077 J3 A C 1 1/4 CK-S SA C SKID Q ENG DR FUEL P 1 SUPPLY, RG-01,____ ___ ___ ___ __ ___ ____ ___ ___ __ ___ ___TP-05 DGDO-CV-15CV 2077 J4 A C 1 1/4 CK-S SA C SKID Q ENG DR FUEL P 2 SUPPLY, RG-01,_______ ____ _ _____ ______ TP-05 DGDO-CV-16CV 2077 H3 A C 1 1/4 CK-S SA C SKID Q FUEL BOOSTER P 1 DISCH, RG-01,____ _ ______ ___ ___TP-05 DGDO-CV-17CV 2077 H4 A C 1 1/4 CK-S SA C SKID Q FUEL BOOSTER P 2 DISCH, RG-01,____ ___ ___ ____ _____ __ ____ _____ ___ ___ ___ ___TP-05 DGDO-PRV-PRV101 2077 E3 A C 3/4 PRV SA C SKID Q DG1 FUEL INJECT HDR PRESS REG____ __ ______ _ _ ___ __ ___V, RG-01, TP-05 DGDO-PRV-PRV102 2077 E4 A C 3/4 PRV SA C SKID Q DG2 FUEL INJECT HDR PRESS REG____ ___ ___ __ ___ _ _ ___ ___ ___ ___ _ _ ___ _ _ ___ ___ __ ___ _ _ ___ __ ___ V, RG-01, TP-05 Revision 0 Page 191 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: DIESEL GENERATOR DIESEL OIL (DGDO)VALVE CIC P&ID 1P&ID ISI IST 1VALVE TVALVE ACT INORM TEST TEST [ NOTES/DESCRIPTION COOR CLASS CATj SIZE TYPE TYPE POS RQMT FREQ DGDO-RV-10RV 2077 H2 A C 3/4 RV SA C RVT APP I DGDO PUMP 1 SUCTION RELIEF,_____ARV-02, RG-01 DGDO-RV-1 1RV 2077 H4 A C 3/4 RV SA C RVT APP I DGDO PUMP 2 SUCTION RELIEF,____ ___ ___ _ _ ___ ___ _ _ ___ARV-02, RG-01 DGDO-SOV-2077 112 A B 3/4 SOV SO 0/C FSO Q DGDO DAY TANK 1 INLET FUEL SSV5028 ________________SAFETY VALVE, ARV-Ol, RG-01 DGDO-SOV-2077 116 A B 3/4 SOV SO 0/C FSO Q DGDO DAY TANK 2 INLET FUEL SSV5029 SAFETY VALVE, ARV-01l, RG-01 Revision 0 Pg 9 Page 192 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: DIESEL GENERATOR STARTING AIR (DGSA)VALVE CIC P&ID P&ID 1SI IST VALVE IVALVE ACT INORM TEST ITEST NOTES/DESCRIPTION COO CLAS CAT SIZE ITYPE TYPE POS RQMT IFREQ_____ __ ~R ___S_ _ I _ _ I _ _ _ _ _DGSA-AOV-AV5 117.10 E3 A B 2 1/2A GB AO C SKID Q DG-l LEFT BANK STARTING AIR-IC-09 _______________VALVE, RG-01, TP-05 DGSA-AOV-AV6 117.10 E3 A B 2 1/2A GB AO C SKID Q DG-2 LEFT BANK STARTING AIR-IC-09 ____ ___ _________VALVE, RG-01, TP-05 DGSA-AOV-AV7 117.10 El A B 2 1/2A GB AO C SKID Q DG-1 RIGHT BANK STARTING-IC=09 ________AIR VALVE, RG-01, TP-05 DGSA-AOV-AV8 117.10 El A B 2 1/2A GB AO C SKID Q DG-2 RIGHT BANK STARTING-IC-09 ____AIR VALVE, RG-01, TP-05 DGSA-CV-10CV 2077 D7 A C 2 CK-L- SA C FSC Q STARTING AIR COMPRESSOR 1A_______ ______FSO Q DISCHARGE, RG-01, TP-01 DGSA-CV-11CV 2077 D9 A C 2 CK-L SA C FSC Q STARTING AIR COMPRESSOR lB____ ______ ___ ___FSO Q DISCHARGE, RG-01, TP-01 DGSA-CV-12CV 2077 D10 A C 2 CK-L SA C FSC Q STARTING AIR COMvPRESSOR 2B__________F__ SO Q DISCHARGE, RG-01, TP-01 DGSA-CV-13CV 2077 D12 A C 2 CK-L SA C FSC Q STARTING AIR COMPRESSOR 2A F___ SO Q DISCHARGE, RG-01, TP-01 DGSA-CV-14CV 2077 C8 A C 2 CK-S SA C CCL CVCM AIR RECEIVER IA INLET, RG-.01,_______ __ ___ ___COF CVCM TP-01 DGSA-CV-15CV 2077 C9 A C 2 CK-S SA C CCL CVCM AIR RECEIVER lB INLET, RG-01, COF CVCM TP-01 Revision 0 Page 193 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: DIESEL GENERATOR STARTING AIR (DGSA)VALVE CIC IP&ID IP&ID 1SI IST 1VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION _______j jCOO CLAS CAT SIZE TYPE TYPE POS RQMT FREQ R S DGSA-CV- 16CV 2077 Cl10 A C 2 CK-S SA C CCL CVCM AIR RECEIVER 2A INLET, RG-01, COF CVCM TP-01 DGSA-CV-17CV 2077 C11 A C 2 CK-S SA C CCL CVCM AIR RECEIVER 2B INLET, RG-01, COF CVCM TP-01 DGSA-CV-18CV 2077 B7 A C 3 CK-S SA C FSO Q AIR RECEIVER 1A OUTLET, RG-01 FSC Q DGSA-CV-19CV 2077 B8 A C 3 CK-S SA C FSO Q AIR RECEIVER lB OUTLET, RG-01 FSC Q DGSA-CV-20CV 2077 B 11 A C 3 CK-S SA C FSO Q AIR RECEIVER 2A OUTLET, RG-01 FSC Q DGSA-CV-21CV 2077 B12 A C 3 CK-S SA C FSO Q AIR RECEIVER 2B OUTLET, RG-01 FSC Q DGSA-CV-30CV 117.10-Ic-D3 A C 1/4 CK-S SA 0 SKID Q DG1 125 PSIG STRT AIR SUPPLY 09 SHUTTLE V, RG-0l, TP-05 DGSA-CV-31CV 117.10-IC-D3 A C 1/4 CK-S SA 0 SKID Q DG2 125 PSIG STRT AIR SUPPLY 09 SHUTTLE V, RG-01, TP-05 DGSA-CV-32CV 117.10-IC-F5 A C 1/4 CK-S SA C SKID Q DG1 LEFT BANK AIR START 09 SHUTTLE VALVE, RG-01, TP-05 DGSA-CV-33CV 1 17.10-ic-F1 A C 1/4 CK-S SA C SKID Q DG1 RIGHT BANK AIR START 09 SHUTTLE VALVE, RG-01, TP-05 Revision 0 Page 194 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: DIESEL GENERATOR STARTING AIR (DGSA)VALVE CIC P&ID P&ID IISI IST VALVE VALVE IACT NORM TEST TEST I NOTES/DESCRIPTION COO jCLAS CAT SIZE TYPE ITYPE POS RQMT FREQ_ _ _ _ _ _ _ _ _ _ R j S _ _ _ _ _ _ _ _ _ _ _DGSA-CV-34CV 117.10-Ic-F5 A C 1/4 CK-S SA C SKID Q DG2 LEFT BANK AIR START o9___SHUTTLE VALVE, RG-01, TP-05 DGSA-CV-35CV 117.10-IC-F1 A C 114 CK-S SA C SKID Q DG2 RIGHT BANK AIR START 09 SHUTTLE VALVE, RG-01, TP-05 DGSA-SOV-SPV1 117.10 F4 A B 1/2A GB SO C SKID Q SOLENOID PILOT VALVE FOR-IC-09 DGSA-AOV-AV5 (DG1 LEFT_____ ______BANK), RG-01, TP-05 DGSA-SOV-SPV2 117.10 F4 A B 1/2A GB SO C SKID Q SOLENOID PILOT VALVE FOR-IC-O9 DGSA-AOV-AV6 (DG2 LEFT______________ _____BANK), RG-01, TP-05 DGSA-SOV-SPV3 117.10 Fl A B 1/2A GB SO C SKID Q SOLENOID PILOT VALVE FOR-IC-O9 DGSA-AOV-AV7 (DG1 RIGHT___________________ ____ ____________BANK), RG-01, TP-05 DGSA-SOV-SPV4 117.10 Fl A B 1/2A GB SO C SKID Q SOLENOID PILOT VALVE FOR-IC-O9 DGSA-AOV-AV8 (DG2 RIGHT_____ ___________BANK), RG-0I, TP-05 DGSA-RV-14RV 2077 C7 A RV 1 RV SA C RVT APP I DGSA AIR RECEIVER 1A RELIEF,____ ___ ___ __ _ ___RG-01 DGSA-RV-15RV 2077 C9 A RV 1 RV SA C RVT APP I DGSA AIR RECEIVER lB RELIEF,_____RG-01 DGSA-RV-16RV 2077 Cl V 1 RV SA C RVT APP I DGSA AIR RECEIVER 1C RELIEF, Revision 0 Page 195 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: DIESEL GENERATOR STARTING AIR (DGSA)VALVE CIC P&ID P&ID IS 1 I ST VALVE VALVE ACT NORM TEST TEST J NOTES/DESCRIPTION COO CLAS CAT SIZE TYPE }TYPE POS RQMT FREQ____ ___ ___ _ __ ___ ___ _ IRG-01 DGSA-RV-17RV 2077 C12 A RV 1 RV SA C RVT APP I -DGSA AIR RECEIVER 1D RELIEF,____ ___ ___ _ _ ___ __ __ ___ ___ __ ______ _j ___ __ ___ RG-01 Revision 0 Page 196 Revision 0 Page 196 Cooper Nuclear Station Fifth Interval Inserviee Testing Pro gram for Pumps and Valves SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPC1)1P&ID IISI IST VALVE VALVE IACT NORM TEST TEST VALVE CIC P&ID COOR _CLASS CAT SIZE TYPEJTYPE POS RQMT FREQ [ NOTES/DESCRIPTION HPCI-AOV-AO42 2041 J6 2 B 1 GB AO 0 FSC Q STEAM LINE DRIPLEG DRAIN, FST Q RG-01, TP-04___ ___ __ ___ __ _____ __ PIT 2Y HPCI-AOV-AO70 2044 E9 2 A 1 BAL AO C LJ-1 OPB HPCI EXHAUST BOOTLEG DRAIN FSC Q INBOARD, RG-01, TP-04 FST Q_______________ ______ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _HPCI-AOV-AO71 2044 E9 2 A 1 BAL AO C LI-i OPB HPCI EXHAUST BOOTLEG DRAIN FSC Q OUTBOARD, RG-01, TP-04 FST Q_____ _____ __ __ __ __ __ __ ___ _ ___ _ ____ _____ PIT 2Y_ _ _ _ _ _ _ _ _ _ _ _ _HPCI-AOV-PCV50 2044 H3 2 B 2 GB AO 0 PS0 Q I-IPCI AUXILIARY COOLING FSO CS SUPPLY PRESSURE CONTROL,________FST CS CSJ-02, RG-01, TP-04 I-PCI-CV-10CV 2044 13 2 C 16 CK-S SA C COF CVCM ECST SUPPLY TO HPCI PUMP,___________CCL CVCM RG-01 HPCI-CV-1 1CV 2044 1110 2 C 16 CK-S SA C COD CVCM HPCI PUMP SUCTION FROM___ ______CCD CVCM SUPPRESSION POOL, RG-0 1 HPCI-CV-13CV 2044 15 2 C 4 CK-S SA C COF CVCM HPCI AUXILIARY COOLING_________ ___ ___CCD CVCM RETURN, RG-01 Revision 0 Page 197 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)IP&ID ISI IST VALVE VLE AT NORM TEST TEST VALVE CIC P&ID COOR CLASS CAT SIZE TYPE TYP POS RQMT FREQ NOTES/DESCRIPTION HPCI-CV-14CV 2044 H5 2 C 2 CK-P SA C COF CVCM HPCI CONDENSATE PUMP CCF CVCM DISCHARGE TO AUXILIARY____ ___ ____ ___ _ __ __ ____ __ __ _ ___COOLING RETURN, RG-01, TP-01 HPCI-CV-15CV 2044 D8 2 A/C 20 CK-L SA C LJ-i OPB HPCI TURBINE EXHAUST, RG-01 COF CVCM____ ____CCL CVCM HPCI-CV-16CV 2044 F9 2 C 2 CK-P SA C COD CVCM HPCI TURBINE EXHAUST DRAIN CCL CVCM TO SUPPRESSION POOL, RG-01,____ ___ ____ ___ ____ _ _ ____ __ ___TP-01 HPCI-CV-17CV 2044 C6 2 C 4 CK-S SA C COF CVCM HPCI PUMP MINIMUM FLOW LINE,___ __ ___ __ __ __ __ _____ __ ____ CCL CVCM RG-01 HPCI-CV-18CV 2044 B8 2 C 2 CK-P SA C COF CVCM CONDENSATE SUPPLY TO HPCI CCF CVCM SYSTEM, RG-01, TP-01 CCD/ CVCM______ _____CCR _ _ _ _ _ _ _ _ _ _ _ _HPCI-CV-19CV 2044 B8 A C 2 CK-P SA C COF CVCM CONDENSATE SUPPLY TO HPCI__________ _____________ ____CCF CVCM SYSTEM, RG-01, TP-01 HPCI-CV-24CV 2044 Eli1 C 3 CK-S SA C FSO SD HPCI VACUUM BREAKER, RG-01,__ __ __ __ __ _ _________ __ __ _ __ _ __ __ FSC SD TP-02 HPCI-CV-25CV 2044 El0 C 3 CK-S SA C FSO SD HPCI VACUUM BREAKER, RG-01,__ __ __ __ __ _ ____ _____ __ __ _FSC SD TP-02 HPCI-CV-26CV 204 Ei 2 C 3 CK-S SA C FSO SD HPCI VACUUM BREAKER, RG-01, FSC SD TP-02 Revision 0 Page 198 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)[IP&ID 1ISI IST VALVE [VALVE 1AC T NORM TEST TEST1 VALVE CIC _P&ID COOR jCLASS CAT SIZE TYPE jTYPE jPOS jRQMT FREQ NOTES/DESCRIPTION HPCI-CV-27CV 2044 El0 2 C 3 CK-S SA C ESO SD HPCI VACUUM BREAKER, RG-01,__________ ____ ____FSC SD TP-02 HPCI-CV-29CV 2044 B9 .1 A/C 14 CK-S SA C LJ-1 RF INJECTION CHECK VALVE, ROJ-09, FSO RE RG-01 FSC RE___PIT 2Y HPCI-HOV-HOV10 2041 H5 2 B 10 STOP HO C SKID Q TU STOP V, RG-01, TP-05 HPCI-MOV-MO014 2041 H5 2 B 10 GT MO C FSO Q STEAM SUPPLY TO TURBINE, RG-_______PIT 2Y 01 HPCI-MOV-MO015 2041 D5 1 A 10 GT MO 0 J-i1 OPB STEAM SUPPLY INBOARD FSC Q/CS ISOLATION, CSJ-10, RG-01_____ _____ ___ _ ______ ______PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _HPCI-MOV-MO16 2041 D4 1 A 10 GT MO 0 LJ-i OPB STEAM SUPPLY OUTBOARD FSC Q ISOLATION, RG-01____PIT 2Y HPCI-MOV-MO17 2044 J2 2 B 16 GT MO 0 FSO Q PUMP SUCTION FROM FSC Q EMERGENCY CONDENSATE __________PIT 2Y STORAGE TANK, RG-01 HPCI-MOV-MO19 2044 B8 2 B 14 GT MO C FSO Q HPCI INJECTION, RG-01__ __ __ ____ __PIT 2Y HPCI-MOV-MO20 2044 B7 2 B 14 GT MO 0 PIT 2Y HPCI PUMP DISCHARGE, RG-01 Revision 0 Page 199 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: HIGH PRESSURE COOLANT INJECTION (IIPC1)P&ID SI 15 IST VALVE VALVE IACT NORM TEST TEST VALVE CIC [P&IDJCOOR _CLASS CAT SIZE TYP__JTYPE P0OS RQMT FREQ NOTES/DESCRIPTION HPCI-MOV-MO21 2044 D2 2 B 10 GB MO C PIT 2Y HPCI PUMP TEST BYPASS TO EMERGENCY CONDENSATE ____ _____ ___ _____ ____ _____STORAGE TANK, PASSIVE, RG-01 HPCI-MOV-MO24 2044 D2 A B 10 GT MO C PIT 2Y HPCI PUMP TEST BYPASS REDUNDANT SHUTOFF, PASSIVE,_____RG-01 HPCI-MOV-MO25 2044 C7 2 B 4 GB MO C FSO Q HPCI PUMP MINIMUM FLOW FSC Q BYPASS LINE ISOLATION, RG-01_____________ _____ ______PIT 2Y_ _ _ _ _ _ _ _ _ _ _ _ _HPCI-MOV-MO58 2044 G10 2 B 16 GT MO C FSO Q HPCI PUMP SUCTION FROM FSC Q SUPPRESSION POOL, RG-01_____ ~~~~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _HPCI-RD-S241 2044 D5 2 D 16 RD SA C RD 5Y EXHAUST LINE RUPTURE DISK,____ ____ ___ ____ __ __ _ ___RG-01 HPCI-RV-10RV 2044 J2 2 C 1 RV SA C RVT APP I HPCI PUMP SUCTION RELIEF, RG-____ ___ ___ ___ __ ___ __ _ ___ __ ___ ___ ___ ___01 HPCI-RV-12RV 2044 H4 2 C 1 RV SA C RVT APP I HPCI AUXILIARY COOLING__________ ____ ____WATER SUPPLY, RG-01 HIPCI-SOV-SSV64 2044 G5 2 B 1 SOV SOV C FSC Q/PB HPCI EXHAUST DRIP LEG DRAIN,_____________ _ ____FST Q RV-01, RG-01, TP.-04 HPCI-SOV-SSV87 2044 G4 2 B 1 SOV SOV C FSC Q/PB HPCI TURBINE DRIP LEG DRAIN,__ _ __ _ __ _ ___ ____ _ __ _ __ _ __ _ ____ _F_ ST Q RG-01, RV-01, TP-04 Revision 0 Page 200 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)IP&ID ISI I ST VALVE VALVE ACT NORM TEST 1TEST VALVE CIC P&IDj COOR CLASS jCAT _SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION HPCI-TU-VGR 2041 H5 A B 9 CTL HO C SKID Q HPCI TU V GEAR ASSY W / PILOT,____ ___ ____ ___ ___ __ _ ___ _ __ ___ _ _ ____ ___ ___TP-05 HPCI-V-44 2044 D8 2 A/C 20 S-CK MA 0 LU-i OPB HPCI TURBINE EXHAUST TO COF CVCM SUPPRESSION POOL ISOLATION,__________ ___ ____CCL CVCM RG-01 HPCI-V-50 2044 F9 2 C 2 S-CK MA 0 COD CVCM TURBINE DRAIN TO SUPPRESSION CCD CVCM POOL ISOLATION, RG-.01, TP-01 Revision 0 Page 201 Revision 0 Page 201 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves SYSTEM: HEATING AND VENTILATION (HV)VALVE CIC P&ID P&ID 1SI I ST VALVE [VALVE IACT NORM ITEST TEST NOTES/DESCRIPTION COOR CLASSJCAT SIZE [TYPE TYPE POS JRQMT FREQ HV-AOV-257AV 2020 D4 A B 72 IN BTF AO 0 FSC Q REACTOR BUILDING VENTILATION FST Q OUTBOARD SUPPLY, RG-01, TP-04_____ ~~~~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _IIV-AOV-259AV 2020 B 12 A B 72 IN BTF AO 0 FSC Q REACTOR BUILDING VENTILATION FST Q INBOARD EXHAUST, RG-01, TP-04 PIT 2Y __ _ _ _ _ _ _ _ _ _ _ _ _ttIV-AOV-261AV 2020 B12 A B 72 IN BTF AO 0 FSC Q REACTOR BUILDING VENTILATION FST Q INBOARD EXHAUST, RG-01, TP-04__ __ ____ PIT 2Y HV-AOV-263AV 2020 E9 A B 72 IN BTF AO 0 FSC Q REACTOR MG SET 1A FST Q VENTILATION OUTBOARD SUPPLY,____ __________PIT 2Y RG-01, TP-04 HV-AOV-265AV 2020 E9 A B 72 IN BTF AO 0 FSC Q REACTOR MG SET lB FST Q VENTILATION OUTBOARD SUPPLY,____ ____PIT 2Y RG-01, TP-04 IIV-AOV-267AV 2020 Dl1 A B 72 IN BTF AO 0 FSC Q *REACTOR MG SET lA EST Q VENTILATION INBOARD EXHAUST,_________ ____ PIT 2Y RG-0l, TP-04 IIV-AOV-269AV 2020 El11 A B 72 IN BTF AO 0 FSC Q REACTOR MG SET lB FST Q VENTILATION OUTBOARD_____________PIT 2Y EXHAUST, RG-01, TP-04 Revision 0 Page 202 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: HEATING AND VENTILATION (HV)VALVE GIGC P&ID P&ID ISI IST VALVE VALVE jACT NORM TEST TEST NOTES/DESCRIPTION _______ICOOR CLASS CAT SIZE TYPE TYPE P0S RQMT FREQ ____________ HV-AOV-270AV 2019 B2 A B 20 IN BTF AO 0 FSC Q CONTROL RM HVAC INLET VALVE,________SH 1___ FST Q RG-01, TP-04 HV-AOV-271AV 2019 Al A B 12 IN BTF AO C FSO Q CONTROL RM HVAC EMERGENCY_________SHi 1______ FST Q BYPASS, RG-01, TP-04 HV-AOV-272AV 2019 B6 A B 8 IN BTF AO 0 FSC Q CONTROL RM PANTRY EXHAUST SH 1 ______ __ __ FST Q ISOLATION, RG-01, TP-04 HV-MOV-258MV 2020 B 12 A B 48 IN BTF MO 0 FSC Q REACTOR BUILDING VENTILATION __________________________PIT 2Y OUTBOARD EXHAUST, RG-01 HV-MOV-260MV 2020 B12 A B 48 IN BTF MO 0 FSC Q REACTOR BUILDING VENTILATION __________ ___ ___ ____________PIT 2Y OUTBOARD EXHAUST, RG-01 HV-MOV-262MV 2020 El0 A B 48 IN BTF MO 0 FSC CS RR MG SET 1A VENTILATION __________ ___ ____PIT 2Y INBOARD SUPPLY, CSJ-08, RG-01 HV-MOV-264MV 2020 El0 A B 48 IN BTF MO 0 FSC CS RR MG SET 1B VENTILATION __________ ___ ____PIT 2Y INBOARD SUPPLY, CSJ-08, RG-01 HV-MOV-266MV 2020 Dl1 A B 48 IN BTF MO 0 FSC CS RR MG SET 1A VENTILATION PIT 2Y OUTBOARD EXHAUST. CSJ-08, RG-____ ___ ___ ___ __ ___ ___ ____ __ ___ ___ ___ ___ ___ ___ 01 HV-MOV-268MV 2020 El11 A B 48 IN BTF MO 0 FSC CS RR MG SET lB VENTILATION PIT 2Y OUTBOARD EXHAUST. CSJ-08, RG-____ __ ___ ___ ___ ___ _______ ___ ___ ___ ___ 01 HV-MOV-272MV 2020 D4 A B 72 IN BTF MO 0 FSC Q REACTOR BUILDING VENTILATION PIT 2Y INBOARD SUPPLY, RG-0l Revision 0 Pg 0 Page 203 Cooper Nuclear Station Ffifh Interval lnservice Testing Pro gram for Pumps and Valves SYSTEM: INSTRUMENT AIR (IA)VALVE CIC P&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ ___________ IA-CV-17CV 2010 E8 A A/C 1/2A CK-L SA 0 LT-1 2Y MS-RV-71A ACCUMULATOR SH 2 FSC RF SUPPLY, ROJ-07, RG-01, TP-01___ __ __ ___ __ __ ___ _ __ _FSO RF IA-CV-I 8CV 2010 E8 A A/C 1/2A CK-L SA 0 LT-1 2Y MS-RV-71B ACCUMULATOR SH 2 FSC RE SUPPLY, ROJ-07, RG-01, TP-01______ _________________ _____ ____ FSO REF_ _ _ _ __ _ _ _ _ _IA-CV- 19CV 2010 D8 A A/C 1/2 CK-L SA 0 LT- 1 2Y MS-RV-71C ACCUMULATOR SH 2 FSC RE SUPPLY, ROJ-07, RG-01, TP-01___ __ __ __ _ ___ __ ____ __ FSO RE IA-CV-20CV 2010 D9 A A/C 1/2A CK-L SA 0 LT-1 2Y MS-RV-71E ACCUMULATOR SH 2 _FSC RE SUPPLY, ROJ-07, RG-01, TP-01______ ______ __________ _____FSO REF_ _ _ _ _ _ _ _ _ _ _IA-CV-21CV 2010 E9 A A/C 1/2A CK-L SA 0 LT-1 2Y MS-RV-71G ACCUMULATOR SH 2 FSC RE SUPPLY, ROJ-07, RG-01, TP-01___ ___ __ __ __ _ _ __ __ _____ _ FSO RE IA-CV-22CV 2010 E9 A A/C 1/2A CK-L SA 0 LT-1 2Y MS-RV-71H ACCUMULATOR SH 2 FSC RE SUPPLY, ROJ-07, RG-01, TP-01_____________ _____ FSO REF_ _ _ _ __ _ _ _ _ _IA-CV-28CV 2010 F9 A A/C 1/2A CK-L SA 0 LT-1 2Y MSIV-AO8OA SUPPLY SH 2 FSC RE ROJ-08, RG-01, TP-01________ ____ __ __ ___________FSO REF_ _ _ _ __ _ _ _ _ _Revision 0 Page 204 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: INSTRUMENT AIR (IA)VALVE CIC P&ID P&ID ISI IST VALVE VALVE [ACT NORM TEST TEST [ NOTES/DESCRIPTION ___COOR] CLASS CAT SIZE TYPE [TYPE POS RQMT FREQ IA-CV-29CV 2010 F8 A A/C 1/2A CK-L SA 0 LT-1 2Y MSIV-AO8OB SUPPLY SH 2 FSC RF ROJ-08, RG-01, TP-01_____ _____ __ __ __ __ __ __ ___ _ ____ ____ ____ FSO RE _ _ _ _ _ _ _ _ _ _ _IA-CV-30CV 2010 F8 A A/C 1/2A CK-L SA 0 LT-1 2Y MSIV-AO8OC SUPPLY SH 2 FSC RF ROJ-08, RG-01, TP-01___ __ ___ __ __ __ ___ _ ___ __ ___ _ ___ __ ___ __ __ FSO RE IA-CV-31CV 2010 F9 A A/C 1/2A CK-L SA 0 LT- 1 2Y MSIV-AO8OD SUPPLY SH 2 FSC RE ROJ-08, RG-01, TP-01___ ___FSO RE IA-CV-32CV 2010 D12 A A/C 1/2A CK-L SA 0 LT-1 2Y MSIV-AO86A SUPPLY SH 2 FSC RE ROJ-08, RG-01, TP-01__ ____ __ FSO RE IA-CV-33CV 2010 D12 A A/C 1/2A CK-L SA 0 LT-1 2Y MSIV-AO86B SUPPLY SH 2 FSC RE ROJ-08, RG-01, TP-01___ __ __ ___ _ __ __ ____ __ FSO RE IA-CV-34CV 2010 E12 A A/C 1/2A CK-L SA 0 LT-1 2Y MSIV-AO86C SUPPLY SH 2 FSC RE ROJ-08, RG-01, TP-01___ _ __ _ __ _ _ __ _ __ _ FSO RE IA-CV-35CV 2010 E12 A A/C 1/2A CK-L SA 0 LT-1 2Y MSIV-AO86D SUPPLY SH 2 FSC RE ROJ-08, RG.-01, TP-01___FSO RE IA-CV-36CV 2010 E8 A A/ 1/2' CK-L OA LT- 1 2Y MS-RV-71D ACCUMULATOR SH 2 FSC RE SUPPLY, ROJ-07, RG-01, TP-01 FSO RE Revision 0 Page 205 Cooper Nuclear Station Ffifh Interval lnservice Testing Program for Pumps and Valves SYSTEM: INSTRUMENT AIR (IA)VALVE CIC P&ID jP&ID IS 1ST VALVE VALVE ACT NORM ITEST TEST NOTES/DESCRIPTION __COOR_____CLASS___ CAT SIZE ITYPE TYPEj POS jRQMT FREQ ____________ IA-CV-37CV 2010 El0 A A/C 1A CK-.L SA 0 LT-1 2Y MS-RV-71F ACCUMULATOR SIT 2 FSC RF SUPPLY, ROJ-07, RG-01, TP-01__ ____ _ _ __ _FSO RF IA-CV-50CV 2010 G5 A A/C 1/4 CK-L SA 0 LT-1 2Y ACCUMULATOR AO-82 IA CHECK SH 2 CCL CVCM VALVE, RG-01, TP-01____ ____COF CVCM IA-CV-51CV 2010 G5 A A/C 1/4 CK-L SA 0 LT-1 2Y ACCUMULATOR AO-83 IA CHECK SH 2 CCL CVCM VALVE, RG-01, TP-01____ ___ ____ ____COF CVCM IA-CV-52CV 2010 G6 *A A/C 1/4 CK-L SA 0 LT-1 2Y ACCUMULATOR AO94 IA CHECK SH 2 CCL CVCM VALVE, RG-01, TP-.01____ ____COF CVCM IA-CV-53CV 2010 G6 A A/C 1/4 CK-L SA 0 LT-1 2Y ACCUMULATOR AO95 IA CHECK SH 2 CCL CVCM VALVE, RG-01, TP-01____ ____ ____COF CVCM IA-CV-54CV 2010 A7 A A/C 1/4 CK-L SA 0 LT-1 2Y HV-AOV-257AV ACCUMULATOR SH 2 CCL CVCM CHECK, RG.-01, TP-01____ ____COF CVCM IA-CV-55CV 2010 A6 A A/C 1/4 CK-L SA 0 LT-1 2Y HV-AOV-259AV ACCUMULATOR SH 2 CCL CVCM CHECK, RG-01, TP-01____ ___ ____COF CVCM Revision 0 Page206 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: INSTRUMENT AIR (IA)VALVE CIC P&ID P&ID ISI 1ST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS C AT SIZE TYPE TYPE POS RQMT FREQ ___________ IA-CV-56CV 2010 B9 A A/C 1/4 CK-L SA 0 LT-1 2Y HV-AOV-261AV ACCUMULATOR SH2 CL CVM CHECKRJ1, RG-01, TP-O1 IA-CV-57CV 2010 B9 A A/C 1/4 CK-L SA 0 LT-1 2Y HV-AOV-263AV ACCUMULATOR SH 2 FSC RF CHECK, ROJ-13, RG-01, TP-01___ __ __ ___ __ __ ___ _ __ __ __ __FSO RE IA-CV-58CV 2010 B9O A A/C 1/4 CK-L SA 0 LT-1 2Y IIV-AOV-265AV ACCUMULATOR Sf12 FSC RE CHECK, ROJ-13, RG-01, TP-01_____ _____ _ ___ _ __ __ __ ___ _ ___ _ ___ FSO REF_ _ _ _ __ _ _ _ _ _IA-CV-59CV 2010 B80 A A/C 1/4 CK-L SA 0 LT-1 2Y HV-AOV-267AV ACCUMULATOR SH 2 FSC RE CHECK, ROJ-13, RG-01, TP-01___ __ __ __ _ ___ __ __ _ __ _FSO RE IA-CV-60CV 2010 F8 2 A/C 2/ CK-L SA 0 LT-1 2PY XV-22I2BOARV IOATIONLATRG SH 2 CFS CVCMKRJ13 G01 P0 IA-CV-65CV 2010 F8 2 A/C 2 CK-L SA 0 LJ-1 OPB X-22 OUTBOARD ISOLATION,RG0 SH 2 COF CVCM RG0___ __ ___ __ __ __ ___ __ ___ _____ __ CCL CVCM IA-CV- 111CV 2010 Ei A AC 1/4 CK-L OA LT-1 2Y IA-271AV ACCUMULATOR Sf11 CCL CVCM CHECK, RG-01, TP-01 COF CVCM Revision 0 Page 207 Cooper Nuclear Station Fifth Interval lnservice Testing Program for Pumps and Valves SYSTEM: INSTRUMENT AIR (IA)VALVE CIC P&ID IP&ID 1S1 J ST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION PC-V-559 2028 C8 2 A 1 GB MA C LI-i OPB PASSIVE MANUAL ISOLATION PC-V-560 2028 D8 2 A 1 GB MA C LI-i OPB PASSIVE MANUAL ISOLATION PC-V-561 2028 C8 2 A 1 GB MA C LJ-i OPB PASSIVE MANUAL ISOLATION PC-V-562 2028 C8 2 A 1 GT MA C LI-i OPB PASSIVE MANUAL ISOLATION PC-V-563 2028 C8 2 A 1 GT MA C LI-i OPB PASSIVE MANUAL ISOLATION PC-V-564 2028 C8 2 A 1 GT MA C LJ-1 OPB PASSIVE MANUAL ISOLATION PC-V-565 2028 D8 2 A 1 ' GT MA C LI-i OPB PASSIVE MANUAL ISOLATION PC-V-566 2028 D8 2 A 1 GT MA C LJ-i OPB PASSIVE MANUAL ISOLATION PC-V-569 2027 17 2 A 1 GT MA C LI-i OPB PASSIVE MANUAL ISOLATION SHl 1_ __ ___ _ _PC-V-570 2027 17 2 A 1/2A GB MA C LI-i OPB PASSIVE MANUAL ISOLATION Sil 1__PC-V-571 2027 17 2 A 1 GB MA C LI-i OPB PASSIVE MANUAL ISOLATION Sil __ __ ___ _ _PC-V-572 2027 17 2 A 1/2A GB MA C LI-i OPB PASSIVE MANUAL ISOLATION SHi 1_ _ _ _PC-V-573 2027 17 2 A 1 GT MA C LI-i OPB PASSIVE MANUAL ISOLATION___________ Sil 1 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _Rev&ion 0 Page 208 Cooper Nuclear Station Fifth Interval Inserviee Testing Program for Pumps and Valves SYSTEM: INSTRUMENT AIR (IA)VALVE CIC P&ID P&ID IS1I IST 1VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR _CLASS CAT SIZE TYPE TYPE POS RQMT FREQ ____________ PC-V-574 2027 J7 2 A 1/2A GB MA C LJ-1 OPB PASSIVE MANUAL ISOLATION___________ SHi 1 _ __ _ _ _ _ _ _ _PC-V-575 2027 J7 2 A 1 GB MA C LU-1 OPB PASSIVE MANUAL ISOLATION________________ SHl_ 1 ___ ________PC-V-576 2027 J7 2 A 1/2A GB MA C LJ-i OPB PASSIVE MANUAL ISOLATION Sil 1_ _ _ __ _ ___ _ _ _ _ _ _ _ _PC-V-577 2027 19 2 A 1 GB MA C LJ-1 OPB PASSIVE MANUAL ISOLATION Sil 1__PC-V-578 2027 J9 2 A 1/2A GB MA C LI-1 OPB PASSIVE MANUAL ISOLATION Sil 1 _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _PC-V-579 2027 J9 2 A 1 GB MA C LJ-i OPB PASSIVE MANUAL ISOLATION___________ _ Sil_ 1 _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _PC-V-580 2027 J9 2 A 1/2/ GB MA C LJ-i OPB PASSIVE MANUAL ISOLATION Sil 1 __PC-V-581 2027 J9 2 A 1 GB MA C LJ-i OPB PASSIVE MANUAL ISOLATION SH 2 __ ___ ___ _ __ _ _PC-V-582 2027 J3 2 A 'A GB MA C LI-i OPB PASSIVE MANUAL ISOLATION________ ___ 5SH 2 _ __ _PC-V-583 2027 13 2 A 1 GB MA C LI-1 OPB PASSIVE MANUAL ISOLATION____________ S11 2 _ __ _ _ _ __ _ _ _ _ _ _ _ _PC-V-584 2027 13 2 A 1/2 GB MA C LJ-i OPB PASSIVE MANUAL ISOLATION SH 2 Revision 0 Page 209 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves[SYSTEM: INSTRUMENT AIR (IA)VALVE CIC IP&ID P&ID 1SI 1ST VALVE VALVE ACT NORM TEST [TEST NOTES/DESCRIPTION PC-V-585 2027 J4 2 A 1 GB MA C LJ-1 OPB PASSIVE MANUAL ISOLATION PC-V-586 2027 J4 2 A 1/2A GB MA C LJ-i OPB PASSIVE MANUAL ISOLATION SH2 2 _ __ _ _ _PC-V-587 2027 J4 2 A 1 GB MA C LJ-1 OPB PASSIVE MANUAL ISOLATION SH 2 ____ ____ ____ ____PC-V-588 2027 J4 2 A 1/2A GB MA C LJ-1 OPB PASSIVE MANUAL ISOLATION SH2 2_ __ __ _ _PC-V-589 2027 J5 2 A 1 GB MA C LJ-1 OPB PASSIVE MANUAL ISOLATION___________ S_ 5 2 _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _PC-V-590 2027 J5 2 A 1/2A GB MA C LJ-1i OPB PAssIVE MANUAL ISOLATION S112 Revision 0 Page 210 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: MAIN STEAM (MS)P&ID ISI IST VALVE VALVE ACT NORM TEST TEST VALVE CIC P&ID COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION MS-AOV-A0SOA 2041 B5 1 A 24 GB AO 0 LJ-1 RE MSIV A INBOARD ISOLATION PSC Q CSJ-03, RG-01 FSC CS FST CS___PIT 2Y MS-AOV-AO80B 2041 B5 I A 24 GB AO 0 UJ-1 RE MSIV B INBOARD ISOLATION PSC Q CSJ-03, RG-01 FSC CS FST CS___PIT 2Y MS-AOV-AO80C 2041 B7 1 A 24 GB AO 0 LJ-1 RE MSIV C INBOARD ISOLATION PSC Q CSJ-03, RG-01 FSC CS FST CS______ ~~~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _MS-AOV-AO80D 2041 B7 1 A 24 GB AO 0 LJ-i RE MSIV D INBOARD ISOLATION PSC Q CSJ-03, RG-01 FSC CS FST CS PIT 2Y __ _ _ _ _ _ _ _ _ _ _ _ _Revision 0 Page 211 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: MAIN STEAM (MS)IP&ID ISI] IST VALVE IVALVE ACT INORM] TEST 1TEST VALVE CIC P&ID COOR CLASS CAT SIZE jTYPE TYPEj POS R'QMT JFREQ NOTES/DESCRIPTION MS-AOV-AO86A 2041 B4 1 A 24 GB AO 0 LJ-i RF MSIV A OUTBOARD ISOLATION PSC Q CSJ-03, RG-01 FSC CS FST CS____PIT 2Y MS-AOV-AO86B 2041 A4 1 A 24 GB AO 0 LJ-i RF MSIV B OUTBOARD ISOLATION PSC Q CSJ-03, RG-01 FSC CS FST CS__ __ __ __ _ __ _PIT 2Y MS-AOV-AO86C 2041 A8 1 A 24 GB AO 0 UJ-1 RF MSIV C OUTBOARD ISOLATION PSC Q CSJ-03, RG-01 FSC CS FST CS___PIT 2Y MS-AOV-AO86D 2041 B8 1 A 24 GB AO 0 LJ-i RF MSIV D OUTBOARD ISOLATION PSC Q CSJ-03, RG-01 FSC CS FST CS__ __ _ __ _PIT 2Y MS-AOV-738AV 2028 C9 1 B 1/2A GB AO C PIT 2Y RPV HEAD VENT DRAIN TO SUMP,_____ _____PASSIVE ISOLATION, RG-01 MS-AOV-739AV 2028 C9 1 B 1/2A GB AO C PIT 2Y RPV HEAD VENT DRAIN TO SUMP,____ ___ ___ __ _ ___ ___ _ __ ___ ___ _ __ ___ ___ __ ____ __ ___PASSIVE ISOLATION, RG-01 Revision 0 Page 212 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: MAIN STEAM (MS)P&ID ISI IST 1VALVE 1VALVE ACT 1NORM TEST ITEST VALVE CIC P&ID COOR CLASS CAT jSIZE TYPE TYPE jPOS RQMT jFREQ NOTES/DESCRIPTION MS-CV-20CV 2028 Cl11 C 10 CK-S SA C VBT RF 71A RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 Al, RG-01, TP-08___FSC 2Y MS-CV-21CV 2028 Cll C 10 CK-S SA C VBT RE 71A RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 A2, RG-01I, TP-08___ ____ __FSC 2Y MS-CV-22CV 2028 C11 C 10 CK-S SA C VBT RF 71B RV DISCHARGE VACUUM FSO 2Y BREAKER VB 7iBl,RG-0i, TP-08___ _ __ ____ __FSC 2Y MS-CV-23CV 2028 Ci11 C 10 CK-S SA C VBT RF 71B RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 B2, RG-01, TP-08___ __ ___ __ _ _____ _____ __FSC 2Y MS-CV-24CV 2028 Dii1 C 10 CK-S SA C VBT RF 71C RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 C1, RG-01, TP-08___ ____ __FSC 2Y MS-CV-25CV 2028 Di 31 C 10 CK-S SA C VBT RF 71C RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 C2, RG-01, TP-08___ _ __ ____ __FSC 2Y MS-CV-26CV 2028 D11 C 10 CK-S SA C VBT RE 71D RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71D1, RG-01, TP-08___ _ __ ____ __FSC 2Y MS-CV-27CV 208 Di 3 C 10 CK-S SA C VBT RE 71D RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 D2, RG-01, TP-08 FSC 2Y Revision 0 Page 213 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: MAIN STEAM (MS)P&ID IISI IST VALVE VALVE ACT 1NORM TEST 1TES T VALVE CIC [P&ID COOR JCLASS CAT SIZE TYPE TYPE jPOS RQMT FRLEQ JNOTES/DESCRI1PTION MS-CV-28CV 2028 D9 3 C 10 CK-S SA C VBT RE 71E RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 El,RG-01, TP-08___ ___ __ ___ __ __ __ __FSC 2Y MS-CV-29CV 2028 D9 3 C 10 CK-S SA C VBT RF 71E RV DISCHARGE VACUUMI~FSO 2Y BREAKER VB 71 E2, RG-01, TP-08___ _ __ ____ __FSC 2Y MS-CV-30CV 2028 D9 3 C 10 CK-S SA C VBT RF 71F RV DISCHARGE VACUUM FSO 2Y BREAKER VB71 F1, RG-01, TP-08_____ ______ _____ ~ FSC 2Y _ _ _ _ _ _ _ _ _ _ _ _ _MS-CV-31CV 2028 D9 3 C 10 CK-S SA C VBT RF 71F RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 F2, RG-01, TP-08____ ____FSC 2Y MS-CV-32CV 2028 E9 3 C 10 CK-S SA C "VBT RF 71G RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 G1, RG-01, TP-08___ _ __ ____ __FSC 2Y MS-CV-33CV 2028 E9 3 C 10 CK-S SA C VBT RF 71G RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 G2, RG-01, TP-08___ _ __ ____ __FSC 2Y MS-CV-34CV 2028 E9 3 C 10 CK-S SA C VBT RF 71H RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 HI,RG-01, TP-08____FSC 2Y Revision 0 Page 214 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: MAIN STEAM (MS)IP&ID IISI 1ST VALVE VALVE ACT NORM TEST TESTJ VALVE CIC _P&ID COOR _CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION MS-CV-35CV 2028 E9 3 C 10 CK-S SA C VBT RF 7111 RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 112, RG-01, TP-08___FSC 2Y MS-MOV-MO74 2041 C7 1 A 3 GT MO 0 LJ-1 OPB MS L1NE DRAIN INBOARD FSC Q/CS ISOLATION, CSJ-10, RG-01____ ___PIT 2Y MS-MOV-MO77 2041 C11 A 3 GT MO 0/C LU-1 OPB MS LINE DRAIN OUTBOARD FSC Q ISOLATION, RG-01______ ~~~~PIT 2Y __ _ _ _ _ _ _ _ _ _ _ _ _MS-RV-70ARV 2028 C10 C 6 RV SA C RVT APP I SAFETY VALVE MS LINE A, RV-02,______ _____RG-01 MS-RV-70BRV 2028 E9 1 C 6 RV SA C RVT APP I SAFETY VALVE MS LINE D, RV-02,____ ___ __ _ ___RG-01 MS-RV-70CRV 2028 E9 1 C 6 RV SA C RVT APP I SAFETY VALVE MS LINE D, RV-02,____ ___ ____ __ ___RG-01 MS-RV-71ARV 2028 Cli1 B/C 6 RV AO C RVT APP I SAFETY RELIEF VALVE MS LINE A, FSO RF RV-03, RG-01 FSC RF___ __ __ ____ __PIT RF MS-RV-71BRV 208 Ci 1 BC 6 RV AO C RVT APP I SAFETY RELIEF VALVE MS LINE A, FSO RF RV-03, RG-01 FSC RF PIT RF Revision 0 Page 215 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: MAIN STEAM (MS)P&ID ISI 1ST VALVE VALVE ACT NORM TEST TEST VALVE CIC P&ID [COOR CLASS CAT SIZE TYPE TYPE POS jRQMT FREQ NOTES/DESCRIPTION MS-RV-71CRV 2028 D11 B/C 6 RV AO C RVT APP I SAFETY RELIEF VALVE MS LINE B, FSO RE RV-03, RG-01 FSC RE_____ _____ ~~PIT RF _ _ _ _ _ _ _ _ _ _ _ _ _MS-RV-71DRV 2028 DI11 B/C 6 RV AO C RVT APP I SAFETY RELIEF VALVE MS LINE B, FSO RE RV-03, RG-01 FSC RE PIT RE MS-RV-71ERV 2028 D9 1 B/C 6 RV AO C RVT APP I SAFETY RELIEF VALVE MS LINE C, FSO RE RV-03, RG-01 FSC RE__ __ __ _PIT RE MS-RV-71FRV 2028 D9 1 B/C 6 RV AO C RVT APP I SAFETY RELIEF VALVE MS LINE C, FSO RE RV-03, RG-01 FSC RE___ _ __ __PIT RE MS-RV-71GRV 2028 E9 1 B/C 6 RV AO C RVT APP I SAFETY RELIEF VALVE MS LINE D, FSO RE RV-03, RG-01 FSC RE___ _ _ ___PIT RE MS-RV-71HRV 2028 E9 1 B/C 6 RV AO C RVT APP I SAFETY RELIEF VALVE MS LINE D, FSO RE RV-03, RG-01 FSC RE PIT RE Revision 0 Page216 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: NEUTRON MONITORING TRAVERSING INCORE PROBE (NMT)VALVE CIC P&ID P&ID ISI IST VALVE IVALVE ACT NORM TEST TEST NOTES/DESCRIPTION ________jCOOR CLASS jCAT SIZE jTYPE TYPE _POSj RQMT FREQ NM-CV-CV2 2083 J2 2 A!C 3/8 CK-P SA C LU-1 OPB N2 PURGE ISOLATION, RG-01, COF CVCM TP-01, TP-06___ ___ __ ___ _ ___ ___ _______ CCL CVCM NM-CV-CV4 2083 J2 2 A/C 3/8 CK-P SA C LU-i OPB N2 PURGE ISOLATION, RG-01, COF CVCM TP-01, TP-06___________ ____ __ __CCL CVCM NMT-NVA-104AX 2083 J2 2 D 3/8 SHR EX 0 EX 2Y TIP A SHEAR VALVE, RG-01 NMT-NVA-104BX 2083 J2 2 D 3/8 SHR EX 0 EX 2Y TIP B SHEAR VALVE, RG-01 NMT-NVA-104CX 2083 J2 2 D 3/8 SHR EX 0 EX 2Y TIP C SHEAR VALVE, RG-01 NMT-NVA-104DX 2083 J2 2 D 3/8 SHR EX 0 EX 2Y TIP D SHEAR VALVE,, RG-01 NMT-NVA-104A 2083 J2 2 A 3/8 BAL 50 0 LJ-1 OPB TIP A BALL VALVE, RG-01, FSC Q TP-04 FST Q__ __ __ __ _ __ _PIT 2Y NMT-NVA-104B 2083 J2 2 A 3/8 BAL 50 0 LJ-i OPB TIP BBALL VALVE, RG-0i, FSC Q TP-04 FST Q PIT 2Y Revision 0 Page 217 Cooper Nuclear Station F~fih Interval Inservice Testing Program for Pumps and Valves SYSTEM: NEUTRON MONITORING TRAVERSING INCORE PROBE (NMT)VALVE CIC IP&ID P&ID IISI IST 1VALVE VALVE ACT NORM TEST 1TEST NOTES/DESCRIPTION ___________I____COOR _C LASS CAT SIZE TYPE TYPE POS RQMT] FREQ ____________ NMT-NVA-104C 2083 J2 2 A 3/8 BAL S0 0 LU-i OPB TIP C BALL VALVE, FSC Q RG-01, TP-04 FST Q______ _____ ~PIT 2Y __ _ _ _ _ _ _ _ _ _ _ _NMT-NVA-104D 2083 J2 2 A 3/8 BAL S0 0 LJ-i OPB TIP D BALL VALVE, FSC Q RG-01, TP-04 FST Q PIT 2Y Revision 0 Page 218 Revision 0 Page 218 Cooper Nuclear Station F~fih Interval Inservice Testing Program for Pumps and Valves SYSTEM: NUCLEAR BOILER INSTRUMENTATION (NBI) AND VARIOUS OTHER EXCESS FLOW CHECK VALVES VALVE CIC P&ID P&ID ISI IST 1VALVE VALVE ACT 1NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT jSIZE TYPE TYPE POS RQMT jFREQ ____________ NBI-AOV-736AV 2028 C9 1 B 1/2A GB AO 0 PIT 2Y RPV FLANGE LEAKOFF PASSIVE_____ISOLATION, RG-01 NBI-AOV-737AV 2028 C9 1 B 1/2A GB AO C PIT 2Y RPV FLANGE LEAKOFF PASSIVE ISOLATION, RG-01 NBI-CV-10BCV 2026 B5 1 C 1 CK-B SA 0 CCL CVCM UPPER CHECK FOR SH 1___ __ COF CVCM LI-61, RG-01, TP-01 NBI-CV-1 1BCV 2026 B5 1 C 1 CK-B SA 0 CCL CVCM CONDENSING CHAMBER 2A UPPER________Sil 1__ COF CVCM TO 25-5 RACK, RG-01, TP-01 NBI-CV-12BCV 2026 C5 1 C 1 CK-B SA 0 CCL CVCM CONDENSING CHAMBER 2A Sil ____1__ COF CVCM LOWER TO 25-5 RACK, RG-01, TP-01 NBI-CV-13BCV 2026 C5 1 C 1 CK-B SA 0 CCL CVCM CONDENSING CHAMBER 3A TO SH 1___ __ COF CVCM LITS-73A, RG-01, TP-01 NBI-CV-14BCV 2026 D5 1 C 1 CK-B SA 0 CCL CVCM CONDENSING CHAMBER 3A TO 25-Sil 1__ COF CVCM 5 & 25-5-1 RACK, RG-01, TP-01 NBI-CV-15BCV 2026 D5 1 C 1 CK-B SA 0 CCL CVCM LT-52A; LIS-83A & C; LT-61; DPT-65;Sil 1__ COF CVCM LIS-101A &B, RG-01, TP-01 NBI-CV-16BCV 2026 D5 1 C 1 CK-B SA 0 CCL CVCM LOW SIDE, FT-64T, RG-01, TP-01_______Sil 1______ __ COF CVCM NBI-CV-17BCV 2026 D5 1 C 1 CK-B SA 0 CCL CVCM LOW SIDE FT-64R;__ _ _ _ _ _ _ Sil 1 _ _ ______ ____ ___ _ _ _ COF CVCM FT-63D, RG-01, TP-01 Revision 0 Page 219 Cooper Nuclear Station F~ifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: NUCLEAR BOILER INSTRUMENTATION (NBI) AND VARIOUS OTHER EXCESS FLOW CHECK VALVES VALVE GIG P&ID P&D II IT VLEVLE ACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE jTYPE P05 RQMT FREQ _____________ NBI-CV- 18BCV 2026 E5 1 C 1 CK-B SA 0 CCL CVCM HIGH SIDE FT-63D, RG-01, TP-01__ _ _ _ _ _ _ SHi 1 _ _ _ _ __ _ _ _ _ _ COF CVCM _ _ _ _ _ _ _ _ _ _ _ _ _NBI-CV-19BCV 2026 G5 1 C 1 CK-B SA 0 CCL CVCM ABOVE CORE PLATE PRESS, RG-01,________SHi1 COF CVCM TP-01 NBI-CV-20BCV 2026 F5 1 C 1 CK-B SA 0 CCL CVCM BELOW CORE PLATE TO SHi1 COF CVCM INSTRUMENT RACK 25-51, RG-01,____ ___ ____ ___TP-01 NBI-CV-21BCV 2026 C8 1 C 1 CK-B SA 0 CCL CVCM CONDENSING CHAMBER 2B UPPER________SH 1___ COF CVCM TO 25-6 RACK, RG-01, TP-01 NBI-CV-22BCV 2026 B8 1 C 1 CK-B SA 0 CCL CVCM CONDENSING CHAMBER 2B UPPER________SHi 1__ ______ COF CVCM TO 25-6 RACK, RG-01, TP-01 NBI-CV-23BCV 2026 C8 1 C 1 CK-B SA 0 CCL CVCM CONDENSING CHAMBER 3B LT-70;________SHi 1__ COF CVCM LITS-73B, RG-01, TP-01 NBI-CV-24BCV 2026 D8 1 C 1 CK-B SA 0 CCL CVCM CONDENSING CHAMBER 3B TO 25-________SH 1 _______ ___COF CVCM 6 & 25-6-1 RACKS, RG-01, TP-01 NBI-CV-25BCV 2026 D8 1 C 1 CK-B SA 0 CCL CVCM LOW SIDE LIS-83B;________SHl ___ COF CVCM LT-52B; LIS-IO1C & D, RG-01, TP-01 NBI-CV-26BCV 2026 D8 1 C 1 CK-B SA 0 CCL CVCM LOW SIDE FT-64L, RG-01, TP-01________SHl 1 ___ _____ COF CVCM NBI-CV-27BCV 2026 D8 1 C 1 CK-B SA 0 CCL CVCM LOW SIDE FT-63C;________Sil 1 __ ______ _____ COF CVCM FT-64N, RG-01, TP-01 NBI-CV-28BCV 26 E8 1 C 1 CK-B OA CCL CVCM HIGH SIDE FT-63C, RG-01, TP-01 Revision 0 Page 220 Cooper Nuclear Station F~fih Interval Inservice Testing Program for Pumps and Valves SYSTEM: NUCLEAR BOILER INSTRUMENTATION (NBI) AND VARIOUS OTHER EXCESS FLOW CHECK VALVES VALVE CIC P&ID P&ID ISI IST VALVE VALVE IACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ _____________ SHl 1___ __ COF CVCM _____________ NBI-CV-29BCV 2026 P8 1 C 1 CK-B SA 0 CCL CVCM BELOW CORE PLATE TO Sil1 COF CVCM INSTRUMENT RACK 25-52, RG-01,________TP-0 1 NBI-CV-30BCV 2026 H5 1 C 1 CK-B SA 0 CCL CVCM FT-63B; FT-64B LOW SIDE, RG-01,________Sil 1__ __ COF CVCM TP-01 NBI-CV-31BCV 2026 H5 1 C 1 CK-B SA 0 CCL CVCM FT-63B; HIGH SIDE, RG-01, TP-01 SHi ___1_ _ __ _ _ ___ COP CVCM _ _ _ _ _ _ _ _ _ _ _ _ _NBI-CV-32BCV 2026 H5 1 C 1 CK-B SA 0 CCL CVCM FT-64D; LOW SIDE, RG-01, TP-01_ _ _ _ _ _ _ _ SHl 1_ _ ___ COF CVCM _ _ _ _ _ _ _ _ _ _ _ _ _NBI-CV-33BCV 2026 H5 1 C 1 CK-B SA 0 CCL CVCM FT-64F; LOW SIDE, RG-01, TP-01 5S1 1 ____COF CVCM NBI-CV-34BCV 2026 H5 1 C 1 CK-B SA 0 CCL CVCM FT-64K; LOW SIDE, RG-01, TP-01 SH 1__ ___ COF CVCM _____________ NBI-CV-35BCV 2026 H5 1 C 1 CK-B SA 0 CCL CVCM FT-64M; LOW SIDE, RG-01, TP-0l_ _ _ _ _ _ _ _ Sil 1 __ _ __ _ _ _ _ _ _ _ COF CVCM _ _ _ _ _ _ _ _ _ _ _ _ _NBI-CV-36BCV 2026 115 1 C 1 CK-B SA 0 CCL CVCM FT-64V; LOW SIDE, RG-01, TP-01 Sil 1 _ _ _ _ _ _ _ _ _ _ _ _ _ COP CVCM _ _ _ _ _ _ _ _ _ _ _ _ _NBI-CV-37BCV 2026 115 1 C 1 CK-B SA 0 CCL CVCM FT-64X; LOW SIDE, RG-01, TP-01_________ Sil ___ ____ _ _ COP CVCM _ _ _ _ _ _ _ _ _ _ _ _ _ _NBI-CV-38BCV 26 J5 1 C 1 CK-B OA CCL CVCM FT-64Z; LOW SIDE, RG-01, TP-01 SillCOP CVCM Revision 0 Page221 Cooper Nuclear Station F~ifh lnterval Inservice Testing Program for Pumps and Valves SYSTEM: NUCLEAR BOILER INSTRUMENTATION (NB1) AND VARIOUS OTHER EXCESS FLOW CHECK VALVES VALVE CIC IPP&ID &ID ISI IST VALVE VLE ACT NORM TEST TEST NOTES/DESCRIPTION _______jCOOR CLASS CAT_]_SIZE TYPE_[TYPE POS RQMT FREQ ____________ NBI-CV-39BCV 2026 J5 1 C 1 CK-B SA 0 CCL CVCM FT-64A; FT-63A LOW SIDE, RG-01, SHl1 COF CVCM TP-01 NBI-CV-40BCV 2026 J5 1 C 1 CK-B SA 0 CCL CVCM FT-63A HIGH SIDE, RG-01, TP-01 Sil 1 __ _____________ COF CVCM NBI-CV-41BCV 2026 J5 1 C 1 CK-B SA 0 CCL CVCM FT-64C LOW SIDE, RG-01, TP-01 Sil 1______ COF CVCM NBI-CV-42BCV 2026 J5 1 C 1 CK-B SA 0 CCL CVCM FT-64E LOW SIDE, RG-01, TP-01 Sil 1______ __ COF CVCM NBI-CV-43BCV 2026 J5 1 C 1 CK-B SA O CCL CVCM FT-64J LOW SIDE, RG-01, TP-0l Sil 1__ COF CVCM NBI-CV-44BCV 2026 J5 1 C 1 CK-B SA 0 CCL CVCM FT-64S LOW SIDE, RG-01, TP-0l Sil 1 _ _ __ _ _ _ _ _ _ COF CVCM _ _ _ _ _ _ _ _ _ _ _ _NBI-CV-45BCV 2026 J5 1 C 1 CK-B SA 0 CCL CVCM FT-64U LOW SIDE, RG-01, TP-01 Sil 1______ COF CVCM ____________ NBI-CV-46BCV 2026 J5 1 C 1 CK-B SA 0 CCL CVCM FT-64W LOW SIDE, RG-01, TP-01 Sil 1___ __ COF CVCM NBI-CV-47BCV 2026 J5 1 C 1 CK-B SA 0 CCL CVCM FT-64Y LOW SIDE, RG-01, TP-01 Sil 1__ COF CVCM NBI-CV-48BCV 2028 C8 1 C 1 CK-B SA 0 CCL CVCM VESSEL FLANGE LEAKOFF LINE,__ __ __ __ __ _ _ __ _ __ __ _ __ _ __ _ __ _ COF CVCM RG-01, TP-01 NBI-CV-49BCV 2063 C 1/2/ CK-S SA C FSO RF REFERENCE LEG LOOP A Rev&ion 0 Page 222 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: NUCLEAR BOILER INSTRUMENTATION (NBI) AND VARIOUS OTHER EXCESS FLOW CHECK VALVES VALVE CIC IP&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION _______JCOOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ _____________ Sill C2 ___ ___FSC RE INJECTION, ROJ-0l, RG-01, TP-0l NBI-CV-50BCV 2026 C3 3 C 1/2A CK-S SA C FSO RF REFERENCE LEG LOOP A SH 1 __ ________________ FSC RE INJECTION, ROJ-01, RG-01, TP-01 NBI-CV-51BCV 2026 Cl11 C 1/2A CK-S SA C FSO RF REFERENCE LEG LOOP B________SH 1 __________FSC RE INJECTION, ROJ-01, RG-01, TP-01 NBI-CV-52BCV 2026 Cl11 C 1/2A CK-S SA C FSO RF REFERENCE LEG LOOP B Sil 1___ __ FSC RF INJECTION, ROJ-01, RG-0i, TP-0i NBI-CV-55CV 2026 Gil1 A/C 3/8 CK-S SA 0 LT2 RF REFERENCE LEG LOOP 3A SHll FSC RE OUTBOARD INJECTION, ROJ-02, FSO RE RG-01,TP-01 NBI-CV-56CV 2026 HI 1 3 A/C 3/8 CK-S SA 0 LT2 RE REFERENCE LEG LOOP 3B SH l FSC RE OUTBOARD INJECTION, ROJ-02,_____ ____________F SO RE RG-0l, TP-01 NBI-SOV-SSV738 2026 HI 1 2 B 1/2A SOV SOV C FSO RE REFERENCE LEG LOOP A SHi 1 ST RE INJECTION, ROJ-01, RG-01, TP-04___ _ __ __ __ ____ __PIT 2Y NBI-SOV-SSV739 2026 C12 2 B 1/2A SOV SOV C FSO RE REFERENCE LEG LOOP B SHl 1EST RE INJECTION, ROJ-01, RG-01, TP-04___ __ __ ____ __PIT 2Y CS-CV-16BCV 2045 A8 1 C 1 CK-B SA 0 CCL CVCM DPIS-43A LOW SIDE EXCESS FLOW, SHi 1__ COF CVCM RG-0l, TP-01 CS-CV-l7BCV 24 A8 1 C 1 CK-B OA CCL CVCM DPIS-43B LOW SIDE EXCESS FLOW, Revision 0 Page 223 Cooper Nuclear Station F~ifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: NUCLEAR BOILER INSTRUMENTATION (NBI) AND VARIOUS OTHER EXCESS FLOW CHECK VALVES VALVE CIC P&ID 1P&ID ISI 1IST VALVE VALVE ACT TNORM 1TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ _____________ SH 1 COF CVCM RG-01, TP-01 HPCI-CV-10BCV 2041 E4 1 C 1 CK-B SA 0 CCL CVCM HIGH SIDE HPCI-DPIS-76; -77; PS-___________COF CVCM 68A; PS-68C, RG-01, TP-01 HPCI-CV-1 1BCV 2041 E4 1 C 1 -CK-B SA 0 CCL CVCM LOW SIDE HIPCI-DPIS-76; DPIS-77;___________COF CVCM PS-68B; PS-68D, RG-01, TP-01 MS-CV-10BCV 2041 .C4 1 C 1 CK-B SA 0 CCL CVCM FT-51A; DPIS-116A, B, C, & D HIGH___ __ ___COF CVCM SIDE, RG-01, TP-01 MS-CV-11BCV 2041 C4 1 C 1 CK-B SA 0 CCL CVCM FT-51A; DPIS-116A, B, C,& D LOW____ ___ ________COF CVCM SIDE, RG-01, TP-01 MS-CV-12BCV 2041 C4 1 C 1 CK-B SA 0 CCL CVCM FT-51B; DPIS-117A, B,C, &D LOW____________ ___ ___COF CVCM SIDE, RG-01, TP-01 MS-CV-13BCV 2041 C4 1 C 1 CK-B SA 0 CCL CVCM FT-51B; DPIS--117A, B, C, &D HIGH___ ___COF CVCM SIDE, RG-01, TP-01 MS-CV-14BCV 2041 C8 1 C 1 CK-B SA 0 CCL CVCM FT-51C; DPIS-118A, B, C, & D HIGH________ ____COF CVCM SIDE, RG-01, TP-01 MS-CV-15BCV 2041 C8 1 C 1 CK-B SA 0 CCL CVCM FT-51D; DPIS-119A, B, C, & DHIGH__________ ___ ____COF CVCM SIDE, RG-01, TP-01 MS-CV-16BCV 2041 C8 1 C 1 CK-B SA 0 CCL CVCM FT-51C; DPIS-118A, B, C, & D LOW____ ____ ____ ____COF CVCM SIDE, RG-01, TP-01 MS-CV-17BCV 24 C8 1 C 1 CK-B SA 0 CCL CVCM FT-51D; DPIS-1 19A, B, C, & D LOW COF CVCM SIDE, 'RG-01, TP-01 Revision 0 Page 224 Cooper Nuclear Station F~flh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: NUCLEAR BOILER INSTRUMENTATION (NBI) AND VARIOUS OTHER EXCESS FLOW CHECK VALVES VALVE CIC IP&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION ___________COOR CLASS CAT SIZE TYPEJTYPE POS RQMT FREQ _____________ RCIC-CV-10BCV 2041 E8 1 C 1 CK-B SA 0 CCL CVCM RCIC-DPIS-84 HIGH SIDE EFCV, RG-______________COF CVCM 01, TP-01 RCIC-CV-1 1BCV 2041 E8 1 C 1 CK-B SA 0 CCL CVCM RCIC-DPIS-84 LOW SIDE EXCESS___________ _________COF CVCM FLOW CHECK VALVE, RG-0 1, TP-0 1 RCIC-CV-12BCV 2041 E8 1 C 1 CK-B SA 0 CCL CVCM RCIC-DPIS-83 HIGH SIDE EFCV, RG-___ ___COF CVCM 01, TP-01 RCIC-CV-13BCV 2041 E8 1 C 1 CK-B SA 0 CCL CVCM RCIC-DPIS-83 LOW SIDE EXCESS___________ ___ ____COF CVCM FLOW CHECK VALVE, RG-01, TP-01 RR-CV-10CV 2027 D5 1 C 1 CK-B SA 0 CCL CVCM PS-128A SENSING LINE, RG-01, TP-Sil 1 __ __ ___ __ __ COF CVCM 01 RR-CV-11CV 2027 D5 1 C 1 CK-B SA 0 CCL CVCM PS-128A SENSING LINE, RG-01, TP-Sil 1 __ ______ COF CVCM 01 RR-CV-12CV 2027 C5 1 C 1 CK-B SA 0 CCL CVCM DPT-1 11A LOW SIDE, RG-01, TP-01_ _ _ _ _ _ _ _ Sil 1 _ _ _ _ _ _ __ _ _ COF CVCM _ _ _ _ _ _ _ _ _ _ _ _ _RR-CV-13CV 2027 C5 1 C 1 CK-B SA 0 CCL CVCM DPT-1 11A HIGH SIDE, RG-01, TP-01 SH 1 __ ______ COF CVCM _____________ RR-CV-15CV 2027 F5 1 C 1 CK-B SA 0 CCL CVCM PT-25A SENSING LINE, RG-01, TP-01________ SH 1 _ _ ___ COF CVCM _ _ _ _ _ _ _ _ _ _ _ _ _RR-CV- 16CV 2027 F5 1 C 1 CK-B SA 0 CCL CVCM PT-24A SENSING LINE, RG-01, TP-01________Sill COF CVCM RR-CV-17CV 22 C5 1 C 1 CK-B SA O CCL CVCM -FT-110A AND BHIGH SIDE, RG-01, Revision 0 Page 225 Cooper Nuclear Stajion F~fih Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: NUCLEAR BOILER INSTRUMENTATION (NBI) AND VARIOUS OTHER EXCESS FLOW CHECK VALVES VALVE CIC IP&ID P&ID IISI IST 1VALVE VALVE [ACT NORM TEST TEST NOTES/DESCRIPTION __________COOR _CLASS CAT_]_SIZE TYPE TYPE POS RQMT FREQ________SH 1 __ ___ COF CVCM TP-01 RR-CV-18CV 2027 C5 1 C 1 CK-B SA 0 CCL CVCM FT-110A AND BLOW SIDE, RG-01,__ _ _ _ _ _ _ Sil ___1_ _ __ _ ____ COF CVCM TP-O1 RR-CV-27CV 2027 C6 1 C 1 CK-B SA 0 CCL CVCM DPT-l111B LOW SIDE, RG-0l, TP-01__ _ _ _ _ _ _ SH 2 __ _ _ _ _ _ _COF CVCM _ _ _ _ _ _ _ _ _ _ _ _ _RR-CV-28CV 2027 C6 1 C 1 CK-B SA 0 CCL CVCM DPT-111lB HIGH SIDE, RG-0l, TP-01 SH 2 ____ _ _ __ _ ____ COF CVCM _ _ _ _ _ _ _ _ _ _ _ _ _RR-CV-30CV 2027 G6 1 C 1 CK-B SA 0 CCL CVCM PT-25B REACTOR RECIRC PUMP lB SH 2 __________COF CVCM SEAL CAVITY LINE, RG-01, TP-0l RR-CV-31CV 2027 G6 1 C 1 CK-B SA 0 CCL CVCM PT-24B SENSING LllNE, RG-01, TP-01 SH 2 ______COF CVCM____________ RR-CV-32CV 2027 B6 1 C 1 CK-B SA 0 CCL CVCM FT-i110C AND D HIGH SIDE, RG-01, SH 2 ___ _________ ___COF CVCM TP-01 RR-CV-33CV 2027 B6 1 C 1 CK-B SA 0 CCL CVCM FT-110C AND DLOW SIDE, RG-01, S_______ H 2 ___ ___ _____ _________ COF CVCM TP-01 Revision 0 Page 226 Revision 0 Page 226 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: PRIMARY CONTAINMENT (PC)VALVE CIC P&ID P&ID ISI 1ST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ PC-AOV-237AV 2022 F9 2 A 24 BTF AO C LJ-i RF SUPPRESSION CHAMBER INLET SHi1 FSC Q OUTBOARD ISOLATION, RG-01, FST Q TP-04___ __ __ ___ _ __ __ __ __ _____ __PIT 2Y PC-AOV-238AV 2022 E8 2 A 24 BTF AO C LJ-i RF DRY WELL INLET OUTBOARD SHi1 FSC Q ISOLATION, RG-01, TP-04 FST Q______________ _____PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _PC-AOV-243AV 2022 G10 A 20 BTF AO C LJ-i RE SUPPRESSION CHAMBER SH 1 FSO "Q VACUUM RELIEF, RG-01, TP-04 FST Q___ _ __ _ __ ____ __PIT 2Y PC-AOV-244AV 2022 H10 2 A 20 BTF AO C LI-i RE SUPPRESSION CHAMBER SHl 1 SO Q VACUUM RELIEF, RG-01, TP-04 FST Q___ __ __ __ _ __ _ __ ____ __PIT 2Y PC-AOV-245AV 2022 Fl1 A 24 BTF AO C LI-i RE SUPPRESSION CHAMBER SHi1 FSC Q EXHAUST OUTBOARD FST Q ISOLATION, RG-01, TP-04___ __ __ ___ _ __ __ ____ __PIT 2Y PC-AOV-246A'V 22 C2 2 A 24 BTF AO C LIi RE DRYWELL EXHAUST OUTBOARD SH1FSC Q ISOLATION, RG-01, TP-04 FST Q PIT 2Y Revision 0 Page 227 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: PRIMARY CONTAINMENT (PC)VALVE CIC P&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE P0S RQMT FREQ ____________ PC-AOV-247AV 2022 D3 2 A 0.5 GL AO C LJT-1 OPB DRYWELL EXHAUST OUTBOARD SHi1 FSC Q ISOLATION, RG-01, TP-04___ _ ___ _ __ ___ __ __FST Q PC-AOV-248AV 2022 D3 2 A 0.5 GL AO C LJT-1 OPB DRYWELL EXHAUST OUTBOARD SH 1 FSC Q ISOLATION, RG-01, TP-04___ _ _ _____ __ ___ _ FST Q PC-AOV-NRV20 2027 H7 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SH 1 PIT 2Y DRYWELL VACUUM BREAKER, VBT RF RG-01, TP-08 FSC 2Y___FSO 2Y PC-AOV-NRV21 2027 H7 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SHi1 PIT 2Y DRYWELL VACUUM BREAKER, VBT RF RG-01, TP-08 FSC 2Y___ _ __ ____ _____ __ __ __ FSO 2Y PC-AOV-NRV22 2027 H7 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SH 1 VBT RE DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y____ _ _ __ ___ _ __ ___ ___ _ _ ___ FSO 2Y _ _ _ _ _ _ _ _ _ _ _PC-AOV-NRV23 2027 H7 2 AC 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SHi1 VBT RE DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y Revision 0 Page 228 Cooper Nuclear Station FUifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: PRIMARY CONTAINMENT (PC)VALVE CIC P&ID P&ID 1SI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION ________COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ ____________ FSO 2Y PC-AOV-NRV24 2027 H9 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO Sill VBT RE DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y__ __ __ __ _ __ _ __ _ FSO 2Y PC-AOV-NRV25 2027 119 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO Sill VBT RE DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y___ _ __ __ __ __ __ __ FSO 2Y PC-AOV-NRV26 2027 H3 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SH 2 VBT RE DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y_____ FSO 2Y __ _ _ _ _ _ _ _ _ _ _PC-AOV-NRV27 2027 H3 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SH 2 VBT RE DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y_____ FSO 2Y __ _ _ _ _ _ _ _ _ _ _PC-AOV-NRV28 2027 14 2 AC 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SH 2 VBT RE DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y ESO 2Y Revision 0 Page 229 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: PRIMARY CONTAINMENT (PC)VALVE CIC P&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ ____________ PC-AOV-NRV29 2027 H4 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SH 2 VBT RF DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y FSO 2Y PC-AOV-NRV30 2027 115 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SH 2 VBT RF DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y___ _ ____FSO 2Y PC-AOV-NRV3 1 2027 H6 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SH 2 VBT .RF DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y_____ __ ___ ____ _ _____FSO 2Y _ _ _ _ _ _ _ _ _ _ _ _PC-CV-13CV 2022 G10 A/C 20 CK-D SA C LJ-1 RE SUPPRESSION CHAMBER SH 1 VBT 2Y VACUUM RELIEF, RG-01, TP-08 FSO 2Y___ __ ___ __ _ _ __ _ _ __ _ _ __ __ __ __FSC 2Y PC-CV-14CV 2022 H10 A/C 20 CK-D SA C LJ-1 RE SUPPRESSION CHAMBER Sil1 VBT 2Y VACUUM RELIEF, RG-01, TP-08 FSO 2Y___ ___FSC 2Y PC-CV-21CV 2022 El E122 IA/C 1/4 CK-S [SA 0/C COL-1i VC OPB R-1N2 SUPPLY TO H1202 MONITORS,TP0 Revision 0 Page 230 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: PRIMARY CONTAINMENT (PC)VALVE CIC P&ID P&ID !SI IST VALVE VALVE ,ACT NORM TEST TEST NOTES/DESCRIPTION ________COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ ____________ _____ ______ _____ ______CCL CVCM _ _ _ _ _ _ _ _ _ _ _ _PC-CV-22CV 2022 D3 2 A/C 1/4 CK-S SA 0/C LJ-1 OPB H2 CAL SUPPLY TO H 2 0 2 SH 2 COF CVCM MONITORS, RG-01, TP-01____CCL CVCM PC-CV-23CV 2022 C3 2 A/C 114 CK-S SA 0/C LJ-1 OPB 02 SUPPLY TO H 2 0 2 MONITORS, SH 2 COF CVCM RG-01, TP-01____ ____CCL CVCM PC-CV-25CV 2022 E5 2 A/C 1/4 CK-S SA 0/C LJ-i OPB N2 SUPPLY TO H 2 O 2 MONITORS, S112 COF CVCM RG-01, TP-01____ ____ ____ ____CCL CVCM PC-C V-26C V 2022 D6 2 A/C 1/4 CK-S SA 0/C LJ-1 OPB H2 CAL SUPPLY TO H 2 0 2 SH 2 COF CVCM MONITORS, RG-01, TP-01___ __ ___ __ ____ __ _____ __CCL CVCM PC-CV-27CV 2022 C6 2 A/C 1/4 CK-S SA 0/C LJ-i OPB 02 SUPPLY TO H 2 0 2 MONITORS, SH 2 COF CVCM RG-01, TP-01___ ___ ___ _____ _ _ __ _ _ ___ _ ___ ___ __ CCL CVCM PC-C V-33CV 2027 A5 2 A/C 3/8 CK-S SA C LJ-1 OPB N2 SUPPLY TO RR-AOV-741AV, SHi1 COF CVCM RG-01, TP-01____ ____ _____CCL CVCM PC-CV-34CV 2027 A4 2 A/C 3/8 CK-S SA C LJ-1 OPB N2 SUPPLY TO RR-AOV-741AV, SHi1 COF CVCM RG-01, TP-01___ __ ___ __ _ _____ __ ____ CCL CVCM PC-CV-35CV 2028 D8 2 A/C 3/8 CK SA C LJ-i OPB X-45D PASSIVE ISOLATION Revision 0 Page 231 Cooper Nuclear Station Fjifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: PRIMARY CONTAINMENT (PC)VALVE CIC P&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION ________COOR CLASS CAT SIZE TYPE TYPE P0S RQMT FREQ ____________ PC-CV-36CV 2028 *D7 2 A/C 3/8 CK SA C LJ-1 OPB X-45D PASSIVE ISOLATION PC-MOV-230MV 2022 F2 2 A 24 BTF MO C LJ-i RF SUPPRESSION CHAMBER SHi1 FSC Q EXHAUST INBOARD ISOLATION,__ _ __ _ __ ___ _ _ _ _ _ _PIT 2Y RG-01 PC-MO V-23 1MV 2022 C2 2 A 24 BTF MO C LJ-1 RF DRYWELL EXHAUST INBOARD SH 1 FSC Q ISOLATION, RG-01___ __ ___ __ _ _ __ _ _____ __ __ __ PIT 2Y PC-MOV-232MV 2022 E8 2 A 24 BTF MO C UJ-1 RF DRYWELL INLET INBOARD SHi1 FSC Q ISOLATION, RG-01__ __ __ _PIT 2Y PC-MO V-233MV 2022 E8 2 A 24 BTF MO C LJ-1 RF SUPPRESSION CHAMBER INLET SHi1 FSC Q INBOARD ISOLATION, RG-01___ _ __ _ __ ____ __PIT 2Y PC-MOV-305MV 2022 G2 2 A 2 GT MO C LU-1 OPB PC-MOV-23OMV BYPASS, RG-01 SHi1 FSC Q___ _ __ _ __ ____ __PIT 2Y PC-MOV-306MV 2022 C2 2 A 2 GT, MO C UJ-1 OPB PC-MOV-231MV BYPASS, RG-01 Sil1 FSC Q___ _ __ _ __ ____ __PIT 2Y PC-MOV-1301MV 2084 C9 2 A 1 GT MO 0 LJ-i OPB SUPPRESSION CHAMBER FSC Q ISOLATION SYSTEM B, RG-01___ ___ ___ __ ___ __ __ ___ _ ___ __ ___ ___ __ ___ __ __ PIT 2Y PC-MOV-1302MV 204 CG T MO 0O- OPB SUPPRESSION CHAMBER Revision 0 Page 232 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: PRIMARY CONTAINMENT (PC)VALVE CIC P&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ ____________ FSC Q ISOLATION SYSTEM B, RG-01______ ______ _ ____ ________________PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _PC-MOV-1303MV 2084 E2 2 A 1 GT MO C LJ-1 OPB SUPPRESSION CHAMBER FSC Q ISOLATION SYSTEM A, RG-01___ __ ___ __ _ _ __ __ __PIT 2Y PC-MOV-1304MV 2084 E3 2 A 1 GT MO C LJ-1 OPB SUPPRESSION CHAMB3ER FSC Q ISOLATION SYSTEM A, RG-01___ __ __ ___ _ __ __ ____ __PIT 2Y PC-MOV-1305MV 2084 C2 2 A 1 GT MO C LJ-I OPB DRYWELL ISOLATION SYSTEM FSC Q A, RG-01______ ~~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _PC-MOV-1306MV 2084 C3 2 A 1 GT MO C LJ-1 OPB DRYWELL ISOLATION SYSTEM FSC Q A, RG-01_____ ______ ~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _PC-MO V-1308MV 2084 G9 2 A 1 GT MO C LJ-i OPB BLEED ISOLATION FOR FSC Q SUPPRESSION CHAMBER, RG-01_____ _____ ~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _PC-MOV-1310MV 2084 B9 2 A 1 GT MO C LJ-1 OPB BLEED ISOLATION FOR FSC Q DRYWELL, RG-01______ _ ____ ___ _ ______ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _PC-MOV-1311iMV 2084 C8 2 A 1 GT MO 0 LJ-1 OPB DRYWELL DILUTION SUPPLY FSC Q ISOLATION SYSTEM B, RG-01_____ _ ____ __ __ ____ _ _____ ______ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _Revision 0 Pg 3 Page 233 Cooper Nuclear Station F~ifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: PRIMARY CONTAINMENT (PC)VALVE CIC P&ID P&ID IS1 1ST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION ___________COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ ____________ PC-MOV-1312MV 2084 C9 2 A 1 GT MO 0 LJ-1 OPB DRYWELL DILUTION SUPPLY FSC Q ISOLATION SYSTEM B, RG-01 PIT 2Y Revision 0 Page 234 Revision 0 Page 234 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: RADIATION MONITORING (RMV)VALVE CIC P&ID P&ID ISI 1ST VALVE VALVE IACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE jTYPE POS RQMT FREQ_____________ RMV-AOV- 10AV 2022 D3 2 A 3/4 GB AO 0 LU-i1 OPB RM-4 CONTAIN4MENT ISOLATION, Sill FSC Q INBOARD, RG-0l, TP-04 FST Q___ __ __ __ ____ __PIT 2Y RMV-AOV-l11AV 2022 D2 2 A 3/4 GB AO 0 LU-I OPB RM-4 CONTAINMENT ISOLATION, SHi1 FSC Q OUTBOARD, RG-01, TP-04 FST Q___ __ __ __ ____ __PIT 2Y RMV-AOV-12AV 2022 E3 2 A 3/4 GB AO 0 LJ-i OPB RM-4 CONTAINMENT ISOLATION, SHl 1 FSC Q INBOARD, RG-0l, TP-04 FST Q___ __ __ ___ _ __ __PIT 2Y RMV-AOV--13AV 2022 E2 2 A 3/4 GB AO 0 LJ-i OPB RM-4 CONTAINMENT ISOLATION, SHll FSC Q OUTBOARD, RG-0l, TP-04 FST Q PIT 2Y Revision 0 Pg 3 Page 235 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: RADIOACTIVE WASTE (RW)1P&ID ISI 1ST VALVE VALVE IACT NORM TEST TEST VALVE CIC P&ID COOR CLASS CAT SIZE TYPE jTYPE P0S RQMT FREQ NOTES/DESCRIIPTION RW-AOV-AO82 2038 G5 2 A 3 GT AO 0 LJ-i OPB DRYWELL FLOOR DRAIN SUMP FSC Q DISCHARGE, RG-01, TP-04 EST Q________ ______ _ ____ ___ _ ______ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RW-AOV-AO83 2038 G5 2 A 3 GT AO 0 LJ-i OPB DRYWELL FLOOR DRAIN SUMP FSC Q DISCHARGE, RG-01, TP-04 FST Q________ ______ ____ ___ __ ____ ______ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RW-AOV-AO94 2028 Gil1 2 A 3 GT AO 0 LU-i1 OPB DRYWELL EQUIPMENT DRAIN FSC Q SUMP DISCHARGE, RG-O1, TP-04 FST Q__________ ___ _____ ______ __ __ ______ PIT 2Y_ _ _ _ _ _ _ _ _ _ _ _ _RW-AOV-AO95 2028 G12 2 A 3 GT AO 0 LI-1 OPB DRYWELL EQUIPMENT DRAIN FSC Q SUMP DISCHARGE, RG-01, TP-04 FST Q______________ _____ ______ PIT 2Y_ _ _ _ _ _ _ _ _ _ _ _ _RW-CV-58CV 2005 F10 A C 2 CK-S SA C FSO Q Z SUMP PUMP A DISCHARGE S_______ H 2 _____________FSC Q' CHECK, RG-01 RW-CV-59CV 2005 F10 A C 2 CK-S SA C FSO Q Z SUMP PUMP B DISCHARGE SH 2 FSC Q CHECK, RG-01 OG-CV-8CV "2037 D9 A C 1/2 CK-P SA C COF CVCM OFFGAS HOLD-UP DRAIN LINE CCD CVCM VENT, RG-01, TP-01 Revision 0 Pg 3 Page 236 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: RADIOACTIVE WASTE (RW)P&ID ISI 1ST VALVE VALVE ACT NORM TEST TEST VALVE CIC P&ID_ COOR CLASS jCAT SIZEJTYPE TYPE POSj RQMT FREQ J NOTES/DESCRIPTION OG-CV- 12CV 2037 D9 A C 1/2 CK-P SA C COF CVCM IOFFGAS HOLD-UP D1AIN LINE______________ ___ ____J__________________ ___JCCD CVCM VENT, RG-01, TP-01 Revision 0 Page 237 Revision 0 Page 237 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: REACTOR CORE ISOLATION COOLING (RCIC)____ _

  • NOR IP&ID ISI IST VALVE IVALVE ACT M TEST TEST VALVE CIC P&D COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION RCIC-AOV-AO34 2041 [H110 2 B 1 GB AO 0 FSC Q RCIC STEAM LINE DRIPLEG DRAIN IjFST Q RG-01, TP-04___ __ __ __ __ ___
  • __ __.1___ __ ______ ___ __ ______.1______

__ __IL __ __ PI 2 __ __ __ __ __ __PIT__ ___2Y_RCIC-CV-1 0CV 2043 115 12 1C F6 CK-S S C 1COF CVCM 1ECST SUPPLY TO RCIC PUMP, RG-_______ ________ IKS .1. CCL IICVCM 101 RCIC-CV-11CV 2043 1110 [2 1C [6 CK-S 1SA C 1COD CVCM 1RCIC SUPPLY FROM SUPPRESSION _________ i I J ___[ j __ ___CCD CVCM ]CHAMBER, RG-01 RCIC-CV-12CV 2043 G9 2 C 2 CK-S SA C COF CVCM VACUUM PUMP DISCHARGE TO[CCL CVCM SUPPRESSION CHAMBER, RG-01,_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ TP-01 RCIC-CV- 13CV [04 D5 2 C 2 [CK-P SA] C COF CM MIMUM FLOW LINE, RG-01______ ______ ____ _ __ __ __ ii L .___ ____ CCL CVCM _ _ _ _ _ _ _ _ _ _ _ _RCIC-CV-15CV 2043 D8 2 A/C 8 CK-S SA C LJ-1 OPB RCIC TURBINE EXHAUST TO iFCOF CVCM ISUPPRESSION CHAMBER, RG-01___ ___ __ ___ ___ _ ___ _ ___ ___ _ ___ __ __ ___ _ CCL CVCM RCIC-CV-16CV 2043 F9 2 C 2 CK-S SA -C COF 1CVCM VACUUM PUMP DISCHARGE TO CCL CVCM SUPPRESSION CHAMBER, RG-01, RCIC-CV-18CV 2043 A8 12 C 2 CK-P SA C CO CVM CNEST SUPPLY TO RI Revision 0 Page 238 Revision 0 Page 238 Cooper Nuclear Station Ffifh Interval lnservice Testing Pro gram for Pumps and Valves SYSTEM: REACTOR CORE ISOLATION COOLING (RCIC)NOR P&ID ISI IST VALVE VALVE ACT M TEST TEST VALVE CIC P&ID COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION RCIC-CV-19CV 2043 A8 A C 2 CK-P SA -C COF CVCM -CONDENSATE SUPPLY TO RCIC CCF CVCM PRESSURE MAINTENANCE, RG-01, RCIC-CV-20CV 2043 118 A irC 2 CK-P ][SA [C ][COD [CVCM [RCIC CONDENSATE PLivIP__________ ___ ___ _______ ____ ________ ___J[CCD [CVCM [DISCHARGE, RG-01, TP-01 RCCC- 1C 2043 H 17 1 2 1C 2 [CK-P ][SA IC [COD [CVCM [RCIC CONDENSATE PM_______________ ___I ____B.___ ____ [ I____I____CCD CVCM [DISCHARGE, RG-01 rRCIC-CV-22CV 12043 [F9 1 2 [C 1 1/2 [CK-L ][SA C ][COD CVCM IRCIC VACUU BREAKER,RG0 -] ___ [ J_____ ____ ____ I J[___ I ][CCD CVCM __________ RCIC-CV-23CV 12043 IFO10 2 C 1 1/2A CK-L ][SA IC COD CVCM RCIC VACUUM BREAKER, RG-01______________ ___I i ____ I J____ ____CCD CVCM __________ RCIC-CV-24CV 12043 F9 2 C 1 1/2 CK-L SA C COD CVCM CIVACUUM BREAKER, RG-01___________I ___ [___ ____J[___ ____ ____ ___ ____CCD CVCM _____________ RCIC-CV-25CV 12043 [F10 C 1 'A1/2 CK-L ][SA IC ][COD ICVCM [RCIC VACUU BREAKER, RG-01__________ ___ [___ ___Jr________[___ ____[ J[CCD CVCM __________ RCIC-CV-26CV 2043 B8 1 A/C 4 CK-S SA 0 LJ-1 RE INJECTION CHECK VALVE, iFSO RE ROJ-09, RG-01__________________ ______ ______________I, _____ _______ _______I______ ______ii_______ ______ ______________FSC________RF___ Revision 0 Pg 3 Page 239 Cooper Nuclear Station F/ifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: REACTOR CORE ISOLATION COOLING (RCIC)NOR P&ID ISI IST VALVE VALVE ACT M TEST TEST VALVE CIC P&I COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION ___ ___ __ ___ __ __ __ __ __ ____ __ i __ __ ___ __ __ _ _ ]] _ _ _ ][ [ PiT ][ 2Y _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _RCIC-HOV-HO 11 2041 G8 A B 2 1/2 GL HO C SKID Q ]RCIC TU GOV V, RG-01, TP-05 RCIC-MOV-MO14 2041 G8 2 B f2 x3 GL HO C SKID fQ I RCIC TU TRIP & THROTTFILE V, RCIC-MOV-MO 15 2041 D7 1 A 3 GT MO 0 LJ-I OPB IRCIC STEAM INBOARD 111FSC Q/CS BISOLATION, CSJ-10,RG0 RCIC-MOV-MO16 2041 D9 1 A 3 GT MO 0 LJ-i OPB RCIC STEAM OUTBOARD FSC Q ISOLATION, RG-01 PIT 2Y RCIC-MOV-MO018 2043 H4 2 B 6 GT MO 0 FSO Q RCIC SUPPLY FROM CONDENSATE FSC Q STORAGE, RG-01_____ _____ _____ _____PIT 2Y_ _ _ _ _ _ _ _ _ _ _ _ _RCIC-MOV-MO20 2043 B7 2 B 4 GT MO 0 PIT 2Y RCIC PUMP DISCHARGE, RG-01 RCIC-MOV-MO21 2043 B8 2 B 4 GT MO C FSO Q RCIC INJECTION TO REACTOR,___ __ __ __ __ ___ _ __ _ __ __ __ _ __ __ ___ __ __ __ __ __ PIT 2Y RG-01 RCIC-MOV-MO27 2043 C7 2 B ]GB MO C FSO Q 1RCIC PUMP MINIMUM FLOW___PIT 2Y CHAMBER, RG-01 Revision 0 Page 240 Revision 0 Page 240 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: REACTOR CORE ISOLATION COOLING (RCIC)[ NOR P&ID ISI IST VALVE VALVE ACT M TEST TEST VALVE CIC P&ID COOR CLASS CAT [SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION RCIC-MOV-MO30 [2043 1D1 2 B 4 GB IMO 1C PIT J2Y 1RCIC TEST RETURN ROOT,____ ____ ____ _ I____ .1 1 _ ____ ____ __ ___ ___ ___ ___ 1 _ ____ ___ __ PASSIVE, PAS I RE 0 RCCMVM3 20431 El1 A 1B 4[ 1GT MO] C PIT ]2Y 1RCIC TEST RETURN SHUTOFF,____ ____ ____ _ _ ___ ____ I.____ _ I ____31 ____ 1 _ ___ _ __ __ _ ____ ___ __ PASSIVEPASSVE, -G01 RCIC-MOV-MO41 2043 Hl0 2 B 6 GT MO C FSO Q RCIC SUPPLY FROM SUPPRESSION _____ _____ ____ ___ _ __ _ _ ___ ____ ____ ____ PIT [2Y_ _ _ _ _ _ _ _ _ _RCIC-MO V-MO 131 2041 1G9 F2 1B F3 1 GB MO 1~C FSO 1 RCIC STEAM SUPPLY TO RCIC_______________ _____I _____ I______ I ____ I ______ I_____PIT___2Y _TURBPII2YETRB RE, -G01 RCIC-MO V-MO 132 2043] E4 [2 1B F2 1 B MO C FSO 1 AUXILIARY COOLING SUPPLY,___ ___ ___ ___ __ _ _ __ __ ii___ __ __ ___ [ _ __ PIT ]2Y RG-01_ __ __ __ _RCCR-20 20431 D6 F2 1Df 8 1RD SA 1[C 1RD ]5Y 1 EXHAUST LINE RUPTURE DISC, CI-V10ORV 20431 G5 F2 1C 1 RV SA] C 1 RVT [APP I]RCIC PUMP SUCTION RELIEF,____ ___ ___ _ _ ___J _ ___[ __ __ __ _ _ ___1 _ ___ ___ _ I. _ ___J _ ___j [ J RG-01 RCIC-RV-11RV 2043 G6 2 ]C 1 RV SA C RVT A(PPI ]CI AUX~ILAY COOLING RCIC-V-37 2043 D9 2 A/C 8 -CK MA 0 LJ-i1 OPB RCIC TURBINE EXHAUST TO___________ ____ __ __ ____ __ _ ____ ____ ___ ____ CCL CVCM ISOLATION, RG-01 Revision 0 Pg 4 Page 241 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: REACTOR EQUIPMENT COOLING (REC)VALVE CIC 1P&ID TP&ID ISI IST 1VALVE] VALVE ACT] NORM] TEST TTEST [ NOTES/DESCRIPTION ________I jCOOR CLASS CAT SIZE jTYPE TYPE jP0S RQMT jFREQ REC-CV-10CV 2031 F2 3 C 8 CK-T SA 0, FSO Q REC-P-A DISCHARGE, RG-01 SH 2 ___ __FSC Q REC-CV- 11 CV 2031 F3 3 C 8 CK-T SA 0 FSO Q REC-P-B DISCHARGE, RG-0 1__ _ _ _ _ _ SH 2 __ _FSC Q REC-CV-12CV 2031 F4 3 C 8 CK-T SA 0 FSO Q REC-P-C DISCHARGE, RG-01 S______ H 2 FSC Q REC-CV-13CV 2031 F5 3 C 8 CK-T SA 0 FSO Q REC-P-D DISCHARGE, RG-01 SH 2 __ _FSC Q _ _ _ _ _ _ _ _ _ _ _REC-CV- 1 6CV 2031 H2 3 C 12 CK-S SA 0 COF CVCM NON-CRITICAL HEADER RETURN S_______ H 2 ___ ___CCF CVCM TO REC PUMPS, RG-01, TP-01, TP-06 REC-MOV-694MV 2031 H2 3 B 4 GT MO 0 PIT 2Y PASSIVE CRITICAL LOOP RETURN S_______ H 2 _______ ___CROSSTIE, RG-01 REC-MOV-695MV 2031 C5 3 B 4 GT MO 0 PIT 2Y PASSIVE CRITICAL LOOP SUPPLY S H2 ___ ______ ___CROSSTIE, RG-01 REC-MOV-697MV 2031 H3 3 B 6 GT MO 0 FSO Q NORTH CRITICAL LOOP RETURN,______ SH 2 PIT 2Y RG-01 REC-MOV-698MV 2031 Hi1 B 6 GT MO 0 FSO Q SOUTH CRITICAL LOOP RETURN, S_____ H 2 PIT 2Y RG-01 REC-MOV-700MV 2031 B2 3 B 10 GT MO 0 FSC Q NON-CRITICAL LOOP SUPPLY__ _ _ _ _ _ SH 2 _ _ __ _ __ _ _ _ _ _ __ _ __ _ __ PIT 2Y SHUTOFF, RG-01 Revision 0 Page 242 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: REACTOR EQUIPMENT COOLING (REC)VALVE CIC P&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION ________COOR C LASS CAT SIZE TYPE _TYPE POS RQMT FREQ REC-MOV-702MV 2031 A4 2 A 8 GT MO 0 LJ-1 OPB DRYWELL SUPPLY ISOLATION SH 2 FSC Q VALVE, RG-01 S_ __ ____ PIT 2Y REC-MOV-709MV 2031 G3 2 A 8 GT MO 0 LJ-1 OPB DRYWELL RETURN ISOLATION SHf1 FSC Q VALVE, RG-01 PIT 2Y REC-MOV-7 11 MV 2031 D4 3 B 6 GT MO 0 FSC Q NORTH CRITICAL LOOP SUPPLY, SH 2 FSO Q RG-01___ _ __ _ _ __ _ ___ __ __ __ __ __ PIT 2Y REC-MOV-712MV 2031 Dl B 12 BFL MO 0 FSC Q REC HEAT EXCHANGER A SH 2 _____ ___ ___PIT 2Y OUTLET, RG-01 REC-MOV-713MV 2031 C1 B 12 BFL MO 0 FSC Q REC HEAT EXCHANGER B OUTLET, Sf 2 _ __ PIT 2Y RG-01 REC-MOV-714MV 2031 C5 3 B 6 GT MO 0 FSC Q SOUTH CRITICAL LOOP SUPPLY, SH 2 FSO Q RG-01___ __ __ ___ _ __ _ ___ _ __ __ PIT 2Y REC-MOV-1329MV 2031 B3 3 B 8 GT MO 0 FSC Q AUXILIARY RADIOACTIVE WASTE SHf2 PIT 2Y BUILDING SUPPLY, RG-01 Revision 0 Page 243 Revision 0 Page 243 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: REACTOR FEEDWATER (RF)P&ID ISI IST VALVE VALVE ACT NORM TEST TEST VALVE CIC P&ID COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION RF-CV-13CV 2044 B9 1 A/C 18 CK-S SA 0 LU-i RF FEEDWATER LINE B TO REACTOR FSC RF OUTBOARD, ROJ-03, RG-.01, TP-01_____ ______ ~~FSO OP __ _ _ _ _ _ _ _ _ _ _ _RF-CV- 14CV 2044 Cl 10 A/C 18 CK-S SA 0 U -i1 RF FEEDWATER LINE B TO REACTOR FSO OP INBOARD, ROJ-03, RG-01_______________ ______ _____ FSC RF_ _ _ _ _ _ _ _ _ _ _ _ _ _RF-CV-15CV 2043 A9 1 A/C 18 CK-S SA 0 UJ-i RF FEEDWATER LINE A TO REACTOR FSC RF OUTBOARD, ROJ-03, RG-01, TP-01 FSO OP __ _ _ _ _ _ _ _ _ _ _ _RF-CV- 16CV 2043 Cl 10 A/C 18 CK-S SA 0 U -i1 RF FEEDWATER LINE A TO REACTOR FSO OP INBOARD, ROJ-03, RG-01 FSC RE Revision 0 Pg 4 Page 244 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: REACTOR RECIRCULATION (RR)P&ID 1ISI IST VALVE VALVE ACT NORM TEST TEST VALVE CIC [P&ID COOR C LASS CAT SIZE TYPE [TYPE POS RQMT FREQ NOTES/DESCRIPTION RR-AOV-740AV 2027 A3 1 A 3/4 GB AO 0 LI-i OPB SP-1 OUTBOARD ISOLATION, SHi1 FSC Q RG-0l, TP-04 FST Q PIT 2Y RR-AOV-741AV 2027 A7 1 A 3/4 GB AO 0 LI-i OPB SP-1 INBOARD ISOLATION, RG-01, SHi1 FSC Q TP-04 FST Q___ _ __ _ __ __PIT 2Y RR-MOV-MO53A 2027 B7 1 B 28 GT MO 0 FSC CS RR PUMP A DISCHARGE SHi ________ PIT 2Y CSJ-04, RG-01 RR-MOV-MO53B 2027 C4 1 B 28 GT MO 0 FSC CS RR PUMP B DISCHARGE__________ SHi 1 ___ _______ ________ PIT 2Y CSJ-04, RG-01 Revision 0 Page 245 Revision 0 Page 245 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: REACTOR WATER CLEANUP (RWCU)IP&1ID ISI IST VALVE VALVE ACT NORM TEST TEST VALVE CIC jP&ID COOR CLASS CAT SIZE TYPE jTYPE POS RQMT FREQ NOTES/DESCRIPTION RWCU-CV-15CV 2042 C4 1 A/C 4 CK-S SA 0 LU-1 RE RWCU RETURN TO REACTOR SHi1 FSC RE VESSEL, ROJ-04, RG-01, TP-01, TP-06___ _ _ __ __ __ __ __ FSO RE RWCU-MOV-MO15 2042 E2 1 A 6 GT MO 0 LU-i OPB RWCU SUPPLY INBOARD SHi1 FSC RF ISOLATION, ROJ-05, RG-01________ _____ _ ___ ___ __ ____ ______ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RWCU-MOV-MO018 2042 B4 1 A 6 "GT MO 0 LJ-1 OPB RWCUJ SUPPLY OUTBOARD Sill FSC RE ISOLATION, ROJ-05, RG-01 PIT 2Y Revision 0 Page 246 Revision 0 Page 246 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RIHR)VALVE CIC P&ID COOR CLASS CAT SIZE TYPE TYPE P0S RQMT FREQ T NOTES/DESCRIPTION RIIR-CV-10CV 2040 F3 2 C 3 CK-S SA C COD CVCM RHR PUMP A MINIMUM FLOW, SH 1 __ ___ CCD CVCM RG-01 RHR-CV-1 1CV 2040 F8 2 C 3 CK-S SA C COD CVCM RI{R PUMP B MINIMUM FLOW, S_______ H 2 __________CCD CVCM RG-0l RH-R-CV-12CV 2040 H6 2 C 3 CK-S SA C COD CVCM RHR PUMP C MINIMUM FLOW,________Sil 1 __ ___ __ CCD CVCM RG-01 RHR-CV-13CV 2040 H7 2 C 3 CK-S SA C COD CVCM RH-R PUMP D MINIMUM FLOW, S_______ H 2 ___ ______CCD CVCM RG-01 RHR-CV-14CV 2040 F5 2 C 16 CK-T SA C FSO Q RLER PUMP A DISCHARGE, RG-01__ _ _ _ _ _ SHi 1 _ _ _ _ __ _ FSC Q RHR-CV- 1 5CV 2040 F8 2 C 16 CK-T SA C ESO Q RHR PUMP B DISCHARGE, RG-0 1 SH 1-_ _ __ _ ___ _ FSC Q RHR-CV-16CV 2040 H5 2 C 16 CK-T SA C FSO Q RHR PUMP C DISCHARGE, RG-01 SHl ___________ FSC Q RHiR-CV-17CV 2040 H8 2 C 16 CK-T SA C FSO Q RIHR. PUMP D DISCHARGE, RG-01__ _ _ _ _ _ SHl _ _ __ _ FSC Q RHIR-CV-18CV 2040 A9 A C 4 CK-S SA C COF CVCM RH-R LOOP B OUTBOARD SH 2 CCF CVCM PRESSURE MAINTENANCE ____ ___ ____ _ _ ___ ____ _ _ ___SUPPLY, RG-01, TP-01 RI-R-CV-19CV 2040 A9 2 C 4 CK-S SA C COF CVCM RHR LOOP B INBOARD PRESSURE Revision 0 Page 247 Cooper Nuclear Station FUifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RIIR)P&ID ISI IST VALVE VALVE ACT NORM TEST TEST VALVE CIC P&ID COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION SH 2 CCF CVCM MAINTENANCE SUPPLY, RG-01, TP-CCD/ CVCM 01 CCR RiHR-CV-24CV 2040 C7 A C 4 CK-S SA C COF CVCM RHR LOOP A OUTBOARD SH 2 CCF CVCM PRESSURE MAINTENANCE __ _ _______ ____ ___SUPPLY, RG-01, TP-01 RHIR-CV-25CV 2040 C7 2 C 4 CK-S SA C COF CVCM R!HR LOOP A INBOARD PRESSURE SHi1 CCF CVCM MAINTENANCE SUPPLY, RG-01, TP-CCD/ CVCM 01______ ~~~~~~CCR__ _ _ _ _ _ _ _ _ _ _ _RHR-CV-26CV 2040 BI 10 A/C 24 CK-S SA C LT-2 OPE LOOP A INJECTION LINE SF11 FSO RF TESTABLE CHECK, PRESSURE FSC RF ISOLATION VALVE, ROJ-1 1, RV-05,____PIT RE RG-01 RHlR-CV-27CV 2040 B4 1 A/C 24 CK-S SA C LT-2 OPB LOOP B INJECTION LINE SF1 1 FSO RE TESTABLE CHECK, PRESSURE FSC RE ISOLATION VALVE, ROJ-1 1, RV-05,____PIT RE RG-01 RHR-MOV-MO12A 2040 E3 2 B 16 GT MO 0 FSO Q RHR HEAT EXCHANGER A SH 1 PIT 2Y OUTLET, RG-01 RH{R-MOV-MO 12B 2040 F10 B 16 GT MO 0 FSO Q RIR HEAT EXCHANGER B S_____ H 2 ___ __PIT 2Y OUTLET, RG-01 RitIR-MOV-MO13A 2040 FO 2 B 20 GT MO 0 FSO Q RUR PUMP A SUCTION FROM SF1 1 FSC Q SUPPRESSION CHAMBER, RG-01 Revision 0 Page 248 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RHR)P&ID IS I5 ST VALVE VALVE fACT NORM TEST TEST VALVE CIC [P&ID COOR {CLASS CAT SIZE TYPE TYPE P0S RQMT "FREQ NOTES/DESCRIPTION ___________ ___ _____ _ ___ ______ PIT 2Y_ _ _ _ _ _ _ _ _ _ _ _ _RHR-MOV-MO13B 2040 F4 2 B 20 GT MO 0 FSO Q RHIR PUMP B SUCTION FROM SH 2 FSC Q SUPPRESSION CHAMBER, RG-01_____ ______ ~~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RHR-MOV-MO13C 2040 F10 B 20 GT MO 0 FSO Q RHR PUMP C SUCTION FROM SH 1 FSC Q SUPPRESSION CHAMBER, RG-01______ _____ ~~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RHR-MOV-MO 13D 2040 F3 2 B 20 GT MO 0 FSO Q RH-R PUMP D SUCTION FROM SH12 FSC Q SUPPRESSION CHAMBER.. RG-01___ __ __ ___ ____ __PIT 2Y RHR MOV-MO15A 2040 F9 2 B 20 GT MO C FSC Q RHR PUMP A SDC SUCTION, RG-0l SHl 1 _ __ _ _ _ PIT 2Y RHR-MO V-MO 15B 2040 F4 2 B 20 GT MO C FSC Q RI-R PUMP B SDC SUCTION, RG-01I SH 2 __ __ _ _ _ PIT 2Y _ _ _ _ _ _ _ _ _ _RHIR-MOV-MOI15C 2040 G8 2 B 20 GT MO C FSC Q RHR PUMP C SDC SUCTION, RG-01 Sil 1 ____ ___ PIT 2Y______ _____________ RHR-MOV-MO15D 2040 G5 2 B 20 GT MO C FSC Q RI-R PUMP D SDC SUCTION, RG-01 SH 2 ___PIT 2Y RI-IR-MOV-MO 16A 2040 E7 2 B 4 GT MO 0 FSO Q PUMP AAND CMINIMUM FLOW, SH 1 FSC Q RG-01__ _ __PIT 2Y RHR-MOV-MO16B 24 E7 2 B 4 GT MO O j FSO Q PUMP B AND DMIMINUM FLOW, Revision 0 Page 249 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RHR)P&ID ISI IST VALVE VALVE ACT NORM TEST TEST VALVE CIC P&ID COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION SH 2 FSC Q RG-01___ __ __ ___ ____ ___ __ __PIT 2Y RH{R-MOV-MO17 2040 C8 1 A 20 GT MO C LJ-1 OPB RIIR SDC SUPPLY OUTBOARD SH11 LT-2 PB PRESSURE ISOLATION VALVE, FSC CS CSJ-05, RV-05, RG-01___ ___ ___ ___ _ ______ __ __ _ _ ___ __ __PIT 2Y RHR-MOV-MO18 2040 C10 A 20 GT MO C LT-2 OPB RUR SDC SUPPLY INBOARD 5111 FSC CS PRESSURE ISOLATION VALVE,___________ _________________PIT 2Y CSJ-05, RV-05, RG-01 RHR-MOV-MO20 2040 H3 2 B 20 GT MO C PIT 2Y RHR PASSIVE CROSSHEADER __________ SHi 1___ ______ SHUTOFF, RG-01 RHR-MOV-MO25A 2040 B8 1 A 24 GT MO C LJ-1 OPB RHR LOOP A INJECTION INBOARD SH 1 LT-2 PB ISOLATION, RV-05, RG-01 FSO Q FSC Q___ __ ___ __ _ _ __ __ __ __ __PIT 2Y RHR-MOV-MO25B 2040 B5 1 A 24 GT MO C LJ-1 OPB RHR LOOP B INJECTION INBOARD 5112 LT-2 PB ISOLATION, RV-05, RG-01 FSO Q FSC Q___ __ ___ __ _ _ __ _ _ __ __ __PIT 2Y RHR-MOV-MO26A 24 A7 2 B 10 GT MO C FSO Q DRYWELL SPRAY LOOP A Revision 0 Page 250 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RHLR)P&ID ISI IST VALVE VALVE [ACT NORM TEST TEST VALVE CIC _P&ID COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION S111 FSC Q OUTBOARD ISOLATION, RG-01_____ _____ ____ ______ ______ IT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RHR-MOV-MO26B 2040 B6 2 B 10 GT MO C FS0 Q DRYWELL SPRAY LOOP B SH 2 ESC Q OUTBOARD ISOLATION, RG-01 PIT 2Y RHR-MOV-MO27A 2040 B8 2 B 24 ANG MO 0 FSO Q LOOP A INJECTION OUTBOARD SHi 1 __ PIT 2Y THROTTLE, RG-01 RHR-MOV-MO27B 2040 B6 2 B 24 ANG MO 0 FSO Q LOOP B INJECTION OUTBOARD SH 2 ___ __PIT 2Y THROTTLE, RG-01 RI-R-MOV-MO31A 2040 A8 2 A 10 GT MO " C LJ-1 OPB DRYWELL SPRAY LOOP A SHi1 FSO Q INBOARD ISOLATION, RG-01 FSC Q_____ _____ ______ ~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RHR-MOV-MO3 lB "2040 B5 2 A 10 GT MO C LJ-i OPB DRYWELL SPRAY LOOP B SH 2 FSO Q INBOARD ISOLATION, RG-01 FSC Q PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RHR-MOV-MO34A 2040 E7 2 B 18 GB MO C FSO Q SUPPRESSION CHAMBER COOLING SHi1 FSC Q LOOP A INBOARD THROTTLE,________PIT 2Y RG-01 RHR-MOV-MO34B 2040 E6 2 B 18 GB MO C FS0 Q SUPPRESSION CHAMBER COOLING SH 2 FSC Q LOOP B INBOARD THROTITLE,___ __ __ __ __ __ _ __ _ _ __ _ __ _ _ __ _ __ __ __ _ __ __ PIT 2Y RG-01 Revision 0 Page 251 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RHR)P&D IS ST VALVE VAVEACT NORM TEST TEST VALVE CIC [P&ID COOR CLASS CAT SIZE TYPE [TYPE POS RQMT FREQ NOTES/DESCRIPTION RHR-MOV-MO38A 2040 D9 2 A 6 GB MO C UI_- OPB SUPPRESSION CHAMBER SPRAY SH 1 FSO Q LOOP A INBOARD THROTTLE, FSC Q RG-01_____ ______ _____ ______ PIT 2Y_ _ _ _ _ _ _ _ _ _ _ _ _RHR-MOV-MO38B 2040 D6 2 A 6 GB MO C LI-1 OPB SUPPRESSION CHAMBER SPRAY SH 2 ESO Q LOOP B INBOARD THROTTLE, FSC Q RG-01_____ ______ ______ ~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RHR-MOV-MO39A 2040 D6 2 B 18 GT MO C FSO Q SUPPRESSION CHAMBER COOLING SHi1 FSC Q LOOP A OUTBOARD ISOLATION, PIT 2Y RG-01 RHR-MOV-MO39B 2040 D7 2 B 18 GT MO C FSO Q SUPPRESSION CHAMBER COOLING SH 2 FSC Q LOOP B OUTBOARD ISOLATION,____PIT 2Y RG-01 RHR-MOV-MO57 2040 H2 2 B 4 GB MO C LT-2 RF RHR DISCHARGE TO RAD WASTE SH 2 FSC Q INBOARD THROTTLE, RG-01______ ~~~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RHR-MOV-M065A 2040 F3 2 B 16 GT MO 0 PIT 2Y RHR HEAT EXCHANGER A INLET, SH 1__ __ RG-01 RHR-MOV-MO65B 2040 Fl0 2 B 16 GT MO 0 PIT 2Y RHR HEAT EXCHANGER B INLET, SH 2 _ _ __ _ __ _ __ _RG-01 RHR-MOV-MO66A 24 F4 2 B 20 GB MO O FSO Q RIIR HEAT EXCHANGER A BYPASS SH1FSC Q THROTTLE, RG-01 PIT 2Y Revision 0 Page 252 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RHR)FIP&ID ISI 1ST VALVE VALVE ACT NORM TEST TEST VALVE.CIC _P&ID COOR CLASS CAT SIZE TYPE _TYPE POS RQMT FREQ NOTES/DESCRIPTION RHR-MOV-MO66B 2040 F9 2 B 20 GB MO 0 FSO Q RHR HEAT EXCHANGER B BYPASS SH 2 FSC Q THROTTLE, RG-01__ _ _ _ _PIT 2Y RIIR-MOV-MO67 2040 H2 2 B 4 GT MO C LT-2 RF RHR DISCHARGE TO RAD WASTE SH 2 FSC Q OUTBOARD SHUTOFF, RG-01_____ ____ ______ __ __ ______ PIT 2Y_ _ _ _ _ _ _ _ _ _ _ _ _RIHR-MOV-MO0166A 2041 H2 2 A 1 GB MO C LJ-i OPB RHR HEAT EXCHANGER A VENT______ ____ __ ___PASSIVE RHR-MOV-MO166B 2041 H3 2 A 1 GB MO C UJ-i OPB RHR HEAT EXCHANGER B VENT______PASSIVE RI-IR-MO V-MO 1 67A 2041 H2 2 A 1 GB MO C UJ- 1 OPB RH-R HEAT EXCHANGER A VENT______PASSIVE RHR-MOV-MO167B 2041 H3 2 A 1 GB MO C UJ-i OPB RHR HEAT EXCHANGER B VENT_____ _____ ___ __ ____ __ ___ PASSIVE RHR-MOV-MO274A 2040 Bl 10 A 2 GB MO C LT-2 OPB RHR-CV-26CV PASSIVE BYPASS SHi1 PIT 2Y PRESSURE ISOLATION VALVE,____ ___ _ __ __ _ ___RV-05, RG-01 RHR-MOV-M0274B 2040 B4 1 A 2 GB MO C LT-2 OPB RHR-CV-27CV PASSIVE BYPASS SH 2 PIT 2Y PRESSURE ISOLATION VALVE,____ ___ __ ____ ___ _ __ ___ RV-05, RG-01 RHR-MOV-920MV 201 D2 B 3 GT M C FSC CS STEAM SUPPLY TO AOG PIT 2Y UPSTREAM SHUTOFF, CSJ-06, Revision 0 Page 253 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RHR)VALVE GIG P&ID COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION ____ ___ ____ __ __ _ ___RG-0 1 RHR-MOV-921MV 2041 D1 A B 3 GT MO C FSC CS STEAM SUPPLY TO AOG PI!T 2Y DOWNSTREAM SHUTOFF, CSJ-06,____ ___ ____ __ __ _ ___RG-01 RHR-RV-10RV 2040 F8 2 C 1 RV SA C FVT APP I RHR PUMP A SUCTION RELIEF, S111 RG-01 RHR-RV-l1RV 2040 F5 2 C 1 RV SA C RVT APP I RJIR PUMP B SUCTION RELIEF, 5112 RG-01 RHR-RV-12RV 2040 118 2 C 1 RV SA C RVT APP I RHR PUMP C SUCTION RELIEF, 5111 RG-01 RHR-RV-13RV 2040 145 2 C 1 RV SA C RVT APP I RHR PUMP D SUCTION RELIEF, SH 2 RG-01 RHR-RV-14RV 2040 C4 2 C 1 RV SA C RVT APP I RHR LOOP A SUPPLY RELIEF, 5111 RG-01 RHR-RV-15RV 2040 C9 2 C 1 RV SA C RVT APP I RIIR LOOP B SUPPLY RELIEF,__ _ _ _ _ SH 2 __ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ RG-01 Revision 0 Pg 5 Page 254 Cooper Nuclear Station Fifthi Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RHLR)VALVE CIC P&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ RHR-RV- 17RV 2040 F8 2 C 1 RV SA C RVT APP I SHUTDOWN COOLING SHi 1 ______ ______ SUPPLY RELIEF, RG-01 RLFR-RV-20RV 2041 G2 2 A/C I RV SA C RVT APP I RIHR HEAT EXCHANGER A SHELL LJ__ -1 OPB SIDE RELIEF, RG-01 RHR-RV-21RV 2041 G2 2 A/C 1 RV SA C RVT APP I RHR HEAT EXCHANGER B SHELL__________ __L ___ -i1 OPB SIDE RELIEF, RG-01 Revision 0 Pg 5 Page 255 Cooper Nuclear Station Fifth Interval lnservice Testing Pro gram for Pumps and Valves SYSTEM: SERVICE WATER (SW)VALVE CIC P&ID P&ID ISI 1ST VALVE VALVE ACT NORM ITEST JTEST I NOTES/DESCRIPTION I_____ __COOR CLASS CAT SIZE jTYPE TYPE _POS RQMT jFREQ SW-AOV-854AV 2036 Fl11 3 B 2 BAL AO C IPIT 2Y SW PASSIVE RAD MONITOR SH 1______ __ SAMPLE RETURN, RG-01 SW-AOV-855AV 2036 F10 B 2 BAL AO C PIT 2Y SW PASSIVE RAD MONITOR SH 1 ____SAMPLE RETURN, RG-01 SW-AOV-TCV451A 2036 E7 3 B 12 GB AO O/T PS0 Q REC HEAT EXCHANGER A OUTLET SHi1 FSO Q/CS TEMPERATURE CONTROL VALVE, FST Q/CS CSJ-07, RG-01, TP-04_____ ~~~PIT 2Y __ _ _ _ _ _ _ _ _ _ _ _SW-AOV-TCV451B 2036 F7 3 B 12 GB AO O/T PSO Q REC HEAT EXCHANGER B OUTLET SHi 1ESO Q/CS TEMPERATURE CONTROL VALVE, FST Q/CS CSJ-07, RG-01, TP-04______ ______ ______ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _SW-AOV-TCV452A KSV- D2 3 B 2 1/2 GL AO 0 SKID Q DG1 TURBO INTER COOL TCV, TP-47-8 ________ ___ __ _ 05, RG-01 SW-AOV-TCV452B KSV- D2 3 B 2 1/22 GL AO 0 SKID Q DG2 TURBO INTER COOL TCV, TP-47-8 ____ ___ ____ __ _ 05, RG-01 SW-AOV-2797AAV KSV- H5 3 B 6 BTF AO 0 SKID Q DG1 SUPPLY, TP-05, RG-01 47-8 ____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _SW-AOV-2797BAV KSV- H5 3 B 6 -BTF AO 0 SKID Q DG2 SUPPLY, TP-05, RG-O1 47-8 _ _ _ ______ _ _ _ _ _ _ _____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _SW-CV-ARA 23 ES 3 C 6 CK SC C COD CVCM SW AIR RELEASE, RG-01, TP-01 Revision 0 Page 256 Cooper Nuclear Station Fifth lnterval Inservice Testing Program for Pumps and Valves SYSTEM: SERVICE WATER (SW)VALVE CIC P&ID P&ID IS1 1IST VALVE VALVE JACT NORM TEST TEST- NOTES/DESCRIPTION ___________COOR CLASSJCAT SIZE TYPE_+/-TYPE POS RQMT FREQ SH 1 ___ ___ __ CL CVCM SW-CV-ARB 2036 C5 3 C 6 CK SC C ,COD CVCM SW AIR RELEASE, RG-01, TP-01_______SHi 1 _______ CCL CVCM SW-C V-10CV 2006 All 3 C 20 CK-D SA 0 ESO Q SW PUMP A DISCHARGE, RG-01__ _ _ _ _ _ SH 1 _ _ _ _ _ _ __FSC Q SW-CV-1 1CV 2006 A8 3 C 20 CK-D SA 0 FSO Q SW PUMP B DISCHARGE, RG-01 Sil __ _ __ _ FSC Q _ _ _ _ _ _ _ _ _ _SW-CV-12CV 2006 A10 C 20 CK-D SA 0 FSO Q SW PUMP C DISCHARGE, RG-01__ _ _ _ __ _ H1-1 __ _ _ _ _ FSC Q _ _ _ _ _ _ _ _ _ _ _ _ _SW-CV-13CV 2006 A7 3 C 20 CK-D SA 0 FSO Q SW PUMP D DISCHARGE, RG-01__ _ _ _ _ _ SHi 1 _ __ _ _ _ FSC Q SW-CV-19CV 2006 F9 3 C 14 CK-T SA C FSO Q RHIR SW BOOSTER PUMP A_______SH 4 _______________FSC Q DISCHARGE, RG-01 SW-CV-20CV 2006 C9 3 C 14 CK-T SA C FSO Q RHR SW BOOSTER PUMP B________SH 4 ______FSC Q DISCHARGE, RG-01 SW-CV-21CV 2006 D9 3 C 14 CK-T SA C FSO Q RI-R SW BOOSTER PUMP C S______ H 4 _____________FSC Q DISCHARGE, RG-01 SW-C V-22CV 2006 A9 3 C 14 CK-T SA C FSO Q RIIR SW BOOSTER PUMP D S_______ H 4 _____________FSC Q DISCHARGE, RG-01 SW-CV-27CV 206 C .14 CK-D OA 0 COF CVCM REC HEAT EXCHANGER B SUPPLY, SiCCD CVCM RG-01, TP-01 Revision 0 Page 257 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: SERVICE WATER (SW)VALVE CIC P&ID 1P&ID ISI IST jVALVE VALVE ACT 1NORM TEST 1TEST NOTES/DESCRIPTION SW-CV-28CV 2036 D2 3 C 14 CK-D SA 0 COF CVCM REC HEAT EXCHANGER A SUPPLY,________SH 1 __________CCD CVCM RG-01, TP-01 SW-CV-35CV 2077 D1 C 10 CK-S SA 0 F:SO Q DG1 SUPPLY, RG-01___ __ ___ __ __FSC Q sw-cv-36cv 2077 D2 3 C 10 CK-S SA 0 FSO Q DG1 SUPPLY, RG-01__ ____ __FSC Q SW-CV-37CV 2077 D5 3 C 10 CK-S SA 0 IFSO Q DG2 SUPPLY, RG-01___ _ _ __ __ __ __ __FSC Q SW-CV-38CV 2077 D5 C 10 CK-S SA 0 FSO Q DG2 SUPPLY, RG-01___ _ __ __ ____ FSC Q SW-CV-86CV 2006 D10 3 C 0.5 CK-P SA 0 COF CVCM SW PUMP A & C CHEMICAL SH 1 __ ___ __ CCL CVCM INJECTION, RG-01, TP-01 SW-CV-87CV 2006 El0 C 0.5 CK-P SA 0 COF CVCM SW PUMP A & C CHEMICAL SH 1___ __ CCL CVCM INJECTION, RG-01, TP-01 SW-C V-88MV 2006 C10 C 0.5 CK-P SA 0 COF CVCM SW PUMP B & D CHEMICAL SH 1___ CCL CVCM INJECTION, RG-01, TP-01 SW-CV-89MV 2006 C10 C 0.5 CK-P SA 0 COF CVCM SW PUMP B & D CHEMICAL SH 1___ __ CCL CVCM INJECTION, RG-01, TP-01 SW-MOV-36MV 2006 El0 B 24 BTF MO 0 FSC Q SW LOOP CRITICAL HEADER_ _ _ _ _ _ _ SHi ___1______ ____ ___ _ _ _ PIT 2Y ISOLATION, RG-01 SW-MOV-37MV 206 EO 3 B 24 BTF MO O FSC Q SW PUMPS CROSSTIE, RG-01 Revision 0 Page 258 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: SERVICE WATER (SW)VALVE CIC IP&ID P&ID 1SI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION j___COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ _____________ SH 1_ _ __ _ _ PIT 2Y _ _ _ _ _ _ _ _ _ _ _SW-MOV-MO89A 2036 C7 3 B 18 GB MO C PSO Q RH{R HEAT EXCHANGER A SW Sill FSO RE OUTLET, ROJ-l0, RG-01_____ _ ___ ___ _ ____ _ _____ ______ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _SW-MOV-MO89B 2036 Cl10 B 18 GB MO C PSO Q RIIR HEAT EXCHANGER B SW Sill FSO RF OUTLET, ROJ-10, RG-0l_____ _____ ______ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _SW-MOV-650MV 2036 E3 3 B 18 BTF MO 0 IFSO Q REC HEAT EXCHANGER A Sill FSC Q OUTLET, RG-01_____ ~~~~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _SW-MOV-651MV 2036 C3 3 B 18 BTF MO 0 FSO Q REC HEAT EXCHANGER B OUTLET, Sill FSC Q RG-01_____ ______ ~~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _SW-MOV-886MV 2036 D1 B 4 GT MO C FSO Q EMERGENCY SUPPLY TO REC Sil 1___ __ PIT 2Y NORTH CRITICAL LOOP, RG-01 SW-MOV-887MV 2036 D1 B 4 GT MO C FSO Q EMERGENCY SUPPLY TO REC Sil 1 __ ___ __ PIT *2Y SOUTH CRITICAL LOOP, RG-01 SW-MOV-888MV 2036 E4 3 B 4 GT MO C FSO Q EMERGENCY RETURN FROM REC Sil 1__ PIT 2Y NORTH CRITICAL LOOP, RG-01 SW-MOV-889MV 2036 C4 3 -B 4 GT MO C FSO Q EMERGENCY RETURN FROM REC________SHl 1 __ ________________ _ PIT 2Y SOUTH CRITICAL LOOP, RG-01 SW-RV-12RV 20 G8 3 C 3/4 RV SA C RVT APP I SWBP 1A SEAL WATER RELIEF, Revision 0 Page 259 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: SERVICE WATER (SW)VALVE CIC P&ID P&ID 1ISI IST VALVE VALVE fACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE [TYPE POS RQMT FREQ_____________ _ _ _ _ _ _ SH 4 _ __ __ _ __ RG-01 SW-RV-13RV 2006 C8 3 C 3/4 RV SA C RVT APP I SWBP 1C SEAL WATER RELIEF, S__ _ __ H 4 __ __ _ _ _ _ _ _ _ _ _ _ RG-01 SW-RV-14RV 2006 E8 C 3/4 RV SA d RVT APP I SWBP 1B SEAL WATER RELIEF, SH 4 __ _ _ _ _ _ __ _ _ RG-01 SW-RV-15RV 2006 B8 3 C 3/4 RV SA C RVT APP I SWBP 1D SEAL WATER RELIEF, SH 4 __ _ _ _ _ _ RG-01 SW-V-640 2006 E7 3 B 1 BAL MA 0 FSC 2Y SWBP C SEAL WATER RIVERWELL SH 4 ____ _______SHUTOFF, RG-01 SW-V-649 2006 G7 3 B 1 BAL MA 0 FSC 2Y SWBP A SEAL WATER RIVERWELL SH 4 _______ ________ SHUTOFF, RG-01 SW-V-656 2006 B7 3 B 1 BAL MA 0 FSC 2Y SWBP D SEAL WATER RI VER WELL SH 4 ____ _______SHUTOFF, RG-01 SW-V-665 2006 D7 3 B 1 BAL MA 0 FSC 2Y SWBP B SEAL WATER RIVERWELL SH 4 _ _ __ _ __ _ _ _ _ _ _ _ _ _ __ _ __ _ _ _ _ SHUTOFF, RG-01 SW-V-1422 2006 F8 3 C 3/4 GB MA C ESO 2Y SWBP A GLAND WATER SUPPLY, SH 4 __ __ _ _ _ _ _ _ _ _RG-01 SW-V-1424 2006 G8 3 C 3/4 GB MA C FSO 2Y SWBP A GLAND WATER FLOW SH 4 _______ __CONTROL, RG-01 SW-V-1426 2006 C8 3 C 3/4 GB MA C FSO 2Y SWBP B GLAND WATER SUPPLY, S__ _ __ H 4 _ _ __ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ __ RG-01 Revision 0 Page 260 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: SERVICE WATER (SW)* VALVE CIC P&ID P&ID IS I5 ST VALVE VALVE ACT INORM TEST ITEST NOTES/DESCRIPTION COOR jCLASS* CAT SIZE TYPE TYPE POS RQMT FREQ _________SW-V-1428 2006 D8 3 C 3/4 GB MA C ESO 2Y SWBP B GLAND WATER FLOW S______ H 4 _____________CONTROL, RG.-01 SW-V-1430 2006 E8 3 C 3/4 GB MA C FSO 2Y SWBP C GLAND WATER SUPPLY, SH14 ___ __RG-01 SW-.V-1432 2006 E8 C 3/4 GB MA C FSO 2Y SWBP C GLAND WATER FLOW SH 4 CONTROL, RG-01 SW-V-1434 2006 B8 3 C 3/4 GB MA C FSO 2Y SWBP D GLAND WATER SUPPLY, SH 4 ___ __RG-01 SW-V-1436 2006 B8 3 C 3/4 GB MA C FSO 2Y SWBP D GLAND WATER FLOW SH 4 CONTROL, RG-01 Revision 0 Pg 6 Page 261 Cooper Nuclear Station Fifth lnterval Inservice Testing Program for Pumps and Valves SYSTEM: STANDBY GAS TREATMENT (SGT)VALVE CIC P&ID JP&IDj ISI IST IVALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COO CLASS CAT SIZE TYPE TYPE POS RQMT FREQ SGT-AOV-249AV 2037 C2 A B 12 BTF AO C FSO Q SGT UNIT A INLET, RG-O1, TP-04Q FST q___ __ ___ __ _ _ __ _ ___ _ _____ __PIT 2Y SGT-AOV-250AV 2037 G2 A B 12 BTF AO C FSO Q SGT UNIT B INLET, RG-01, TP-04 FSC Q FST Q___ _ __ ____ __PIT 2Y SGT-AOV-251IAV 2037 C6 A B 12 BTF AO C FSO Q SGT UNIT A DISCHARGE, RG-01, FST Q TP-04___ _ __ __ __ ____ __ PIT 2Y SGT-AOV-252AV 2037 G6 A B 12 BTF AO C FSO Q SGT UNIT B DISCHARGE. RG-01, FST Q TP-04__ ___ __ _PIT 2Y SGT-AOV-255AV 2037 A4 A B 10 BTF AO C PIT 2Y SOT UNIT A PASSIVE BYPASS,__________RG-0 1 SGT-AOV-256AV 2037 E4 A B 10 BTF AO C PIT 2Y SGT UNIT B PASSIVE BYPASS,____ ___ ___ __ __ ____ __ ______ _ _ ___ RG-01 SGT-AOV-270AV 2037 C1 A B 10 BTF AO C FSO Q SGT UNIT A DILUTION AIR FST Q SHUTOFF, RG-01, TP-04___ _ __ __ __ __ __ __ __ __PIT 2Y SGT-AOV-271AV 23 01 A B 10 BTF AO C FO Q SGT UNIT B DILUTION AIR Revision 0 Page 262 Cooper Nuclear Station Ffith lnterval Inservice Testing Program for Pumps and Valves SYSTEM: STANDBY GAS TREATMENT (SGT)VALVE CIC P&ID P&ID 1 S 1 I ST VALVE VALVE ACT NORM TEST TEST 1 NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ _____________ FST Q SHUTOFF, RG-.0I, TP-04___ __ __ __ __ __ _ __ __PIT 2Y SGT-AOV- 2037 C7 A B 10 BTF AO C FSO Q SGT UNIT A DISCHARGE DPCV546A FST Q DIFFERENTIAL PRESSURE___ __ __ ___ _ __ __ __ ____________ ____ __ PIT 2Y CONTROL, RG-01, TP-04 SGT-AOV- 2037 E7 A B 10 BTF AO C FzSO Q SGT UNIT B DISCHARGE DPCV546B FST Q DIFFERENTIAL PRESSURE___ ________PIT 2Y CONTROL, RG-01, TP-04 SGT-CV-14CV 2037 C6 A C 10 CK-D SA C FSO CS SGT UNIT AFAN EXHAUST,____ ____ ___ ___ ___ FSC CS CSJ-01, RG-01 SGT-CV-15CV 2037 G6 A C 10 CK-D SA C FSO CS SGT UNIT B FAN EXHAUST, FSC CS CSJ-01, RG-01 Revision 0 Page 263 Revision 0 Page 263 Cooper Nuclear Station Ffifh Interval lnservice Testing Program for Pumps and Valves SYSTEM: STANDBY LIQUID CONTROL (SLC)1 IP&ID ISI IST 1VALVE VALVE ACT NORM TEST TEST VALVE CIC P&D COOR CLASS CAT jSIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION SLC-CV- 10CV 2045 E9 A C 11/2 CK-L SA C COF CVCM SLC PUMP A DISCHARGE______ _SH 2 _______CCR CVCM CHECK, RG-01 SLC-CV-11CV 2045 F9 A C 11/2/ CK-L SA C COF CVCM SLC PUMP B DISCHARGE S______ H 2 _ __CCR CVCM CHECK, RG-01 SLC-CV- 12CV 2045 E8 1 A/C 11/2 CK-L SA C J-i1 OPB SLC INJECTION LINE SH 2 COF CVCM OUTBOARD CHECK, RG-01___________ ____ ____CCL CVCM SLC-CV-13CV 2045 E7 1 A/C 11/2 CK-L SA C UJ-i OPB SLC INJECTION LINE INBOARD SH 2 COF CVCM CHECK, RG-01___________ CL CVCM SLC-RV-10RV 2045 D1O A C 3/4 RV SA C RVT APP 12 SLC PUMP A DISCHARGE________SH 2 ____RELIEF, RG-01 SLC-RV-1 1RV 2045 G9 A C 3/4 RV SA C 7RVT APP I 2 SLC PUMP B DISCHARGE S_______ H 2 _______RELIEF, RG-01 SLC-SQBV-14A 2045 E8 A D 11/2 SHR CH C EX 2Y SLC EXPLOSIVE VALVE A, RG-01 SH2 2_ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _SLC-SQBV-14B 2045 E8 A D 11/2 SHR CH C EX 2Y SLC EXPLOSIVE VALVE A, RG-01 5112 Revision 0 Pg 6 Page 264 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: STATION AIR (SA)VALVE CIC IP&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR________ __ CLASS CAT SIZE TYPE -TYPE POS RQMT FREQ SA-V-647 2010 G4 2 A 1GB MA C LJ-1 OPB PASSIVE DRYWELL OUTBOARD________ SH 3 ___ ___ _______ ___ ____ ___ ___SUPPLY ISOLATION SA-V-648 2010 G4 2 A 1GB MA C LJ-i OPB PASSIVE DRYWELL INBOARD SH 3 SUPPLY ISOLATION 1. License Amendment 234 to the CNS Facility Operating License requires that the SLC system be maintained in the Augmented IST Program for the Alternate Source Term Function 2. License Amendment 176 to the CNS Facility Operating License requires that the control of the SLC relief valve testing (and as-left set point) be maintained in the Augmented 1ST Program. This amendment requires that the USAR describe the as-left set point (1540 +/-- 1%) and that future changes be evaluated per 10CFR50.59. The as-found criteria is 1478 to 1602 psig per USAR section III-9.4.Revision 0 Page 265 NLS2016012 Page 1 of 7 Enclosure 2 Cooper Nuclear Station Fifth Interval Inservice Examination and Testing Program for Snubbers Nebraska Public Power District Cooper Nuclear Station Fifth Interval*Inservice Examination &Testing Program for Snubbers Revision 0 Cooper Nuclear Station P0 BOX 98 Brownville, NE 68321-0098 Commercial Operation Date: July 1, 1974 Snubber Engineer: Program Backup / Supervisor: Date: _-______Date: __ -_ ____Date: 9,46 EP & C Manager: I Cooper Nuclear Station Fifth Interval Inservice Examination & Testing Pro gram for Snubbers TABLE OF CONTENTS SECTION TITLE

1.0 INTRODUCTION

1.1 Purpose 1.2 Scope 2.0 VISUAL EXAMINATION REQUIREMENTS 2.1 General 2.2 Preservice Examinations 2.3 Inservice Examinations 3.0 FUNCTIONAL TESTING REQUIREMENTS 3.1 Preservice Operational Readiness Testing 3.2 Inservice Operational Readiness Testing 4.0 SERVICE LIFE MONITORING REQUIREMENTS 5.0 CNS PROCEDURES Revision 0 Page 2 Revision 0 Page 2 Cooper Nuclear Station Ffith Interval Inservice Examination & Testing Program for Snubbers 1.0 IN~TRODUCTION 1.1 Purpose This document establishes the Cooper Nuclear Station (CNS) Fifth 120-Month Interval Inservice Testing (IST) Program requirements for the American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 dynamic restraints (snubbers). The Snubber Program requirements include Visual Examinations, Functional testing, and Service Life Monitoring. The CNS Fifth 120-Month Interval Snubber Inservice Testing Program Plan will be applicable during the following time period.Begin: 03/01/2016 End: 02/28/2026 The objective of the Snubber Program is to provide reasonable assurance that the snubbers within this plan are capable of performing their intended function(s) during normal operations and seismic events.1.2 Scope The 1ST Program scope for dynamic restraints applies to the Class 1, 2, and 3 dynamic restraints that meet the scope statements outlined in the ASME OM Code of record. The CNS Code of record for the Fifth 120-Month Interval is the 2004 Edition through the 2006 addenda of the ASME OM Code.This document is based on the following Subsections of the 2004 Edition through the 2006 Addenda of the ASME OM Code.*Subsection ISTA, "General Requirements" ISTA-1 100 states that "These requirements apply to" (c) "dynamic restraints (snubbers) used in systems that perform one or more of the three functions identified in subpara. ISTA-1 100(a) [shutting down a reactor to the safe shutdown condition, in maintaining the safe shutdown condition, or in mitigating the consequences of an accident], or to ensure the integrity of the reactor coolant pressure boundary"* Subsection ISTD, "Preservice and inservice Examination and Testing of dynamic Restraints (Snubbers) in Light-Water Reactor Nuclear Power Plants." Cooper Nuclear Station was designed and licensed to operate with the Hot Shutdown condition defined as the "safe" shutdown condition. CNS currently has a total of 208 snubbers in this program, 155 Mechanical snubbers and 53 hydraulic snubbers. These snubbers and their locations are listed in CNS Maintenance Procedure, 7.2.34.1, "Snubber Examination". Revision 0 Page 3 Revision 0 Page 3 Cooper Nuclear Station Ffifth Interval Inservice Examination & Testing Pro gram for Snubbers 2.0 VISUAL EXAMINATIONS 2.1 General The examination boundary shall include the snubber assembly from pin to pin, inclusive (ISTD-31 10).Typical preservice or inservice examination checklist items to be considered are listed in Nonmandatory Appendix B of the ASME OM Code.2.2 Preservice Examinations For new and modified systems, preservice examinations shall be performed after placing the systems in service. These examinations should be identified within the modification package implementing the new and/or modified system. The preservice examination requirements of ISTD-4 100 shall be met.2.3 Inservice Examinations Snubbers shall be visually examined on the required schedule and evaluated to dctcrmine their operational readiness (ISTD-4200). The inservice examination shall be a visual examination to identify physical damage, leakage, corrosion, or degradation that may have been caused by environmental exposure or operating conditions. External characteristics that may indicate operational readiness of the snubber shall be examined. An examination checklist shall be used (ISTD-42 10).Snubber installations shall meet all the requirements of ISTD-423 1 (restrained movement), ISTD-4232 (thermal movement), and ISTD-4233 (design-specific characteristics). CNS is utilizing approved Code Case OMN-1 3 (2004 Edition), "Requirements for Extending Snubber Inservice Visual Examination Interval at LWR Power Plants," as listed in Table 1 of Regulatory Guide 1.192, Revision 1 (August 2014), "Operation and Maintenance Code Case Acceptability, ASME OM Code." This Code Case allows for the extension of visual examinations beyond the frequencies specified in ISTD for mechanical and hydraulic snubbers to at least once every 10 years. All requirements of this NRC approved version of OMN-1 3 must be met.Snubbers at CNS are examined in accordance with CNS procedures 7.2.34.1 or 7.2.34.2.A snubber that requires further evaluation or is classified as unacceptable during inservice examination may be tested in accordance with the requirements ISTD-521 0.3.0 FUNCTIONAL TESTING REQUIREMENTS 3.1 Preservice Operational Readiness Testing Preservice Operational readiness testing shall be performed on all snubbers. Testing may be performed at the manufacturer's facility (ISTD-5 110).Revision 0 Pg Page 4 Cooper Nuclear Station Fifth Interval Inservice Examination & Testing Pro gram for Snubbers Test parameters shall meet the requirements of ISTD-5 120. If a failure occurs, corrective action shall meet the requirements of ISTD-5 130.3.2 Inservice Operational Readiness Testing Snubbers shall be tested for operational readiness during each fuel cycle. Functional tests at CNS are completed in accordance with the 10% Testing Sample Plan (ISTD-5300). Test parameters are in accordance with ISTD-52 10.The CNS snubber population consists of hydraulic and mechanical snubbers. These snubbers are separated into Design Test Plan Groups (DTPG) by snubber type (hydraulic versus mechanical) and size in accordance with ISTD-5252 (see below).DTPG Population Description Number of Snubbers Sample Plan 1 PSA-3 Mechanical Snubbers 18 10%2 PSA-l0 Mechanical Snubbers 89 10%3 PSA-35 Mechanical Snubbers 48 10%4 ANVIIL/Grinnell Hydraulic 53 10%_______Snubbers _________________ Testing shall be performed during normal operation, or during system or plant outages (ISTD-5200). However, snubber testing may begin no earlier than 60 days before a scheduled refueling outage (ISTD-5240). The initial sample shall be 10% of the DTPG, composed according to either ISTD-5311l(a) or ISTD-531 1(b).Snubbers that do not meet test requirements specified in ISTD-52 10 or ISTD-5230 shall be evaluated to determine the cause of the failure (ISTD-527 1). Failure mode groupings (FMGs) should be determined in accordance with ISTD-5272, as applicable. The FMG boundaries shall be applied per ISTD-5273 and utilized with the 10% sample plan per ISTD-5300. Snubbers will generally be tested at CNS in accordance with CNS procedures 7.2.34.7 and 7.2.34.8. However, snubbers can be functionally tested by vendors, if necessary. 4.0 SERVICE LIFE MONITORING REQUIREMENTS Initial snubber service life shall be predicted based on manufacturer's reconmnendation or design review (ISTD-6 100).Service life shall be evaluated at least once each fuel cycle, and increased or decreased, if warranted (ISTD-6200). This is typically done by reviewing the examination and functional test results at the completion of the refueling outage campaign.Snubbers shall be replaced or reconditioned, as required, to ensure that the service life is not exceeded before the next scheduled system or plant outage.If necessary, technical justification shall be documented for extending the service life to or beyond the next scheduled system or plant outage.Revision 0 Pg Page 5 Cooper Nuclear Station Fifthi Interval Inservice Examination & Testing Program for Snubbers Causes for any examination or testing failures shall be deternined and considered in establishing or re-establishing service life.5.0 CNS PROCEDURES 5.1 Engineering Procedure 3.39, "Snubber Program" 5.2 Maintenance Procedure 7.2.34.1, "Snubber Examination" 5.3 Maintenance Procedure 7.2.34.2, "Pipe Snubbers Removal and Installation" 5.4 Maintenance Procedure 7.2.34.3, "Grinnell Figure 200/201 Hydraulic Snubber Maintenance" 1 5.5 Maintenance Procedure 7.2.34.4, "Pacific Scientific PSA-3 and PSA-10 Snubber Maintenance 5.6 Maintenance Procedure 7.2.34.5, "Pacific Scientific PSA-35 Snubber Maintenance 5.7 Maintenance Procedure 7.2.34.7, "Grinnell Figure 200/201 Hydraulic Snubber Functional Test" 5.8 Maintenance Procedure 7.2.34.8, "Pacific Scientific Snubber Functional Test" 5.9 Administrative Procedure 0.30, "ASME Section XI Repair/Replacement and Temporary Code and Non-Code Repair Procedure" 5.10 Surveillance Procedure 6.SNUB.601, "Snubber Operability" 5.11 Surveillance Procedure 6.SNUB.602, "Snubber Service Life Evaluation" Revision 0 Page 6 Revision 0 Page 6 N!Nebraska Public Power District Always the're when you need us NLS2016012 March 9, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Fifth Ten-Year Interval Inservice Testing Program Cooper Nuclear Station, Docket No. 50-298, DPR-46

Dear Sir or Madam:

The purpose of this correspondence is to provide Nebraska Public Power District's Inservice Testing (IST) Program Plan for the Fifth Ten-Year Interval. Submittal of the plan is in accordance with the requirements of the American Society of Mechanical Engineers Code for Operations and Maintenance of Nuclear Power Plants, Subsection ISTA-3200(a),"Administrative Requirements." As documented within the IST Program, the relief requests included in the 1ST Program have been previously submitted to the Nuclear Regulatory Commission and approved for use. The Fifth Ten-Year Interval for Cooper Nuclear Station began on March 1, 2016 and concludes on February 28, 2026.Enclosure 1 to this letter contains the "Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves." Enclosure 2 contains the "Cooper Nuclear Station Fifth Interval Inservice Examination and Testing Program for Snubbers." There are no regualatory commitments contained in this letter.Should you have any questions regarding the information contained in this submittal, please contact me at (402) 825-2788.Sincerely, Licensin Manager-/dv COOPER0 NUCLEARSTATION P.O. Box 98 / Browrivi/Ie, NE 68321-0098 Telephone: (402) 825-3811 / Fax:-(402) 825-5211 www.nppd.comn NLS2016012 Page 2 of 2

Enclosures:

1. Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves 2. Cooper Nuclear Station Fifth Interval Inservice Examination and Testing Program for Snubbers cc: Regional Administrator w/enclosures USNRC -Region IV Cooper Project Manager w/enclosures USNRC -NRR Plant Licensing Branch IV-2 Senior Resident Inspector w/enclosures NPG Distribution wlo enclosures CNS Records w/enclosures ATTACHMENT 9.4 REGULATORY SUBMITTAL REVIEW Sheet I of 2 Letter #: NLS2016012 Response Due: 3/1/2016

Subject:

Fifth Ten-Year Interval IST Program Submittal to NRC Date Issued for Review: 2/22/2016 Correspondence Preparer I Phone #: David Van Der Kamp / x2904 Section I Letter Concurrence and Aareement to Perform Actions POSITION I NAME Action Signature/Date (concurrence, (sign, interoffice menmo, e-mail, or telecom)certification, etc.)EP&C Eng -Tom Robinson Validation LIC Spec -David Van Der Kemp Validation / ,¢EP&C Supv -Stan Domikaitis Concurrence Z./_6/ =EP&C Mgr -Troy Barker Concurrence _ /DOE -Dan Buman Concurrence ' _ -// // , COMMENTS Section II Correspondence Screening Does this letter contain commitments? If "yes," identify the commitments with due Yes LI dates in the submittal and in Section III. No [If "yes," and the due date is > 3 months, is tracking of interim milestones required? Yes LI No LI Does this letter contain any information or analyses of new safety issues performed at NRC Yes LI request or to satisfy a regulatory requirement? If "yes," reflect requirement to update the No [USAR in Section II1. __ ___Does this letter require any document changes (e.g., procedures, DBDs, USAR, TS Bases, Yes LI etc.), if approved? If "yes," indicate in Section III an action for the responsible No [department to revise the affected documents. (The Correspondence Preparer may indicate the specific documents requiring revision, if known or may initiate an action for review.)Does this letter contain information certified accurate? If "yes," identify the information Yes LI and document certification in an attachment. (Attachment 9.5 must be used.) No ____ ATTACHMENT 9.4 REGULATORY SUBMITTAL REVIEW Sheet 2 of 2 Does this letter require posting per 1OCFR1 9? If "yes," ensure posting after submittal. Yes LI No [Does this letter contain Safeguards Information? If "yes," do not scan to CNS intranet Yes LI and ensure handling per Procedure 1.2. No [Does this letter contain information to be withheld from public disclosure (e.g., Proprietary or Yes LI Non-Safeguards Security-Related Information)? If "yes," do not scan to CNS intranet No [and ensure appropriate marking and handling. ___Section III Actions and Commitments Required Actions/Tracking Numbers Due Date Responsible Dept.Note: Actions needed upon approval should be captured in the appropriate action tracking system None CommitmentslCommitment Numbers Due Date Responsible Dept.Note: Enter the commitments into the commitment management system.None Section IV Final Document Signoff for Submittal Correspondence Preparer David Van Der Kamp /v Final Submittal Review (optional) See F&F v L Responsible Department Head SeLte Director of Nuclear Safety N/A Assurance (as applicable) ATTACHMENT 9.4 REGULATORY SUBMITTAL REVIEW Sheet 1 of 2 Letter#: NLS2016012 Response Due: 3/1/2016

Subject:

Fifth Ten-Year Interval IST Program Submittal to NRC Date Issued for Review: 2/22/2016 Correspondence Preparer / Phone #: David Van Der Kamp / x2904 Section I Letter Concurrence and Aareement to Perform Actions POSITION / NAME Action Signature/Date (concurrence, (sign, interoffice memo, e-mail, or telecom)certification, etc.)EP&C Eng -Tom Robinson Validation , LIC Spec -David Van Der Kamp Validation EP&C Supv -Stan Domikaitis Concurrence z/. /EP&C Mgr -Troy Barker Concurrence DOE -Dan Buman Concurrence __ _// / , COMMENTS Section II Correspondence Screenin~Does this letter contain commitments? If "yes," identify the commitments with due Yes LI dates in the submittal and in Section III. No [If "yes," and the due date is > 3 months, is tracking of interim milestones required? Yes LI No LI Does this letter contain any information or analyses of new safety issues performed at NRC Yes LI request or to satisfy a regulatory requirement? If "yes," reflect requirement to update the No [USAR in Section IlL.Does this letter require any document changes (e.g., procedures, DBDs, USAR, TS Bases, Yes LI etc.), if approved? If "yes," indicate in Section III an action for the responsible No [department to revise the affected documents. (The Correspondence Preparer may indicate the specific documents requiring revision, if known or may initiate an action for review.)Does this letter contain information certified accurate? If "yes," identify the information Yes LI and document certification in an attachment. (Attachment 9.5 must be used.) No ____ .-.ATTACHMENT 9.4 REGULATORY SUBMITTAL REVIEW Sheet 2 of 2 Does this letter require posting per 1OCFR1 9? If "yes," ensure posting after submittal. Yes LI No [Does this letter contain Safeguards Information? If "yes," do not scan to CNS intranet Yes Ii and ensure handling per Procedure 1.2. No ____Does this letter contain information to be withheld from public disclosure (e.g., Proprietary or Yes LI Non-Safeguards Security-Related Information)? If "yes," do not scan to CNS intranet No [and ensure appropriate marking and handling. ___ __Section III Actions and Commitments Required Actions/Tracking Numbers Due Date Responsible Dept.Note: Actions needed upon approval should be captured in the appropriate action tracking system None CommitmentslCommitment Numbers Due Date Responsible Dept.Note: Enter the commitments into the commitment management system.__________ None Section IV Final Document Signoff for Submittal Correspondence Preparer David Van Der Kamp ( f Final Submittal Review (optional) See F&F .Responsible Department Head SeLte Director of Nuclear Safety N/A Assurance (as applicable) NLS2016012 Page 1 of 266 Enclosure 1 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Nebraska Public Power District Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Revision 0 Cooper Nuclear Station P.O. Box 98 Brownville, NE 68321-0098 Commercial Operation Date: July 1, 1974___________-_ d Date: _ _ _ __01#-/IST Engineer: Program Backup / Supervisor: Date: )j-________ Date: _-_/_____EP & C Manager:-I Cooper Nuclear Station Ffifh Interval Inserviee Testing Pro gram for Pumps and Valves TABLE OF CONTENTS SECTION

1.0 INTRODUCTION

1.1 Purpose 1.2 Scope 2.0 INSERVICE TESTING PLAN FOR PUMPS 2.1 Pump Inservice Testing Plan Description 2.2 Pump Plan Table Description 2.3 Measurement of Test Quantities 2.4 Allowable Ranges of Test Quantities 2.5 Instrument Accuracy 3.0 INSERVICE TESTING PLAN FOR VALVES 3.1 Valve Inservice Testing Plan Description 3.2 Valve Plan Table Description 4.0 ATTACHMENTS

1. System and P&ID Listing 2. Pump Relief Requests 3. Augmented Pump Relief Requests 4. Valve Relief Requests 5. Augmented Valve Relief Requests 6. General Relief Requests 7. Cold Shutdown Justifications
8. Refuel Outage Justifications
9. Technical Positions 10. Inservice Testing Pump Table 11. Inservice Testing Valve Table Revision 0 Pg Page 2 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves

1.0 INTRODUCTION

1.1 Purpose This document establishes the Cooper Nuclear Station (CNS) Fifth 120-Month Interval Inservice Testing (IST) Program requirements for the American Society of Mechanical Engineers (ASME)Code Class 1, 2, and 3 pumps and valves whose specific functions are required to either:* Shutdown the reactor to the safe shutdown condition,* Maintain the safe shutdown condition, and/or* To mitigate the consequences of an accident.This document will also establish the CNS Augmented Inservice Testing Program requirements for ASME non-Code Class pumps and valves whose specific functions will meet one or more of those listed above. Other pumps and valves may be included in the augmented scope at the discretion of CNS.The CNS Fifth 120-Month Interval Pump and Valve Inservice Testing Program Plan will be applicable during the following time period.Begin: 03/01/2016 End: 02/28/2026 1.2 Scope Per 10OCFR 50.55a(f)(4), pumps and valves that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the inservice test requirements set forth in the ASME OM Code and Addenda. Therefore, the regulatory IST Program scope for pumps and valves applies to the Class 1, 2, and 3 pumps and valves that meet the scope statements outlined in the ASME OM Code of record. The CNS Code of record for the Fifth 120-Month Interval is the 2004 Edition through the 2006 addenda of the ASME OM Code.This document is based on the following Subsections and Appendices of the 2004 Edition through the 2006 Addenda of the ASME OM Code:* Subsection ISTA, "General Requirements" ISTA-1 100 states that "These requirements apply to (a) pumps and valves that are required to perform a specific function in shutting down a reactor to the safe shutdown condition, in maintaining the safe shutdown condition, or in mitigating the consequences of an accident." Cooper Nuclear Station was designed and licensed to operate with the Hot Shutdown condition defined as the "safe" shutdown condition.

  • Subsection ISTB, "Inservice Testing of Pumps in Light-Water Reactor Nuclear Power Plants"* Subsection ISTC, "Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants" Revision 0 Pg Page 3 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves*Mandatory Appendix I, "Inservice Testing of Pressure Relief Devices in Light- Water Reactor Nuclear Power Plants"* Mandatory Appendix II, "Check Valve Condition Monitoring Program" Those pumps and valves that are ASME non-Code Class 1, 2, or 3, that meet the scope of the ASME OM Code, will be identified as augmented components within this document.

Other pumps and valves may be included in the augmented scope at the discretion of CNS. NRC approval for any deviation from the ASME OM Code, for the augmented components, is not required.The Cooper Nuclear Station Inservice Testing Program Basis Document includes the justification for inclusion of pumps and valves that are in the IST or Augmented IST Program scope and also many justifications for pumps and valves excluded from the IST and/or Augmented 1ST Program scope.The IST Check Valve Condition Monitoring (CVCM) Program Document contains the details on the check valves selected for this program and all the necessary requirements for implementation of this Program.The CNS IST Program Basis Document, IST CVCM Program Document, administrative procedures, surveillance testing procedures, and other records required to define and execute the Inservice Testing Program are all retained and available at Cooper Nuclear Station.Revision 0 Page 4 Revision 0 Page 4 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves 2.0 INSERVICE TESTING PLAN FOR PUMPS 2.1 Pump Inservice Testing Plan Description The CNS testing program for pumps meets the requirements of the ASME OM Code 2004 Edition through 2006 addenda, Subsection ISTB, "Inservice Testing of Pumps in Light- Water Reactor Nuclear Power Plants". Where these requirements have been determined to be impractical, specific requests for relief were written and are included in Attachment 2 (Pump Relief Requests) and Attachment 3 (Augmented Pump Relief Requests). NUREG 1482, Revision 2 has been used as guidance in the development of the 1ST Program Plan for pumps.2.2 Pump Plan Table Description The pumps included in the Cooper Nuclear Station IST Plan are listed in Attachment

10. The information contained within these tables identifies those pumps which are to be tested to the requirements of Subsection ISTB of the ASME GM Code, 2004 Edition through 2006 Addenda, the testing parameters and frequencies, and associated relief requests.

The headings for the pump tables are delineated below._System: The plant system in which the pump is located.Pump CIC: The pump component identification code.ISI Class: The ASME Inservice Inspection (ISI) classification of the component. Augmented components are classified "A".P&TD: The associated piping and instrumentation drawing number.P&ID Coor: The P&ID coordinate location of the pump.IST Group The Pump group as defined in ISTB-2000 Group A Continuous or routinely operated pumps Group B Standby pumps not operated routinely Parameters: The pump test quantities to be measured or observed. The test designators are as follows: DP Differential Pressure N Speed Pd Discharge Pressure Q Flow Rate V Vibration Frequency: The frequency of testing each pump. The following test designators are used: Q Once every 92 days (Quarterly) 6M Once every 6 months 2Y Once every 2 years Revision 0 Page 5 Rev&ion 0 Page 5 Cooper Nuclear Station Ffith Interval Inservice Testing Pro gram for Pumps and Valves 2.2 Pump Plan Table Description (continued) Notes: This column contains a brief component description, reference to any applicable relief request(s), and contains any other component-related infornation. Relief Requests are designated RP-XX for pumps or RG-XX for general Program Requirements. Augmented Relief Requests are designated ARP-XX. Station Technical Positions are designated TP-XX.2.3 Measurement of Test Quantities Speed (N) Per ASME OM Code ISTB-3530, rotational speed measurement of variable speed pumps shall be taken by a method which meets the requirements of paragraph ISTB-35 10.Pressure (DP, Pd) Differential pressure across a pump will be calculated from inlet and discharge pressure measurements or by direct differential pressure measurement. Discharge pressure will be by direct measurement of discharge pressure. Per NUREG 1482, revision 2, section 5.5.3, suction pressure may be calculated based on inlet tank or bay level.Flow Rate (0) Flow rate of the pump will be measured using a rate or quantity meter installed in the pump test circuit.Vibration (V) Pump vibration will be measured with a digital vibration meter in accordance with the applicable section of ASME CM Code ISTB -3540.2.4 Allowable Ranges of Test Quantities The applicable allowable ranges specified in ASME CM Code ISTB, Tables ISTB-5 121-1, ISTB-5221-1, and ISTB-5321 -1 will be used for differential pressure, flow and vibration measurements except where specific relief is requested and/or approved or design/licensing acceptance criteria is more restrictive than that prescribed in the tables. Should a measured test quantity fall outside the allowable range, corrective action per ASME OM Code ISTB-6200 shall be followed. Records shall be maintained in accordance with ASME OM Code ISTB-9000. 2.5 Instrument Accuracy and Range Requirements Allowable instrument (and loop, where applicable) accuracy's for pressure, flow rate, speed, vibration, and differential pressure are provided in ASME CM Code ISTB Table ISTB-35 10-1 and paragraph ISTB-35 10(a). If the accuracies of the station's instruments do not meet the requirements of this table/section, temporary instruments meeting those requirements in ASME OM Code ISTB Table ISTB-3510-1 will be used or approved relief shall be received.In determining instrument accuracy, the Code does not explicitly require the Licensee to consider physical attributes (such as orifice plate tolerances), tap locations, environmental effects (such as temperature, radiation or humidity), vibration effects (such as seismic), or process effects (such as temperature). This position is documented in NUJREG 1482, revision 2, paragraph 5.5.4.Revision 0 Page 6 Revision 0 Page 6 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Additionally, ASME OM Code ISTB-35 10(b) requires that the full-scale range of analog instruments be no more than 3 times the reference value. Digital instruments shall be selected such that the reference value does not exceed 90% of the calibrated range of the instrument. Revision 0 Page 7 Revision 0 Page 7 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves 3.0 INSERVICE TESTING PLAN FOR VALVES 3.1 Valve Inservice Testing Plan Description The CNS testing program for valves meets the requirements of the ASME OM Code 2004 Edition through 2006 addenda, Subsection ISTC "Inservice Testing of Valves in Light- Water Reactor Nuclear Power Plants"; Mandatory Appendix I "Inservice Testing of Pressure Relief Devices in Light- Water Reactor Nuclear Power Plants "; and Mandatory Appendix II "Check Valve Condition Monitoring Program. " Where these requirements are determined to be impractical, specific requests for relief have been written and are included in Attachment 4 (Valve Relief Requests) and Attachment 5 (Augmented Valve Relief Requests). 3.2 Valve Plan Table Description The table in Attachment 11 lists all ASME Class 1, 2, 3, and Augmented (A) Valves that have been scoped to be within this Program Plan, and have been assigned Valve Categories. In general, valves exempt per ASME OM Code ISTC- 1200 are not listed. The following information is included for each valve.System: The plant system in which the valve is located.Valve GIG: The valve component identification code.P&ID: The associated piping and instrumentation drawing number.P&ID Coor: The drawing coordinates location on the P&ID for the valve.ISI Class: The ASME Inservice Inspection (ISI) classification of the component. Augmented components are classified "A".IST Cat: The category(s) assigned to the valve based on the definitions per ASME GM Code ISTC-1300. The following categories are defined in the Code: Category A -Valves for which seat leakage is limited to a specific maximum amount in the closed position for fulfillment of their required function(s), as specified in ISTA-1 100.Category B -Valves for which seat leakage in the closed position is inconsequential for fulfillment of the required function(s), as specified in ISTA-1 100.Category C -Valves that are self-actuating in response to some system characteristic, such as pressure (relief valves) or flow direction (check valves) for fulfillment of the required function(s), as specified in ISTA-1100.Category D -Valves that are actuated by an energy source capable of only one operation, such as rupture disks or explosively actuated valves.Valve Size: The nominal size of the valve in inches Revision 0 Page 80 Page 8 Cooper Nuclear Station F~/lh Interval Inservice Testing Program for Pumps and Valves 3.2 Valve Plan Table Description (continued) Valve Type: The Valve Body Design as indicated by the following abbreviations: ANGLE BALL BUTTERFLY BALL CHECK DUAL DISK CHECK LIFT CHECK PISTON CHECK SWING CHECK DIAPHRAGM FLOAT VALVE GATE GLOBE PLUG PRESSURE REGULATING RUPTURE DISK RELIEF/SAFETY SOLENOID VALVE STOP VALVE STOP CHECK TILTING DISK CHECK SQUIB ANG BAL BTF CK-B CK-D CK-L CK-P CK-S DIA FOV GT GL PLG PRV RD RV SOV STOP S-CK CK-T SHR ACT Type: The type of Valve Actuator as indicated by the following abbreviations: AIR OPERATOR EXPLOSIVE CHARGE HYDRAULIC OPERATED MANUAL MOTOR OPERATION SELF ACTUATED SOLENOID OPERATOR PILOT ACTUATED AO EX HO MA MO SA SO PA NORMAL POS: The normal position of the valve during regular plant operation, specified as follows: 0 C T OPEN CLOSED THROTITLED TEST ROMT The test(s) that will be performed to fulfill the requirements of ASME OM Code ISTC. The definitions and abbreviations are identified below: Revision 0 Page 9 Revision 0 Page 9 Cooper Nuclear Station Fiflh Interval Inservice Testing Program for Pumps and Valves 3.2 Valve Plan Table Description (continued) LJ-1 Contaimnment Isolation Type C Valve Seat Leakage Test in accordance with the CNS 10CFR50 Appendix J Program (ISTC-3 620).LT- 1 Accumulator Check Valve Leakage Test LT-2 Category A Leak tests other than those already specified COD Check Valve Condition Monitoring open test per Disassembly and Examination CCD Check Valve Condition Monitoring closure test per Disassembly and Examination COF Check Valve Condition Monitoring open test per flow indication measurement CCF Check Valve Condition Monitoring backflow / closure test CCL Check Valve Condition Monitoring closure test per Leakage method of testing CCR Check Valve Condition Monitoring closure test per radiography method of testing ESO Full stroke exercise test to the open position (includes stroke time measurement except for check valves and manual valves)FSC Full stroke exercise test to the closed position (includes stroke time measurement except for check valves and manual valves)PSO Partial stroke exercise to the open position PSC Partial stroke exercise to the closed position FST Fail safe position test (includes stroke time measurement unless otherwise noted)PIT Position indication test RD Rupture Disc test RVT Test of safety/relief valves EX Explosive valve test Revision 0 Page 10 Revision 0 Page 10 Cooper Nuclear Station F/ifh Interval Inservice Testing Pro gram for Pumps and Valves 3.2 Valve Plan Table Description (continued) VBT Vacuum breaker testing SKID Component integral to or that supports operation of major component and is adequately tested as part of the major component. TEST FREO The frequency at which the applicable test will be performed. The definitions and abbreviations are identified below: Test Frequency Frequency of Testing Q At least once per 92 days CS Cold Shutdown RF Refueling Cycle 6M At least once every 6 months 2Y At least once every 2 years 5Y At least once every 5 years l0Y At least once per 10 years App I Relief Valve Frequency (OM Code Appendix I)CVCM IST CVCM Frequency (Appendix II). Refer to the 1ST CVCM Program Document for details.OPB l0CFR50 Appendix J Option B Leakage Rate Frequency PB Performance based frequency other than Option B TS In accordance with Technical Specifications SD Sample Disassembly and Examination NOTES: This column contains a brief component description and references to any applicable cold shutdown justifications(s), refueling outage justification(s), relief request(s), and any other component related information. All valves are considered "active" unless otherwise noted in this column. Cold shutdown justifications are designated CSJ-XX;Refueling Outage Justifications are designated ROJ-XX; Relief Requests are designated RV-XX for valves or RG-XX for general Program Requirements; Augmented Relief Requests are designated ARV-XX;station Technical Positions are designated TP-XX; and the Check Valve Condition Monitoring Plan Bases are designated in the 1ST CVCM Program Document.Revision 0Pae1 Page 11 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves 4.0 ATTACHMENTS: Attachment 1 Attachment 2 Attachment 3 Attachment 4 Attachment 5 Attachment 6 Attachment 7 Attachment 8 Attachment 9 Attachment 10 Attachment 11 System and P&ID Listing Pump Relief Requests Augmented Pump Relief Requests Valve Relief Requests Augmented Valve Relief Requests General Relief Requests Cold Shutdown Justifications Refuel Outage Justifications Technical Positions Inservice Testing Pump Table Inservice Testing Valve Table Revision 0Pae1 Page 12 Cooper Nuclear Station F~fih Interval Inservice Testing Pro gram for Pumps and Valves ATTACHMENT 1 SYSTEM AND P&ID LISTJING System CRD CS DW DGDO DGSA HPCI'IV IA MS NMT NBI PC RW RCIC REC RF RR RWCU RIHR SW SGT SLC SA System Name Control Rod Drive Core Spray Demineralized Water Diesel Generator Diesel Oil Diesel Generator Starting Air High Pressure Coolant Injection Heating and Ventilation Instrument Air Main Steam Neutron Monitoring Traversing Incore Probe Nuclear Boiler Instrumentation Primary Containment Radioactive Waste Reactor Core Isolation Cooling Reactor Equipment Cooling Reactor Feedwater Reactor Recirculation Reactor Water Cleanup Residual Heat Removal Service Water Standby Gas Treatment Standby Liquid Control Station Air P&LD 2039 2045 2029 2011, 2077 2077, 117.10-IC.09 2041, 2044 2019, 2020 2010, 2027, 2028 2028, 2041 2083 2026, 2027, 2028, 2041,, 2045 2022, 2027, 2028, 2084 2005, 2037, 2038 2041, 2043 2031 2043. 2044 2027 2042 2040, 2041 2006, 2036, 2077 2037 2045 2010 Revision 0 Page 13 Revision 0 Page 13 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves ATTACHMENT 2 PUMP RELEIEF REQUESTS________- PUMP RELIEF REQUEST INDEX _______Relif DecripionNRC Approval Date Request No.Decito RP-01 J[Core Spray Pumps Suction Gauge Range 2-12-16'RP-02 1[Residual Heat Removal Pumps Suction Gauge Range [[ 2-12-16'R-3 High Pressure Coolant Injection Pumps Suction Gauge 1 2-12-16~'~*~ Range __________ RP-04 J[Reactor Core IsOlation Cooling Pump Suction Gauge Range ]j 2-12-16'RP-05 I[ Pump Loop Accuracy Requirements 2-12-16' 1 RiP-06 ][Reactor Equipment Cooling Flow Gauge Range [[ 2-12-16'RP-07 j[Core Spray Pump B Vibration Alert Limits II 2-12-16'RP-08 J[Comprehensive Pump Test Upper Limit 2-12-16' 1 RP-09 1[Variance Around the Reference Values 2-12-16'(1) Approved by NRC letter, dated 2-12-16, from Meena K. Khanna, NRC, to Mr. Oscar A. Limpias, Vice President of Nuclear and CNO for CNS Revision 0 Page 14 Revision 0 Page 14 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-01 Core Spray Pump Suction Gauge Range Requirements Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected CS-P-A Core Spray Pump A CS-P-B Core Spray Pump B 2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004 Edition through 2006 Addenda 3. Applicable Code Reqiuirement ISTB-3510O(b)(1) -The full-scale range of each analog instrument shall not be greater than three times the reference value.4. Reason for Reqzuest Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(l), relief is requested from the requirement of ASME GM Code ISTB-35 10(b)(l). The proposed alternative would provide an acceptable level of quality and safety.The installed suction pressure gauge range of the core spray pumps is 30" Hg (inches Mercury) to 30.0 pounds per square inch (psig). The actual values for suction pressure during inservice testing are approximately 4.0 psig. As a result, the instrument range exceeds the requirement of ISTB-35 10(b)(1).5. Proposed Alternative and Basis for Use Pump suction pressure is used along with pump discharge pressure to determine pump differential pressure. Pump suction pressure actual values for the core spray pumps during inservice testing are approximately 4.0 psig. Based on ISTB-3510(b)(l), this would require, as a maximum, a gauge with a range of 0 to 12.0 psig (3 X 4.0 psig) to bound the actual value for suction pressure.Applying the accuracy requirement of + 2% of full scale (+- 6% of reference) for the quarterly Group B pump test, the resulting inaccuracies due to pressure effects would be +/-- 0.24 psig (0.02 X 12 psig).Pump discharge pressure actual values for the core spray pumps during inservice testing are approximately 300 psig. Based on ISTB-351 0(b)(1), this would require, as a maximum, a gauge with a range of 0 to 900 psig (3 X 300.0 psig) to bound the actual value for discharge pressure.Applying the accuracy requirement of + 2% of full scale (4- 6% of reference) for the quarterly Group B pump test, the resulting inaccuracies due to pressure effects would be +/-- 18 psig (0.02 X Revision 0Pae1 Page 15 Cooper Nuclear Station F~ifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-O1 Core Spray Pump Suction Gauge Range Requirements (Continued) 900 psig). Therefore, the maximum inaccuracies due to the suction and discharge pressure indications allowed by the code would be approximately +/- 18.24 psig.The Cooper Nuclear Station (CNS) installed suction pressure gauges (PI-36A/B), which were designed to have an accuracy of+ 0.5% of full scale, have a range of approximately 45 psig. The 45 psig gauge range is derived from the 30" Hg portion of the gauge range that is in a vacuum, which converts to approximately 15 psig, added to the 30 psig positive portion of the gauge. The+/- 0.3 psig current calibration tolerance is essentially a tolerance of approximately 0.66% of full scale (0.0066 X 45 psig = -~ +/- 0.3 psig). Currently, the installed discharge pressure indicators (PI-48A/B) are 0 to 500 psig indicators that are calibrated in a loop with corresponding pressure transmitters (PT-3 8A/B). These loops are being calibrated to +/- 10 psig, or + 2% of full scale (0.02 X 500 psig =+/- 10.0 psig).As an alternative, for the Group B quarterly test, CNS will use the installed suction pressure gauge (30" Hg to 30.0 psig), currently calibrated to within a tolerance of+/- 0.3 psig, together with the installed discharge pressure gauge (0 psig to 500 psig), currently calibrated in a loop to within a tolerance of+/- 10 psig. This results in a combined maximum inaccuracy of+ 10.3 psig due to the installed suction and discharge pressure indications, which is less than the code-allowed +/- 18.24 psig.Although the permanently installed suction pressure gauges (PI-36A/B) are above the maximum range limits of ASME OM Code ISTB-35 10(b)(1), they, in conjunction with the permanently installed discharge pressure gauges (PI-48A/B), yield a better accuracy for differential pressure than the minimum requirements dictated by the code and are, therefore, suitable for the test. The range and accuracy of the instruments used to determine differential pressure will be within +- 6%of the differential pressure reference value. Reference NUJREG 1482, Revision 2, Section 5.5.1.Although not anticipated, if any revisions to the current tolerance information provided occurs within the CNS fifth ten-year interval or actual suction and discharge pressure readings were to change significantly, this relief request will remain valid as long as the combination of range and accuracy will be less than the +/- 6% of the differential pressure reference value.Using the provisions of this relief request as an alternative to the specific requirements of ISTB-351 0(b)(1), identified above, will provide adequate indication of pump performance and continue to provide an acceptable level of quality and safety.Therefore, pursuant to 10 CFR 50.55a(z)(l), Nebraska Public Power District (NPPD) requests relief from the specific ISTB requirements identified in this request.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.Revision 0Pae1 Page 16 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves Relief Request RP-01 Core Spray Pump Suction Gauge Range Requirements (Continued)

7. Precedents This relief request was previously approved for the fourth ten-year interval at CNS as Relief Request RP-01 (TAC Nos. MC8837, MC8975, MC8976, MC8977, MC8978, MC8979, MC8980, MC8981, MC8989, MC8990, MC8991, and MC8992, June 14, 2006).Revision 0 Page 17 Revision 0 Page 17 Cooper Nuclear Station F~ith Interval Inservice Testing Program for Pumps and Valves Relief Request RP-02 Residual Heat Removal Pump Suction Gauge Range Requirements Proposed Alternative in Accordance with 10 CFR 50.55a@z)(1)

Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected RHR-P-A Residual Heat Removal (RHR) Pump A .RHR-P-B Residual Heat Removal Pump B RHR-P-C Residual Heat Removal Pump C RHR-P-D Residual Heat Removal Pump D 2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTB-35 10(b)(1) -The full-scale range of each analog instrument shall not be greater than three times the reference value.4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(l), relief is requested from the requirement of ASME OM Code ISTB-3510(b)(1). The proposed alternative would provide an acceptable level of quality and safety.The installed suction pressure gauge range of the residual heat removal pumps is 30" Hg to 150.0 psig. The actual values for suction pressure during inservice testing are approximately 5.0 psig. As a result, the instrument range exceeds the requirement of ISTB-3510(b)(1).

5. Proposed Alternative and Basis for Use Pump suction pressure is used along with pump discharge pressure to determine pump differential pressure.

Pump suction actual values for the residual heat removal pumps during inservice testing is approximately 5.0 psig. Based on ISTB-35 10(b)(1), this would require, as a maximum, a gauge with a range of 0 to 15.0 psig (3 X 5.0 psig) to bound the actual value for suction pressure. Applying the accuracy requirement of+/- 2% of full scale (+ 6% of reference) for the quarterly Group A pump test, the resulting inaccuracies due to pressure effects would be +/- 0.3 psig (0.02 X 15.0 psig).Pump discharge pressure actual values for the RHR pumps during inservice testing are approximately 170 to 195 psig. Conservatively basing it on the lowest of these discharge pressure readings, ISTB-35 10(b)(1) would require, as a maximum, a gauge with a range of 0 to 510 psig (3 X 170.0 psig) to bound the actual value for discharge pressure.Revision 0Pae1 Page 18 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-02 Residual Heat Removal Pump Suction Gauge Range Requirements (Continued) Applying the accuracy requirement of+/- 2% of full scale (+/- 6% of reference) for the quarterly Group A pump test, the resulting inaccuracies due to pressure effects would be +/- 10.2 psig (0.02 X 510 psig). Therefore, the maximum inaccuracies due to the suction and discharge pressure indications allowed by the code would be approximately +/- 10.5 psig.The CNS-installed suction pressure gauges (PI-10O6A/B/C/D), which were designed to have an accuracy of +/- 0.5% of full scale, have a range of approximately 165 psig. The 165 psig gauge range is derived from the 30" Hg portion of the gauge range that is in a vacuum, which converts to approximately 15 psig, added to the 150 psig positive portion of the gauge. The 4- 1.0 psig current calibration tolerance at the 5 psig suction pressure point is essentially a tolerance of approximately 0.6% of full scale (0.006 X 165 psig = -+ 1.0 psig). Currently, the installed discharge pressure indicators (PI-107AIB/C/D) are 0 to 400 psig indicators. The discharge indicators are being calibrated to +/- 5 psig, or+/- 1.25% of full scale (0.0125 X 400 psig = +/- 5.0 psig).As an alternative, for the Group A quarterly test, CNS will use the installed suction pressure gauge (30" Hg to 150.0 psig), currently calibrated to within a tolerance of 1 psig at the 5 psig point, together with the installed discharge pressure gauge (0 psig to 400 psig), currently calibrated to within a tolerance of+ 5 psig. This results in a combined maximum inaccuracy of+/- 6 psig due to the installed suction and discharge pressure indications, which is less than the code-allowed +/- 10.5 psig.Although the permanently installed suction pressure gauges (PI-1 06A/B/C/D) are above the maximum range limits of ASME OM Code ISTB -351 0(b)(1), they, in conjunction with the permanently installed discharge pressure gauges (PI-10O7A/B/C/D), yield a better accuracy for differential pressure than the minimum requirements dictated by the code and are, therefore, suitable for the test. The range and accuracy of the instruments used to determine differential pressure will be within + 6% of the differential pressure reference value. Reference NUJREG 1482, "Guidelines for Inservice Testing at Nuclear Power Plants," Revision 2, Section 5.5.1.Although not anticipated, if any revisions to the current tolerance information provided occurs within the CNS fifth ten-year interval or actual suction and discharge pressure readings were to change significantly, this relief request will remain valid as long as the combination of range and accuracy will be less than the +/- 6% of the differential pressure reference value.Using the provisions of this relief request as an alternative to the specific requirements of ISTB-351l0(b)(1), identified above, will provide adequate indication of pump performance and continue to provide an acceptable level of quality and safety.Therefore, pursuant to 10 CFR 50.55a(z)(l), NIPPD requests relief from the specific ISTB requirements identified in this request.Revision 0 Page 19 Revision 0 Page 19 Cooper Nuclear Station Fifthi Interval Inserviee Testing Program for Pumps and Valves Relief Request RP-02 Residual Heat Removal Pump Suction Gauge Range Requirements (Continued)

6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth 10-year interval.7. Precedents This relief request was previously approved for the fourth 10-year interval at CNS as Relief Request RP-02 (TAC Nos. MC8837, MC8975, MC8976, MC8977, MC8978, MC8979, MC8980, MC8981, MC8989, MC8990, MC8991, and MC8992, June 14, 2006).Revision 0 Page 20 Revision 0 Page 20 Cooper Nuclear Station Fifthi Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-03 High Pressure Coolant Injection Pump Suction Gauge Range Requirements Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected HTPCI-P-MP High Pressure Coolant Injection (HIPCD) Main Pump HPCI-P-BP High Pressure Coolant Injection Booster Pump 2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTB-3510 (b)(1) -The full-scale range of each analog instrument shall not be greater than three times the reference value.4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(l), relief is requested from the requirement of ASME OM Code ISTB-351 0(b)(1). The proposed alternative would provide an acceptable level of quality and safety.The installed suction pressure gauge range of the high pressure coolant injection pumps is 30" Hg to 150.0 psig. The actual value for suction pressure during inservice testing is approximately 15.0 psig. As a result, the instrument range exceeds the requirement of ISTB-35 10(b)(1).5. Proposed Alternative and Basis for Use Pump suction pressure is used along with pump discharge pressure to determine pump differential pressure. Pump suction actual values for the high pressure coolant injection pumps during inservice testing are approximately 15.0 psig. Based on ISTB-3510(b)(1) this would require, as a maximum, a gauge with a range of 0 to 45.0 psig (3 X 15.0 psig) to bound the actual value for suction pressure. Applying the accuracy requirement of+/- 2% of full scale (+/- 6% of reference) for the quarterly Group B pump test, the resulting inaccuracies due to pressure effects would be+ 0.9 psig (0.02 X 45.0 psig).The pump discharge pressure actual value for the HPCI pump during inservice testing is approximately 1200 psig. Based on ISTB-3510(b)(1), this would require, as a maximum, a gauge with a range of 0 to 3600 psig (3 X 1200.0 psig) to bound the actual value for discharge pressure.Applying the accuracy requirement of + 2% of full scale (+ 6% of reference) for the quarterly Group B pump test, the resulting inaccuracies due to pressure effects would be + 72 psig (0.02 X 3600 psig). Therefore, the maximum inaccuracies due to the suction and discharge pressure indications allowed by the code would be approximately + 72.9 psig.Revision 0Pae2 Page 21 Cooper Nuclear Station FifthI Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-03 High Pressure Coolant Injection Pump Suction Gauge Range Requirements (Continued) The CNS-installed suction pressure gauge (PI-99), which was designed to have an accuracy of-4 0.5% of full scale, has a range of approximately 165 psig. The 165 psig gauge range is derived from the 30" Hg portion of the gauge range that is in a vacuum, which converts to approximately 15 psig, added to the 150 psig positive portion of the gauge. The +i 1.0 psig current calibration tolerance is essentially a tolerance of approximately 0.6% of full scale (0.006 X 165 psig =+ 1.0 psig). Currently, the installed discharge pressure indicator (P1-81) is a 0 to 1500 psig indicator. The discharge indicator is currently being calibrated to + 7.5 psig, or + 0.5% of full scale (0.005 X 1500 psig = + 7.5 psig).As an alternative, for the Group B quarterly test, CNS will use the installed suction pressure gauge (30" Hg to 150.0 psig), currently calibrated to within a tolerance of+/-: 1 psig, together with the installed discharge pressure gauge (0 psig to 1500 psig), currently calibrated to within a tolerance of+/- 7.5 psig. This results in a combined maximum inaccuracy of+/- 8.5 psig due to the installed suction and discharge pressure indications, which is less than the code-allowed + 72.9 psig.Although the permanently installed suction pressure gauge (P1-99) is above the maximum range limits of ASME OM Code ISTB-35 10(b)(1), it, in conjunction with the permanently installed discharge pressure gauge (P1-81), yields a better accuracy for differential pressure than the minimum requirements dictated by the code and is, therefore, suitable for the test. The range and accuracy of the instruments used to determine differential pressure will be within +/-- 6% of the differential pressure reference value. Reference NUREG 1482, Revision 2, Section 5.5.1.Although not anticipated, if any revisions to the current tolerance information provided occurs within the CNS fifth ten-year interval or actual suction and discharge pressure readings were to change significantly, this relief request will remain valid as long as the combination of range and accuracy will be less than the + 6% of the differential pressure reference value.Using the provisions of this relief request as an altemnative to the specific requirements of ISTB-35 10(b)(1), identified above, will provide adequate indication of pump performance and continue to provide an acceptable level of quality and safety.Therefore, pursuant to 10 CFR 50.55a(z)(1), NPPD requests relief from the specific ISTB requirements identified in this request.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth 10-year interval.7. Precedents This relief request was previously approved for the fourth 10-year interval at CNS as Relief Request RP-03 (TAC Nos. MC8837, MC8975, MC8976, MC8977, MC8978, MC8979, MC8980, MC8981, MC8989, MC8990, MC8991, and MC8992, June 14, 2006).Revision 0Pae2 Page 22 Cooper Nuclear Station Ffith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-04 Reactor Core Isolation Cooling Pump Suction Gauge Range Requirements Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected RCIC-P-MP Reactor Core Isolation Cooling (RCIC) Main Pump 2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTB-3 510O(b)(1) -The full-scale range of each analog instrument shall not be greater than three times the reference value.4. Reason for Reqjuest Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(1), relief is requested from the requirement of ASME OM Code ISTB-35 10(b)(1). The proposed alternative would provide an acceptable level of quality and safety.The installed suction pressure gauge range of the reactor core isolation cooling pump is 30" Hg to 150.0 psig. The actual value for suction pressure during inservice testing is approximately 15.0 psig. As a result, the instrument range exceeds the requirement of ISTB-35 10(b)(1).5. Proposed Alternative and Basis for Use Pump suction pressure is used along with pump discharge pressure to determine pump differential pressure. Pump suction actual values for the reactor core isolation cooling pump during inservice testing is approximately 15.0 psig. Based on ISTB-351l0(b)(1) this would require, as a maximum, a gauge with a range of 0 to 45.0 psig (3 X 15.0 psig) to bound the lowest actual value for suction pressure. Applying the accuracy requirement of+ 2% of full scale (+ 6% of reference) for the quarterly Group B pump test, the resulting inaccuracies due to pressure effects would be +/- 0.9 psig (0.02 X 45.0 psig).The discharge pressure actual value for the RCIC pump during inservice testing is approximately 1250 psig. Based on ISTB-3510(b)(1), this would require, as a maximum, a gauge with a range of 0 to 3750 psig (3 X 1250.0 psig) to bound the actual value for discharge pressure. Applying the accuracy requirement of + 2% of full scale (+ 6% of reference) for the quarterly Group B pump test, the resulting inaccuracies due to pressure effects would be +/- 75 psig (0.02 X 3750 psig). Therefore, the maximum inaccuracies due to the suction and discharge pressure indications allowed by the code would be approximately +/- 75.9 psig.Revision 0 Page 23 Revision 0 Page 23 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-04 Reactor Core Isolation Cooling Pump Suction Gauge Range Requirements (Continued) The CNS-installed suction pressure gauge (PI-66), which was designed to have an accuracy of+/- 0.5% of full scale, has a range of approximately 165 psig. The .165 psig gauge range is derived from the 30" Hg portion of the gauge range that is in a vacuum, which converts to approximately 15 psig, added to the 150 psig positive portion of the gauge. The 4- 1.0 psig current calibration tolerance is essentially a tolerance of approximately 0.6% of full scale (0.006 X 165 psig =+ 1.0 psig). Currently, the installed discharge pressure indicator (PI-5 9) is a 0 to 1500 psig indicator. The discharge indicator is being calibrated to +/- 15 psig, or + 1.0% of full scale (0.01 X 1500 psig= + 15.0 psig).As an alternative, for the Group B quarterly test, CNS will use the installed suction pressure gauge (30" Hg to 150.0 psig), currently calibrated to within a tolerance of++/- 1 psig, together with the installed discharge pressure gauge (0 psig to 1500 psig), currently calibrated to within a tolerance of+ 15.0 psig. This results in a combined maximum inaccuracy of+ 16.0 psig due to the installed suction and discharge pressure indications, which is less than the code-allowed +/- 75.9 psig.Although the permanently installed suction pressure gauge (P1-66) is above the maximum range limits of ASME OM Code ISTB-3510(b)(l), it, in conjunction with the permanently installed discharge pressure gauge (PI-59), yields a better accuracy for differential pressure than the minimum requirements dictated by the code and is, therefore, suitable for the test. The range and accuracy of the instruments used to determine differential pressure will be within + 6% of the differential pressure reference value. Reference NUREG 1482, Revision 2, Section 5.5.1.Although not anticipated, if any revisions to the current tolerance information provided occurs within the CNS fifth ten-year interval or actual suction and discharge pressure readings were to change significantly, this relief request will remain valid as long as the combination of range and accuracy will be less than the +/- 6% of the differential pressure reference value.Using the provisions of this relief request as an alternative to the specific requirements of ISTB-351 0(b)(1), identified above, will provide adequate indication of pump performance and continue to provide an acceptable level of quality and safety.Therefore, pursuant to 10 CFR 50.55a(z)(1), NPPD requests relief from the specific ISTB requirements identified in this request.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was previously approved for the fourth ten-year interval at CNS as Relief Request RP-04 (TAC Nos. MC8837, MC8975, MC8976, MC8977, MC8978, MC8979, MC8980, MC8981, MC8989, MC8990, MC8991, and MC8992, June 14, 2006).Revision 0Pae2 Page 24 Cooper Nuclear Station F~ilh Interval Inservice Testing Pro gram for Pumps and Valves Loop Accuracy Requirements Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternate Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected CS-P-A CS-P-B HIPCI-P-MP HiPCI-P-BP SW-P-BPA SW-P-BPB SW-P-BPC*SW-P-BPD Core Spray (CS) Pump A Core Spray Pump B High Pressure Coolant Injection Main Pump High Pressure Coolant Injection Booster Pump Service Water Booster (SWB) Pump A Service Water Booster Pump B Service Water Booster Pump C Service Water Booster Pump D 2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement Table ISTB-35 10-1, "Required Instrument Accuracy" 4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(1), relief is requested from the requirement of ASME OM Code ISTB Table ISTB-3510-l for Group A and B Pump Pressure accuracy (A- 2%) and for flow rate accuracy (A- 2%). The proposed alternative would provide an acceptable level of quality and safety.The installed instrumentation for the subject pumps yield the following loop accuracies: Pump Parameter Equip. Loop Accuracy (%)Calibration Loop Accuracy (%)CS Pump Discharge Pressure CS Pump Flowrate FTPCI Pump Flowrate SWB Pump Flowrate 2.06 2.02 2.03 2.03<2 200%< 2.00%< 2.00%< 2.00%As a result, the equipment loop accuracies do not meet the A- 2% requirements of Table ISTB-35 10-1, "Required Instrument Accuracy." 5. Proposed Alternative and Basis for Use The difference between the code required and presently installed instrument loop accuracies is 0.06%, at a maximum, as presented above. This difference is insignificant when applied to the quantitative measured values for these parameters during the respective Group A or Group B Revision 0 Page 25 Cooper Nuclear Station Ffith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-05 Loop Accuracy Requirements (Continued) quarterly tests. Additionally, all calibration tolerances of the loops involved meet or exceed the code-allowed accuracies of+ 2% or better.CS pump discharge pressure loop is made up of a pressure indicator (range of 0 to 500 psig) and a pressure transmitter. The pressure indicator (PI-48A/B) has a nameplate accuracy of+ 2%, and the pressure transmitter (PT-38A/B) has a nameplate accuracy oftq 0.5%. Therefore, based on the nameplate accuracies alone, the equipment loop accuracy for discharge pressure indication is-4 2.06% (square root of the sum of the squares), which exceeds the code requirement of+/- 2%.The variation from the code of 0.06%, with a gauge range of 0 to 500 psig, would amount to a potential deviation of only 0.3 psig (0.0006 X 500). However, CNS is currently calibrating this discharge pressure loop to within +- 10 psig, which is equivalent to a +- 2% of full scale tolerance (0.02 X 500 psig =+ 10 psig), which meets the accuracy requirements of the code.CS pump flow rate loop is made up of a flow indicator (range of 0 to 6000 gallons per minute[gpm]), and a flow transmitter. The flow indicator (FI-50A/B) has a nameplate accuracy of+/- 2%, and the flow transmitter (FT-40A/B) has a nameplate accuracy of+ 0.25%. Therefore, based on the nameplate accuracies alone, the equipment loop accuracy for discharge pressure indication is+ 2.02% (square root of the sum of the squares), which exceeds the code requirement of+/- 2%.The variation from the code of 0.02%, with a gauge range of 0 to 6000 gpm, would amount to potential deviation of only 1.2 gpm (6000 X .0002). However, CNS is currently calibrating this flow loop to within +/- 50 gpm (at the Inservice Testing (1ST) reference value of 5000 gpm) or approximately +/- 0.83% of full scale (+/-- 0.0083 X 6000 = +/- 50 gpm), which is better than the+/-- 2% of full scale accuracy requirements of the code. If a preservice test were to be run, CNS would ensure that the loop was calibrated to < 2% over the full range of the test prior to performing it.HiPCI pump flow rate loop is made up of a flow indicating controller (range of 0 to 5000 gpm), a flow transmitter, and a flow square rooter. The flow indicating controller (FIC-1 08) has a nameplate accuracy of+ 0.25%, the flow transmitter (FT-82) has a nameplate accuracy of+- 0.25%, and the flow square rooter (SQRT-l118) has a nameplate accuracy of +/- 2% from approximately 0 to 1000 gpm and +/- 0.5% from approximately 1000 to 5000 gpm. Therefore, based on the nameplate accuracies alone, the equipment loop accuracy for flow indication is approximately

  • 2.03% (square root of the sum of the squares) from 0 to 1000 gpm, which does not meet the code requirement of +/- 2%, and approximately

+- 0.6 1% from 1000 to 5000 gpm, which does meet the code requirement of + 2%. The variation from the code of 0.03% in the range of 0 to 1000 gpm, with a gauge range of 0 to 5000 gpm, would amount to a potential deviation of only 1.5 gpm (5,000 X .0003). However, CNS is currently calibrating this flow loop to within + 100 gpm (at the 1ST reference of 4000 gpm and at other points from 1000 gpm to 5000 gpm) or + 2% of full scale (+ 0.02 X 5000 = ~-+ 100 gpm), which is equivalent to the + 2%of full scale accuracy requirements of the code. If a preservice test were to be run, CNS would ensure that the loop was calibrated to < 2% over the full range of the test prior to performing it.The SWB flow rate is made up of a flow indicator (range of 0 to 10,000 gpm), a flow transmitter, and a flow square rooter. The flow indicator (FI-l132A/B) has a nameplate accuracy of+/- 2%, the flow transmitter (FT-97) has a nameplate accuracy of+/- 0.25%, and the flow square rooter (SQRT-132AiB) has a nameplate accuracy of+/- 0.25%. Therefore, based on the nameplate accuracies alone, the equipment loop accuracy for flow indication is approximately + 2.03%Revision 0 Page 26 Cooper Nuclear Station Ffith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-05 Loop Accuracy Requirements (Continued)(square root of the sum of the squares), which exceeds the code requirement of + 2%. The variation from the code of 0.03%, with a gauge range of 0 to 10,000 gpm, would amount to a potential deviation of only 3 gpm (0.0003 X 10,000). However, CNS is currently calibrating this flow loop to within +/-- 100 gpm, which is equivalent to a +/-: 1% of full scale tolerance (0.01 X 10,000 gpm = +/-- 100 gpm), which is better than the +/-- 2% of full scale accuracy requirements of the code.As an alternative for the Group A or Group B quarterly test, CNS will use the installed instruments calibrated such that the loop accuracies are as indicated in the above table. No adjustments to acceptance criteria will be made as the calibrated loop accuracies will meet or exceed the code tolerances. Although the permanently installed instrument loops do not meet the accuracy requirements of ASME OM Code ISTB Table ISTB-3510-1 when looking at nameplate accuracies, the effects of these small inaccuracies are insignificant when compared to the measured values, and credit will be taken for the ability to calibrate the loop within the code-allowed tolerance. Although not anticipated, if any revisions to the current tolerance information provided occurs within the CNS fifth ten-year interval, this relief request will remain valid as long as the calibrated loop accuracies meet the code required tolerances of < 2.00% of full scale.Using the provisions of this relief request as an alternative to the specific requirements of ISTB Table 3510-1, identified above, will provide adequate indication of pump performance and continue to provide an acceptable level of quality and safety.Therefore, pursuant to 10 CFR 50.55a(z)(1), NPPD requests relief from the specific ISTB requirements identified in this request.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was previously approved for the fourth ten-year interval at CNS as Relief Request RP-05 (TAC Nos. MC8837, MC8975, MC8976, MC8977, MC8978, MC8979, MC8980, MC8981, MC8989, MC8990, MC8991, and MC8992, June 14, 2006).Revision 0 Page 27 Revision 0 Page 27 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RP-06 Reactor Equipment Cooling Pump Flow Rate Range Requirements Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected REC-P-A Reactor Equipment Cooling (REC) Pump A REC-P-B Reactor Equipment Cooling Pump B REC-P-C Reactor Equipment Cooling Pump C REC-P-D Reactor Equipment Cooling Pump D 2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTB-35 10(b)(1) -The full-scale range of each analog instrument shall not be greater than three times the reference value.4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(1), relief is requested from the requirement of ASME OM Code ISTB-35 10(b)(l). The proposed alternative would provide an acceptable level of quality and safety.The installed flow rate instrument range of the reactor equipment cooling pumps is 0 to 4000 gpm. The reference values for flow rate during inservice testing are 1100 gpm. As a result, the instrument range exceeds the requirement of ISTB-35 10(b)(1).5. Proposed Alternative and Basis for Use The permanent plant flow Instruments REC-FI-450A and REC-FI-450B are calibrated such that their accuracy is 1.25% of full scale. This yields a total inaccuracy of 50 gpm (0.0125 X 4000 gpm). Reference flow rates for the reactor equipment cooling pumps are 1100 gpm. Based on ISTB-3510(b)(1) this would require, as a maximum, a gauge with a range of 0 to 3300 gpm (3 X 1100 gpm) to bound the lowest reference value for flow.Applying the accuracy requirement of+~ 2% for the pump test, the resulting inaccuracies due to flow would be + 66 gpm (0.02 X 3300 gpm).Revision 0Pae2 Page 28 Cooper Nuclear Station Ffith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-06 Reactor Equipment Cooling Pump Flow Rate Range Requirements (Continued) As an alternative, for the reactor equipment cooling pump inservice tests, CNS will use the installed flow rate instrumentation (0 to 4000 gpm) calibrated to less than + 2% such that the inaccuracies due to flow will be less than or equal to that required by the code (+ 66 gpm). This will ensure that the installed flow rate instrumentation is equivalent to the code, or better, in terms of measuring flow rate.Although the permanently installed flow gauges are above the maximum range limits of ASME OM Code ISTB-35 1O(b)(1), they are within the accuracy requirements and are, therefore, suitable for the test. Reference NUJREG 1482, Revision 2, Section 5.5.1.Using the provisions of this relief request as an alternative to the specific requirements of ISTB-351 0(b)(1), identified above, will provide adequate indication of pump performance and continue to provide an acceptable level of quality and safety.Therefore, pursuant to 10 CFR 50.55a(z)(1), NPPD requests relief from the specific ISTB requirements identified in this request.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was previously approved for the fourth ten-year interval at CNS as Relief Request RP-06 (TAG Nos. MC8837, MC8975, MC8976, MC8977, MC8978, MC8979, MC8980, MC8981, MC8989, MC8990, MC8991, and MC8992, June 14, 2006).Revision 0 Page 29 Revision 0 Page 29 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2) Alternative Demonstrates a Hardship Without a Compensating Increase in Quality and Safety 1. ASME Code Component(s) Affected CS-P-B Core Spray Pump B 2. Applicable Code Edition and Addenda ASMEi OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTB Table ISTB-5 121-1, "Centrifugal Pump Test Acceptance Criteria" 4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(2), relief is requested from the requirement of ASME OM Code ISTB Table ISTB-5121-1 during the biennial comprehensive pump test or any other time vibrations are taken to determine pump acceptability (i.e., post-maintenance testing, other periodic testing, etc.). The proposed alternative demonstrates a hardship without a compensating increase in quality and safety.The 1ST Program has consistently required (prior to obtaining relief per RP-06 of the third interval program) that CS Pump B (CS-P-B) be tested on an increased frequency due to vibration values at Points 1H and 5H, as shown in Figure 1, periodically being in the alert range. Relief is requested from ISTB Table ISTB-5 121-1 requirements to test the pump on an increased periodicity due to vibration levels for Points 1H and/or 5H exceeding the ISTB alert range absolute limit for the comprehensive pump test. This request is based on analysis of vibration and pump differential pressure data indicating that no pump degradation is taking place. CNS is proposing to use alternative vibration alert range limits for vibration Points 1H and 5H. This provides an alternative method that continues to meet the intended function of monitoring the pump for degradation over time while keeping the required action level unchanged.

5. Proposed Alternative and Basis for Use Pump Testing Methodology CS-P-B at CNS is tested using a full flow recirculation test line back to the suppression pool each quarter. CS-P-B has a minimum flow line which is used only to protect the pump from overheating when pumping against a closed discharge valve. The minimum flow line isolation valve for CS-P-B is initially open when the pump is started, and flow is initially recirculated through the minimum flow line back to the suppression pool.Revision 0Pae3 Page 30 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued)

Then, the full-flow test line isolation valve is throttled open to establish flow through the full-flow recirculation test line. The minimum flow line is then isolated automatically, and all flow remains through the full-flow test line for the IST test.The B train of the CS system is operated in the same manner and under the same conditions for each test of CS-P-B, regardless of whether CNS is operating or shut down. Consequently, the pump will experience the same potential for flow-induced, low frequency vibration whenever it is tested, whether CNS is operating or shut down. As a result, this relief is requested for the comprehensive pump testing of CS-P-B when vibration measurements are required or any other time vibrations are recorded to determine pump acceptability (i.e., post-maintenance testing, other periodic testing, etc.).CNS considers full-flow testing to be preferable to minimum flow testing due to the ability to evaluate overall pump performance at post-accident flow design conditions. Minimum flow testing would provide only limited information about the pump.Nuclear Regulatory Commission (NRC) Staff Document NUREG/CP-O0152 NRC Staff document NUREG/CP-O0152, entitled "Proceedings of the Fourth NRC/ASME Symposium on Valve and Pump Testing," dated July 15-18, 1996, included a paper entitled Nuclear Power Plant Safety Related Pump Issues, by Joseph Colaccino of the NRC staff. That paper presented four key components that should be addressed in a relief request of this type to streamline the review process. These four key components are as follows: I. The licensee should have sufficient vibration history from inservice testing which verifies that the pump has operated at this vibration level for a significant amount of time, with any "spikes" in the data justified. II. The licensee should have consulted with the pump manufacturer or vibration expert about the level of vibration the pump is experiencing t~o determine if pump operation is acceptable. III. The licensee should describe attempts to lower the vibration below the defined code absolute levels through modifications to the pump.IV. The licensee should perform a spectral analysis of the pump-driver system to identify all contributors to the vibration levels.The following is a discussion of how these four key components are addressed for this relief request.Revision 0Pae3 Page 31 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) I. Vibration History (Key Component No. 1)A. Testing Methods and Code Reqiuirements Inconsistent higher vibrations on CS-P-B have been a condition that has existed since original installation of this pump in 1973. During the construction and preoperational testing, vibrations were measured in "mils" at the top and side of the motor outboard (farthest from the pump), the side of the motor inboard (nearest the pump), and pump inboard (nearest the motor). The vibration signals were tape recorded along with the dynamic pressure pulsations in the suction and discharge of the pump as thc flow was varied. The intention was to see if hydraulic disturbances were responsible for the observed phenomena. Observation of the vibration signals on the oscilloscope showed conclusively that the motor was vibrating with randomly distributed bursts of energy at the natural frequency of the total system. Therefore, it was determined that the hydraulic disturbances found in the piping was the source of the energy. Pipe restraints were added that reduced the piping system vibrations. The monitoring of multiple vibration points over the years had not been a requirement of Section XI of the ASME Code until the adoption of the OM Standards/Codes. Therefore, at CNS, the first and second ten-year interval IST code requirements did not include the monitoring of multiple vibration points. The CNS second interval 1ST Program was committed to the 1980 Edition, Winter 1981 Addenda of Section XI. Paragraph IWP-4510 of this code required that "at least one displacement vibration amplitude shall be read during each inservice test." This code was in effect at CNS until the start of the third ten-year interval, which began on March 1, 1996. The CNS third interval IST Program was committed to the 1989 Edition of Section XI, which required multiple vibration points to be recorded during IST pump testing in accordance with the ANSI/ASME Operations and Maintenance Standard, Part 6, 1987 Edition with the 1988 Addenda.However, CNS proactively began monitoring vibration on pumps in the IST Program in velocity units (inches per second) at multiple vibration points in 1990 in accordance with an approved relief request. Therefore, data exists for vibration Points 11H and 5H from April 1,990 to the present. This data is included in the figures provided in this relief request. In April 1990, an analog velocity meter was utilized to begin measuring five different points in units of velocity. These are the same points measured today. Further technological advances resulted in the utilization of more reliable vibration meters beginning in late 1996. For the fourth interval, which began on March 1, 2006, the 2001 Edition through 2003 Addenda of the ASME OM Code was the code of record.Vibration measurements were required to be taken only during the comprehensive test since the CS-P-B pump is considered a Group B pump. The same will be true for the fifth interval, beginning on March 1, 2016, in which the 2004 Edition through the 2006 Addenda of the ASME OM Code will be the code of record.Revision 0 Page 32 Revision 0 Page 32 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) B. Review of Vibration History Data Beginning in April 1990, five vibration points (lV, 1H, 2H, 3H, 5H) were recorded for CS-P-B. However, the pump was tested at 4720 gpm from April 1990 to April 1992, then at 4800 gpm from April 1992 through December 1994, and finally at 5000 gpm from January 1995 to the present. The January 1995 test was also a post-maintenance test following the work that replaced the restricting orifice in the test return line. The last re-baseline occurred on November 6, 1996, due to the implementation of a new vibration meter with new instrument settings. Therefore, it would be appropriate to review the data from this date forward to track for degradation. This would be over eighteen years of data at the same reference points.CS-P-B IST vibration trend graphs for vibration points 5H, 1V, 2H, and 3H (Figures 3a, 4a, 5a, and 6a in this relief request), which include data from November 6, 1996, to the present, show flat or slightly downward trends. Vibration point 1H shows an essentially flat trend from -~2002 to the present (Figure 2a) and when including the data since 1990 (Figure 2b). These observations indicate that CS-P-B vibrations are not increasing in magnitude. These trends also show that Points 1H and 5H occasionally exceed the alert range criteria (Figures 2a and 3a). Figure 12 illustrates the trend for CS-P-B differential pressure (D/P) readings from January 1995 (re-baselined pump at 5000 gpm) to the present. This represents approximately twenty years of data for pump D/P with the testing at 5000 gpm. As can be seen from Figure 12, no degradation in pump D/P has occurred.Trend Graphs 2b, 3b, 4b, 5b, and 6b illustrate vibration data dating back to April 1990 for all vibration points. The data prior to 1996 represents data taken with analog, less reliable vibration instruments and, as discussed previously, at differing flows. However, it does clearly indicate that the piping-induced vibrations for vibration Points 1 H and 5H were present in the early 1990s. This condition was also documented in the 1980s. In July 1985, CNS work item #85-2497 documented high vibration readings on the horizontal motor position. A pipe resonance problem was suspected at that time.Vibrational readings varied between 0.3 and 0.5 in/sec with spikes to 0.7 ir/sec every few seconds. This 1985 documentation, available vibration data since 1990, along with the testing performed during the preoperational time period, substantiates that the piping-induced vibrations have been in existence since the pump was installed. These graphs indicate that the vibration point trends since April 1990 are essentially flat or slightly downward. Therefore, based on the available data at CNS, this pump has experienced essentially no degradation in vibration levels for ~-24.5 years or in DiP for -20 years.Revision 0 Page 33 Revision 0 Page 33 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) C. Review of "Spikes" in Vibration Data In reviewing the trend data for vibration points 1 H (Figures 2a and 2b) and 5H (Figures 3a and 3b), which includes the code-required frequency ranges (one-third pump running speed to 1000 Hertz [Hz].), random spikes were observed throughout the data that resulted in values above the alert range. These spikes are best described in a 2001 report by Machinery Solutions, Inc., an industry expert on vibrations, as follows: Most of the vibration that is measured on the motor casing is due to excitation of the structural resonances of the motor/pump by turbulent flow. These structural resonances are poorly damped and can be easily excited. Most vertical pumps have similar types of behavior, and it is not necessarily problematic by itself. A problem occurs when a pump has a continuous forcing function whose frequency coincides with a resonance (i.e., running speed). The forcing function in this case is flow turbulence caused in large part by the S-curve in the piping just off the pump discharge. The flow through this area generates lateral broadband forces, due to elbow effects, that excite the resonances in a non-continuous fashion.This is why the amplitude swings so dramatically on the motor case (the location of vibration points 1 H and 5H). The system goes from brief periods of excitation to brief periods of no excitation. The discharge riser is also moving side to side from the same forces. Although the discharge piping configuration is both non-standard and less than optimum for this application, it poses no threat to the long-term reliability of either the pump or the motor. The only negative impact is on vibration levels relative to a generic standard.As illustrated previously, there have been no degrading trends associated with vibration data points 1H and 5H for -24.5 years (Figures 2b and 3b). Since June 2002, filtered data (removal of one-third pump running speed to one-half pump running speed frequencies) has been recorded in addition to the current code-required values for vibration points 111 and 5H (reference Figures 2c and 3c for data since 2010). In reviewing this data, the trends are lower in value, steady, and without the spikes that the code-required data contains. This further supports the fact that the spikes in the original code data are due to the piping-induced, non-detrimental vibration occurring at the one-third to one-half pump running speed.Revision 0Pae3 Page 34 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) II. Consultation -Pump Manufacturer/Vibration Expert (Key Component No. 2)A. Pump Manufacturer Evaluation of CS-P-B Vibrations Byron Jackson is the pump manufacturer for CS-P-B. The pump is an 8 x 14 x 30 DVSS, vertical mount, single stage centrifugal pump. The pump impeller is mounted on the pump motor's extended shaft. As outlined in the Core Spray System Summary of Preoperational Test, the data obtained for the B Core Spray Pump indicated high vibration. The high yibration had been recognized early in the construction testing phase, and Byron Jackson sent a representative to the site to investigate. In a letter dated February 16, 1973, the Byron Jackson representative indicated the following:

1. Tests indicated that the natural frequency of the pump was 940 revolutions per minute (rpm) (approximately one-half pump speed) in the direction of the piping and 720 rpm (between one-third and one-half of pump speed) in the direction perpendicular to the piping.2. Observation of the test signals on the oscilloscope showed very conclusively that the motor was vibrating with randomly distributed bursts of energy, the frequency of which matched the natural frequency of the total system. This can only mean that the energy is coming from the hydraulic disturbances found in the piping.3. Whenever large flows are carried in piping, there is usually considerable turbulence associated with the elbows, tees, etc., of the piping configuration, all of which results in piping reactions and motion. Apparently, the vibrating piping was, in turn, vibrating the pump.4. When jacks were installed between the top of the pump and the bottom of the motor flange in an effort to stiffen the motor pump system, the motor vibrations went up due to more energy being transmitted from the pipe-pump system into the motor.5. Testing was performed to determine any weaknesses in the pump-motor mechanical system. The vibration amplitude using the IRD instrument, with the filter set at operating speed, sampled many points vertically along the pump-motor structure.

Plots of the data (along with phase angle determined by means of the strobe light) showed very clearly that the total structure was vibrating as a rigid assembly from the floor mounting. Examination of the high amplitude vibration signals showed them to be at the extremely low system natural frequencies as determined earlier.6. Such low acceleration levels, along with the system acting as a rigid structure (between motor and pump), means that the motor and pump can operate Revision 0Pae3 Page 35 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) with these levels of vibration with absolutely no impairment of operating life.This is the picture that seems very clearly described by the data obtained during these tests. There is absolutely no reason to restrict the operation of these pumps in any way.Although the vibration was found to be acceptable, CNS took actions to install new pipe supports as an attempt to reduce these piping-induced vibrations. This action was successful as will be discussed in a later section of this relief request.B. CNS Expert Analysis of CS-P-B Vibrations As the Vibration Monitoring Program expanded in the early 1 990s, it became evident that the low frequency, piping-induced vibrations still remained in CS-P-B. Design Change (DC) 94-046 resulted in the replacement of the orifices in the test return line. A March 16, 1995, memo to the CNS 1ST Engineer from the CNS Lead Civil/Structural Engineer discussed the CS-P-B vibration measurements obtained during DC 94-046 acceptance testing.The vibration data was collected using peak velocity measuring instrumentation as required for the performance of the IST test and with instrumentation that provides displacement and velocity versus frequency data. It was observed that the significant vibrations in the 1H direction were occurring around 700 cycles per minute (cpm), while the pump speed is at 1780 cpm (i.e., rpm). Given the piping movement of the system, and the knowledge that piping vibrations can commonly occur in the 700 cpm (12 Hz)range, CNS concluded that the pump vibrations were piping dependent. The CNS Lead Civil/Structural Engineer concluded that the significant pump vibrations are occurring at less than one-half of the pump operating speed. The pumps are rigidly mounted at their bases, and any impeller-induced vibrations would occur at the pump running speed or at the vane passing frequency. Therefore, the sub-synchronous pump vibrations are clearly piping induced, non-detrimental to pump/motor service or reliability, and should not be used as a basis for pump degradation. This is because the purpose of pump in-service testing is to diagnose and trend internal pump degradation. The memo further states that the vibration data collection requirement specified in the 1ST procedure consists of peak velocity recordings, which may be masked by piping-induced vibrations, negating internal pump degradation diagnosis and trending. Based on the historical trending data for both CS pumps, the vibration has remained at a consistent amplitude, trending neither upward nor downward, indicating that the induced vibrations are not impairing pump operability, nor capable of preventing the pump from fulfilling its safety function. The piping vibration is present when flow is present through the test return line. It was visually observed during DC 94-046 acceptance testing that piping vibrations were minimal when flow was directed through the minimum flow line.Revision 0Pae3 Page 36 Cooper Nuclear Station Fifth Interval lnservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Following the DC 94-046 testing, CNS noted that the deflections observed in the discharge piping were significantly reduced. Based on these results, it was determined by the Nuclear Engineering Department, Civil/Structural Group, that the CS Loop B piping vibration stresses are less than the endurance limit of the piping.On October 17, 2002, a Plant Engineering Supervisor at CNS, knowledgeable in the area of pump vibration analysis, issued a memo to the CNS Risk & Regulatory Affairs Manager discussing the low frequency vibration issue with the CS-P-B.In the memo, it is stated that the pipe is vibrating as a reaction to flow turbulence, which in turn is causing the pump to vibrate. The memo documents the basis for why the low frequency vibration (less than one-half pump running speed) experienced during CS-P-B operation is not indicative of degrading pump performance and is not expected to adversely impact pump operability. To summarize, in the area of pump performance, aside from the randomness of the low frequency peaks, the spectral data shows no degrading trend in performance over several years of data. The low frequency piping-induced vibrations are not expected to adversely impact pump operability. C. Independent Industry Vibration Expert Evaluation of CS-P-B In 2001, Machinery Solutions, Inc. was retained to perform an independent study of the CS-P-B vibrations. The following discussion was obtained from their report, issued in September of 2001. Machinery Solutions, Inc. utilized seven transducers and acquired data from CS-P-B continuously while it was operating, and data was stored every 3 seconds. Orbit plots, spectrum plots, bode and polar plots, cascade/waterfall plots, overall amplitude plots, trend plots, XY graph plots, and tabular lists were utilized to analyze the data. The data obtained by Machinery Solutions, Inc., indicated that the vibration amplitudes during the run were much higher at the top of the motor than they were at the bottom of the motor. The amplitudes decreased even further on the pump.The spectrum plots showed that most of the vibration was occurring below running speed. They also showed that the low frequency vibration is a different frequency in each direction. The predominant peaks occur at approximately 870 cpm (less than one-half pump running speed) in line with discharge and at approximately 630 cpm (less than one-half pump running speed) perpendicular to discharge. The amplitude of each of these peaks varied significantly from second to second. The natural frequency of the pump-motor-piping structure was determined via impact testing prior to starting the pump. The natural frequencies were determined to be approximately 830 cpm in line with discharge and 670 cpm perpendicular to discharge. Such a vibration response is typical for vertical pumps.Revision 0 Page 37 Revision 0 Page 37 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Machinery Solutions, Inc. concluded the following:

1. Most of the vibration that is measured on the motor casing is due to excitation of the structural resonances of the motor/pump by turbulent flow. These structural resonances are poorly damped and can be easily excited. Most vertical pumps have similar types of behavior, and it is not necessarily problematic by itself. A problem occurs when a pump has a continuous forcing function whose frequency coincides with a resonance (i.e., running speed). The forcing function in this case is flow turbulence caused in large part by the S-curve in the piping just off the pump discharge.

The flow through this area generates lateral broadband forces, due to elbow effects, that excite the resonances in a non-continuous fashion. This is why the amplitude swings so dramatically on the motor case (the location of vibration points 1 H and 5H). The system goes from brief periods of excitation to brief periods of no excitation. The discharge riser is also moving side to side from the same forces. Although the discharge piping configuration is both non-standard and less than optimum for this application, it poses no threat to the long-term reliability of either the pump or the motor. The only negative impact is on vibration levels relative to a generic standard.2. The balance condition of the motor and pump are acceptable with no corrective action required at this time.3. The shaft alignment between the motor and the pump is acceptable for long-term operation.

4. There is no evidence of motor bearing wear.Machinery Solutions, Inc. recommended the following actions: 1. Create a new IST vibration data point configuration within the data collector database to use an overall level that is generated from spectral data above 950 cpm. This will eliminate the energy from the resonances from the data set and still allow for protection from bearing degradation, impeller degradation, and motor malfunctions.

The only potential failure mode that could occur within this excluded frequency range would be a fundamental train pass frequency generated by a rolling element bearing. This frequency only occurs with increased bearing clearance. On vertical machines, this increased bearing clearance causes increased bearing compliance and the 1X component will become larger. The 1X change will be evident in the monitored data set.Revision 0Pae3 Page 38 Cooper Nuclear Station Fifth Interval lnservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued)

2. Continue to acquire the old data points with the low-frequency data "for information only" to verify that the system response does not change.III. Attempts to Lower Vibration (Key Component No. 3)CNS installed additional pipe restraints during the preoperational period in order to reduce piping-induced vibrations.

Testing on October 26 and 27, 1973, following the installation of these new supports, demonstrated significantly reduced vibrations. Low-frequency piping-induced vibrations continued, but with reduced amplitude following the installation of the pipe restraints. However, the issue resurfaced in the early 1990s when additional vibration points were recorded, more strict acceptance criteria were adopted for vibrations, and new technology was incorporated into the CNS vibration program.These new points were more influenced by the low-frequency piping-induced vibrations than the one or two points recorded in the 1 980s. It was evident that the piping-induced vibrations were still pievalent with thc CS-P-B pump.In 1993, a deficiency report was written to address increased frequency IST testing of CS-P-B due to vibration. It was suspected that the pump vibrations were piping induced.Preliminary investigation of the vibration issue concluded that cavitation at the CS test return line throttle valve and/or restriction orifices was likely causing the elevated piping vibration in both CS System loops. Vibration testing of the CS piping confirmed this conclusion. To reduce these flow-induced vibrations, DC 94-046 was developed to replace the existing simple, single-stage orifices on both CS subsystem test return lines with multi-stage orifices. Post-installation testing with these multi-stage orifices demonstrated lower vibration levels on CS-P-A, but higher vibration levels on CS-P-B. A multi-hole single-stage orifice was fabricated and installed in the CS-P-B test return line (and later in the CS-P-A test return line) with significantly improved results. Visual observation and vibration data collected during acceptance testing determined that CS-P-B pump vibrations had been reduced, but one direction (location 1H in Figure 1) still demonstrated peak velocity reading in the alert range. The pump vibrations in the 1H direction were occurring at frequencies much lower than the pump operating speed.The major vibration peaks were occurring at approximately 700 (cpm), while the pump speed is at 1780 cpm, indicating that the vibration was piping induced. It was also observed during acceptance testing that vibrations were minimal during operation in the minimum flow condition. IV. Spectral Analysis (Key Component No. 4)Figures 7 through 11 in this relief request show spectrum plots for CS-P-B, as well as spectrum trends. These plots show that the peak energy spikes for points 1H and 5H Revision 0 Page 39 Revision 0 Page 39 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) remain below one-half pump running speed and that the pump vibration signature remains fairly uniform. Figure 12 shows that pump differential pressure is consistently acceptable. This data validates the analysis performed by Machinery Solutions, Inc., and the earlier conclusions that the elevated vibrations are piping induced, and not indicative of degraded pump performance. No pump or motor faults and/or degradation are evident in the spectral analysis for this pump. This test data also shows that the vibrations experienced remain in the region of the CS-P-B pump-motor-piping system natural frequency, at less than half the pump's operating speed.Vibrations occurring at these low frequencies are not expected to be detrimental to the long-term reliability of either the pump or the motor. Typical pump faults, i.e., impeller wear, bearing problems, alignment problems, shaft bow, etc., would result in measurable vibration response in frequencies equal to or greater than one-half of the pump's running speed. Such faults would also be evident in pump trends. However, the vibrations are being expe ienced below one-half pump operating speed, have existed since initial operation, and are not trending higher. Visual inspection by Machinery Solutions, Inc., in 2001 of the pump base plate, soleplate, and grout, identified no visible cracks or degradation. Further, they concluded that the balance condition and shaft alignment of the pump and motor were acceptable, and detected no evidence of motor bearing wear.D. Maintenance History The maintenance history for CS-P-B reflects that there have been no significant work items applicable to CS-P-B due to the low-frequency vibrations that have been experienced since the construction phase of the plant. A review of maintenance history for the CS-P-B pump and motor was performed. The search consisted of a historical review of CS-P-B pump and motor maintenance in addition to a more general search of CS System vibrational issues. This search identified that the pump and motor installed in the plant today is the same combination that was installed during the construction phase of the plant. Some of the key items reviewed arc summarized below: 1. 1973: Additional supports installed on "B" CS System during pre-operational stage. As discussed previously, this resulted in lowering CS-P-B vibrations.

2. January 1977: Vibration eliminator on "B" CS test line, CS-VE7, required tightening of wall plate bolts per Maintenance Work Request (MWR) 77-1-10.Bolts in pipe clamp were replaced and clamp was realigned.

Design was determined to be adequate, but lock washers should be used to prevent recurrence of the problem. MWR 77-1-262 completed this action.Revision 0 Page 40 Revision 0 Page 40 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued)

3. April 1989 (Work Item [WI] 89-0269);

November 1991 (WI 91-1507), February 1993 (MWR #92-2876): CS-P-B stator end turn bracing brackets inspected for stress corrosion cracking or unusual conditions such as loose bolts or bending.No cracks, loose bolts, or other unusual conditions were observed.4. March 1993: A magnetic particle examination of CS-P-B support attachment weld revealed an indication at Lug #5 of the pump support. The indication was ground out, repaired, and retested satisfactorily. The indication was very small and would not have affected the overall stiffness of the pump. In 2003, no recurrence of this indication was identified.

5. April 1993: Work Order #93-1631 was initiated due to mechanical seal leakage.A complete inspection of the pump/motor was also completed.

The pump was found with the keyway not properly aligned with the mechanical seal, causing the leakagc. Thc impeller was found to have minor pitting at the base of the wear ring area. The pump casing and cover had minor erosion and pitting. No significant problems with the pump or motor were noted.6. July 1994: Bolt torque checked for lower end bell and lower bearing housing on CS-P-B motor due to a loose bolt found on the "A" RHIR pump motor. No movement on lower bearing housing bolts. Movement of lower end bell bolts were as follows: 1/16 flat on #1, 3, 4, and 5 and no movement on #2, 6, 7, and 8.These were very minor adjustments.

7. Late 1994: DC 94-046 installs new orifices in CS-P-B test line. As previously discussed, this reduced piping deflections in the test line.8. Oil Samples (Dates: 09-22-95, 10-22-95, 11-24-95, 02-28-97, 03-26-98, 04-05-99, 01-24-00, 12-26-00, 10-28-02, 08-30-04, 01-05-05, 08-14-06, 02-28-07, 08-14-07, 02-11-08, 08-14-08, 02-19-09, 08-12-09, 02-09-10, 08-25-10, 03-11-11, 09-02-11, 12-13-11, 03-02-12, 08-24-12, 02-12-13, 08-13-13, 02-11-14, 08-13-14): Periodic Oil Sample Analysis of the upper and lower motor bearings in accordance with Preventive Maintenance Program. Results of CS-P-B Motor oil analysis were satisfactory with no corrective actions required.9. Numerous Visual Motor Inspections completed satisfactory (i.e., January of 2002): Visual motor inspection satisfactory per Work Order #4199724.10. February 2003: Notification
  1. 10225272 identified an indication approximately 3/8" on a CS-P-B integral attachment (CS-PB-Al).

The indication is at the top of one of the small gusset supports where the gusset is welded to the cast pump Revision 0Pae4 Page 41 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) bowl extension (different spot than the 1993 indication). Within Engineering Evaluation 03-030, the indication was determined to be on the gusset side of the weld and appears to be an incomplete fusion of the weld and not a service load-induced flaw. Poor accessibility was the most likely cause. Engineering Calculation 03-007 demonstrated that, even if the five minor gusset plates were ignored, the pump support is still qualified under the most severe design loads.This search of the maintenance history, covering a time period of approximately forty years, identified no significant maintenance or corrective actions that had to be implemented for the "B" CS pump and motor due to the piping-induced vibrations. Only minor indications were noted on the pump impeller and casing during the last significant motor/pump disassembly in 1993.No other documentation of pump/motor disassembly inspection results was found during this review. Oil analyses of the CS-P-B lower and upper motor bearing housings were found to be satisfactory for all the results documented since 1995 to the present. Wear metals, contaminants, additives, etc., were all at acceptable levels. The addition of pipe supports in 1973 and new orifices in the test lines were necessary modifications and were previously discussed. Other than these modifications, only minor corrections have been made with pipe and/or pump supports (tightening bolts, minor indication, etc.), none of which were found to be significant. Therefore, the maintenance history supports the basis of this relief request in that the piping-induced vibrations occurring on CS-P-B have not degraded the pump or motor in any way.E. Basis for Code Alternative Alert Values for Points 1H and 5H By this relief request, NPPD is proposing to increase the absolute alert limit for vibration points 1H and 5H from 0.325 in/s to 0.400 in/s. The piping-induced vibration, which occurs at low frequencies, occasionally causes the overall vibration value for these two points to exceed 0.325 mI/s, resulting in CS-P-B being on an increased test frequency. However, several expert analyses and maintenance history reviews have shown that this piping-induced vibration has not resulted in degradation to the pump. Additionally, the overall vibration levels have remained steady over the past ~-24.5 years. Therefore, it has been demonstrated that doubling the test frequency under the current conditions does not provide additional assurance as to the condition of the pump and its ability to perform its safety function.These new values are reasonable as they represent an alternative method that still meets the intended function of monitoring the pump for degradation over time while keeping the required action level unchanged. The proposed values encompass the majority of the historical values, but not all of them (reference Figures 2a, 2b, 3a, 3b). With these new values, a reading above 0.400 in/s would require NPPD to place the pump on an increased testing frequency and to evaluate the pump performance to determine the cause Revision 0Pae4 Page 42 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps" and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) of the reading. It is expected that a small amount of degradation occurring in the pump or a slight increase in the piping-induced vibration would be quickly identified with these new parameters. The new alert limits will still allow for early detection of pump degradation or piping-induced vibration increases prior to component failure, while the required action absolute limit will remain at the code value of 0.700 in/s. Therefore, the intent of the code will be maintained. Conclusions Several expert evaluations have documented that no internal pump or motor degradation is occurring due to the piping-induced vibration, which has been present since the pre-operational testing time period. The available vibration data over the past years and differential pressure data over nearly the past -~20 years supports this fact as essentially no degradation has been indicated. A maintenance history review and review of oil analyses results further supports these conclusions. Based on this information, CNS concludes that doubling the test frequency for CS-P-B does not provide additional information nor does it provide additional assurance as to the condition of the pump and its ability to perform its safety function. Testing of this pump on an increased frequency places an unnecessary burden on CNS resources. All four key components discussed in NUREG/CP-01 52 have been addressed in detail, supporting the alternative testing recommended in this relief request.CNS concludes that CS-P-B is operating acceptably and will perform its safety function as required during normal and accident conditions. The increased alert limits proposed for vibration points 111 and 511 in this relief request will continue to assure long-term reliability of CS-P-B.During the performance of CS-P-B inservice comprehensive pump testing, or any other time vibrations are recorded to determine pump acceptability (i.e., post-maintenance testing, other periodic testing, etc.), pump vibration shall be monitored in accordance with ISTB-3510(e) and ISTB-3540(a). The acceptance criteria for vibration points 2H, 3H, and 1V will follow the criteria specified in ISTB Table ISTB-5121-1. The acceptance criteria of vibration points 1 H and 5H1 will have increased absolute alert limit values of 0.400 in/s. The absolute required action limits for all points will continue to be 0.700 in/s in accordance with ISTB Table ISTB-5121-1. The absolute alert and required action limits for all vibration points associated with CS-P-B are summarized in the table below.Revision 0Pae4 Page 43 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Absolute Vibration Acceptance Criteria for CS-P-B: Vibration Acceptable Range Alert Range Required Action Parameter Range 111 < 0.400 in./sec. > 0.400 in./sec. > 0.700 in./sec.5H1 < 0.400 in./sec. > 0.400 in./sec. > 0.700 in./sec.1V < 0.325 in./sec. > 0.325 in./sec. > 0.700 in./sec.2H1 < 0.325 in./sec. > 0.325 in./sec. > 0.700 in./sec.3H < 0.325 in./sec. > 0.325 in./sec. > 0.700 in./sec.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was previously approved for the fourth ten-year interval at CNS as Relief Request RP-07 (TAC Nos. MC8837, MC8975, MC8976, MC8977, MC8978, MC8979, MC8980, MC8981, MC8989, MC8990, MC8991, and MC8992, June 14, 2006).Revision 0 Page 44 Revision 0 Page 44 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) CS-P-B Figures Figure Description Attach. 2 Number Page Number 1 CS-P-B Vibration Monitoring Points 46 2a CS-P-B Vibration Point lH from November 1996 to the Present 47 2b CS-P-B Vibration Point 11H from April 1990 to the Present 48 2c Trend of Vibration Point 1 H with Data Below One-Half Pump 49 Running Speed Filtered from May 2010 to the Present 3a CS-P-B Vibration Point 5H from November 1996 to the Present 50 3b CS-P-B Vibration Point 5H from April 1990 to the Present 51 3c Trend of Vibration Point 5H with Data Below One-Half Pump 52 Running Speed Filtered from February 2010 to Present 4a CS-P-B Vibration Point 1V from November 1996 to the Present 53 4b CS-P-B Vibration Point 1V from April 1990 to the Present 54 5a CS-P-B Vibration Point 2H from November 1996 to the Present 55 5b CS-P-B Vibration Point 2H from April 1990 to the Present 56 6a CS-P-B Vibration Point 3H from November 1996 to the Present 57 6b CS-P-B Vibration Point 3H from April 1990 to the Present 58 7 Spectral Trend for Vibration Point 111 59 8 Spectral Trend for Vibration Point 5H 60 9 Spectral Trend for Vibration Point 1 V 61 10 Spectral Trend for Vibration Point 2H 62 11 Spectral Trend for Vibration Point 3H 63 12 CS-P-B Differential Pressure since January 1995 to the Present 64 Revision 0Pae4 Page 45 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) 1V 5H-~3 H--I DISCHARGE,----SUCTION 62CS101A Figure 1 CS-P-B Vibration Monitoring Points Revision 0 Page 46 Revision 0 Page 46 Cooper Nuclear Station F~fih Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) ~56 9 V~(1H1~*~t ~~~14~2O0O 01 4~6Q#II i Date Figure 2a CS-P-B Vibration Point 1H1 from November 1996 to the Present Revision 0 Page 47 Revision 0 Page 47 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Figure 2b CS-P-B Vibration Point 11H from April 1990 to the Present Revision 0 Page 48 Revision 0 Page 48 Cooper Nuclear Station JFih Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) CORE SPRAY PUMP MOTOR B I 1H- -MOTOR UPPR HORIZO'NTAL SOUTIfH(H0) 14 83-1K]Hz 1ST-Baseline-Value: 0.1 35 5118/1198* AM r- 2f12.12015 15 i 20:11PMd Amn0: 1r 20 O02e-1112412012 5:32:33 PM 0392 V -DG Pk = 0 394 LOA=D -100.00 RPN-= 1780.0 (2967liz)021 -I ,,"0.14-0.07 -L I 0 -~ _____________________________________ Frm7V113 I 0 1o00o 20000 30000 Frequency (CPM)00000 o00000 rct32.10 Amp: 0.00019 List of Trend Points Station: REACTOR BUILDING Machine: CS-MOT-B --> CORE SPRAY PUMP MOTOR B Meas Point: 1H --> MOTOR UPPR HORIZONTAL SOUTH (H01)Parameter: 14.83-1KHZ (PK Velocity in In/Sec)Date Time 10-May-10 14:21 16-Nov-10 13:09 21-Apr-11 15:06 24-May-11 13:43 23-Nov-11 13:32 23-May-12 10:51 24-Nov-12 17:32 Value.103.109.135.110.122.125.113 Date Time Value 16-May-13 11:28 .109 13-Nov-13 12:35 .108 13-May-14 22:51 .124 14-Aug-14 10:55 .112 12-Oct-14 02:08 .130 10-Nov-14 23:21 .120 12-Feb-15 23:20 .120 Figure 2c Early Warning Limits--Alert Limit Values--Fault Limit Values--.169.300.700 Trend of Vibration Point 1H1 with Data Below One-Half Pump Running Speed Filtered from May 2010 to the Present0 Page 49 Cooper Nuclear Station F/ith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Figure 3a CS-P-B Vibration Point 511 from November 1996 to the Present Revision 0 Page 50 Revision 0 Page 50 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RIP-07 Core Spray Pump B Vibration Alert Limits (Continued) Figure 3b CS-P-B Vibration Point 5H from April 1990 to the Present Revision 0 Page 51 Revision 0 Page 51 Cooper Nuclear Station Ffi~h Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) CORE SPRAY PUMP MOTOR B81 5H -MOTOR UPPR HORIZONTAl. W.'EST (H05)40 C ED 0~14.83-1KI-4z 1ST-Baseline -Value: 0.269 5/11811 998 1 2:00 AM 211 212 015 1 1:21:36 Ph Amp. 0.266 021-211212015 1121;36 PM 0.374 V -ID Pk 0 0374 LOAO- 100 00 RPM = 1780.J0 (29.67 lIZ).0-14-.1 01 I .~10000 20000 30000 Frequency (CPM)80000-I F~eci31~4 80000 04d.1 ~'9Amp 0 0O~504 List of Trend Points Station: REACTOR BUILDING Machine: CS-MOT-B --> CORE SPRAY PUMP MOTOR B3 Meas Point: 5H --> MMOTOR UPPR HORIZONTAL WEST (H05)Parameter: 14.83-1KHZ (PK Velocity in in/Sec)Date Time 09-Feb-10 14:48 10-Feb-10 10:04 10-May-10 14:21 16-Nov-10 13:09 21-Apr-11 15:06 24-May-11 13:43 23-Nov-11 13:33 23-May-12 10:52 Value.252.252.238.252.238.201.215.232 Date Time 24-Nov-12 17:32 16-May-13 11:30 13-Nov-13 12:35 13-May-14 22:52 14-Aug-14 10:56 12-Oct-14 02:09 10-Nov-14 23:22 12-Feb-15 23:21 Value.229.223.243.262.224.229.250.266 Early Warning Limits--Alert Limit Values--Fault Limit Values--.309.300.700 Figure 3c Trend of Vibration Point 5H with Data Below One-Half Pump Running Speed Filtered from February 2010 to the Present Revision 0 Page 52 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Figure 4a CS-P-B Vibration Point 1V from November 1996 to the Present Revision 0Pae3 Page 53 Cooper Nuclear Station F~th Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Figure 4b CS-P-B Vibration Point 1V from April 1990 to the Present Revision 0 Page 540 Page 54 Cooper Nuclear Station F~fih Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Figure 5a CS-P-B Vibration Point 2H from November 1996 to the Present Revision 0 Page 55 Revision 0 Page 55 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Figure 5b CS-P-B Vibration Point 2H from April 1990 to the Present Revision 0 Page 56 Revision 0 Page 56 Cooper Nuclear Station F~ith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Figure 6a CS-P-B Vibration Point 3H from November 1996 to the Present Revision 0 Page 57 Revision 0 Page 57 Cooper Nuclear Station Ffifh Interval Inser-vice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued)

  • A Vb~3N)A*P I U v'~ ~O(~sc)0 ror.miniq c~-p ~ I95~' ~o~.I 041 S 11~U1S96 I6fl4~OOO 0tiZ~O04 04~~I201I t1fl~Z0t4Figure 6b CS-P-B Vibration Point 3H from April 1990 to the Present Revision 0 Page 58 Revision 0 Page 58 Cooper Nuclear Station Ffi~h Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued)

CORE SPRAY PUMP MOTOR B / 1H -MOTOR UPPR HORIZONTAL SOUTH(H01) r-) 0 2112/2015 11:20:11 PM SRoute 0.302 V-DG PR= 0.302 LOAD= 100.00 RPMl= 1700.0.. .. (29.67 Hz)-i Freq:1 26060 60000 Ord.0 708 mp: 0.00644 30000 Frequency (CPM)CORE SPRAY PUMP MOTOR B / 1H -MOTOR UPPR HORIZONTAL SOUTH(H01) a)CC, C C C a)0~l-0.28;I-I k!0.28', 2/12/2016~11/10/2014 _/1/o,,20?14 6//13/2014-I 1/13/2013 616/2013/11/24/2012 6/¢523/2012 lLi , ,11/23/'2011 ,/24/2011 42/2011/11/16/2010 6110/20"10 60000l -- m¸ = I m I It-LL ..t 16000 30000 Frequency (CPM)46000 Figure 7 Spectral Trend for Vibration Point 1H Revision 0 Page 59 Revision 0 Page 59 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) CORE SPRAY PUMP MOTOR B / 5H -MOTOR UPPR HORIZONTAL WEST (H05)028.021 C_).'0.14 0.07.2/12/2015 11:21:36 PM Route 0.374 v-Do Pk= 0.374 LOAD = 100.00 RPM=1 780.0..... (29.67 Hz)L...Ut. ., ..a 10000 20000 306oo Frequency (CPM)40000 506o0 i~~ Freq:1 250.0 80000 0'rd.0.08 Amp; 0 00770 CORE SPRAY PUMP MOTOR B 1 5H -MOTOR UPPR HORIZONTAL WEST (HO5)t-0.28 U)C,)C C 0 U)V~1L~1 I to *t'U)-~I~1=~ I.. I*Illi .1 I.1i~C.11 t 1 .1 III I I I LI:1 t1.~,.0.28 -II b i J A l.1m I I-I 211212015/ 11/1012014 S10112120 14.. 5/1312014.'I 11/13.r2013 ,5/11/2013 .11124/20125,2312012 111231"2011 j"5,2412011 S412112011 11/1612010 S5/101'2010 L L II ,J h 1= .11 S II A '.i~ -Ph KI..J.. I I, I 2/1212015 11:21 PM RPM: 1780.0 Freq: 59810.6 Ord:33.60 Amp: 0.0001 II [ ............ v r I 0 100003 20000 ao00o Frequency (CPM)400(o 6o6000( 60000 Figure 8 Spectral Trend for Vibration Point 5H Revision 0 Page 60 Revision 0 Page 60 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) CORE SPRAY PUMP MOTOR B /1V -MOTOR UPPR AXL, .(V01)0.2.2/12/2015 11:19.19 PM Route 0.191 V-DO Pk= 0.191 LOAD =100 00 RPM= 1-I780.0 (29.67 Hz)0.15-5, C U 0 15 0~cl.1 0.05 LL t.J~-. -.10000 3000 0~requency (CPM)40600 so6oo Freq5e6~,8 60000 Ord.32.96 Aii~p 0000~CORE SPRAY' PUMP MOTOR B / lV -MOTOR UPPR AXLAL. (V0l)-0.2-0O.1 0.1-I'F-~r I---- -&L .1 Ii I I*1 I 2/12f2Qi~11/10/2014 10/12/2014 8/14'2014 9/12/2014 1111 ~'2O 1 Z~9/19/2013 11/2412012 6t22t2012 11/2312011 ~/24t2O1 i 4/21/2011 11119/2010 0/10/2010 I.1 1 11 i II~---~- I ViY..--.L, I I I1'-~L.2t12.,201S 11.19 PM RPM: 0 Freq:5901 0 600001-- -I i ii J i [ i f 0 ICMXX0 20000 30X000 Frequency (CPMI 40000 50000 500000 Figure 9 Spectral Trend for Vibration Point 1V Revision 0 Page 61 Revision 0 Page 61 Cooper Nuclear Station F~fih Interval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) CORE SPRAY PUMP MOTOR 0 / 2H -MOTOR LWR HORIZONTAL SOUTH(H02) 2/12/2015 11:22:21 PM 0.08--0.67-S0.04-S00027-0.013.]Route 0.104 V-DG Pk=0.105 LOAD = 10000 RPM = 1780.0 (29.67 Hz)L L4~L I[II ; ---_+. ...... ., ,L , + , ,+,.,.0 ioo 20000 30000 40000 5000 60000 Frequency (CPM)CORE SPRAY PUMP MOTOR B ! 2H -MOTOR LWdR HORIZONTAL SOUTH(H02) ._)O0¢.#'0 0 U)2/i41 2/1 2015 11/10/2014 7012/2014 812014 2/1 2/2015 11:22 PM RPM: 1780.0 Freq:59968.1 Ord:33.69 Amp: 0 0005 3 30000 40000 50000 50000 Frequency (CPM)Figure 10 Spectral Trend for Vibration Point 2H Revision 0 Page 62 Revision 0 Page 62 Cooper Nuclear Station Ffi~h Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) RX -CORE SPRAY PUMP MOTOR B CS-MOT-B -3H MOTOR LWR HORIZONTAL WEST (H03)0.15-017-2112(2015 11:23:07 PM... Route 0.183 V -DG Pk = 0.183.LOAD = 100 00 RPM = 1780.0 29,67 Hz)o0.06-(]3 0.03-Frecj £9F5:3 0 D~d 33 62 Amp C ['6039 Frequency (C PM)CORE SPRAY PUMP MOTOR B I 3H -MOTOR LWR HORIZONTAL WEST (H03)U)C,, C C 0 U)2112/2015/, t 810/1212014 611412014 5/132014 11/13/'201 -0,1 S1112412012 512312012/1112312011 4J/2/24/2011I 4212011"11/1182010 211 2"2015 11:23 PM RPM: 1780.0 Freq: 59968 .1 0rd:33 69 Amp: 00004 Frequency (CPM)Figure 11 Spectral Trend for Vibration Point 311 Revision 0 Page 63 Revision 0 Page 63 Cooper Nuclear Station Fifth lnterval Inservice Testing Program for Pumps and Valves Relief Request RP-07 Core Spray Pump B Vibration Alert Limits (Continued) Cooper Nuclear Station F~ih Interval Inservice Testing Program for Pumps and Valves Relief Request RLP-07 Core Spray Pump B Vibration Alert Limits (Continued) Figure 12 CS-P-B Differential Pressure Since January 1995 to the Present Revision 0 Page 64 Revision 0 Page 64 Cooper Nuclear Station Ffifh Interval Inser-vice Testing Pro gram for Pumps and Valves Relief Request RP-08 Comprehensive Pump Test Upper Limit Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected REC-P-A/B/C/D Reactor Equipment Cooling Pumps RHR-P-A/B/C/D Residual Heat Removal Pumps SW-P-A/B/C/D Service Water Pumps 2. Applicable Code Edition and Addenda ASME CM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTB-5123 "Comprehensive Test Procedure," (e), refers to Table ISTB-5121-1, which utilizes a multiplier of 1.03 times the reference value for the comprehensive pump test's upper "Acceptahie, Range" and "Required Action Range High" criteria.ISTB-5223 "Comprehensive Test Procedure," (e), refers to Table ISTB-5221-1, which utilizes a multiplier of 1.03 times the reference value for the comprehensive pump test's upper "Acceptable Range" and "Required Action Range High" criteria.ISTB-5323 "Comprehensive Test Procedure," (e), refers to Tables ISTB-5321-1 and ISTB-5321-2, both of which utilize a multiplier of 1.03 times the reference value for the comprehensive pump test's upper "Acceptable Range" and "Required Action Range High" criteria.4. Reason for Request Occasionally, NPPD has had some difficulty with implementing the high required action range limit of 1.03% above the established hydraulic parameter reference value due to normal data scatter. NPPD has had to address an inoperability of a pump on at least two occasions during the fourth ten-year interval in which a pump was declared inoperable during a comprehensive pump test due to exceeding this upper limit. The result was that the plant had to enter (or remain in) an applicable Technical Specification Limiting Condition for Operation (LCO) for reasons other than a pump degradation issue.Based on the similar difficulties experienced by other Owners, ASME OM Code Case OMN-1 9 was developed and has been published in the 2011 Addenda of the ASME CM Code. The white paper for this code case, Standards Committee Ballot 09-6 10, record 09-657, discussed the impact of instrument inaccuracies, human factors involved with setting and measuring test parameters, readability of gauges and other miscellaneous factors on the ability to meet the 1.03% acceptance criteria. Industry operating experience is also discussed in the white paper.Revision 0 Page 65 Revision 0 Page 65 Cooper Nuclear Station Ffith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-08 Comprehensive Pump Test Upper Limit (Continued) Code Case OMN-1 9 has not yct bcen approved for use in RG 1.192, "Operations and Maintenance Code Case Acceptability, ASME OM Code." Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(1), relief is requested from the multiplier of 1.03 times the reference value for the comprehensive pump test's upper "Acceptable Range" and "Required Action Range High" criteria, referenced in Tables ISTB-5 121-1, ISTB-5221-1, ISTB-5321-1, and ISTB-5321-2. The proposed alternative would provide an acceptable level of quality and safety.5. Proposed Alternative and Basis for Use CNS proposes to use the ASME OM Code Case OMN-19 as published in the 2011 Addenda of the ASME OM Code for the fifth ten year interval IST Program. The ASME OMN-19 Code Case allows for the use of a multiplier of 1.06 times the reference value in lieu of the 1.03 multiplier for the comprehensive pump test's upper "Acceptable Range" criteria and "Required Action Range, High" criteria referenced in the applicable ISTB test acceptance criteria tables ISTB-5121-1, ISTB-5221-1, ISTB-5321I-1 and ISTB-5321-2. The bases for the approval of OMN-1 9, as discussed in the Standards Committee Ballot white paper, are summarized below: 1) Instrument inaccuracies of measured hydraulic value.2) Instrument inaccuracies of set value and its effect on measured value.3) Instrument inaccuracies and allowed tolerance for speed.4) Human factors involved with setting and measuring flow, DiP, and speed.5) Readability of Gauges based on the smallest gauge increment.

6) Miscellaneous Factors.These inaccuracies may cause the measured value to exceed the existing code allowed comprehensive pump test's upper "Acceptable Range" criteria and the "Required Action Range, High" criteria of 3%. The new upper limit of 6%, as approved in Code Case OMN-19, will eliminate declaring the pump inoperable and entering an unplanned Technical Specification LCO or will eliminate the extension of an existing LCO.As a condition for using OMN- 19, CNS will implement a pump periodic verification (PPV) test program to verifyr that a pump can meet the required differential (or discharge) pressure, as applicable, at its highest design basis accident flow rate, as discussed in Mandatory Appendix V, which was published in the 2012 Edition of the ASME OM Code. CNS will not be required to perform a PPV test if the design basis accident flow rate in the licensee's safety analysis is bounded by the comprehensive pump test or Group A test. Also, if a pump does not have a design basis accident flow rate, then a PPV test is not required.

Therefore, any IST pump that is utilizing the 1.06 multiplier for the comprehensive pump test will meet this condition. On June 30, 2015, in a response to the Nuclear Regulatory Commission's Request for Additional Information, CNS provided the following list of pumps that RP-08 is applicable to along with the requested information. The last column of the table also indicates which pumps will have a pump periodic verification (PPV) test based on the current design basis accident flow rate and the Revision 0 Page 66 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RP-08 Comprehensive Pump Test Upper Limit (Continued) current comprehensive pump test flow rate. As is required by the 2012 ASME GM Code, Mandatory Appendix V, Section V-3000(e), the basis for the PPV test parameters will be documented by the owner.Instrument inaccuracies associated with the PPV test parameters will be accounted for within the safety analyses and/or within the test acceptance criteria. CNS considers this a clarification to Mandatory Appendix V, Section V-3000(f), which states that the owner shall account for the pump periodic verification test instrument accuracies in the test acceptance criteria. Although not expected, any flow rate changes associated with this table would be available for NRC inspection, upon request.RP-08 IST Class 1, 2, and 3 Applicable Pumps Pump Pump Pump Type ASME ASME Design Basis IST PPV Test Name Number .Code CM Code Accident Flow Comprehensive Required Class Category Rate (gallons Pump Test (Yes/No)per minute) Flow Rate (gallons per________minute) Reactor Horizontal Equipment REC-P- centrifugal 3 GruA41 Cooling A/B/C/D pump GopA461100 No Pumps Residual Vertical Heat RHR-P- centrifugal 2 Group A 7700 7800 No Removal A/BIC/D pump Pumps ____Service Vertical line Water 5500-Ye Pumps A/B/C/D shaft pump 3 Group A 5846 50 e Using the upper limit of 1.06 times the reference value in lieu of the 1.03 multiplier for the comprehensive pump test's upper "Acceptable Range" criteria and "Required Action Range, High" criteria referenced in the applicable ISTB test acceptance criteria tables will provide an acceptable level of quality and safety.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was approved for the fourth ten-year interval at Columbia Generating Station as RP-06 (TAC Nos. MF3847, MF3848, MF3849, MF3851, MF3852, MF3853, MF3854, MF3855, MF3856, MF3857, and MF3858, December 9 and February 9, 2015).Revision 0 Page 67 Cooper Nuclear Station Fifth Interval Inserviee Testing Program for Pumps and Valves Relief Request RP-09 Variance Around the Reference Values Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected CS-P-A/B HPCI-P-MP &HPCI-P-BP RCIC-P-MP REC-P-A/B/C/D RHR-P-A/B/C/D SW-P-AIB/C/D SW-P-BPA/B/C/D Core Spray Pumps High Pressure Coolant Injection Main & Booster Pumps Reactor Core Isolation Cooling Main Pump Reactor Equipment Cooling Pumps Residual Heat Removal Pumps Service Water Pumps Residual Heat Removal Service Water Booster Pumps 2. Applicable Code Edition and Addenda ASME GM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTB-5 121 Group A Test Procedure ISTB-5 122 Group B Test Procedure ISTB-5 123 Comprehensive Test Procedure ISTB-5221 Group A Test Procedure ISTB-5222 Group B Test Procedure ISTB-5223 Comprehensive Test Procedure ISTB-5321 Group A Test Procedure ISTB-5322 Group B Test Procedure ISTB-5323 Comprehensive Test Procedure 4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(l), an alternative is proposed to the pump testing reference value requirements of the ASME OM Code. The basis of the request is that the proposed alternative would provide an acceptable level of quality and safety.For pump testing, there is difficulty adjusting system throttle valves with sufficient precision to achieve exact flow reference values during subsequent IST exams. Section ISTB of the ASME GM Code does not allow for variance from a fixed reference value for pump testing. However, Revision 0 Page 68 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RP-09 Variance Around the Reference Values (Continued) NUJREG-1482, Revision 2, Section 5.3, acknowledges that certain pump system designs do not allow for the licensee to set the flow at an exact value because of limitations in the instruments and controls for maintaining steady flow.ASME OM Code Case OMN-21 provides guidance for adjusting reference flow/DP to within a specified tolerance during Inservice Testing. The Code Case states "It is the opinion of the Colmmittee that when it is impractical to operate a pump at a specified reference point and adjust the resistance of the system to a specified reference point for either flow rate, differential pressure or discharge pressure, the pump may be operated as close as practical to the specified reference point with the following requirements. The Owner shall adjust the system resistance to as close as practical to the specified reference point where the variance from the reference point does not exceed +2% or -1% of the reference point when the reference point is flow rate, or +1% or -2% of the reference point when the reference point is differential pressure or discharge pressure." 5. Proposed Alternative and Basis for Use CNS seeks to perform Inservice Pump testing in a manner consistent with the requirements as stated in ASME OM Code Case OMN-2 1. Specifically, testing will be performed such that flow rate is adjusted as close as practical to the reference value and within proceduralized limits not to exceed +2%/-l1% of the reference value. Or, if differential pressure or discharge pressure is set, then it will be set as close as practical to the reference value and within proceduralized limits not to exceed +l%/-2% of the reference value.CNS plant operators will still strive to achieve the exact test flow reference values during testing.Typical test guidance will be to adjust flow to the specific reference value. If necessary, additional guidance will be provided such that if the reference value cannot be achieved with reasonable effort, the test will be considered valid if the steady state reference value is within the procedural limits. The procedural limits will be carefully determined on a case by case basis, and will not exceed the limits provided in Code Case OMN-2 1. The test will be considered valid if the steady state reference value is within the proceduralized limits of the procedure. On June 30, 2015, in response to the Nuclear Regulatory Commission's Request for Additional Information, CNS provided the following list of pumps that RP-09 is applicable to along with the requested information. Revision 0 Page 69 Revision 0 Page 69 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RP-09 Variance Around the Reference Values (Continued) RP-09 IST Class 1, 2, and 3 Applicable Pumps Pump Name Pump Number Pump Type ASME ASME OM Code Code Category____ ___ ___ ___ ___ _ __ ___ ___ ___ ___Class Core Spray Pumps CS-P-A/B Vertical centrifugal pump 2 Group B High Pressure Coolant HCIPM Tubndrvnhizta2GopB Injection Main & HPCI-P-BP cerbnt rivega pump ota Booster Pumps HC--P cnrfglpm ru Rctorolin M soatinoum Turbine driven horizontal2GruB ReCtororen IsoationCPCmpM centrifugal pump Reactor Equipment REC-P-AIB/C/D Horizontal centrifugal pump 3GruA Cooling Pumps3GruA Residual Heat Removal RIR.-ABCD Vertical centrifugal pump2GruA Pumps ______Service Water Pumps SW-P-.A/B/CiD Vertical line shaft pump 3 Group A Residual Heat Removal Service Water Booster SW-P-BPA/B/C/D Horizontal centrifugal pump 3 Group A Pumps Using the provisions of this request as an alternative to the specific requirements of ISTB-5 121, ISTB-5122, ISTB-5123, ISTB-5221, ISTB-5222, ISTB-5223, ISTB-5321, ISTB-5322, and ISTB-5323 as described above will provide adequate indication of pump performance and continue to provide an acceptable level of quality and safety.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was approved for Callaway for their fourth ten-year interval as PR-06 (TAC Nos. MF2784, MF2785, MF2786, MF2787, MF2788, and MF2789, July 15, 2014).Revision 0 Page 70 Revision 0 Page 70 Cooper Nuclear Station F~fih Interval Inservice Testing Pro gram for Pumps and Valves ATTACHMENT 3 AUGMENTED PUMP RELIEF REQUESTS AUGMENTED PUMP RELIEF REQUEST INDEX Relief 1Description 1CNS Approval Request No. JJDate ARP-01 ]jElevated Release Point Sump Pump Testing ]I 3-1-2016 ALRP -02 ][Standby Liquid Control Pump Vibration Accuracy ][ 3-1-2016 ARP-03 ]jStandby Liquid Control Pump Testing ]j 3-1-2016 ARP-04 ][Diesel Generator Fuel Oil Transfer Pump Testing ]j 3-1-2016 Revision 0Pae1 Page 71 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request ARP-O1 Elevated Release Point Sump Pump Testing Alternative Provides Acceptable Level of Quality and Safety 1. Augmented Code Component(s) Affected RW-P-Zl Elevated Release Point Sump Pump A RW-P-Z2 Elevated Release Point Sump Pump B 2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004 Edition through 2006 Addenda 3. Applicable Code Reouirement ISTB-3000, "General Testing Requirements." This includes Preservice and Inservice testing of pumps.ISTB-5000, "Specific Testing Requirements." This includes the requirements of a Group A tesit, Comprehensive Pump Test, and Preservice Test.ISTB-6000. "Monitoring, Analysis, and Evaluation. This includes actions to take based on alert and required action levels.4. Reason for Req~uest Augmented Relief is requested from the requirements of ISTB-3000, IST-5000, and ISTB-6000. The proposed alternative provides an acceptable level of quality and safety. This Augmented Relief Request does not require NRC approval.5. Proposed Alternative and Basis for Use Elevated release point sump pumps (Z-Sumps) remove water from the sump supporting the Standby Gas Treatment system drains. Pump failure could cause SGT drain lines to backup and interfere with SGT operation. These pumps operate intermittently depending on sump level. Testing requires manually providing sufficient sump inventory to facilitate pump operation. The pumps receive automatic actuation signals from sump level switches. Pump testing with insufficient inventory would result in damage to the pumps. Additionally, the pumps are submerged and water cooled.Suction pressure, differential pressure, flowrate, and vibration measurements are not feasible due to inaccessibility, the short time the pump runs, lack of available test instrumentation, and the change in suction pressure throughout the test.Testing shall be performed quarterly, utilizing pump start and stop level switches and measuring the time (pump run time) it takes to pump a specified quantity of fluid from the sump. Since the quantity of fluid is constant, pump run time, TM, shall be the test parameter that is measured each test and compared with the corresponding acceptable, alert and action limits. The acceptable Revision 0 Page 72 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request ARP-O1 Elevated Release Point Sump Pump Testing (Continued) range shall be 0.80 to 1.20'1TMr (T1Mrreference run time); the alert range low shall be <0.80 TMr; the alert range high shall be >1.20 TMr to 1.50 TMr; and the required action range shall be>1.50 TMr, not to exceed the maximum pump run time value documented in the Cooper Nuclear Station Inservice Testing Program Basis Document. If run time falls within the alert range, the test frequency shall be doubled until the cause of the deviation is determined and the condition corrected by repair, replacement or an evaluation which resolves the condition. If run time falls into the required action range, the pump shall be declared Non-Functional until the cause of the deviation has been determined and the condition corrected by repair, replacement or an evaluation which resolves the condition. These pumps are ASME non-code class pumps outside the scope of the IST Program. This method of monitoring these pumps provides a level of testing that is commensurate to the level of safety for these components.

6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents A similar relief request was previously approved by CNS for the third and fourth ten-year intervals as ARP-0 1.Revision 0 Page 73 Revision 0 Page 73 Cooper Nuclear Station Ffith Interval Inser-vice Testing Pro gram for Pumps and Valves Relief Request ARP-02 Standby Liquid Control Pump Vibration Accuracy.Alternative Provides Acceptable Level of Quality and Safety 1. Aug~mented Code Component(s)

Affected SLC-P-A Standby Liquid Control Pump A SLC-P-B Standby Liquid Control Pump B 2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004 Edition through 2006 Addenda 3. Applicable Code Requirement Table ISTB-35 10-1, "Required Instrument Accuracy" and ISTB-35 10(a), "Accuracy." Vibration instruments require an accuracy of+/- 5% over the calibrated range.4. Reason for Request Augmented Relief is requested from the requirements of table ISTB-35 10-1 and paragraph ISTB-3510(a) for vibration measurement accuracy. The proposed alternative provides an acceptable level of quality and safety. This Augmented Relief Request does not require NRC approval.5. Proposed Alternative and Basis for Use The SLC pumps function to pump a boron neutron absorber solution into the reactor if the reactor cannot be shut down or kept shut down with control rods. The SLC System also has a safety-related accident mitigation function based on the implementation of the Alternate Source Term (AST) methodology as the radiological source term design basis accident analysis (in accordance with RG 1.183). SLC is credited for controlling the pH of the water in the Suppression Pool, Reactor Vessel (Rx), and Core Cooling systems following a Design Basis LOCA.The Code requires vibration equipment to meet the accuracy of +/- 5% across the frequency response range, which includes the minimum frequency response of 1/3 pump shaft speed. For the SLC pumps, this is 173.3 rpm or 2.8 Hz. The vibration meters used at CNS are calibrated to meet the code down to and including 5 Hz. Currently, there are no calibration points being taken below 5 Hz. Therefore, the accuracy below 5 Hz. may not meet the code tolerance. The average velocity for an IST test is a single, average energy reading. The effect of this potential change in accuracy below 5 Hz., when averaged into the overall reading, is quite small.It would only be a concern if a single frequency in the spectrum were being evaluated between 2 -5 Hz. Furthermore, detection of pump degradation via vibration data is based on changes in vibration measurement from one test to another. Thus, if the calibration accuracy is consistent, then the change in vibration measurement from one test to another is appropriate information for trending purposes. Therefore, existing vibration equipment will provide adequate trending information and may be used for SLC pump vibration data collection. These pumps are ASME non-code class pumps outside the scope of the 1ST Program.Revision 0 Page 74 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Relief Request ARP-02 Standby Liquid Control Pump Vibration Accuracy (Continued) Vibration data for the SLC pumps will be taken with equipment calibrated fr'om 5 Hz. to at least 1000 Hz. at +/- 5% or better, and will not be calibrated to +/- 5% or better below 5 Hz.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents A similar relief request was previously approved by CNS for the third and fourth ten-year intervals as ARIP-02.Revision 0Pae7 Page 75 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request ARP-03 Standby Liquid Control Pump Testing Alternative Provides Acceptable Level of Quality and Safety 1. Augmented Code Component(s) Affected SLC-P-A Standby Liquid Control Pump A SLC-P-B Standby Liquid Control Pump B 2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTB-3400, "Frequency of Inservice Tests" JSTB-53 10, "Preservice Testing" ISTB 5323, "Comprehensive Test Proccdurc" 4. Reason for Request Augmented Relief is requested from the requirements of ISTB-3400, ISTB-53 10, and ISTB-5323. The proposed alternative provides an acceptable level of quality and safety. This Augmented Relief Request does not require NRC approval.5. Proposed Alternative and Basis for Use The SLC pumps function to pump a boron neutron absorber solution into the reactor if the reactor cannot be shut down or kept shut down with control rods. The SLC System also has a safety-related accident mitigation function based on the implementation of the Alternate Source Term (AST) methodology as the radiological source term design basis accident analysis (in accordance with RG 1.183). SLC is credited for controlling the pH of the water in the Suppression Pool, Reactor Vessel (Rx), and Core Cooling systems following a Design Basis LOCA.Each SLC system pump shall be capable of delivering no less than 38.2 gal/min against a system head of 1300 psig to be considered operable. This flow rate is based on the original system design requirement that a single standby liquid control pump be capable of shutting down the reactor from the most reactive condition at any time in core life and maintaining it subcritical during cooldown with all control rods withdrawn in the rated power pattern.The Standby Liquid Control pumps are categorized as Group B pumps since they are standby emergency pumps and are only operated for testing. They are horizontally-mounted reciprocating positive displacement pumps.As an alternative to the code requirement for performing a comprehensive pump test, each of these pumps will have a modified Group A test performed each quarter in place of the Group B quarterly test and the 2-yr Comprehensive pump test. The pumps will be operated at a reference Revision 0 Page 76 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Relief Request ARP-03 Standby Liquid Control Pump Testing (Continued) discharge pressure (Pr) of 1300 psig with pump flow rate (Q) measured and compared to the required action and alert range requirements of Table ISTB-5321 -2 for the group A test, which is more stringent than the Group B testing that would normally be applied each quarter. In addition, vibration measurements will be recorded every 6 months (every other quarter) rather than only once every 2 years during the comprehensive pump test. Vibration measurements will be compared to the range requirements of Table ISTB-532 1-2 for the Group A test. Corrective actions will be taken in accordance with ISTB-6200. Permanently installed plant instrumentation will be used to determine discharge pressures and flow rates. Portable vibration instruments will be used to determine vibration measurements. All instrumentation will meet the accuracy requirements of a Group A test unless specific relief is requested. This level of accuracy is sufficient for these augmented pumps, especially based on the large margin to the minimum flow of 38.2 gpm per pump. However, the current discharge pressure gauge is calibrated to +/-1/2%, which meets the accuracy requirements of the Group A test (+/-2%) and the Comprehensive pump test (+/-1/2%), so no variabilities between the Group A and Comprehensive Pump Test instrumentation currently exist.One of thie requiirements of the comprehensive test is to perform the test at oubotoantial flow (+/-20% of design flow). CNS will meet this requirement each quarter by performing the test at the design flow discharge pressure: Design Point: 1300 psig with a minimum of 38.2 gpm Test Point: 1300 psig with a minimum of 38.2 gpm Although these are Group B pumps, the OM Code allows the substitution of a Group A or comprehensive test. CNS will perform a modified Group A test as stated above such that the acceptance criteria for hydraulic performance will meet the code requirements for a Group A test.Additionally, CNS will perform vibration monitoring on these Group B pumps on a frequency of once every 6 months.The Standby Liquid Control pumps are tested at a set discharge pressure of 1300 psig. Per Table ISTB-5321 -2, the required action range for the Group A flow measurement would be <0.93Qr and > 1.lOQr, with an alert range of 0.93Qr to < 0.95Qr. This is the same as the comprehensive test requirement with the exception that the upper range is >1l.O3Qr. With reference values of approximately only 53-54 gpm, this upper limit for the comprehensive pump test may not encompass the normal data scatter associated with acceptable SLC pump operation. Therefore, CNS will monitor the flow of these pumps at the Group A test criteria each quarter as follows: Acceptable Range 0.95 to 1.10 Qr Alert Range 0.93 to <0.95 Qr Required Action <0.93 Qr or >1.10 Qr CNS will evaluate all ranges against the design conditions to ensure that all procedure lower limits bound the more conservative of the design or ASME GM Code ranges delineated above.Revision 0Pae7 Page 77 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request ARP-03 Standby Liquid Control Pump Testing (Continued) Performance of a substantial flow test each quarter would result in eight sets of data over a two-year period instead of the required one comprehensive test. Monitoring of vibration on these pumps every six months will result in four sets of mechanical data versus the required one every two years. CNS believes this testing regime provides an overall better assessment of pump mechanical and hydraulic health and will determine operational readiness on a quarterly frequency. Additionally, this modified group A positive displacement pump test performed with vibrations will verify that the pump is operating acceptably and may be utilized as the post-maintenance test following significant maintenance. Multi-point preservice testing for positive displacement pumps is not required by the OM Code per ISTB-5323. These pumps are ASME non-code class pumps outside the scope of the IST Program. This method of monitoring these pumps provides a level of testing that is commensurate to the level of safety for these components.

6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was previously approved by CNS for the fourth ten-year interval as ARP-03.Revision 0 Page 78 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request ARP-04 Diesel Generator Fuel Oil Transfer Pump Testing Alternative Provides Acceptable Level of Quality and Safety 1. Augmented Code Component(s)

Affected DGDO-P-DOTA Diesel Generator Fuel Oil Transfer Pump A DGDO-P-DOTB Diesel Generator Fuel Oil Transfer Pump B 2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004 Edition through 2006 Addenda 3. Applicable Code Req~uirement ISTB-3 400, "Frequency of Inservice Tests" ISTB-51 10, "Preservice Testing" ISTB-5121, 'Group A Test Procedure" ISTB-5 123, "Comprehensive Test Procedure" 4. Reason for Request Augmented Relief is requested from the requirements of ISTB-3400, ISTB-51 10, ISTB-5 121, and and ISTB-5 123. The proposed alternative provides an acceptable level of quality and safety.This Augmented Relief Request does not require NRC approval.5. Proposed Alternative and Basis for Use Thc diesel fuel oil transfer pumps 'have an active safety function to transfer fuel oil to the respective diesel fuel oil day tank during normal diesel generator operation. The pumps will automatically start and stop under certain low and high levels in their respective day tanks.Each DGDO system pump must provide sufficient fuel flow to one diesel engine to meet consumption requirements. The design flow rate of the diesel fuel oil transfer pumps is 15 gpm.To support continuous Diesel Generator operation at full load, the transfer pump must be capable of delivering 4.64 gpm. Therefore, significant margin exists for these pumps.The DGDO pumps are conservatively categorized as group A pumps since they are operated routinely in support of diesel generator operation in addition to testing for IST purposes. The pumps are considered vertically mounted centrifugal pumps. Some plants consider these pumps skid-mounted in support of the diesel generators. CNS has decided to individually test these pumps in order to better monitor their performance. As an alternative to the code requirements for performing a Preservice test, Comprehensive test, and a Group A test, a modified Group A test will be performed. Revision 0Pae7 Page 79 Cooper Nuclear Station F/ifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request ARP-04 Diesel Generator Fuel Oil Transfer Pump Testing (Continued) A Comprehensive pump test will not be performed, which would be essentially the same test as the Group A test. During the modified Group A test, the pumps will be operated at a reference flow rate (Qr) of 6 gpm (+/-0.3 gpm) with pump differential pressure (P) measured and compared to its reference value. Deviations from the reference value will be compared to the required action and alert range requirements of Table ISTB-5121-1 for the Group A test. In addition, vibration measurements will be recorded once every 6 months (every other quarter). This 6 month frequency for vibrations is more than adequate based on the steady vibration trends observed over the past several years. The vibration measurements will be compared to the range requirements of Table ISTB-5 121-1 for the Group A test. Corrective actions will be taken in accordance with ISTB-6200. Permanently installed plant instrumentation will be used to determine differential pressures and flow rates. Portable vibration instruments will be used to determine vibration measurements. All instrumentation will meet the accuracy requirements of a Group A test unless specific relief is requested. This level of accuracy is sufficient for these augmented pumps, especially based on the large margin to the minimum flow of 4.64 gpm per pump. The current discharge pressure gauge is calibrated to < +1%, which is lower than the Group A accuracy requirement (+/-2%) and nearly mccts thc Comprchcnsivc pump tcst accuracy requirement (+1/2%), su nijujhnal variabilities between the Group A and Comprehensive Pump Test instrumentation currently exist.One of the requirements of the comprehensive test is to perform the test at substantial flow (4-20% of design flow). Since the accident design flow is such a low value, CNS will continue to test the pump at a value slightly higher than 20% above the design value, which allows for plenty of margin for inaccuracies in instrumentation. Design Point: 4.64 gpm Test Point: 6 gpm CNS will perform a modified Group A test as stated above such that the acceptance criteria for hydraulic performance will meet the code requirements for a Group A test. Additionally, CNS will perform vibration monitoring once every six months (every other quarter).The DGDO pumps are tested at a set flow of 6 gpm (+/-0.3 gpm) and differential pressure is measured. Per Table ISTB-5 121-1, the required action range for the Group A differential measurement would be <0.90 A Pr and > 1.10 A Pr. There is no alert range. This is the same as the comprehensive test requirement with the exception that the upper range is >1.03 A Pr and a lower alert range is in place for the measured differential pressure. With reference values in the low twenties for differential pressure, this upper limit for the comprehensive pump test may not encompass the normal data scatter associated with acceptable DGDO pump operation. Also, since the measured values are low and there is significant design margin built into the pump design, alert ranges for differential pressure are not necessary. Therefore, CNS will monitor the differential pressure of these pumps utilizing the Group A test criteria each quarter as follows: Acceptable Range 0.90 to 1.10 APr Required Action <0.9OA Pr or >l.10A Pr Revision 0 Page 80 Cooper Nuclear Station F~ifh Interval Inservice Testing Program for Pumps and Valves Relief Request ARP-04 Standby Liquid Control Pump Testing (Continued) CNS will evaluate all ranges against the design conditions to ensure that all procedure lower limits bound the more conservative of the design or ASME OM Code ranges delineated above.Performance of a substantial flow test each quarter would result in eight sets of data over a two-year period instead of the required one comprehensive test. CNS believes that there would be no benefits added to implementing a 2-year comprehensive test, which would essentially be identical to the quarterly test with vibrations. Therefore, CNS believes that the proposed testing regime establishes an acceptable assessment of pump mechanical and hydraulic health and will determine operational readiness on a quarterly frequency. Additionally, this modified group A centrifugal pump test performed with vibrations will verify that the pump is operating acceptably and may be utilized as the post-maintenance test (in place of a preservice test) following significant maintenance. This one point test is adequate to verify acceptable pump operation due to the simplification of the function the DGDO pumps perform. The fuel oil is being delivered from one large tank to a smaller tank and the minimum amount of oil is verified to be met. No other modes of function occur. As long as this function may be met, and pump performance is trended within the IST Program, this testing will adequately verify pump operability. Thcsc pumps arc ASME non-codc class pumps outsidc thc scope of the IST Program. This method of monitoring these pumps provides a level of testing that is commensurate to the level of safety for these components.

6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fourth ten-year interval.7. Precedents A similar relief request was previously approved by CNS for the fourth ten-year interval as ARP-04.Revision 0 Page 810 Page 81 Cooper Nuclear Station F~/iIi Interval Inservice Testing Pro gram for Pumps and Valves ATTACHMENT 4 VALVE RELIEF REQUESTS VALVE RELIEF REQUEST INDEX Relief Description 1NRC Approval Date Request No. _____________________

RV-0 1 HPCI Solenoid Operated Drain Valve Testing 21 2-2161 RV-02 ftMain Steam Safety Valve Testing per Code Case OMN- 17 ] 2412416'RV-03___ [Main Steam Safety Relief Valve Testing ] 2-12-16'RV-04 [Control Rod Drive (CRD) Technical Specification Testing 2-12-16'RV-05 ftPerformance-Based Scheduling of PIV Leakage Tests ] 2-12-16'(1) Approved by NRC letter, dated 2-12-16, from Meena K. Khanna, NRC, to Mr. Oscar A. Limpias, Vice President of Nuclear and CNO for CNS Revision 0Pae2 Page 82 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RY-Ol HiPCI Solenoid Operated Drain Valve Testing Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected Valve Class Category System HPCI-SOV-SSV-64 2 B HPCI HPCI-SOV-SSV-87 2 B 2. Applicable Code Edition and Addenda ASME CM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTC-3300 Reference Values -Reference values shall be determined from the results of preservice testing or from the results of inservice testing.ISTC-33 10 Effects of Valve Repair, Replacement, or Maintenance on Reference Values -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... ISTC-3 500 Valve Testing Requirements -Active and passive valves in the categories defined in ISTC-l1300 shall be tested in accordance with the paragraphs specified in Table ISTC-3500-1 and the applicable requirements of ISTC-5 100 and ISTC-5200. ISTC-35 10 Exercising Test Frequency -Active Category A, Category B, and Category C check valves shall be exercised nominally every 3 months except as provided by ISTC-3520, ISTC-3540, ISTC-3550, ISTC-3570, ISTC-5221, and ISTC-5222. ISTC-3560 Fail-Safe Valves -Valves with fail-safe actuators shall be tested by observing the operation of the actuator upon loss of valve actuating power in accordance with the exercising frequency of ISTC-35 10.ISTC-5 151 Valve Stroke Testing -(a) Active valves shall have their stroke times measured when exercised in accordance with ISTC-3500.(b) The limiting value(s) of full-stroke time of each valve shall be specified by the Owner.(c) Stroke time shall be measured to at least the nearest second.ISTC-5 152 Stroke Test Acceptance Criteria -Test results shall be compared to reference values established in accordance with ISTC-3300, ISTC-33 10, or IsTc-3320. ISTC-5 153 Stroke Test Corrective Action.Revision 0 Page 83 Revision 0 Page 83 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-01 IIPCI Solenoid Operated Drain Valve Testing (Continued)

4. Reason for Requnest Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(1), relief is requested from the listed requirements of the ASME OM Code. The proposed alternative would provide an acceptable level of quality and safety.The HPCI turbine and exhaust steam drip leg drain to gland condenser (HiPCI-SO V-S SV-64) and HPCI turbine and exhaust steam drip leg drain to equipment drain isolation valve (HPCI-SOV-SSV-87) have an active safety function in the closed position to maintain pressure boundary integrity of the IIPCI turbine exhaust line. These valves serve as a Class 2 to non-code boundary barrier.These valves are rapid acting, encapsulated, solenoid-operated valves. Their control circuitry is provided with a remote manual switch for valve actuation to the open position and an auto function which allows the valves to actuate from signals received from the associated level switches I{PCI-LS-98 and HPCI-LS-680.

Both valves receive a signal to change disc position during testing of drain pot level switches. However, remote position indication is not provided for positive verification of disc position. Additionally, their encapsulated design prohibits the ability to visually verify the physical position of the operator, stem, or internal components. Modification of the system to verify valve closure capability and stroke timing is not practicable nor cost beneficial since no commensurate increase in safety would be derived.5. Proposed Alternative and Basis for Use CNS has been performing a robust exercise test for these two valves that verifies obturator movement since 1998 on a quarterly basis. In 2001, this test identified some leakage past IJPCI-SOV-SSV64 and the valve was removed and refurbished. For the past -44 years, the exercise test has been completed without any issues. This test is accomplished through the performance of surveillance procedure, 6.H!PCI.204, HPCI-SOV-SSV64 and HPCI-SOV-SSV87 1ST Closure Test. With HPCI not in operation, a demineralized water source is utilized to verify that HPCI-SOV-SSV64 opens when level switch HPCI-LS-680 (turbine exhaust drain pot high level) trips, allowing level in the gland seal condenser to start to rise due to water flow through HPCL-SOV-SSV64. After HPCI-LS-680 resets and I-IPCI-SOV-SSV64 closes, the gland seal condenser level is verified to be steady.Similarly, CNS verifies that HPCI-SOV-SSV87 opens when level switch I-IPCI-LS-98 (turbine exhaust drip leg high) trips, allowing the observation of water flow to a floor drain from a drain pipe downstream of HPCI-SOV-SSV87. After HPCI-LS-98 resets and HIPCI-SOV-SSV87 closes, CNS observes the drain pipe downstream of HIPCI-SOV-SSV87 for gross leakage past the valve.Therefore, CNS verifies valve obturator movement for both valves open and closed while simultaneously verifying the calibration of two level switches.Typically, tests that involve hooking up pressure sources and various amounts of test tubing are not performed on a quarterly basis due to their complexities (i.e. local leak rate tests). In addition, each time this "quarterly" test has been performed, IJPCI unavailability time ('-.5 Revision 0 Page 84 Rev&ion 0 Page 84 Cooper Nuclear Station JFifh Interval Inser-vice Testing Program for Pumps and Valves Relief Request RY-01 Hi-PCI Solenoid Operated Drain Valve Testing (Continued) hours) is consumed in addition to some minor radiological dose. Finally, this exercise test is actually a much better method of determining the valve's operational readiness than a quarterly fast acting stroke time test would have been. Therefore, based on the complexities of the test, consuming unnecessary HIPCI unavailability time and personnel radiation exposure, the exceptional test history dating back to 2001, and the fact that this is a robust test that verifies obturator movement, CNS proposes to exercise each valve to the full closed position, as described, on a 6 month basis.In addition to performing this robust exercise test every 6 months, each solenoid valve will be disassembled and examined for degradation on a periodic basis per the Preventative Maintenance Program. The valve body, insert, piston, plunger/stem assembly, and stem spring will all be examined per criteria outlined in surveillance procedure 6.IHPCI.404. In addition, continuity and the physical condition of the coil will also be checked. The valve and/or valve parts will be refurbished and/or replaced, as necessary, based on this examination. This maintenance shall be performed at an optimized frequency, not to exceed 48 months (2 cycles). The purpose of this enhanced preventative maintenance is to ensure the long term reliability of the components and to monitor for internal degradation. This is consistent with NUREG 1482, Section 4.2.3. The 6 month exercise tests will ensure that the valves are operational and will fulfill thcir safcty function when called upon.On June 9, 2015, in response to the Nuclear Regulatory Commission's Request for Additional Information, CNS provided an explanation as to how the frequency of the preventative maintenance task of disassembly, inspection, and refurbishment is developed, maintained, and optimized. The following several paragraphs and Table 1 were submitted to the NRC.The frequency of the preventative maintenance (PM) task was developed after reviewing the*maintenance and test histories for these two solenoid valves and after reviewing the Electric Power Research Institute (EPRI) recommendations for PMs on solenoid valves. The maintenance history for these valves since 2005 is documented in Table 1 located at the end of this response.A revicw of this data demonstrates that each valve has had two examinations that resulted in minor issues resulting in the replacement of parts (March of 2011 and April of 2014 for HPCI-SOV-SSV64; April of 2012 and April of 2014 for HPCI-SOV-SSV87). For these cases, the PM was doing its job by identifying parts that had minor issues and replacing them prior to them becoming a major issue and impacting the safety function of the valve. The exercise testing performed prior to and after these examinations was completed with acceptable results. As long as the exercise testing of the valves continues to demonstrate acceptable performance and the examination PMs do not identify any major issues that could have impacted the closure safety function, then the maximum frequency of 48 months may be utilized for these HPCI PMs.The maximum frequency of 48 months (2 cycles) is conservative when comparing this frequency to the EPRI PM recommendations; The EPRI recommended task for elastomer replacement and internal inspection of a solenoid valve is 5 years for a severe environment and up to ten years for a mild environment. Cooper Nuclear Station (CNS) considers the location of these valves to be a severe environment, so the maximum frequency allowed by CNS would be one year less than what is recommended by EPRI.Revision 0 Page 85 Revision 0 Page 85 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RV-O1 HLPCI Solenoid Operated Drain Valve Testing (Continued) The frequency will be maintaincd through the CNS work management system and the PM process. A maintenance plan has been established with the necessary tasks required to satisfy the PM. A PM work order with these required tasks is automatically created well ahead of the scheduled due date and is scheduled based on the CNS work schedule process. Any frequency changes must be approved by the IST Engineer.The monitoring of these valves will be done by tracking the proposed six month exercise testing and the results of the internal examinations. As was described in the relief request, CNS has had excellent results with the exercise tests. Based on internal valve degradation, October of 2001 was the last time one of these valves (HPCI-SOV-SSV64) failed its closure acceptance criteria.For clarification purposes, however, there was a system issue in June of 2002 in which foreign material was causing HPCI-SOV-SSV64 to leak. An internal examination identified that there was foreign material found under the valve disc of HPCI-SOV-SSV64, but the valve itself, was examined and found to be in an acceptable condition. The CNS corrective action program addressed the issue and no other foreign material issues have impacted the closure function of these valves since then. Therefore, no internal valve degradation issue has impacted the closure function of these components since October of 2001 and no system issue has impacted the closure function of these components since June of 2002.The frequency of the PMs is optimized by balancing the component reliability with the correct PM frequency. The goal is to ensure that the solenoid valves continue to perform their closure function in a reliable manner without performing the internal examination PMs too frequently. As long as the PM ensures that any minor issue is taken care of prior to it becoming an issue with the closure function of the valve meeting its acceptance criteria, then the frequency is set at an acceptable duration. This, in ,conjunction with acceptable exercise tests, justifies the acceptability of the frequency. If the exercise testing results in a failure of the closure acceptance criteria of one of the solenoid valves, or the examination PM of one of the solenoid valves identifies a significant component issue that may have resulted in the respective valve not bcing able to perform its closure function, then the examination frequency of both solenoid valves shall be moved from 48 month frequencies to 24 month frequencies. From this point, two periodic examinations would have to be performed and completed satisfactorily at the 24 month frequency prior to returning the frequency to the 48 month frequency. In conclusion, the PM was developed based on a review of the maintenance and test history results, and review of EPRI recommendations. The existing frequency will be monitored as acceptable as long as the exercise testing is completed satisfactory and the internal examinations are either satisfactory or identify parts for replacement prior to when the parts issue would have caused a failure with the closure exercise testing. The frequency of internal examinations will be reduced from 48 months to 24 months for both valves if one valve were to fail its acceptance criteria for the closure exercise testing or if the findings of an internal examination of one of the valves results in the determination that it would not have met its closure function. Two successful examinations at the 24 month frequency would be required in order to return the PM(s)to a 48 month frequency. This is how the frequency of the preventive maintenance task of disassembly, inspection, and refurbishment was developed, and how it will be maintained, monitored, and optimized, if approved.Revision 0 Page 86 Cooper Nuclear Station Ffifh Interval Inser-vice Testing Program for Pumps and Valvesi Relief Request RV-01 HPCI Solenoid Operated Drain Valve Testing (Continued) Table 1 : Maintenance historics for HPCI-SOV-SSV64 and HPCI-SOV-SSV87 HPCI-SOV-SSV64 HPCI-SOV-SSV87 02-10-05: Visual exam satisfactory 02-10-05: Visual exam satisfactory (PM work order #4363336) (PM work order #4363336)N/A 06-21-05: Replaced valve at same time as non-essential valve, HPCI-SOV-SSV88, was replaced. Valves are in close proximity. New valve allows parts to be procured.(Corrective Maintenance [CM] work order#4211944)11-7-06: Visual exam satisfactory 11-14-06: Visual exam satisfactory (PM work order #4446767) (PM work order #4446767). 03-18-08: Visual exam satisfactory 03-18-08: Visual exam satisfactory (PM work order #4569097) (PM work order #4569097)08-19-09: Visual exam satisfactory 08-19-09: Visual exam satisfctory (PM work order #4626047) ____ (PM work order #462 6047)03-21-l- 1: Valve replaced for parts reasons 03-22-11: Visual exam satisfactory with a valve upgrade to match that of HPCI- (PM work order #4750715)SOV-SSV87 (CM work order #479 1033) ____________________ 04-24-12: Visual exam satisfactory 04-25-12: Seat plug on the bottom was found (PM work order #4803 767) curled around the edges and was replaced.This issue did not impact the valve's closure function as the previous closure testing was performed successfully (PM work order#4803 767)4-23-13: Visual exam satisfactory 04-23-13: Visual exam satisfactory (PM work order #4895831). (PM work order #4895831). 4-22-14: Plunger found slightly corroded and 4-22-14: Insert showed minor stem assembly was scored in the seating area. erosion/corrosion and plunger/stem has a small Both parts were replaced. Did not impact the groove around seating area. Both parts were valve's closure function as the previous closure replaced. Did not impact the valve's closure testing was performed successfully. function as the previous closure testing was (PM work order #493 8492) performed successfully. ________________________________(PM work order #493 8492)2-10-15: Visual exam satisfactory 2-10-15: Visual exam satisfactory (PM work order 5003464). (PM work order 5003464).Revision 0Pae8 Page 87 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RV-01 ILPCI Solenoid Operated Drain Valve Testing (Continued) The robust 6 month exercise testing and the enhanced preventative maintenance will provide an adequate indication of valve performance and will continue to provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(z)(1), NPPD requests relief from the specific ISTC requirements identified in this request.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents A version of this relief request was previously approved for the fourth ten-year interval at CNS as Relief Request RV-01, Revision 1 (TAG NO. ME7021, August 28, 2012) and Revision 0 (TAG Nos. MC8837, MC8975, MC8976, MC8977, MC8978, MC8979, MC8980, MC8981, MC8989, MC8990, MC8991, and MC8992, June 14, 2006).A version of this relief request was previously approved for the fifth ten-year interval at Dresden Nuclear Power Station as Relief Request RV-23H (TAG Nos. ME9865, ME9866, ME9869, ME9870, ME9871, and ME9872, October 31, 2013).Revision 0 Page 88 Revision 0 Page 88 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RV-02 Main Steam Safety Valve Testing per Code Case OMN-17 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected Valve Class Category System MS-RV-70ARV 1 C Main Steam (MS)MS-RV-70BRV 1 C MS MS-RV-70CRV 1 C MS 2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Codk Requirement ISTC-5240 -Safety and Relief Valves. Safety and relief valves shall meet the inservice test requirements of Mandatory Appendix I.ASME OM Code Mandatory Appendix I, "Inservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants," Section I-1320, "Test Frequencies, Class 1 Pressure Relief Valves," paragraph (a), "5-Year Test Interval," states that Class 1 pressure relief valves shall be tested at least once every 5 years.4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(l), relief is requested from the requirements of ASME GM Code Appendix I, 1-1320(a). The proposed alternative would provide an acceptable level of quality and safety.Section ISTC-3200, "Inservice Testing," states that inservice testing shall commence when the valves are required to be operable to fulfill their required function(s). Section ISTC-5240, "Safety and Relief Valves," directs that safety and relief valves meet the inservice testing requirements set forth in Appendix I of the ASME GM C ode. Appendix I, Section 1-1320(a), of the ASME GM Code states that Class 1 pressure relief valves shall be tested at least once every 5 years, starting with initial electric power generation. This section also states a minimum of 20 percent of the pressure relief valves are tested within any 24-month interval and that the test interval for any individual valve shall not exceed 5 years. Prior to Cycle 28, CNS had refueling cycles of 18 months. With three safety valves, CNS has been meeting the ASME GM Code by removing, Revision 0Pae8 Page 89 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-02 Main Steam Safety Valve Testing per Code Case OMN-17 (Continued) testing, rebuilding, and re-installing one valve per refiueling outage. All three of these safety valves have an acceptable test history since 1997 as will be described in section 5.However, after Refueling Outage (RE) 27 (Fall/20 12), CNS began the current 24-month refueling cycle. The five year frequency was met for the safety valve due in RE28 (Fall/20 14), but a relief request, requesting the use of Code Case OMN-l17, will be necessary in order to continue with the process of testing only one valve each refueling outage for the fifth ten-year interval, beginning March 1, 2016. Without this relief request, CNS would be required to remove and test all three valves within a two cycle frequency (two one outage and one the next) in order to ensure that all three valves are removed and tested in accordance with the ASME OM Code requirements. This testing pattern would ensure compliance with the ASME GM Code requirements for testing Class 1 pressure relief valves within a 5 year interval.Extending the test interval to 6 years, as described in Code Case OMN- 17, would allow CNS to continue with the current method of removing, testing, rebuilding, and re-installing one safety valve per outage so that all three safety valves would be replaced over three refuel cycles (i.e., years).Without Code relief, thc incremenital outage work due to the inclusion of an additional safety valve every other outage would be contrary to the principle of maintaining radiation dose As Low As Reasonably Achievable (ALARA). The removal and replacement of the additional safety valve every other outage results in an additional exposure of approximately 450 millirem (mrem) to 726 mrem. This estimate is based on the actual radiation received to remove and re-install a safety valve each of the last three refueling outages.In accordance with 10 CFR 50.55a(z)(1), NPPD requests approval of an alternative to the 5 year test interval requirement of the ASME GM Code, Appendix I, Section 1-1320(a) for the safety valves at CNS.5. Proposed Alternative and Basis for Use NPPD requests that the test interval be increased from 5 years to 6 years in accordance with Code Case OMN-17. All aspects of Code Case OMN-17 will be followed for the MS safety valves.As an alternative to the Code required 5-year test interval per Appendix I, paragraph I-1320(a), NPPD proposes that the subject Class 1 safety valves be tested at least once every three refueling cycles (approximately 6 years/72 months) with a minimum of 20% of the valves tested within any 24-month interval. This 20% would consist of valves that have not been tested during the current 72-month interval, if they exist. The test interval for any individual valve would not exceed 72 months except that a 6-month grace period is allowed to coincide with refueling outages to accommodate extended shutdown periods and certification of the valve prior to installation. This is all in accordance with OMN-1 7, paragraph (a).After as-found set-pressure testing, the valves shall be disassembled and inspected to verify that parts are free of defects resulting from time-related degradation or service induced wear. As-left Revision 0Pae9 Page 90 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RV-02 Main Steam Safety Valve Testing per Code Case OMN-17 (Continued) set-pressure testing shall be performed following maintenance and prior to returning thle valve to service. Each valve shall have been disassembled and inspected prior to the start of the 72-month interval. Disassembly and inspection performed prior to the implementation of Code Case OMN-17 may be used.Each refueling outage, CNS will remove one safety valve to be sent off-site to a test facility.Upon receipt at the off-site facility, the valves are subject to an as-found inspection, as-found seat leakage test, and as-found set pressure test in accordance with Appendix I of the ASME OM Code. Prior to the returning the valve to the plant for re-installation, the safety valve is disassembled and inspected to verify that internal surfaces and parts are free 'from defects or service induced wear. During this process, anomalies or damage are identified for resolution. Damaged or worn parts (i.e. springs, gaskets and seals) are replaced or repaired, as necessary. Following reassembly, the valve's set pressure is recertified. This existing process is in accordance with ASME OM Code Case OMN-1 7, paragraphs (d) and (e). Alternatively, CNS may elect to replace the removed valve with a spare valve that has previously already been through the process just described. Up to three spare valves may be used in accordance with paragraph (b) of OMN-17.NPPD has reviewed the as-found set point tcst results for all three safety valves tested since 1997 as detailed in Table 1. Since 1997, all as found lift tests have been within a +/-3% tolerance (maximum of +2.02%). The current Technical Specification requirements are that the as found test results fall within a +/-3% tolerance. Technical Specifications require the as left certification of the valves to meet a +/-1% tolerance. If an as found test is found to be outside of the +/-3 %tolerance, the other 2 safety valves will be removed and tested in accordance with Code Case OMN-17, paragraph (c).Accordingly, the proposed alternative of implementing all aspects of OMN-1 7, which will increase the test interval for the subject Class 1 safety valves from 5 years to 3 fuel cycles (approximately 6 years/72 months), will provide an acceptable level of quality and safety. This will also restore the operational and maintenance flexibility that was lost when the 24-month fuel cycle created the unintended consequences of more frequent testing. This proposed alternative will continue to provide assurance of the valves' operational readiness and provides an acceptable level of quality and safety pursuant to 10 CFR 50.55a(z)(1).

6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents A similar relief was previously approved at Peach Bottom for the fourth ten-year interval as Relief Request 01A-VRR-3 (TAC Nos. MF2509 and MF25 10, April 30, 2014).Revision 0Pae9 Page 91 Cooper Nuclear Station FJith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-02 Main Steam Safety Valve Testing per Code Case OMN-17 (Continued)

Monticello Nuclear Generating Plant Relief Request VR-04 was approved in a NRC Safety Evaluation Report dated September 26, 2012 (ML 12244A272). Quad Cites Nuclear Power Station, Units 1 and 2 Relief Request RV-05 was approved in a NRC Safety Evaluation Report dated February 14, 2013 (ML13042A348). Table 1: Cooper Nuclear Station Safety Valve Test History Safety Valve AF Test Date Set Pressure AsFudSt Dvainfo Pressure Set Pressure 4/9/1997 1240 1217 -1.85%10/9/1998 1240 1252 +0.97%MS-RV-70ARV 3/8/2003 1240 1226 -1.13%4/19/2008 1240 1232 -0.65%10/21/2012 1240 1255 +1.21%4/10/1997 1240126-.3 3/12/2000 1240 13 07 MS-RV-7OBRV 1/25/2005 1240121+.8 10/3/2009 1240 1260 +1.61%10/8/2014 1240 1253 +1.05%4/10/1997 1240 1262 + 1.77%11/12/2001 1240 1237 -0.24%MS-RV-70CRV 10/26/2006 1240 1265 +2.02%3/21/2011 1240 1262 +1.77%Revision 0Pae9 Page 92 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RV-03 Main Steam Safety Relief Valve Testing Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected Valve Class Category System MS-RV-71ARV 1 B/C MS MS-RV-71BRV 1 B/C MS MS-RV-71CRV 1 B/C MS MS-RV-7lDRV 1 B/C MS MS-RV-71ERV 1 B/C MS MS-RV-71FRV 1 B/C MS MS-RV-71GRV 1 B/C MS M S-RV-71HIRV 1 B/C M S 2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTC-5 240 -Safety and Relief Valves. Safety and relief valves shall meet the inservice test requirements of Mandatory Appendix I.ASME OM Code Mandatory Appendix I, "lnservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants," Section I-1320, "Test Frequencies, Class 1 Pressure Relief Valves," paragraph (a), "5-Year Test Interval," states that Class 1 pressure relief valves shall be tested at least once every 5 years.ASME OM Code Mandatory Appendix I, 1-3310 Class 1 Main Steam Pressure Relief Valves with Auxiliary Actuation Devices -Tests before maintenance or set-pressure adjustment, or both, shall be performed for 1-3310(a), (b) and (c) in sequence. The remaining shall be performed after maintenance or set-pressure adjustments:

a. visual examination; Revision 0Pae3 Page 93 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RV-t03 Main Steam Safety Relief Valve Testing (Continued)
b. seat tightness determination, if practicable;
c. set-pressure determination;
d. determination of electrical characteristics and pressure integrity of solenoid valve(s);e. determination of pressure integrity and stroke capability of air actuator;f. determination of operation and electrical characteristics of position indicators;
g. determination of operation and electrical characteristics of bellows arm switch;h. determination of actuating pressure of auxiliary actuating device sensing element, where applicable, and electrical continuity;
i. determination of compliance with the Owner's seat tightness criteria.4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(l), relief is requested from the requirements of ASME GM Code Appendix I, sections 1-1320(a) and 1-3310. The proposed alternative would provide an acceptable level of quality and safety.Section ISTC-5240, "Safety and Relief Valves," directs that safety and relief valves meet the inservice testing requirements set forth in Appendix I of the ASME OM Code.Appendix I, Section 1-1320(a), of the ASME GM Code states that Class 1 pressure relief valves shall be tested at least once every 5 years, starting with initial electric power generation.

This section also states a minimum of 20 percent of the pressure relief valves are tested within any 24-month interval and that the test interval for any individual valve shall not exceed 5 years.CNS has eight MS safety relief valves (SRV). The approach for the past several years has been to remove either 2 or 3 of the entire valves (i.e. main body and pilot assembly) every refueling outage and send them off for as found testing, refurbishment, rebuilding, and re-certification in preparation for the next time they are re-installed into the plant. Those 2 or 3 entire valves have been replaced with refurbished valves that were recertified just prior to the outage. The schedule is planned so that all eight entire valves get sent off, as found tested, refurbished, and re-certified within a three cycle frequency. In addition, CNS has replaced the remainder of the pilot assemblies (5 or 6 per outage) and sent them off for testing, refurbishment, and re-certification in preparation for the next time they are re-installed into the plant. These 5 or 6 additional pilot assemblies are replaced with refurbished and recertified pilot assemblies that were recertified just prior to the outage. Therefore, the pilot assemblies for the full complement of 8 valves have been set pressure tested every outage for several years.Revision 0Pae9 Page 94 Cooper Nuclear Station Ftifh Interval Inservice Testing Program for Pumps and Valves Relief Request RV-03 Main Steam Safety Relief Valve Testing (Continued) CNS plans to continue this approach into the fifth ten-year interval. However, refucling outage 27 (Fall/20 12) was the last refueling outage under an 18-month cycle. CNS is now operating with 24-month cycles. With this in mind, the refurbishment of the entire valves will eventually align with a six year frequency, which is consistent with Code Case OMN-1 7. However, all eight of the pilot assemblies are being removed, tested and replaced with refurbished/recertified spare pilot assemblies every refueling outage, which means a full complement of the set pressure portion of the valves are being tested every refueling outage. Therefore, although this approach is very conservative, documenting acceptability of this approach is being pursued per this relief request.Additionally, since 5-6 pilot assemblies, alone, are being replaced every outage (versus the entire valve), documenting acceptability of how portions of Appendix 1-3310 are being satisfied is also being pursued per this relief request.5. Proposed Alternative and Basis for Use These eight SRVs are considered Class 1 main steam pressure relief valves with auxiliary actuating devices. They are located on the main steam lines. In addition to their automatic function of opening to prevent over pressurization of the reactor vessel, six of these valves are associated with the Automatic Depressurization System and two are associated with the Low Low Set logic. The valves are two-stage Target Rock valves, each equipped with a main body, a pilot assembly for set pressure control, a solenoid valve, and an air operator assembly.CNS proposes to follow the Code Case OMN-1 7, paragraph (d), recommendations for Maintenance on these eight valves. Therefore, on a three cycle (up to 6 year) frequency, CNS proposes to remove the entire valve unit (i.e. main body and pilot assembly) for each one of these valves and ship it off for as found testing, refurbishment, and re-certification. CNS will replace these entire valve units with spare refurbished and re-certified entire valve units.As mentioned earlier, each valve is equipped with a pilot valve assembly that controls the set pressure. The remainder of the pilot valve assemblies (5 or 6 per refueling outage) will be removed from the main body and sent off site for examination, as found testing, refurbishment, and re-qualification testing (set point, reseat, and pilot stage seat tightness). The test facility has a main body slave for this purpose. The removed pilot valve assemblies are replaced with previously refurbished and re-qualified pilot valve assemblies. By testing all of the pilot valve assemblies every outage, the potential need to expand to test additional valves due to set pressure failures is alleviated and the future valve reliability is improved. Test results are being monitored by serial numbers. Any as found set pressure failure will be addressed via the CNS Corrective Action Program.ASME OM Code Interpretation, 98-8, clarifies that a pilot operated relief valve with an auxiliary actuating device is not required to be tested as a unit. Furthermore, it clarifies that set pressure determination on the pilot operator may be perforned after the pilot operator is removed from the valve body.Revision 0Pae9 Page 95 Cooper Nuclear Station Fith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-03 Main Steam Safety Relief Valve Testing (Continued) Appendix I, 1-3310(a) visual examination is completed at the test facility for those main bodies and pilot assemblies being sent there for examination, testing and refurbishment. With the removal of the pilot assemblies from the main bodies at the plant, the accessible portions of the main bodies will be examined in place without further disassembly as permitted by I-1310(c). Appendix I, 1-3310(b) seat tightness, and 1-3310(c) set pressure, is satisfied through as found seat leakage and set pressure testing at the offsite test facility for those main valves and pilot valve assemblies being sent there for inspection, testing and refurbishment. Paragraph 1-3310(i) is satisfied through as left seat leakage testing at the facility. Seat leakage of installed main valves is continuously monitored and also satisfies 1-3310(i). Pressure switches in the SRV discharge lines annunciate in the control room and indicate when the main valve seat is open. In addition, there are temperature elements on the valve discharge lines which provide leakage indication. During startup, the main valve and Auxiliary Actuation Devices are verified to function properly by being full stroke exercised open and closed. Successfully exercising these valves open and closed verifies the electrical characteristics and pressure integrity of the solenoid valve and air actuator (satisfying Appendix I, paragraphs (d) and (e)). During this exercise, Appendix I, paragraph 1-3310(f), is also satisfied through the use of the valve indicating lights, discharge pressure switches, and temperature elements.Finally, Appendix I, paragraphs 1-3310(g) and 1-3310(h), are not applicable to the CNS MS safety relief valves.This proposed alternative is conservative in nature and will continue to provide an acceptable level of quality and safety pursuant to 10 CFR 50.55a(z)(1).

6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents A version of this relief request was previously approved for the fourth ten-year interval at CNS as Relief Request RV-04 (TAC Nos. MC8 837, MC8975, MC8976, MC8977, MC8978, MC8979, MC8980, MC8981, MC8989, MC8990, MC8991, and MC8992, June 14, 2006).Revision 0 Page 96 Revision 0 Page96 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves Relief Request RV-04 Control Rod Drive (CRD) Technical Specification Testing Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

Alternative Provides Acceptable Level of Quality and Safety 1. ASME Code Component(s) Affected Valve Class Category System CRD-SOV-SO120* 2 B CRD CRD-SOV-SO121* 2 B CRD CRD-SOV-SO122* 2 B CRD CRD-SOV-SO123* 2 B CRD CRD-AOV-CV126* 2 B CRD CRD-AOV-CV127* 2 B CRD CRD-CV-114CV* 2 C CRD CRD-CV-138CV* 2 C CRD SOV=Solenoid Operated Valve AOV=Air Operated Valve CV==Check Valve*Typical of 137 Hydraulic Control Units (HCU)2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ASME OM Code ISTC-3500 Valve Testing Requirements -Active and passive valves in the categories defined in ISTC-1 300 shall be tested in accordance with the paragraphs specified in Table ISTC-3500-1 and the applicable requirements of ISTC-5 100 and ISTC-5200. ISTC-35 10 Exercising Test Frequency -Active Category A, Category B, and Category C check valves shall be exercised nominally every three (3) months, except as provided by ISTC-3520, ISTC-3540, ISTC-3550, ISTC-3570, ISTC-5221, and ISTC-5222. ISTC-3560 Fail-Safe Valves -Valves with fail-safe actuators shall be tested by observing the operation of the actuator upon loss of valve actuating power in accordance with the exercising frequency of ISTC-35 10.ISTC-5 131 (a) Valve Stroke Testing -Active valves shall have their stroke times measured when exercised in accordance with ISTC-3500. ISTC-5 151 (a) Valve Stroke Testing -Active valves shall have their stroke times measured when exercised in accordance with ISTC-3500. Revision 0 Page 97 Revision 0 Page 97 Cooper Nuclear Station Fdith Interval Inservice Testing Program for Pumps and Valves Relief Request RV-04 Control Rod Drive (CRD) Technical Specification Testing (continued) ISTC-5221 (a) Valve Obturator Movement -The necessary valve obturator movement during exercise testing shall be demonstrated by performing both an open and a close test.4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(l), relief is requested from the requirements of ASME GM Code ISTC-3500, ISTC-3510, ISTC-3560, ISTC-5131 (a), ISTC-5151 (a), and ISTC-5221 (a). The proposed alternative would provide an acceptable level of quality and safety.This relief is needed to make the fifth ten-year inservice test program consistent with NTJREG 1482, Revision 2.5. Proposed Alternative and Basis for Use Background Information It is typical for Boiling Water Reactors (BWR) to perform the subject CRD) testing per their respective plant Technical Specifications. This originated from Generic Letter (GL) 89-04, Position 7. Per section 1.3 of NUIREG 1482, Revision 2, specific relief is required to implement the guidance derived from GL 89-04, which is why this testing is being documented under a relief request. The proposed alternatives and the basis for use are discussed in further detail below.CRD-CV-13 8CV: CRD-SOV-SO1 20, SO121. SO 122. SO 123: The CRD cooling water header check valve, CRD-CV-138CV (typical of 137 HCUs), has a safety function to close in the event of a scram to prevent diversion of pressurized HCU accumulator water to the cooling water header. The exhaust water withdrawal/settle (CRD-SOV-S0120), exhaust water insert (CRD-SOV-SO121), drive water withdrawal (CRD-SOV-SO122), and drive water insert (CRD-SO V-SO 123) solenoid valves (typical of 137), have a safety function to close in order to provide a boundary to non-code class piping.Normal control rod motion will verify that the associated cooling water check valve has moved to its safety function position of closed. Industry experience has shown that rod motion may not occur if this check valve were to fail in the open position.The solenoid valves listed above have a safety function to close in order to provide a class 2 to non-code class boundary isolation. During normal operation, these solenoid valves are used for control rod insertion and withdrawal. They are exercised open and closed during normal operation of the associated CRD. They are not equipped with position indication or control switches. They automatically change position to affect control rod movement.Therefore, control rod exercising in accordance with the CNS Technical Specifications, Surveillance Requirement (SR) 3.1.3.3, will provide an acceptable level of quality and safety for these valves. This testing method is consistent with GL 89-04, Position 7, and NUREG 1482, Revision 2, Section 4.4.6.Revision 0 Page 98 Cooper Nuclear Station F~ith Interval Inservice Testing Program for Pumps and Valves Relief Request RV-04 Control Rod Drive (CRD) Technical Specification Testing (continued) CRD-AOV-CV126, CRD-AOV-CV127. and CRD-CV-1 14CV: These valves operate as an integral part of their respective HCU to rapidly insert the control rods in support of a scram. The CRD scram inlet valve, CRD-AOV-CV 126 (typical of 137), opens with a scram signal to pressurize the lower side of the Control Rod Drive Mechanism (CRDM)pistons from the accumulator or from the charging water header. The CRD outlet isolation valve, CRD-AOV-CV 127 (typical of 137), opens with scram signal to vent the top of the CRDM piston to the scram discharge header. The CRD scram outlet check valve, CRD-CV-1 14CV (typical of 137), opens to allow flow from the top of the CRDM piston to the scram discharge header.Individual stroke time measurements of air-operated valves CRD-AOV-CV1 26 and CRD-AOV-CV127 are impractical due to their rapid acting operation and they are not equipped with position indication. '[herefore, valve stroke times will not be measured. Additionally, the air-operated valves fail-open on a loss of air or power. Normal opening removes power to the pilot solenoid valve, simulating a loss of power. On loss of power, the solenoid vents the air operator and CRD-AOV-CV126 and CRD-AOV-CV 127 are spring-driven open. Thus, each time a scram signal is given, the valves "experience" a loss of air/power to verity' each valve's fail-safc open feature.Testing these valves simultaneously would result in a full reactor scram. An excess number of scrams performed routinely could cause thermal and reactivity transients, which could lead to fuel, vessel, CRD, or piping damage. The CRDs cannot be tested during cold shutdown because the control rods are inserted and must remain inserted.Therefore, control rod scram time testing in accordance with the CNS Technical Specifications, SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4, will provide an acceptable level of quality and safety for these valves. This testing method for these valves is consistent with GL 89-04, Position 7, and NUJREG 1482, Revision 2, Section 4.4.6.'6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was previously approved for the fourth ten-year interval at CNS as relief request RV-06 (TAC No. ME1 521, April 26, 2010). A similar alternative was approved at Perry-1 for relief request VR-1, revision 1 (TAC No. ME7380, February 22, 2012).Revision 0 Page 99 Revision 0 Page 99 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-O5 Performance-Based Scheduling of Pressure Isolation Valve Leakage Tests Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Alternative Provides Acceptable Level of Quality and Safety 1' ASME Code Component(s) Affected Valve Class Category System RHR-MOV-MO25A 1 A RHR RHR-MOV-MO25B 1 A RHR RHR-MOV-MO274A 1 A RHR RHIR-MOV-MO274B 1 A RHR RHR-CV-26CV 1 A/C RHIR RHR-CV-27CV 1 A/C RHR RHR-MOV-MO017 1 A RHR RHR-MOV-MO018 1 A RHR CS-MOV-MO012A 1 A CS CS-MO V-MO 12B 1 A CS CS-C V-18CV 1 A/C CS CS-C V-19CV 1 A/C CS MOV=Motor Operated Valve 2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTC-3630 -Leakage Rate for Other Than Containment Isolation Valves.ISTC-3630(a) -Frequency. Tests shall be conducted at least once every two years.4. Reason for Request Pursuant to 10 CER 50.55a, "Codes and standards," paragraph (z)(1), relief is requested from the requirement of ASME OM Code ISTC-3630(a). ISTC-3630(a) requires that leakage rate testing (water) for pressure isolation valves (PIV) be performed at least once every two years. Data from RE25 and RE26 was used to identify that PIV testing alone each refueling outage incurs a total dose of at least 600 mRem. The reason for this relief request is to reduce outage dose. The basis of this relief request is that the proposed alternative would provide an acceptable level of quality and safety.5. Proposed Alternative and Basis for Use The RHR and CS systems at CNS contain valves that function as PIVs. PI~s are defined as two normally closed valves in series at the reactor coolant system boundary that isolate the reactor Revision 0 Pg 0 Page 100 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RV-05 Performance-Based Scheduling of Pressure Isolation Valve Leakage Tests (continued) coolant system from an attached low pressure system. These affected valves, listed in Section 1, are located on the 'A' and 'B' CS and RHR injection lines and the RHR shutdown cooling line.PI~s are not specifically included in the scope for performance-based testing as provided for in 10 CFR 50 Appendix J, Option B. The concept behind the Option B alternative for containment isolation valves is that licensees should be allowed to adopt cost effective methods for complying with regulatory requirements. Additionally, NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," describes the risk-informed basis for the extended test intervals under Option B. That justification shows that for valves which have demonstrated good performance by passing their leak rate tests (air) for two consecutive cycles, further failures appear to be governed by the random failure rate of the component. NEI 94-01 also presents the results of a comprehensive risk analysis, including the statement that "the risk impact associated with increasing [leakrate] test intervals is negligible (less than 0.1 percent of total risk)." The valves identified in this relief request are in water applications. The PIV testing is performed with water pressurized to normal plant operating pressures. This relief request is intended to provide for a performance-based scheduling of PIV tests at CNS.As stated in the previous section, the reason for requesting this relief is dose reduction. Data reviewed from RE25 and RE26 identified that PIV testing alone incurred a total dose of approximately 600 mrem in RE26, which benefited from the chemical decontamination that was performed, and approximately 1600 mrem in RE25. Therefore, assuming the PIVs remain classified as good performers, extended test intervals of three refueling outages would provide a savings of at least 1200 mrem over a three-cycle period.NUREG 0933, "Resolution of Generic Safety Issues," Issue 105, discusses the need for PIV leak rate testing based primarily on three pre-1980 historical failures of applicable valves industry-wide. These failures involved human errors in either operations or maintenance. None of these failures involved inservice equipment degradation. The performance of PlY leak rate testing provides assurance of acceptable seat leakage with the valve in a closed condition. Typical PIV testing does not identify functional problems which may inhibit the valves ability to re-position from open to closed. For check valves, such functional testing is accomplished per ASME OM Code ISTC-3522 and ISTC-3520. Power-operated valves are routinely full stroke tested per ASME OM Code to ensure their functional capabilities. The periodic functional testing of the PI~s is adequate to identify abnormal conditions that might affect closure capability. Performance of the separate 24-month PIV leak rate testing does not contribute any additional assurance of functional capability; it only determines the seat tightness of the closed valves.Revision 0 Pg 0 Page 101 Cooper Nuclear Station F~fhz Inteival Inserviee Testing Program for Pumps and Valves Relief Request RV-05 Performance-Based Scheduling of Pressure Isolation Valve Leakage Tests (continued) The functional test and position indication test (PIT) frequencies are as follows: Valve Functional Test PIT RITR-MOV-MO25A Quarterly 2 years RHR-MOV-MO25B Quarterly 2 years RITR-MOV-M0274A Normally De-energized Closed Refueling Outage (exercised during PIT test)RIIR-MOV-MO274B Normally De-energized Closed Refueling Outage (exercised during PIT test)RIIR-CV-26CV Refueling Outage Refueling Outage RHR-CV-27CV Refueling Outage Refueling Outage RETR-MOV-MO017 Cold S/D Refueling Outage RHR-MO V-MO 18 Cold SiD Refueling Outage CS-MO V-MO 12A Cold S/D Refueling Outage CS-MOV-MO12B Cold S/D Refueling Outage CS-C V-i18CV Refueling Outage Refueling Outage CS-C V-19CV Refueling Outage Refueling Outage CNS proposes to perform PlV testing at intervals ranging from every refueling outage to every third refueling outage. The specific interval for each valve would be a function of its performance and would be established in a manner consistent with the containment isolation valve (CIV) process under 10 CFR 50 Appendix J, Option B. Five of the 12 valves listed in Section 1 (RHR-MOV-MO25A, RHIR-MOV-M025B, CS-MOV-MO12A, CS-MOV-MO12B, RHR-MO V-MO 17) are also classified as CIVs and are leak rate tested with air at intervals determined by 10 CFR 50 Appendix J, Option B. Appendix J and inservice leak testing program guidance will be established such that if any of those five valves fail either their as found CIV test or their PIV test, the test interval for both tests will be reduced to every refueling outage until they can be re-classified as good performers per Appendix J, Option B requirements. The test intervals for the seven remaining valves with a PIV-only function will be determined in the same manner as is done under Option B. That is, the test interval may be extended to every three refueling outages (not to exceed a nominal six year period) upon completion of two consecutive, periodic PIV tests with results within prescribed acceptance criteria. Any test failure will require a return to the initial interval (every refueling outage) until good performance can again be established. The primary basis for this relief request is the historically good performance of the PI~s. There have been no PIV seat leakage failures since PlV testing began at CNS in 1995 through the present. Leakages recorded have been a very small percentage of the overall allowed leakage.Revision 0 Pg 0 Page 102 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RV-05 Performance-Based Scheduling of Pressure Isolation Valve Leakage Tests (continued) The test results for the PI~s listed in Section 1 have been exceptional. For example, a plot of the RHR-MO V-MO 17 test results is shown below: This graph is typical of the affected PI~s listed in Section 1; however, there have been cases where the CIV air testing has indicated a failure with components identified in this relief request.There is a general industry-wide consensus that CIV air testing is a more challenging and accurate measurement of seat condition, and more likely to identify any seat condition degradation. PIV testing has also been utilized at CNS as a post-maintenance test following packing replacements on the CS and RHR injection check valves to ensure the packing is adjusted adequately at normal system pressure. Therefore, PIV testing will continue to be utilized as post-maintenance testing, as necessary. On June 8, 2012, the NRC staff reviewed and endorsed NEI 94-01, Revision 3 (see the safety evaluation at Accession No. ML12 1030286), which allows for up to a 75-month frequency for"Type C tests." Per the NRC safety evaluation report (SER) for the fourth interval IST Program (TAC No. ME7021, dated August 28, 2012), to obtain a frequency extension beyond 60 months (up to 75 months), licensees should provide additional information, such as maintenance history, acceptance tests criteria, condition monitoring programs, etc., to justify the acceptability of the extension. In order to further justify the proposed maximum frequency of 3 cycles (72 months)with a standard grace period of 6 months, additional information is being provided.Table 1 of this relief request contains the maintenance history and Local Leak Rate Test (LLRT)/PIV test history for all 12 of these pressure isolation valves for the past 10 years (since 01/01/2005). The table includes the as found and as left LLRT and PIV test results with the Revision 0 Page 103 Revision 0 Page 103 Cooper Nuclear Station Ffifh Interval lnservice Testing Pro gram for Pumps and Valves Relief Request RV-05 Performance-Based Scheduling of Pressure Isolation Valve Leakage Tests (continued) associated operability limits. Note that corrective and preventative maintenance has been performed over the past 10 years (and beyond) in order to maintain the acceptable performance of the components. For instance, the MOV Program requires regular inspections and diagnostic tests of the motor operators to ensure that they continue to be relied upon throughout the life of the plant and the check valves have preventative maintenance plans to replace the valve packing on a periodic basis to ensure the packing material is properly maintained. Note that not all of the maintenance performed impacts the seating ability of the components or the test boundary of the associated LLRT/PIV tests, so pre- or post-LLRT/P1V testing may not have been required to be performed. Exercise testing, stroke time testing, and position indication testing was not listed in Table 1.As can be observed from Table 1, the As Found LLRT test results have been excellent with no failures associated with these valves over the past 10 years and a significant amount of margin has been maintained to the administrative component operability limit. Even more so, a very large margin exists between the PIV test results and the operability limit for each PTV test. With a limit of 5 gpm, the highest recorded PIV leakage in the last 10 years was 0.43 5 gpm, which is only 8.7% of thc allowcd leakage. Historically, since 1995, all of the PIV valves have maintained this much or more of a margin to the 5 gpm acceptance criteria as shown below.Test # Components Maximum PIV Percent of Percent of leakage recorded allowed leakage margin to 5 gpm since 1995 (gpm) limit 1 RHiR-MOV-MO25A 0.299 5.98% 94.02%2 RHIR-MOV-MO25B 0.272 5.44% 94.56%3 RHR-CV-26CV /RH-O-07A0.1224 2.45% 97.55%4 RHR-CV-27CV /RH-O-07B0.326 6.52% 93.48%5 RHiR-MOV-MO017 0.0272 0.54% 99.46%6 RI-R-MOV-M0 18 0.218 4.36% 95.64%7 CS-MO V-MO 12A 0.435 8.70% 91.30%8 C S-MOV-MO 12B 0.082 1.64% 98.36%9 CS-CV-18CV 0.3264 6.53% 93.47%10 CS-CV-19CV 0.082 1.64% 98.36%The NRC SER for NEI TR 94-01, Revision 3, resulted in a condition that the licensee report the margin between the Type B and Type C leakage rate summation and its regulatory limit and maintain an acceptable margin to the regulatory limit. A second condition requires the licensee to include considerable extra margin in order to extend the LLRT intervals beyond 5 years to a 75-month interval. In comparison, for these PIV tests, CNS will establish an administrative limit of<1 gpm for each of the PlY tests in order to maintain each test on an extended frequency. This administrative limit is only 20% of the allowed leakage and will provide considerable extra margin to the limit of 5 gpm when looking at the historical test results.Revision 0 Pg 0 Page 104 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-05 Performance-Based Scheduling of Pressure Isolation Valve Leakage Tests (continued) NUREG/CR-5928, "ISLOCA Research Program Final Report," evaluated the likelihood and potential severity of inter-system loss-of-coolant accident (ISLO CA) events in BWR and pressurized water reactors. The BWR design used as a reference for this analysis was a BWR/4 with a Mark 1 containment. CNS was listed in Section 4.1 of NUREG/CR-5928 as one of the applicable plants. The applicable BWR systems were individually analyzed and in each case, this report concluded that the system was "...judged to not be a concern with respect to ISLOCA risk." Section 4.3 concluded the BWR portion of the analysis by saying "ISLOCA is not a risk concern for the BWR plant examined here." Summary of bases / rationale for this relief request:* Performance-based PIV testing would yield a dose reduction of up to 1200 mrem over a three-cycle period.* Performance of separate functional testing of PIVs per ASME Code.a Excellent historical performance results from PlV testing for the applicable valves.* Low likelihood of valve mispositioning during power operations (procedures, interlocks).

  • Air testing versus water testing -degrading seat conditions are identified much sooner with air testing.* Relief valves in the low pressure piping -these relief valves may not provide ISLOCA mitigation for inadvertent PIV mispositioning (gross leakage), but their relief capacity can easily accommodate conservative PIV seat leakage rates.* Alarms that identify high pressure to low pressure leakage -Operators are highly trained to recognize symptoms of a present or incipient ISLOCA and to take appropriate actions.The intent of this relief request is simply to allow for a performance-based approach to the scheduling of PTV leakage testing. It has been shown that ISLOCA represents a small risk impact to BWRs such as CNS. CNS PIVs have an excellent performance history in terms of seat leakage testing. The risks associated with extending the leakage test interval to a maximum of three refueling outages (nominal 24 months) are extremely low. The performance-based interval shall not exceed 72 months. Standard scheduling practice may extend the program interval by 25%, not to exceed six months. This relief will provide significant reductions in radiation dose.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents A version of this relief request was previously approved for the fourth ten-year interval at CNS as relief request RV-07 (TAC No ME7021, dated 8-28-20 12). Fermi 2 received a Safety Evaluation by the NRC, dated September 28, 2010, on a similar relief request for the performance-based testing of PI~s (TAC No. ME2558, ME2557, and ME2556).Revision 0 Page 105 Revision 0 page lO5 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005

_____Component(s) [ Date hOutage jWork Order Work Order Description f AF Tests [ AL Tests Comments RIHR-MOV-MO25A (Test #1)Sprnn/2005 RE22 N/A N/A LLRT: 2.02 scfhi (< 30 scfh)PIV: 0.299 gpm (< 5 nom)N/A LLRT and PIV test due.10/02/2005 Online CM 4464719 Remove insulation and N/A N/A No impact on validate leak; tightened LLRT or PIV cap screws on pressure testing.seal; slowed leakage. _______02/10/2006 Online CM 4465302 Efforts were made to stop N/A N/A No impact on bonnet seal leak. LLRT or PIV testing.05/30/2006 Online PM 4390882 Clean & Lubricate Stem N/A N/A No impact on LLRT or PIV testing.Fall/2006 RE23 CM 4464912 Repaired pressure seal LLRT: LLRT (Final AL): Major CM 4534360 leak, refurbed motor 0.83 scfha (< 30 scflh) 7.95 scfh ( < 30 maintenance operator, disassembled scfh)) resets LLRT and examined valve, and PIV: Freq. to every diagnostically tested. 0.08 gpm (AL) refueling outage.04/03/2007 Online PM 4542913 Perform Motor Pinion N/A N/A No impact on Inspection LLRT or PIV testing.10/04/2007 Online PM 4498618 Examine MO-Mech N/A N/A No impact on PM 4498668 Examine MO-Elect LLRT or PIV testing.Spring/2008 RiE24 N/A N/A LLRT: N/A 1 st periodic 1.25 scfh (< 50 scflh) test for LLRT PIV: (and PIV)0.109 gpm (< 5 test.gpm)_________________ 03/30/2009 Online PM 4625205 PM 4625262 PM 4625267 Clean & Lubricate Stem Examine MO-Mech Examine MO-Elec N/A N/A No impact on LLRT or PIV testing.Revision 0 Pg 0 Page 106 Cooper Nuclear Station Fifth Interval lnservice Testing Program for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 Component(s) '-Date Outage Work Order Work Order Description AF Tests AL Tests C~mments FalL'2009 RE25 CM 4641890 MOV Program Diagnostic LLRT: PIV: No impact to test/motor replacement. 1.75 scfha (< 50 scfha) 0.109 gpm (< 5 gpm) LLRT or PIV No AL LLRT required testing. 2nd due to minimal seat thrust periodic test change, for LLRT (and PIV)test.Spring/2011I RE26 N/A N/A LLRT: N/A 3rd (extra)2.1 scfha (< 50 scfha) periodic test PIV: for LLRT 0.136 gpm (< 5 (and PIV)gpm) test.06/05/2012 Online PM 4802964 Clean and Lubricate Stem N/A N/A *No impact on PM 4803040 Examine MO-Mech LLRT or PIV PM 4803052 Examine MO-Elec testing.Fall/2012 RE27 N/A N/A N/A N/A No tests due to Option B /approved PIV relief request.Fall/2014 RE28 N/A N/A LLRT: N/A No PIV test 3.82 scfha (< 50 scfha) due to approved PIV_____________relief request.RHR-MOV-M025B Spring/2005 RE22 CM 4335229 Votes diagnostic test LLRT: LLRT: MOV (Test #2) 24 scfli (< 30 scfth) 23.8 scflh (< 30 sofia) periodic test.PIV: 0.0544 gpm (< 5 gpm)04/11/2005 Online PM 4381354 Clean and Lubricate Stem N/A N/A No impact on LLRT or PIV testing.10/17/2006 Online CM 4531030 Examine Torque Switch N/A N/A No impact on LLRT or PIV testing.10/18/2006 Online CM 4531090 Replace Motor Pinion N/A N/A No impact on Gear LLRT or PIV testing.Revision 0 Pg 0 Page 107 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 _____Component(s) [ Date [Outage ]Work Order ]Work Order Description AF Tests AL Tests [Comments Fall/2 006 RE23 CM 4537229: Packing Leak; No AL LLRT required due to minimal packing/seat load change.LLRT: 17.5 scfl ( 50 scf)PIV: 0 gpm (< 5 gpm)No impact on LLRT or PIV testing.Monitoring LLRT;assume 1st periodic test for PIV.04/18/2007 Online PM 4485530 Examine MO-Elec N/A N/A No impact on PM 4498698 Examine MO LLRT or PIV testing.Spring/2008 RE24 CM 4632406 Repack valve LLRT: N/A No impact on CM 4631163 Adjust Packing/Viper Test 32.14 scfih (< 50 LLRT or PIV CM 4531210 Refurbed AO; scfh) testing.No AL LLRT/PIV PIV: Monitoring required due to minimal 0.0544 gpm (< 5 LLRT; 2nd packing/seat load change. gpm) periodic test__________for PIV.10/14/2008 N/A PM 4600595 Clean and Lubricate Stem N/A N/A No impact on LLRT or PIV testing.Fall/2009 RE25 N/A N/A LLRT: N/A Monitoring 12.74 sceti (< 50 LLRT; 3rd scfh) periodic test PIV: for PIV.0.054 gpm (< 5gpm)07/13/2010 Online PM 4664227 Examine MO-Elec N/A N/A No impact on PM 4664250 Examine MO-Mech LLRT or PIV____________testing. Spring/201 1 RE26 N/A N/A LLRT: N/A Monitoring 23.16 scifh (< 50 LLRT (1st scflh) periodic test);PIV: 4th periodic 0.136 gpm (< 5 test for PIV.gpm) _ _ _07/19/2011 Online PM 4749837. Clean and Lubricate Stem N/A N/A No impact on LLRT or PIV testing.Revision 0 Pg 0 Page 108 Cooper Nuclear Station Ffifh Interv'al Inservice Testing Pro gram for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 Component(s) '-Date Outage Work Order Work Order Description AF Tests AL Tests Comments Fall/2012 RE27 CM 4842207 Viper diagnostic test; No LLRT: N/A No impact on AL LLRT/PIV required 0.17 scfha (< 50 sctfh) LLRT or PIV due to minimal seat load testing.change. Mc~nitorinlg LLRT (2nd periodic test);No PIV test due to approved PIV relief request.01/14/20 13 Online PM 4864090 Examine MO-Mech N/A N/A No impact on PM 4864089 Examine MO-Elec LLRT or PIV__________ ________________ ________________ testing.01/16/2014 Online CM 4996016 Reterminate Motor Wiring N/A N/A No impact on LLRT or PIV testing.Fall/2014 RE28 N/A N/A N/A N/A No tests due to Option B /approved PIV relief request.01/12/2015 Online PM 4953672 Clean and Lubricate stem N/A N/A No impact on LLRT or PIV_______________________testing. RHIR-MOV-Spring/2005 RE22 RiHR-MO- Examine Motor Operator LLRT: N/A No impact on MO274A MO274A 7.5 sctlh (< 35 scfh) LLRT or PIV& PM 4363586 PIV: testing.RHR-CV-26CV 0.1224 gpm (< 5 (Test #3) ______gpm) Fall/2006 RE23 RHR-MOV- Evaluate Packing -Adjust LLRT: LLRT: Routine PM MO274A: or Repack -Repacked 0.75 scfh (< 35 scfh) 4.7 scth (< 35 scflh)PM 4446728 valve PIV: 0.041 gpm (< 5 gpm)RHR-MO- Examine Motor Operator MO274A______________________ ________PM 4446878 ____________________ Revision 0 Page 109 Revision 0 Page 109 Cooper Nuclear Station Ffith Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 Component(s) ~ -Date Outage Work Order Work Order Description -AF Tests AL Tests Comments Spring/2008 RE24 N/A N/A -LLRT: N/A Assume 1 st 4.27 scfha (< 35 scfh) periodic test PIV: for LLRT 0.054 gpm (< 5 (and PIV)______gpm) test.Fall/2009 RE25 RHR-MO- Examine MO-Mech LLRT: N/A No impact on MO274A 8.31 scfha(< 35 scth) LLRT or PIV PM 4645290 PIV: testing. 2nd 0.082 gpm (< 5 periodic test RHR-CV- Adjust Reed Switches gpm) for LLRT 26CV (and PIV)CM 4723494 test.Spring/201 1 RE26 RHR-MOV- Evaluate Packing -Adjust _N/A PIV: No impact on MO274A: or Repack -Tightened 0.082 gpm (< 5 gpm) PIV test.PM 4744619: Packing LLRT no longer required due to closed ioop analysis. 3rd periodic test_______________________________________ ____________________________________________or__PforPIVsest Revision 0 Page 110 Revision 0 Page 110 Cooper Nuclear Station Fifth interval Inservice Testing Pro gram for Pumps and V'alves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 Component(s) -Date Outage Work Order Work Order Description AF Tests AL Tests Comments Fall/2012 RE27 RHR-CV- Repack Valve; performed N/A N/A No PIV test 26CV: exercise test and no due to PM 4848060 extemnal leakage at approved PIV pressure as PMT; PlV test relief request.not required as had no impact to seating ability and located inside Drywell.RIIR-CV- Adjust Limit Switch 26CV: CM 4918074 RHR-MO- Examine MO-Mech MO274A: PM 4848151 __________ Fai1/2014 RE28 N/A N/A N/A N/A No PIV test due to approved PIV___________________relief request.RIIR-MOV-Spring/2005 RE22 RiHR-MO- Refurbish MO LLRT: LLRT: AF ~AL MO274B MO274B Examine MO 9.6 scflh (< 35 scflh) 9.4 scflh (< 35 scfha) LLRT.& CM 4299766 PIV: RHR-CV-27CV PM 4363585 0.136 gpm (< 5 gpm)(Test #4) Fall/2006 RF23 RHR-MOV- Evaluate Packing -Adjust LLRT: PIV: No impact on MO274B or Repack -No packing 9.8 scfha (< 35 scflh) 0.109 gpm (< 5 gpm) LLRT or PIV PM 4446729 adjustment required, testing.Examine MO PM 4446875 Spring/2008 RE24 RHR-CV- Repacked Valve LLRT: LLRT: Assume 27CV: 14.5 scfha (< 35 scfih) 28.59 scfha (< 35 scfh) resets LLRT PM 4541360 PIV: frequency. _____________ ________ _______ ________ ________________ _____________0.163 gpm (< 5 gpm) ______Revision 0 Page 111 Revision 0 Page 111 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 Component(s) '-Date Outage Work Order Work Order Description AF Tests AL Tests " Comments Fall/2009 RE25 RHR-MO- Examine MO-Mech LLRT: N/A No impact on 274B: 15.7 scfha (< 35 scth) LLRT or PIV PM 4645289 PIV: tests.0.218 gpm ( <5 Assume 1st gpm) periodic test for LLRT and PIV test.Spring/201 1 RE26 RHR-MOV- Evaluate Packing -Adjust N/A PIV: LLRT no MO274B: or Repack -Tightened 0.326 gpm ( < 5 gpm) longer PM 4744620 packing required due to closed loop analysis. No impact on PLV test; 2nd periodic PIV_____________test. Fall/2012 RE27 PM 4848150 Examine MO-Mech N/A N/A No PIV test due to approved PIV relief request.Fall/2014 RE28 N/A N/A N/A N/A No PIV test due to approved PIV relief request.RHIR-MOV-M017 Spring/2005 RE22 PM 4363507 Examine MO and Verify LLRT: No impact on (Test #5) Indication 2.95 scfha (< 30 scfha) LLRT or PIV PIV: testing.PM 4363526 Examine MO Ci.027 gpm (< 5 Assume 1st gpm) periodic test for LLRT and________PIV. Fall/2006 RE23 PM 4446718 Clean and Lube Stem LLRT: No impact on CM 4535994 Perform Motor Pinion 1.75 scfh (< 30 scflh) LLRT or PIV Inspection PIV: testing. 2nd 0 gpm (< 5 gpm) periodic test for LLRT and____________________________________________PIV. Revision 0 Page ll2 Cooper Nuclear Station Ffifh Inter'al Inservice Testing Program for Pumps and VZ~lves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 Component(s) -'Date Outage Work Order Work Order Description AF Tests AL Tests Comments Spring/2008 RE24 CM 4546759 Replace MO and LLRT: PIV: No impact on diagnostic test; AL LLRT 0.68 scfha (< 30 scflh) 0 gpm (< 5 gpm) LLRT or PIV not required due to testing. 3rd minimal change in seat periodic test load, for LLRT and PlV.Fall/2009 RE25 PM 4645142 Clean and Lubricate PIV: LLRT not 0' gpm (< 5 gpm) required due to Option B;4th periodic PIV test.Spring/201 1 RE26 PM 4744696 Examine MO -Mech PIV: LLRT not CM 4740307 Motor Pinion Inspection C.027 gpm (< 5 required due PM 4744691 Examine MO -Elect gpm) to Option B;5th periodic PIV test.Fall/2012 RE27 PM 4848600 Examine MO LLRT: No PIV test 5.32 scfih (< 30 scfha) due to approved PIV relief request.Fa11/2014 RE28 PM 4983676 Examine MO N/A No tests due to Option B /PIV relief reqtuest.RHR-MOV-MO18 Spring/2005 RE22 PM 4363506 Examine MO and Verify LLRT: LLRT: Assume (Test #6) Indication 1.96 scflh (< 30 scflh) 2.06 scfha (< 30 scfih) resets LLRT PIV: frequency. CM 4212544 Refurb MO and diagnostic 0.0408 gpm (< 5 test gpm)PM 4363568 Examine MO Revision 0 Page 113 Revision 0 Page 113 Cooper Nuclear Station Ffifh Interv'a[Inservice Testing Pro gram for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 _____Component(s) Outage Work Order Work Order Description AF Tests AL Tests Comments Fall/2006 RE23 PM 4446727 Evaluate Packing Adjust or Repack -(tightened 2 flats); AL LLRT not required due to minimal change in packing and seating forces.Examine Motor Operator Examine MO and Verify Indication LLRT: 2.6scth(<30scfh) PIV: 0.109 gpm (<5 5gpm)No impact on LLRT or PIV testing. 1st periodic test for LLRT and PIV.PM 4446867 PM 4446860 Spring/2008 RE24 PM 4549525 Examine Motor Operator LLRT: No impact on 0.7 scfha (< 30 scflh) LLRT or PIV CM 4531750 Motor Pinion Gear PIV: testing. 2nid Inspection '. 109 gpm (< 5 periodic test gpm) for LLRT and PlV.Fall/2009 RE25 CM 4640553 Motor Pinion Gear ILLRT: PIV: LLRT no Inspection N'/A 0.218 gpm (< 5 gpm) longer required due to closed loop analysis. No impact on PIV test. 3rd periodic PIV test.Spring/201 1 RE26 PM 4744618 Evaluate Packing Adjust LLRT: PIV: No impact on or Repack -(1 flat): N/A 0.027 gpm (< 5 gpm) PIV test. 4th periodic PIV PM 4746148 Examine MO-Mech test._________ ________PM 4744690 Examine MO-Elec___________________ ______Pall/LULL KILL/I'M 48480U1 tExammne MU-Mechi SN!A N/A No impact on PIV test. No PIV test due to approved PIV relief request..1 _____________ 1 ____________ C ______________ C ________________________________________________ i ______________________ .1 Revision 0 Page 114 Revision 0 Page 114 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRTr Test History' Since 01/01/2005 Component(s) -Date Outage Work Order Work Order Description AF Tests AL Tests Comments Fali12014 RE28 CM 4945389 Viper Test -no torque N/A N/A No PIV test switch or packing due to adjustments required approved PIV (AF=AL); PIV not relief request.required PM 4949440 Evaluate Packing Adjust or Repack -Not needed PM 4950106 Examine MO (Mech/Elec) CS-MOV-MO12A Spring/2005 RE22 N/A N/A LLRT: N/A Periodic (Test #7) 0.004 scfha (< 10 LLRT andPIV test.PIV: Assume 1st (1.299 gpm (< 5 periodic PIV________gpm) test.08/02/2005 Online PM 4387217 Clean, Lubricate, Partial N/A N/A No impact on Stroke LLRT or PIV testing.08/02/2006 Online CM 4447691 Adjust Packing (tightened N/A N/A No impact on 2 flats); no AL LLRT/PIV LLRT or PIV required. testing.Fall/2006 RE23 CM 4531453 Motor Pinion Gear N/A PIV: .No impact on Inspection 0.435 gpm (< 5 gpm) LLRT or PIV testing.LLRT not required due to option B.2nd periodic PIV test.02/05/2008 Online PM 4532685 Examine MO -Mech N/A N/A No impact on LLRT or PIV__________________ ___________ _________ ___________ _______________________________________ _____________ testitesigg Revision 0 Pg 1 Page 115 Cooper Nuclear Station Fifth Interva!Inservice Testing Program for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 _____Component(s) ~ -Date jjOutage 1Work Order Work Order Description AF Tests AL Tests TComments Spring/2008 RE24 CM 4547083 CM 4561 197 Refurb and test MO Install ETT/QSS LLRT: 0.86 scfth (< 10 scflh)LLRT: 1.05 sofhi ( < 10 scfhi)PIV: 0.19 gpm (< 5 gpm)AF-AL LLRT. No impact on LLRT or PIV testing. 3rd periodic PlV test.Fall/2009 RE25 CM 4723418 Adjust close limit switch PIV: N/A No impact on 0 gpm (< 5 gpm) LLRT or PIV testing.LLRT not required due to option B.4th periodic PIV test.Spring/201 1 RE26 N/A N/A PIV: N/A LLRT not C, gpm (< 5 gpm) required due to option B.5th periodic PIV test.08/10/2011 Online PM 4749833 Clean, Lubricate, and N/A N/A No impact on Partial Stroke LLRT or PIV testing.Fall/2012 RE27 PM 4848626 Examine MO (Mech & LLRT: N/A No impact on Elec); 0.1528 scfh (< 10 LLRT or PlV sefh) testing. 6th PIV: periodic PIV_______ ______ _______0 gpm (< 5 gpmn) __________test. Fall/2014 RE28 CM 4945454 PM 4950123 Viper Test; AL LLRT/PIV tests not required due to minimal change in packing and seating forces.Examine MO (Clean/Lube Stem)LLRT: 1.1 scth (< 10 scfhi)No impact on LLRT or PIV testing. PIV test not required due to approved relief request.Revision 0 Pg 1 Page 116 Cooper Nuclear Station Ffifh Interval lnservice Testing Pro gram for Pumps and Vailves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 _____Component(s) ~ -Date jOutage Work Order LWork Order Description ] AF Tests ] AL Tests Comments CS-MOV-MO12B (Test #8)Spring/2005 RE22 N/A N/A LLRT: 1 .23 scfih (< 10 scflh)PIV: 0 gpm (< 5 gpm)N/A Periodic LLRT and assume 1st periodic PIV test.08/14/2006 Online PM 4465037 Clean/Lube/Partial Stroke -N/A N/A No impact on LLRT or PIV CM 4334765 Periodic Diagnostic Test; testing.no AL LLRT/PIV_______ Fall/2006 RE23 CM 4534089 Motor Pinion Gear PIV: N/A No impact on Inspection .C gpm (< 5 gpm) LLRT or PIV testing.LLRT not required due to option B.2nd periodic PIV test.Spring/2008 RE24 CM 4561198 Install ETT/QSS PIV: N/A No impact on 0.082 gpm (< 5 LLRT or PIV gpm) testing.LLRT not required due to option B.3rd periodic PIV test.Fa11/2009 RE25 PM 4658094 Examine & Clean LLRT: N/A No impact on Operator 1.67 scfha (< 10 scfha) LLRT or PIV PIV: testing. 4th 0 gpm (< 5 gpm) periodic PIV test.11/12/2009 Online PM 4625209 Clean/Lube/Partial Stroke N/A N/A No impact on LLRT or PIV testing.Revision 0 Pg 1 Page ll7 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 Component(s) -'Date Outage Work Order Work Order Description AF Tests AL Tests Comments Spring/201 1 RE26 PM 4767601 Examine Motor Operator PIV: N/A No impact on 0Ogpm (< 5gpm) LLRT or PIV testing.LLRT not required due to option B.4th periodic___________PIV test.Fall/2012 RE27 CM 4840074 Viper Test LLRT: LLRT: No impact on 1.82 scfih (< 10 scfh) 2.02 scfh LLRT or PIV PM 4848541 Examine MO (Mech/Elec) PIV: testing. 5th 0.0 136 gpm (< 5 periodic PIV________gpm) test.Fall/2014 RE28 PM 4950054 Examine MO (Clean/Lube N/A N/A No impact on Stem) LLRT or PIV testing. No LLRT test performed due to option B and no PIV test performed due to an approved relief request.CS-CV-18CV Spring/2005 RE22 N/A N/A PIV: N/A LLRT not (Test #9) 0 3264 gpm (< 5 required due gpem) to Option B;periodic PIV______________test. Fa11/2006 RE23 N/A N/A PIV: N/A LLRT not 0.326 gpm (< 5 required due gpm) to Option B;periodic PIV___________ __________ ___________ _______________________________________ _______________ testest Revision 0 Page 118 Revision 0 Page 118 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 Component(s) '-Date Outage Work Order Work Order Description AF Tests AL Tests Comments Spring/2008 RE24 N/A N/A LLRT: N/A Periodic 1.19 scflh (< 15 scifi) LLRT and PIV: PIV tests.0.136 gpm (< 5____________ ~~~gpm) __________ ______Fall/2009 RE25 PM 4645121 Repack Valve LLRT: LLRT (Final AL): Significant CM 4724012 Disassemble and Repair 0.13 1 scfih (< 15 0.79 scfha (< 15 scfla) Maint. Resets following issues during scfh) PIV: LLRT/PIV repack 0 gpm (< 5 gpm) frequency. Spring/2011 RE26 N/A N/A PIV: N/A LLRT no 0 gpm (< 5 gpm) longer required due to closed loop analysis.First periodic PIV test.Fall/2012 RE27 N/A N/A PIV: N/A Second o gpm (< 5 gpm) periodic PIV test.Fall/2014 RE28 N/A N/A N/A PIV test not required due to approved relief request.CS-CV-19CV Spring/2005 RE22 N/A N/A LLRT: N/A Periodic (Test #10) 0.95 scfh (< 15 scflh) LLRT and PIV: PIV tests.0 gpm (< 5 gpm)Fall/2006 RE23 N/A N/A PIV: N/A LLRT not 0 gpm (< 5 gpm) required due to Option B;periodic PIV____________test. Spring/2008 RE24 CM 4631924 Adjust/add packing LLRT: LLRT (Final AL): Elected to PM 4541346 Repack valve 0.65 scfha (< 15 sctha) 1.4 scfha Reset PIV (Final AL): LLRT/PIV______________________________ ________0.05 gpm (< 5 gpm) frequency. Revision 0 Pg 1 Page 119 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request RV-05: Table 1: Maintenance and PIV/LLRT Test History Since 01/01/2005 _____Component(s) '-Date Outage Work Order Work Order Description AF Tests AL Tests Comments Fall/2009 RE25 N/A N/A LLRT: N/A First periodic 1.37 scih (< 15 setha) LLRT and PIV: PIV test.o gpm (< 5 gpm)Spring/2011 RE26 N/A N/A PIV: N/A LLRT no o gpm (< 5 gpm) longer required due to closed ioop analysis.Second periodic PIV test.Fall/2012 RE27 N/A N/A N/A N/A No PIV test required due to approved relief request.Fall!2014 RE28 N/A N/A N/A N/A No IPIV test required due to approved__________________ ___________ _________ ____________ _____________________ _________________ _________re___ie__ rlie ureqest AF =As Found AL = As Left CM = Corrective Maintenance PM = Preventative Maintenance Revision 0 Pg 2 Page 120 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves ATTACHMENT 5 AUGMENTED VALVE RELIEF REQUESTS AUGMENTED VALVE RELIEF REQUEST INDEX Relief [Description 1CNS Approval Date Request No._____ARV-01 DGDO Day Tank Valve Test Method ] 3-1-20 16 ARV-02 Diesel Fuel Oil Relief Valve Test Media ] 3-1-20 16 ARV-03 IfDGDO Check Valve Closure Tests j 3-1-20 16 Revision 0 Pg 2 Page 121 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request ARV-01 DGDO Day Tank Valve Test Method Alternative Provides Acceptable Level of Quality and Safety 1. Augmented Code Component(s) Affected Valve Class Category System DGDO-SOV-SSV5028 A B DGDO DGDO-SOV-SSV5029 A B DGDO 2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTC-5 151 (a), Active valves shall have their stroke times measured when exercised in accordance with ISTC-3500. ISTC-3560, "Fail-Safe Valves." Valves with fail-safe actuators shall be tested by observing the operation of the actuator upon loss of valve actuating power in accordance with the exercising frequency of ISTC-35 10.4. Reason for Request Augmented Relief is requested from the requirements of table ISTC-5 151l(a) and ISTC-3560, as applicable, for the valves listed above. The proposed alternative provides an acceptable level of quality and safety. This Augmented Relief Request does not require NRC approval.5. Proposed Alternative and Basis for Use The DG Day Tank Fuel Safety Solenoid Valves (DGDO-SOV-SSV5028, SSV5029) have a passive safety function in the open position to permit filling of the associated Day Tank during fuel oil transfer operations. These solenoid operated valves also have an active defense-in-depth (non-safety) function in the closed position to prevent overfilling of the associated Diesel Fuel Tank. However, this is only a back-up function to the safety-related day tank high level switch function to stop the fuel oil transfer pump in the same division on a high level in the day tank.DGDO-SOV-SSV-5028, 5029 are encapsulated solenoid valves and are not provided with remote position indication or remote manual switches. The design of these valves prohibits visual verification of the physical position of the valve operator, stem, or internal components. If the safety-related high level switch fails, the solenoid valve will close at the high-high level.Revision 0 Pg 2 Page 122 Cooper Nuclear Station Fifth Interval lnservice Testing Program for Pumps and Valves Relief Request ARV-01 DGDO Day Tank Valve Test Method (Continued) Modification of the system to verify individual valve exercising capability is not practicable nor cost beneficial since no commensurate increase in safety would be derived. These valves are ASME non-code class valves and are not within the scope of the IST Program.Since the Diesel Fuel Oil transfer system is an Augmented IST System, the components within this system are not required to follow the requirements of class 1, 2, or 3 components. However, components are required to be tested in a manner that is commensurate with the level of safety that they provide. Although these valves are considered to be passive open valves, CNS will conservatively test these valves periodically in the open position. At least quarterly, the diesel fuel oil day tank level alarms and transfer pump control level switches are functionally tested during diesel runs. The solenoid valves shall be verified open by observing restoration of day tank levels and measurement of sufficient flows during the IST transfer pump test surveillances. The testing being performed will ensure that these valves may be relied upon to fulfill their open safety function.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents A version of this relief request was previously approved by CNS for the fourth ten-year interval as ARV-O 1.Revision 0 Page 123 Revision 0 Page 123 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Relief Request ARV-02 Diesel Fuel Oil Relief Valve Test Media Alternative Provides Acceptable Level of Quality and Safety 1. Augmented Code Component(s) Affected Valve Class Category System DGDO-RV-l10RV A C DGDO DGDO-RV-l11RV A C DGDO 2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004 Edition through 2006 Addenda 3. Applicable Code Requirement Mandatory Appendix I, paragraph 1-4130(a), "Test Media." Valves shall be tested with the normal system operating fluid and temperature for which they are designed. Alternative liquids and different temperatures may be used, provided the requirements of 1-43 00 are met.4. Reason for Request Augmented Relief is requested from the requirements of Mandatory Appendix I, paragraph I-4130(a), concerning the liquid test media. The proposed alternative provides an acceptable level of quality and safety. This Augmented Relief Request does not require NRC approval.5. Proposed Alternative and Basis for Use These augmented relief valves provide over pressure protection of the engine driven fuel oil pump suction due to thermal expansion. These relief valves are periodically tested onsite at CNS. However, performing this relief valve set pressure testing with diesel fuel oil would be an unsafe practice due to the dangerous characteristics associated with this media. The fuel oil is flammable, and could be damaging to the eyes and/or skin. Utilizing the fuel oil as the test media would also contaminate the test equipment, making it difficult to clean and re-use this test equipment for other relief valve testing.In order to alleviate the concerns with utilizing diesel fuel oil as the test media for these relief valves, water may be used with negligible effects on the results of the relief valve testing performed to satisfy Mandatory Appendix I. The diesel fuel oil is slightly more viscous than water, but this should not affect the 1ST set pressure or seat leakage tests for the following reasons. The set pressure is determined by applying a static force against the upstream portion of the valve disc and recording the initial gush or first sign of continuous flow through the valve.Revision 0 Pg 2 Page 124 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request ARV-02 Diesel Fuel Oil Relief Valve Test Media (Continiued) Utilizing fuel oil or water should have a negligible effect on this test. In support of this statement, the Anderson Greenwood Crosby Test Report #5595, which compared opening pressures with water, lube oil and fuel oil, concluded that a one-to-one correlation exists between Fuel oil and water. The test results indicated no significant differences in the opening pressures. The seat leakage testing criteria is set at zero leakage at 90% of set pressure. If water passes this test, then diesel fuel oil should also pass the test since water is less viscous than diesel fuel.These valves are Seismic Class IS and are ASME non-Code Class. Therefore, they are outside the scope of the 1ST requirements of 10CFR50.55a. The valves are included in the 1ST Augmented Program.Relief Valve testing will be performed per Mandatory Appendix I, with the exception that water will be utilized as the test media rather than diesel fuel oil.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was previously approved by CNS for the fourth ten-year interval as ARV-02.Revision 0 Page 125 Revision 0 Page 12 5 Cooper Nuclear Station Fifth Interval Inser-vice Testing Pro gram for Pumps and Valves Relief Request ARV-03 DGDO Check Valve Closure Tests Alternative Provides Acceptable Level of Quality and Safety 1. Augmented Code Component(s) Affected Valve Class Category System DGDO-CV-l0CV A C DGDO DGDO-CV-1lCV A C DGDO DGDO-CV-12CV A C DGDO DGDO-CV-13CV A C DGDO 2. Applicable Code Edition and Addenda American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004 Edition through 2006 Addenda 3. Applicable Code Requirement ISTC-5221 (a), "The necessary valve obturator movement during exercise testing shall be demonstrated by performing both an open and a close test." 4. Reason for Request Augmented Relief is requested from the requirements of ISTC-522 1(a), concerning the closure testing of these check valves. The proposed alternative provides an acceptable level of quality and safety. This Augmented Relief Request does not require NRC approval.5. Proposed Alternative and Basis for Use DGDO-CV- 10CV and DGDO-C V- 11CV are the fuel oil transfer pump discharge check valves and DGDO-CV-12CV and DGDO-CV-13CV are the day tank inlet check valves. These augmented check valves have a safety function in the open position to allow transfer of fuel oil to their respective diesel fuel oil day tank during normal diesel generator operation. These check valves have no safety function to close, but the code requires bi-directional testing of check valves.Some plants may have elected to classify the Diesel Fuel Oil Transfer pumps and their associated check valves as being skid-mounted in accordance with the ISTA-2000 definition for skid-mounted pumps/valves and the ISTC-l1200 exemptions. This position would allow these components to be considered adequately tested with an acceptable diesel generator run.However, CNS has elected to include the pumps within the augmented 1ST Program to better track the performance of these pumps. As a part of this pump testing, the check valves associated with each transfer pump are tested to the open position by verification that the valves can deliver Revision 0 Pg 2 Page 126 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Relief Request ARV-03 DGDO Check Valve Closure Tests (Continued) > 4.64 gpm, which is the flow required to support continuous operation of a diesel generator at full load. The closure function, however, cannot be tested within this surveillance. The fuel oil transfer pump headers are isolated downstream of the check valves via normally closed cross-tie manual valve, DGDO-V-l19. With this manual valve in the passive closed position, the potential of a diversion of fuel oil from one division to the other is eliminated. Satisfactory surveillance testing of the diesel generators and fuel oil transfer pumps ensure that sufficient fuel flow is transferred to the day tanks. Finally, the diesel fuel oil discharge header cross-tie valve, DGDO-V-1 6, is a normally closed passive manual valve that is never opened during normal operations or during transient or accident conditions, which alleviates the conccrn of backflow through this line.Since the Diesel Generator Diesel Fuel Oil transfer System is an Augmented IST System, the components within this system are not required to follow the requirements of class 1, 2, and 3 components. However, the components are required to be tested in a manner that is commensurate with the level of safety that they provide. Testing DGDO-CV-10CV, DGDO-CV 11CV, DGDO-CV-l12CV, and DGDO-CV-l13CV in the open position clearly demonstrates that the check valves within the diesel fuel oil transfer system are adequately tested and may be relied upon to fulfill their safety functions. This is consistent with CNS Engineering Evaluation, EE 08-026, which reconfigured DGDO-V-l19 from open to closed. For these reasons, the closure testing of these check valves are not required.6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was previously approved by CNS for the fourth ten-year interval as ARV-03.Revision 0 Page 127 Revision 0 Page 127 Cooper Nuclear Station F~fih Interval Inservice Testing Program for Pumps and Valves ATTACHMENT 5 GENERAL RELIEF REQUESTS VALVE RELIEF REQUEST INDEX Relief Request Tecito R prvlDt No.DeritoNRAprvlDe RG-O1 [ASME OM Code Test Frequencies Pending Revision 0 Pg 2 Page 12 8 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves ASME OM Code Test Frequencies Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2) Hardship or Unusual Difficulty without a Compensating Increase in Level of Quality and Safety 1. ASME Code Component(s) Affected All Pumps and Valves contained within the Inservice Testing Program (IST) scope.2. Applicable Code Edition and Addenda ASME OM Code 2004 Edition through 2006 Addenda 3. Applicable Code Requirement This request for relief applies to the frequency specification of the ASME OM Code for all pump and valve testing contained within the IST Program scope. The applicable ASME OM Code (2004 Edition through the 2006 Addenda) sections include the following: Code Paragraph Description ISTA-320(a)The frequency for inservice testing shall be in accordance with the ISTA3 12(a)requirements of Section 1ST ISTB-3400 Frequency of I~nservice Tests ISTB-6200 Corrective Action ISTC-35 10 Exercising Test Frequency ISTC-3540 Manual Valves ISTC-3560 Fail-Safe Valves ISTC-3630(a) Frequency ISTC-3700 Position Verification Testing At least one valve from each group shall be disassembled and examined ISTC-522 1 (c)(3) at each refueling outage; all valves in a group shall be disassembled and examined at least once every 8 years.ISTC-5222 Condition-Monitoring Program ISTC-5230 Vacuum Breaker Valves ISTC-5 240 Safety and Relief Valves ISTC-5260 Explosively Actuated Valves Appendix 1*, 1-1320 Test Frequencies, Class 1 Pressure Relief Valves Appendix I, 1-1330 Test Frequency, Class 1 Nonreclosing Pressure Relief Devices Test Frequency, Class 1 Pressure Relief Valves That Are Used for Appedix , 1-340 Thermal Relief Applications Appendix I, 1-1350 Test Frequency, Classes 2 and 3 Pressure Relief Valves Appendix I, 1-1360 Test Frequency, Classes 2 and 3 Nonreclosing Pressure Relief Devices Test Frequency, Classes 2 and 3 Primary Containment Vacuum Relief Appendix I, 1-1370 Vle Test Frequency, Classes 2 and 3 Vacuum Relief Valves, Except for Appedix , 1-380 Primary Containment Vacuum Relief Valves Revision 0 Pg 2 Page 129 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RG-01 ASME OM Code Test Frequencies (continued) Code Paragraph Description Test Frequency, Classes 2 and 3 Pressure Relief Devices that are Used Appedix , 1-390 for Thermal Relief Application. Appendix II**, II- Performance Improvement Activities 4000(a)Appendix II, II- Optimization of Condition-Monitoring Activities 4000(b) ______________________________

  • Appendix I is for Pressure Relief Devices** Appendix II is for the Check Valve Condition Monitoring Program (CVCM)4. Reason for Request Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph (z)(2), relief is requested from the frequency specification of the ASME OM Code. The basis of the Relief Request is that the Code requirement presents an undue hardship without a compensating increase in the level of quality and safety.The ASME GM Code, 2004 Edition through the 2006 Addenda, establishes the inservice test frequency for all components within the scope of the Code. The frequencies (e.g., quarterly) have always been interpreted as "nominal" frequencies (generally as defined in Table 3.2 of NUREG 1482, Revision 2) and if necessary, owners applied the surveillance extension time period (i.e.grace period) contained in the plant Technical Specifications (TS) SRs. The CNS TS SR 3.0.2 states that the specified frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency.

This would allow an extension of up to 25% of the surveillance test interval to accommodate plant conditions that may not be suitable for conducting the surveillance. However, regulatory issues have been raised concerning the applicability of the TS grace period to ASME GM Code required inservice test frequencies. The lack of a tolerance band (grace period) on the ASME GM Code IST frequency restricts operational flexibility. There may be a conflict where an IST test could be required (i.e., its frequency could expire), but it is not possible or not desired that it be performed until sometime after a plant condition or associated TS is applicable. Therefore, to avoid this conflict, the IST test intervals should be allowed to be extended by up to 25%.Thus, just as with TS required surveillance testing, some tolerance is needed to allow adjusting GM Code testing intervals to suit the plant conditions and other maintenance and testing activities. This assures operational flexibility when scheduling 1ST tests that minimize the conflicts between the need to complete the test and plant conditions.

5. Proposed Alternative and Basis for Use Code Case OMN-20 is included in the ASME GM Code, 2012 Edition, and will be used as an alternative to the frequencies of the ASME GM Code. The requirements of Code Case OMN-20 are described below.Revision 0 Pg 3 Page 13 0 Cooper Nuclear Station Fifthi Interval Inservice Testing Pro gram for Pumps and Valves Relief Request RG-01 ASME OM Code Test Frequencies (continued)

ASME OM, Division 1, Section 1ST and all earlier editions and addenda specify 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.).(a) 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 as shown in Table 1. The specified time period between tests may be reduced or extended as follows: (1) For periods specified as fewer than 2 years, the period may be extended by up to 25% for any given test.(2) For periods specified as greater than or equal to 2 years, the period may be extended by up to 6 months for any given test.(3) All periods specified may be reduced at the discretion of the owner (i.e., there is no minimum period requirement). Period extension is to facilitate test scheduling and considers plant operating conditions that may not be suitable for performance of the required testing (e.g., performance of the test would cause an unacceptable increase in the plant risk profile due to transient conditions or other ongoing surveillance, test, or maintenance activities). Period extensions are not intended to be used repeatedly merely as an operational convenience to extend test intervals beyond those specified. Period extensions may also be applied to accelerated test frequencies (e.g., pumps in alert range)and other fewer than 2 year test frequencies not specified in Table 1.Period extensions may not be applied to the test frequency requirements specified in Subsection ISTD, Preservice and Inservice Examination and Testing of Dynamic Restraints (Snubbers) in Light-Water Reactor Nuclear Power Plants, as Subsection ISTD contains its own rules for period extensions.(b) Components whose test frequencies are based on the occurrence of plant conditions or events may not have their period between t ests extended except as allowed by ASME OM, Division 1, Section IST, 2009 Edition through OMa-20 11 Addenda and all earlier editions and addenda.Revision 0 Page 131 Revision 0 Page 131 Cooper Nuclear Station F~fih Interval Inservice Testing Program for Pumps and Valves Relief Request RG-01 ASME OM Code Test Frequencies (continued) Table 1 Specified Test Freauencies Frequency Specified Time Period Between Tests Quarterly92dy (or every 3 months)92dy Semiannually 14dy (or every 6 months) 14dy Annually36 (or every year) 36days x yearsx calendar years where x is a whole number of x_____years_____________years > 2 6. Duration of Proposed Alternative This proposed alternative will be utilized for the entire fifth ten-year interval.7. Precedents This relief request was previously approved for the Fermi-2 third ten-year interval as Relief Request PVRR-001 (TAG No. MF2967, dated July 16, 2014).Three Mile Island Nuclear Station, Unit 1 -Relief Requests PR-0 1, PR-02, and VR-02, Associated With The Fifth 10-Year Inservice Test Interval (TAG Nos. MF0046, MF0047 and MF0048, dated August 15, 2013).Revision 0 Pg 3 Page 132 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves ATTACHMENT 7 COLD SHUTDOWN JUSTIFICATIONS COLD SHUTDOWN JUSTIFICATION INDEX Cold Shutdown Derito Justification Derito No.[CSJ-02 ]jLPCI-AOV-PCV50 Exercising K CSJ-03 ]jMS-AO V-AO80AiB/C/D and MS-AGOV-AO86A/B/C/D Exercising -CSJ-04 ]~RR-MOV-MO53AIB Exercising [CSJ-05 ]~RHR-MOV-MO17/18 Exercising [CSJ-06 ][RHR-MOV.-920MV/921MV Exercising CSJ-07 ][SW-AOV-TCV451lA/B Exercising CSJ-08 tt IV-MOV-262/264/266/268MV Exercising (Augmented) CSJ-09 ]~CS-MOV-MO 12AIB Exercising-CSJ-1 0 ][Exercising of Backseated Valves Revision 0 Pg 3 Page 133 Cooper Nuclear Station F~ifth Interval Inservice Testing Pro gram for Pumps and Valves Cold Shutdown Justification CSJ-O1 (Augmented) Valve Number System Class Categ~ory, SGT-CV-14CV SGT A C SGT-CV-15CV SGT A C Function These check valves open to provide flow paths from their respective SGT filter trains and close to prevent back flow from the operating SGT train discharge. Justification The required full flowrate for these valves is > 1602 cfmn. During plant power operations, system conditions exist that prevent the SGT system from achieving a flowrate of > 1602 cfm. Reactor building differential pressure and back pressure from the Off Gas Dilution Fans act against SGT system pressure restricting SGT system flowrate. Therefore, it is impracticable to perform a fall flowrate test of these valves at power operations. These valves are ASME non-code class and are not within the scope of the IST Program.Alternative Test A full flow open test will be performed on these valves during cold shutdown periods. The full flow test shall also satisfy the closure exercise requirements of the check valve in the idle train.Revision 0 Page 134 Revision 0 Page 134 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Cold Shutdown Justification CSJ-02 Valve Number System Class Category, IIPCI-AO V-PC V50 HIPCI 2 B Function Air operated, pressure regulating valve for the cooling water supply line to the HPCI lube oil cooler. The valve perforns an active safety function in the open/throttled position to allow cooling water flow to the lube oil cooler.Justification Thifs valve functions to control pressure in the cooling water supply line to the H{PCI turbine lube oil cooler. Cooling water is supplied from the HPCI booster pump discharge. The valve is normally maintained in the closed position as a result of the HPCI pump being idle and pressure maintenance, supplied by the auxiliary condensate system, maintaining pressure above the control valve's set point.The valve travels to a throttled position when the HPCI pump starts to automatically maintain pressure im the cooling water line at 50 psig. The valve is designed to fail to the open position on a loss of instrument air to ensure continuity of cooling water flow to the lube oil cooler. Fail open travel is limited by a travel stop in order to prevent the downstream relief valve from lifting. The travel stop is set such that the required flow rate to the lube oil cooler is met under all operating conditions. Stroke timing and fail-safe testing to the open position online would require HIPCI to be declared inoperable and the LIPCI pressure controller, HTPCI-PC-50, to be taken out of service. Then, with instrument air isolated, a test rig would be utilized to apply air to the HPCI-AO V-PC V50 actuator in order to close the valve. Upon completion of valve closure, air to the actuator would then be removed, allowing the valve to be timed as it travels to the open position. This complex process to time this valve open is impracticable to perform every quarter independently during power operation and is also impracticable to perform in conjunction with the HLPCI run. HIPCI testing duration is limited by the resulting suppression pool heat up and Tech Spec temperature limitations. Alternative Test This pressure control valve shall be stroke timed and fail-safe tested open during cold shutdowns by manipulating the valve position by controlling air to the actuator, as discussed above. A partial stroke exercise test will be considered to be performed quarterly with the HPCI pump test.Revision 0 Page 135 Revision 0 Page 13 5 Cooper Nuclear Station F/ifth Interval Inservice Testing Pro gram for Pumps and Valves Cold Shutdown Justification CSJ-03 Valve Number System Class Category.MS-AGOV-AO8OA MS 1 A MS-AO V-AO80B MS 1 A MS-AGOV-AO80C MS 1 A MS-AOV-AO80D MS 1 A MS-AOV-AO86A MS 1 A MS-AO V-A086B MS 1 A MS-AO V-A086C MS 1 A MS-AOV-AO86D MS. 1 A Function The inboard and outboard main steam isolation valves (MSIVs) must be capable of automatic closure to limit the release of radioactivity during a reactor transient or accident condition and to prevent damage to the fuel barrier by limiting the loss of reactor coolant. water in case of a major leak from the steam piping outside of primary containment. Juotification Quarterly full closure testing of the MSJVs during 100% power operation is impracticable due to the potential for reactor transients and scrams. Also, full MSIV closure could create the potential of lifting the main steam safety relief valves (SRVs) due to an increase in steam line pressure. Failure of an SRV to re-close could result in reactor vessel depressurization. These affects may be minimized by performance at a reduced power level of < 70%, but then this activity contributes to the financial burden of reducing reactor power levels to facilitate valve testing. More importantly, the full stroke testing of MSIVs, even at reduced power, places the plant in an abnormal operating condition and introduces an unnecessary challenge to plant equipment. For example, the MSIVs are challenged to close and then re-open with steam in the lines, the plant must stabilize following the isolation and un-isolation of a Main Steam Line. Also, the testing has the potential to cause the plant to remain at a reduced power level and/or cause the initiation of a shutdown in order to make repairs.This would introduce additional equipment cycling and plant thermal transients. Therefore, the testing conditions associated with full-stroke testing the MSJVs online meets the deferral criteria outlined in NUJREG-1 482 Revision 2, section 2.4.5.Finally, per Technical Specification surveillance requirement, SR 3.6.1.3.6, which verifies that the isolation time of each MSIV is between 3 and 5 seconds, the frequency of the test is to be determined by the IST Program (via this cold shutdown). Technical Specification surveillance requirement, SR 3.3.1.1.9, requires a channel functional test to be performed every 92 days and may be satisfied by a partial stroke test closed.Alternative Test The MSJVs shall be partially exercised closed from the full open position, at least once per quarter, to satisfy Technical Specification requirement, SR 3.3.1.1.9, and in accordance with ASME GM Code ISTC-3520. Stroke timing to the closed position and fail-safe testing closed shall be performed on a cold shutdown basis in accordance with ASME GM Code ISTC-3 520.Revision 0 Page 136 Cooper Nuclear Station F~ith Interval Inservice Testing Program for Pumps and Valves Cold Shutdown Justification CSJ-04 Valve Number System Class Category RR-MOV-MO53A RR 1 B RR-MOV-M053B RR 1 B Function These valves are the Reactor Recirculation Pump 1 A and 1 B Discharge Isolation valves and have an active safety function to close in order to prevent diversion of LPCI injection flow following a LOCA.Justification Closure of either of the RR pump discharge valves at power would reduce recirculation flow and result in reactor water temperature transients and reactivity transients. These transients would reduce control of power distribution and fuel usage, and increase the risk of other plant transients. This could lead to decreased fuel reliability and increase the possibility of a fuel element failure. In addition, failure of these valves during operation would require reactor shutdown due to inaccessibility. Alternative Test These valves will be exercised (and stroke timed) to the closed position during cold shutdowns when the reactor recirculation system is not required to be mn service.Revision 0 Pg 3 Page 137 Cooper Nuclear Station F~ifh Interval Inser-vice Testing Program for Pumps and Valves Cold Shutdown Justification CSJ-05 Valve Number System Class Category.RHR-MOV-MO017 RI-ll 1 A RHR-MOV-MO018 RNR 1 A Function These valves are the reactor vessel return to the RHIR pump suction and containment isolation valves during reactor operations. These valves are only opened for low pressure shutdown cooling.Justification Valves RHR-MO V-MO 17 and RHIR-MO V-MO 18 are interlocked for pressure isolation during plant operation. Opening these valves during normal operation could possibly allow high pressure reactor coolant water into the low pressure suction lines of the RHR system. Therefore, it is essential that these valves remain closed during plant operations. Alternative Test These valves will be exercised (and stroke timed) to the closed position during cold shutdowns when reactor pressure isolation is not required.Revision 0 Pg 3 Page 13 8 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Cold Shutdown Justification CSJ-06 Valve Number System Class Category.RHR-MOV-920MV RER 2 B RHR-MOV-921MV RHR A B Function These valves provide isolation of main steam to the Augmented Off Gas (AOG) system.Justification The steam supply cannot be isolated during normal plant operation without causing significant Augmented Off Gas (AOG) system transients. Transients could include a fast or uncontrolled burn of hydrogen gas in the AOG piping buried underground and leading outside the plant. Also, routine quarterly testing of either of these valves could cause a release of radioactive material several orders of magnitude above normal release activities. Alternative Test These valves will be exercised (and stroke timed) to the closed position during cold shutdowns when the steam supply to the augmented off gas system may be isolated.Revision 0 Pg 3 Page 139 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Cold Shutdown Justification CSJ-07 Valve Number System Class Category.SW-AOV-TCV451A SW 3 B SW-AOV-TCV451B SW 3 B Function These valves open to provide a flow path for cooling water to the REC heat exchangers. Justification One temperature control valve is normally open to control flow to the associated REC heat exchanger. During the hot summer months both heat exchangers are in service. Placing either valve in the closed position for an exercise test during this period would interrupt the flow to the associated heat exchanger. The REC heat exchangers provide cooling water for a variety of essential and non-essential components. Therefore, it is essential that both of these valves remain open during plant operations. During cold shutdowns, when the heat load is reduced, one REC heat exchanger can be removed from service. The associated temperature control valve can then be closed and exercised to the full open position.Alternative Test Valve exercising (and stroke timing) to the open position will be performed quarterly except when both heat exchangers are in service. When both heat exchangers are in service, valve exercising (and stroke timing) to the open position will be performed during cold shutdowns, when the heat load is reduced.When a refueling outage falls within the seasonal period in which these valves may be tested quarterly, then an additional test during the refueling outage will not be required (i.e. valves may be tested online just prior to and after the refueling outage, being maintained on a quarterly frequency). Revision 0 Pg 4 Page 140 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Cold Shutdown Justification CSJ-08 (Augmented) Valve Number System Class Category.HV-MOV-262MV HV A B HV-MOV-264MV I-IV A B HV-MOV-266MV H-V A B HV-MOV-268MV LIV A B Function These valves open to provide flow paths for ventilation to be supplied through the casing of both MG sets for cooling purposes, and close upon receipt of a PCIS Group VI signal to provide secondary containment isolation. Justification These valves are required to remain in the open position during power operation to support reactor recirculation pump operation. Closure of these valves during reactor operation could result in overheating the MG set which would compromise reactor recirculation pump operation causing plant shutdown.Also, the valves' control circuitry does not provide for partial stroke capability. These valves are ASME non-code class valves and are not within the scope of the 1ST Program.Alternative Test These valves will be exercised (and stroke timed) to the closed position during cold shutdowns when the reactor recirculation pumps are in an idle state. During unscheduled cold shutdowns when the recirculation pumps are required to remain in operation, valve exercising will be deferred until the next available opportunity when the recirculation pumps can be removed from service.Revision 0 Pg 4 Page141 Cooper Nuclear Station F~ith Interval Inservice Testing Program for Pumps and Valves Cold Shutdown Justification CSJ-09 Valve Number System ClassCaeoy C S-MOV-MO 12A C s 1 A CS-MOV-MO12B CS 1 A Function Open to admit Core Spray water to the reactor vessel to mitigate the consequences of a LOCA and close for primary containment / pressure isolation valve functions. Justification These valves are normally closed for primary containment isolation and to isolate the Core Spray System from the reactor vessel pressure. Operating these valves at RCS pressures above 450 psig (quarterly) would require them to be manually cracked off their seat in the open direction. This action serves to equalize pressure across the valves since downstream of the valves may be pressurized by minor leak-by of the inboard check valves, CS-C V-i 8CV and CS-CV-19CV. Pressure equalization greatly decreases the forces required to pull the valve disk out of the seat and makes it possible to electrically stroke time the valve.Design calculation, NEDC 95-003, lists the MOV limiting components for the CS valves. The stem to disc T-head is the most limiting sub-component for the opening direction at normal operating conditions. The forces affecting the valve stem T-head are equivalent whether the valve is manually opened or electrically opened. Pressure forces acting on the disks must still be overcome by tension on the T-head.Manipulation of the equations in NEDC 9 5-003 that calculate opening forces indicate that the valve stem T-head limits could be exceeded under maximum reactor pressure and worse-case seat friction coefficients. In order to eliminate the possibility of overstressing the valve T-head during the quarterly surveillances, the stroke time test should be limited to periods during cold shutdowns when the reactor pressure is below 450 psig.Alternative Test Valve exercising (and strke timing) to the open and closed position for CS-MO V-MO 12A/B will be performed during cold shutdowns when the reactor pressure is below 450 psig.Revision 0 Pg 4 Page 142 Cooper Nuclear Station Ffith Interval Inser-vice Testing Program for Pumps and Valves Cold Shutdown Justification CSJ-IO Valve Number System Class Category.H{PCI-MOV-MO015 HPCI 1 A MS-MOV-MO74 MS 1 A RCIC-MOV-MO015 RCIC 1 A Function HPCI-MO V-MO 15: Normally open to provide steam to the HPCI turbine and closes automatically on high flow, low pressure, or high temperature signals.MS-MOV-M074: Normally open valve provided to prevent condensed steam from accumulating in the main steam lines and closes to isolate the main steam line inboard drain penetration. RCIC -MO V-MO 15: Normally open to provide steam to the RCIC turbine and closes automatically on high flow, low pressure, or high temperature signals.Justification This cold shutdown justification only applies if one or more of the motor operated valves listed have been power backseated in an effort to reduce Drywell leakage. This method of backseating is performed outside of the Drywell at the motor control centers for the valves. The valves, themselves, are located inside the Drywell. However, the backseated valves will be verified to meet any analyzed closed stroke time limits from the power backseated position prior to declaring the valve operable. The valve would then be placed back into the same backseated configuration and returned to service. If one or more of these valves have been backseated in this manner, it would be impracticable to perform subsequent stroke time testing each quarter based on the following discussion. Repeat backseating each quarter introduces the potential for causing component damage each quarter in addition to potentially re-initiating a Drywell packing leak that had previously been stopped. Power backseating of motor operated valves entails bypassing the normal valve open limit control and using the actuator motor to stall the valve into its backseat. The end result is that the valve stem seals off leakage past the packing by interference with a beveled backseat area in the valve bonnet. Valve vendors have supplied the maximum allowable backseat forces that can be applied. Although there are some risks involved, power backseating has been successfully performed on a limited basis in the Entergy fleet after engineering analysis.Once the motor has been stalled, it is desirable to not repeat this process more than necessary. The backseating process intentionally takes the motor out of the normal control circuit and causes motor heating that with repeated events would be severely damaging to the motor and motor cabling. In addition, if the backseating was previously successful in stopping a Drywell packing leak, the potential exists to re-initiate the packing leak with each backseating evolution. Finally, quarterly testing should not be pursued on backseated valves due to the significant evolution that would be required every quarter. If quarterly testing did continue, the valve would first be stroke timed.Then, an LCO would have to be entered and the backseating test equipment would have to be set up. In order to backseat the valve, maintenance must access the motor control center and auxiliary power. A variable transformer is connected to an alternate power supply, then motor cables are lifted at the motor control center. Outputs of the leads of the variable transformer are connected to the motor power cables, and this current is monitored with meters and the diagnostic test system. The circuit is closed and the Revision 0 Pg 4 Page 143 Cooper Nuclear Station Ffith Interval Inserviee Testing Program for Pumps and Valves Cold Shutdown Justification CSJ-1O (Augmented)(Continued) valve is first stroked from thc normally open position to partially closed. Opening the valve from the mid range position allows for the technicians to start motor movement at full voltage and then dial the voltage down to a predetermined reduced voltage level that controls the stall forces that impact the valve backseat. Therefore, it would be a significant evolution every quarter if the valve were to be stroke timed and re-backseated at this frequency. In conclusion, any one of the three valves listed may be backseated at some time in the future as a mitigating strategy associated with reducing Drywell leakage. Once one or more of these valves are backseated, the stroke time frequency will be changed from quarterly to a cold shutdown frequency. Quarterly testing is impracticable due to the increased potential for causing component damage, the potential of re-initiating a Drywell leak, and the significant evolution that would need to be undertaken every quarter to stroke time the valve and re-backseat it. There would be little to gain from this process, especially since MOVs are characteristically very steady with their stroke times over long periods of time.Therefore, the stroke time testing for these valves would require a re-classification from testing quarterly to a cold shutdown frequency for an interim period until the presence of a packing leak can be ruled out or the packing repaired.Alternative Test If backseated, valve stroke time testing for IHPCI-MO V-MO 15, MS-MO V-MO74, or RCIC -MO V-MO 15 will be performed during cold shutdowns. Revision 0 Pg 4 Page 144 Cooper Nuclear Station F~fih Interval Inservice Testing Program for Pumps and Valves REFUELING OUTAGE JUSTIFICATIONSIDE REFUELING OUTAGE JUSTIFICATION Th~DEX Refueling Outage Description Justification No.ROJ-O1 IINBI-CV-49B/50B/5 1B/52BCV and NBI-SOV-SSV73 8/739 Exercising ROJ-02 NBI-CV-55/56CV Exercising ROJ-03 RF-CV-1 3/14/15/1 6CV Exercising ROJ-04 [j RWCU-CV-15CV Exercising ROJ-05 IRWCU-MO-1 5/18 Exercising ROJ-06 CRD-CV-CV1 15 (Typical of 137) Exercising ROJ-07 IA-C V-17/1 8/1 9/20/21/22/36/37CV Exercising (Augmented) ROJ-08 J] IA-CV-28/29/30/3 1/32/33/34/35CV Exercising (Augmented) ROJ-09 ][ HIPCI-CV-29CV and RCIC-CV-26CV Exercising ROJ-10 ]~SW-MOV-MO89A/B Exercising ROJ-1 1 IICore Spray and RHRE Injection Check Valve Exercising,.. ROJ- 12 ]1 CRD-CV-25/26CV Exercising (Augmented) ROJ-13 ]j IA-CV-57/58/59/60CV Exercising (augmented) Revision 0 Pg 4 Page 145 Cooper Nuclear Station Fifthi Interval Inservice Testing Program for Pumps and Valves Refueling Outage Justification ROJ-O1 Valve Number System Class Category.NBI-CV-49BCV NBI 3 C NBI-CV-J0BCV NBI 3 C NBI-CV-51BCV NBI 3 C NBI-CV-52BCV NBI 3 C NBI-SOV-SSV738 NBI 3 B NBI-SOV-SSV739 NBI 3 B Function The reference leg injection check valves and solenoid operated valves have an active safety function in the open position to inject Core Spray water to the reactor vessel level instrumentation lines in case the reference leg water has flashed or boiled off due to accident conditions in the drywell.Justification This system provides the capability for the Core Spray System to supply a backfill of water for maintaining inventory of the Nuclear Boiler Instrumentation System cold reference legs (condensing chambers 3A and 3B) during accident conditions in the drywell where the reference leg inventory could be compromised. Exercising these valves to the open position, full or partial, would require manually isolating and venting the Cold Reference Leg Backfill System. This is not practicable during power operation or cold shutdown, other than refueling, due to the possible introduction of air into the system.This could cause a spurious reactor vessel level indication which could cause a reactor trip during power operation. During cold shutdown spurious level indications could interrupt the operation of systems required for decay heat removal, thereby placing the reactor in an unsafe condition. During refueling outages, sufficient time exists for decay heat to be reduced to a level which minimizes the impact of momentary interruption in the operation of systems required for decay heat removal such that testing can be performed. Alternative Test Exercising these check valves to the full open and closed positions, and full exercising with stroke timing to the open position of the solenoid operated valves, shall be performed during refueling outages when the cold reference leg backfill system may be isolated and vented to allow testing.Revision 0 Pg 4 Page 146 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Refueling Outage Justification ROJ-02 Valve Number System Class g r NBI-CV-55CV NBI 3 C NBI-CV-56CV NBI 3 C Function These Cold Reference Leg Continuous Backfill System check valves have an active safety function in the closed position to isolate the Class 3 instrumentation piping from the Seismic uIS non-class CRD piping.Justification This system provides for a continuous flow of water from the CRD drive water pumps to prevent noncondensible gases from building up in the Nuclear Boiler Instrumentation System cold reference legs (condensing chambers 3A and 3B). Exercising these valves to the closed position would require manually isolating and venting of the Cold Reference Leg Continuous Backfill System upstream of the check valves. This is not practicable during power operation or cold shutdown, other than refueling, due to the possibility of causing a spurious reactor vessel level indication from entrained air in the system.False level indications resulting from entrained air in the system may either cause a reactor trip during power operation or interrupt the operation of systems required during cold shutdown for decay heat removal, thereby placing the reactor in an unsafe condition. During refueling outages, sufficient time exists for decay heat to be reduced to a level which minimizes the impact of momentary interruption in the operation of systems required for decay heat removal such that testing can be performed. Alternative Test Exercising these check valves to the closed position shall be performed during refueling outages when the cold reference leg backfill system may be isolated and vented to allow testing. Exercise testing shall be accomplished by performing a seat leakage test. The open test will also be credited during each reactor refueling as is allowed per ISTC-3522(a). Revision 0 Pg 4 Page 147 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Refueling Outage Justification ROJ-03 Valve Number System Class Category.R+/--CV-13CV RF 1 A/C RF-CV-14CV RF 1 A/C RF-CV-15CV RF 1 A/C RF-CV- 16CV RF 1 A/C Function These valves are the main feedwater check valves which open to allow normal feedwater flow.Additionally, RF-CV-14CV and RF-CV-l6CV open to allow HPCI and RCIC flow to the vessel, respectively. All four of these valves must be capable of closure to provide containment isolation. Additionally, RF-C V-i13CV and RE-C V-l5CV must be capable of closure to prevent diversion of HPCI and RCIC flow, respectively. Justification These valves are normally open and must remain open during reactor operations to ensure adequate feedwater flow. Exercising these valves closed during plant operation could cause a transient in reactor water level resulting in a reactor scram. The observation of specified leakage during local leak-rate testing provides the only means for verification to the closed position.With the installation of highly accurate ultrasonic flow meters on the feedwater injection lines in RE24, per CED 6023681, individual flow rates for each feedwater injection line may now be determined at any time during normal operation. Therefore, the safety-related required flows through RF-CV-14CV and RE-C V-l6CV (in addition to the non-safety related flows through RE-C V-13CV and RE-C V-15C V), may be verified during normal operations. If the flow meters were to become unavailable, the open test for all the valves may be satisfied via an IST valve disassembly and examination (which requires manual exercising of the valve and visual inspection), or a torque test, both of which would be completed during a refueling outage due to their location and the allowance provided by ISTC-3522(a). In addition, the non-safety open test for RE-C V-i13CV and RE-CV-1 5CV may be satisfied through Operation logs, in which verification of being at full power would verify that adequate feedwater flows exist in supporting this power.Alternative Test These valves will be exercised to the closed position during the Type C leak rate test performed each refueling outage in accordance with the requirements of ASME OM Code ISTC-3 522 for category C check valves and ISTC-3620 for Containment Isolation valves. The open direction test will be credited at least once each reactor refueling as is allowed per ISTC-3522(a). Normally, the open flow test will be satisfied through the recording of acceptable flows through each injection line during normal operation per ISTC-3550. If the flow meters become unavailable, disassembly and examinations (all), torque testing (all), or verification of full power (1 3CV, 1 5CV only) may be utilized.Revision 0 Page 148 Revision 0 Page 148 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves Refueling Outage Justification ROJ-04 Valve Number System Class Category.RWCU-CV- 15 CV RW 1 A/C Function This check valve is normally open to allow Reactor Water Cleanup (RWCU) return and must close to provide containment isolation. Justification This valve cannot be verified as being closed upon reversal or stopping of flow without opening and venting the line upstream of the check valve. Opening or venting the RWCU line during operations could cause a leak of high pressure reactor coolant and potentially lead to the release of radioactive material.Stopping RWCU flow during normal operations or cold shutdown for an extended period would lead to a degradation of reactor water purity. This would add to the radioactive contamination in the reactor coolant system and could lead to additional exposure of site personnel. It is essential that RWCU remain in operation as much as possible and RWCU-C V-i15CV be exercised to the closed position only during refueling outages.Alternative Test The open capability of RWCU-C V-i15CV is verified during normal operations. The closure capability will be verified during refueling outages by performing a Type C local leak rate test per the requirements of ASME OM Code JSTC-3 522 and the CNS Primary Containment Leakage Rate Testing Program. The open direction test will be credited at least once each reactor refueling as is allowed per ISTC-3522(a). Revision 0 Pg 4 Page 149 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Refueling Outage Justification ROJ-05 Valve Number System Class Category.RWCU-MO- 15 RW 1 A RWCU-MO -18 RW 1 A Function These normally open Reactor Water Cleanup (RWCU) valves must close for containment isolation. The open function is not required for safe shutdown of the plant and is considered an operational function only.Justification These normally open containment isolation valves have a function during normal operation to provide a path for reactor coolant to and from the reactor water cleanup system to maintain high reactor water purity. In order to exercise these valves during normal plant operation, the RWCU system would need to be shutdown. Shutdown of the RWCU system induces chemistry transients in the reactor and should be minimized in order to maintain consistent reactor water chemistry. Additionally, shutdown of the RWCU system can lead to hydraulic transients and crud bursts that will result in increases in radiation levels and higher worker dose.Failure of one of these RWCU valves in the closed position would result in the complete loss of the RWCU system. This, in turn, could result in a plant shutdown to repair the valve due to the loss of reactor coolant chemistry parameters. Additionally, shutting down the RWCU system every quarter cycles the equipment without a compensating increase in safety. Shutdown of the RWCU system in a forced outage will also inhibit the ability to cleanup the vessel and result in an increase in radiation levels and personnel dose. Additionally, RWCU-MO-1 5 is located inside the primary contaimnment and is inaccessible during power operation due to high radiation levels and the inerted atmosphere. It is also impractical to de-inert containment for repair of this valve if it fails during cold shutdown testing.Refueling outages have sufficient duration to allow the RWCU System to adequately cleanup the primary coolant prior to being shutdown for testing. Additionally, the refueling outage schedules include periods in which RWCU must be shutdown while maintenance is performed on its support systems. If a tested RWCU valve does fail in the closed position during a refueling outage, adequate time is available to correct the condition without impacting unit availability and without adverse ALARA effects.Alternative Test The closure capability of RWCU-MO-15 and RWCU-MO-l18 will be verified during refueling outages by performing a stroke time test in the closed direction. Revision 0 Pg 5 Page 150 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Refueling Outage Justification ROJ-06 Valve Number System Class Category.CRD-CV-CV 115 CRD 2 C (Typical of 137)Function These valves have a safety function to close to prevent the loss of water pressure in the event that charging supply pressure is lost which prevents bypassing scram water (from the accumulator) to the charging water header (if depressurized). This valve has a function to open to allow charging water to pass from the control rod drive pumps to the hydraulic control units. Flow to the accumulators is required only during scram reset or system startup. This valve has no safety function to open.Justification Exercising these valves requires the depressurization of the charging water header. The header is depressurized by either stopping the CRD pumps or by valving out and depressurizing the charging water header. Stopping the CRD pumps could result in seal damage to the control rod drive mechanisms (CRDM) from a loss of seal cooling water. Additionally, stopping the pumps would interrupt seal cooling water flow to the reactor recirculation pumps resulting in shaft seal damage. This is impracticable during normal plant operation since valving out and depressurizing the charging water header would render the CRD accumulators inoperable and stopping the CRD pumps could cause pressure variations in the CRLD System during the test evolution. Exercising these valves during cold shutdown is not possible due to the interruption of shaft seal cooling water flow as previously discussed. If the recirculation pump became idle stopping the CRD pumps or manually isolating the charging water header for reverse exercise testing could delay plant startup due to the necessity of depressurizing upstream of each individual valve (137 each) in order to accomplish an adequate test. This additional test activity during cold shutdown represents an unusual burden without a compensating increase in the level of quality and safety.Alternative Test These valves will be tested during each reactor refueling outage. Proper closure shall be verified by isolating each of the CRD scram accumulators and venting pressure on the upstream side of the check valve. Accumulator pressure decay would be observed should the respective valve fail to close properly.The open test will also be credited during each reactor refueling as is allowed per ISTC-3522(a). Revision 0 Pg 5 Page 151 Cooper Nuclear Station F/ifh Interval Inservice Testing Program for Pumps and Valves Refueling Outage Justification ROJ-07 (Augmented) Valve Number IA-CV- 1 7CV IA-C V-i18CV IA-C V-i 19CV IA-C V-20C V IA-C V-21 CV IA-C V-22C V IA-C V-36C V IA-C V-37C V System IA IA IA IA IA IA IA IA Class A A A A A A A A Cateaorv A/C A/C A/C A/C A/C A/C A/C A/C Function These valves are the Instrument Air/Nitrogen supply inlet check valves for Main Steam Relief Valve (SRV) accumulators. These check valves must be capable of closure to maintain accumulator integrity in the event of a loss of normal actuating air supply.The check valves must open to allow flow to their respective accumulators. Justification These valves are located inside the drywell and are inaccessible during normal operations or cold shutdowns. They cannot be exercised during each cold shutdown because the drywell is not routinely de-inerted each cold shutdown. Valve exercising during cold shutdown when the drywell is de-inerted could delay plant restart due to the necessity of using portable test equipment inside the drywell. The additional test activity during cold shutdown represents an unusual burden without a compensating increase in the level of quality and safety. Testing these valves during refueling outages is consistent with NUR.EG 1482, revision 2, section 3.1.1.3.These valves are ASME non-code class valves that are not within the scope of the IST Program.Alternative Test An extended time/pressure decay procedure will be used to verify each valve's closure. This will be done by venting the upstream side of the check valve and monitoring accumulator pressure to ensure each check valve functions properly. The above valves will be tested each refueling outage to verify valve closure. The open test will also be credited during each reactor refueling as is allowed per ISTC-3522(a). Revision 0 Page 152 Revision 0 Page 152 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Refueling Outage Justification ROJ-08 (Augmented) Valve Number System Class Ctgr IA-C V-28CV IA A A/C IA-CV-29CV IA A A/C IA-C V-30CV IA A A/C IA-CV-31CV IA A A/C IA-C V-32CV IA A A/C IA-CV-33CV IA A A/C IA-C V-34CV IA A A/C IA-C V-35CV IA A A/C Function These check valves must close to isolate individual Main Steam Isolation Valve accumulators for emergency gas supply.The check valves must open to allow flow to their respective accumulators. Justification These check valves do not have position indication devices. The only practicable method to verify valve closure is a pressure decay test. The valves are located in the steam tunnel and the drywell. They are inaccessible during operation and normal cold shutdowns. The complexity of the pressure decay test could delay plant startup after a cold shutdown when the drywell is de-inerted. Since these emergency air supply accumulators are a backup to the normal pneumatic supply, performing the test at refueling outages is adequate to assess valve operational readiness. Testing these valves during refueling outages is consistent with NUREG 1482, revision 2, section 3.1.1.3.These valves are ASME non-code class valves and are not within the scope of the IST Program.Alternative Test A pressure decay test will be performed each refueling outage to verify valve closure. The open test will also be credited during each reactor refueling as is allowed per ISTC-3522(a). Revision 0 Page 153 Revision 0 Page 153 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Refueling Outage Justification ROJ-09 Valve Number System Class Catgory, HIPCI-CV-29CV HPCI 1 A/C -RCIC-CV-26CV RCIC 1 A/C Function HTPCI-CV-29CV -Opens to provide a flow path from the HIPCI pump to the reactor vessel via the feedwater system; closes for primary containment isolation. RCIC-CV-26CV -Opens to provide a flow path from the RCIC pump to the reactor vessel via the feedwater system; closes for primary containment isolation. Justification These valves are normally closed to isolate the reactor coolant system and the HPCI and RCIC systems.Exercising these check valves to the open position during normal plant operation would require HPCI or RCIC injection to the reactor vessel. This would result in a perturbation of normal feedwater flow and unnecessary thermal cycling of the feedwater nozzles. It would also cause severe power fluctuations due to the relatively cold water from the Emergency Condensate Storage Tanks. Furthermore, these valves are located in the Steam Tunnel. During power operations, this area experiences temperatures of approximately 130 -1 40°F, and high radioactivity. In general, plant personnel are prohibited from entering this area during power operation due to these conditions. Testing during Cold Shutdowns is impractical. This testing would be performed from the Steam Tunnel and would require support from Radiological Protection, Operations, Mechanical Maintenance, and possibly Engineering. During a typical forced outage, entry into the Steam Tunnel would not be made.Therefore, for the sole purpose of performing this test, multiple departments would need to support the evolution. Radiological support would be required to obtain the necessary survey information in the steam tunnels and/or be in attendance with the personnel performing the testing. To exercise these testable check valves, the system is required to be removed from service such that pressure is equalized across the valves prior to exercise testing. In order to accomplish this, this test requires operations personnel to enter the steam tunnel in order to remove pipe caps on vent and test connections on either side of the check valves, connect hoses, and open these valves to equalize pressure.In comparison, set up of this type is required with the performance of an Appendix J leak test, which is specifically allowed to be extended to a refueling outage per section 4.1.6 of NUREG 1482, revision 2.Then, Mechanical Maintenance personnel would perform the required mechanical exercising of the check valves, using appropriate torque wrenches. Following the testing, Operations would be required to restore the system. If any questions arose concerning the testing, which could occur, as described in section 4.1.7 of NUJREG 1482, revision 2, Engineering Personnel could be required to get involved to analyze the data, possibly prolonging the cold shutdown, unnecessarily. A refueling outage would be of a significant duration that this would not be a concern. In addition, personnel safety concerns are heightened due to the lack of lighting in the steam tunnel during a forced outage. In conclusion, because of the required test setup and complexities described, this testing is impracticable to perform during cold shutdowns and may delay unit startup.Revision 0 Page 154 Revision 0 Page 154 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Alternative Test These valves will be mechanically exercised, verifying open and closure capability, during refueling outages when the IIPCI and RCIC systems are not required to be inservice. Revision 0 Pg 5 Page 155 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Refueling Outage Justification ROJ-10 Valve Number System Class Category.SW-MOV-MO89A SW 3 B SW-MO V-MO89B SW 3 B Function These are the Loop A and Loop B outlet isolation for the Service Water booster pump cooling water to the RHiR heat exchangers. These normally closed valves have an active safety function in the throttled position to provide a flow path for cooling water flow through the RI-R heat exchangers during transient and accident conditions. Justification These valves are exercised during quarterly Service Water Booster Pump flow testing to a throttled position required to satisfy Technical Specification flow requirements. Valve stroke timing to the fully opened position is impracticable to perform at power due to the potential to cause RHR Service Water Booster Pump run out. Each Service Water Booster Pump is electrically interlocked with its respective RHIR heat exchanger outlet valve (SW-MOV-MO89A/B). When the pump control switch is taken to START, its respective RHR heat exchanger outlet valve receives an open signal. When the valve reaches a position that ensures pump minimum flow requirements can be met, the pump receives a start signal.The RHIR heat exchanger outlet valve is throttled to obtain the desired flow. Each RHIR heat exchanger outlet valve is electrically interlocked to close when both associated RIHR Service Water Booster pumps are shutdown. These valves are also utilized to maintain RHR-SW shell pressure higher than the RHR system pressure to prevent the potential release of radioactivity into the river.It would also be impracticable to perform full stroke testing at a cold shutdown frequency since Service Water and Service Water Booster pumps are essential for providing cooling water to the RHR heat exchangers during this time period. Defeating the Service Water Booster pump interlock and windmilling the Service Water Booster pumps would not be a desirable activity. It would be more appropriate to perform this testing once the decay heat has lowered during a refueling outage.For the reasons and since this valve is verified to be capable of performing its safety function each quarter, it would be impractical to defeat the interlocks associated with this valve on a quarterly or cold shutdown frequency to obtain a full open stroke time test. It would be practical, however, on a once per refueling outage basis, to full stroke open these valves for trending either through the use of control room hand switches or MOV diagnostic equipment once decay heat has lowered. If hand switches are utilized, flow through the subsystem being tested should be manually isolated and interlock defeated to allow the full stroking of the valve.Alternative Test These valves will be exercised to their safety-related throttled position quarterly, but stroke times will not be measured. During refueling outages, these valves will be stroke time tested to the full open position through the use of the SW-MO V-MO89A/B hand switches with flow isolated and interlocks defeated. If desired, the stroke time may be obtained during MOV diagnostic testing.Revision 0 Pg 5 Page 156 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Refueling Outage Justification ROJ-11 Valve Number System Class. Category, CS-CV-l18CV CS 1 A/C CS-CV-19CV Cs 1 A/C RHR-CV-26CV RIHR 1 A/C RHR-CV-27CV RHR 1 A/C Function These valves open for Core Spray or LPCI injection and close for primary containment isolation. Justification These valves are normally closed for primary containment isolation. They are also closed to isolate the related low pressure systems from the Reactor Recirculation system and the reactor vessel. Opening these valves during power operation is not possible due to the downstream side being exposed to reactor pressure. A drywell entry during cold shutdown would be necessary to facilitate testing, thereby requiring de-inerting, which could potentially delay restart. Also, cold shutdown conditions in the Drywell have previously resulted in high radiation fields near the RHR lines and other areas of the Drywell. Finally, personnel safety concerns due to the Core Spray lines being located in the upper drywell elevations with little to no lighting may also be an issue. The reasons presented are acceptable reasons for deferral to a refueling outage basis per NUREG 1482, rev. 2, sections 2.4.5 and 3.1.1.3.Alternative Test These valves will be mechanically exercised, verifying open and closure capability during refueling outages in accordance with ISTC-3522 and ISTC-5221. Revision 0 Pg 5 Page 15 7 Cooper Nuclear Station Ffifhi Interval Inservice Testing Pro gram for Pumps and Valves Refueling Outage Justification ROJ-12 (Augmented) Valve Number System Class Category.CRD-CV-25CV CRD A A/C CRD-CV-26CV CRD A A/C Function These valves close to prevent possible CRD bypass leakage from exiting secondary containment. They open to supply drive water and charging water to the Hydraulic Control Units (HCUs) and seal water to the Reactor Recirculation pumps.Justification It is impracticable to perform a closure test on these valves during power operations or cold shutdowns. These valves are located in the line from the CRD pumps supplying drive water and charging water to the control rod's HCUs.During power operations, these valves are open since drive water is constantly supplied to the HCUs.Closure testing would require the CRD pumps to be secured and the portion of the system containing these valves to be isolated. Isolating the valves or securing the CRD pumps will terminate the constant drive water supply to the HCUs, causing all control rods to be inoperable. Also, without a continuous charging water supply, HCU accumulators would eventually depressurize and administrative controls require a scram initiation upon depressurization of two accumulators, which is a highly undesirable situation. The interruption of the drive water flow would also interrupt CRLD seal water cooling to the Reactor Recirculation pumps. The stopping of flow would impose a severe thermal transient on the RR pump seals, which could possibly lead to premature seal failure.The method of testing these valves for closure would be through a local leak rate test, which involves the draining of the system, establishment of test boundaries, equipment setup, etc., which is more suitable during a refueling outage. This test method, alone, could delay a plant startup from a cold shutdown.In conclusion, testing these valves during refueling outages is consistent with NUJREG 1482, rev. 2, sections 2.4.5, and 3.1.1.4.Alternative Test The closure capability will be verified during refueling outages by performing a local leak rate test. The open direction test will be credited at least once each reactor refueling as is allowed per ISTC-3522(a). Revision 0 Pg 5 Page 158 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Cold Shutdown Justification ROJ-13 (Augmented) Valve Number System Class Category.IA-C V-57CV IA A C IA-C V-58CV IA A C IA-C V-5 9CV IA A C IA-CV -6OCV IA A C Function These valves are required to close to maintain pressure in the associated air operated valve's accumulator in the event of a loss of the Instrument Air (IA) supply. These IA supplies are for various safety related I-&V system AOV's.Justification It is impracticable to perform a closure test on these valves during power operations or cold shutdowns. IA-CV-57CV, -58CV, -59CV, AND 6OCV are in the IA supply lines to the H&V supply and exhaust isolation dampers for the RRMG Set IA and lB. Isolation and testing of these valves potentially affects the operation of the RRMG Sets. This would be undesirable during power operations and during cold shutdowns when the recirculation pumps are required to be operational. Also, the closure test for these check valves requires that the IA supply piping upstream of the check valves to the associated accumulators be isolated and depressurized. The accumulator pressure is then monitored for one hour via installed test equipment to verify that the check valve will hold.Depressurization of the accumulator below 70 psig with the respective H&V valve open will place the component in an inoperable condition. Due to the complexities of this test and the potential to delay a cold shutdown, it would be more appropriate to perform this testing during refueling outages.These valves are ASME non-code class valves and are not within the scope of the IST Program.Alternative Test A reverse flow leakage test of these check valves will be performed during refueling outages by performing a pressure decay test. The open test will also be credited during refueling outages shutdown as is allowed per ISTC -3522 (a).Revision 0 Page 1590 Page 159 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves ATTACHMENT 9 TECHNICAL POSITIONS TECIHNICAL POSITION INDEX Technical Decito Position No. Decito TP-O01 JfBi-directional Testing of Check Valves TP-02 ]I Sample Disassembly of HPCI Vacuum Breaker Check Valves TP-03 ][Passive Valves without Test Requirements TP-04 ]IFail Safe Testing of Valves TP-05 ]~Classification of Skid-Mounted Components TP-06 ]jCheck Valves in Regular Use TP-07 ]ICategorization of IST Pumps (Group A or B)TP-08 ]jVacuum Breaker Testing Revision 0 Page 160 Revision 0 Page 160 Cooper Nuclear Station Fifthi Interval Inservice Testing Pro gram for Pumps and Valves Technical Position TP-O1 (Page 1 of 3)Bi-directional Testing of Check Valves with Non-Safety Positions Purpose The purpose of this Technical Position is to establish the station position for the verification of the non-safety direction exercise testing of check valves by normal plant operations. Applicability. This Technical Position is applicable to those valves which are included in the Inservice Testing Program that are required to be exercise tested in their non-safety related direction of flow. This position applies to those check valves required to be tested in accordance with Subsection ISTC (ASME OM Code 2004 Edition through 2006 Addenda) and Appendix II. This Technical Position does not apply to testing of the safety function (direction) of check valves included in the Inservice Testing Program.Backg~round The ASME OM Code 2004 Edition through 2006 Addenda section ISTC-3550, "Valves in Regular Use", states: "Valves that operate in the course of plant operation at a frequency that would satisfy the exercising requirements of this Subsection need not be additionally exercised, provided that the observations otherwise required for testing are made and analyzed during such operation and recorded in the plant record at intervals no greater than specified in ISTC-3510." Section ISTC-35 10 requires that check valves shall be exercised nominally every 3 months with exceptions (for extended periods) referenced. Section ISTC-522 1 (a)(2) states: "Check valves that have a safety function in only the open direction 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 perform its intended function(s) (see ISTC-1 100), and verify closure." Section ISTC-5221 (a)(3) states: "Check valves that have a safety function in only the close direction shall be exercised by initiating flow and observing that the obturator has traveled [to] at least the partially open position, 3 and verify that on cessation or reversal of flow, the obturator has traveled to the seat."" 3 The partially open position should correspond to the normal or expected system flow." Normal and/or expected system flow may vary with plant configuration and alignment. Cooper Nuclear Station Operations staff is trained in recognizing normal plant conditions. For check Revision 0 Page 161 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves Technical Position TP-O1.(Page 2 of 3)valves that have a non-safety function in the open position, Operator judgment has been deemed acceptable in determining whether or not the normal or expected flow rates for plant operation has been obtained. For check valves that have a non-safety related function in the closed position, Operator judgment is also deemed acceptable in determining whether or not flow has occurred at a normal or expected flow rate, in order to cause obturator travel.Position Typically, Cooper Nuclear Station will verify the non-safety position of check valves included in the Jnservice Testing Program using a periodic activity within the plant surveillance or preventative maintenance program. In lieu of a dedicated surveillance or preventative maintcnance activity to perform the non-safety direction testing, the following alternate verifications may be performed as follows: 1. An appropriate means shall be determined which establishes the method for determining the open/closed non-safety function of the check valve during normal operations. The position determination may be by direct indicator, or by other positive means such as changes in system pressure, flow rate, level, temperature, seat leakage, etc. This determination shall be documented in the respective Condition Monitoring Plan for the specific check valve group. For check valves included in the Inservice Testing Program and not included in the Condition Monitoring Plan, this determination shall be documented in the 1ST Bases Document for the specific check valve group.2. Observation and analysis of plant processes that a check valve is satisfying its non-safety direction function may be used. For an example, consider a check valve that has a safety function only in the closed direction and normally provides a flow path to maintain plant operations. If this check valve does not open to pass flow when required, an alarm or indication would identify a problem to the operator. The operator would respond by takcing the appropriate actions. A Condition Report would then be generated for the abnormal plant condition which would identify the check valve failure.3. Observation and analysis of plant logs (i.e. Operations or Chemistry) and other records may be an acceptable method for verifying a check valve's non-safety direction function verification during normal plant operations. The open/closed non-safety function shall be recorded at a frequency required by ISTC-35 10, nominally every 3 months, (with exceptions as allowed), in plant records such as Cooper Nuclear Station Operating or Chemistry Logs, Electronic Rounds, chart recorders, automated data loggers, etc. The safety function direction testing requires a periodic Quality Record typically documented within the surveillance and/or preventative maintenance program.Records as indicated above in 1 through 3 are satisfactory for the non-safety direction testing. A condition report shall be generated for any issues regarding check valve operability. Justification This Technical Position establishes the acceptability of the methods used in determining the ability of a check valve to satisfy its non-safety function. Typically, Cooper Nuclear Station will verify the non-safety position of check valves included in the Inservice Testing Program using a Ravision 0 Page 162 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Technical Position TP-O1 (Page 3 of 3)periodic activity within the plant surveillance or preventative maintenance program.Alternatively, through normal plant system operation and operator actions, a valve's non-safety function is verified through either observation or analysis of plant records and logs. Additionally, the recording of parameters which demonstrate valve position is satisfied at a frequency in accordance with ISTC-35 10. These actions collectively demonstrate the non-safety position of Jnservice Testing Program check valves in regular use as required by ISTC-3550. Revision 0 Pg 6 Page 163 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Technical Position TP-02 (Page 1 of 2)Sample Disassembly of HIPCI Vacuum Breaker Check Valves The purpose of this Technical Position is to establish the station position for testing the HPCI Vacuum Breaker Check Valves.Applicabilit, This Technical Position is applicable to the IIPCI Vacuum Breaker Check Valves, HPCI-CV-24CV, 25CV, 26CV, and 27CV.Position These valves are normally closed checkc valves with two in series cross connected to two in parallel (H pattern). The four HPCI valves are swing check valves. The valves are located in the suppression pool fre~e space. In the closed position, they prevent steam from the exhaust line from entering the free space of the suppression chamber. Either two inboard valves or two outboard valves must be closed to perform this function. Two valves in series provide added assurance that steam will not enter the suppression chamber.The valves open to prevent siphoning suppression pooi water into the exhaust line due to steam condensing when the associated HiPCI system is isolated. Each pair of valves is cross connected to the parallel pair of valves (H pattern) so that a single failure will not prevent the vacuum relief function. These valves are not required to be leak tight and are not equipped with position indication or pressure sensing devices. During power operation, the suppression chamber is inerted and inaccessible. Due to the location and configuration of these valves (located on the open ended turbine exhaust lines) a typical closure or open test cannot be performed. These vacuum breaker check valves are not capacity certified. Therefore, they may be tested in accordance with ISTC-5220 (as a check valve). In accordance with the requirements of ISTC-5221 (c), a sample disassembly examination program is being utilized to satisfy 1ST testing for the HPCI vacuum breaker check valves.These valves are grouped together because they perform the same function, have the same manufacturer, design, service conditions and have the same size, materials of construction and orientation. A different valve from this group will be disassembled, inspected and manually exercised during each refueling outage until the entire group has been tested. If the disassembled valve selected is not capable of being full-stroke exercised or there is binding or failure of valve internals, the remaining valves in the group will be disassembled, inspected, and manually exercised during the same outage.Procedural requirements ensure the valve is re-installed correctly, so no open/closed testing is required following reassembly. Revision 0 Page 164 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves Technical Position TP-D2 (Page 1 of 2)Justification All requirements of ISTC-5221(c) and ISTC-9200(c) are met.Revision 0 Page 1650 Page 165 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Technical Position TP-03 (Page 1 of 1)Passive Valves Without Test Requirements Purpose The purpose of this Technical Position is to establish the station position for valves which perform a passive safety function. However, no testing is required in accordance with LSTC.Applicability This Technical Position is applicable to valves that perform a passive function in accordance with ISTA-2000 and do not have inservice testing requirements per Table ISTA-3500-1. This position is typical of Category B, passive valves that do not have position indication. 'An example is a manual valve which must remain in its normal position during an accident, to perform its intended function.' Typically, manual valves that perform a safety function are locked in their safety position and administratively controlled by Cooper Nuclear Station procedures. These valves would be considered passive. If they do not have remote position indicating systems and categorized as B, they would not be subjected to any test requirements in accordance with Table ISTC-3500-l. Position The Cooper Nuclear Station Inservice Testing Program, Valve Tables -Attachment 11, will not list valves that meet the following criteria.* The valve is categorized B (seat leakage in the closed position is inconsequential for fulfillment of the valves' required function(s)) in accordance with ISTC- 1300.* The valve is considered passive (valve maintains obturator position and is not required to change obturator position to accomplish the required function(s)) in accordance with ISTA-2000.

  • The valve does not have a remote position indicating system which detects and indicates valve position.Justification Valves that meet this position will not be listed in the Cooper Nuclear Station Inservice Testing Program, Valve Tables -Attachment 11, however, the basis for categorization and consideration of active/passive functions should be documented in the 1ST Program Basis Document.Revision 0 Pg 6 Page 166 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Technical Position TP-04 (Page 1 of 1)Fail Safe Testing of Valves Purpose The purpose of this Technical Position is to establish the station position for fail safe testing of valves in conjunction with stroke time exercising or position indication testing.Applicabilit.

This Technical Position is applicable to valves with fail-safe actuators required to be tested in accordance with ISTC-3560.

Background

The ASME OM Code 2004 through 2006 Addenda section ISTC-3560 requires;"Valves with fail-safe actuators shall be tested by observing the operation of the actuator upon loss of valve actuating power in accordance with the exercising frequency of ISTC-35 10." Section ISTC-3510 states;"Active Category A, Category B, and Category C check valves shall be exercised nominally every 3 months..." Position In cases where the valve operator moves the valve to the open or closed position following de-energizing the operator electrically, by venting air, or both, the resultant valve exercise will satisfy the fail-safe test requirements and an additional test specific for fail safe testing will not be performed. Cooper Nuclear Station will also use remote position indication as applicable to verify proper fail-safe operation, provided that the indication system for the valve is periodically verified in accordance with ISTC-3700. Justification Fail-Safe Testing tests the ability of the fail-safe mechanism of the valves to go to its fail safe condition. Whether or not the actuation of this fail-safe mechanism is due to Operator Action of failure of either the valve's air or electric power source, the resultant action of the valve will be the same. Therefore, the verification of a valve's fail safe ability can be taken credit for with the performance of either a stroke time exercising or position indication test.'Revision 0 Pg 6 Page 167 Cooper Nuclear Station Ffifh Interval Inserviee Testing Program for Pumps and Valves Technical Position TP-05 (Page 1 of 2)Classification of Skid-Mounted Components Purpose The purpose of this technical position is to clarify requirements for classification of various skid-mounted components, and to clarify the testing requirements of these components.

Background

The ASME Code allows classification of some components as skid-mounted when their satisfactory operation is demonstrated by the satisfactory performance of the associated major components. Testing of the major component is sufficient to satisfy Inservice Testing requirements for skid-mounted components. In section 3.4 of NUREG 1482 Rev l, the NRC supports the designation of components as skid mounted: "The staff has determined that testing the major component is an acceptable means to verify the operational readiness of the skid-mounted components and component subassemblies if the licensee discusses this approach in the IST program document. Licensees should consider and document the specific measurements and attributes of major component testing which relate to the assessment of skid-mounted component condition. In addition, various continuous and periodic observations of the major components (such as System Monitoring Walkdowns or Operator Logs) may also support assurance of skid-mounted component readinesss. This is acceptable for both Code class components and non-Code class components that are tested and tracked by the IST Program." The 2004 Edition through the 2006 Addenda of the ASME Cm Code, Subsection ISTA-2000 provides the following definition. Skid mounted pumps and valves -pumps and valves integral to or that support operation of major components, even though these pumps and valves may not be located directly on the skid. In general, these pumps and valves are supplied by the manufacturer of the major component. Examples include: (a) diesel fuel oil pumps and valves;(b) steam admission and trip throttle valves for high-pressure coolant injection turbine-driven pumps;(c) steam admission and trip throttle valves for auxiliary feedwater turbine driven pumps;(d) solenoid-operated valves provided to control an air-operated valve.Revision 0 Pg 6 Page 16 8 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Technical Position TP-05 (Page 2 of 2)Additionally the Subsections pertaining to pumps (ISTB) and valves (ISTC) include exclusions/exemptions for skid-mounted components; ISTB-1 200(c) Exclusions Skid-mounted pumps that are tested as part of the major component and are justified by the Owner to be adequately tested.ISTC-1 200 Exemptions Skid-mounted valves are excluded from this Subsection, provided they are tested as part of the major component and are justified by the Owner to be adequately tested.Position The 2004/2006a ASME OM Code definition of skid-mounted will be used for classification of components in the Cooper Nuclear Station Inservice Testing Program. In addition, for a component to be considered skid-mounted:

  • The major component associated with the skid-mounted component must be surveillance tested at a frequency sufficient to meet ASME Code test frequency for the skid mounted component.
  • Satisfactory operation of the skid-mounted component must be demonstrated by satisfactory operation of the major component.

If the skid mounted component is a check valve, it does not have to be exercised in both directions if both direction testing is not required to indicate satisfactory operation.

  • The IST Bases Document should describe the bases for classifying a component as skid-mounted, and the IST Program Plan should reference this technical position for the component, if listed.Justification Recognition and classification of components as skid-mounted eliminates the need for the redundant testing of the sub component(s) as the testing of major (parent) component satisfactory demonstrates operation of the "skid mounted" component(s).

Revision 0 Page 169 Revision 0 Page 169 Cooper Nuclear Station F~ifh lnterval Inser-vice Testing Program for Pumps and Valves Technical Position TP-06 (Page 1 of 2)Check Valves in Regular Use Purpose The purpose of this Technical Position is to establish the station position for check valves that are in regular use during normal plant operations. Applicability. This Technical Position is applicable to check valves that are capable of being demonstrated to be open during routine operations.

Background

The ASME OM Code 2004 through 2006 Addenda section ISTC-3550, "Valves in Regular Use", states: "Valves that operate in the course of plant operation at a frequency that would satisfy the exercising requirements of this Subsection need not be additionally exercised, provided that the observations otherwise required for testing are made and analyzed during such operation and recorded in the plant record at intervals no greater than specified in ISTC-3510." Section ISTC-35 10 requires that check valves shall be exercised nominally every 3 months with exceptions (for extended periods) referenced. Check valves that are a part of the 1ST Check Valve Condition Monitoring Program shall be tested per the frequency requirements of that program.Normal and/or expected system flow may vary with plant configuration and alignment. The open"safety function" of a check valve typically requires a specified design accident flow rate. For these subject valves, the normal system flow is above the design accident flow rates. Since the Cooper Nuclear Station Operations staff is trained so as to be able to recognize normal plant conditions, Operator judgment has been deemed acceptable for the purpose of determining check valve open demonstration by observing either normal or expected flow rates for the plant operating condition. Position Cooper Nuclear Station will verify the open position of these subject check valves by observing plant logs, computer systems, strip chart recorders, etc., during normal plant operations. The open/closed safety function shall be recorded at a frequency required by ISTC -3510 [or ISTC-5222, Condition Monitoring Program, if applicable], nominally every 3 months, (with exceptions as provided), in plant records such as Cooper Nuclear Station Operating Logs, Electronic Rounds, chart recorders, automated data loggers, etc.Revision 0 Page 170 Revision 0 Page 170 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves Technical Position TP-06 (Page 2 of 2)Justification Normal plant systems operation and operator actions provide for the observations and analysis that these subject valves are capable of satisfying their open safety function. Additionally, the recording of parameters which demonstrate valve position is satisfied at a frequency in accordance with JSTC-35 10 or ISTC-5222. These actions collectively demonstrate the open safety function of Inservice Testing Program check valves in regular use as required by ISTC-3550.Revision 0 Pg 7 Page 171 Cooper Nuclear Station F~fh~ Interval Inservice Testing Program for Pumps and Valves Technical Position TP-07 (Page 1 of 3)Categorization of IST Pumps (Group A or B)Position Cooper Nuclear Station has categorized the pumps required to be included in the Inservice Testing Program or Augmented (class "~A") Inservice Testing Program as either Group A or B in accordance with the requirements of ISTB-2004/2006a. Group A pumps are pumps that are operated continuously or routinely during normal operation, cold shutdown, or refueling operations. The following pumps are categorized as Group A at Cooper Nuclear Station: Pump CIC Class Group Type Function DGDO-P-DOTA A A Centrifugal Diesel Fuel Oil Transfer DGDO-P-DOTA A A Centrifugal Diesel Fuel Oil Transfer RW-P-Zl1 A A Centrifugal Elevated Release Point Sump RW-P-Z2 A A Centrifugal Elevated Release Point Sump REC-P-A 3 A Centrifugal 'Reactor Equipment Cooling REC-P-B 3 A Centrifugal Reactor Equipment Cooling REC-P-C 3 A Centrifugal Reactor Equipment Cooling REC-P-D 3 A Centrifugal Reactor Equipment Cooling RIIR-P-A 2 A Centrifugal Residual Heat Removal RHR-P-B 2 A Centrifugal Residual Heat Removal RHIR-P-C 2 A Centrifugal Residual Heat Removal RHIR-P-D 2 A Centrifugal Residual Heat Removal SW-P-A 3 A Vertical Service Water SW-P-B 3 A Vertical Service Water SW-P-C 3 A Vertical Service Water SW-P-D 3 A Vertical Service Water SW-P-BPA 3 A Centrifugal Service Water SW-P-BPB 3 A Centrifugal Service Water SW-P-BPC 3 A Centrifugal Service Water SW-P-BPD 3 A Centrifugal Service Water Revision 0 Pg 7 Page 172 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Technical Position TP-07 (Page 2 of 3)Group B pumps are those pumps in standby systems that are not operated routinely except for testing. The following pumps are categorized as Group B at Cooper Nuclear Station: Pump Number Class Group Type Function CS-P-A 2 B Centrifugal Core Spray CS-P-B 2 B Centrifugal Core Spray HTPCI-P-MP 2 B Centrifugal High Pressure Coolant min HiPCI-P-BP 2 B Centrifugal High Pressure Coolant mIn RCIC-P-MP 2 B Centrifugal Reactor Core Iso Cooling SLC-P-A A B Positive Disp Standby Liquid SLC-P-B A B Positive Disp Standby Liquid The following summarizes the Group A, B, and Comprehensive Pump Test requirements as specified by the ASME OM Code Subsection JSTB. This testing must be performed unless relief for alternate testing has been approved. The design flow rate is defined as the maximum accident flow rate for the pump.Group A Pump Tests -Group A tests are performed quarterly for each pump categorized as A.Reference values are established within +/--20% of pump design flow rate, if practicable. If not practicable, the reference point flow rate shall be established at the highest practical flow rate.For centrifugal pumps, the pump is operated at a nominal motor speed for constant speed drives or at a speed adjusted to the reference point (+/--1%) for variable speed drives. The resistance of the system is varied until the flow rate equals the reference point. Then, differential pressure and vibration measurements are determined and compared to their reference values.For positive displacement pumps, the pump is operated at a nominal motor speed for constant speed drives or at a speed adjusted to the reference point (+/-1%) for variable ,speed drives. The resistance of the system is varied until the discharge pressure equals the reference point. Then, flow and vibration measurements are determined and compared to their reference values.Group B Pump Tests -Group B tests are performed quarterly for each pump categorized as B.Reference values are established within +20% of pump design flow rate, if practicable. If not practicable, the reference point flow rate shall be established at the highest practical flow rate.For centrifugal pumps, the pump is operated at a nominal motor speed for constant speed drives or at a speed adjusted to the reference point (+ 1%) for variable speed drives.Then, the differential pressure or flow rate is determined and compared to its reference value.For positive displacement pumps, the pump is operated at a nominal motor speed for constant speed drives or at a speed adjusted to the reference point (+/--1%) for variable speed drives. Then, the flow rate is determined and compared to its reference value.Revision 0 Page 17 3 Cooper Nuclear Station F~fih Interval Inservice Testing Program for Pumps and Valves Technical Position TP-07 (Page 2 of 3)Comprehensive Pump Tests -Comprehensive pump tests are performed biennially for all pumps in the Inservice Testing Program. Reference values are established within +/-20% of pump design flow rate. The procedure to perform a comprehensive pump test is similar to the Group A test.The following instrument accuracy requirements apply to each test type of test: Parameter Group A Group B Comprehensive Pressure +/- 2.0% +/- 2.0% +/- 0.5%Flow Rate +/- 2.0% +/-/- 2.0% +/- 2.0%Speed +/- 2.0% +/- 2.0% +/- 2.0%Vibration +/- 5.0% +/- 5.0% +/- 5.0%Differential Pressure +/- 2.0% +/- 2.0% +/- 0.5%Revision 0 Pg 7 Page 174 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves Technical Position TP-8 (Page 1 of 2)Vacuum Breaker Testing Purpose The purpose of this Technical Position is to establish the station position for testing vacuum breakers.Applicabilit. This Technical Position is applicable to vacuum breakers that are included in the Inservice Testing Program.Position In accordance with the requirements of ISTC-5230 for vacuum breaker valves, vacuum breakers shall meet the applicable inservice test requirements of ISTC-5220 and Mandatory Appendix I.ISTC-5220 is the requirement for the valve obturator movement of check valves and Mandatory Appendix I is the requirement for the Inservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants. The testing performed at Cooper will meet both requirements. Check valves that are capacity certified and are functioning as a vacuum breaker, will be tested in accordance with Mandatory Appendix I. Per 1-3370, Class 2 and 3 Vacuum Relief Valves shall be actuated to verify open and close capability, set-pressure, and performance of any pressure and position-sensing accessories. Seat tightness shall be in compliance with the owner's seat tightness criteria. The disposition after testing or maintenance shall be in accordance with 1-3470 for class 2 and 3 vacuum relief valves.Per 1-1370, the frequency of testing Class 2 and 3 containment vacuum relief valves shall be at each refueling outage or every 2 years, whichever is sooner, unless historical data requires more frequent testing. This code required frequency recognizes that this testing may be able to be performed online by having the "or every 2 years" criteria and "unless historical data requires more frequent testing." Leak test frequencies are designated by the owner in accordance with Table ISTC-3500-l. Per 1-13 80, those Class 2 and 3 vacuum relief valves, other than Primary Containment Vacuum Relief Valves, shall be tested every 2 years, unless performance data suggest the need for a more appropriate test interval.Mandatory Appendix I requirements contain all the necessary requirements to adequately test 1ST vacuum relief valves. However, in order to satisfy ISTC-5220, as required by JSTC-5230, exercise testing in the open (FSO) and closed (FSC) positions have also been indicated within the Attachment 11 Inservice Testing Valve Table in addition to the VBT designation for the vacuum breaker test. The exercise testing requirements of ISTC-5220 (FSO/FSC) are being met within the testing performed per Mandatory Appendix I. No further testing is required.Revision 0 Pg 7 Page 175 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves Technical Position TP-8 (Page 2 of 2)Justification This position was only established to clearly define the basis for how the vacuum breakers are being tested at CNS and at what frequency. All code requirements are being met. As stated above, the vacuum breaker testing fully encompasses the exercise testing requirements of IST-5220. The required code frequencies for testing are clearly being met.Revision 0 Page 176 Revision 0 Page 176 Cooper Nuclear Station Fifthi Interval Inservice Testing Pro gram for Pumps and Valves ATTACHMENT 10 INSERVICE TESTING PUMP TABLE Revision 0 Page 177 Revision 0 Page 177 Cooper Nuclear Station F/ifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: CORE SPRAY (CS), P&ID No. 2045 Sheet 1 Pump P&IID ISI Pump __Parameters CIC Coor Class Group Q dP V N Notes CS-P-A F-3 2 B Q Q 2Y (1) cs Pump A________ RP-01, RP-05, RP-09, RG-01, TP-07 CS-P-B D-3 2 B Q Q 2Y (1) CS Pump B RP-01, RP-05, RP-07, RP-09, RG-01,____ ___ _ __ ___ ___ __ _ ___ ___ ___ ___TP-07 NOTES: (1) Pump is directly coupled to a constant speed synchronous or induction type driver.SYSTEM: DIESEL GENERATOR FUEL OIL TRANSFER (DGDO), P&1ID No. 2077 and 2011 Sheet 1 Pump P&ID ISI Pump Parameters CIC Coor Class Group Q dP V N Notes DGDO-P- G-3 A NA NA NA NA NA Skid-Mounted, RG-01, TP-05 EDF 1 DGDO-P- G-4 A NA NA NA NA NA Skid-Mounted, RG-Ol, TP-05 EDF 1 DGDO-P- A-8 A A Q Q 6M (1) DGDO Pump A, ARP-04, RG-01, DOTA __TP-07 DGDO-P- A-10 A A Q Q 6M (1) DGDO Pump B, ARP-04, RG-01, DOTB TP -07 NOTES: (1) Pump is directly coupled to a constant speed synchronous or induction type driver.SYSTEM: HIGH PRESSURE COOLANT INJECTION (IIPCI), P&llD No. 2044 Pump P&ID ISI Pump Parameters__ CIC Coor Class -Group Q dP V N Notes HPCI-P-MP E-4 2 B Q Q 2Y -Q -HiPCI Pump Main (1)_________ ____ ____ _________ ___ ___ ___RP-03, RP-05, RP-09, RG-01, TP-07 HPCI-P-BP E-3 2 B Q Q -2Y -Q HPCI Pump Booster(l) _________ ____ ____ _________ ___ ___ ___RP-03, RP-05, RP-09, RG-01, TP-07 NOTES: (1) HPCI main and booster pumps will be tested simultaneously. Revision 0 Page 178 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: RADIOACTIVE WASTE (RW), P&LD No. 2005 Sheet 2 Pump dCI P&ID ISI Coor Class Pump Group Q Parameters JdP [V N INotes RW-P-Z1 G-10 A A Q NA NA (1) Elevated Release Point Sump Pump, RWPZ2 G1O A A (2) N NAARP-01, RG-01, TP-07 RWPZ2 G10 A A Q N NA()Elevated Release Point Sump Pump,___________ ____ ____(2) _____ ARIP-01, RG-O1, TP-07 NOTES: (1) Pump is directly coupled to a constant speed synchronous or induction type driver.(2) The time (TM) to pump a specified quantity of water from the sump will be measured and trended.SYSTEM: REACTOR CORE ISOLATION COOLING (RCIC), P&ID No. 2043 Pump P&ID ISI Pump Parameters CIC Coor Class Group Q dP V N Notes RCIC.-P-MP G-3 2 B Q Q 2Y Q RCJC Pump 1A RP-04, RP-09, RG-01, TP-07 SYSTEM: REACTOR EQUIPMENT COOLING (REC), P&ID No. 2031 Sheet 2 Pump P&LD ISI Pump Parameters CIC Coor Class Group Q dP V N Notes REC-P-A G-l 3 A Q Q Q (1) REC Pump lA, Loop A RP-06. RP-08, RP-09, RG-01, TP-07 REC-P-B G-2 3 A Q Q Q (1) REC Pump 1B, Loop A______________________________RP-06, RP-08, RP-09, RG-01, TP-07 REC-P-C G-3 3 A Q Q Q (1) REC Pump 1C, Loop B RP-06, RP-08, RP-09, RG-01, TP-07 REC-P-D G-3 3 A Q Q Q (1) REC Pump 1D, Loop B_____________________________ __________RP-06, RP-08, RP-09, RG-O1, TP-07 NOTES: (1) Pump is directly coupled to a constant speed synchronous or induction type driver.Revision 0 Page 179 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RHR), P&ll) No. 2040 Pump P&LD ISI Pump Parameters CIC Coor Class Group Q dP V N Notes RHR.-P-A G-4 2 A Q Q Q (1) RHR Pump 1A, Loop A RP-02, RP-08, RP-09, RG-01, TP-07 RHiR-P-B G-9 2 A Q Q Q (1) RHR Pump iB, Loop B RP-02, RP-08, RP-09, RG-Ol, TP-07 RHR-P-C H-4 2 A Q Q Q (1) RHR Pumpl1C, Loop A_____ ______RP-02, RP-08, RP-09, RG-O1, TP-07 RHR-P-D H-9 2 A Q Q Q (1) RHR Pump 1D, Loop B_______________RP-02, RP-08, RP-09, RG-Ol, TP-07 NOTES: (1) Pump is directly coupled to a constant speed synchronous or induction type driver.SYSTEM: SERVICE WATER (SW), P&1D No. 2006 Sheets 1 and 4'Pump P&ID 1S1 Pump Parameters CIC Coor Class Group Q dP V N Notes SW-P-A B3-10 3 A Q Q Q (1) sw Pump lA, RP-08, RP-09, RG-01, TP-07 SW-P-B B3-9 3 A Q Q Q (1) sw Pump 11B, RP-08, RP-09, RG-01,______TP-07 SW-P-C B3-8 3 A Q Q Q (1) SW Pump 1C, RP-08, RP-09, RG-01, TP-07 SW-P-D 13-7 3 A Q Q Q (1) sw Pump 1D, RP-08, RP-09, RG-01, TP-07 SW-P-BPA F-7 3 A Q Q Q (1) SW Booster Pump 1A RP-05, RP-09, RG-01, TP-07 SW-P-BPB C-7 3 A Q Q Q (1) sw Booster Pump lB RP-05, RP-09, RG-0l, TP-07 SW-P-BPC E-7 3 A Q Q Q (1) SW Booster Pump IC RP-05, RP-09, RG-01, TP-07 SW-P-BPD A-7 3 A Q Q Q (1) SW Booster Pump 1D____________RP-05, RP-09, RG-01, TP-07 NOTES: (1) Pump is directly coupled to a constant speed synchronous or induction type driver.Revision 0 Page 180 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: STANDBY LIQUID CONTROL (SLC), P&IID No. 2045 Sheet 2 Pump P&ID ISI Pump ___Parameters __CIC Coor Class Group Q Pd V N Notes SLC-P-A E-10 A B Q Q 6M (1) SLC Pump lA_________ ____ _____ _____ ___ ___ ARP-02, ARP-03, RG-01, TP-07 SLC-P-B F-10 A B Q -Q -6M (1) SLC Pump 1A_________ ____ _____ _____ ___ ___ ___ ARP-02, ARP-03, RG-01, TP-07 NOTES: (1) Positive displacement pump Revision (9 Page 181 Revision 0 Page 181 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves ATTACHMENT 11 INSERVICE TESTING VALVE TABLE Revision 0 Page 182 Revision 0 Page 182 Cooper Nuclear Station Fifth Interval lnservice Testing Pro gram for Pumps and Valves SYSTEM: CONTROL ROD DRIVE (CR]))VALVE CIC P&IDP&ID ISI IST 1VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR _CLASS CAT SIZE TYPE TYPE POS RQMT FREQ CRD-AOV-CV32A 2039 J2 2 B 1 GB AO 0 FSC Q SOUTH SDIV INBOARD VENT FST Q ISOLATION VALVE, RG-01, TP-04_________ ____ ___ __ _____PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _CRD-AOV-CV32B 2039 J4 2 B 1 GB AO 0 FSC Q NORTH SDIV INBOARD VENT FST Q ISOLATION VALVE, RG-01, TP-04_________ ____ ___ __ _____PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _CRD-AOV-CV33' 2039 H3 2 B 2 GB AO 0 FSC Q SOUTH SDIV INBOARD DRAIN FST Q ISOLATION VALVE, RG-01, TP-04______________ ______PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _CRD-AOV-CV34 2039 H4 2 B 2 GB AO 0 FSC Q NORTH SDIV INBOARD DRAIN FST Q ISOLATION VALVE, RG-01, TP-04_________ ____ ___ __ _____PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _CRD-AOV-CV35 2039 H3 2 B 2 GB AO 0 FSC Q SOUTH SDIV OUTBOARD DRAIN FST Q ISOLATION VALVE, RG-01, TP-04_________ ____ ____ _ ______PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _CRD-AO V-C V36 -2039 H4 2 B 2 GB AO 0 FSC Q NORTH SDIV OUTBOARD DRAIN FST Q ISOLATION VALVE, RG-01, TP-04"_________ ____ ___ __ _____PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _CRD-AOV-CV38A 2039 J3 2 B 1 GB AO 0 FSC Q SOUTH SD1V OUTBOARD VENT FST Q ISOLATION VALVE, RG-01, TP-04_________ ____ ____ _ ______PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _FSC Q CRD-AOV-CV38B 2039 J4 2 B 1 GB AO 0 FST Q NORTH SDIV OUTBOARD VENT______________ ___ ____ ___ ____ ____ ________ PIT 2Y ISOLATION VALVE, RG-01, TP-04 Revision 0 Page 183 Revision 0 Page 183 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: CONTROL ROD DRIVE (CRD)VALVE GIG P&ID P&ID IS I5 ST 1VALVE VALVE ACT 1NORM TEST 1TEST NOTES/DESCRIPTION COOR _CLASS CATj SIZE TYPE TYPE jPOS RQMT FREQ CRD-AOV-CV126 2039 C10 B 1 DIA AO C FSO TS SCRAM INLET, (TYP. OF 137) ________FST TS RV-04, RG-01*CRD-AOV-CV127 2039 B10 B 3/4 DIA AO C FSO TS SCRAM OUTLET, (TYP. OF 137) ____FST TS RV-04, RG-01 CRD-CV-13CV 2039 A9 1 A/C 3/4 CK-P SA 0 LJ-1 OPB CRD WATER TO REACTOR CCL CVCM RECIRCULATION PUMP A, COF CVCM RG-01, TP-01, TP-06 CRD-CV-14CV 2039 B9 1 A/C 3/4 CK-P SA 0 LJ-1 OPB CRD WATER TO REACTOR CCL CVCM RECIRCULATION PUMP A,_________ ____ ___COP CVCM RG-0i, TP-01, TP-06 CRD-CV-15CV 2039 A8 1 A/C 3/4 CK-P SA 0 LJ-1 OPB CRD WATER TO REACTOR*CCL CVCM RECIRCULATION PUMP B,_________ ____ ___COF CVCM RG-01, TP-01, TP-06 CRD-CV-16CV 2039 B8 1 A/C 3/4 CK-P SA 0 LJ-1 OPB CRD WATER TO REACTOR CCL CVCM RECIRCULATION PUMP B,__________COF CVCM RG-01, TP-01, TP-06 CRD-CV-25CV 2039 B4 A A/C 1 1/2A CK-S SA 0 LT-2 RF CRD SYSTEM ISOLATION CHECK FSC RF VALVE, RG-01, TP-01, ROJ-12___ ___ __ ___ __ __ ___ _ ___ _ _FSO5 RF CRD-CV-26CV 2039 B4 A A/C 1 1/22 CK-S SA 0 LT-2 RF CRD SYSTEM ISOLATION CHECK FSC RE VALVE, RG-01, TP-01, ROJ-12_____ ____ ___ ____ ____ ___ __ _ ___ ____ ___ __ _ ___ ____ FSO RF _ _ _ _ _ _ _ _ _ _ _Revision 0 Page 184 Cooper Nuclear Station Ffith Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: CONTROL ROD DRIVE (CRD)VALVE GIGC P&ID {P&ID ISI{ 1ST VALVE VALVE 1ACT[ NORM ITEST[ TEST NOTES/DESCRIPTION CORCASCAT SIZE TYETYPE POSRQMT[FE CRD-CV-28CV 2039 H2 A C 1/2A CK-P SA C SKID Q SOUTH SDIV DRAIN VALVE AIR__________SUPPLY BYPASS CV, RG-01, TP-05 CRD-CV-29CV 2039 H5 A C 1/2A CK-P SA C SKID Q NORTH SDIV DRAIN VALVE AIR____ _____ _____SUPPLY BYPASS CV, RG-01, TP-05 CRD-CV-30CV 2039 H3 A C 1/2A CK-P SA C SKID Q SOUTH SDIV VENT V AIR____ ____ _________ ______SUPPLY BYPASS CV, RG-01, TP-05 CRD-CV-31CV 2039 H4 A C 1/2A CK-P SA C SKID Q NORTH SDV VENT VAIR SUPPLY____ ____ _____BYPASS CV, RG-01, TP-05 CRD-CV-32CV 2039 H3 A C 1/2A CK-P SA C SKID Q SOUTH SDIV DRAIN VALVE AIR____ ____ ______SUPPLY BYPASS CV, RG-01, TP-05 CRD-CV-33CV 2039 H5 A C 1/2A CK-P SA C SKID Q NORTH SDIV DRAIN VALVE AIR____ ____ ______SUPPLY BYPASS CV, RG-01, TP-05 CRD-CV-34CV 2039 J3 A C 1/2A CK-P SA C SKID Q SOUTH SDV VENT V AIR SUPPLY_____ _____BYPASS CV, RG-01, TP-05 CRD-CV-35CV 2039 J4 A C 1/2A CK-P SA C SKID Q NORTH SDV VENT V AIR SUPPLY____ ____ ____ ______BYPASS CV, RG-01, TP-05 CRD-CV-CV114 2039 Bl 21 C 3/4 CK-B SA 0 FSO TS SCRAM OUTLET CHECK VALVE, (TYP. OF 137) ____ ___RV-04, RG-01, TP-05 CRD-CV-CV1 15 2039 D9 2 C 1/2A CK-B SA 0 FSC RF SCRAM INLET CHECK VALVE, (TYP. OF 137) ____________FSO RjF RG-01, TP-01, ROJ-06 Revision 0 Page 185 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: CONTROL ROD DRIVE (CRD)VALVE GIG P&1ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ CRD-CV-138 2039 C9 2 C 1/22 CK-B SA 0 FSC TS COOLING/SCRAM HEADER (TYP. OF 137) CHECK VALVE, RV-04, RG-01, TP-_________________ ______ _____ ______ __ ___ ___ 05 CRD-SOV-SO120 2039 C9 2 B 114 SOV SO C FSC TS CRD WITHDRAWAL EXHAUST (TYP. OF 137) F___ __ _______ ST TS VALVE, RV-04, RG-01, TP-05 CRD-SOV-SO121 2039 B10 B 1/4 SOV SO C FSC TS CRD INSERT EXHAUST (TYP. OF 137) _______ ___ __ ___ ___FST TS VALVE, RV-.04, RG-01, TP-05 CRD-SOV-SO 122 2039 B10 2 B 1/4 SOV SO C FSC TS CRD WITHDRAWAL VALVE, (TYP. OF 137) ____ ___ _______ ____FST TS RV-04, RG-01, TP-05 CRD-SOV-SO0123 2039 B10 2 B 1/4 SOV SO C FSC TS CRD INSERT VALVE, (TYP. OF 137) ____ _______ _______ ___ _______ FST TS RV-04, RG-01, TP-05 Revision 0 Page 186 Revision 0 Page 186 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: CORE SPRAY (CS){P&ID ISI IST VALVE VALVE fACT NORM TEST TEST VALVE CIC P& COO CLASS CAT SIZE TYPE [TYPE POS -RQMT FREQ NOTES/DESCRIPTION CS-CV-10CV 2045 F5 2 C 12 CK-S SA C FSO Q CS PUMP A DISCHARGE CHECK SH 1 ______FSC Q R.G-01, TP-01 CS-CV-11CV 2045 C5 C 12 CK-S SA C FSO Q CS PUMP B DISCHARGE CHECK SH 1 ___ __ __ FSC Q RG-01, TP-01 CS-CV-12CV 2045 G7 A C 2 CK-P SA C COF CVCM CS LOOP A OUTBOARD PRESSURE SH 1 CCF CVCM MAINTENANCE SUPPLY, RG-01, TP-_______01 CS-CV-13CV 2045 G7 2 C 2 CK-P SA C COF CVCM CS LOOP A INBOARD PRESSURE SH 1 CCF CVCM MAINTENANCE SUPPLY, RG-01, TP-CCD/ CVCM 01______ _____ ~~~~CCR _ _ _ _ _ _ _ _ _ _ _ _ _CS-CV-14CV 2045 D7 A C 2 CK-P SA C COF CVCM CS LOOP B OUTBOARD PRESSURE SHi1 CCF CVCM MAINTENANCE SUPPLY, RG-01, TP-______ _______01 CS-CV-15CV 2045 D7 2 C 2 CK-P SA C COF CVCM CS LOOP B INBOARD PRESSURE SH 1 CCF CVCM MAINTENANCE SUPPLY, RG-0 1, TP-CCD/ CVCM 01 CCR__ _ _ _ _ _ _ _ _ _ _ _CS-CV-18CV 205 D0 1 AC 10 CK-S SA C LT-2 OPB CS SYSTEM A TESTABLE CHECK SH1FSC RF ROJ-11, RV-05, RG-01 FSO RF PIT RE Revision 0 Page 187 Cooper Nuclear Station Ffifti Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: CORE SPRAY (CS)1P&ID ISI IST VALVE VALVE TACT NORM TEST TEST VALVE CIC P&ID COOR CLASS [CAT SIZE TYPE _TYPE P0S RQMT FREQ NOTES/DESCRIPTION CS-CV- 19CV 2045 Bl 10 A/C 10 CK-S SA C LT-2 OPB CS SYSTEM B TESTABLE CHECK SHi1 FSC RF ROJ-11, RV-05, RG-01 FSO RE_____PIT RF CS-MOV-MO5A 2045 E5 2 B 3 GT MO 0 FSO Q CS PUMP A MINIMUM FLOW SH 1 FSC Q RECIRCULATION ISOLATION, RG-____PIT 2Y 01 CS-MOV-MO5B 2045 B5 2 B 3 GT MO 0 FSO Q CS PUMP BMINIMUM FLOW SHi1 FSC Q RECIRCULATION ISOLATION, RG-____PIT 2Y 01 CS-MOV-MO7A 2045 F2 2 B 14 GT MO 0 FSO Q CS PUMP A SUCTION, RG-01 SHi1 FSC Q_____ _____ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _CS-MOV-MO7B 2045 C2 2 B 14 GT MO 0 FSO Q cs PUMP B SUCTION, RG-01 SHi1 FSC Q______________ _____ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _CS-MOV-MO11lA 2045 F8 2 B 10 GT MO 0 PIT 2Y LOOP A INJECTION THROTITLE, RG-__ _ _ _ _ _ _ SH 1_ __ 01 CS-MOV-MO11B 2045 C8 2 B 10 GT MO 0 PIT 2Y LOOP B INJECTION THROTTPLE, RG-__ _ _ _ _ _ _ SH 1 __ _ _ _ _ _ _ _ _ _ _ _ _01 CS-MOV-MO 12A 24 FS 1 A 10 GT MO C LJ-l1 OPB LOOP A INJECTION BLOCK SiLT-2 OPB CSJ-09, RV-05, RG-01 FSO CS Rev&ion 0 Page 188 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: CORE SPRAY (CS)P&ID ISI IST VALVE VALVE ACT 1NORM TEST TTEST VALVE CIC P&ID COOR [C LASS CAT SIZE TYPE TYPE POS RQMT jFREQ NOTES/DESCRIPTION FSC CS_____________ _____ _____ PIT 2Y_ _ _ _ _ _ _ _ _ _ _ _ _CS-MO V-MO 12B 2045 C8 1 A 10 GT MO C LJ-1 OPB LOOP B INJECTION BLOCK Sill LT-2 OPB CSJ-09, RV-05, RG-01 FS0 CS FSC CS___ __ __ __ _ __ _ __ __PIT 2Y CS-MOV-MO26A 2045 F7 2 B 10 GB MO C PIT 2Y CS PUMP A TEST LINE PASSIVE__ _ _ _ _ _ _ Sil 1 _ _ _ _ __ _ __ _ _ ISOLATION, RG-01 CS-MOV-MO26B 2045 C7 2 B 10 GB MO C PIT 2Y CS PUMP B TEST LINE PASSIVE_ _ _ _ _ _ _ SI-I 1 __ _ ___ISOLATION, RG-01 CS-RV-10RV 2045 H3 2 C 3/4 RV SA C RVT APP I CS PUMP A SUCTION RELIEF, RG-__ _ _ _ _ _ _ Sil 1 _ _ _ _ __ _ _ 01 CS-RV-11RV 2045 F6 2 C 2 RV SA C RVT APP I CS PUMP A DISCHARGE RELIEF,_ _ _ _ _ _ Sil 1__ RG-01 CS-RV-12RV 2045 E3 2 C 3/4 RV SA C RVT APP I CS PUMP B SUCTION RELIEF, RG-01___________ Sil 1 _ _ _ _ __ _ _ _ _ _ _ _ _ _CS-RV-13RV 2045 C6 2 C 2 RV SA C RVT APP I CS PUMP B DISCHARGE RELIEF, Sill RG-01 Revision 0 Pg 8 Page 189 Cooper Nuclear Station F~ith Interval lnservice Testing Program for Pumps and Valves SYSTEM: DEMINERALIZED WATER SYSTEM (DW)VALVE CIC P&ID 1P&ID ISI IST 1VALVE IVALVE ACT INORM TEST JTEST f NOTES/DESCRIPTION COOR CLASS CAT jSIZEJTYPE TYPE POS RQMT FREQ DW-V-133 2029 G8 2 A 4 GT MA C LJ-1 OPB PASSIVE DRYWELL OUTBOARD____ ___ ___ __ __ ___ __ ___ ___ ___ _ _ ___ ___ _ _ ___ ___ _ _ ___ SUPPLY VALVE DW-V-219 2029 G8 2 A 4 GT MA C L -i OPB PASSIVE DRYWELL INBOARD SUPPLY VALVE Revision 0 Page 190 Revision 0 Page 190 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: DIESEL GENERATOR DIESEL OIL (DGDO)VALVE CIC P&ID P&ID ISI IST VALVE VALVE IACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ_____________ DGDO-CV-°10CV 2011 A8 A C 2 CK-P SA C FSO Q DGDO TRANSFER PUMP A SH 1______ __ __ DISCHARGE, ARV-03, RG-01 DGDO-CV-1 1CV 2011 A10 A C 2 CK-P SA C FSO Q DGDO TRANSFER PUMP B SHi 1______ __ __ DISCHARGE, ARV-03, RG-01 DGDO-CV-12CV 2011 A4 A C- 2 CK-P SA C FSO Q DGDO DAY TANK 1 INLET, ARV-03,_ _ _ _ _ _ SHi 1_ _ _ _ _ _ _ _ _ _ RG-01 DGDO-CV-13CV 2011 B4 A C 2 CK-P SA C FSO Q DGDO DAY TANK 2 INLET, ARV-03, SHi 1 _ _ _ _ _ _ _ _ _ _ __ _ RG-01 DGDO-CV-14CV 2077 J3 A C 1 1/4 CK-S SA C SKID Q ENG DR FUEL P 1 SUPPLY, RG-01,____ ___ ___ ___ __ ___ ____ ___ ___ __ ___ ___TP-05 DGDO-CV-15CV 2077 J4 A C 1 1/4 CK-S SA C SKID Q ENG DR FUEL P 2 SUPPLY, RG-01,_______ ____ _ _____ ______ TP-05 DGDO-CV-16CV 2077 H3 A C 1 1/4 CK-S SA C SKID Q FUEL BOOSTER P 1 DISCH, RG-01,____ _ ______ ___ ___TP-05 DGDO-CV-17CV 2077 H4 A C 1 1/4 CK-S SA C SKID Q FUEL BOOSTER P 2 DISCH, RG-01,____ ___ ___ ____ _____ __ ____ _____ ___ ___ ___ ___TP-05 DGDO-PRV-PRV101 2077 E3 A C 3/4 PRV SA C SKID Q DG1 FUEL INJECT HDR PRESS REG____ __ ______ _ _ ___ __ ___V, RG-01, TP-05 DGDO-PRV-PRV102 2077 E4 A C 3/4 PRV SA C SKID Q DG2 FUEL INJECT HDR PRESS REG____ ___ ___ __ ___ _ _ ___ ___ ___ ___ _ _ ___ _ _ ___ ___ __ ___ _ _ ___ __ ___ V, RG-01, TP-05 Revision 0 Page 191 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: DIESEL GENERATOR DIESEL OIL (DGDO)VALVE CIC P&ID 1P&ID ISI IST 1VALVE TVALVE ACT INORM TEST TEST [ NOTES/DESCRIPTION COOR CLASS CATj SIZE TYPE TYPE POS RQMT FREQ DGDO-RV-10RV 2077 H2 A C 3/4 RV SA C RVT APP I DGDO PUMP 1 SUCTION RELIEF,_____ARV-02, RG-01 DGDO-RV-1 1RV 2077 H4 A C 3/4 RV SA C RVT APP I DGDO PUMP 2 SUCTION RELIEF,____ ___ ___ _ _ ___ ___ _ _ ___ARV-02, RG-01 DGDO-SOV-2077 112 A B 3/4 SOV SO 0/C FSO Q DGDO DAY TANK 1 INLET FUEL SSV5028 ________________SAFETY VALVE, ARV-Ol, RG-01 DGDO-SOV-2077 116 A B 3/4 SOV SO 0/C FSO Q DGDO DAY TANK 2 INLET FUEL SSV5029 SAFETY VALVE, ARV-01l, RG-01 Revision 0 Pg 9 Page 192 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: DIESEL GENERATOR STARTING AIR (DGSA)VALVE CIC P&ID P&ID 1SI IST VALVE IVALVE ACT INORM TEST ITEST NOTES/DESCRIPTION COO CLAS CAT SIZE ITYPE TYPE POS RQMT IFREQ_____ __ ~R ___S_ _ I _ _ I _ _ _ _ _DGSA-AOV-AV5 117.10 E3 A B 2 1/2A GB AO C SKID Q DG-l LEFT BANK STARTING AIR-IC-09 _______________VALVE, RG-01, TP-05 DGSA-AOV-AV6 117.10 E3 A B 2 1/2A GB AO C SKID Q DG-2 LEFT BANK STARTING AIR-IC-09 ____ ___ _________VALVE, RG-01, TP-05 DGSA-AOV-AV7 117.10 El A B 2 1/2A GB AO C SKID Q DG-1 RIGHT BANK STARTING-IC=09 ________AIR VALVE, RG-01, TP-05 DGSA-AOV-AV8 117.10 El A B 2 1/2A GB AO C SKID Q DG-2 RIGHT BANK STARTING-IC-09 ____AIR VALVE, RG-01, TP-05 DGSA-CV-10CV 2077 D7 A C 2 CK-L- SA C FSC Q STARTING AIR COMPRESSOR 1A_______ ______FSO Q DISCHARGE, RG-01, TP-01 DGSA-CV-11CV 2077 D9 A C 2 CK-L SA C FSC Q STARTING AIR COMPRESSOR lB____ ______ ___ ___FSO Q DISCHARGE, RG-01, TP-01 DGSA-CV-12CV 2077 D10 A C 2 CK-L SA C FSC Q STARTING AIR COMvPRESSOR 2B__________F__ SO Q DISCHARGE, RG-01, TP-01 DGSA-CV-13CV 2077 D12 A C 2 CK-L SA C FSC Q STARTING AIR COMPRESSOR 2A F___ SO Q DISCHARGE, RG-01, TP-01 DGSA-CV-14CV 2077 C8 A C 2 CK-S SA C CCL CVCM AIR RECEIVER IA INLET, RG-.01,_______ __ ___ ___COF CVCM TP-01 DGSA-CV-15CV 2077 C9 A C 2 CK-S SA C CCL CVCM AIR RECEIVER lB INLET, RG-01, COF CVCM TP-01 Revision 0 Page 193 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: DIESEL GENERATOR STARTING AIR (DGSA)VALVE CIC IP&ID IP&ID 1SI IST 1VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION _______j jCOO CLAS CAT SIZE TYPE TYPE POS RQMT FREQ R S DGSA-CV- 16CV 2077 Cl10 A C 2 CK-S SA C CCL CVCM AIR RECEIVER 2A INLET, RG-01, COF CVCM TP-01 DGSA-CV-17CV 2077 C11 A C 2 CK-S SA C CCL CVCM AIR RECEIVER 2B INLET, RG-01, COF CVCM TP-01 DGSA-CV-18CV 2077 B7 A C 3 CK-S SA C FSO Q AIR RECEIVER 1A OUTLET, RG-01 FSC Q DGSA-CV-19CV 2077 B8 A C 3 CK-S SA C FSO Q AIR RECEIVER lB OUTLET, RG-01 FSC Q DGSA-CV-20CV 2077 B 11 A C 3 CK-S SA C FSO Q AIR RECEIVER 2A OUTLET, RG-01 FSC Q DGSA-CV-21CV 2077 B12 A C 3 CK-S SA C FSO Q AIR RECEIVER 2B OUTLET, RG-01 FSC Q DGSA-CV-30CV 117.10-Ic-D3 A C 1/4 CK-S SA 0 SKID Q DG1 125 PSIG STRT AIR SUPPLY 09 SHUTTLE V, RG-0l, TP-05 DGSA-CV-31CV 117.10-IC-D3 A C 1/4 CK-S SA 0 SKID Q DG2 125 PSIG STRT AIR SUPPLY 09 SHUTTLE V, RG-01, TP-05 DGSA-CV-32CV 117.10-IC-F5 A C 1/4 CK-S SA C SKID Q DG1 LEFT BANK AIR START 09 SHUTTLE VALVE, RG-01, TP-05 DGSA-CV-33CV 1 17.10-ic-F1 A C 1/4 CK-S SA C SKID Q DG1 RIGHT BANK AIR START 09 SHUTTLE VALVE, RG-01, TP-05 Revision 0 Page 194 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: DIESEL GENERATOR STARTING AIR (DGSA)VALVE CIC P&ID P&ID IISI IST VALVE VALVE IACT NORM TEST TEST I NOTES/DESCRIPTION COO jCLAS CAT SIZE TYPE ITYPE POS RQMT FREQ_ _ _ _ _ _ _ _ _ _ R j S _ _ _ _ _ _ _ _ _ _ _DGSA-CV-34CV 117.10-Ic-F5 A C 1/4 CK-S SA C SKID Q DG2 LEFT BANK AIR START o9___SHUTTLE VALVE, RG-01, TP-05 DGSA-CV-35CV 117.10-IC-F1 A C 114 CK-S SA C SKID Q DG2 RIGHT BANK AIR START 09 SHUTTLE VALVE, RG-01, TP-05 DGSA-SOV-SPV1 117.10 F4 A B 1/2A GB SO C SKID Q SOLENOID PILOT VALVE FOR-IC-09 DGSA-AOV-AV5 (DG1 LEFT_____ ______BANK), RG-01, TP-05 DGSA-SOV-SPV2 117.10 F4 A B 1/2A GB SO C SKID Q SOLENOID PILOT VALVE FOR-IC-O9 DGSA-AOV-AV6 (DG2 LEFT______________ _____BANK), RG-01, TP-05 DGSA-SOV-SPV3 117.10 Fl A B 1/2A GB SO C SKID Q SOLENOID PILOT VALVE FOR-IC-O9 DGSA-AOV-AV7 (DG1 RIGHT___________________ ____ ____________BANK), RG-01, TP-05 DGSA-SOV-SPV4 117.10 Fl A B 1/2A GB SO C SKID Q SOLENOID PILOT VALVE FOR-IC-O9 DGSA-AOV-AV8 (DG2 RIGHT_____ ___________BANK), RG-0I, TP-05 DGSA-RV-14RV 2077 C7 A RV 1 RV SA C RVT APP I DGSA AIR RECEIVER 1A RELIEF,____ ___ ___ __ _ ___RG-01 DGSA-RV-15RV 2077 C9 A RV 1 RV SA C RVT APP I DGSA AIR RECEIVER lB RELIEF,_____RG-01 DGSA-RV-16RV 2077 Cl V 1 RV SA C RVT APP I DGSA AIR RECEIVER 1C RELIEF, Revision 0 Page 195 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: DIESEL GENERATOR STARTING AIR (DGSA)VALVE CIC P&ID P&ID IS 1 I ST VALVE VALVE ACT NORM TEST TEST J NOTES/DESCRIPTION COO CLAS CAT SIZE TYPE }TYPE POS RQMT FREQ____ ___ ___ _ __ ___ ___ _ IRG-01 DGSA-RV-17RV 2077 C12 A RV 1 RV SA C RVT APP I -DGSA AIR RECEIVER 1D RELIEF,____ ___ ___ _ _ ___ __ __ ___ ___ __ ______ _j ___ __ ___ RG-01 Revision 0 Page 196 Revision 0 Page 196 Cooper Nuclear Station Fifth Interval Inserviee Testing Pro gram for Pumps and Valves SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPC1)1P&ID IISI IST VALVE VALVE IACT NORM TEST TEST VALVE CIC P&ID COOR _CLASS CAT SIZE TYPEJTYPE POS RQMT FREQ [ NOTES/DESCRIPTION HPCI-AOV-AO42 2041 J6 2 B 1 GB AO 0 FSC Q STEAM LINE DRIPLEG DRAIN, FST Q RG-01, TP-04___ ___ __ ___ __ _____ __ PIT 2Y HPCI-AOV-AO70 2044 E9 2 A 1 BAL AO C LJ-1 OPB HPCI EXHAUST BOOTLEG DRAIN FSC Q INBOARD, RG-01, TP-04 FST Q_______________ ______ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _HPCI-AOV-AO71 2044 E9 2 A 1 BAL AO C LI-i OPB HPCI EXHAUST BOOTLEG DRAIN FSC Q OUTBOARD, RG-01, TP-04 FST Q_____ _____ __ __ __ __ __ __ ___ _ ___ _ ____ _____ PIT 2Y_ _ _ _ _ _ _ _ _ _ _ _ _HPCI-AOV-PCV50 2044 H3 2 B 2 GB AO 0 PS0 Q I-IPCI AUXILIARY COOLING FSO CS SUPPLY PRESSURE CONTROL,________FST CS CSJ-02, RG-01, TP-04 I-PCI-CV-10CV 2044 13 2 C 16 CK-S SA C COF CVCM ECST SUPPLY TO HPCI PUMP,___________CCL CVCM RG-01 HPCI-CV-1 1CV 2044 1110 2 C 16 CK-S SA C COD CVCM HPCI PUMP SUCTION FROM___ ______CCD CVCM SUPPRESSION POOL, RG-0 1 HPCI-CV-13CV 2044 15 2 C 4 CK-S SA C COF CVCM HPCI AUXILIARY COOLING_________ ___ ___CCD CVCM RETURN, RG-01 Revision 0 Page 197 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)IP&ID ISI IST VALVE VLE AT NORM TEST TEST VALVE CIC P&ID COOR CLASS CAT SIZE TYPE TYP POS RQMT FREQ NOTES/DESCRIPTION HPCI-CV-14CV 2044 H5 2 C 2 CK-P SA C COF CVCM HPCI CONDENSATE PUMP CCF CVCM DISCHARGE TO AUXILIARY____ ___ ____ ___ _ __ __ ____ __ __ _ ___COOLING RETURN, RG-01, TP-01 HPCI-CV-15CV 2044 D8 2 A/C 20 CK-L SA C LJ-i OPB HPCI TURBINE EXHAUST, RG-01 COF CVCM____ ____CCL CVCM HPCI-CV-16CV 2044 F9 2 C 2 CK-P SA C COD CVCM HPCI TURBINE EXHAUST DRAIN CCL CVCM TO SUPPRESSION POOL, RG-01,____ ___ ____ ___ ____ _ _ ____ __ ___TP-01 HPCI-CV-17CV 2044 C6 2 C 4 CK-S SA C COF CVCM HPCI PUMP MINIMUM FLOW LINE,___ __ ___ __ __ __ __ _____ __ ____ CCL CVCM RG-01 HPCI-CV-18CV 2044 B8 2 C 2 CK-P SA C COF CVCM CONDENSATE SUPPLY TO HPCI CCF CVCM SYSTEM, RG-01, TP-01 CCD/ CVCM______ _____CCR _ _ _ _ _ _ _ _ _ _ _ _HPCI-CV-19CV 2044 B8 A C 2 CK-P SA C COF CVCM CONDENSATE SUPPLY TO HPCI__________ _____________ ____CCF CVCM SYSTEM, RG-01, TP-01 HPCI-CV-24CV 2044 Eli1 C 3 CK-S SA C FSO SD HPCI VACUUM BREAKER, RG-01,__ __ __ __ __ _ _________ __ __ _ __ _ __ __ FSC SD TP-02 HPCI-CV-25CV 2044 El0 C 3 CK-S SA C FSO SD HPCI VACUUM BREAKER, RG-01,__ __ __ __ __ _ ____ _____ __ __ _FSC SD TP-02 HPCI-CV-26CV 204 Ei 2 C 3 CK-S SA C FSO SD HPCI VACUUM BREAKER, RG-01, FSC SD TP-02 Revision 0 Page 198 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)[IP&ID 1ISI IST VALVE [VALVE 1AC T NORM TEST TEST1 VALVE CIC _P&ID COOR jCLASS CAT SIZE TYPE jTYPE jPOS jRQMT FREQ NOTES/DESCRIPTION HPCI-CV-27CV 2044 El0 2 C 3 CK-S SA C ESO SD HPCI VACUUM BREAKER, RG-01,__________ ____ ____FSC SD TP-02 HPCI-CV-29CV 2044 B9 .1 A/C 14 CK-S SA C LJ-1 RF INJECTION CHECK VALVE, ROJ-09, FSO RE RG-01 FSC RE___PIT 2Y HPCI-HOV-HOV10 2041 H5 2 B 10 STOP HO C SKID Q TU STOP V, RG-01, TP-05 HPCI-MOV-MO014 2041 H5 2 B 10 GT MO C FSO Q STEAM SUPPLY TO TURBINE, RG-_______PIT 2Y 01 HPCI-MOV-MO015 2041 D5 1 A 10 GT MO 0 J-i1 OPB STEAM SUPPLY INBOARD FSC Q/CS ISOLATION, CSJ-10, RG-01_____ _____ ___ _ ______ ______PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _HPCI-MOV-MO16 2041 D4 1 A 10 GT MO 0 LJ-i OPB STEAM SUPPLY OUTBOARD FSC Q ISOLATION, RG-01____PIT 2Y HPCI-MOV-MO17 2044 J2 2 B 16 GT MO 0 FSO Q PUMP SUCTION FROM FSC Q EMERGENCY CONDENSATE __________PIT 2Y STORAGE TANK, RG-01 HPCI-MOV-MO19 2044 B8 2 B 14 GT MO C FSO Q HPCI INJECTION, RG-01__ __ __ ____ __PIT 2Y HPCI-MOV-MO20 2044 B7 2 B 14 GT MO 0 PIT 2Y HPCI PUMP DISCHARGE, RG-01 Revision 0 Page 199 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: HIGH PRESSURE COOLANT INJECTION (IIPC1)P&ID SI 15 IST VALVE VALVE IACT NORM TEST TEST VALVE CIC [P&IDJCOOR _CLASS CAT SIZE TYP__JTYPE P0OS RQMT FREQ NOTES/DESCRIPTION HPCI-MOV-MO21 2044 D2 2 B 10 GB MO C PIT 2Y HPCI PUMP TEST BYPASS TO EMERGENCY CONDENSATE ____ _____ ___ _____ ____ _____STORAGE TANK, PASSIVE, RG-01 HPCI-MOV-MO24 2044 D2 A B 10 GT MO C PIT 2Y HPCI PUMP TEST BYPASS REDUNDANT SHUTOFF, PASSIVE,_____RG-01 HPCI-MOV-MO25 2044 C7 2 B 4 GB MO C FSO Q HPCI PUMP MINIMUM FLOW FSC Q BYPASS LINE ISOLATION, RG-01_____________ _____ ______PIT 2Y_ _ _ _ _ _ _ _ _ _ _ _ _HPCI-MOV-MO58 2044 G10 2 B 16 GT MO C FSO Q HPCI PUMP SUCTION FROM FSC Q SUPPRESSION POOL, RG-01_____ ~~~~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _HPCI-RD-S241 2044 D5 2 D 16 RD SA C RD 5Y EXHAUST LINE RUPTURE DISK,____ ____ ___ ____ __ __ _ ___RG-01 HPCI-RV-10RV 2044 J2 2 C 1 RV SA C RVT APP I HPCI PUMP SUCTION RELIEF, RG-____ ___ ___ ___ __ ___ __ _ ___ __ ___ ___ ___ ___01 HPCI-RV-12RV 2044 H4 2 C 1 RV SA C RVT APP I HPCI AUXILIARY COOLING__________ ____ ____WATER SUPPLY, RG-01 HIPCI-SOV-SSV64 2044 G5 2 B 1 SOV SOV C FSC Q/PB HPCI EXHAUST DRIP LEG DRAIN,_____________ _ ____FST Q RV-01, RG-01, TP.-04 HPCI-SOV-SSV87 2044 G4 2 B 1 SOV SOV C FSC Q/PB HPCI TURBINE DRIP LEG DRAIN,__ _ __ _ __ _ ___ ____ _ __ _ __ _ __ _ ____ _F_ ST Q RG-01, RV-01, TP-04 Revision 0 Page 200 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: HIGH PRESSURE COOLANT INJECTION (HPCI)IP&ID ISI I ST VALVE VALVE ACT NORM TEST 1TEST VALVE CIC P&IDj COOR CLASS jCAT _SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION HPCI-TU-VGR 2041 H5 A B 9 CTL HO C SKID Q HPCI TU V GEAR ASSY W / PILOT,____ ___ ____ ___ ___ __ _ ___ _ __ ___ _ _ ____ ___ ___TP-05 HPCI-V-44 2044 D8 2 A/C 20 S-CK MA 0 LU-i OPB HPCI TURBINE EXHAUST TO COF CVCM SUPPRESSION POOL ISOLATION,__________ ___ ____CCL CVCM RG-01 HPCI-V-50 2044 F9 2 C 2 S-CK MA 0 COD CVCM TURBINE DRAIN TO SUPPRESSION CCD CVCM POOL ISOLATION, RG-.01, TP-01 Revision 0 Page 201 Revision 0 Page 201 Cooper Nuclear Station Ffith Interval Inservice Testing Program for Pumps and Valves SYSTEM: HEATING AND VENTILATION (HV)VALVE CIC P&ID P&ID 1SI I ST VALVE [VALVE IACT NORM ITEST TEST NOTES/DESCRIPTION COOR CLASSJCAT SIZE [TYPE TYPE POS JRQMT FREQ HV-AOV-257AV 2020 D4 A B 72 IN BTF AO 0 FSC Q REACTOR BUILDING VENTILATION FST Q OUTBOARD SUPPLY, RG-01, TP-04_____ ~~~~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _IIV-AOV-259AV 2020 B 12 A B 72 IN BTF AO 0 FSC Q REACTOR BUILDING VENTILATION FST Q INBOARD EXHAUST, RG-01, TP-04 PIT 2Y __ _ _ _ _ _ _ _ _ _ _ _ _ttIV-AOV-261AV 2020 B12 A B 72 IN BTF AO 0 FSC Q REACTOR BUILDING VENTILATION FST Q INBOARD EXHAUST, RG-01, TP-04__ __ ____ PIT 2Y HV-AOV-263AV 2020 E9 A B 72 IN BTF AO 0 FSC Q REACTOR MG SET 1A FST Q VENTILATION OUTBOARD SUPPLY,____ __________PIT 2Y RG-01, TP-04 HV-AOV-265AV 2020 E9 A B 72 IN BTF AO 0 FSC Q REACTOR MG SET lB FST Q VENTILATION OUTBOARD SUPPLY,____ ____PIT 2Y RG-01, TP-04 IIV-AOV-267AV 2020 Dl1 A B 72 IN BTF AO 0 FSC Q *REACTOR MG SET lA EST Q VENTILATION INBOARD EXHAUST,_________ ____ PIT 2Y RG-0l, TP-04 IIV-AOV-269AV 2020 El11 A B 72 IN BTF AO 0 FSC Q REACTOR MG SET lB FST Q VENTILATION OUTBOARD_____________PIT 2Y EXHAUST, RG-01, TP-04 Revision 0 Page 202 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: HEATING AND VENTILATION (HV)VALVE GIGC P&ID P&ID ISI IST VALVE VALVE jACT NORM TEST TEST NOTES/DESCRIPTION _______ICOOR CLASS CAT SIZE TYPE TYPE P0S RQMT FREQ ____________ HV-AOV-270AV 2019 B2 A B 20 IN BTF AO 0 FSC Q CONTROL RM HVAC INLET VALVE,________SH 1___ FST Q RG-01, TP-04 HV-AOV-271AV 2019 Al A B 12 IN BTF AO C FSO Q CONTROL RM HVAC EMERGENCY_________SHi 1______ FST Q BYPASS, RG-01, TP-04 HV-AOV-272AV 2019 B6 A B 8 IN BTF AO 0 FSC Q CONTROL RM PANTRY EXHAUST SH 1 ______ __ __ FST Q ISOLATION, RG-01, TP-04 HV-MOV-258MV 2020 B 12 A B 48 IN BTF MO 0 FSC Q REACTOR BUILDING VENTILATION __________________________PIT 2Y OUTBOARD EXHAUST, RG-01 HV-MOV-260MV 2020 B12 A B 48 IN BTF MO 0 FSC Q REACTOR BUILDING VENTILATION __________ ___ ___ ____________PIT 2Y OUTBOARD EXHAUST, RG-01 HV-MOV-262MV 2020 El0 A B 48 IN BTF MO 0 FSC CS RR MG SET 1A VENTILATION __________ ___ ____PIT 2Y INBOARD SUPPLY, CSJ-08, RG-01 HV-MOV-264MV 2020 El0 A B 48 IN BTF MO 0 FSC CS RR MG SET 1B VENTILATION __________ ___ ____PIT 2Y INBOARD SUPPLY, CSJ-08, RG-01 HV-MOV-266MV 2020 Dl1 A B 48 IN BTF MO 0 FSC CS RR MG SET 1A VENTILATION PIT 2Y OUTBOARD EXHAUST. CSJ-08, RG-____ ___ ___ ___ __ ___ ___ ____ __ ___ ___ ___ ___ ___ ___ 01 HV-MOV-268MV 2020 El11 A B 48 IN BTF MO 0 FSC CS RR MG SET lB VENTILATION PIT 2Y OUTBOARD EXHAUST. CSJ-08, RG-____ __ ___ ___ ___ ___ _______ ___ ___ ___ ___ 01 HV-MOV-272MV 2020 D4 A B 72 IN BTF MO 0 FSC Q REACTOR BUILDING VENTILATION PIT 2Y INBOARD SUPPLY, RG-0l Revision 0 Pg 0 Page 203 Cooper Nuclear Station Ffifh Interval lnservice Testing Pro gram for Pumps and Valves SYSTEM: INSTRUMENT AIR (IA)VALVE CIC P&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ ___________ IA-CV-17CV 2010 E8 A A/C 1/2A CK-L SA 0 LT-1 2Y MS-RV-71A ACCUMULATOR SH 2 FSC RF SUPPLY, ROJ-07, RG-01, TP-01___ __ __ ___ __ __ ___ _ __ _FSO RF IA-CV-I 8CV 2010 E8 A A/C 1/2A CK-L SA 0 LT-1 2Y MS-RV-71B ACCUMULATOR SH 2 FSC RE SUPPLY, ROJ-07, RG-01, TP-01______ _________________ _____ ____ FSO REF_ _ _ _ __ _ _ _ _ _IA-CV- 19CV 2010 D8 A A/C 1/2 CK-L SA 0 LT- 1 2Y MS-RV-71C ACCUMULATOR SH 2 FSC RE SUPPLY, ROJ-07, RG-01, TP-01___ __ __ __ _ ___ __ ____ __ FSO RE IA-CV-20CV 2010 D9 A A/C 1/2A CK-L SA 0 LT-1 2Y MS-RV-71E ACCUMULATOR SH 2 _FSC RE SUPPLY, ROJ-07, RG-01, TP-01______ ______ __________ _____FSO REF_ _ _ _ _ _ _ _ _ _ _IA-CV-21CV 2010 E9 A A/C 1/2A CK-L SA 0 LT-1 2Y MS-RV-71G ACCUMULATOR SH 2 FSC RE SUPPLY, ROJ-07, RG-01, TP-01___ ___ __ __ __ _ _ __ __ _____ _ FSO RE IA-CV-22CV 2010 E9 A A/C 1/2A CK-L SA 0 LT-1 2Y MS-RV-71H ACCUMULATOR SH 2 FSC RE SUPPLY, ROJ-07, RG-01, TP-01_____________ _____ FSO REF_ _ _ _ __ _ _ _ _ _IA-CV-28CV 2010 F9 A A/C 1/2A CK-L SA 0 LT-1 2Y MSIV-AO8OA SUPPLY SH 2 FSC RE ROJ-08, RG-01, TP-01________ ____ __ __ ___________FSO REF_ _ _ _ __ _ _ _ _ _Revision 0 Page 204 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: INSTRUMENT AIR (IA)VALVE CIC P&ID P&ID ISI IST VALVE VALVE [ACT NORM TEST TEST [ NOTES/DESCRIPTION ___COOR] CLASS CAT SIZE TYPE [TYPE POS RQMT FREQ IA-CV-29CV 2010 F8 A A/C 1/2A CK-L SA 0 LT-1 2Y MSIV-AO8OB SUPPLY SH 2 FSC RF ROJ-08, RG-01, TP-01_____ _____ __ __ __ __ __ __ ___ _ ____ ____ ____ FSO RE _ _ _ _ _ _ _ _ _ _ _IA-CV-30CV 2010 F8 A A/C 1/2A CK-L SA 0 LT-1 2Y MSIV-AO8OC SUPPLY SH 2 FSC RF ROJ-08, RG-01, TP-01___ __ ___ __ __ __ ___ _ ___ __ ___ _ ___ __ ___ __ __ FSO RE IA-CV-31CV 2010 F9 A A/C 1/2A CK-L SA 0 LT- 1 2Y MSIV-AO8OD SUPPLY SH 2 FSC RE ROJ-08, RG-01, TP-01___ ___FSO RE IA-CV-32CV 2010 D12 A A/C 1/2A CK-L SA 0 LT-1 2Y MSIV-AO86A SUPPLY SH 2 FSC RE ROJ-08, RG-01, TP-01__ ____ __ FSO RE IA-CV-33CV 2010 D12 A A/C 1/2A CK-L SA 0 LT-1 2Y MSIV-AO86B SUPPLY SH 2 FSC RE ROJ-08, RG-01, TP-01___ __ __ ___ _ __ __ ____ __ FSO RE IA-CV-34CV 2010 E12 A A/C 1/2A CK-L SA 0 LT-1 2Y MSIV-AO86C SUPPLY SH 2 FSC RE ROJ-08, RG-01, TP-01___ _ __ _ __ _ _ __ _ __ _ FSO RE IA-CV-35CV 2010 E12 A A/C 1/2A CK-L SA 0 LT-1 2Y MSIV-AO86D SUPPLY SH 2 FSC RE ROJ-08, RG.-01, TP-01___FSO RE IA-CV-36CV 2010 E8 A A/ 1/2' CK-L OA LT- 1 2Y MS-RV-71D ACCUMULATOR SH 2 FSC RE SUPPLY, ROJ-07, RG-01, TP-01 FSO RE Revision 0 Page 205 Cooper Nuclear Station Ffifh Interval lnservice Testing Program for Pumps and Valves SYSTEM: INSTRUMENT AIR (IA)VALVE CIC P&ID jP&ID IS 1ST VALVE VALVE ACT NORM ITEST TEST NOTES/DESCRIPTION __COOR_____CLASS___ CAT SIZE ITYPE TYPEj POS jRQMT FREQ ____________ IA-CV-37CV 2010 El0 A A/C 1A CK-.L SA 0 LT-1 2Y MS-RV-71F ACCUMULATOR SIT 2 FSC RF SUPPLY, ROJ-07, RG-01, TP-01__ ____ _ _ __ _FSO RF IA-CV-50CV 2010 G5 A A/C 1/4 CK-L SA 0 LT-1 2Y ACCUMULATOR AO-82 IA CHECK SH 2 CCL CVCM VALVE, RG-01, TP-01____ ____COF CVCM IA-CV-51CV 2010 G5 A A/C 1/4 CK-L SA 0 LT-1 2Y ACCUMULATOR AO-83 IA CHECK SH 2 CCL CVCM VALVE, RG-01, TP-01____ ___ ____ ____COF CVCM IA-CV-52CV 2010 G6 *A A/C 1/4 CK-L SA 0 LT-1 2Y ACCUMULATOR AO94 IA CHECK SH 2 CCL CVCM VALVE, RG-01, TP-.01____ ____COF CVCM IA-CV-53CV 2010 G6 A A/C 1/4 CK-L SA 0 LT-1 2Y ACCUMULATOR AO95 IA CHECK SH 2 CCL CVCM VALVE, RG-01, TP-01____ ____ ____COF CVCM IA-CV-54CV 2010 A7 A A/C 1/4 CK-L SA 0 LT-1 2Y HV-AOV-257AV ACCUMULATOR SH 2 CCL CVCM CHECK, RG.-01, TP-01____ ____COF CVCM IA-CV-55CV 2010 A6 A A/C 1/4 CK-L SA 0 LT-1 2Y HV-AOV-259AV ACCUMULATOR SH 2 CCL CVCM CHECK, RG-01, TP-01____ ___ ____COF CVCM Revision 0 Page206 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: INSTRUMENT AIR (IA)VALVE CIC P&ID P&ID ISI 1ST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS C AT SIZE TYPE TYPE POS RQMT FREQ ___________ IA-CV-56CV 2010 B9 A A/C 1/4 CK-L SA 0 LT-1 2Y HV-AOV-261AV ACCUMULATOR SH2 CL CVM CHECKRJ1, RG-01, TP-O1 IA-CV-57CV 2010 B9 A A/C 1/4 CK-L SA 0 LT-1 2Y HV-AOV-263AV ACCUMULATOR SH 2 FSC RF CHECK, ROJ-13, RG-01, TP-01___ __ __ ___ __ __ ___ _ __ __ __ __FSO RE IA-CV-58CV 2010 B9O A A/C 1/4 CK-L SA 0 LT-1 2Y IIV-AOV-265AV ACCUMULATOR Sf12 FSC RE CHECK, ROJ-13, RG-01, TP-01_____ _____ _ ___ _ __ __ __ ___ _ ___ _ ___ FSO REF_ _ _ _ __ _ _ _ _ _IA-CV-59CV 2010 B80 A A/C 1/4 CK-L SA 0 LT-1 2Y HV-AOV-267AV ACCUMULATOR SH 2 FSC RE CHECK, ROJ-13, RG-01, TP-01___ __ __ __ _ ___ __ __ _ __ _FSO RE IA-CV-60CV 2010 F8 2 A/C 2/ CK-L SA 0 LT-1 2PY XV-22I2BOARV IOATIONLATRG SH 2 CFS CVCMKRJ13 G01 P0 IA-CV-65CV 2010 F8 2 A/C 2 CK-L SA 0 LJ-1 OPB X-22 OUTBOARD ISOLATION,RG0 SH 2 COF CVCM RG0___ __ ___ __ __ __ ___ __ ___ _____ __ CCL CVCM IA-CV- 111CV 2010 Ei A AC 1/4 CK-L OA LT-1 2Y IA-271AV ACCUMULATOR Sf11 CCL CVCM CHECK, RG-01, TP-01 COF CVCM Revision 0 Page 207 Cooper Nuclear Station Fifth Interval lnservice Testing Program for Pumps and Valves SYSTEM: INSTRUMENT AIR (IA)VALVE CIC P&ID IP&ID 1S1 J ST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION PC-V-559 2028 C8 2 A 1 GB MA C LI-i OPB PASSIVE MANUAL ISOLATION PC-V-560 2028 D8 2 A 1 GB MA C LI-i OPB PASSIVE MANUAL ISOLATION PC-V-561 2028 C8 2 A 1 GB MA C LJ-i OPB PASSIVE MANUAL ISOLATION PC-V-562 2028 C8 2 A 1 GT MA C LI-i OPB PASSIVE MANUAL ISOLATION PC-V-563 2028 C8 2 A 1 GT MA C LI-i OPB PASSIVE MANUAL ISOLATION PC-V-564 2028 C8 2 A 1 GT MA C LJ-1 OPB PASSIVE MANUAL ISOLATION PC-V-565 2028 D8 2 A 1 ' GT MA C LI-i OPB PASSIVE MANUAL ISOLATION PC-V-566 2028 D8 2 A 1 GT MA C LJ-i OPB PASSIVE MANUAL ISOLATION PC-V-569 2027 17 2 A 1 GT MA C LI-i OPB PASSIVE MANUAL ISOLATION SHl 1_ __ ___ _ _PC-V-570 2027 17 2 A 1/2A GB MA C LI-i OPB PASSIVE MANUAL ISOLATION Sil 1__PC-V-571 2027 17 2 A 1 GB MA C LI-i OPB PASSIVE MANUAL ISOLATION Sil __ __ ___ _ _PC-V-572 2027 17 2 A 1/2A GB MA C LI-i OPB PASSIVE MANUAL ISOLATION SHi 1_ _ _ _PC-V-573 2027 17 2 A 1 GT MA C LI-i OPB PASSIVE MANUAL ISOLATION___________ Sil 1 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _Rev&ion 0 Page 208 Cooper Nuclear Station Fifth Interval Inserviee Testing Program for Pumps and Valves SYSTEM: INSTRUMENT AIR (IA)VALVE CIC P&ID P&ID IS1I IST 1VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR _CLASS CAT SIZE TYPE TYPE POS RQMT FREQ ____________ PC-V-574 2027 J7 2 A 1/2A GB MA C LJ-1 OPB PASSIVE MANUAL ISOLATION___________ SHi 1 _ __ _ _ _ _ _ _ _PC-V-575 2027 J7 2 A 1 GB MA C LU-1 OPB PASSIVE MANUAL ISOLATION________________ SHl_ 1 ___ ________PC-V-576 2027 J7 2 A 1/2A GB MA C LJ-i OPB PASSIVE MANUAL ISOLATION Sil 1_ _ _ __ _ ___ _ _ _ _ _ _ _ _PC-V-577 2027 19 2 A 1 GB MA C LJ-1 OPB PASSIVE MANUAL ISOLATION Sil 1__PC-V-578 2027 J9 2 A 1/2A GB MA C LI-1 OPB PASSIVE MANUAL ISOLATION Sil 1 _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _PC-V-579 2027 J9 2 A 1 GB MA C LJ-i OPB PASSIVE MANUAL ISOLATION___________ _ Sil_ 1 _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _PC-V-580 2027 J9 2 A 1/2/ GB MA C LJ-i OPB PASSIVE MANUAL ISOLATION Sil 1 __PC-V-581 2027 J9 2 A 1 GB MA C LJ-i OPB PASSIVE MANUAL ISOLATION SH 2 __ ___ ___ _ __ _ _PC-V-582 2027 J3 2 A 'A GB MA C LI-i OPB PASSIVE MANUAL ISOLATION________ ___ 5SH 2 _ __ _PC-V-583 2027 13 2 A 1 GB MA C LI-1 OPB PASSIVE MANUAL ISOLATION____________ S11 2 _ __ _ _ _ __ _ _ _ _ _ _ _ _PC-V-584 2027 13 2 A 1/2 GB MA C LJ-i OPB PASSIVE MANUAL ISOLATION SH 2 Revision 0 Page 209 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves[SYSTEM: INSTRUMENT AIR (IA)VALVE CIC IP&ID P&ID 1SI 1ST VALVE VALVE ACT NORM TEST [TEST NOTES/DESCRIPTION PC-V-585 2027 J4 2 A 1 GB MA C LJ-1 OPB PASSIVE MANUAL ISOLATION PC-V-586 2027 J4 2 A 1/2A GB MA C LJ-i OPB PASSIVE MANUAL ISOLATION SH2 2 _ __ _ _ _PC-V-587 2027 J4 2 A 1 GB MA C LJ-1 OPB PASSIVE MANUAL ISOLATION SH 2 ____ ____ ____ ____PC-V-588 2027 J4 2 A 1/2A GB MA C LJ-1 OPB PASSIVE MANUAL ISOLATION SH2 2_ __ __ _ _PC-V-589 2027 J5 2 A 1 GB MA C LJ-1 OPB PASSIVE MANUAL ISOLATION___________ S_ 5 2 _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _PC-V-590 2027 J5 2 A 1/2A GB MA C LJ-1i OPB PAssIVE MANUAL ISOLATION S112 Revision 0 Page 210 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: MAIN STEAM (MS)P&ID ISI IST VALVE VALVE ACT NORM TEST TEST VALVE CIC P&ID COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION MS-AOV-A0SOA 2041 B5 1 A 24 GB AO 0 LJ-1 RE MSIV A INBOARD ISOLATION PSC Q CSJ-03, RG-01 FSC CS FST CS___PIT 2Y MS-AOV-AO80B 2041 B5 I A 24 GB AO 0 UJ-1 RE MSIV B INBOARD ISOLATION PSC Q CSJ-03, RG-01 FSC CS FST CS___PIT 2Y MS-AOV-AO80C 2041 B7 1 A 24 GB AO 0 LJ-1 RE MSIV C INBOARD ISOLATION PSC Q CSJ-03, RG-01 FSC CS FST CS______ ~~~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _MS-AOV-AO80D 2041 B7 1 A 24 GB AO 0 LJ-i RE MSIV D INBOARD ISOLATION PSC Q CSJ-03, RG-01 FSC CS FST CS PIT 2Y __ _ _ _ _ _ _ _ _ _ _ _ _Revision 0 Page 211 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: MAIN STEAM (MS)IP&ID ISI] IST VALVE IVALVE ACT INORM] TEST 1TEST VALVE CIC P&ID COOR CLASS CAT SIZE jTYPE TYPEj POS R'QMT JFREQ NOTES/DESCRIPTION MS-AOV-AO86A 2041 B4 1 A 24 GB AO 0 LJ-i RF MSIV A OUTBOARD ISOLATION PSC Q CSJ-03, RG-01 FSC CS FST CS____PIT 2Y MS-AOV-AO86B 2041 A4 1 A 24 GB AO 0 LJ-i RF MSIV B OUTBOARD ISOLATION PSC Q CSJ-03, RG-01 FSC CS FST CS__ __ __ __ _ __ _PIT 2Y MS-AOV-AO86C 2041 A8 1 A 24 GB AO 0 UJ-1 RF MSIV C OUTBOARD ISOLATION PSC Q CSJ-03, RG-01 FSC CS FST CS___PIT 2Y MS-AOV-AO86D 2041 B8 1 A 24 GB AO 0 LJ-i RF MSIV D OUTBOARD ISOLATION PSC Q CSJ-03, RG-01 FSC CS FST CS__ __ _ __ _PIT 2Y MS-AOV-738AV 2028 C9 1 B 1/2A GB AO C PIT 2Y RPV HEAD VENT DRAIN TO SUMP,_____ _____PASSIVE ISOLATION, RG-01 MS-AOV-739AV 2028 C9 1 B 1/2A GB AO C PIT 2Y RPV HEAD VENT DRAIN TO SUMP,____ ___ ___ __ _ ___ ___ _ __ ___ ___ _ __ ___ ___ __ ____ __ ___PASSIVE ISOLATION, RG-01 Revision 0 Page 212 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: MAIN STEAM (MS)P&ID ISI IST 1VALVE 1VALVE ACT 1NORM TEST ITEST VALVE CIC P&ID COOR CLASS CAT jSIZE TYPE TYPE jPOS RQMT jFREQ NOTES/DESCRIPTION MS-CV-20CV 2028 Cl11 C 10 CK-S SA C VBT RF 71A RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 Al, RG-01, TP-08___FSC 2Y MS-CV-21CV 2028 Cll C 10 CK-S SA C VBT RE 71A RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 A2, RG-01I, TP-08___ ____ __FSC 2Y MS-CV-22CV 2028 C11 C 10 CK-S SA C VBT RF 71B RV DISCHARGE VACUUM FSO 2Y BREAKER VB 7iBl,RG-0i, TP-08___ _ __ ____ __FSC 2Y MS-CV-23CV 2028 Ci11 C 10 CK-S SA C VBT RF 71B RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 B2, RG-01, TP-08___ __ ___ __ _ _____ _____ __FSC 2Y MS-CV-24CV 2028 Dii1 C 10 CK-S SA C VBT RF 71C RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 C1, RG-01, TP-08___ ____ __FSC 2Y MS-CV-25CV 2028 Di 31 C 10 CK-S SA C VBT RF 71C RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 C2, RG-01, TP-08___ _ __ ____ __FSC 2Y MS-CV-26CV 2028 D11 C 10 CK-S SA C VBT RE 71D RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71D1, RG-01, TP-08___ _ __ ____ __FSC 2Y MS-CV-27CV 208 Di 3 C 10 CK-S SA C VBT RE 71D RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 D2, RG-01, TP-08 FSC 2Y Revision 0 Page 213 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: MAIN STEAM (MS)P&ID IISI IST VALVE VALVE ACT 1NORM TEST 1TES T VALVE CIC [P&ID COOR JCLASS CAT SIZE TYPE TYPE jPOS RQMT FRLEQ JNOTES/DESCRI1PTION MS-CV-28CV 2028 D9 3 C 10 CK-S SA C VBT RE 71E RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 El,RG-01, TP-08___ ___ __ ___ __ __ __ __FSC 2Y MS-CV-29CV 2028 D9 3 C 10 CK-S SA C VBT RF 71E RV DISCHARGE VACUUMI~FSO 2Y BREAKER VB 71 E2, RG-01, TP-08___ _ __ ____ __FSC 2Y MS-CV-30CV 2028 D9 3 C 10 CK-S SA C VBT RF 71F RV DISCHARGE VACUUM FSO 2Y BREAKER VB71 F1, RG-01, TP-08_____ ______ _____ ~ FSC 2Y _ _ _ _ _ _ _ _ _ _ _ _ _MS-CV-31CV 2028 D9 3 C 10 CK-S SA C VBT RF 71F RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 F2, RG-01, TP-08____ ____FSC 2Y MS-CV-32CV 2028 E9 3 C 10 CK-S SA C "VBT RF 71G RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 G1, RG-01, TP-08___ _ __ ____ __FSC 2Y MS-CV-33CV 2028 E9 3 C 10 CK-S SA C VBT RF 71G RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 G2, RG-01, TP-08___ _ __ ____ __FSC 2Y MS-CV-34CV 2028 E9 3 C 10 CK-S SA C VBT RF 71H RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 HI,RG-01, TP-08____FSC 2Y Revision 0 Page 214 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: MAIN STEAM (MS)IP&ID IISI 1ST VALVE VALVE ACT NORM TEST TESTJ VALVE CIC _P&ID COOR _CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION MS-CV-35CV 2028 E9 3 C 10 CK-S SA C VBT RF 7111 RV DISCHARGE VACUUM FSO 2Y BREAKER VB 71 112, RG-01, TP-08___FSC 2Y MS-MOV-MO74 2041 C7 1 A 3 GT MO 0 LJ-1 OPB MS L1NE DRAIN INBOARD FSC Q/CS ISOLATION, CSJ-10, RG-01____ ___PIT 2Y MS-MOV-MO77 2041 C11 A 3 GT MO 0/C LU-1 OPB MS LINE DRAIN OUTBOARD FSC Q ISOLATION, RG-01______ ~~~~PIT 2Y __ _ _ _ _ _ _ _ _ _ _ _ _MS-RV-70ARV 2028 C10 C 6 RV SA C RVT APP I SAFETY VALVE MS LINE A, RV-02,______ _____RG-01 MS-RV-70BRV 2028 E9 1 C 6 RV SA C RVT APP I SAFETY VALVE MS LINE D, RV-02,____ ___ __ _ ___RG-01 MS-RV-70CRV 2028 E9 1 C 6 RV SA C RVT APP I SAFETY VALVE MS LINE D, RV-02,____ ___ ____ __ ___RG-01 MS-RV-71ARV 2028 Cli1 B/C 6 RV AO C RVT APP I SAFETY RELIEF VALVE MS LINE A, FSO RF RV-03, RG-01 FSC RF___ __ __ ____ __PIT RF MS-RV-71BRV 208 Ci 1 BC 6 RV AO C RVT APP I SAFETY RELIEF VALVE MS LINE A, FSO RF RV-03, RG-01 FSC RF PIT RF Revision 0 Page 215 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: MAIN STEAM (MS)P&ID ISI 1ST VALVE VALVE ACT NORM TEST TEST VALVE CIC P&ID [COOR CLASS CAT SIZE TYPE TYPE POS jRQMT FREQ NOTES/DESCRIPTION MS-RV-71CRV 2028 D11 B/C 6 RV AO C RVT APP I SAFETY RELIEF VALVE MS LINE B, FSO RE RV-03, RG-01 FSC RE_____ _____ ~~PIT RF _ _ _ _ _ _ _ _ _ _ _ _ _MS-RV-71DRV 2028 DI11 B/C 6 RV AO C RVT APP I SAFETY RELIEF VALVE MS LINE B, FSO RE RV-03, RG-01 FSC RE PIT RE MS-RV-71ERV 2028 D9 1 B/C 6 RV AO C RVT APP I SAFETY RELIEF VALVE MS LINE C, FSO RE RV-03, RG-01 FSC RE__ __ __ _PIT RE MS-RV-71FRV 2028 D9 1 B/C 6 RV AO C RVT APP I SAFETY RELIEF VALVE MS LINE C, FSO RE RV-03, RG-01 FSC RE___ _ __ __PIT RE MS-RV-71GRV 2028 E9 1 B/C 6 RV AO C RVT APP I SAFETY RELIEF VALVE MS LINE D, FSO RE RV-03, RG-01 FSC RE___ _ _ ___PIT RE MS-RV-71HRV 2028 E9 1 B/C 6 RV AO C RVT APP I SAFETY RELIEF VALVE MS LINE D, FSO RE RV-03, RG-01 FSC RE PIT RE Revision 0 Page216 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: NEUTRON MONITORING TRAVERSING INCORE PROBE (NMT)VALVE CIC P&ID P&ID ISI IST VALVE IVALVE ACT NORM TEST TEST NOTES/DESCRIPTION ________jCOOR CLASS jCAT SIZE jTYPE TYPE _POSj RQMT FREQ NM-CV-CV2 2083 J2 2 A!C 3/8 CK-P SA C LU-1 OPB N2 PURGE ISOLATION, RG-01, COF CVCM TP-01, TP-06___ ___ __ ___ _ ___ ___ _______ CCL CVCM NM-CV-CV4 2083 J2 2 A/C 3/8 CK-P SA C LU-i OPB N2 PURGE ISOLATION, RG-01, COF CVCM TP-01, TP-06___________ ____ __ __CCL CVCM NMT-NVA-104AX 2083 J2 2 D 3/8 SHR EX 0 EX 2Y TIP A SHEAR VALVE, RG-01 NMT-NVA-104BX 2083 J2 2 D 3/8 SHR EX 0 EX 2Y TIP B SHEAR VALVE, RG-01 NMT-NVA-104CX 2083 J2 2 D 3/8 SHR EX 0 EX 2Y TIP C SHEAR VALVE, RG-01 NMT-NVA-104DX 2083 J2 2 D 3/8 SHR EX 0 EX 2Y TIP D SHEAR VALVE,, RG-01 NMT-NVA-104A 2083 J2 2 A 3/8 BAL 50 0 LJ-1 OPB TIP A BALL VALVE, RG-01, FSC Q TP-04 FST Q__ __ __ __ _ __ _PIT 2Y NMT-NVA-104B 2083 J2 2 A 3/8 BAL 50 0 LJ-i OPB TIP BBALL VALVE, RG-0i, FSC Q TP-04 FST Q PIT 2Y Revision 0 Page 217 Cooper Nuclear Station F~fih Interval Inservice Testing Program for Pumps and Valves SYSTEM: NEUTRON MONITORING TRAVERSING INCORE PROBE (NMT)VALVE CIC IP&ID P&ID IISI IST 1VALVE VALVE ACT NORM TEST 1TEST NOTES/DESCRIPTION ___________I____COOR _C LASS CAT SIZE TYPE TYPE POS RQMT] FREQ ____________ NMT-NVA-104C 2083 J2 2 A 3/8 BAL S0 0 LU-i OPB TIP C BALL VALVE, FSC Q RG-01, TP-04 FST Q______ _____ ~PIT 2Y __ _ _ _ _ _ _ _ _ _ _ _NMT-NVA-104D 2083 J2 2 A 3/8 BAL S0 0 LJ-i OPB TIP D BALL VALVE, FSC Q RG-01, TP-04 FST Q PIT 2Y Revision 0 Page 218 Revision 0 Page 218 Cooper Nuclear Station F~fih Interval Inservice Testing Program for Pumps and Valves SYSTEM: NUCLEAR BOILER INSTRUMENTATION (NBI) AND VARIOUS OTHER EXCESS FLOW CHECK VALVES VALVE CIC P&ID P&ID ISI IST 1VALVE VALVE ACT 1NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT jSIZE TYPE TYPE POS RQMT jFREQ ____________ NBI-AOV-736AV 2028 C9 1 B 1/2A GB AO 0 PIT 2Y RPV FLANGE LEAKOFF PASSIVE_____ISOLATION, RG-01 NBI-AOV-737AV 2028 C9 1 B 1/2A GB AO C PIT 2Y RPV FLANGE LEAKOFF PASSIVE ISOLATION, RG-01 NBI-CV-10BCV 2026 B5 1 C 1 CK-B SA 0 CCL CVCM UPPER CHECK FOR SH 1___ __ COF CVCM LI-61, RG-01, TP-01 NBI-CV-1 1BCV 2026 B5 1 C 1 CK-B SA 0 CCL CVCM CONDENSING CHAMBER 2A UPPER________Sil 1__ COF CVCM TO 25-5 RACK, RG-01, TP-01 NBI-CV-12BCV 2026 C5 1 C 1 CK-B SA 0 CCL CVCM CONDENSING CHAMBER 2A Sil ____1__ COF CVCM LOWER TO 25-5 RACK, RG-01, TP-01 NBI-CV-13BCV 2026 C5 1 C 1 CK-B SA 0 CCL CVCM CONDENSING CHAMBER 3A TO SH 1___ __ COF CVCM LITS-73A, RG-01, TP-01 NBI-CV-14BCV 2026 D5 1 C 1 CK-B SA 0 CCL CVCM CONDENSING CHAMBER 3A TO 25-Sil 1__ COF CVCM 5 & 25-5-1 RACK, RG-01, TP-01 NBI-CV-15BCV 2026 D5 1 C 1 CK-B SA 0 CCL CVCM LT-52A; LIS-83A & C; LT-61; DPT-65;Sil 1__ COF CVCM LIS-101A &B, RG-01, TP-01 NBI-CV-16BCV 2026 D5 1 C 1 CK-B SA 0 CCL CVCM LOW SIDE, FT-64T, RG-01, TP-01_______Sil 1______ __ COF CVCM NBI-CV-17BCV 2026 D5 1 C 1 CK-B SA 0 CCL CVCM LOW SIDE FT-64R;__ _ _ _ _ _ _ Sil 1 _ _ ______ ____ ___ _ _ _ COF CVCM FT-63D, RG-01, TP-01 Revision 0 Page 219 Cooper Nuclear Station F~ifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: NUCLEAR BOILER INSTRUMENTATION (NBI) AND VARIOUS OTHER EXCESS FLOW CHECK VALVES VALVE GIG P&ID P&D II IT VLEVLE ACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE jTYPE P05 RQMT FREQ _____________ NBI-CV- 18BCV 2026 E5 1 C 1 CK-B SA 0 CCL CVCM HIGH SIDE FT-63D, RG-01, TP-01__ _ _ _ _ _ _ SHi 1 _ _ _ _ __ _ _ _ _ _ COF CVCM _ _ _ _ _ _ _ _ _ _ _ _ _NBI-CV-19BCV 2026 G5 1 C 1 CK-B SA 0 CCL CVCM ABOVE CORE PLATE PRESS, RG-01,________SHi1 COF CVCM TP-01 NBI-CV-20BCV 2026 F5 1 C 1 CK-B SA 0 CCL CVCM BELOW CORE PLATE TO SHi1 COF CVCM INSTRUMENT RACK 25-51, RG-01,____ ___ ____ ___TP-01 NBI-CV-21BCV 2026 C8 1 C 1 CK-B SA 0 CCL CVCM CONDENSING CHAMBER 2B UPPER________SH 1___ COF CVCM TO 25-6 RACK, RG-01, TP-01 NBI-CV-22BCV 2026 B8 1 C 1 CK-B SA 0 CCL CVCM CONDENSING CHAMBER 2B UPPER________SHi 1__ ______ COF CVCM TO 25-6 RACK, RG-01, TP-01 NBI-CV-23BCV 2026 C8 1 C 1 CK-B SA 0 CCL CVCM CONDENSING CHAMBER 3B LT-70;________SHi 1__ COF CVCM LITS-73B, RG-01, TP-01 NBI-CV-24BCV 2026 D8 1 C 1 CK-B SA 0 CCL CVCM CONDENSING CHAMBER 3B TO 25-________SH 1 _______ ___COF CVCM 6 & 25-6-1 RACKS, RG-01, TP-01 NBI-CV-25BCV 2026 D8 1 C 1 CK-B SA 0 CCL CVCM LOW SIDE LIS-83B;________SHl ___ COF CVCM LT-52B; LIS-IO1C & D, RG-01, TP-01 NBI-CV-26BCV 2026 D8 1 C 1 CK-B SA 0 CCL CVCM LOW SIDE FT-64L, RG-01, TP-01________SHl 1 ___ _____ COF CVCM NBI-CV-27BCV 2026 D8 1 C 1 CK-B SA 0 CCL CVCM LOW SIDE FT-63C;________Sil 1 __ ______ _____ COF CVCM FT-64N, RG-01, TP-01 NBI-CV-28BCV 26 E8 1 C 1 CK-B OA CCL CVCM HIGH SIDE FT-63C, RG-01, TP-01 Revision 0 Page 220 Cooper Nuclear Station F~fih Interval Inservice Testing Program for Pumps and Valves SYSTEM: NUCLEAR BOILER INSTRUMENTATION (NBI) AND VARIOUS OTHER EXCESS FLOW CHECK VALVES VALVE CIC P&ID P&ID ISI IST VALVE VALVE IACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ _____________ SHl 1___ __ COF CVCM _____________ NBI-CV-29BCV 2026 P8 1 C 1 CK-B SA 0 CCL CVCM BELOW CORE PLATE TO Sil1 COF CVCM INSTRUMENT RACK 25-52, RG-01,________TP-0 1 NBI-CV-30BCV 2026 H5 1 C 1 CK-B SA 0 CCL CVCM FT-63B; FT-64B LOW SIDE, RG-01,________Sil 1__ __ COF CVCM TP-01 NBI-CV-31BCV 2026 H5 1 C 1 CK-B SA 0 CCL CVCM FT-63B; HIGH SIDE, RG-01, TP-01 SHi ___1_ _ __ _ _ ___ COP CVCM _ _ _ _ _ _ _ _ _ _ _ _ _NBI-CV-32BCV 2026 H5 1 C 1 CK-B SA 0 CCL CVCM FT-64D; LOW SIDE, RG-01, TP-01_ _ _ _ _ _ _ _ SHl 1_ _ ___ COF CVCM _ _ _ _ _ _ _ _ _ _ _ _ _NBI-CV-33BCV 2026 H5 1 C 1 CK-B SA 0 CCL CVCM FT-64F; LOW SIDE, RG-01, TP-01 5S1 1 ____COF CVCM NBI-CV-34BCV 2026 H5 1 C 1 CK-B SA 0 CCL CVCM FT-64K; LOW SIDE, RG-01, TP-01 SH 1__ ___ COF CVCM _____________ NBI-CV-35BCV 2026 H5 1 C 1 CK-B SA 0 CCL CVCM FT-64M; LOW SIDE, RG-01, TP-0l_ _ _ _ _ _ _ _ Sil 1 __ _ __ _ _ _ _ _ _ _ COF CVCM _ _ _ _ _ _ _ _ _ _ _ _ _NBI-CV-36BCV 2026 115 1 C 1 CK-B SA 0 CCL CVCM FT-64V; LOW SIDE, RG-01, TP-01 Sil 1 _ _ _ _ _ _ _ _ _ _ _ _ _ COP CVCM _ _ _ _ _ _ _ _ _ _ _ _ _NBI-CV-37BCV 2026 115 1 C 1 CK-B SA 0 CCL CVCM FT-64X; LOW SIDE, RG-01, TP-01_________ Sil ___ ____ _ _ COP CVCM _ _ _ _ _ _ _ _ _ _ _ _ _ _NBI-CV-38BCV 26 J5 1 C 1 CK-B OA CCL CVCM FT-64Z; LOW SIDE, RG-01, TP-01 SillCOP CVCM Revision 0 Page221 Cooper Nuclear Station F~ifh lnterval Inservice Testing Program for Pumps and Valves SYSTEM: NUCLEAR BOILER INSTRUMENTATION (NB1) AND VARIOUS OTHER EXCESS FLOW CHECK VALVES VALVE CIC IPP&ID &ID ISI IST VALVE VLE ACT NORM TEST TEST NOTES/DESCRIPTION _______jCOOR CLASS CAT_]_SIZE TYPE_[TYPE POS RQMT FREQ ____________ NBI-CV-39BCV 2026 J5 1 C 1 CK-B SA 0 CCL CVCM FT-64A; FT-63A LOW SIDE, RG-01, SHl1 COF CVCM TP-01 NBI-CV-40BCV 2026 J5 1 C 1 CK-B SA 0 CCL CVCM FT-63A HIGH SIDE, RG-01, TP-01 Sil 1 __ _____________ COF CVCM NBI-CV-41BCV 2026 J5 1 C 1 CK-B SA 0 CCL CVCM FT-64C LOW SIDE, RG-01, TP-01 Sil 1______ COF CVCM NBI-CV-42BCV 2026 J5 1 C 1 CK-B SA 0 CCL CVCM FT-64E LOW SIDE, RG-01, TP-01 Sil 1______ __ COF CVCM NBI-CV-43BCV 2026 J5 1 C 1 CK-B SA O CCL CVCM FT-64J LOW SIDE, RG-01, TP-0l Sil 1__ COF CVCM NBI-CV-44BCV 2026 J5 1 C 1 CK-B SA 0 CCL CVCM FT-64S LOW SIDE, RG-01, TP-0l Sil 1 _ _ __ _ _ _ _ _ _ COF CVCM _ _ _ _ _ _ _ _ _ _ _ _NBI-CV-45BCV 2026 J5 1 C 1 CK-B SA 0 CCL CVCM FT-64U LOW SIDE, RG-01, TP-01 Sil 1______ COF CVCM ____________ NBI-CV-46BCV 2026 J5 1 C 1 CK-B SA 0 CCL CVCM FT-64W LOW SIDE, RG-01, TP-01 Sil 1___ __ COF CVCM NBI-CV-47BCV 2026 J5 1 C 1 CK-B SA 0 CCL CVCM FT-64Y LOW SIDE, RG-01, TP-01 Sil 1__ COF CVCM NBI-CV-48BCV 2028 C8 1 C 1 CK-B SA 0 CCL CVCM VESSEL FLANGE LEAKOFF LINE,__ __ __ __ __ _ _ __ _ __ __ _ __ _ __ _ __ _ COF CVCM RG-01, TP-01 NBI-CV-49BCV 2063 C 1/2/ CK-S SA C FSO RF REFERENCE LEG LOOP A Rev&ion 0 Page 222 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: NUCLEAR BOILER INSTRUMENTATION (NBI) AND VARIOUS OTHER EXCESS FLOW CHECK VALVES VALVE CIC IP&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION _______JCOOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ _____________ Sill C2 ___ ___FSC RE INJECTION, ROJ-0l, RG-01, TP-0l NBI-CV-50BCV 2026 C3 3 C 1/2A CK-S SA C FSO RF REFERENCE LEG LOOP A SH 1 __ ________________ FSC RE INJECTION, ROJ-01, RG-01, TP-01 NBI-CV-51BCV 2026 Cl11 C 1/2A CK-S SA C FSO RF REFERENCE LEG LOOP B________SH 1 __________FSC RE INJECTION, ROJ-01, RG-01, TP-01 NBI-CV-52BCV 2026 Cl11 C 1/2A CK-S SA C FSO RF REFERENCE LEG LOOP B Sil 1___ __ FSC RF INJECTION, ROJ-01, RG-0i, TP-0i NBI-CV-55CV 2026 Gil1 A/C 3/8 CK-S SA 0 LT2 RF REFERENCE LEG LOOP 3A SHll FSC RE OUTBOARD INJECTION, ROJ-02, FSO RE RG-01,TP-01 NBI-CV-56CV 2026 HI 1 3 A/C 3/8 CK-S SA 0 LT2 RE REFERENCE LEG LOOP 3B SH l FSC RE OUTBOARD INJECTION, ROJ-02,_____ ____________F SO RE RG-0l, TP-01 NBI-SOV-SSV738 2026 HI 1 2 B 1/2A SOV SOV C FSO RE REFERENCE LEG LOOP A SHi 1 ST RE INJECTION, ROJ-01, RG-01, TP-04___ _ __ __ __ ____ __PIT 2Y NBI-SOV-SSV739 2026 C12 2 B 1/2A SOV SOV C FSO RE REFERENCE LEG LOOP B SHl 1EST RE INJECTION, ROJ-01, RG-01, TP-04___ __ __ ____ __PIT 2Y CS-CV-16BCV 2045 A8 1 C 1 CK-B SA 0 CCL CVCM DPIS-43A LOW SIDE EXCESS FLOW, SHi 1__ COF CVCM RG-0l, TP-01 CS-CV-l7BCV 24 A8 1 C 1 CK-B OA CCL CVCM DPIS-43B LOW SIDE EXCESS FLOW, Revision 0 Page 223 Cooper Nuclear Station F~ifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: NUCLEAR BOILER INSTRUMENTATION (NBI) AND VARIOUS OTHER EXCESS FLOW CHECK VALVES VALVE CIC P&ID 1P&ID ISI 1IST VALVE VALVE ACT TNORM 1TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ _____________ SH 1 COF CVCM RG-01, TP-01 HPCI-CV-10BCV 2041 E4 1 C 1 CK-B SA 0 CCL CVCM HIGH SIDE HPCI-DPIS-76; -77; PS-___________COF CVCM 68A; PS-68C, RG-01, TP-01 HPCI-CV-1 1BCV 2041 E4 1 C 1 -CK-B SA 0 CCL CVCM LOW SIDE HIPCI-DPIS-76; DPIS-77;___________COF CVCM PS-68B; PS-68D, RG-01, TP-01 MS-CV-10BCV 2041 .C4 1 C 1 CK-B SA 0 CCL CVCM FT-51A; DPIS-116A, B, C, & D HIGH___ __ ___COF CVCM SIDE, RG-01, TP-01 MS-CV-11BCV 2041 C4 1 C 1 CK-B SA 0 CCL CVCM FT-51A; DPIS-116A, B, C,& D LOW____ ___ ________COF CVCM SIDE, RG-01, TP-01 MS-CV-12BCV 2041 C4 1 C 1 CK-B SA 0 CCL CVCM FT-51B; DPIS-117A, B,C, &D LOW____________ ___ ___COF CVCM SIDE, RG-01, TP-01 MS-CV-13BCV 2041 C4 1 C 1 CK-B SA 0 CCL CVCM FT-51B; DPIS--117A, B, C, &D HIGH___ ___COF CVCM SIDE, RG-01, TP-01 MS-CV-14BCV 2041 C8 1 C 1 CK-B SA 0 CCL CVCM FT-51C; DPIS-118A, B, C, & D HIGH________ ____COF CVCM SIDE, RG-01, TP-01 MS-CV-15BCV 2041 C8 1 C 1 CK-B SA 0 CCL CVCM FT-51D; DPIS-119A, B, C, & DHIGH__________ ___ ____COF CVCM SIDE, RG-01, TP-01 MS-CV-16BCV 2041 C8 1 C 1 CK-B SA 0 CCL CVCM FT-51C; DPIS-118A, B, C, & D LOW____ ____ ____ ____COF CVCM SIDE, RG-01, TP-01 MS-CV-17BCV 24 C8 1 C 1 CK-B SA 0 CCL CVCM FT-51D; DPIS-1 19A, B, C, & D LOW COF CVCM SIDE, 'RG-01, TP-01 Revision 0 Page 224 Cooper Nuclear Station F~flh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: NUCLEAR BOILER INSTRUMENTATION (NBI) AND VARIOUS OTHER EXCESS FLOW CHECK VALVES VALVE CIC IP&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION ___________COOR CLASS CAT SIZE TYPEJTYPE POS RQMT FREQ _____________ RCIC-CV-10BCV 2041 E8 1 C 1 CK-B SA 0 CCL CVCM RCIC-DPIS-84 HIGH SIDE EFCV, RG-______________COF CVCM 01, TP-01 RCIC-CV-1 1BCV 2041 E8 1 C 1 CK-B SA 0 CCL CVCM RCIC-DPIS-84 LOW SIDE EXCESS___________ _________COF CVCM FLOW CHECK VALVE, RG-0 1, TP-0 1 RCIC-CV-12BCV 2041 E8 1 C 1 CK-B SA 0 CCL CVCM RCIC-DPIS-83 HIGH SIDE EFCV, RG-___ ___COF CVCM 01, TP-01 RCIC-CV-13BCV 2041 E8 1 C 1 CK-B SA 0 CCL CVCM RCIC-DPIS-83 LOW SIDE EXCESS___________ ___ ____COF CVCM FLOW CHECK VALVE, RG-01, TP-01 RR-CV-10CV 2027 D5 1 C 1 CK-B SA 0 CCL CVCM PS-128A SENSING LINE, RG-01, TP-Sil 1 __ __ ___ __ __ COF CVCM 01 RR-CV-11CV 2027 D5 1 C 1 CK-B SA 0 CCL CVCM PS-128A SENSING LINE, RG-01, TP-Sil 1 __ ______ COF CVCM 01 RR-CV-12CV 2027 C5 1 C 1 CK-B SA 0 CCL CVCM DPT-1 11A LOW SIDE, RG-01, TP-01_ _ _ _ _ _ _ _ Sil 1 _ _ _ _ _ _ __ _ _ COF CVCM _ _ _ _ _ _ _ _ _ _ _ _ _RR-CV-13CV 2027 C5 1 C 1 CK-B SA 0 CCL CVCM DPT-1 11A HIGH SIDE, RG-01, TP-01 SH 1 __ ______ COF CVCM _____________ RR-CV-15CV 2027 F5 1 C 1 CK-B SA 0 CCL CVCM PT-25A SENSING LINE, RG-01, TP-01________ SH 1 _ _ ___ COF CVCM _ _ _ _ _ _ _ _ _ _ _ _ _RR-CV- 16CV 2027 F5 1 C 1 CK-B SA 0 CCL CVCM PT-24A SENSING LINE, RG-01, TP-01________Sill COF CVCM RR-CV-17CV 22 C5 1 C 1 CK-B SA O CCL CVCM -FT-110A AND BHIGH SIDE, RG-01, Revision 0 Page 225 Cooper Nuclear Stajion F~fih Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: NUCLEAR BOILER INSTRUMENTATION (NBI) AND VARIOUS OTHER EXCESS FLOW CHECK VALVES VALVE CIC IP&ID P&ID IISI IST 1VALVE VALVE [ACT NORM TEST TEST NOTES/DESCRIPTION __________COOR _CLASS CAT_]_SIZE TYPE TYPE POS RQMT FREQ________SH 1 __ ___ COF CVCM TP-01 RR-CV-18CV 2027 C5 1 C 1 CK-B SA 0 CCL CVCM FT-110A AND BLOW SIDE, RG-01,__ _ _ _ _ _ _ Sil ___1_ _ __ _ ____ COF CVCM TP-O1 RR-CV-27CV 2027 C6 1 C 1 CK-B SA 0 CCL CVCM DPT-l111B LOW SIDE, RG-0l, TP-01__ _ _ _ _ _ _ SH 2 __ _ _ _ _ _ _COF CVCM _ _ _ _ _ _ _ _ _ _ _ _ _RR-CV-28CV 2027 C6 1 C 1 CK-B SA 0 CCL CVCM DPT-111lB HIGH SIDE, RG-0l, TP-01 SH 2 ____ _ _ __ _ ____ COF CVCM _ _ _ _ _ _ _ _ _ _ _ _ _RR-CV-30CV 2027 G6 1 C 1 CK-B SA 0 CCL CVCM PT-25B REACTOR RECIRC PUMP lB SH 2 __________COF CVCM SEAL CAVITY LINE, RG-01, TP-0l RR-CV-31CV 2027 G6 1 C 1 CK-B SA 0 CCL CVCM PT-24B SENSING LllNE, RG-01, TP-01 SH 2 ______COF CVCM____________ RR-CV-32CV 2027 B6 1 C 1 CK-B SA 0 CCL CVCM FT-i110C AND D HIGH SIDE, RG-01, SH 2 ___ _________ ___COF CVCM TP-01 RR-CV-33CV 2027 B6 1 C 1 CK-B SA 0 CCL CVCM FT-110C AND DLOW SIDE, RG-01, S_______ H 2 ___ ___ _____ _________ COF CVCM TP-01 Revision 0 Page 226 Revision 0 Page 226 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: PRIMARY CONTAINMENT (PC)VALVE CIC P&ID P&ID ISI 1ST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ PC-AOV-237AV 2022 F9 2 A 24 BTF AO C LJ-i RF SUPPRESSION CHAMBER INLET SHi1 FSC Q OUTBOARD ISOLATION, RG-01, FST Q TP-04___ __ __ ___ _ __ __ __ __ _____ __PIT 2Y PC-AOV-238AV 2022 E8 2 A 24 BTF AO C LJ-i RF DRY WELL INLET OUTBOARD SHi1 FSC Q ISOLATION, RG-01, TP-04 FST Q______________ _____PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _PC-AOV-243AV 2022 G10 A 20 BTF AO C LJ-i RE SUPPRESSION CHAMBER SH 1 FSO "Q VACUUM RELIEF, RG-01, TP-04 FST Q___ _ __ _ __ ____ __PIT 2Y PC-AOV-244AV 2022 H10 2 A 20 BTF AO C LI-i RE SUPPRESSION CHAMBER SHl 1 SO Q VACUUM RELIEF, RG-01, TP-04 FST Q___ __ __ __ _ __ _ __ ____ __PIT 2Y PC-AOV-245AV 2022 Fl1 A 24 BTF AO C LI-i RE SUPPRESSION CHAMBER SHi1 FSC Q EXHAUST OUTBOARD FST Q ISOLATION, RG-01, TP-04___ __ __ ___ _ __ __ ____ __PIT 2Y PC-AOV-246A'V 22 C2 2 A 24 BTF AO C LIi RE DRYWELL EXHAUST OUTBOARD SH1FSC Q ISOLATION, RG-01, TP-04 FST Q PIT 2Y Revision 0 Page 227 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: PRIMARY CONTAINMENT (PC)VALVE CIC P&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE P0S RQMT FREQ ____________ PC-AOV-247AV 2022 D3 2 A 0.5 GL AO C LJT-1 OPB DRYWELL EXHAUST OUTBOARD SHi1 FSC Q ISOLATION, RG-01, TP-04___ _ ___ _ __ ___ __ __FST Q PC-AOV-248AV 2022 D3 2 A 0.5 GL AO C LJT-1 OPB DRYWELL EXHAUST OUTBOARD SH 1 FSC Q ISOLATION, RG-01, TP-04___ _ _ _____ __ ___ _ FST Q PC-AOV-NRV20 2027 H7 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SH 1 PIT 2Y DRYWELL VACUUM BREAKER, VBT RF RG-01, TP-08 FSC 2Y___FSO 2Y PC-AOV-NRV21 2027 H7 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SHi1 PIT 2Y DRYWELL VACUUM BREAKER, VBT RF RG-01, TP-08 FSC 2Y___ _ __ ____ _____ __ __ __ FSO 2Y PC-AOV-NRV22 2027 H7 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SH 1 VBT RE DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y____ _ _ __ ___ _ __ ___ ___ _ _ ___ FSO 2Y _ _ _ _ _ _ _ _ _ _ _PC-AOV-NRV23 2027 H7 2 AC 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SHi1 VBT RE DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y Revision 0 Page 228 Cooper Nuclear Station FUifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: PRIMARY CONTAINMENT (PC)VALVE CIC P&ID P&ID 1SI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION ________COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ ____________ FSO 2Y PC-AOV-NRV24 2027 H9 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO Sill VBT RE DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y__ __ __ __ _ __ _ __ _ FSO 2Y PC-AOV-NRV25 2027 119 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO Sill VBT RE DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y___ _ __ __ __ __ __ __ FSO 2Y PC-AOV-NRV26 2027 H3 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SH 2 VBT RE DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y_____ FSO 2Y __ _ _ _ _ _ _ _ _ _ _PC-AOV-NRV27 2027 H3 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SH 2 VBT RE DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y_____ FSO 2Y __ _ _ _ _ _ _ _ _ _ _PC-AOV-NRV28 2027 14 2 AC 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SH 2 VBT RE DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y ESO 2Y Revision 0 Page 229 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: PRIMARY CONTAINMENT (PC)VALVE CIC P&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ ____________ PC-AOV-NRV29 2027 H4 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SH 2 VBT RF DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y FSO 2Y PC-AOV-NRV30 2027 115 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SH 2 VBT RF DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y___ _ ____FSO 2Y PC-AOV-NRV3 1 2027 H6 2 A/C 20 CK-S AO C LT-2 2Y SUPPRESSION CHAMBER TO SH 2 VBT .RF DRYWELL VACUUM BREAKER, PIT 2Y RG-01, TP-08 FSC 2Y_____ __ ___ ____ _ _____FSO 2Y _ _ _ _ _ _ _ _ _ _ _ _PC-CV-13CV 2022 G10 A/C 20 CK-D SA C LJ-1 RE SUPPRESSION CHAMBER SH 1 VBT 2Y VACUUM RELIEF, RG-01, TP-08 FSO 2Y___ __ ___ __ _ _ __ _ _ __ _ _ __ __ __ __FSC 2Y PC-CV-14CV 2022 H10 A/C 20 CK-D SA C LJ-1 RE SUPPRESSION CHAMBER Sil1 VBT 2Y VACUUM RELIEF, RG-01, TP-08 FSO 2Y___ ___FSC 2Y PC-CV-21CV 2022 El E122 IA/C 1/4 CK-S [SA 0/C COL-1i VC OPB R-1N2 SUPPLY TO H1202 MONITORS,TP0 Revision 0 Page 230 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: PRIMARY CONTAINMENT (PC)VALVE CIC P&ID P&ID !SI IST VALVE VALVE ,ACT NORM TEST TEST NOTES/DESCRIPTION ________COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ ____________ _____ ______ _____ ______CCL CVCM _ _ _ _ _ _ _ _ _ _ _ _PC-CV-22CV 2022 D3 2 A/C 1/4 CK-S SA 0/C LJ-1 OPB H2 CAL SUPPLY TO H 2 0 2 SH 2 COF CVCM MONITORS, RG-01, TP-01____CCL CVCM PC-CV-23CV 2022 C3 2 A/C 114 CK-S SA 0/C LJ-1 OPB 02 SUPPLY TO H 2 0 2 MONITORS, SH 2 COF CVCM RG-01, TP-01____ ____CCL CVCM PC-CV-25CV 2022 E5 2 A/C 1/4 CK-S SA 0/C LJ-i OPB N2 SUPPLY TO H 2 O 2 MONITORS, S112 COF CVCM RG-01, TP-01____ ____ ____ ____CCL CVCM PC-C V-26C V 2022 D6 2 A/C 1/4 CK-S SA 0/C LJ-1 OPB H2 CAL SUPPLY TO H 2 0 2 SH 2 COF CVCM MONITORS, RG-01, TP-01___ __ ___ __ ____ __ _____ __CCL CVCM PC-CV-27CV 2022 C6 2 A/C 1/4 CK-S SA 0/C LJ-i OPB 02 SUPPLY TO H 2 0 2 MONITORS, SH 2 COF CVCM RG-01, TP-01___ ___ ___ _____ _ _ __ _ _ ___ _ ___ ___ __ CCL CVCM PC-C V-33CV 2027 A5 2 A/C 3/8 CK-S SA C LJ-1 OPB N2 SUPPLY TO RR-AOV-741AV, SHi1 COF CVCM RG-01, TP-01____ ____ _____CCL CVCM PC-CV-34CV 2027 A4 2 A/C 3/8 CK-S SA C LJ-1 OPB N2 SUPPLY TO RR-AOV-741AV, SHi1 COF CVCM RG-01, TP-01___ __ ___ __ _ _____ __ ____ CCL CVCM PC-CV-35CV 2028 D8 2 A/C 3/8 CK SA C LJ-i OPB X-45D PASSIVE ISOLATION Revision 0 Page 231 Cooper Nuclear Station Fjifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: PRIMARY CONTAINMENT (PC)VALVE CIC P&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION ________COOR CLASS CAT SIZE TYPE TYPE P0S RQMT FREQ ____________ PC-CV-36CV 2028 *D7 2 A/C 3/8 CK SA C LJ-1 OPB X-45D PASSIVE ISOLATION PC-MOV-230MV 2022 F2 2 A 24 BTF MO C LJ-i RF SUPPRESSION CHAMBER SHi1 FSC Q EXHAUST INBOARD ISOLATION,__ _ __ _ __ ___ _ _ _ _ _ _PIT 2Y RG-01 PC-MO V-23 1MV 2022 C2 2 A 24 BTF MO C LJ-1 RF DRYWELL EXHAUST INBOARD SH 1 FSC Q ISOLATION, RG-01___ __ ___ __ _ _ __ _ _____ __ __ __ PIT 2Y PC-MOV-232MV 2022 E8 2 A 24 BTF MO C UJ-1 RF DRYWELL INLET INBOARD SHi1 FSC Q ISOLATION, RG-01__ __ __ _PIT 2Y PC-MO V-233MV 2022 E8 2 A 24 BTF MO C LJ-1 RF SUPPRESSION CHAMBER INLET SHi1 FSC Q INBOARD ISOLATION, RG-01___ _ __ _ __ ____ __PIT 2Y PC-MOV-305MV 2022 G2 2 A 2 GT MO C LU-1 OPB PC-MOV-23OMV BYPASS, RG-01 SHi1 FSC Q___ _ __ _ __ ____ __PIT 2Y PC-MOV-306MV 2022 C2 2 A 2 GT, MO C UJ-1 OPB PC-MOV-231MV BYPASS, RG-01 Sil1 FSC Q___ _ __ _ __ ____ __PIT 2Y PC-MOV-1301MV 2084 C9 2 A 1 GT MO 0 LJ-i OPB SUPPRESSION CHAMBER FSC Q ISOLATION SYSTEM B, RG-01___ ___ ___ __ ___ __ __ ___ _ ___ __ ___ ___ __ ___ __ __ PIT 2Y PC-MOV-1302MV 204 CG T MO 0O- OPB SUPPRESSION CHAMBER Revision 0 Page 232 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: PRIMARY CONTAINMENT (PC)VALVE CIC P&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ ____________ FSC Q ISOLATION SYSTEM B, RG-01______ ______ _ ____ ________________PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _PC-MOV-1303MV 2084 E2 2 A 1 GT MO C LJ-1 OPB SUPPRESSION CHAMBER FSC Q ISOLATION SYSTEM A, RG-01___ __ ___ __ _ _ __ __ __PIT 2Y PC-MOV-1304MV 2084 E3 2 A 1 GT MO C LJ-1 OPB SUPPRESSION CHAMB3ER FSC Q ISOLATION SYSTEM A, RG-01___ __ __ ___ _ __ __ ____ __PIT 2Y PC-MOV-1305MV 2084 C2 2 A 1 GT MO C LJ-I OPB DRYWELL ISOLATION SYSTEM FSC Q A, RG-01______ ~~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _PC-MOV-1306MV 2084 C3 2 A 1 GT MO C LJ-1 OPB DRYWELL ISOLATION SYSTEM FSC Q A, RG-01_____ ______ ~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _PC-MO V-1308MV 2084 G9 2 A 1 GT MO C LJ-i OPB BLEED ISOLATION FOR FSC Q SUPPRESSION CHAMBER, RG-01_____ _____ ~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _PC-MOV-1310MV 2084 B9 2 A 1 GT MO C LJ-1 OPB BLEED ISOLATION FOR FSC Q DRYWELL, RG-01______ _ ____ ___ _ ______ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _PC-MOV-1311iMV 2084 C8 2 A 1 GT MO 0 LJ-1 OPB DRYWELL DILUTION SUPPLY FSC Q ISOLATION SYSTEM B, RG-01_____ _ ____ __ __ ____ _ _____ ______ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _Revision 0 Pg 3 Page 233 Cooper Nuclear Station F~ifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: PRIMARY CONTAINMENT (PC)VALVE CIC P&ID P&ID IS1 1ST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION ___________COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ ____________ PC-MOV-1312MV 2084 C9 2 A 1 GT MO 0 LJ-1 OPB DRYWELL DILUTION SUPPLY FSC Q ISOLATION SYSTEM B, RG-01 PIT 2Y Revision 0 Page 234 Revision 0 Page 234 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: RADIATION MONITORING (RMV)VALVE CIC P&ID P&ID ISI 1ST VALVE VALVE IACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE jTYPE POS RQMT FREQ_____________ RMV-AOV- 10AV 2022 D3 2 A 3/4 GB AO 0 LU-i1 OPB RM-4 CONTAIN4MENT ISOLATION, Sill FSC Q INBOARD, RG-0l, TP-04 FST Q___ __ __ __ ____ __PIT 2Y RMV-AOV-l11AV 2022 D2 2 A 3/4 GB AO 0 LU-I OPB RM-4 CONTAINMENT ISOLATION, SHi1 FSC Q OUTBOARD, RG-01, TP-04 FST Q___ __ __ __ ____ __PIT 2Y RMV-AOV-12AV 2022 E3 2 A 3/4 GB AO 0 LJ-i OPB RM-4 CONTAINMENT ISOLATION, SHl 1 FSC Q INBOARD, RG-0l, TP-04 FST Q___ __ __ ___ _ __ __PIT 2Y RMV-AOV--13AV 2022 E2 2 A 3/4 GB AO 0 LJ-i OPB RM-4 CONTAINMENT ISOLATION, SHll FSC Q OUTBOARD, RG-0l, TP-04 FST Q PIT 2Y Revision 0 Pg 3 Page 235 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: RADIOACTIVE WASTE (RW)1P&ID ISI 1ST VALVE VALVE IACT NORM TEST TEST VALVE CIC P&ID COOR CLASS CAT SIZE TYPE jTYPE P0S RQMT FREQ NOTES/DESCRIIPTION RW-AOV-AO82 2038 G5 2 A 3 GT AO 0 LJ-i OPB DRYWELL FLOOR DRAIN SUMP FSC Q DISCHARGE, RG-01, TP-04 EST Q________ ______ _ ____ ___ _ ______ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RW-AOV-AO83 2038 G5 2 A 3 GT AO 0 LJ-i OPB DRYWELL FLOOR DRAIN SUMP FSC Q DISCHARGE, RG-01, TP-04 FST Q________ ______ ____ ___ __ ____ ______ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RW-AOV-AO94 2028 Gil1 2 A 3 GT AO 0 LU-i1 OPB DRYWELL EQUIPMENT DRAIN FSC Q SUMP DISCHARGE, RG-O1, TP-04 FST Q__________ ___ _____ ______ __ __ ______ PIT 2Y_ _ _ _ _ _ _ _ _ _ _ _ _RW-AOV-AO95 2028 G12 2 A 3 GT AO 0 LI-1 OPB DRYWELL EQUIPMENT DRAIN FSC Q SUMP DISCHARGE, RG-01, TP-04 FST Q______________ _____ ______ PIT 2Y_ _ _ _ _ _ _ _ _ _ _ _ _RW-CV-58CV 2005 F10 A C 2 CK-S SA C FSO Q Z SUMP PUMP A DISCHARGE S_______ H 2 _____________FSC Q' CHECK, RG-01 RW-CV-59CV 2005 F10 A C 2 CK-S SA C FSO Q Z SUMP PUMP B DISCHARGE SH 2 FSC Q CHECK, RG-01 OG-CV-8CV "2037 D9 A C 1/2 CK-P SA C COF CVCM OFFGAS HOLD-UP DRAIN LINE CCD CVCM VENT, RG-01, TP-01 Revision 0 Pg 3 Page 236 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: RADIOACTIVE WASTE (RW)P&ID ISI 1ST VALVE VALVE ACT NORM TEST TEST VALVE CIC P&ID_ COOR CLASS jCAT SIZEJTYPE TYPE POSj RQMT FREQ J NOTES/DESCRIPTION OG-CV- 12CV 2037 D9 A C 1/2 CK-P SA C COF CVCM IOFFGAS HOLD-UP D1AIN LINE______________ ___ ____J__________________ ___JCCD CVCM VENT, RG-01, TP-01 Revision 0 Page 237 Revision 0 Page 237 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: REACTOR CORE ISOLATION COOLING (RCIC)____ _

  • NOR IP&ID ISI IST VALVE IVALVE ACT M TEST TEST VALVE CIC P&D COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION RCIC-AOV-AO34 2041 [H110 2 B 1 GB AO 0 FSC Q RCIC STEAM LINE DRIPLEG DRAIN IjFST Q RG-01, TP-04___ __ __ __ __ ___
  • __ __.1___ __ ______ ___ __ ______.1______

__ __IL __ __ PI 2 __ __ __ __ __ __PIT__ ___2Y_RCIC-CV-1 0CV 2043 115 12 1C F6 CK-S S C 1COF CVCM 1ECST SUPPLY TO RCIC PUMP, RG-_______ ________ IKS .1. CCL IICVCM 101 RCIC-CV-11CV 2043 1110 [2 1C [6 CK-S 1SA C 1COD CVCM 1RCIC SUPPLY FROM SUPPRESSION _________ i I J ___[ j __ ___CCD CVCM ]CHAMBER, RG-01 RCIC-CV-12CV 2043 G9 2 C 2 CK-S SA C COF CVCM VACUUM PUMP DISCHARGE TO[CCL CVCM SUPPRESSION CHAMBER, RG-01,_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ TP-01 RCIC-CV- 13CV [04 D5 2 C 2 [CK-P SA] C COF CM MIMUM FLOW LINE, RG-01______ ______ ____ _ __ __ __ ii L .___ ____ CCL CVCM _ _ _ _ _ _ _ _ _ _ _ _RCIC-CV-15CV 2043 D8 2 A/C 8 CK-S SA C LJ-1 OPB RCIC TURBINE EXHAUST TO iFCOF CVCM ISUPPRESSION CHAMBER, RG-01___ ___ __ ___ ___ _ ___ _ ___ ___ _ ___ __ __ ___ _ CCL CVCM RCIC-CV-16CV 2043 F9 2 C 2 CK-S SA -C COF 1CVCM VACUUM PUMP DISCHARGE TO CCL CVCM SUPPRESSION CHAMBER, RG-01, RCIC-CV-18CV 2043 A8 12 C 2 CK-P SA C CO CVM CNEST SUPPLY TO RI Revision 0 Page 238 Revision 0 Page 238 Cooper Nuclear Station Ffifh Interval lnservice Testing Pro gram for Pumps and Valves SYSTEM: REACTOR CORE ISOLATION COOLING (RCIC)NOR P&ID ISI IST VALVE VALVE ACT M TEST TEST VALVE CIC P&ID COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION RCIC-CV-19CV 2043 A8 A C 2 CK-P SA -C COF CVCM -CONDENSATE SUPPLY TO RCIC CCF CVCM PRESSURE MAINTENANCE, RG-01, RCIC-CV-20CV 2043 118 A irC 2 CK-P ][SA [C ][COD [CVCM [RCIC CONDENSATE PLivIP__________ ___ ___ _______ ____ ________ ___J[CCD [CVCM [DISCHARGE, RG-01, TP-01 RCCC- 1C 2043 H 17 1 2 1C 2 [CK-P ][SA IC [COD [CVCM [RCIC CONDENSATE PM_______________ ___I ____B.___ ____ [ I____I____CCD CVCM [DISCHARGE, RG-01 rRCIC-CV-22CV 12043 [F9 1 2 [C 1 1/2 [CK-L ][SA C ][COD CVCM IRCIC VACUU BREAKER,RG0 -] ___ [ J_____ ____ ____ I J[___ I ][CCD CVCM __________ RCIC-CV-23CV 12043 IFO10 2 C 1 1/2A CK-L ][SA IC COD CVCM RCIC VACUUM BREAKER, RG-01______________ ___I i ____ I J____ ____CCD CVCM __________ RCIC-CV-24CV 12043 F9 2 C 1 1/2 CK-L SA C COD CVCM CIVACUUM BREAKER, RG-01___________I ___ [___ ____J[___ ____ ____ ___ ____CCD CVCM _____________ RCIC-CV-25CV 12043 [F10 C 1 'A1/2 CK-L ][SA IC ][COD ICVCM [RCIC VACUU BREAKER, RG-01__________ ___ [___ ___Jr________[___ ____[ J[CCD CVCM __________ RCIC-CV-26CV 2043 B8 1 A/C 4 CK-S SA 0 LJ-1 RE INJECTION CHECK VALVE, iFSO RE ROJ-09, RG-01__________________ ______ ______________I, _____ _______ _______I______ ______ii_______ ______ ______________FSC________RF___ Revision 0 Pg 3 Page 239 Cooper Nuclear Station F/ifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: REACTOR CORE ISOLATION COOLING (RCIC)NOR P&ID ISI IST VALVE VALVE ACT M TEST TEST VALVE CIC P&I COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION ___ ___ __ ___ __ __ __ __ __ ____ __ i __ __ ___ __ __ _ _ ]] _ _ _ ][ [ PiT ][ 2Y _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _RCIC-HOV-HO 11 2041 G8 A B 2 1/2 GL HO C SKID Q ]RCIC TU GOV V, RG-01, TP-05 RCIC-MOV-MO14 2041 G8 2 B f2 x3 GL HO C SKID fQ I RCIC TU TRIP & THROTTFILE V, RCIC-MOV-MO 15 2041 D7 1 A 3 GT MO 0 LJ-I OPB IRCIC STEAM INBOARD 111FSC Q/CS BISOLATION, CSJ-10,RG0 RCIC-MOV-MO16 2041 D9 1 A 3 GT MO 0 LJ-i OPB RCIC STEAM OUTBOARD FSC Q ISOLATION, RG-01 PIT 2Y RCIC-MOV-MO018 2043 H4 2 B 6 GT MO 0 FSO Q RCIC SUPPLY FROM CONDENSATE FSC Q STORAGE, RG-01_____ _____ _____ _____PIT 2Y_ _ _ _ _ _ _ _ _ _ _ _ _RCIC-MOV-MO20 2043 B7 2 B 4 GT MO 0 PIT 2Y RCIC PUMP DISCHARGE, RG-01 RCIC-MOV-MO21 2043 B8 2 B 4 GT MO C FSO Q RCIC INJECTION TO REACTOR,___ __ __ __ __ ___ _ __ _ __ __ __ _ __ __ ___ __ __ __ __ __ PIT 2Y RG-01 RCIC-MOV-MO27 2043 C7 2 B ]GB MO C FSO Q 1RCIC PUMP MINIMUM FLOW___PIT 2Y CHAMBER, RG-01 Revision 0 Page 240 Revision 0 Page 240 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: REACTOR CORE ISOLATION COOLING (RCIC)[ NOR P&ID ISI IST VALVE VALVE ACT M TEST TEST VALVE CIC P&ID COOR CLASS CAT [SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION RCIC-MOV-MO30 [2043 1D1 2 B 4 GB IMO 1C PIT J2Y 1RCIC TEST RETURN ROOT,____ ____ ____ _ I____ .1 1 _ ____ ____ __ ___ ___ ___ ___ 1 _ ____ ___ __ PASSIVE, PAS I RE 0 RCCMVM3 20431 El1 A 1B 4[ 1GT MO] C PIT ]2Y 1RCIC TEST RETURN SHUTOFF,____ ____ ____ _ _ ___ ____ I.____ _ I ____31 ____ 1 _ ___ _ __ __ _ ____ ___ __ PASSIVEPASSVE, -G01 RCIC-MOV-MO41 2043 Hl0 2 B 6 GT MO C FSO Q RCIC SUPPLY FROM SUPPRESSION _____ _____ ____ ___ _ __ _ _ ___ ____ ____ ____ PIT [2Y_ _ _ _ _ _ _ _ _ _RCIC-MO V-MO 131 2041 1G9 F2 1B F3 1 GB MO 1~C FSO 1 RCIC STEAM SUPPLY TO RCIC_______________ _____I _____ I______ I ____ I ______ I_____PIT___2Y _TURBPII2YETRB RE, -G01 RCIC-MO V-MO 132 2043] E4 [2 1B F2 1 B MO C FSO 1 AUXILIARY COOLING SUPPLY,___ ___ ___ ___ __ _ _ __ __ ii___ __ __ ___ [ _ __ PIT ]2Y RG-01_ __ __ __ _RCCR-20 20431 D6 F2 1Df 8 1RD SA 1[C 1RD ]5Y 1 EXHAUST LINE RUPTURE DISC, CI-V10ORV 20431 G5 F2 1C 1 RV SA] C 1 RVT [APP I]RCIC PUMP SUCTION RELIEF,____ ___ ___ _ _ ___J _ ___[ __ __ __ _ _ ___1 _ ___ ___ _ I. _ ___J _ ___j [ J RG-01 RCIC-RV-11RV 2043 G6 2 ]C 1 RV SA C RVT A(PPI ]CI AUX~ILAY COOLING RCIC-V-37 2043 D9 2 A/C 8 -CK MA 0 LJ-i1 OPB RCIC TURBINE EXHAUST TO___________ ____ __ __ ____ __ _ ____ ____ ___ ____ CCL CVCM ISOLATION, RG-01 Revision 0 Pg 4 Page 241 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: REACTOR EQUIPMENT COOLING (REC)VALVE CIC 1P&ID TP&ID ISI IST 1VALVE] VALVE ACT] NORM] TEST TTEST [ NOTES/DESCRIPTION ________I jCOOR CLASS CAT SIZE jTYPE TYPE jP0S RQMT jFREQ REC-CV-10CV 2031 F2 3 C 8 CK-T SA 0, FSO Q REC-P-A DISCHARGE, RG-01 SH 2 ___ __FSC Q REC-CV- 11 CV 2031 F3 3 C 8 CK-T SA 0 FSO Q REC-P-B DISCHARGE, RG-0 1__ _ _ _ _ _ SH 2 __ _FSC Q REC-CV-12CV 2031 F4 3 C 8 CK-T SA 0 FSO Q REC-P-C DISCHARGE, RG-01 S______ H 2 FSC Q REC-CV-13CV 2031 F5 3 C 8 CK-T SA 0 FSO Q REC-P-D DISCHARGE, RG-01 SH 2 __ _FSC Q _ _ _ _ _ _ _ _ _ _ _REC-CV- 1 6CV 2031 H2 3 C 12 CK-S SA 0 COF CVCM NON-CRITICAL HEADER RETURN S_______ H 2 ___ ___CCF CVCM TO REC PUMPS, RG-01, TP-01, TP-06 REC-MOV-694MV 2031 H2 3 B 4 GT MO 0 PIT 2Y PASSIVE CRITICAL LOOP RETURN S_______ H 2 _______ ___CROSSTIE, RG-01 REC-MOV-695MV 2031 C5 3 B 4 GT MO 0 PIT 2Y PASSIVE CRITICAL LOOP SUPPLY S H2 ___ ______ ___CROSSTIE, RG-01 REC-MOV-697MV 2031 H3 3 B 6 GT MO 0 FSO Q NORTH CRITICAL LOOP RETURN,______ SH 2 PIT 2Y RG-01 REC-MOV-698MV 2031 Hi1 B 6 GT MO 0 FSO Q SOUTH CRITICAL LOOP RETURN, S_____ H 2 PIT 2Y RG-01 REC-MOV-700MV 2031 B2 3 B 10 GT MO 0 FSC Q NON-CRITICAL LOOP SUPPLY__ _ _ _ _ _ SH 2 _ _ __ _ __ _ _ _ _ _ __ _ __ _ __ PIT 2Y SHUTOFF, RG-01 Revision 0 Page 242 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: REACTOR EQUIPMENT COOLING (REC)VALVE CIC P&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION ________COOR C LASS CAT SIZE TYPE _TYPE POS RQMT FREQ REC-MOV-702MV 2031 A4 2 A 8 GT MO 0 LJ-1 OPB DRYWELL SUPPLY ISOLATION SH 2 FSC Q VALVE, RG-01 S_ __ ____ PIT 2Y REC-MOV-709MV 2031 G3 2 A 8 GT MO 0 LJ-1 OPB DRYWELL RETURN ISOLATION SHf1 FSC Q VALVE, RG-01 PIT 2Y REC-MOV-7 11 MV 2031 D4 3 B 6 GT MO 0 FSC Q NORTH CRITICAL LOOP SUPPLY, SH 2 FSO Q RG-01___ _ __ _ _ __ _ ___ __ __ __ __ __ PIT 2Y REC-MOV-712MV 2031 Dl B 12 BFL MO 0 FSC Q REC HEAT EXCHANGER A SH 2 _____ ___ ___PIT 2Y OUTLET, RG-01 REC-MOV-713MV 2031 C1 B 12 BFL MO 0 FSC Q REC HEAT EXCHANGER B OUTLET, Sf 2 _ __ PIT 2Y RG-01 REC-MOV-714MV 2031 C5 3 B 6 GT MO 0 FSC Q SOUTH CRITICAL LOOP SUPPLY, SH 2 FSO Q RG-01___ __ __ ___ _ __ _ ___ _ __ __ PIT 2Y REC-MOV-1329MV 2031 B3 3 B 8 GT MO 0 FSC Q AUXILIARY RADIOACTIVE WASTE SHf2 PIT 2Y BUILDING SUPPLY, RG-01 Revision 0 Page 243 Revision 0 Page 243 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: REACTOR FEEDWATER (RF)P&ID ISI IST VALVE VALVE ACT NORM TEST TEST VALVE CIC P&ID COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION RF-CV-13CV 2044 B9 1 A/C 18 CK-S SA 0 LU-i RF FEEDWATER LINE B TO REACTOR FSC RF OUTBOARD, ROJ-03, RG-.01, TP-01_____ ______ ~~FSO OP __ _ _ _ _ _ _ _ _ _ _ _RF-CV- 14CV 2044 Cl 10 A/C 18 CK-S SA 0 U -i1 RF FEEDWATER LINE B TO REACTOR FSO OP INBOARD, ROJ-03, RG-01_______________ ______ _____ FSC RF_ _ _ _ _ _ _ _ _ _ _ _ _ _RF-CV-15CV 2043 A9 1 A/C 18 CK-S SA 0 UJ-i RF FEEDWATER LINE A TO REACTOR FSC RF OUTBOARD, ROJ-03, RG-01, TP-01 FSO OP __ _ _ _ _ _ _ _ _ _ _ _RF-CV- 16CV 2043 Cl 10 A/C 18 CK-S SA 0 U -i1 RF FEEDWATER LINE A TO REACTOR FSO OP INBOARD, ROJ-03, RG-01 FSC RE Revision 0 Pg 4 Page 244 Cooper Nuclear Station Ffifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: REACTOR RECIRCULATION (RR)P&ID 1ISI IST VALVE VALVE ACT NORM TEST TEST VALVE CIC [P&ID COOR C LASS CAT SIZE TYPE [TYPE POS RQMT FREQ NOTES/DESCRIPTION RR-AOV-740AV 2027 A3 1 A 3/4 GB AO 0 LI-i OPB SP-1 OUTBOARD ISOLATION, SHi1 FSC Q RG-0l, TP-04 FST Q PIT 2Y RR-AOV-741AV 2027 A7 1 A 3/4 GB AO 0 LI-i OPB SP-1 INBOARD ISOLATION, RG-01, SHi1 FSC Q TP-04 FST Q___ _ __ _ __ __PIT 2Y RR-MOV-MO53A 2027 B7 1 B 28 GT MO 0 FSC CS RR PUMP A DISCHARGE SHi ________ PIT 2Y CSJ-04, RG-01 RR-MOV-MO53B 2027 C4 1 B 28 GT MO 0 FSC CS RR PUMP B DISCHARGE__________ SHi 1 ___ _______ ________ PIT 2Y CSJ-04, RG-01 Revision 0 Page 245 Revision 0 Page 245 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: REACTOR WATER CLEANUP (RWCU)IP&1ID ISI IST VALVE VALVE ACT NORM TEST TEST VALVE CIC jP&ID COOR CLASS CAT SIZE TYPE jTYPE POS RQMT FREQ NOTES/DESCRIPTION RWCU-CV-15CV 2042 C4 1 A/C 4 CK-S SA 0 LU-1 RE RWCU RETURN TO REACTOR SHi1 FSC RE VESSEL, ROJ-04, RG-01, TP-01, TP-06___ _ _ __ __ __ __ __ FSO RE RWCU-MOV-MO15 2042 E2 1 A 6 GT MO 0 LU-i OPB RWCU SUPPLY INBOARD SHi1 FSC RF ISOLATION, ROJ-05, RG-01________ _____ _ ___ ___ __ ____ ______ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RWCU-MOV-MO018 2042 B4 1 A 6 "GT MO 0 LJ-1 OPB RWCUJ SUPPLY OUTBOARD Sill FSC RE ISOLATION, ROJ-05, RG-01 PIT 2Y Revision 0 Page 246 Revision 0 Page 246 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RIHR)VALVE CIC P&ID COOR CLASS CAT SIZE TYPE TYPE P0S RQMT FREQ T NOTES/DESCRIPTION RIIR-CV-10CV 2040 F3 2 C 3 CK-S SA C COD CVCM RHR PUMP A MINIMUM FLOW, SH 1 __ ___ CCD CVCM RG-01 RHR-CV-1 1CV 2040 F8 2 C 3 CK-S SA C COD CVCM RI{R PUMP B MINIMUM FLOW, S_______ H 2 __________CCD CVCM RG-0l RH-R-CV-12CV 2040 H6 2 C 3 CK-S SA C COD CVCM RHR PUMP C MINIMUM FLOW,________Sil 1 __ ___ __ CCD CVCM RG-01 RHR-CV-13CV 2040 H7 2 C 3 CK-S SA C COD CVCM RH-R PUMP D MINIMUM FLOW, S_______ H 2 ___ ______CCD CVCM RG-01 RHR-CV-14CV 2040 F5 2 C 16 CK-T SA C FSO Q RLER PUMP A DISCHARGE, RG-01__ _ _ _ _ _ SHi 1 _ _ _ _ __ _ FSC Q RHR-CV- 1 5CV 2040 F8 2 C 16 CK-T SA C ESO Q RHR PUMP B DISCHARGE, RG-0 1 SH 1-_ _ __ _ ___ _ FSC Q RHR-CV-16CV 2040 H5 2 C 16 CK-T SA C FSO Q RHR PUMP C DISCHARGE, RG-01 SHl ___________ FSC Q RHiR-CV-17CV 2040 H8 2 C 16 CK-T SA C FSO Q RIHR. PUMP D DISCHARGE, RG-01__ _ _ _ _ _ SHl _ _ __ _ FSC Q RHIR-CV-18CV 2040 A9 A C 4 CK-S SA C COF CVCM RH-R LOOP B OUTBOARD SH 2 CCF CVCM PRESSURE MAINTENANCE ____ ___ ____ _ _ ___ ____ _ _ ___SUPPLY, RG-01, TP-01 RI-R-CV-19CV 2040 A9 2 C 4 CK-S SA C COF CVCM RHR LOOP B INBOARD PRESSURE Revision 0 Page 247 Cooper Nuclear Station FUifh Interval Inservice Testing Program for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RIIR)P&ID ISI IST VALVE VALVE ACT NORM TEST TEST VALVE CIC P&ID COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION SH 2 CCF CVCM MAINTENANCE SUPPLY, RG-01, TP-CCD/ CVCM 01 CCR RiHR-CV-24CV 2040 C7 A C 4 CK-S SA C COF CVCM RHR LOOP A OUTBOARD SH 2 CCF CVCM PRESSURE MAINTENANCE __ _ _______ ____ ___SUPPLY, RG-01, TP-01 RHIR-CV-25CV 2040 C7 2 C 4 CK-S SA C COF CVCM R!HR LOOP A INBOARD PRESSURE SHi1 CCF CVCM MAINTENANCE SUPPLY, RG-01, TP-CCD/ CVCM 01______ ~~~~~~CCR__ _ _ _ _ _ _ _ _ _ _ _RHR-CV-26CV 2040 BI 10 A/C 24 CK-S SA C LT-2 OPE LOOP A INJECTION LINE SF11 FSO RF TESTABLE CHECK, PRESSURE FSC RF ISOLATION VALVE, ROJ-1 1, RV-05,____PIT RE RG-01 RHlR-CV-27CV 2040 B4 1 A/C 24 CK-S SA C LT-2 OPB LOOP B INJECTION LINE SF1 1 FSO RE TESTABLE CHECK, PRESSURE FSC RE ISOLATION VALVE, ROJ-1 1, RV-05,____PIT RE RG-01 RHR-MOV-MO12A 2040 E3 2 B 16 GT MO 0 FSO Q RHR HEAT EXCHANGER A SH 1 PIT 2Y OUTLET, RG-01 RH{R-MOV-MO 12B 2040 F10 B 16 GT MO 0 FSO Q RIR HEAT EXCHANGER B S_____ H 2 ___ __PIT 2Y OUTLET, RG-01 RitIR-MOV-MO13A 2040 FO 2 B 20 GT MO 0 FSO Q RUR PUMP A SUCTION FROM SF1 1 FSC Q SUPPRESSION CHAMBER, RG-01 Revision 0 Page 248 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RHR)P&ID IS I5 ST VALVE VALVE fACT NORM TEST TEST VALVE CIC [P&ID COOR {CLASS CAT SIZE TYPE TYPE P0S RQMT "FREQ NOTES/DESCRIPTION ___________ ___ _____ _ ___ ______ PIT 2Y_ _ _ _ _ _ _ _ _ _ _ _ _RHR-MOV-MO13B 2040 F4 2 B 20 GT MO 0 FSO Q RHIR PUMP B SUCTION FROM SH 2 FSC Q SUPPRESSION CHAMBER, RG-01_____ ______ ~~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RHR-MOV-MO13C 2040 F10 B 20 GT MO 0 FSO Q RHR PUMP C SUCTION FROM SH 1 FSC Q SUPPRESSION CHAMBER, RG-01______ _____ ~~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RHR-MOV-MO 13D 2040 F3 2 B 20 GT MO 0 FSO Q RH-R PUMP D SUCTION FROM SH12 FSC Q SUPPRESSION CHAMBER.. RG-01___ __ __ ___ ____ __PIT 2Y RHR MOV-MO15A 2040 F9 2 B 20 GT MO C FSC Q RHR PUMP A SDC SUCTION, RG-0l SHl 1 _ __ _ _ _ PIT 2Y RHR-MO V-MO 15B 2040 F4 2 B 20 GT MO C FSC Q RI-R PUMP B SDC SUCTION, RG-01I SH 2 __ __ _ _ _ PIT 2Y _ _ _ _ _ _ _ _ _ _RHIR-MOV-MOI15C 2040 G8 2 B 20 GT MO C FSC Q RHR PUMP C SDC SUCTION, RG-01 Sil 1 ____ ___ PIT 2Y______ _____________ RHR-MOV-MO15D 2040 G5 2 B 20 GT MO C FSC Q RI-R PUMP D SDC SUCTION, RG-01 SH 2 ___PIT 2Y RI-IR-MOV-MO 16A 2040 E7 2 B 4 GT MO 0 FSO Q PUMP AAND CMINIMUM FLOW, SH 1 FSC Q RG-01__ _ __PIT 2Y RHR-MOV-MO16B 24 E7 2 B 4 GT MO O j FSO Q PUMP B AND DMIMINUM FLOW, Revision 0 Page 249 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RHR)P&ID ISI IST VALVE VALVE ACT NORM TEST TEST VALVE CIC P&ID COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION SH 2 FSC Q RG-01___ __ __ ___ ____ ___ __ __PIT 2Y RH{R-MOV-MO17 2040 C8 1 A 20 GT MO C LJ-1 OPB RIIR SDC SUPPLY OUTBOARD SH11 LT-2 PB PRESSURE ISOLATION VALVE, FSC CS CSJ-05, RV-05, RG-01___ ___ ___ ___ _ ______ __ __ _ _ ___ __ __PIT 2Y RHR-MOV-MO18 2040 C10 A 20 GT MO C LT-2 OPB RUR SDC SUPPLY INBOARD 5111 FSC CS PRESSURE ISOLATION VALVE,___________ _________________PIT 2Y CSJ-05, RV-05, RG-01 RHR-MOV-MO20 2040 H3 2 B 20 GT MO C PIT 2Y RHR PASSIVE CROSSHEADER __________ SHi 1___ ______ SHUTOFF, RG-01 RHR-MOV-MO25A 2040 B8 1 A 24 GT MO C LJ-1 OPB RHR LOOP A INJECTION INBOARD SH 1 LT-2 PB ISOLATION, RV-05, RG-01 FSO Q FSC Q___ __ ___ __ _ _ __ __ __ __ __PIT 2Y RHR-MOV-MO25B 2040 B5 1 A 24 GT MO C LJ-1 OPB RHR LOOP B INJECTION INBOARD 5112 LT-2 PB ISOLATION, RV-05, RG-01 FSO Q FSC Q___ __ ___ __ _ _ __ _ _ __ __ __PIT 2Y RHR-MOV-MO26A 24 A7 2 B 10 GT MO C FSO Q DRYWELL SPRAY LOOP A Revision 0 Page 250 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RHLR)P&ID ISI IST VALVE VALVE [ACT NORM TEST TEST VALVE CIC _P&ID COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION S111 FSC Q OUTBOARD ISOLATION, RG-01_____ _____ ____ ______ ______ IT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RHR-MOV-MO26B 2040 B6 2 B 10 GT MO C FS0 Q DRYWELL SPRAY LOOP B SH 2 ESC Q OUTBOARD ISOLATION, RG-01 PIT 2Y RHR-MOV-MO27A 2040 B8 2 B 24 ANG MO 0 FSO Q LOOP A INJECTION OUTBOARD SHi 1 __ PIT 2Y THROTTLE, RG-01 RHR-MOV-MO27B 2040 B6 2 B 24 ANG MO 0 FSO Q LOOP B INJECTION OUTBOARD SH 2 ___ __PIT 2Y THROTTLE, RG-01 RI-R-MOV-MO31A 2040 A8 2 A 10 GT MO " C LJ-1 OPB DRYWELL SPRAY LOOP A SHi1 FSO Q INBOARD ISOLATION, RG-01 FSC Q_____ _____ ______ ~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RHR-MOV-MO3 lB "2040 B5 2 A 10 GT MO C LJ-i OPB DRYWELL SPRAY LOOP B SH 2 FSO Q INBOARD ISOLATION, RG-01 FSC Q PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RHR-MOV-MO34A 2040 E7 2 B 18 GB MO C FSO Q SUPPRESSION CHAMBER COOLING SHi1 FSC Q LOOP A INBOARD THROTTLE,________PIT 2Y RG-01 RHR-MOV-MO34B 2040 E6 2 B 18 GB MO C FS0 Q SUPPRESSION CHAMBER COOLING SH 2 FSC Q LOOP B INBOARD THROTITLE,___ __ __ __ __ __ _ __ _ _ __ _ __ _ _ __ _ __ __ __ _ __ __ PIT 2Y RG-01 Revision 0 Page 251 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RHR)P&D IS ST VALVE VAVEACT NORM TEST TEST VALVE CIC [P&ID COOR CLASS CAT SIZE TYPE [TYPE POS RQMT FREQ NOTES/DESCRIPTION RHR-MOV-MO38A 2040 D9 2 A 6 GB MO C UI_- OPB SUPPRESSION CHAMBER SPRAY SH 1 FSO Q LOOP A INBOARD THROTTLE, FSC Q RG-01_____ ______ _____ ______ PIT 2Y_ _ _ _ _ _ _ _ _ _ _ _ _RHR-MOV-MO38B 2040 D6 2 A 6 GB MO C LI-1 OPB SUPPRESSION CHAMBER SPRAY SH 2 ESO Q LOOP B INBOARD THROTTLE, FSC Q RG-01_____ ______ ______ ~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RHR-MOV-MO39A 2040 D6 2 B 18 GT MO C FSO Q SUPPRESSION CHAMBER COOLING SHi1 FSC Q LOOP A OUTBOARD ISOLATION, PIT 2Y RG-01 RHR-MOV-MO39B 2040 D7 2 B 18 GT MO C FSO Q SUPPRESSION CHAMBER COOLING SH 2 FSC Q LOOP B OUTBOARD ISOLATION,____PIT 2Y RG-01 RHR-MOV-MO57 2040 H2 2 B 4 GB MO C LT-2 RF RHR DISCHARGE TO RAD WASTE SH 2 FSC Q INBOARD THROTTLE, RG-01______ ~~~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _RHR-MOV-M065A 2040 F3 2 B 16 GT MO 0 PIT 2Y RHR HEAT EXCHANGER A INLET, SH 1__ __ RG-01 RHR-MOV-MO65B 2040 Fl0 2 B 16 GT MO 0 PIT 2Y RHR HEAT EXCHANGER B INLET, SH 2 _ _ __ _ __ _ __ _RG-01 RHR-MOV-MO66A 24 F4 2 B 20 GB MO O FSO Q RIIR HEAT EXCHANGER A BYPASS SH1FSC Q THROTTLE, RG-01 PIT 2Y Revision 0 Page 252 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RHR)FIP&ID ISI 1ST VALVE VALVE ACT NORM TEST TEST VALVE.CIC _P&ID COOR CLASS CAT SIZE TYPE _TYPE POS RQMT FREQ NOTES/DESCRIPTION RHR-MOV-MO66B 2040 F9 2 B 20 GB MO 0 FSO Q RHR HEAT EXCHANGER B BYPASS SH 2 FSC Q THROTTLE, RG-01__ _ _ _ _PIT 2Y RIIR-MOV-MO67 2040 H2 2 B 4 GT MO C LT-2 RF RHR DISCHARGE TO RAD WASTE SH 2 FSC Q OUTBOARD SHUTOFF, RG-01_____ ____ ______ __ __ ______ PIT 2Y_ _ _ _ _ _ _ _ _ _ _ _ _RIHR-MOV-MO0166A 2041 H2 2 A 1 GB MO C LJ-i OPB RHR HEAT EXCHANGER A VENT______ ____ __ ___PASSIVE RHR-MOV-MO166B 2041 H3 2 A 1 GB MO C UJ-i OPB RHR HEAT EXCHANGER B VENT______PASSIVE RI-IR-MO V-MO 1 67A 2041 H2 2 A 1 GB MO C UJ- 1 OPB RH-R HEAT EXCHANGER A VENT______PASSIVE RHR-MOV-MO167B 2041 H3 2 A 1 GB MO C UJ-i OPB RHR HEAT EXCHANGER B VENT_____ _____ ___ __ ____ __ ___ PASSIVE RHR-MOV-MO274A 2040 Bl 10 A 2 GB MO C LT-2 OPB RHR-CV-26CV PASSIVE BYPASS SHi1 PIT 2Y PRESSURE ISOLATION VALVE,____ ___ _ __ __ _ ___RV-05, RG-01 RHR-MOV-M0274B 2040 B4 1 A 2 GB MO C LT-2 OPB RHR-CV-27CV PASSIVE BYPASS SH 2 PIT 2Y PRESSURE ISOLATION VALVE,____ ___ __ ____ ___ _ __ ___ RV-05, RG-01 RHR-MOV-920MV 201 D2 B 3 GT M C FSC CS STEAM SUPPLY TO AOG PIT 2Y UPSTREAM SHUTOFF, CSJ-06, Revision 0 Page 253 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RHR)VALVE GIG P&ID COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION ____ ___ ____ __ __ _ ___RG-0 1 RHR-MOV-921MV 2041 D1 A B 3 GT MO C FSC CS STEAM SUPPLY TO AOG PI!T 2Y DOWNSTREAM SHUTOFF, CSJ-06,____ ___ ____ __ __ _ ___RG-01 RHR-RV-10RV 2040 F8 2 C 1 RV SA C FVT APP I RHR PUMP A SUCTION RELIEF, S111 RG-01 RHR-RV-l1RV 2040 F5 2 C 1 RV SA C RVT APP I RJIR PUMP B SUCTION RELIEF, 5112 RG-01 RHR-RV-12RV 2040 118 2 C 1 RV SA C RVT APP I RHR PUMP C SUCTION RELIEF, 5111 RG-01 RHR-RV-13RV 2040 145 2 C 1 RV SA C RVT APP I RHR PUMP D SUCTION RELIEF, SH 2 RG-01 RHR-RV-14RV 2040 C4 2 C 1 RV SA C RVT APP I RHR LOOP A SUPPLY RELIEF, 5111 RG-01 RHR-RV-15RV 2040 C9 2 C 1 RV SA C RVT APP I RIIR LOOP B SUPPLY RELIEF,__ _ _ _ _ SH 2 __ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ RG-01 Revision 0 Pg 5 Page 254 Cooper Nuclear Station Fifthi Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: RESIDUAL HEAT REMOVAL (RHLR)VALVE CIC P&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ RHR-RV- 17RV 2040 F8 2 C 1 RV SA C RVT APP I SHUTDOWN COOLING SHi 1 ______ ______ SUPPLY RELIEF, RG-01 RLFR-RV-20RV 2041 G2 2 A/C I RV SA C RVT APP I RIHR HEAT EXCHANGER A SHELL LJ__ -1 OPB SIDE RELIEF, RG-01 RHR-RV-21RV 2041 G2 2 A/C 1 RV SA C RVT APP I RHR HEAT EXCHANGER B SHELL__________ __L ___ -i1 OPB SIDE RELIEF, RG-01 Revision 0 Pg 5 Page 255 Cooper Nuclear Station Fifth Interval lnservice Testing Pro gram for Pumps and Valves SYSTEM: SERVICE WATER (SW)VALVE CIC P&ID P&ID ISI 1ST VALVE VALVE ACT NORM ITEST JTEST I NOTES/DESCRIPTION I_____ __COOR CLASS CAT SIZE jTYPE TYPE _POS RQMT jFREQ SW-AOV-854AV 2036 Fl11 3 B 2 BAL AO C IPIT 2Y SW PASSIVE RAD MONITOR SH 1______ __ SAMPLE RETURN, RG-01 SW-AOV-855AV 2036 F10 B 2 BAL AO C PIT 2Y SW PASSIVE RAD MONITOR SH 1 ____SAMPLE RETURN, RG-01 SW-AOV-TCV451A 2036 E7 3 B 12 GB AO O/T PS0 Q REC HEAT EXCHANGER A OUTLET SHi1 FSO Q/CS TEMPERATURE CONTROL VALVE, FST Q/CS CSJ-07, RG-01, TP-04_____ ~~~PIT 2Y __ _ _ _ _ _ _ _ _ _ _ _SW-AOV-TCV451B 2036 F7 3 B 12 GB AO O/T PSO Q REC HEAT EXCHANGER B OUTLET SHi 1ESO Q/CS TEMPERATURE CONTROL VALVE, FST Q/CS CSJ-07, RG-01, TP-04______ ______ ______ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _SW-AOV-TCV452A KSV- D2 3 B 2 1/2 GL AO 0 SKID Q DG1 TURBO INTER COOL TCV, TP-47-8 ________ ___ __ _ 05, RG-01 SW-AOV-TCV452B KSV- D2 3 B 2 1/22 GL AO 0 SKID Q DG2 TURBO INTER COOL TCV, TP-47-8 ____ ___ ____ __ _ 05, RG-01 SW-AOV-2797AAV KSV- H5 3 B 6 BTF AO 0 SKID Q DG1 SUPPLY, TP-05, RG-01 47-8 ____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _SW-AOV-2797BAV KSV- H5 3 B 6 -BTF AO 0 SKID Q DG2 SUPPLY, TP-05, RG-O1 47-8 _ _ _ ______ _ _ _ _ _ _ _____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _SW-CV-ARA 23 ES 3 C 6 CK SC C COD CVCM SW AIR RELEASE, RG-01, TP-01 Revision 0 Page 256 Cooper Nuclear Station Fifth lnterval Inservice Testing Program for Pumps and Valves SYSTEM: SERVICE WATER (SW)VALVE CIC P&ID P&ID IS1 1IST VALVE VALVE JACT NORM TEST TEST- NOTES/DESCRIPTION ___________COOR CLASSJCAT SIZE TYPE_+/-TYPE POS RQMT FREQ SH 1 ___ ___ __ CL CVCM SW-CV-ARB 2036 C5 3 C 6 CK SC C ,COD CVCM SW AIR RELEASE, RG-01, TP-01_______SHi 1 _______ CCL CVCM SW-C V-10CV 2006 All 3 C 20 CK-D SA 0 ESO Q SW PUMP A DISCHARGE, RG-01__ _ _ _ _ _ SH 1 _ _ _ _ _ _ __FSC Q SW-CV-1 1CV 2006 A8 3 C 20 CK-D SA 0 FSO Q SW PUMP B DISCHARGE, RG-01 Sil __ _ __ _ FSC Q _ _ _ _ _ _ _ _ _ _SW-CV-12CV 2006 A10 C 20 CK-D SA 0 FSO Q SW PUMP C DISCHARGE, RG-01__ _ _ _ __ _ H1-1 __ _ _ _ _ FSC Q _ _ _ _ _ _ _ _ _ _ _ _ _SW-CV-13CV 2006 A7 3 C 20 CK-D SA 0 FSO Q SW PUMP D DISCHARGE, RG-01__ _ _ _ _ _ SHi 1 _ __ _ _ _ FSC Q SW-CV-19CV 2006 F9 3 C 14 CK-T SA C FSO Q RHIR SW BOOSTER PUMP A_______SH 4 _______________FSC Q DISCHARGE, RG-01 SW-CV-20CV 2006 C9 3 C 14 CK-T SA C FSO Q RHR SW BOOSTER PUMP B________SH 4 ______FSC Q DISCHARGE, RG-01 SW-CV-21CV 2006 D9 3 C 14 CK-T SA C FSO Q RI-R SW BOOSTER PUMP C S______ H 4 _____________FSC Q DISCHARGE, RG-01 SW-C V-22CV 2006 A9 3 C 14 CK-T SA C FSO Q RIIR SW BOOSTER PUMP D S_______ H 4 _____________FSC Q DISCHARGE, RG-01 SW-CV-27CV 206 C .14 CK-D OA 0 COF CVCM REC HEAT EXCHANGER B SUPPLY, SiCCD CVCM RG-01, TP-01 Revision 0 Page 257 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: SERVICE WATER (SW)VALVE CIC P&ID 1P&ID ISI IST jVALVE VALVE ACT 1NORM TEST 1TEST NOTES/DESCRIPTION SW-CV-28CV 2036 D2 3 C 14 CK-D SA 0 COF CVCM REC HEAT EXCHANGER A SUPPLY,________SH 1 __________CCD CVCM RG-01, TP-01 SW-CV-35CV 2077 D1 C 10 CK-S SA 0 F:SO Q DG1 SUPPLY, RG-01___ __ ___ __ __FSC Q sw-cv-36cv 2077 D2 3 C 10 CK-S SA 0 FSO Q DG1 SUPPLY, RG-01__ ____ __FSC Q SW-CV-37CV 2077 D5 3 C 10 CK-S SA 0 IFSO Q DG2 SUPPLY, RG-01___ _ _ __ __ __ __ __FSC Q SW-CV-38CV 2077 D5 C 10 CK-S SA 0 FSO Q DG2 SUPPLY, RG-01___ _ __ __ ____ FSC Q SW-CV-86CV 2006 D10 3 C 0.5 CK-P SA 0 COF CVCM SW PUMP A & C CHEMICAL SH 1 __ ___ __ CCL CVCM INJECTION, RG-01, TP-01 SW-CV-87CV 2006 El0 C 0.5 CK-P SA 0 COF CVCM SW PUMP A & C CHEMICAL SH 1___ __ CCL CVCM INJECTION, RG-01, TP-01 SW-C V-88MV 2006 C10 C 0.5 CK-P SA 0 COF CVCM SW PUMP B & D CHEMICAL SH 1___ CCL CVCM INJECTION, RG-01, TP-01 SW-CV-89MV 2006 C10 C 0.5 CK-P SA 0 COF CVCM SW PUMP B & D CHEMICAL SH 1___ __ CCL CVCM INJECTION, RG-01, TP-01 SW-MOV-36MV 2006 El0 B 24 BTF MO 0 FSC Q SW LOOP CRITICAL HEADER_ _ _ _ _ _ _ SHi ___1______ ____ ___ _ _ _ PIT 2Y ISOLATION, RG-01 SW-MOV-37MV 206 EO 3 B 24 BTF MO O FSC Q SW PUMPS CROSSTIE, RG-01 Revision 0 Page 258 Cooper Nuclear Station Fifth Interval Inservice Testing Program for Pumps and Valves SYSTEM: SERVICE WATER (SW)VALVE CIC IP&ID P&ID 1SI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION j___COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ _____________ SH 1_ _ __ _ _ PIT 2Y _ _ _ _ _ _ _ _ _ _ _SW-MOV-MO89A 2036 C7 3 B 18 GB MO C PSO Q RH{R HEAT EXCHANGER A SW Sill FSO RE OUTLET, ROJ-l0, RG-01_____ _ ___ ___ _ ____ _ _____ ______ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _SW-MOV-MO89B 2036 Cl10 B 18 GB MO C PSO Q RIIR HEAT EXCHANGER B SW Sill FSO RF OUTLET, ROJ-10, RG-0l_____ _____ ______ PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _SW-MOV-650MV 2036 E3 3 B 18 BTF MO 0 IFSO Q REC HEAT EXCHANGER A Sill FSC Q OUTLET, RG-01_____ ~~~~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _SW-MOV-651MV 2036 C3 3 B 18 BTF MO 0 FSO Q REC HEAT EXCHANGER B OUTLET, Sill FSC Q RG-01_____ ______ ~~PIT 2Y _ _ _ _ _ _ _ _ _ _ _ _ _SW-MOV-886MV 2036 D1 B 4 GT MO C FSO Q EMERGENCY SUPPLY TO REC Sil 1___ __ PIT 2Y NORTH CRITICAL LOOP, RG-01 SW-MOV-887MV 2036 D1 B 4 GT MO C FSO Q EMERGENCY SUPPLY TO REC Sil 1 __ ___ __ PIT *2Y SOUTH CRITICAL LOOP, RG-01 SW-MOV-888MV 2036 E4 3 B 4 GT MO C FSO Q EMERGENCY RETURN FROM REC Sil 1__ PIT 2Y NORTH CRITICAL LOOP, RG-01 SW-MOV-889MV 2036 C4 3 -B 4 GT MO C FSO Q EMERGENCY RETURN FROM REC________SHl 1 __ ________________ _ PIT 2Y SOUTH CRITICAL LOOP, RG-01 SW-RV-12RV 20 G8 3 C 3/4 RV SA C RVT APP I SWBP 1A SEAL WATER RELIEF, Revision 0 Page 259 Cooper Nuclear Station Ffifh Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: SERVICE WATER (SW)VALVE CIC P&ID P&ID 1ISI IST VALVE VALVE fACT NORM TEST TEST NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE [TYPE POS RQMT FREQ_____________ _ _ _ _ _ _ SH 4 _ __ __ _ __ RG-01 SW-RV-13RV 2006 C8 3 C 3/4 RV SA C RVT APP I SWBP 1C SEAL WATER RELIEF, S__ _ __ H 4 __ __ _ _ _ _ _ _ _ _ _ _ RG-01 SW-RV-14RV 2006 E8 C 3/4 RV SA d RVT APP I SWBP 1B SEAL WATER RELIEF, SH 4 __ _ _ _ _ _ __ _ _ RG-01 SW-RV-15RV 2006 B8 3 C 3/4 RV SA C RVT APP I SWBP 1D SEAL WATER RELIEF, SH 4 __ _ _ _ _ _ RG-01 SW-V-640 2006 E7 3 B 1 BAL MA 0 FSC 2Y SWBP C SEAL WATER RIVERWELL SH 4 ____ _______SHUTOFF, RG-01 SW-V-649 2006 G7 3 B 1 BAL MA 0 FSC 2Y SWBP A SEAL WATER RIVERWELL SH 4 _______ ________ SHUTOFF, RG-01 SW-V-656 2006 B7 3 B 1 BAL MA 0 FSC 2Y SWBP D SEAL WATER RI VER WELL SH 4 ____ _______SHUTOFF, RG-01 SW-V-665 2006 D7 3 B 1 BAL MA 0 FSC 2Y SWBP B SEAL WATER RIVERWELL SH 4 _ _ __ _ __ _ _ _ _ _ _ _ _ _ __ _ __ _ _ _ _ SHUTOFF, RG-01 SW-V-1422 2006 F8 3 C 3/4 GB MA C ESO 2Y SWBP A GLAND WATER SUPPLY, SH 4 __ __ _ _ _ _ _ _ _ _RG-01 SW-V-1424 2006 G8 3 C 3/4 GB MA C FSO 2Y SWBP A GLAND WATER FLOW SH 4 _______ __CONTROL, RG-01 SW-V-1426 2006 C8 3 C 3/4 GB MA C FSO 2Y SWBP B GLAND WATER SUPPLY, S__ _ __ H 4 _ _ __ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ __ RG-01 Revision 0 Page 260 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: SERVICE WATER (SW)* VALVE CIC P&ID P&ID IS I5 ST VALVE VALVE ACT INORM TEST ITEST NOTES/DESCRIPTION COOR jCLASS* CAT SIZE TYPE TYPE POS RQMT FREQ _________SW-V-1428 2006 D8 3 C 3/4 GB MA C ESO 2Y SWBP B GLAND WATER FLOW S______ H 4 _____________CONTROL, RG.-01 SW-V-1430 2006 E8 3 C 3/4 GB MA C FSO 2Y SWBP C GLAND WATER SUPPLY, SH14 ___ __RG-01 SW-.V-1432 2006 E8 C 3/4 GB MA C FSO 2Y SWBP C GLAND WATER FLOW SH 4 CONTROL, RG-01 SW-V-1434 2006 B8 3 C 3/4 GB MA C FSO 2Y SWBP D GLAND WATER SUPPLY, SH 4 ___ __RG-01 SW-V-1436 2006 B8 3 C 3/4 GB MA C FSO 2Y SWBP D GLAND WATER FLOW SH 4 CONTROL, RG-01 Revision 0 Pg 6 Page 261 Cooper Nuclear Station Fifth lnterval Inservice Testing Program for Pumps and Valves SYSTEM: STANDBY GAS TREATMENT (SGT)VALVE CIC P&ID JP&IDj ISI IST IVALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COO CLASS CAT SIZE TYPE TYPE POS RQMT FREQ SGT-AOV-249AV 2037 C2 A B 12 BTF AO C FSO Q SGT UNIT A INLET, RG-O1, TP-04Q FST q___ __ ___ __ _ _ __ _ ___ _ _____ __PIT 2Y SGT-AOV-250AV 2037 G2 A B 12 BTF AO C FSO Q SGT UNIT B INLET, RG-01, TP-04 FSC Q FST Q___ _ __ ____ __PIT 2Y SGT-AOV-251IAV 2037 C6 A B 12 BTF AO C FSO Q SGT UNIT A DISCHARGE, RG-01, FST Q TP-04___ _ __ __ __ ____ __ PIT 2Y SGT-AOV-252AV 2037 G6 A B 12 BTF AO C FSO Q SGT UNIT B DISCHARGE. RG-01, FST Q TP-04__ ___ __ _PIT 2Y SGT-AOV-255AV 2037 A4 A B 10 BTF AO C PIT 2Y SOT UNIT A PASSIVE BYPASS,__________RG-0 1 SGT-AOV-256AV 2037 E4 A B 10 BTF AO C PIT 2Y SGT UNIT B PASSIVE BYPASS,____ ___ ___ __ __ ____ __ ______ _ _ ___ RG-01 SGT-AOV-270AV 2037 C1 A B 10 BTF AO C FSO Q SGT UNIT A DILUTION AIR FST Q SHUTOFF, RG-01, TP-04___ _ __ __ __ __ __ __ __ __PIT 2Y SGT-AOV-271AV 23 01 A B 10 BTF AO C FO Q SGT UNIT B DILUTION AIR Revision 0 Page 262 Cooper Nuclear Station Ffith lnterval Inservice Testing Program for Pumps and Valves SYSTEM: STANDBY GAS TREATMENT (SGT)VALVE CIC P&ID P&ID 1 S 1 I ST VALVE VALVE ACT NORM TEST TEST 1 NOTES/DESCRIPTION COOR CLASS CAT SIZE TYPE TYPE POS RQMT FREQ _____________ FST Q SHUTOFF, RG-.0I, TP-04___ __ __ __ __ __ _ __ __PIT 2Y SGT-AOV- 2037 C7 A B 10 BTF AO C FSO Q SGT UNIT A DISCHARGE DPCV546A FST Q DIFFERENTIAL PRESSURE___ __ __ ___ _ __ __ __ ____________ ____ __ PIT 2Y CONTROL, RG-01, TP-04 SGT-AOV- 2037 E7 A B 10 BTF AO C FzSO Q SGT UNIT B DISCHARGE DPCV546B FST Q DIFFERENTIAL PRESSURE___ ________PIT 2Y CONTROL, RG-01, TP-04 SGT-CV-14CV 2037 C6 A C 10 CK-D SA C FSO CS SGT UNIT AFAN EXHAUST,____ ____ ___ ___ ___ FSC CS CSJ-01, RG-01 SGT-CV-15CV 2037 G6 A C 10 CK-D SA C FSO CS SGT UNIT B FAN EXHAUST, FSC CS CSJ-01, RG-01 Revision 0 Page 263 Revision 0 Page 263 Cooper Nuclear Station Ffifh Interval lnservice Testing Program for Pumps and Valves SYSTEM: STANDBY LIQUID CONTROL (SLC)1 IP&ID ISI IST 1VALVE VALVE ACT NORM TEST TEST VALVE CIC P&D COOR CLASS CAT jSIZE TYPE TYPE POS RQMT FREQ NOTES/DESCRIPTION SLC-CV- 10CV 2045 E9 A C 11/2 CK-L SA C COF CVCM SLC PUMP A DISCHARGE______ _SH 2 _______CCR CVCM CHECK, RG-01 SLC-CV-11CV 2045 F9 A C 11/2/ CK-L SA C COF CVCM SLC PUMP B DISCHARGE S______ H 2 _ __CCR CVCM CHECK, RG-01 SLC-CV- 12CV 2045 E8 1 A/C 11/2 CK-L SA C J-i1 OPB SLC INJECTION LINE SH 2 COF CVCM OUTBOARD CHECK, RG-01___________ ____ ____CCL CVCM SLC-CV-13CV 2045 E7 1 A/C 11/2 CK-L SA C UJ-i OPB SLC INJECTION LINE INBOARD SH 2 COF CVCM CHECK, RG-01___________ CL CVCM SLC-RV-10RV 2045 D1O A C 3/4 RV SA C RVT APP 12 SLC PUMP A DISCHARGE________SH 2 ____RELIEF, RG-01 SLC-RV-1 1RV 2045 G9 A C 3/4 RV SA C 7RVT APP I 2 SLC PUMP B DISCHARGE S_______ H 2 _______RELIEF, RG-01 SLC-SQBV-14A 2045 E8 A D 11/2 SHR CH C EX 2Y SLC EXPLOSIVE VALVE A, RG-01 SH2 2_ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _SLC-SQBV-14B 2045 E8 A D 11/2 SHR CH C EX 2Y SLC EXPLOSIVE VALVE A, RG-01 5112 Revision 0 Pg 6 Page 264 Cooper Nuclear Station Fifth Interval Inservice Testing Pro gram for Pumps and Valves SYSTEM: STATION AIR (SA)VALVE CIC IP&ID P&ID ISI IST VALVE VALVE ACT NORM TEST TEST NOTES/DESCRIPTION COOR________ __ CLASS CAT SIZE TYPE -TYPE POS RQMT FREQ SA-V-647 2010 G4 2 A 1GB MA C LJ-1 OPB PASSIVE DRYWELL OUTBOARD________ SH 3 ___ ___ _______ ___ ____ ___ ___SUPPLY ISOLATION SA-V-648 2010 G4 2 A 1GB MA C LJ-i OPB PASSIVE DRYWELL INBOARD SH 3 SUPPLY ISOLATION 1. License Amendment 234 to the CNS Facility Operating License requires that the SLC system be maintained in the Augmented IST Program for the Alternate Source Term Function 2. License Amendment 176 to the CNS Facility Operating License requires that the control of the SLC relief valve testing (and as-left set point) be maintained in the Augmented 1ST Program. This amendment requires that the USAR describe the as-left set point (1540 +/-- 1%) and that future changes be evaluated per 10CFR50.59. The as-found criteria is 1478 to 1602 psig per USAR section III-9.4.Revision 0 Page 265 NLS2016012 Page 1 of 7 Enclosure 2 Cooper Nuclear Station Fifth Interval Inservice Examination and Testing Program for Snubbers Nebraska Public Power District Cooper Nuclear Station Fifth Interval*Inservice Examination &Testing Program for Snubbers Revision 0 Cooper Nuclear Station P0 BOX 98 Brownville, NE 68321-0098 Commercial Operation Date: July 1, 1974 Snubber Engineer: Program Backup / Supervisor: Date: _-______Date: __ -_ ____Date: 9,46 EP & C Manager: I Cooper Nuclear Station Fifth Interval Inservice Examination & Testing Pro gram for Snubbers TABLE OF CONTENTS SECTION TITLE

1.0 INTRODUCTION

1.1 Purpose 1.2 Scope 2.0 VISUAL EXAMINATION REQUIREMENTS 2.1 General 2.2 Preservice Examinations 2.3 Inservice Examinations 3.0 FUNCTIONAL TESTING REQUIREMENTS 3.1 Preservice Operational Readiness Testing 3.2 Inservice Operational Readiness Testing 4.0 SERVICE LIFE MONITORING REQUIREMENTS 5.0 CNS PROCEDURES Revision 0 Page 2 Revision 0 Page 2 Cooper Nuclear Station Ffith Interval Inservice Examination & Testing Program for Snubbers 1.0 IN~TRODUCTION 1.1 Purpose This document establishes the Cooper Nuclear Station (CNS) Fifth 120-Month Interval Inservice Testing (IST) Program requirements for the American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 dynamic restraints (snubbers). The Snubber Program requirements include Visual Examinations, Functional testing, and Service Life Monitoring. The CNS Fifth 120-Month Interval Snubber Inservice Testing Program Plan will be applicable during the following time period.Begin: 03/01/2016 End: 02/28/2026 The objective of the Snubber Program is to provide reasonable assurance that the snubbers within this plan are capable of performing their intended function(s) during normal operations and seismic events.1.2 Scope The 1ST Program scope for dynamic restraints applies to the Class 1, 2, and 3 dynamic restraints that meet the scope statements outlined in the ASME OM Code of record. The CNS Code of record for the Fifth 120-Month Interval is the 2004 Edition through the 2006 addenda of the ASME OM Code.This document is based on the following Subsections of the 2004 Edition through the 2006 Addenda of the ASME OM Code.*Subsection ISTA, "General Requirements" ISTA-1 100 states that "These requirements apply to" (c) "dynamic restraints (snubbers) used in systems that perform one or more of the three functions identified in subpara. ISTA-1 100(a) [shutting down a reactor to the safe shutdown condition, in maintaining the safe shutdown condition, or in mitigating the consequences of an accident], or to ensure the integrity of the reactor coolant pressure boundary"* Subsection ISTD, "Preservice and inservice Examination and Testing of dynamic Restraints (Snubbers) in Light-Water Reactor Nuclear Power Plants." Cooper Nuclear Station was designed and licensed to operate with the Hot Shutdown condition defined as the "safe" shutdown condition. CNS currently has a total of 208 snubbers in this program, 155 Mechanical snubbers and 53 hydraulic snubbers. These snubbers and their locations are listed in CNS Maintenance Procedure, 7.2.34.1, "Snubber Examination". Revision 0 Page 3 Revision 0 Page 3 Cooper Nuclear Station Ffifth Interval Inservice Examination & Testing Pro gram for Snubbers 2.0 VISUAL EXAMINATIONS 2.1 General The examination boundary shall include the snubber assembly from pin to pin, inclusive (ISTD-31 10).Typical preservice or inservice examination checklist items to be considered are listed in Nonmandatory Appendix B of the ASME OM Code.2.2 Preservice Examinations For new and modified systems, preservice examinations shall be performed after placing the systems in service. These examinations should be identified within the modification package implementing the new and/or modified system. The preservice examination requirements of ISTD-4 100 shall be met.2.3 Inservice Examinations Snubbers shall be visually examined on the required schedule and evaluated to dctcrmine their operational readiness (ISTD-4200). The inservice examination shall be a visual examination to identify physical damage, leakage, corrosion, or degradation that may have been caused by environmental exposure or operating conditions. External characteristics that may indicate operational readiness of the snubber shall be examined. An examination checklist shall be used (ISTD-42 10).Snubber installations shall meet all the requirements of ISTD-423 1 (restrained movement), ISTD-4232 (thermal movement), and ISTD-4233 (design-specific characteristics). CNS is utilizing approved Code Case OMN-1 3 (2004 Edition), "Requirements for Extending Snubber Inservice Visual Examination Interval at LWR Power Plants," as listed in Table 1 of Regulatory Guide 1.192, Revision 1 (August 2014), "Operation and Maintenance Code Case Acceptability, ASME OM Code." This Code Case allows for the extension of visual examinations beyond the frequencies specified in ISTD for mechanical and hydraulic snubbers to at least once every 10 years. All requirements of this NRC approved version of OMN-1 3 must be met.Snubbers at CNS are examined in accordance with CNS procedures 7.2.34.1 or 7.2.34.2.A snubber that requires further evaluation or is classified as unacceptable during inservice examination may be tested in accordance with the requirements ISTD-521 0.3.0 FUNCTIONAL TESTING REQUIREMENTS 3.1 Preservice Operational Readiness Testing Preservice Operational readiness testing shall be performed on all snubbers. Testing may be performed at the manufacturer's facility (ISTD-5 110).Revision 0 Pg Page 4 Cooper Nuclear Station Fifth Interval Inservice Examination & Testing Pro gram for Snubbers Test parameters shall meet the requirements of ISTD-5 120. If a failure occurs, corrective action shall meet the requirements of ISTD-5 130.3.2 Inservice Operational Readiness Testing Snubbers shall be tested for operational readiness during each fuel cycle. Functional tests at CNS are completed in accordance with the 10% Testing Sample Plan (ISTD-5300). Test parameters are in accordance with ISTD-52 10.The CNS snubber population consists of hydraulic and mechanical snubbers. These snubbers are separated into Design Test Plan Groups (DTPG) by snubber type (hydraulic versus mechanical) and size in accordance with ISTD-5252 (see below).DTPG Population Description Number of Snubbers Sample Plan 1 PSA-3 Mechanical Snubbers 18 10%2 PSA-l0 Mechanical Snubbers 89 10%3 PSA-35 Mechanical Snubbers 48 10%4 ANVIIL/Grinnell Hydraulic 53 10%_______Snubbers _________________ Testing shall be performed during normal operation, or during system or plant outages (ISTD-5200). However, snubber testing may begin no earlier than 60 days before a scheduled refueling outage (ISTD-5240). The initial sample shall be 10% of the DTPG, composed according to either ISTD-5311l(a) or ISTD-531 1(b).Snubbers that do not meet test requirements specified in ISTD-52 10 or ISTD-5230 shall be evaluated to determine the cause of the failure (ISTD-527 1). Failure mode groupings (FMGs) should be determined in accordance with ISTD-5272, as applicable. The FMG boundaries shall be applied per ISTD-5273 and utilized with the 10% sample plan per ISTD-5300. Snubbers will generally be tested at CNS in accordance with CNS procedures 7.2.34.7 and 7.2.34.8. However, snubbers can be functionally tested by vendors, if necessary. 4.0 SERVICE LIFE MONITORING REQUIREMENTS Initial snubber service life shall be predicted based on manufacturer's reconmnendation or design review (ISTD-6 100).Service life shall be evaluated at least once each fuel cycle, and increased or decreased, if warranted (ISTD-6200). This is typically done by reviewing the examination and functional test results at the completion of the refueling outage campaign.Snubbers shall be replaced or reconditioned, as required, to ensure that the service life is not exceeded before the next scheduled system or plant outage.If necessary, technical justification shall be documented for extending the service life to or beyond the next scheduled system or plant outage.Revision 0 Pg Page 5 Cooper Nuclear Station Fifthi Interval Inservice Examination & Testing Program for Snubbers Causes for any examination or testing failures shall be deternined and considered in establishing or re-establishing service life.5.0 CNS PROCEDURES 5.1 Engineering Procedure 3.39, "Snubber Program" 5.2 Maintenance Procedure 7.2.34.1, "Snubber Examination" 5.3 Maintenance Procedure 7.2.34.2, "Pipe Snubbers Removal and Installation" 5.4 Maintenance Procedure 7.2.34.3, "Grinnell Figure 200/201 Hydraulic Snubber Maintenance" 1 5.5 Maintenance Procedure 7.2.34.4, "Pacific Scientific PSA-3 and PSA-10 Snubber Maintenance 5.6 Maintenance Procedure 7.2.34.5, "Pacific Scientific PSA-35 Snubber Maintenance 5.7 Maintenance Procedure 7.2.34.7, "Grinnell Figure 200/201 Hydraulic Snubber Functional Test" 5.8 Maintenance Procedure 7.2.34.8, "Pacific Scientific Snubber Functional Test" 5.9 Administrative Procedure 0.30, "ASME Section XI Repair/Replacement and Temporary Code and Non-Code Repair Procedure" 5.10 Surveillance Procedure 6.SNUB.601, "Snubber Operability" 5.11 Surveillance Procedure 6.SNUB.602, "Snubber Service Life Evaluation" Revision 0 Page 6 Revision 0 Page 6}}

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