Text

Fifth Ten-Year Lnservice Testing Plan
	ML12061A045
Person / Time
Site:	Monticello
Issue date:	02/28/2012
From:	O'Connor T; Northern States Power Co, Xcel Energy
To:	; Office of Nuclear Security and Incident Response, Document Control Desk
References
	L-MT-12-020
	Download: ML12061A045 (217)
	v • d • e

@ Xcel Energy. Monticello Nuclear Generating Plant 2807 W County Rd 75 Monticello, MN 55362 February 28,2012 L-MT-12-020 10 CFR 50.55a(f)

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Monticello Nuclear Generating Plant Docket 50-263 Renewed License No. DPR-22

Subject:

Fifth Ten-Year lnservice Testing Plan Pursuant to 10 CFR 50.55a(f)(5)(i), Northern States Power Company, a Minnesota corporation, d/b/a Xcel Energy, the licensee for the Monticello Nuclear Generating Plant (MNGP), submits the fifth ten-year interval Pump and Valve lnservice Testing (IST)

Program Plan (enclosure). This IST plan for the fifth ten-year interval begins September 1, 2012. The IST fifth ten-year interval program plan was developed in accordance with the 2004 Edition, through 2006 Addenda of American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code, as the code of record.

Should you have questions regarding this letter, please contact Mr. Randy Rippy at (612) 330-6911.

ew commitments and no revisions to existing commitments.

Monticello Nuclear Generating Plant Northern States Power Company-Minnesota Enclosure

Document Control Desk Page 2 cc: Administrator, Region Ill, USNRC Project Manager, Monticello Nuclear Generating Plant, USNRC Resident Inspector, Monticello Nuclear Generating Plant, USNRC

ENCLOSURE MONTICELLO NUCLEAR GENERATING PLANT (Unit 1)

PUMP & VALVE IN-SERVICE TESTING PROGRAM PLAN FIFTH INTERVAL - REVISION 0 214 pages follow

Operated by: NSP-Minnesota 414 Nicollet Mall MINNEAPOLIS, MN 55401-1993 Owned by: XCEL ENERGY 414 Nicollet Mall MINNEAPOLIS, MN 55401-1993 MONTICELLO NUCLEAR GENERATING PLANT (Unit 1)

(Provisional License Granted: September 8th, 1970) 2807 WEST HIGHWAY 75 MONTICELLO, MINNESOTA 55362 PUMP & VALVE IN-SERVICE TESTING PROGRAM PLAN FIFTH INTERVAL - REVISION 0 SEPTEMBER 1st, 2012 THROUGH MAY 31st, 2022 Prepared By: SEE PASSPORT AR 1317531 Date IST Program Coordinator Reviewed By: SEE PASSPORT AR 1317531 Date Approved By: SEE PASSPORT AR 1317531 Date

Monticello Nuclear Generating Plant Revision 0 Summary of Change Log Rev. Plan Body/ Page Section Change Summary Attachment 0 Entire Plan All All Complete rewrite for Fifth 10-Year Test Interval. Code of Record shall be ASME OM Code-2004 Edition with Addenda through OMb-2006.

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Monticello Nuclear Generating Plant Revision 0 Table of Contents Section Page Summary of Change Log i Table of Contents ii

1.0 INTRODUCTION

1.1 Purpose 1 1.2 Scope 1 1.3 Program Bases 3 1.4 References 3 2.0 IN-SERVICE TESTING (IST) PLAN FOR PUMPS 2.1 Pump IST Plan Description 4 2.2 Pump Plan Table Description 4 2.3 Pump IST Requirements 5 3.0 IST PLAN FOR VALVES 3.1 Valve IST Plan Description 7 3.2 Valve Plan Table Description 7 3.3 General Valve Testing and Measurement of Test Quantities 12 3.4 Allowable Ranges for Test Quantities 17 ATTACHMENTS 1 IST Plan System Listing and Acronyms 2* IST Plan Notes 3 Index For IST Plan Relief Requests and Deferred Testing Justifications (DTJs) 4 Pump Relief Requests (PR) 5 Valve Relief Requests (VR) 6 Deferred Testing Justifications (DTJ) 7 Code Cases Invoked by IST Program 8 Pump Tables 9 Valve Table 10 NRC Safety Evaluations (SER) for IST Program Relief Requests (Pending NRC Approval)

References to IST Plan notes from Attachment 2 are shown by subscripts ii of ii

Monticello Nuclear Generating Plant Revision 0

1.0 INTRODUCTION

1.1 Purpose Monticello Nuclear Generating Plant (MNGP) Technical Specification 5.5.5, Inservice Testing Program, requires Inservice Testing of ASME Code Class 1, 2 and 3 pumps and valves per the requirements of the ASME Operations and Maintenance Code (OM Code). Title 10, Part 50.55a of the Code of Federal Regulations specifies the rules for the application of ASME Operations and Maintenance Code (OM Code) requirements. These rules include the limitations and modifications of OM Code requirements, additional testing requirements, the edition and addenda of the OM Code that shall be used, provisions for pursuit of relief from OM Code requirements, and alternatives to OM Code requirements.

The OM Code provides the requirements for the IST Program scope, program administration, required tests, test methods, acceptance criteria, and corrective actions.

This Inservice Testing Plan identifies the pumps, valves, and pressure relief devices that are included in the IST Program. The IST Plan identifies the inservice tests to be performed on the components in the program in order to verify their operational readiness and monitor for component degradation.

1.2 Scope The basic scope of the IST Program is defined in Subsection ISTA (general requirements) of the OM Code. Per paragraph ISTA-1100, IST requirements apply to all valves and pumps that are required to perform a specific function in shutting down the reactor to a safe shutdown condition1, maintaining the safe shutdown condition or in mitigating the consequences of an accident. Mandatory Appendix I of the OM Code specifies testing requirements for all pressure relief devices (including vacuum breakers and rupture disks) that provide underpressure

/overpressure protection for systems, parts of systems or components that perform the above functions.

The MNGP IST Program Plan will be in effect through the fifth ten-year interval at MNGP (September 1, 2012 through May 31, 2022). This plan will be updated as required in accordance with 10CFR50.55a(f). This program plan meets the requirements of 10 CFR 50.55a(b) 12 months prior to the fifth interval start date of September 1, 2012, in that it complies with the 2004 Edition, through 2006 Addenda of ASME OM Code.

Attachments to the program plan provide a complete listing of those IST Program in-scope pumps and valves included in the program per the requirements of ISTA, ISTB, ISTC and Mandatory Appendix I of OM Code 2004 Edition through OMb-2006 Addenda.

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Monticello Nuclear Generating Plant Revision 0 Per 10 CFR 50.55a, the scope of IST requirements are limited to only those components classified as ASME Class 1, 2, or 3. However, per 10 CFR 50, Appendix B, Criterion XI, all components must be tested commensurate with the safety function that they perform. Therefore, testing requirements are also applicable to non-Code (NC) Class components although compliance with the OM Code testing requirements is not required. However, licensees are required to demonstrate the adequacy of testing performed on non-Code Class components, as mentioned above, in accordance with Appendix B requirements. Since the OM Code is accepted by the NRC staff as an acceptable standard for pump and valve testing, non-Code Class components that perform functions important to safety are tested to OM Code requirements whenever practical. However, for NC Components, if compliance with the OM Code is not practical, or an alternate test method or test frequency is determined, it is not necessary to obtain relief from the OM Code requirements from the NRC staff. In these cases, the adequacy of the alternative test(s) is supported by DTJs contained in the IST Program and/or other Program controlled documents.

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Monticello Nuclear Generating Plant Revision 0 1.3 Program Basis The IST Program Basis Document defines component safety functions and test requirements for MNGP. Component safety functions are defined in accordance with design and licensing basis documents (USAR, Technical Specifications, NRC commitments) and Nuclear Regulatory Commission guidance (Generic Letters, Information Notices, NUREGs, etc.). Based on the defined component safety functions, testing requirements were assigned in accordance with the OM Code.

The MNGP quality groups and safety classifications along with the references listed in Section 1.4 were used to classify components as ASME Code Class 1, 2, 3 and non-class (NC) for the purposes of IST. All ASME Code Class 1, 2, and 3 components were evaluated for inclusion in the IST Program. The IST Program also identifies MNGP defined augmented testing requirements for NC components that perform functions determined to be important to nuclear safety.

After all components were identified and classified, the safety functions for each component were determined. The safety function reference of each component was identified and documented in the IST Program Basis Document utilizing reference sources such as the USAR, Technical Specifications, and System Design Basis Documents, etc. Valves included in the IST Program were categorized in accordance with paragraph ISTC-1300. Pumps included in the IST Program were identified as either centrifugal, vertical line shaft or positive displacement in accordance with paragraph ISTB-1100 and then grouped as either Group A or Group B in accordance with paragraph ISTB-1300..

Subsequent to determining component safety function, classification and categorization, Subsections ISTB, ISTC, Appendix I and Appendix II were utilized to assign test type and test frequencies for each pump and valve.

Assignment of test frequency was performed on a most limiting basis considering practicality and all Technical Specification, USAR and licensing commitments.

References are provided to support the evaluations of component safety functions and testing requirements as not all components are specifically addressed in the licensing basis documentation and system.

1.4 References MNGP Technical Specifications MNGP Updated Safety Analysis Report 10CFR50.55a(f) 10CFR50 Appendix J Option B, Primary Reactor Containment Leakage Testing for Water-Cooled Power Plants ASME OM Code, 2004 Edition through OMb-2006 Addenda, Code for Operation and Maintenance of Nuclear Power Plants Page 3 of 20

Monticello Nuclear Generating Plant Revision 0 NRC Generic Letter No. 89-04, Guidance on Developing Acceptable In-Service Testing Programs" NRC NUREG-1482 Rev. 0 and Rev. 1, Guidelines for In-Service Testing at Nuclear Power Plants MNGP Color Coded Piping and Instrument Diagrams NRC Safety Evaluation Report, Safety Evaluation (SE) of Relief Requests for the Pump and Valve In-Service Testing Program (TAC No.MB9736), dated December 10, 2003 NRC Safety Evaluation Report, Safety Evaluation (SE) of Relief Request for the Pump and Valve In-Service Testing Program (TAC No.MB6807), dated July 17, 2003 Page 4 of 20

Monticello Nuclear Generating Plant Revision 0 NRC Safety Evaluation Report, Safety Evaluation (SE) of Relief Request for the Pump and Valve In-Service Testing Program (TAC No.MB9550), dated August 7, 2003 NRC Regulatory Guide 1.192, Operation and Maintenance Code Case Acceptability, ASME OM Code 2.0 IST PLAN FOR PUMPS Pumps included within the scope of the IST Plan are identified by the appropriate Group A or Group B designator in Technical Position (TP) 10.

2.1 Pump IST Plan Description This program plan establishes the requirements for the performance, administration, and implementation of the IST Plan for selected pumps at MNGP.

This plan includes those pumps that are provided with an emergency power source and are required in bringing the reactor to the safe shutdown1 condition, maintaining the safe shutdown1 condition, or mitigating the consequences of an accident.

This program plan meets the requirements of the ASME OM Code-2004 Edition through 2006 Addenda, Subsections ISTA, General Requirements, and Subsection ISTB, Inservice Testing of Pumps in Light-Water Reactor Nuclear Power Plants. Where these requirements are determined to be impractical, specific requests for relief have been written and included in Attachment 4.

2.2 Pump Plan Table Description The pumps included in the MNGP Nuclear Generating Station IST Plan are listed in Attachment 8. The information contained in this table identifies those pumps required to be tested to the requirements of ISTB, the testing parameters and frequency of testing, and associated relief requests. The heading for the pump tables are delineated below:

2.2.1 Component - The unique pump identification 2.2.2 Description - Functional/noun name for the pump.

2.2.3 P&ID - The Piping and Instrument Diagram on which the pump is depicted 2.2.4 Class - The ASME Code classification of the pump (1, 2, 3, or NC for ASME non-Code Class) 2.2.5 Group - The ISTB pump group A - Group A (those pumps in standby systems that are operated continuously or routinely during normal operation, safe shutdown1, or refueling operations)

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Monticello Nuclear Generating Plant Revision 0 B - Group B (those pumps in standby systems that are not operated continuously or routinely except for testing) 2.2.6 Type - The type of pump Centrifugal Positive Displacement Vertical Line Shaft 2.2.7 Test - Test parameters N - Speed DP - Differential Pressure P - Discharge Pressure Q - Flow V - Vibration 2.2.8 Frequency - Test frequency Q - Quarterly Y2 - Once every two years Q/Y2 - Quarterly and once every two years 2.2.9 PR - Pump Relief Request 2.3 Pump IST Requirements 2.3.1 Frequency and scheduling of Pump IST Pump IST are conducted on each pump in the program each quarter (for Group A and B tests) and once every two years (for comprehensive pump tests) unless the pump is declared inoperable or not required to be operable in accordance with Technical Specifications. This is in accordance with ISTB-3420.

2.3.2 Test Parameters Note: Test Parameters are stated per Table ISTB-3000-1.

Speed (N) - Pump speed is a reference variable for variable speed pumps Differential Pressure (DP) - Differential pressure is a trended variable measured at a given reference flow value for centrifugal and vertical line shaft pumps Discharge Pressure (P) - Discharge pressure is a reference value used to measure the trended variable of flow for positive displacement pumps. Per Table ISTB-5321-2, discharge pressure is not required to be measured for reciprocating positive displacement pumps during the Group B pump test.

Flow Rate (Q) - Flow rate is a reference variable for centrifugal and vertical line shaft pumps. Flow rate is a trended variable for positive displacement pumps.

External recirculated flow is not required to be measured if it is not practical to isolate, has a fixed resistance, and has been evaluated to not have a substantial effect on the results of the test.

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Monticello Nuclear Generating Plant Revision 0 NOTE: Per Tables ISTB-5121-1, 5221-1 and 5321-2, vibration measurement is not required during the Group B pump test, but is required during the Group A test and Comprehensive pump test.

Vibration (V) - Pump bearing vibration is a trended variable measured at a given reference value during pump testing. Pump vibration will be measured at the locations described below, dependent upon the type of pump. Pump vibration measurements will be in units of velocity (inches per second).

Centrifugal Pumps: Per ISTB-3540(a), measurements shall be taken in a plane approximately perpendicular to the rotating shaft in two approximately orthogonal directions on each accessible pump bearing housing, and in the axial direction on each accessible pump thrust bearing housing.

Vertical Line Shaft Pumps: Per ISTB-3540(b), measurements shall be taken on the upper motor bearing housing in three approximately orthogonal directions, one of which is the axial direction.

Reciprocating Pumps: Per ISTB-3540(c), measurements shall be taken on the bearing housing of the crankshaft, approximately perpendicular to both the crankshaft and the line of plunger travel.

2.3.3 Allowable Ranges for Test Parameters The allowable ranges specified in Tables ISTB-5121-1, 5221-1 or 5321-2, as appropriate, will be used, unless exception has been granted by relief request.

Should a measured test quantity fall outside the allowable range, action will be taken in accordance with ISTB-6200.

2.3.4 Instrument Accuracy Allowable instrument accuracies are given in Table ISTB-3510-1, Required Instrument Accuracy. If the accuracies of the plant installed instruments are not acceptable, temporary instruments meeting the requirements of Table ISTB-3510-1 will be used. When determining differential pressure by calculating the difference between individual suction and discharge pressure measurements, the range and accuracy of an individual instrument may be greater than allowed as long as the overall determination of differential pressure can be shown to be better than the Code required accuracy and addressed by relief request (Refer to NUREG-1482, Rev.1 Section 5.5.1 for further discussion).

Digital instruments may be selected such that the reference value does not exceed 90% of the calibrated range per ISTB-3510(b)(2) of OMb-2006.

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Monticello Nuclear Generating Plant Revision 0 3.0 IST PLAN FOR VALVES 3.1 Valve IST Plan Description This program establishes the requirements for the performance, administration and implementation of the IST Plan for valves at MNGP. This plan includes those valves that are required to perform a specific function in bringing the reactor to the safe shutdown1 condition, in maintaining the safe shutdown1 condition, or in mitigating the consequences of an accident.

This plan establishes the test intervals, parameters to be measured and meets the requirements of ISTC, Appendix I and Appendix II of the 2004 Edition through 2006 Addenda OM Code with the exception of the specific relief requests contained in Attachment 5.

Where the frequency requirements for valve testing have been determined to be impracticable, cold shutdown or refueling outage Deferred Testing Justifications (DTJs) have been identified and documented. These justifications are provided in Attachment 6.

3.2 Valve Plan Table Description The valves included in the MNGP IST Plan are listed in Attachment 9. The information contained in these tables identifies those valves required to be tested to the requirements of ISTC and/or Appendix I, the test parameters, the frequency of testing, associated relief requests and associated DTJs. Valves excluded per ISTC-1200 are not listed. The column headings for the valve tables are delineated below:

3.2.1 System - The MNGP system for which valve is associated with.

3.2.2 Component - The unique valve identification. Valve descriptions identified with the symbol (*) indicates that the valve is typical for additional valves.

3.2.3 Description - Functional/noun name for the valve.

3.2.4 P&ID - The Piping and Instrument Diagram on which the valve is represented.

3.2.5 Coord - The coordinate location of the valve on the P&ID.

3.2.6 Size - The nominal pipe size for the valve in inches.

3.2.7 Type - The type of valve is indicated by the following abbreviations:

Valve Type Description CK Check BF Butterfly GT Gate GL Globe RV Relief Page 8 of 20

Monticello Nuclear Generating Plant Revision 0 RD Rupture Disk PL Plug XP Explosive AR Air Relief FV Excess Flow Check BA Ball AN Angle SC Stop Check 2W 2-Way Solenoid 3W 3-Way Solenoid 4W 4-Way Solenoid 3.2.8 Act - The valve actuator type is indicated by the following abbreviations:

Actuator Description MO Motor Operated AO Air Operated SO Solenoid Operated MA Manual SA Self Actuated XP Explosive 3.2.9 Class - The ASME Code classification of the valve (1, 2, 3, or NC for ASME non-Code Class).

3.2.10 Cat - The category assigned to the valve per the definitions of ISTC-1300.

Category Description A Valves (manual or power-operated) for which seat leakage is limited to a specific amount in the closed position for fulfillment of their required safety function.

B Valves (manual or power-operated) for which seat leakage in the closed position is inconsequential for fulfillment of the required safety function.

C Self actuating (e.g., check valves, relief valves)

D Actuated by an energy source capable of only one operation (e.g., rupture disks and explosive actuated valves).

Split categories assigned to the valve can be as follows Category Description BC Valves that are self-actuating but are also equipped with power actuators that are capable of remote operation.

AC Check valves for which seat leakage is limited to a specific amount in the closed position for fulfillment of their required safety function.

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Monticello Nuclear Generating Plant Revision 0 3.2.11 Norm Pos - The position(s) of the valve during normal power operations is indicated as follows:

Position Description O Open C Closed O/C Open or Closed LO Locked Open LC Locked Closed 3.2.12 Safe Pos - The safety function position(s) for valves is indicated as follows:

Position Description O Open C Closed O/C Open and Closed 3.2.13 A/P - Active or Passive valve function as indicated below:

A/P Description A Active P Passive A/P Active and Passive Page 10 of 20

Monticello Nuclear Generating Plant Revision 0 3.2.14 Test - The tests performed to fulfill the requirements of ISTC, Table ISTC-3500-1, as indicated below.

Test Description Valve Applicability LT Category A Seat Leakage Test Category A Containment, Pressure Isolation Valves (CIV, PIV), & Owner Specified STO Category A/B Stroke Time Test Open Category A/B Power Operated STC Category A/B Stroke Time Test Closed Category A/B Power Operated ET Category A/B Exercise Test (Non- Category A/B Power Operated & Manual Valves Timed)

PS Partial Stroke Exercise Test MSIVs CTO Category C Exercise Test Open Self Actuating (i.e., Check Valves)

CTC Category C Exercise Test Closed Self Actuating (i.e., Check Valves)

DT Category D Test Rupture Disk (RD) & Explosively Actuated Valves (EV)-

Class 2, 3 & Augmented RDs: ISTC-5250 EVs (TIP & SBLC): ISTC-5260 PIT Position Indication Test Passive/Active Valves with Remote Position Indication RT Relief Valve Test ASME Class 1: Appendix I, Section I-3310, I-3410 ASME Class 2, 3 & Augmented: Appendix I, Section I-3350 ASME Class 2 & 3 Vacuum Relief Valves: Appendix I, Section I-3370 ASME Class 2&3 Thermal Relief Valves, Appendix I, Section I-1390 FO Fail Safe Open Test Category A/B Power Operated Valves with Fail-Safe Function FC Fail Safe Closed Test Category A/B Power Operated Valves with Fail-Safe Function DI Disassemble/Inspect (Examination) Check Valves Test XT Excess Flow Check Valve Test Excess Flow Check Valve Closure and Category A leakage test performed concurrently Page 11 of 20

Monticello Nuclear Generating Plant Revision 0 3.2.15 Freq - The frequency at which the valve test is performed to fulfill the requirements of ISTC, Table ISTC-3500-1,. The following abbreviations are used for test frequencies:

Frequency Description Q Quarterly CS Cold Shutdown RF Refueling Y2 Every Two Years Y5 Every Five Years SY6 Every Six Years (72 months)

Class 1 Safety and Relief Valves:

Once per 6 yrs minimum & Group Sampling every 24 months per VR 04 AJ Per Appendix J Program SY5 Class 2, 3 and augmented non reclosing pressure relief device replacement interval.

SY10 Class 2, 3 and Augmented Safety and Relief Valves:

Once per 10 yrs min & Group Sampling every 48 months.

Category D Explosively Actuated Valves, Maximum life based on manufacturers service life or 10 years, whichever is less.

Class 2&3 Thermal Relief Devices (Group Sampling does not apply)

SY8 Applicable Check Valves:

Once per 8 yrs min & Group Sampling every Refueling Outage TS Technical Specification (TS)-

Those components for which credit can be taken for the Applicable TS Test frequency.

CM Check valves included in the Check Valve Condition Monitoring (CVCM)

Program will be tested at the frequency specified in the applicable CVCM Program Plan for the valve group.

PV Periodic Verification. Category A&B MOVs will have their stroke time and position indication tests deferred to the frequency specified in the MOV Program for diagnostic testing when allowed per Valve Relief Request VR 03.

3.2.16 VR/DTJ -Relief request numbers for valves are prefixed with VR. Deferred testing justifications refers to cold shutdown and refueling outage justifications.

These justifications are listed when the test frequency is cold shutdown or refueling instead of quarterly and are prefixed with DTJ.

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Monticello Nuclear Generating Plant Revision 0 3.3 General Valve Testing and Measurement of Test Quantities 3.3.1 Stroke Time:

Stroke time is that time interval from initiation of the actuating signal to the end of the actuating cycle. Stroke time reference values for each power operated valve are specified in the valve test table for the open or closed direction. Stroke time for all power operated valves is measured to the nearest one-tenth of a second.

3.3.2 Position Indication:

Valve obturator movement is determined by exercising the valve while observing an appropriate indicator that signals the required change of obturator position.

Actual valve movement or observing indirect evidence, such as changes in system pressure, flow rate, level or temperature, which reflect stem or disk position will be used to verify that remote valve position indicators agree with valve travel direction, if practicable.

3.3.3 Seat Leakage:

Category A containment isolation valves with a leakage rate requirement based on Appendix J program commitment shall be tested in accordance with the 10 CFR 50, Appendix J Program. Appendix J Program valves are identified in USAR 5.2.

For valves outside the scope of Appendix J; seat leakage is measured by one of the following methods:

a) measuring leakage through a downstream telltale connection while maintaining test pressure on one side of the valve, or b) measuring the feed rate required to maintain test pressure in the test volume or between two seats of a gate valve, provided the total apparent leakage rate is charged to the valve or gate valve seat being tested, and that the conditions required by ISTC-3630(b) are satisfied; or c) determining leakage by measuring pressure decay in the test volume, provided the total apparent leakage rate is charged to the valve or valve combination or gate valve seat being tested, and that the conditions required by ISTC-3630(b) are satisfied.

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Monticello Nuclear Generating Plant Revision 0 3.3.4 Check Valves

Check Valve Exercise - Obturator Movement (ISTC-5221) a) During exercise testing with flow, the necessary obturator movement shall be demonstrated by performing both an open and a close test.

[ISTC-5221(a)].

(1) Check valves that have a safety function in both the open and close directions shall be exercised by initiating flow and observing that the obturator has traveled to either the full open position or the position required to perform its intended function(s) and verify that on cessation or reversal of flow, the obturator has traveled to its seat.

(2). 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 the position required to perform its intended function(s) and verify closure.

(3) 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 at least to the partially open position (normal or expected system flow), and verify that on cessation of reversal of flow, the obturator has traveled to the seat.

Observations shall be made by observing a direct indicator (e.g.

position indicating device) or other positive means (e.g. changes in system pressure, flow rate, level, temperature, seat leakage testing, or non-intrusive testing results.

b) If a mechanical exerciser is used to exercise a valve, the force or torque required to move the obturator and fulfill its safety function(s) shall meet the acceptance criteria specified by MNGP [ISTC-5221(b)].

If practicable, the force(s) or torque(s) required to move the obturator and fulfill any non-safety function should be evaluated to detect abnormality or erratic action for corrective action. The following shall be considered when determining acceptance criteria for mechanical exercising:

(1) Exercise test(s) shall detect a missing obturator, sticking (closed or open), binding (throughout obturator movement), and the loss of any weight(s). Both an open and closed test may not be required.

(2) Acceptance criteria shall consider the specific design, application, and historical performance. (Previous editions of the Code required breakaway force to not vary by more than 50%

from the established reference value).

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Monticello Nuclear Generating Plant Revision 0 (3) If impracticable to detect a missing obturator or the loss or movement of any weight(s) using a mechanical exerciser, other positive means may be used (e.g., seat leakage tests and visual observations to detect obturator loss and the loss or movement of external weight(s), respectively).

c) Per ISTC-5221(c); If the test methods in ISTC-5221(a) (flow exercising) and ISTC-5221(b) (mechanical exercising) are impractical for certain check valves, or if sufficient flow cannot be achieved or verified, a sample disassembly and inspection program shall be used to verify obturator movement. If maintenance is performed on one of these valves that could affect its performance, the post-maintenance testing shall be conducted in accordance with ISTC-5221(c)(4).

Check valves that will be disassembled and inspected shall be grouped by similar design, application, and service condition and require a periodic examination of one valve from each group each refueling outage. The details and bases of the sampling program shall be documented and recorded in the test plan. The following shall be considered when implementing a sample disassembly and inspection program:

(1) Grouping of check valves for the sample disassembly and inspection program shall be technically justified and shall consider, as a minimum, valve manufacturer, design, service, size, materials of construction, and orientation. [ISTC-5221(c)(1)]

Maintenance and modification history should be considered in the grouping process. Valve groupings should also consider potential flow instabilities, required degree of disassembly, and the need for tolerance or critical dimension checks.

(2) During the disassembly process, the full stroke motion of the obturator shall be verified. Full stroke motion of the obturator shall be verified immediately prior to completing reassembly.

Check valves that have their obturator disturbed before full stroke motion is verified shall be examined to determine if a condition exists that could prevent full opening or reclosure of the obturator. Examples of valves that could have their obturators disturbed prior to verifying full stroke motion include; spring loaded check valves or check valves with the obturator supported from the bonnet. [ISTC-5221(c)(2)]

(3) At least one valve from each group shall be disassembled and inspected each refueling outage; and all valves in the group be Page 15 of 20

Monticello Nuclear Generating Plant Revision 0 disassembled and inspected at least once every 8 years. [ISTC-5221(c)(3)]

(4) Before return to service, valves that were disassembled for inspection or that received maintenance that could affect their performance, shall be exercised full- or part-stroke, if practicable, with flow in accordance with ISTC-3520. Those valves shall also be tested for other requirements (e.g., closure verification or leak rate testing) before returning them to service.

[ISTC-5221(c)(4)]

Series Valves in Pairs (ISTC-5223)

If two check valves are in a series configuration without provisions to verify individual reverse flow closure and the plant safety analysis assumes closure of either valve (but not both), the valve pair may be operationally tested closed as a unit. If the plant safety analysis assumes that a specific valve or both valves of the pair close to perform the safety function(s), the required valve(s) shall be tested to demonstrate individual valve closure. Series check valves are identified in Technical Position (TP) 03.

Check Valve Condition Monitoring (ISTC-5222 and Appendix II)

As an alternative to the requirements of paragraphs ISTC-3510, ISTC-3520, ISTC-3530, ISTC-3550, and ISTC-5221, MNGP may establish a Check Valve Condition Monitoring (CVCM) Program per ISTC-5222. The purpose of this program is to both (a) improve check valve performance and to (b) optimize testing, examination, and preventive maintenance activities in order to maintain the continued acceptable performance of a select group of check valves. MNGP may implement this program on a valve or a group of similar valves basis.

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

(1) have an unusually high failure rate during inservice testing or operations (2) cannot be exercised under normal operating conditions or during shutdown (3) exhibit unusual, abnormal, or unexpected behavior during exercising or operation (4) the Owner elects to monitor for improved valve performance b) Examples of candidates for (b) optimization of testing, examination, and preventive maintenance activities are check valves with documented acceptable performance that:

(1) have had their performance improved under the Check Valve Condition Monitoring Program Page 16 of 20

Monticello Nuclear Generating Plant Revision 0 (2) cannot be exercised or are not readily exercised during normal operating conditions or during shutdowns (3) can only be disassembled and examined (4) the Owner elects to optimize all the associated activities of the valve or valve group in a consolidated program.

The program shall be implemented in accordance with Appendix II, Check Valve Condition Monitoring Program, of OM Code-2004 and a site administrative procedure. Site implementing procedures which perform the specified tests are identified in the individual Check Valve Condition Monitoring (CVCM) Program Plans.

If the Appendix II CVCM Program for a valve or group of valves is discontinued then the requirements of ISTC-3510, ISTC-3520, ISTC-3530, ISTC-3550, and ISTC-5221 shall be implemented.

3.3.5 Set Pressure:

Set pressure testing for relief devices is measured by one of the following methods:

a) Pressure Relief Valves - valves requiring a set pressure measurement may be tested in place or removed for bench testing. Valves designed to operate on steam shall be set pressure tested using saturated steam. Valves on systems using other compressible fluids shall be tested with the normal operating fluid. Valves used on liquid service systems shall be tested with the normal system operating fluid for which they were designed.

Alternative test media may be used provided the requirements contained in Appendix I of the OM Code are met.

b) Reclosing Relief Devices (i.e., Vacuum Relief or Set Pressure) - shall be actuated to verify open and close capability, set pressure, and performance of any pressure and position sensing accessories.

NOTE: Non-reclosing pressure relief devices used in BWR scram accumulators are excluded from the requirements of the OM Code.

(ISTC1200) c) Non-Reclosing Rupture Disks - Devices are visually inspected upon receipt, functional testing is not required. Devices are periodically replaced as provided for in Appendix I of the OM Code.

3.3.6 Manual Valves (Re: TP 11)

Pursuant to ISTC-3540 of OMb-2006, manual valves shall be exercised every 2 years. Exercise testing shall be considered acceptable if valve stem travel exhibits unrestricted movement with no abnormal resistance or binding through one complete cycle. Where practical, process parameters may be utilized to verify obturator movement. However, where process parameters are utilized to verify obturator movement it is not necessary to be performed simultaneous to manual exercising. This testing methodology is consistent with the discussion provided in Page 17 of 20

Monticello Nuclear Generating Plant Revision 0 NUREG-1482, Revision 1, Section 4.4.3. If a valve fails to exhibit the required change of obturator position, the valve shall immediately declared inoperable.

The use of a valve persuader (cheater) for additional mechanical advantage will not invalidate the test, as it is recognized that larger valves may exhibit increased packing friction and/or increased friction associated with the disk to seat interface.

In addition, a valve persuader may be used for personnel safety depending on a valve's service application (i.e. main steam).

3.3.7 Skid-Mounted Components Skid-mounted valves are excluded from Subsection ISTC, provided they are tested as part of the major component and are justified by MNGP to be adequately tested. Further discussion pertaining to skid-mounted valves and component subassemblies is provided in NUREG-1482, Rev.1, Sections 3.4 and 4.1.10.

Skid-Mounted pumps and valves are those which are integral to or that support operation of major components, even though these pumps and valves may not be located on the skid. In general, these valves are supplied by the manufacturer of the major component. Examples include: air start valves associated with the emergency diesel generators, and solenoid operated pilot valves used to control air operated valves and ADS/Safety Relief valves. Components considered as skid-mounted are identified in Technical Position (TP) 05.

3.4 Allowable Ranges of Test Quantities 3.4.1 Stroke Time:

Stroke times shall be compared to the initial reference values established per the OM Code. Stroke times shall meet the criteria listed below. Valves not meeting this criteria shall be immediately retested or declared inoperable. Valves declared inoperable may be repaired, replaced, or the data may be analyzed to determine the cause of the deviation and the valve shown to be operating acceptably. Valve operability based on analysis shall have the results of the analysis recorded in the record of tests. (ISTC-5115, ISTC-5123, ISTC-5133, ISTC-5143, ISTC-5153) a) Electric motor operated valves with reference stroke times greater than 10 sec. shall exhibit no more than +/-15% change in stroke time when compared to the reference value.

b) Electric motor operated valves with stroke times less than or equal to 10 sec.

shall exhibit no more than a +/-25% or +/-1 sec. change in stroke time, whichever is greater, when compared to the reference value.

NOTE: As an alternative to the requirements of paragraph ISTC-5120 of the ASME OM Code-2004 through OMb-2006, Code Case OMN-1, Revision 1, "Alternative Rules for Preservice and Inservice Testing of Certain Electric Motor-Operated Valve Assemblies in LWR Power Plants provides an alternative to MOV stroke time testing.

Page 18 of 20

Monticello Nuclear Generating Plant Revision 0 MNGP may adopt the alternative test requirements specified in ASME OM Code Case OMN-1, Revision 1, in lieu of stroke timing certain motor operated valves (MOVs) in accordance with the requirements specified in paragraph ISTC-5120 and position indication testing in accordance with the requirements specified in paragraph ISTC-3700. The MNGP MOV Program satisfies the criteria specified in ASME OM Code Case OMN-1, Revision 1, and the conditional acceptance specified in Reg. Guide 1.192, Operation and Maintenance Code Case Acceptability, ASME OM Code. Paragraph 3.6 of OMN-1 requires MOVs to be full stroke exercised (not timed) to the position(s) required to fulfill their function(s) on an interval not to exceed one year or one refueling cycle (which ever is longer). Full stroke exercising is based on the practicality of exercising during power operation, cold shutdown, or refueling. Justification for extending full stroke exercising of ASME OM Code Case OMN-1 scoped MOVs beyond a quarterly frequency are provided in Attachment 6 of the MNGP IST Program. Further guidance regarding the use of ASME OM Code Case OMN-1 is provided in NUREG-1482, Revision 1, Section 4.2.5, Alternatives to Stoke-Time Testing. Also, refer to Valve Relief Request VR-03.

c) Other power operated valves with reference stroke times greater than 10 sec.

shall exhibit no more than +/-25% change in stroke time when compared to the reference value.

d) Other power operated valves with reference stroke times less than or equal to 10 sec. shall exhibit no more than +/-50% change in stroke time when compared to the reference value.

e) Valves that stroke in less than 2 sec. may be exempted from (d) above. In such cases the maximum limiting stroke time shall be 2 sec.

f) Valve stroke time shall not exceed either the values used to satisfy MNGP Technical Specifications or the owners established stroke times.

3.4.2 Position Indication:

The valve travel direction (open/closed) will agree with remote position indicators.

3.4.3 Seat Leakage:

Valve leakage rates shall not exceed either the value specified by Technical Specifications or the Owners established leakage rates.

3.4.4 Set Pressure:

Set pressure shall not exceed the greater of either the +/- tolerance limit of the Owner-established set pressure acceptance criteria or +/- 3% of the valves nameplate set pressure.

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Monticello Nuclear Generating Plant Revision 0 3.4.5 Instrument Accuracy Instruments used to measure stroke times shall be capable of measurement to the nearest tenth of a second.

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Monticello Nuclear Generating Plant Revision 0 ATTACHMENT 1 IST Plan System Listing & Acronyms SYSTEM ACRONYM COMMON USE (Site 3-Letter ID) ACRONYM Alternate N2 AN2 -

Auto Pressure Relief APR SRV Condensate and Feedwater CFW -

Condensate Storage CST -

Control Rod Drive Hydraulics CRH CRD Core Spray Cooling CSP CS Demineralized Water System DWS -

Diesel Generators DGN -

Diesel Oil Storage DOL -

EDG Emergency Service Water ESW -

EFT Emergency Service Water FSW -

Fuel Pool Cooling and Cleanup FPC -

High Pressure Coolant Injection HPC HPCI Hydrogen Oxygen Analyzer HOA -

Instrument and Service Air AIR -

Liquid Radwaste LRW -

Main Condenser CDR -

Main Steam MST -

Post Accident Sampling PAS PASS Primary Containment PCT -

Reactor and Vessel Assembly RPV -

Reactor Building Closed Cooling Water RBC RBCCW Reactor Core Isolation Cooling RCI RCIC Reactor Recirculation REC -

Reactor Water Clean-up RWC RWCU Residual Heat Removal RHR -

RHR Service Water RSW RHRSW Service Water SSW SW Standby Liquid Control SLC SBLC Traversing In Core Probe TIP -

Page 1 of 1

Monticello Nuclear Generating Plant Revision 0 ATTACHMENT 2 IST Plan Notes (Notes referenced as superscript in IST Plan Document as applicable)

1. MNGP is licensed to meet the Hot Shutdown Plant Condition for Safe Shutdown.

2. Skid-Mounted Components will be detailed in the MNGP IST Program (and/or Administrative related documents) by Technical Position. Skid-mounted pumps and valves will not be included in Attachment 8 and Attachment 9.

Page 1 of 1

Monticello Nuclear Generating Plant Revision 0 ATTACHMENT 3 Index For IST Plan Relief Requests and Deferred Testing Justifications Pump Relief Requests (PR)

PR 01 SLC Test Methods PR 02 RHR Instrumentation Range PR 03 HPCI Pump Vibration PR 04 HPCI/RCIC Instrument Range PR 05 SLC Pump Vibration Frequency Response Range PR 06 HPCI Pump Testing Utilizing Pump Curves Valve Relief Requests (VR)

VR 01 Closure Testing of Scram Discharge Header Check Valves (CRD-114)

VR 02 RHRSW Flow Control Valve Exercising VR 03 Use of Code Case OMN-1, Revision 1, on Various Motor Operated Valves VR 04 Use of Code Case OMN-17, Revision 0, on the Class 1 Main Steam Safety Relief Valves Deferred Testing Justifications (DTJ)

DTJ 01 RBCC-15 DTJ 02 CST-88, CST-90, CST-92, CST-94 DTJ 03 XR-27-1, XR-27-2, XR-25-1, XR-25-2 DTJ 04 (Deleted)

DTJ 05 CRD-115 DTJ 06 RHR-8-1, RHR-8-2 DTJ 07 XP-6, XP-7 DTJ 08 RC-6-1, RC-6-2 DTJ 09 AI-626-1, AI-625 DTJ 10 AO-1825A, AO-1825B DTJ 11 MO-1426, MO-4229, MO-4230 DTJ 12 FW-91-1, FW-91-2, FW-94-1, FW-94-2, FW-97-1, FW-97-2 DTJ 13 AO-2-80A, AO-2-80B, AO-2-80C, AO-2-80D (Fail-Safe testing)

DTJ 14 MO-2-53A, MO-2-53B DTJ 15 AO-23-18, AO-13-22 DTJ 16 MO-2029, MO-2030 DTJ 17 HPCI-9, HPCI-10, HPCI-14, HPCI-15, HPCI-31, HPCI-65, HPCI-71, HPCI-18 DTJ 18 RCIC-9, RCIC-10, RCIC-16, RCIC-17, RCIC-31, RCIC-57, RCIC-59 DTJ 19 AI-598, AI-708, AI-729, AI-730, SV-4235 DTJ 20 PAS-59-5, PAS-59-6 Page 1 of 2

Monticello Nuclear Generating Plant Revision 0 ATTACHMENT 3 Deferred Testing Justifications (DTJ) continued DTJ 21 BF-12, BF-14, BF-24, BF-26, BF-35, BF-37, BF-46, BF-48 DTJ 22 XFV-1 through XFV-89 DTJ 23 AI-571 DTJ 24 AI-613 through AI-619, AI-663, AI-666, AI-669, AI-672, AI-675, AI-678, AI-681, AI- 683, AI-685, AI-694, AI-695 DTJ 25 AI-13-4, AI-13-7, AI-610-1 through AI-610-4, AI-611, AI-612 DTJ 26 AO-10-46A, AO-10-46B DTJ 27 AO-14-13A, AO-14-13B DTJ 28 AO-2382A through AO-2382C, AO-2382E through AO-2382H, AO-2382K DTJ 29 HPCI-32, RCIC-41 DTJ 30 HPCI-42, RCIC-37 DTJ 31 CRD-114, CRD-115, CRD-138, CV-126, CV-127 DTJ 32 PC-20-1, PC-20-2 DTJ 33 AI-12-9 through AI-12-12 DTJ 34 RHR-81 DTJ 35 AI-243-1, AI-243-2, AI-244-1, AI-244-2 DTJ 36 CST-189, CST-98 DTJ 37 XP-3-1, XP-3-2 DTJ 38 CV-3-32A, CV-3-32B, CV-3-32C, CV-3-32D, CV-3-33A, CV-3-33B, CV-3-33C, CV-3-33D, SV-3-31A, SV-3-31B, SV-3-31C, and SV-3-31D.

DTJ 39 (Deleted)

DTJ 40 AO-2-80A through AO-2-80D (Full Stroke Exercise Testing)

AO-2-86A through AO-2-86D (Full Stroke Exercise and Fail-Safe Testing)

DTJ 41 MO-2397 and MO-2398 DTJ 42 MO-1753, MO-1754, MO-2014 and MO-2015 DTJ 43 ESW 1-1 and ESW 1-2 DTJ 44 (Deleted)

DTJ 45 (Deleted)

Page 2 of 2

Monticello Nuclear Generating Plant Revision 0 ATTACHMENT 4 Pump Relief Requests (PR)A,B A

IST Relief Request formats were developed following the guidelines provided in NUREG-1482, Rev.1, Attachment 1, NEI White Paper, Entitled Standard Format for Requests from Commercial Reactor Licensees Pursuant to 10 CFR 50.55a, Revision 1 dated June 2004.

B Reference Attachment 10 for NRC Safety Evaluations for approved Relief Requests. (Pending)

Page 1 of 27

Monticello Nuclear Generating Plant Revision 0 10 CFR 50.55a REQUEST NUMBER - PR 01 SLC Pump Test Method Proposed Alternative In Accordance with 10CFR50.55a(a)(3)(ii)

On the basis that compliance with the OM Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety

1. ASME Code Component(s) Affected P-203A and P-203B, Standby Liquid Control (SLC) Pumps (Class 2) (Group B)

Component/System Function The design objective of the Group B SLC Pump is to deliver a boron solution at a sufficient concentrate to the reactor and bring it to a safe shutdown condition from full power in the event of failure of the withdrawn control rods to insert.

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

3. Applicable Code Requirement(s)

ISTB-3550, Flow Rate, states in part, when measuring flow rate, a rate or quantity meter shall be installed in the pump test circuit. If a meter does not indicate the flow rate directly, the record shall include the method used to reduce the data.

ISTB-5300(a), Duration of Tests (Positive Displacement Pumps), ISTB-5300(a)(1),

states in part, for the Comprehensive test, after pump conditions are stable as the system permits, each pump shall be run at least two minutes. At the end of this time at least one measurement or determination of each of the quantities required by Table ISTB-3000-1 shall be made and recorded.

4. Reason for Request The Group B positive displacement SLC pumps are designed to pump a constant flow rate regardless of system resistance. The SLC system was not designed with a flow meter in the flow loop. The system was designed to be tested using a test tank where the change in level can be measured over time. This test methodology also limits the pump run time based on the size of the test tank. In addition, the installation of a larger test tank to facilitate pump testing would be a burden because of the design, fabrication, and installation changes that would be required.

The Code requirements to use flow rate instrumentation and a two-minute test duration for the Comprehensive pump test are considered a burden, which would result in a hardship without a corresponding increase in the level of quality or safety.

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Monticello Nuclear Generating Plant Revision 0

5. Proposed Alternative and Basis for Use NSPM proposes to determine pump flow rate by measuring changes in tank level over time. After taking an initial pump volume measurement, an SLC pump will be started with suction from the demineralized water system and will discharge to the test tank. After at least two minutes of operation the pump will be stopped and the change in level over the measured time will be converted to flow rate by the following formula:

Q (GPM) = L (In)/t (Sec)

Where is a constant which reflects tank dimensions and unit conversions.

The test tank level will be set at approximately the same level at the beginning of each test to ensure repeatability. Use of the demineralized water system ensures a large volume source at a constant pressure for a stable testing environment. The vibration testing will be performed while recirculating an adequately filled test tank.

Therefore, the duration of test code requirements for vibration testing will be met.

This proposed alternative is consistent with the guidelines provided in NUREG-1482, Revision 1, Section 5.5.2, Use of Tank Level to Calculate Flow Rate for Positive Displacement Pumps.

Using the provisions of this relief request as an alternative to the test requirements specified in ISTB-3550 and ISTB-5300(a)(1) is an acceptable alternative method that provides reasonable assurance of operational readiness of SLC pumps P-203A and P-203B.

A plant IST instrument accuracy calculation determined that the accuracy of SLC flow measurement was better than 2% (approximately 1.4%). This calculation used the dimensions of the tank, the method of measuring the change in depth of water in the tank and the method of measuring time to evaluate accuracy by a square root of the sum of the squares method.

6. Duration of Proposed Alternative The proposed alternative identified in this relief request shall be implemented during the Fifth Ten Year IST Interval beginning September 1, 2012.

Page 3 of 27

Monticello Nuclear Generating Plant Revision 0

7. Precedents NRC Safety Evaluation for Monticello Nuclear Generating Plant, Relief Requests Relating to the Fourth 10-Year Interval Inservice Testing Program, Docket No. 50-263. (TAC No. 6807), July 17, 2003 Page 4 of 27

Monticello Nuclear Generating Plant Revision 0 10 CFR 50.55a REQUEST NUMBER - PR 02 RHR Instrumentation Range Proposed Alternative In Accordance with 10CFR50.55a(a)(3)(i)

On the basis that compliance with the provides an acceptable level of quality and safety

1. ASME Code Component(s) Affected P-202A/B/C/D, 11/12/13/14 Residual Heat Removal (RHR) Pumps (Class 2) (Group A)

P-109A/B/C/D, 11/12/13/14 Residual Heat Removal Service Water (RHRSW)

Pumps (Class 3) (Group A)

Component/System Function The Group A RHR pumps (P-202A/B/C/D) must operate to satisfy Low Pressure Coolant Injection and Containment Spray/Cooling requirements during post-accident conditions as well as performing a function during normal shutdown cooling.

The Group A RHRSW pumps (P-109A/B/C/D) are required to operate to remove the heat rejected by the residual heat removal system during normal shutdown and accident operations. The pumps must also supply a source of water for the RHR-RHR Service Water Intertie.

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

3. Applicable Code Requirement(s)

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

4. Reason for Request Flow transmitters FT-10-111A, FT-10-111B, FT-10-97A, and FT-10-97B are each designed to indicate flow while two parallel pumps are operating (RHR or RHRSW).

During In-Service testing, only one pump operates at a time. The resulting reference value of flow for one pump is less than one-third of the instruments range. The installed flow transmitters have typically had an as-found accuracy of about 0.25% of full scale. In addition, the system is verified to have an as-found accuracy that is within 2% of the Code allowed reference value for analog instruments.

The current relevant data for the instruments is included in Table 4-1 as follows:

Page 5 of 27

Monticello Nuclear Generating Plant Revision 0 Table 4-1 Instrument Pumps Instrument Equivalent Range to Span (Range) Reference Reference Value Value Ratio RHR Pumps FT-10-111A P-202A 4-20 mA 6.4 mA (16/2.4) = 6.67 P-202C FT-10-111B P-202B 4-20 mA 6.4 mA (16/2.4) = 6.67 P-202D RHRSW Pumps FT-10-97A P-109A 10-50 mA 18.39 mA (40/8.39) = 4.77 P-109C FT-10-97B P-109B 10-50 mA 18.57 mA (40/8.57) = 4.67 P-109D Transmitters FT-10-111A, FT-10-111B, and FT-10-97A output signals are read on a mV display with the pump test procedures specifying a reference target range that corresponds one to one mV to mA. The transmitter FT-10-97B output signal is converted from a 10-50 mA range to a 4-20mA range via FY-4105, RHR SERVICE WATER FLOW ISOL, and read on a mV display with the pump test procedures specifying a reference target range that corresponds one to one mV to mA of the converted signal range. The equivalent reference value is the center of this reference flow signal range and is in mA. Dividing the transmitter range by the equivalent reference mA value shows the instrument range to exceed the reference value by more than a factor of 3.

FY-4105 output equivalent reference value is 7.43 mA. Thus the range to reference value ratio is also (16/3.43) = 4.67 when taken at the FY-4105 output which is equivalent to the (40/8.57) = 4.67 at FT-10-97B output.

The installed flow transmitters typically have an as-found accuracy of 0.25 percent of full scale. In addition, the system is verified to have an as-found accuracy in accordance with paragraph ISTB-3510(a). ISTB-3510(a) requires that instrument accuracy be within the limits of Table 3510-1, which specifies an accuracy requirement of +/-2% of full-scale for analog flow instruments. Paragraph ISTB-3510(b)(1) requires that the full-range of each analog instrument be no greater than three times the reference value. The combination of these two requirements (i.e.,

accuracy equal to +/-2% of full-scale and full-scale being up to 3 times the reference value) yields a permissible inaccuracy of +/-6% of the reference value.

Table 4-2 shows the ranges, reference values, range to reference value ratio, and calculated effective accuracies for instruments FT-10-111A/B and FT-10-97A/B. The calculated effective instrument accuracies are much less than the Code required effective accuracy of +/- 6 percent. Therefore, these instruments yield readings at least equivalent to the reading achieved from instruments that meet OM Code requirements (i.e., up to +/- 6 percent) and, thus, provide an acceptable level of quality Page 6 of 27

Monticello Nuclear Generating Plant Revision 0 and safety. The effective accuracy of each instrument is provided in the following table.

Table 4-2 Instrument Pump Instrument Reference Range to Effective Range Value Reference Accuracy with Value +/-0.25%

Instrument Accuracy RHR Pumps FT-10-111A P-202A 16 mA (6.4-4)= (16/2.4)= (6.67 x 0.25%)=

P-202C 2.4 mA 6.67 +/-1.67%

FT-10-111B P-202B 16 mA (6.4-4)= (16/2.4)= (6.67 x 0.25%)=

P-202D 2.4 mA 6.67 +/-1.67%

RHRSW Pumps FT-10-97A P-109A 40 mA (18.39-10)= (40/8.39)= (4.77 x 0.25%)=

P-109C 8.39 mA 4.77 +/-1.2%

FT-10-97B P-109B 40 mA (18.57-10)= (40/8.57)= (4.67 x 0.25%)=

P-109D 8.57 mA 4.67 +/-1.17%

5. Proposed Alternative and Basis for Use Use the existing station instruments to measure pump In-Service test parameters.

Perform a loop check on the flow instrumentation for these systems that verifies the AS FOUND accuracy is within the 2% accuracy requirement given in Table ISTB 3510-1, Required Instrument Accuracy, and within the range required of 3 times the reference value of any RHR or RHRSW pump. This will be done as part of the routine calibration task.

Using the provisions of this relief request as an alternative to the requirements of ISTB-3510(b)(1) provides an acceptable level of quality and safety since their use yields a reading that is as at least equivalent to that achieved using instruments that meet the Code requirements as described in NUREG-1482, Rev.1, Section 5.5.1.

6. Duration of Proposed Alternative The proposed alternative identified in this relief request shall be implemented during the Fifth Ten Year IST Interval beginning September 1, 2012.

7. Precedents NRC Safety Evaluation for Monticello Nuclear Generating Plant, Relief Requests Relating to the Fourth 10-Year Interval Inservice Testing Program, Docket No. 50-263. (TAC No. MB6807), July 17, 2003 Page 7 of 27

Monticello Nuclear Generating Plant Revision 0 10 CFR 50.55a REQUEST NUMBER - PR 03 HPCI Pump Vibration Proposed Alternative In Accordance with 10CFR50.55a(a)(3)(ii)

On the basis that compliance with the OM Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety

1. ASME Code Component(s) Affected P-209, High Pressure Coolant Injection (HPCI) Pump (Class 2) (Group B)

Component/System Function The HPCI System is designed to pump water into the reactor vessel under loss-of-coolant conditions which do not result in rapid depressurization of the pressure vessel. The loss-of-coolant might be due to loss of reactor feedwater or to small line breaks which do not cause depressurization of the reactor vessel [Updated Safety Analysis Report 6.2.4].

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

3. Applicable Code Requirement(s)

Table ISTB-5121-1, Centrifugal Pump Test Acceptance Criteria imposes a vibration alert limit of > 0.325 in/Sec when performing a Comprehensive pump test.

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

4. Reason for Request The Monticello Nuclear Generating Plant (MNGP) Group B HPCI pump consists of a centrifugal main pump, a separate centrifugal booster pump, a speed reducing gear for the booster pump, and a Terry turbine steam driver. The Terry turbine and main pump operates during testing near 3590 RPM. The booster pump operates at a lower RPM due to the reducing gear. All these components are mounted horizontally along the same drive train. Therefore, there are four independently balanced and aligned rotating assemblies that are coupled together. As a result, the normal (baseline) vibration readings in the horizontal direction on the main pump is approximately 0.423 in/Sec. Application of a 0.325 in/Sec alert limit would require Northern States Power - Minnesota (NSPM) to enter accelerated test frequency each time the pump was tested because one or more of these points measured would exceed this limit.

Page 8 of 27

Monticello Nuclear Generating Plant Revision 0 NSPM has many years of In-Service test data showing that baseline vibrations at 0.423 in/Sec represent acceptable pump operation. NSPM has had these vibration levels analyzed by an engineering consultant that specializes in vibration analysis.

The consultants analysis shows that this pump can operate at vibration levels up to 0.700 in/Sec.

Component industry history was reviewed for this type of pump. No failures attributed to extended hours of pump operation at vibration levels exceeding 0.325 in/Sec were found. Implementing the alert limit of 0.325 in/Sec would require NSPM to constantly have the HPCI pump on accelerated test frequency. This would result in an annual comprehensive pump In-Service test instead of biennial. The intent of increased test frequency is to closely monitor a pump that is deteriorating from its baseline values. In this case, the pump is operating at its normal vibration range and no change would be seen. The additional annual test would require a significant amount of time and resources.

Modifying the system in an attempt to reduce the vibration levels, such as installing new shafts and impellers, are extremely expensive and may not reduce the vibration levels. Therefore, requiring an alert limit of 0.325 in/Sec on the HPCI pump is an extreme hardship without a compensating increase in public safety. An appropriate alert limit for these vibration data points is 0.500 in/Sec. This is based on previous test history, a review of industry data and the vibration analysis performed.

5. Proposed Alternative and Basis for Use A vibration alert limit of 0.500 in/Sec will be used for the pump horizontal vibration data points. The Table ISTB-5121-1 required action limit of 0.700 in/Sec will be adhered to.

NSPMs evaluation of the historical HPCI pump vibration data, shows that HPCI pump P-209 normally runs at elevated vibration levels and has not experienced any failure to date. Requiring NSPM to meet the OM Code requirements by increasing the frequency of MNGP HPCI pump testing would result in hardship without a compensating increase in the level of quality and safety. This is because of the additional testing that would need to be performed on a pump that adequately operates at elevated vibration levels.

Using the provisions of this relief request as an alternative to the test requirements specified in Table ISTB-5121-1 and ISTB-6200(a) is an acceptable alternative method that provides reasonable assurance of operational readiness of HPCI pump P-209.

6. Duration of Proposed Alternative The proposed alternative identified in this relief request shall be implemented during the Fifth Ten Year IST Interval beginning September 1, 2012.

Page 9 of 27

Monticello Nuclear Generating Plant Revision 0

7. Precedents NRC Safety Evaluation for Monticello Nuclear Generating Plant, Relief Requests Relating to the Fourth 10-Year Interval Inservice Testing Program, Docket No. 50-263. (TAC No. MB6807), July 17, 2003.

Page 10 of 27

Monticello Nuclear Generating Plant Revision 0 10 CFR 50.55a REQUEST NUMBER - PR 04 HPCI/RCIC Instrumentation Range Proposed Alternative In Accordance with 10CFR50.55a(a)(3)(i)

On the basis that the proposed alternative provides an acceptable level of quality and safety

1. ASME Code Component(s) Affected P-209, High Pressure Coolant Injection (HPCI) Pump (Class 2) (Group B)

P-207, Reactor Core Isolation Cooling (RCIC) Pump (Class 2) (Group B)

Component/System Function The HPCI System is designed to pump water into the reactor vessel under loss- of-coolant conditions which do not result in rapid depressurization of the pressure vessel. The loss-of-coolant might be due to loss of reactor feedwater or to small line breaks which do not cause depressurization of the reactor vessel [Updated Safety Analysis Report (USAR) 6.2.4].

The RCIC System consists of a turbine-driven pump unit capable of delivering makeup water to the reactor vessel during the unlikely event feedwater is isolated from the vessel. [USAR 10.2.5].

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

3. Applicable Code Requirement(s)

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

4. Reason for Request The differential pressure for the HPCI and RCIC pumps is determined by subtracting the indicated suction pressure from the indicated discharge pressure. The HPCI pump suction pressure is read in the Control Room from instrument PI-23-116, which is sent a 10 to 50 mA signal from local transmitter PT-23-100. The RCIC pump suction pressure is read locally from instrument PI-13-66. The current instrument ranges exceed three times the current reference values. The relevant data for the instruments is included in Table 4-1 as follows:

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Monticello Nuclear Generating Plant Revision 0 Table 4-1 Instrument Pump Instrument Reference Range to Range Value Reference Value Ratio PI-23-116 P-209 30 Hg - 100 psi 33.7 psi 114.7/33.7=3.4 (Note 1)

PT-23-100 P-209 10 - 50 mA 11.8 mA* 40/11.8=3.4 (Note 2)

PI-13-66 P-207 30 Hg - 100 psi 33.7 psi 114.7/33.7=3.4 (Note 1)

21.8 mA equates to 11.8 mA on the 40 mA span NOTE 1: The vacuum range for the pressure indicators was converted to PSI for determining the ratio. 30 HG Vacuum = 14.7 PSI; thus the range = 100 + 14.7 PSI. The same principle was applied to the reference value. With a reference value of 19 PSI indicated on the instrument, the reference value used for the ratio determination is 19 + 14.7 = 33.7 PSI.

NOTE 2: The pressure transmitter has a 10 to 50 mA range, or a span of 40 mA. The ratio for this instrument must be determined by reducing the reference value to its value on the 40 mA span.

Paragraph ISTB-3510(a) requires that instrument accuracy be within the limits of Table 3510-1, which specifies an accuracy requirement of +/-2% of full-scale for analog flow instruments. Paragraph ISTB- 3510(b)(1) requires that the full-scale range of each analog instrument be not greater than three times the reference value.

The combination of the two requirements (i.e., accuracy equal to +/-2% of full-scale and full scale being up to 3 times the reference value) yields a permissible inaccuracy of +/-6% of the reference value.

Group B Tests: In accordance with ISTB-3510(b)(1) Northern States Power -

Minnesota (NSPM) proposes to apply three times the reference value for determination of the OM Code equivalent range for the instruments. The +/- 2% Code required instrument accuracy (see ISTB-3510(a)) for the Group B test is determined from this Code equivalent range as described in Table 4-2 below:

Table 4-2 Instrument Reference Code Equivalent 2% of Code Value Range Equivalent Range PI-23-116 33.7 psi 3 x 33.7 = 101 psi +/- 2 psi PT-23-100* 21.8 mA 3 x 11.8 = 35.4 mA +/- 0.7 mA PI-13-66 33.7 psi 3 x 33.7 = 101 psi +/- 2 psi

21.8 mA equates to 11.8 mA on the 40 mA span Page 12 of 27

Monticello Nuclear Generating Plant Revision 0

4. Reason for Request (Cont)

The current instrument calibration tolerances are +/- 2 psi for the pressure indicators and +/- 0.7 mA for the pressure transmitter. The as-found data in the calibration history for these instruments shows that they have been consistently well within these current code equivalent tolerances.

Comprehensive Tests: The full scale range of pressure transmitter PT-23-100 is approximately 3.4 times the reference value, which is greater than the ISTB-3510(b)(1) requirement of three times the reference value.

Therefore, NSPM proposes that the instrument accuracy requirements of ISTB-3510(a) be demonstrated by determining the loop accuracy using both temporary and in-plant installed instrumentation (PT-23-100). The as-found data in the calibration history for these instruments shows that they have consistently been well within the OM Code equivalent tolerances.

5. Proposed Alternative and Basis for Use NSPMs proposed alternatives to the OM Code requirements of paragraph ISTB-3510(b)(1) as described in the following.

Group B Tests: NSPM will calibrate instruments PI-23-116, PT-23-100, and PI 66 to +/- 2 percent of the OM Code equivalent range for Group B tests. The Code equivalent range will be calculated by multiplying the current test parameter reference value by three.

Comprehensive Tests: NSPM will demonstrate the instrument accuracy requirements of paragraph ISTB-3510(a) by determining the loop accuracy using both temporary and in-plant installed instrumentation (PT-23-100).

Using the provisions of this relief request as an alternative to the requirements of ISTB-3510(b)(1) provides an acceptable level of quality and safety since their use yields a reading that is as at least equivalent to that achieved using instruments that meet the Code requirements as described in NUREG-1482, Rev.1, Section 5.5.1.

6. Duration of Proposed Alternative The proposed alternative identified in this relief request shall be implemented during the Fifth Ten Year IST Interval beginning September 1, 2012.

Page 13 of 27

Monticello Nuclear Generating Plant Revision 0

7. Precedents NRC Safety Evaluation for Monticello Nuclear Generating Plant, Relief Requests Relating to the Fourth 10-Year Interval Inservice Testing Program, Docket No. 50-263. (TAC No. MB6807), July 17, 2003 Page 14 of 27

Monticello Nuclear Generating Plant Revision 0 10 CFR 50.55a REQUEST NUMBER - PR 05 SLC Pump Vibration Frequency Response Range Proposed Alternative In Accordance with 10CFR50.55a(a)(3)(ii)

On the basis that compliance with the OM Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety

1. ASME Code Component(s) Affected P-203A and P-203B, Standby Liquid Control Pumps (SLC) (Class 2) (Group B)

Component/System Function The design objective of the SLC Pumps is to deliver a boron solution at a sufficient concentrate to the reactor and bring it to a safe shutdown condition from full power in the event of failure of the withdrawn control rods to insert.

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

3. Applicable Code Requirement(s)

ISTB-3510(e); General, Frequency Response Range; states, The frequency response range of the vibration-measuring transducers and their readout system shall be from one-third minimum pump shaft rotational speed to at least 1000 Hz.

4. Reason for Request The nominal shaft rotational speed of the SLC pumps is 280 RPM, which is equivalent to approximately 4.7 Hz. Based on this frequency and ISTB-3510(e), the required frequency response range of instruments used for measuring pump vibration is to be 1.56 to 1000 Hz. Procurement and calibration of instruments to cover this range to the lower extreme (1.56 Hz) is impractical due to the limited number of vendors supplying such equipment (and replacement parts), the level of equipment sophistication and the difficulty of the instruments to meet the required accuracy at low frequencies.

These pumps are of a simplified reciprocating (piston) positive displacement design with rolling element bearings, Model Number TD-60, manufactured by Union Pump Corporation. Union Pump Corporation has performed an evaluation of the pump design and has determined that there is no probable sub-synchronous failure modes associated with these pumps under normal operating conditions. Furthermore, there are no known failure mechanisms that would be revealed by vibration at frequencies below those related to Page 15 of 27

Monticello Nuclear Generating Plant Revision 0 shaft speed (4.7 Hz.). Based upon the absence of a credible failure mode, no useful information is obtained by testing below the 4 Hz frequency nor will any indication of pump degradation be masked by instrumentation unable to collect data below this frequency. The requirement to measure vibration with instruments with response to 1/3 shaft speed stems from the need to detect oil whip or oil whirl associated with journal bearings. In the case of these pumps, there are no journal bearings to create these phenomena, thus satisfying the Code requirements of ISTB-3510(e) would serve no significant purpose. The significant modes of vibration, with respect to equipment monitoring, are as follows:

1-Times Crankshaft Speed - An increase in vibration at this frequency may be an indication of rubbing between a single crankshaft cheek and rod end, cavitations at a single valve or coupling misalignment.

2-Times Crankshaft Speed - An increase in vibration at this frequency may be an indication of looseness at a single rod bearing or crosshead pin, a loose valve seat in the fluid cylinder, a loose plunger/crosshead stub connection or coupling misalignment.

Other Multiples of Shaft Speed - An increase in vibration at other frequencies may be an indication of cavitation at several valves, looseness at multiple locations or bearing degradation.

On March 2, 2006 the NRC in a request for additional information to Florida Power and Light (FPL) Energy (licensee for the Duane Arnold Energy Center (DAEC)) indicated that new technologies that would make detection of low frequency shaft rotation were available to licensees. In the FPL Energy response, dated May 8, 2006 (ADAMS Accession No. ML061430462) the licensee demonstrated that the 1/3 turning speed for the DAEC SLC pump was lower (1.33 Hz) than a previous licensee who could use the improved instrumentation. In addition, FPL Energy demonstrated that even with additional signal filtering the integration from acceleration to velocity would create a slope that would prevent obtaining reliable data at frequencies this low. The integration creates a ski slope on the low end, so that at the Code-required low-end frequencies for the DAEC SLC pumps, the data is corrupted by the ski slope and the data would be unreliable. The NRC granted approval of the relief request in a letter dated July 21, 2006 (Reference 1).

The DAEC approval is applicable to the MNGPs SLC pump which has a 1/3 turning speed of 1.56 Hz. Further, other knowledgeable industry IST program personnel indicated that currently available instrumentation would not provide data that is reliable for monitoring of speeds at this low value. Finally, if transducers were obtained that can be reliably calibrated down to 1Hz, the errors received by the other meter and cabling electronic errors would continue to be above 5% at these low values. The NRC previously approved this request for the MNGP IST 4th ten-year interval (Reference 2).

Page 16 of 27

Monticello Nuclear Generating Plant Revision 0 Based on the foregoing discussion, it is clear that monitoring pump vibration within the frequency range of 4 to 1000 Hz will provide adequate information for evaluating pump condition and ensuring continued reliability with respect to the pumps function.

5. Proposed Alternative and Basis for Use Vibration levels of the SLC Pumps will be measured in accordance with the applicable portions of Subsection ISTB with the exception of the lower frequency response limit for the instrumentation listed in Paragraph ISTB-3510(e). The frequency response range for vibration measurement for the SLC pumps shall be 4 to1000 Hz.

Using the provisions of this 10CFR50.55a request as an alternative to the test requirements specified in ISTB-3510(e) is an acceptable alternative method that provides reasonable assurance of operational readiness of SLC pumps P-203A and P-203B.

6. Duration of Proposed Alternative The proposed alternative identified in this relief request shall be implemented during the Fifth Ten Year IST Interval beginning September 1, 2012.

7. Precedents

1. NRC Safety Evaluation for Monticello Nuclear Generating Plant, Relief Requests Relating to the Fourth 10-Year Interval Inservice Testing Program, Docket No. 50-263. (TAC No. MB6807), July 17, 2003

2. NRC Safety Evaluation for Duane Arnold Energy Center, Relief Requests Related to the Fourth 10-Year Interval Inservice Testing program (TACNos. MC8713, MC8784 and MC8785), July 21, 2006

8. References

1. Letter from L. Raghavan (NRC) to Gary Van Middlesworth (DAEC),

Subject:

Duane Arnold Energy Center - Relief Requests Related to the Fourth 10-Year Interval Inservice Testing (IST) program (TAC Nos.

MC8713, MC8784 and MC8785), dated July 21, 2006. (ADAMS Accession No. ML061870011)

2. Letter from L. Raghavan (NRC) to David Wilson (NMC),

Subject:

Monticello Nuclear Generating Plant - Evaluation of Relief Request Nos. PR-01, Pr-02, PR-03, PR-04, PR-05 and VR-02 Related to the Fourth 10-Year Interval Inservice Testing Program (TAC No. MB6807),

dated July 17, 2003. (ADAMS Accession No. ML031700209)

Page 17 of 27

Monticello Nuclear Generating Plant Revision 0 10 CFR 50.55a REQUEST NUMBER - PR 06 HPCI Pump Testing Utilizing Pump Curves Proposed Alternative In Accordance with 10CFR50.55a(a)(3)(i)

On the basis that the proposed alternative provides an acceptable level of quality and safety.

1. ASME Code Component(s) Affected P-209, High Pressure Coolant Injection (HPCI) Pump (Class 2) (Group B)

Component/System Function The HPCI System is designed to pump water into the reactor vessel under loss- of-coolant conditions which do not result in rapid depressurization of the pressure vessel. The loss-of-coolant might be due to loss of reactor feedwater or to small line breaks which do not cause depressurization of the reactor vessel [USAR Updated Safety Analysis Report 6.2.4].

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

3. Applicable Code Requirement(s)

ISTB-5122, Group B Test Procedure (Centrifugal Pumps)

ISTB-5122(a), states in part, the pump shall be operated at a speed adjusted to the reference point (+/- 1 percent).

ISTB-5122(c), specifies system resistance may be varied as necessary to achieve the reference point.

ISTB-5123, Comprehensive Test Procedure (Centrifugal Pumps)

ISTB-5123(a), states in part, The pump shall be operated at a speed adjusted to the reference point (+/- 1 percent).

ISTB-5123(b), states in part, the resistance of the system shall be varied until the flow rate equals the reference point. The differential pressure shall then be determined and compared to its reference value.

4. Reason for Request In order to perform accurate trending and data analysis, the use of an accurate reference value is very important. The complexities of the flow control system used for HPCI makes it difficult to exactly duplicate the Page 18 of 27

Monticello Nuclear Generating Plant Revision 0 reference points. Additionally, iterative manipulation of the control system equipment to refine the hydraulic and speed parameters contributes additional wear to system components. As alternative testing is allowed under the provisions of 10 CFR 50.55a, MNGP Northern States Power -

Minnesota (NSPM) proposes an alternative test method for the Comprehensive and Group B HPCI pump test, as required by ASME OM Code, Subsection ISTB.

5. Proposed Alternative and Basis for Use As stated in NUREG-1482, Rev 1, Section 5.2, some system designs do not allow for testing at a single reference point or a set of reference points. In such cases, it may be necessary to plot pump curves to use as the basis for variable reference points. Code Case OMN-16, Use of Pump Curves for Testing, is included in the issuance of ASME Code OMb-2006. This Code Case has not been accepted by NRC staff for inclusion in NRC Regulatory Guide (RG) 1.192; Operation and Maintenance Code Case Acceptability, ASME OM Code; however, Code Case OMN-9 Use of Pump Curves for Testing has been conditionally accepted by NRC staff for inclusion in RG 1.192. The conditions imposed on OMN-9 as stated in RG 1.192 have been incorporated into OMN-16. In addition, Code applicability for the use of OMN-16 includes OM Code-2004 with Addenda through OMb-2006 which is the fifth interval Code of Record for the MNGP IST Program.

As an alternative to the testing requirements of ISTB-5122 and ISTB-5123, NSPM will assess pump performance and operational readiness through the use of reference pump curves per the guidelines provided in Code Case OMN-16. Flow rate and pump differential pressure will be measured during inservice testing in the as found condition of the system and compared to an established reference curve. The following elements will be used in the development of the reference pump curves:

During Comprehensive and Quarterly HPCI Pump Testing, pump differential pressure and flow rate will be evaluated using a reference point derived from a pump curve. Figure 1 and Figure 2 provide the representative graph which NSPM proposes to use for Quarterly and Comprehensive Testing, respectively. The reference point test pump curve will be restricted to an operating range that is representative of accident conditions, or conservative conditions that are the most sensitive indicator of pump degradation.

Appropriate upper and lower acceptance criteria limits for differential pressure will be established for the Required Action and Alert range limits, as applicable, for Group B (Quarterly) and Comprehensive testing.

These limits will be scalar multiples of the reference pump curve. For determination of whether the In-service Testing (IST) Acceptance Criteria is met, Table 1 and Table 2 are proposed to be used to analyze the data.

These acceptance criteria satisfy the requirements specified in Code Case Page 19 of 27

Monticello Nuclear Generating Plant Revision 0 OMN-16, paragraph 16-6200(a) Alert Range and paragraph 16-6200(b)

Required Action Range.

NSPM will follow the stipulations established by Pump Relief Request PR-03, HPCI Pump Vibration for the Vibration Alert Levels and Code established limits for the Action Required Levels over the reference value curve range for Comprehensive testing.

The vibration data (see Figure 3 through Figure 6) from the test was reviewed and no adverse correlation was evident between flow rate and vibration at the nominal reference point speed. Therefore, NSPM will not establish new vibration reference values and related allowable limits over the reference value curve at this time.

If future requirements necessitate the need for re-generation of a new pump reference curve, NSPM will obtain vibration readings across the expected operating test range of the pump.

The alternative testing described above provides an acceptable level of quality and safety because the method will provide increased accuracy in trending and data analysis. Since the methodology utilized is consistent with the NRC staff guidance provided in NUREG-1482, Rev.1, Section 5.2 and Code Case OMN-16 it will provide reasonable assurance of pump operational readiness.

6. Duration of Proposed Alternative The proposed alternative identified in this relief request shall be implemented during the Fifth Ten Year IST Interval beginning September 1, 2012.

7. Precedents NRC Safety Evaluation for Monticello Nuclear Generating Plant, Evaluation of Relief Request PR-06 Relating to the Fourth 10-Year Interval Inservice Testing Program, Docket No. 50-263. (TAC No. MB9550), August 7, 2003.

Page 20 of 27

Monticello Nuclear Generating Plant Revision 0 10 CFR 50.55a REQUEST NUMBER - PR 06 HPCI Pump Testing Utilizing Pump Curves Attachments:

Table 1 - HPCI Pump Group B Quarterly Test Acceptance Criteria Table 2 - HPCI Pump Comprehensive Test Acceptance Criteria Figure 1 - HPCI P-209 Group B Quarterly Testing Limits Figure 2 - HPCI P-209 Comprehensive Testing Limits Figures 3 through Figure 6 - HPCI Vibration Comparison over Reference Curve Flow Range Page 21 of 27

@--w Monticello Nuclear Generating Plant Revision 0 10 CFR 50.55a REQUEST NUMBER - PR 06 HPCl Pump Testing Utilizing Pump Curves HPC Pump Group B Quarterly Test Acceptance Criteria (For Information Only, actual data may change during 5" IST Interval)

Table 1 Pump Diff Press Reference Point Reference Value Lower Required Finure 1 HPCl P-209 Group 8 QuartarlvTeslina Limils (kt W o ~ U o a O ~ , k N ~ I R d e t e n u C vm vd t Unk NY bn~rulrad) 3050 3100 Flow Rate [gpm)

Page 22 of 27

I

-- Monticello Nuclear Generating Plant Revision 0 10 CFR 50.55a REQUEST NUMBER - PR 06 HPCl Pump Testing Utilizing Pump Curves HPC Pump Comprehensive Test Acceptance Criteria (For Information Only, actual data may change during 5~ IST Interval)

Table 2 ComprehensiveTest Pump Flow Pump Diff Press Reference Point Reference Value Upper Required Lower Alert Lower Required (gpm) (psidl Action Range Range Action Range

( > psid) ( < psid) ( < psid) 2900 1061.4 1093.2 . 987.2 955.3 2950 1057.0 1088.7 983.1 951.3 3000 1051.9 1083.4 978.3 946.8 3050 1046.0 1077.3 972.8 941.4 3100 1039.7 1070.8 967.0 935.8 3150 1032.9 1063.8 960.6 929.7 3200 1025.9 1056.6 954.1 923.4 3250 1018.8 1049.3 947.5 917.0 3300 1011.6 1041.9 940.8 910.5 Finure 2 BPCI P-2I-m p (For Inf ormtlon Only, Actuat Reference Curve and Llmits m y be ravlred)

Flow R a t e (gpm)

I-krnp Reference Curve --- .Upper Action Lirrit --- *Lower Alert Lirrit --- Lower Action Lin-it I 10 CFR 50.55a REQUEST NUMBER - PR 06 Page 23 of 27

Monticello Nuclear Generating Plant

,{

Revision 0 HPCl Pump Testing Utilizing Pump Curves Fiaure 3 HPCl Vibration Corn parision over Reference Curve flow Range (2.12.03 Test) 0 2700 2800 2900 3000 3100 3200 3300 3400 3500 Flow Rate (gpmf Turbine Main Pump Gear Booster Page 24 of 27

e' y 0 d -W Monticello Nuclear Generating Plant Revision 0 10 CFR 50.55a REQUEST NUMBER PR 06 -

HPCl Pump Testing Utilizing Pump Curves Fiaure 4 HPCI Vibration Cornparision over Reference Curve Flow Range (2.12.03 Test)

Flow Rate (gpm)

Turbine Main Pump Gear Booster Page 25 of 27

Monticello Nuclear Generating Plant Revision 0 10 CFR 50.55a REQUEST NUMBER - PR 06 HPCl Pump Testing Utilizing Pump Curves Finure 5 HPCl Vibration Cornparisionover Reference Curve Flow Range (2.12.03 Test)

Flow Rate (gpm) 03-A 03-v ~3 -H Turbine Main Pump Gear Booster Page 26 of 27

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Monticello Nuclear Generating Plani Revision 0 10 CFR 50.55a REQUEST NUMBER - PR 06 HPCl Pump Testing Utilizing Pump Curves HPCl Vibration Cornparision over Reference Curve Flow Range (2.12.03 Test) 0.5 1 1 0 !

2700 2800 2900 3000 3100 3200 3300 3400 3500 Flow Rate (gpm) 1 2 3

-4 Turbine Main Pump Gear Booster Page 27 of 27

Monticello Nuclear Generating Plant Revision 0 ATTACHMENT 5 Valve Relief Requests (VR)A,B A

IST Relief Request formats were developed following the guidelines provided in NUREG-1482, Rev.1, Attachment 1, NEI White Paper, Entitled Standard Format for Requests from Commercial Reactor Licensees Pursuant to 10 CFR 50.55a, Revision 1 dated June 2004.

B Reference Attachment 10 for NRC Safety Evaluations for approved Relief Requests (Pending)

Page 1 of 16

Monticello Nuclear Generating Plant Revision 0 10 CFR 50.55a REQUEST NUMBER - VR 01 Closure Testing of Scram Discharge Header Check Valves CRD-114 (Typical of 121 Valves)

Proposed Alternative In Accordance with 10CFR50.55a(a)(3)(i)

On the basis that the proposed alternative provides an acceptable level of quality and safety

1. ASME Code Component(s) Affected CRD-114, Scram Discharge Header Checks Valves (Typical 121 valves, one per Hydraulic Control Unit (HCU) (Class 2)

Component/System Function These check valves are required to open during a reactor scram, providing a flow path for the exhaust water from the Control Rod Drives (CRDs) to the Scram Discharge Volume. The check valves have no safety function in the closed direction.

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

3. Applicable Code Requirement(s)

ISTC-3510, Exercise Test Frequency, requires active Category C check valves to be exercised nominally every 3 months. If exercising every 3 months is not possible then exercising shall be performed during cold shutdowns or refueling outages as permitted by ISTC-3522, Category C Check Valves.

ISTC-3522, Category C Check Valves, states in part, that each check valve exercise test shall include open and close tests.

ISTC-5221(a)(2), Valve Obturator Movement, requires 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 either the full open position or to the position required to perform its intended safety function(s), and verify closure.

4. Reason for Request The subject check valves, CRD-114 (scram discharge header check valves), are simple ball-check design. There are no internal parts in the check valves that are susceptible to rapid degradation and sudden failure.

Page 2 of 16

Monticello Nuclear Generating Plant Revision 0 In addition, the control rods are infrequently scrammed and these valves are thus subjected to few stress/wear cycles. It is not practical to perform the close exercise test on-line.

Furthermore, the valves are welded into the line and it is not practicable to perform a disassembly and inspection of each valve in accordance with ISTC- 5221(c). There is no provision for routine access for direct visual examination of the ball and body seats or for indirect examination of internals using remote viewing aides such as a boroscope. In order to observe that the obturator has traveled would require a complete disassembly and inspection of the check valve, and an additional valve would require disassembly.

NUREG-1482, Rev.1, Section 4.4.6 Testing Individual Scram Valves for Control Rods in Boiling-Water Reactors provides an approved alternative.

NUREG-1482 requires that those ASME Code Class valves that must change position to provide the scram function should be included in the IST Program and be tested in accordance with the requirements of ISTC except where relief has been granted in a previously issued Safety Evaluation. Bi-directional exercising testing of check valves is required by the 1996 Addenda to the ASME Code (and later editions and addenda).

NUREG-1482, Rev.1, Section 4.4.6 further states in part:

The control rod drive system valves that perform an active safety function in scramming the reactor are the scram discharge volume vent and drain valves, the scram inlet and outlet valves, the scram discharge header check valves, the charging water header check valves, and the cooling water header check valves. With the exception of the scram discharge volume vent and drain valves, exercising the other valves quarterly during power operations could result in the rapid insertion of one or more control rods.

For those control rod drive system valves where testing could result in the rapid insertion of one or more control rods, the rod scram test frequency identified in the facility TS may be used as the valve testing frequency to minimize rapid reactivity transients and wear of the control rod drive mechanisms. The alternate test frequency should be clearly stated and documented in the IST Program.

The proper operation of these check valves is demonstrated during scram time testing. During scram time testing, scram insertion time is measured for each CRD. Monticello Nuclear Generating Plants (MNGPs) Technical Specification (TS) 3.1.4 provides a specific time for individual CRD scram insertion. If a particular CRDs scram insertion time is less than the specified time, the above check valves are functioning properly.

Page 3 of 16

Monticello Nuclear Generating Plant Revision 0 MNGPs TS surveillance requirement (SR) 3.1.4.1 requires verification that each control rod scram time is within the limits of TS Table 3.1.4-1 with reactor steam dome pressure 800 psig prior to exceeding 40% Rated Thermal Power (RTP) after each reactor shutdown 120 days.

MNGP TS SR 3.1.4.2 requires verification, for a representative sample, that each tested control rod scram time is within the limits of TS Table 3.1.4-1 with reactor steam dome pressure 800 psig each 200 days cumulative operation in Mode 1.

The Scram Discharge Header Check Valves have a safety function to open. This valve (CRD-114) must open to provide a flow path from the overpiston area of the control rod drive to the scram discharge header during a scram. This check valves closed function is to prevent backflow from the scram discharge volume (SDV) to the overpiston area of the drive when a Scram is reset. Flow from the CRD to the SDV occurs throughout the entire scram stroke of the control rod and continues until volume pressure equals reactor vessel pressure. There would normally be no demand for check valve closure until after the rod is fully inserted and latched. Additionally, any condition that would require check valve closure would prevent further control rod insertion regardless of the position of this check valve. Therefore, failure of the scram outlet check valves to close would not prevent the system from performing its safety function.

5. Proposed Alternative and Basis for Use Northern States Power - Minnesota (NSPM) considers that the proper operation of each of these check valves is demonstrated during scram time testing where each drive scram insertion time is measured. As previously discussed, MNGP TS 3.1.4 provides a specific time for CRD scram insertion. If a particular scram insertion time is less than the specified time scram, then the related valves are functioning properly.

The successful scram time of a CRD also represents the successful full stroke exercising of these check valves (CRD-114). NSPM proposes to perform testing of the CRD scram discharge header check valves consistent with the alternative testing provided in NUREG-1482, Rev.1, Section 4.4.6. Therefore, the closed function of the scram discharge header check valves would not be tested as required by ISTC-3522.

Using the provisions of this relief request as an alternative to the test requirements specified in ISTC-3510, ISTC-3522 and ISTC-5221(a)(2) is an acceptable alternative method of detecting degradation of the check valves and provides an acceptable level of quality and safety for determining the check valves are functioning properly.

Page 4 of 16

Monticello Nuclear Generating Plant Revision 0

6. Duration of Proposed Alternative The proposed alternative identified in this relief request shall be implemented during the Fifth Ten Year IST Interval beginning September 1, 2012.

7. Precedents NRC Safety Evaluation for Monticello Nuclear Generating Plant, Relief Request No. VR-01 Relating to the Fourth 10-Year Interval Inservice Testing Program, Docket No. 50-263. (TAC No. MB9736), December 10, 2003 Page 5 of 16

Monticello Nuclear Generating Plant Revision 0 10 CFR 50.55a REQUEST NUMBER VR 02 RHRSW Flow Control Valve Exercising Proposed Alternative In Accordance with 10CFR50.55a(f)(5)(iii) for the NRC to Grant Relief and Impose an Alternative Requirement in Accordance with 10CFR50.55a(f)(6)(i)

On the basis that compliance with the ASME OM Code requirements are impractical for MNGP

1. ASME Code Component(s) Affected CV-1728, RHR Service Water to RHR Heat Exchanger Flow Control Valve (Class 3)

CV-1729, RHR Service Water to RHR Heat Exchanger Flow Control Valve (Class 3)

Component/System Function These valves must open to the throttled position to provide a flow path from the Residual Heat Removal (RHR) Service Water (RHRSW) system to the RHR heat exchanger tube side during normal shutdown cooling or containment spray/cooling mode.

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

3. Applicable Code Requirement(s)

ISTC-3510, Exercise Test Frequency, requires active Category B to be exercised nominally every 3 months. If exercising every 3 months is not possible then exercising shall be performed during cold shutdowns or refueling outages as permitted by ISTC-3520.

ISTC-5130, Pneumatically Operated Valves ISTC-5131, Valve Stroke Testing, requires active valves to have their stroke times measured when exercised in accordance with ISTC-3500 and limiting values to be specified by the Owner.

ISTC-5132, Stroke Time Acceptance Criteria, requires test results to be compared to the reference values established in accordance with ISTC-3320.

Page 6 of 16

Monticello Nuclear Generating Plant Revision 0 ISTC-5133, Stroke Test Corrective Action, requires valves that exceed their limiting values of full stroke test to be immediately declared inoperable (ISTC-5133(a). In addition, ISTC-5133(b) requires valves with measured stroke times that do not meet the acceptance criteria of ISTC-5132 shall be immediately retested or declared inoperable.

4. Impracticality of Compliance ISTC-5131 requires that a limiting value of full stroke time be established for a power operated valve and that the stroke time be measured whenever such a valve is full stroke tested. Performing full stroke time testing or full stroke exercising of these valves is impractical based on the control scheme design of the valves, adverse plant impact, and the functional requirements of the valves.

ISTC-2000 defines the full-stroke time as the time interval from initiation of the actuating signal to the indication of the end of the operating stroke.

The control scheme design of these valves does not receive an actuation signal (neither by manual hand switch nor by automatic logic) to stroke to the position required to fulfill their safety function.

RHRSW valves CV-1728 and CV-1729 are air operated control valves on the outlet line of the RHRSW side of the A and B RHR heat exchangers, respectively. These control valves maintain a differential pressure between the RHRSW process stream and the RHR process stream during RHRSW system operation. The valves are controlled by a positioner, which is controlled by a differential pressure-indicating controller (DPIC). The DPIC senses pressure on the RHRSW discharge line and the RHR inlet line to the RHR heat exchanger. The desired differential pressure control point, and thus the desired valve position for system flow, is manually set by the operator. The valve positioner modulates the valve position as necessary to maintain this control point.

Stroke time testing or full stroke exercising of these valves on quarterly basis is not consistent with the design of the valves control scheme and is not in the interest of plant safety.

These valves are interlocked to receive a closed signal when the RHRSW pumps are de-energized. This interlock is provided to ensure that system water inventory is not lost during system shutdown. Stroke time testing or full stroke exercising of valves CV-1728 and CV-1729 when the RHRSW pumps are de-energized would result in the loss of liquid fill for a significant portion of the RHRSW system as well as require the bypassing of an interlock designed to minimize the potential for water hammer. Such testing increases the possibility of an adverse water hammer during Page 7 of 16

Monticello Nuclear Generating Plant Revision 0 startup of the RHRSW system as well as requires filling and venting of the system following the stroke time testing. In addition to the adverse impact on plant operation, such testing results in causing system or component damage.

Stroke time testing or full stroke exercising of the valves during RHRSW pump operation negates the loss of system fill concern; however, this testing would also have an adverse impact on plant safety and equipment integrity. Stroke time testing or full stroke exercising during pump operation would require the valves to be initially in the closed position during pump operation. Establishing the initial test conditions of a closed valve during pump operation would result in an undesirable deadheading of the pump. Subsequent opening of the valves to perform stroke time testing or full stroke exercising will result in pump runout if a single RHRSW pump is in operation, an undesirable condition which adversely impacts pump integrity and performance. The pump runout concern can be addressed by stroke timing the valve open during operation of both RHRSW pumps; however, this exacerbates the pump deadheading concerns and would result in undesirable transients on the system and could cause system or component damage.

5. Burden Caused By Compliance:

Proper stroke time testing or full stroke exercising would require the plant to modify the control logic of the valves. The activity associated with performing this modification is not offset by an increase in public safety.

The proposed alternative testing is an effective means to ensure the valves perform their safety function and is consistent with other valve category test requirements, such as check valve exercising. By extension, if stroke time testing is not performed, the requirement of ISTC-5132 for establishing stroke time acceptance criteria is impractical. Similarly, if there are no stroke time limits applicable, then the requirement of ISTC-5133 for corrective action when stroke time limits are exceeded is not applicable when performing Code required testing.

ISTC-3530 provides for demonstrating the necessary valve disk movement by observing indirect evidence (such as changes in system pressure, flow rate, level, or temperature), which reflect stem or disk position. The most representative test of the capability of valves CV-1728 and CV-1729 to perform their intended function is performed during In-Service testing of the RHRSW pumps. Quarterly testing of the RHRSW pumps verifies the capability of the valves to operate properly to pass the maximum required accident flow as well as the valve position necessary to achieve required flow conditions. In addition, Per ISTC-5131(d) any abnormality or erratic action shall be recorded and an evaluation shall be made regarding need for corrective action. Testing of the valves in this Page 8 of 16

Monticello Nuclear Generating Plant Revision 0 manner demonstrates valve performance capability and provides a means to monitor for valve degradation. It should be noted that these valves are within the scope of the Monticello Nuclear Generating Plant (MNGP) Air Operated Valve (AOV) Program. As such, the valves receive diagnostic testing per the AOV Program.

Using the provisions of this relief request as an alternative to the requirements of ISTC-3510 and 5130 provides a reasonable alternative to the Code requirements. Based on the determination that the proposed alternative provides reasonable assurance of the valves operational readiness and is an acceptable alternative method of detecting degradation, Northern States Power - Minnesota (NSPM) requests that relief be granted pursuant to 10CFR50.55a(f)(6)(i).

7. Duration of Proposed Alternative The proposed alternative identified in this relief request shall be implemented during the Fifth Ten Year IST Interval beginning September 1, 2012.

8. Precedents NRC Safety Evaluation for Monticello Nuclear Generating Plant, Relief Requests Relating to the Fourth 10-Year Interval Inservice Testing Program, Docket No. 50-263. (TAC No. MB6807), July 17, 2003 Page 9 of 16

Monticello Nuclear Generating Plant Revision 0 10 CFR 50.55a REQUEST NUMBER - VR 03 Use of Code Case OMN-1, Revision 1, on Various Motor Operated Valves Proposed Alternative In Accordance with 10CFR50.55a(a)(3)(i)

On the basis that the proposed alternative provides an acceptable level of quality and safety

1. ASME Code Component(s) Affected Certain motor-operated valve assemblies currently included in the Monticello Nuclear Generating Plant (MNGP) Motor-Operated Valve (MOV) Program.

Component/System Function The applicable motor operated valves are required to perform a specific function in shutting down the reactor to a safe shutdown condition, maintaining a safe shutdown condition or in mitigating the consequences of an accident.

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

3. Applicable Code Requirement(s)

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

ISTC-5120, Motor-Operated Valves, ISTC-5121(a) states, Active valves shall have their stroke times measured when exercised in accordance with ISTC-3500.

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

4. Reason for Request NUREG-1482, Rev.1, Section 4.2.5 states in part: As an alternative to MOV stroke-time testing, ASME developed Code Case OMN-1 Rev 0, Alternative Rules for Preservice and Inservice Testing of Certain Electric Motor-Operated Valve Assemblies in LWR Power Plants which provides periodic exercising and diagnostic testing for use in assessing the operational readiness of MOVs. Section 4.2.5 recommends that the Page 10 of 16

Monticello Nuclear Generating Plant Revision 0 licensees implement ASME Code Case OMN-1 Rev 0 as accepted by the NRC (with certain conditions) in the regulations or Regulatory Guide (RG) 1.192, as an alternative to the stroke-time testing provisions in the ASME Code for MOVs. RG 1.192 allows licensees to implement ASME Code Case OMN-1, Revision 0, (in accordance with the provisions in the RG) as an alternative to the Code provisions for MOV stroke-time testing in the ASME OM Code 1995 Edition through 2000 Addenda.

The Code of Record for MNGP Fifth 10-Year IST Interval is OM Code-2004 Edition with Addenda through OMb Code-2006.

5. Proposed Alternative and Basis for Use Pursuant to the guidelines provided in NUREG-1482, Rev.1, Section 4.2.5, and the conditions stated in RG 1.192, NSPM proposes to implement Code Case OMN-1, Revision 1, in lieu of the stroke-time provisions specified in ISTC-5120 for MOVs as well as the position verification testing in ISTC-3700.

There are no significant differences between the version of Code Case OMN-1 that is in the 1999 Addenda of the OM Code currently approved for use in RG 1.192, Revision 0 and the Revision 1 of the Code Case (OMN-1-1) in 2009 Edition of the OM Code.

The use of Code Case OMN-1-1 by a licensee permits licensees to replace stroke time and position verification testing of MOVs with a program of exercising MOVs every refueling outage and diagnostically testing on longer intervals.

The proposed alternative is considered to be acceptable because Code Case OMN-1-1 provides a superior method than the stroke-timing method required by the OM code for assessing the operational readiness of MOVs.

Using the provisions of this relief request as an alternative to the MOV stroke-time testing requirements of ISTC-5120 and position indication verification of ISTC-3700 provide an acceptable level of quality for determination of valve operational readiness. Code Case OMN-1-1 should be considered acceptable for use with ASME OM Code-2004 Edition with Addenda through OMb Code-2006 as the Code of Record.

6. Duration of Proposed Alternative Page 11 of 16

Monticello Nuclear Generating Plant Revision 0 The proposed alternative identified in this relief request shall be implemented during the Fifth Ten Year IST Interval beginning September 1, 2012.

7. Precedents

1. NRC Safety Evaluation for Beaver Valley Power Station, Unit 2, Docket No. 50-412. Regarding the Third 10-Year Interval Inservice Testing Program Relief Requests (TAC Nos. MD5595 - MD5604),

February 14, 2008. (ADAMS Accession No. ML080140299)

2. NRC Safety Evaluation for South Texas Project, Unit 1 and 2, Docket No. 50-498 and 50-499. Regarding the Third 10-Year Inservice Testing Program Interval (TAC Nos. ME3515, ME3516, ME3517, ME3518, ME3520, ME3521 and ME3522), September 2, 2010. (ADAMS Accession No. ML102150077)

Page 12 of 16

Monticello Nuclear Generating Plant Revision 0 VALVE RELIEF REQUEST NUMBER - VR 04 Use of Code Case OMN-17, Revision 0, on the Class 1 Main Steam Safety Relief Valves Proposed Alternative In Accordance with 10CFR50.55a(a)(3)(i)

On the basis that the proposed alternative provides an acceptable level of quality and safety.

1. ASME Code Component(s) Affected RV-2-71A, Main Steam Safety/Relief Valve (ADS) (Class 1)

RV-2-71B, Main Steam Safety/Relief Valve (Class 1)

RV-2-71C, Main Steam Safety/Relief Valve (ADS) (Class 1)

RV-2-71D, Main Steam Safety/Relief Valve (ADS) (Class 1)

RV-2-71E, Main Steam Safety/Relief Valve (Low-Low Set) (Class 1)

RV-2-71F, Main Steam Safety/Relief Valve (Class 1)

RV-2-71G, Main Steam Safety/Relief Valve (Low-Low Set) (Class 1)

RV-2-71H, Main Steam Safety/Relief Valve (Low-Low Set) (Class 1)

Component/System Function The Nuclear Boiler System provides Reactor Pressure Vessel (RPV) overpressure protection by opening the Safety/Relief Valves (S/RVs). The valves must open in order to prevent over pressurization of the reactor coolant system thereby preventing failure of the reactor system due to overpressure. The overpressure relief operation is self-actuated. The valves will open automatically or manually by the air operator during depressurization operation.

Certain valve are a designated part of the Automatic Depressurization System (ADS) and must open to provide automatic reactor depressurization as a result of a small break in the nuclear system coincidental with a failure of the High Pressure Coolant Injection (HPCI)

System. Rapid depressurization is necessary so that Low Pressure Coolant Injection (LPCI) and the Core Spray systems can operate to protect the fuel cladding. ADS is automatically actuated after receipt of simultaneous Residual Heat Removal (RHR) or Core Spray pump running and low-low reactor water level signals.

Page 13 of 16

Monticello Nuclear Generating Plant Revision 0 Component/System Function (Cont.)

In addition to the above, certain valves are designated as part of the S/RV Low-Low Set System and are set to open automatically at a set-point lower than the mechanical self-actuated set point to prevent the re-opening of a non-low-low set S/RV following a reactor isolation transient.

The set-points of the low-low set S/RVs ensures that they will be the first S/RVs to open and the last to close. After opening and closing of a low-low set SRV, a time delay relay prevents the operator or the low-low set logic from immediately re-opening the S/RV to allow the water leg in the S/RV discharge line to recede. The valves will open as part of the SRV Low-Low Set System in the event of a reactor SCRAM with reactor pressure greater than the low-low setpoint and the SRV low-low set hand switch in the auto position.

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

3. Applicable Code Requirement(s)

Appendix I, Paragraph I-1320(a), 5-Year Test Interval, specifies that Class 1 pressure relief valves shall be tested at least once every 5 years, starting with initial electric power generation. No maximum limit is specified for the number of valves to be tested within each interval; a minimum of 20% of the valves from each valve group shall be tested within any 24-month interval. This 20% shall consist of valves that have not been tested during the current 5-year interval, if they exist. The test interval for any individual valve shall not exceed 5 years.

4. Reason for Request MNGP transitioned from an 18-month fuel cycle to a 24-month fuel cycle on September 30, 2005 via Licensing Amendment No. 143. Prior to transitioning to the 24-month fuel cycle, ASME Code requirements could be satisfied by removing and testing approximately one-third of the 8 S/RVs each refueling outage in order to comply with the 5-year test interval requirements for Class 1 pressure relief valves imposed by the Code of record during that time. Since transitioning to the 24-month fuel cycle MNGP must remove at least one-half of the subject relief valves each refueling outage for off-site testing. The removal of approximately half of the 8 valves versus approximately a third of the valves each outage requires the removal of additional insulation, instrumentation, and other interferences. This additional work results in an undesirable increase in radiation exposure to maintenance personnel.

Page 14 of 16

Monticello Nuclear Generating Plant Revision 0

4. Reason for Request (Cont.)

The ASME Code committees have recently developed Code Case OMN-17, Alternative Rules for Testing ASME Class 1 Pressure Relief/Safety Valves which was published via ASME OM Code-2009 Edition. This Code Case has not been approved for use in US NRC Regulatory Guide 1.192, Operation and Maintenance Code Case Acceptability, ASME OM Code, dated June 2003. The Code Case allows the Owner to extend the test frequencies for Class 1 pressure relief valves to a 72-month (6-year) test interval providing all the requirements of the Code Case are satisfied.

The Code applicability specified in the Code Case is, in part, ASME OM Code 2001 Edition through the 2006 Addenda of Appendix I, Section I-1320. This is consistent with the 5th Interval Code of record for MNGP.

MNGP currently meets or exceeds all the requirements specified in Code Case OMN-17.

5. Proposed Alternative and Basis for Use As an alternative to the Code required 5-year test interval per Appendix I, paragraph I-1320(a), MNGP proposes that the subject Class 1 pressure relief 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.

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

MNGP submits that the proposed alternative of increasing the test interval for the subject Class 1 pressure relief valves from 5 years to 3 fuel cycles (approximately 6 years/72 months) would continue to provide an acceptable level of quality and safety while restoring 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 10CFR50.55a(a)(3)(i).

Page 15 of 16

Monticello Nuclear Generating Plant Revision 0

6. Duration of Proposed Alternative The proposed alternative identified in this relief request shall be implemented during the Fifth Ten Year IST Interval beginning September 1, 2012.

7. Precedents Perry Nuclear Power Plant, Docket No. 50-440, SER Date October 22, 2009, Safety Evaluation of Relief Requests for Third 10-Year Pump and Valve Inservice Testing Program (TAC Nos. ME0191 through ME0198)

(re: VR-6)

Nine Mile Point Nuclear Station, Docket No. 50-410, SER Date April 17, 2001, Safety Evaluation of the Alternative to ASME Code Regarding Inservice Testing of Main Steam Safety/Relief Valves, (TAC No. MB0290).

8. References Code Case OMN-17, Alternative Rules for Testing ASME Class 1 Pressure Relief/Safety Valves Page 16 of 16

Monticello Nuclear Generating Plant Revision 0 ATTACHMENT 6 Deferred Testing Justifications (DTJ)

Page 1 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 01 System: Reactor Building Closed Cooling Water (RBC)

Valve(s): RBCC-15 P & ID: NH-36042-2 (M-111-1)

Code Class: 2 Category: A/C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: Inboard (located outside of containment) Containment Isolation Valve for RBCCW supply to the drywell must open to provide cooling flow to a number of Primary Containment (PC) heat loads, including PC Ventilation, as well as Recirculation Pump seal and motor coolers.

Should the RBCCW line inside the drywell break during a LOCA, this valve will close to perform its Primary Containment isolation function.

Frequency Change Justification: It is not practical to exercise close the RBCCW supply check valve every 3 months per the requirements of ISTC-3510. This check valve is the inboard (although located outside containment) primary containment isolation valve for a single train non-code support system required to be in service during plant operation. The normally open check valve requires an exercise in the reverse flow direction, which can be verified by some form of leak testing. Interrupting RBCCW flow to within Primary Containment the heat loads there (including Recirculation Pump seal and motor coolers and Drywell Coolers) long enough to perform such a test with the plant on-line would likely result in equipment damage and plant shutdown. Similarly, RBCCW flow through this valve is needed most of the time during cold shutdown for operation of the Recirculation Pumps and for Drywell Cooling to maintain tolerable ambient temperatures for personnel in the Drywell.

Alternate Test Frequency: Test exercise valve closed at a refueling outage frequency in accordance with ISTC-3522(c). As an alternative to reverse exercising each refueling outage, this Category A/C check valve may be placed in the Check Valve Condition Monitoring (CVCM) Program and reverse exercised at the Appendix J, Option B frequency during the performance of Type C seat leakage testing. This extension of frequency for reverse exercising is per the guidelines provided in NUREG-1482, Revision 1, Section 4.4.7, Use of Appendix J, Option B, in Conjunction with ISTC Exercising Tests.

Page 2 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 02 System: Condensate Storage (CST)

Valve(s): CST-88, CST-90, CST-92, CST-94 P & ID: NH-85509 (M-114-1)

Code Class: 2 Category: C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: The Condensate keep fill valves are the boundary valves between the safety-related RHR pumps discharge or suction piping and the non-safety related service condensate keep fill system. These valves supply keep-fill water pressure to various locations in the B Loop of RHR and the Shutdown Cooling (SDC) suction line. Their safety function is to close to ensure diversion of ECCS injection flow cannot occur during a ECCS System initiation (CST- 88, CST-92, CST-94) when reactor vessel pressure falls below pump discharge pressure once the respective injection motor operated injection valves open. CST-90 closes while the plant is in SDC to ensure there is no loss of vessel inventory while in SDC mode.

Frequency Change Justification: It is not practical to exercise close the CST boundary check valves every 3 months per the requirements of ISTC-3510. These valves have a closed safety position since they prevent diversion of RHR flow into the service condensate system. There are no test taps or instrumentation installed that would allow testing that proves by positive means that the disc moves to the seat on cessation or reversal of flow. Installation of test taps and isolation valves to reverse flow test these valves is not practical. Similarly, there is no practical way to measure flow in these lines.

Alternate Test Frequency: These valves will be grouped and tested in accordance with ISTC-5221(c). At least one valve from the group will be disassembled and examined each refueling outage for the open and close exercise test. As an alternative to the disassembly and examination requirements of ISTC-5221(c), these check valves may be placed in the Check Valve Condition Monitoring (CVCM) Program and disassembled and examined per the requirements of Appendix II of the OM Code. If so, the frequency for disassembly and examination shall be specified in the CVCM Program Plan applicable to the valve Group.

Page 3 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 03 System: Reactor Recirculation (REC)

Valve(s): XR-27-1, XR-27-2, XR-25-1, XR-25-2 P & ID: NH-36243-1 (M-117), NH-36244 (M-118)

Code Class: 2 Category: A/C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: These check valves open to provide lubrication and cooling water flow to the Recirculation Pump seals. During closure, they prevent reversal of reactor water flow from the recirculation seals to the CRD System and fulfill their primary containment isolation function.

Frequency Change Justification: It is not practical to exercise the recirculation pump seal check valves closed every 3 months per the requirements of ISTC-3510. These valves are the inboard and outboard Primary Containment isolation valves for the CRD injection flow to the Recirc Pump seals.

To exercise these normally open check valves to the closed position requires temporarily stopping seal injection flow for some sort of leak test which could result in damage to the Reactor Coolant Recirculation pump seals and necessitate a plant shutdown. The reactor coolant recirculation pumps are also normally kept operating during cold shutdown.

Alternate Test Frequency: Test exercise the valves closed at a refueling outage frequency in accordance with ISTC-3522(c). As an alternative to reverse exercising each refueling outage, these Category A/C check valves may be placed in the Check Valve Condition Monitoring (CVCM)

Program and reverse exercised at the Appendix J, Option B frequency during the performance of Type C seat leakage testing. This extension of frequency for reverse exercising is per the guidelines provided in NUREG-1482, Revision 1, Section 4.4.7, Use of Appendix J, Option B, in Conjunction with ISTC Exercising Tests.

Page 4 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 04 (DELETED)

Page 5 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 05 System: Control Rod Hydraulics (CRH)

Valve(s): CRD-115 (121 valves total, one per HCU)

P & ID: NH-36245 (M-119)

Code Class: 2 Category: C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months Function/Description: The Hydraulic Control Unit Charging Water Inlet Check Valves open to admit charging water to the CRD scram accumulators. Their safety function is to close to prevent diversion of scram water from the accumulator if the charging water header pressure drops.

Frequency Change Justification: It is not practical to exercise close the hydraulic control unit charging water inlet check valves every 3 months per the requirements of ISTC-3510. These valves are located on each of the 121 hydraulic control units. These valves can be tested to verify closure only by doing a leak test. This test involves stopping the CRD pump and depressurizing the accumulator charging water header, then monitoring for accumulator low-pressure alarms. A CRD pump is normally operated during cold shutdown for CRD cooling water flow and reactor recirculation pump seal injection. Shutdown of the CRD pump system could cause damage to the reactor recirculation pump seals.

Alternate Test Frequency: Test exercise the valves closed at a refueling outage frequency in accordance with ISTC-3522(c).

Page 6 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 06 System: Residual Heat Removal (RHR)

Valve(s): RHR-8-1, RHR-8-2 P & ID: NH-36247 (M-121), NH-36246 (M-120))

Code Class: 2 Category: C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: The RHR minimum flow check valves open to the torus to provide minimum flow recirculation from the RHR pumps. They close to minimize gross leakage flow diversion during containment cooling under certain conditions.

Frequency Change Justification: It is not practical to exercise the RHR Minimum Flow check valves open or closed every 3 months per the requirements of ISTC-3510. There is no means of measuring flow rate directly through these valves during quarterly pump testing, as instrumentation is not installed. Installation of instrumentation to perform this test is impractical and provides no compensating increase in safety. There are no downstream isolation valves and test connections to permit reverse flow on a seat leakage type test to verify closure. Exercising the valves closed would require de-inerting containment, entering the torus and removal of multiple trains of ECCS systems.

Therefore, disassembly and examination shall be used to meet Code exercising requirements.

Alternate Test Frequency: These valves will be grouped and tested in accordance with ISTC-5221(c). At least one valve from the group will be disassembled and examined each refueling outage for the open and close exercise test. As an alternative to the disassembly and examination requirements of ISTC-5221(c), these check valves may be placed in the Check Valve Condition Monitoring (CVCM) Program and disassembled and examined per the requirements of Appendix II of the OM Code. If so, the frequency for disassembly and examination shall be specified in the CVCM Program Plan applicable to the valve Group.

Page 7 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 07 System: Standby Liquid Control (SLC)

Valve(s): XP-6, XP-7 P & ID: NH-36253 (M-127)

Code Class: 1 Category: A/C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: Standby Liquid Control Injection Check Valves open to allow injection of sodium pentaborate to shutdown the reactor. They close to provide a Primary Containment function.

Frequency Change Justification: It is not practical to exercise the SLC injection check valves open or closed every 3 months per the requirements of ISTC-3510. To verify forward flow operability during normal operation would require firing a squib valve and injecting water into the reactor vessel using the SLC pumps. This is impractical due to the extensive maintenance and cost required to replace squib valves. The SLC system would also be inoperable while changing the squib valves.

Verification of closure ability is performed by seat leakage testing. This requires primary containment de-inerting and entry into the drywell to isolate the injection line, which is not possible with the plant on-line and impractical under cold shutdown.

Alternate Test Frequency: Test exercise valves open and closed at a refueling outage frequency in accordance with ISTC-3522(c). As an alternative to reverse exercising each refueling outage, these Category A/C check valves may be placed in the Check Valve Condition Monitoring (CVCM)

Program and reverse exercised at the Appendix J, Option B frequency during the performance of Type C seat leakage testing. This extension of frequency for reverse exercising is per the guidelines provided in NUREG-1482, Revision 1, Section 4.4.7, Use of Appendix J, Option B, in Conjunction with ISTC Exercising Tests.

Page 8 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 08 System: Reactor Water Clean-up (RWC)

Valve(s): RC-6-1, RC-6-2 P & ID: NH-36254 (M-128)

Code Class: 2 Category: A/C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: These check valves open to allow RWCU water return to the reactor vessel.

These valves close to perform their safety function to prevent flow diversion from HPCI/RCIC injection.

Frequency Change Justification: It is not practical to exercise closed the RWCU return check valves every 3 months per the requirements of ISTC-3510. Exercising these valves with the plant on-line would result in substantial personnel radiation dose from entries into the Steam Chase to manipulate manual isolation valves in series with these parallel check valves. In addition testing on-line requires transitioning the system through conditions with water hammer potential as well as accumulating additional thermal transient stress cycles on system components. Closure exercise testing requires some form of a leakage test necessitating an extensive and burdensome isolation of the HPCI/RCIC, Feedwater and CRD line, which is impractical during cold shutdown.

Alternate Test Frequency: Test exercise valves closed and open at a refueling outage frequency in accordance with ISTC-3522(c).

Page 9 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 09 System: Instrument and Service Air (AIR)

Valve(s): AI-626-1 and AI-625 P & ID: NH-36049-14 (M-131)

Code Class: 2 Category: A/C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: The Transverse In-core Probe Indexer Pneumatic purge line check valves close to perform a containment isolation function. These valves open to admit dry nitrogen/air to the TIP indexers for moisture exclusion.

Frequency Change Justification: It is not practical to exercise closed the TIP Purge Line check valves every 3 months per the requirements of ISTC-3510. Check valves AI-626-1 and AI-625 are normally open check valves that are in service during all modes of operation. Closure position verification requires primary containment de-inerting and entry for system isolation and connection of test equipment, which is impractical during cold shutdown.

Alternate Test Frequency: Test exercise valves closed at a refueling outage frequency in accordance with ISTC-3522(c). As an alternative to reverse exercising each refueling outage, these Category A/C check valves may be placed in the Check Valve Condition Monitoring (CVCM)

Program and reverse exercised at the Appendix J, Option B frequency during the performance of Type C seat leakage testing. This extension of frequency for reverse exercising is per the guidelines provided in NUREG-1482, Revision 1, Section 4.4.7, Use of Appendix J, Option B, in Conjunction with ISTC Exercising Tests.

Page 10 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 10 System: Main Condenser (CDR)

Valve(s): AO-1825A, AO-1825B P & ID: NH-36035-2 (M-104-1)

Code Class: NC Category: B Code Frequency: ISTC-3510, Active Category B valves shall be exercised nominally every 3 months.

Function/Description: The mechanical vacuum pump (hogger) suction isolation valves open to allow non-condensable gases to flow to the mechanical vacuum pump suction. These valves close to isolate the mechanical vacuum pump from the condenser.

Frequency Change Justification: It is not practical to exercise closed the Mechanical Vacuum Pump Isolation valves every 3 months per the requirements of ISTC-3510. These valves are not required to be in-service tested by 10CFR50.55a and are considered augmented quality components within the program scope. They are normally closed during power operation and have a closed safety position. To cycle and stroke time the valves, the valve controls require the mechanical vacuum pump be started and stopped. Plant operating procedures prohibit operation of the mechanical vacuum pump above 5% power for equipment and personnel safety reasons; therefore, it is not practical to exercise these valves during power operation.

Alternate Test Frequency: Test exercise valves closed at a cold shutdown frequency in accordance with ISTC-3521(c)

Page 11 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 11 System: Reactor Building Closed Cooling Water (RBC)

Valve(s): MO-1426, MO-4229, MO-4230 P & ID: NH-36042-2 (M-111-1)

Code Class: 2 Category: A Code Frequency: ISTC-3510, Active Category A valves shall be exercised nominally every 3 months.

Function/Description: These are the RBCCW containment isolation valves, which are open to provide cooling flow to a number of Primary Containment (PC) heat loads, including PC Ventilation, as well as Recirculation Pump seal and motor coolers. The control room operators will close these valves remotely if there is indication of an RBCCW line break inside the drywell and primary containment integrity is needed.

Frequency Change Justification: It is not practical to exercise closed RBCCW containment isolation valves every 3 months per the requirements of ISTC-3510. Interrupting RBCCW flow to Primary Containment and the heat loads there (including Recirculation Pump seal and motor coolers and Drywell Coolers) long enough to perform such a test with the plant on-line would likely result in equipment damage and plant shutdown. Similarly, RBCCW flow through this valve is needed most of the time during cold shutdown for operation of the Recirculation Pumps and for Drywell Cooling to maintain tolerable ambient temperatures for personnel in the Drywell.

Alternate Test Frequency: Test exercise the valves closed at a cold shutdown frequency in accordance with ISTC-3521(c). As an alternative to stroke time testing these valves during cold shutdowns, the valves will be stroke timed at the frequency specified in the MOV Program for diagnostic testing. This proposed alternative is pursuant to the guidelines provided in NUREG-1482, Revision 1, Section 4.2.5, Alternatives to Stroke Time Testing and Valve Relief Request VR 03, Use of Code Case OMN-1, Revision 1, on Various Motor Operated Valves. The valves will continue to be exercised (non-timed) during cold shutdowns.

Page 12 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 12 System: Condensate and Feedwater (CFW)

Valve(s): FW-91-1, FW-91-2, FW-94-1, FW-94-2, FW-97-1, FW-97-2 P & ID: NH-36241 (M-115)

Code Class: 1 (FW-94-1, FW-94-2, FW-97-1, FW-97-2) 2 (FW-91-1, FW-91-2)

Category: A/C (FW-91-1, FW-91-2)

A/C (FW-94-1, FW-94-2, FW-97-1, FW-97-2)

Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: Check valves FW-94-1/2 and FW-97-1/2 close for containment isolation and open to allow HPCI/RCIC injection flow to the reactor vessel. These valves are normally open to allow feedwater and RWCU return flow to the reactor vessel.

Check valves FW-91-1/2 are open during power operation as feedwater injection check valves and have a closed safety position to prevent diversion of HPCI or RCIC injection flow.

Frequency Change Justification: It is not practical to exercise closed the Feedwater check valves every 3 months per the requirements of ISTC-3510. These six valves are verified open quarterly as part of normal plant operation. The only method of cycling these valves closed is to perform a reverse flow test. A reverse flow test requires isolating and venting the system and installing temporary bypasses (hoses, gages, etc). This cannot be performed without shutting down the plant and entering the Drywell. Therefore, it is impractical to verify the closed position of these valves during normal power operations. Further, NUREG 1482, Revision 1, paragraph 3.1.1.3 states that de-inerting containment solely for the purpose of performing cold shutdown testing is impractical. If the Drywell is not de-inerted, then the closure test would be performed during refueling.

Alternate Test Frequency: Test exercise the valves closed at a cold shutdown frequency in accordance with ISTC-3522(b).

Page 13 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 13 System: Main Steam (MST)

Valve(s): AO-2-80A through D P & ID: NH-36241 (M-115)

Code Class: 1 Category: A Code Frequency: ISTC-3560, Fail-Safe Valves at a frequency of every 3 months per ISTC-3510.

Function/Description: Main steam line inboard isolation valves have a closed safety position as primary containment isolation valves. They have a safety related pneumatic supply that acts to both open and assist to close them Frequency Change Justification: It is not practical to fail safe test the Inboard Main Steam Isolation Valves (MSIVs) every 3 months per the requirements of ISTC-3560. The valves have a safety related pneumatic supply that acts to both open and assist to close them. The valves have springs that provide the primary closing force needed. Both the safety related pneumatic supply and actuator springs are credited for closing the valves in accident analyses. Testing the fail-safe function of the valve springs requires venting the safety related pneumatic supply locally and monitoring valve stem movement.

This test cannot be done during power operation since the valves are located inside the Drywell, which is not accessible during power operations. NUREG 1482, Revision 1, paragraph 3.1.1.3 states that de-inerting containment solely for the purpose of performing cold shutdown testing is impractical. If the Drywell is not de-inerted, then the closure test would be performed during refueling.

Alternate Test Frequency: Fail-safe test valves closed at a cold shutdown frequency in accordance with per ISTC-3560.

Page 14 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 14 System: Reactor Recirculation (REC)

Valve(s): MO-2-53A, MO-2-53B P & ID: NH-36243 (M-117)

Code Class: 1 Category: B Code Frequency: ISTC-3510, Active Category B valves shall be exercised nominally every 3 months.

Function/Description: The reactor recirculation pump discharge isolation valves are normally open during power operation to allow reactivity control using the recirculation pumps to change core flow. Their safety position is closed to direct LPCI flow into the reactor vessel.

Frequency Change Justification: It is not practical to exercise closed MO-2-53A/B every 3 months per the requirements of ISTC-3510. The valves cannot be cycled during power operation without a large power reduction, associated Plant transient and reactivity changes that require substantial Operator attention and a net decrease in safety.

Alternate Test Frequency: Test exercise the valves closed at a cold shutdown frequency in accordance with ISTC-3521(c). As an alternative to stroke time testing these valves during cold shutdowns, the valves will be stroke timed at the frequency specified in the MOV Program for diagnostic testing. This proposed alternative is pursuant to the guidelines provided in NUREG-1482, Revision 1, Section 4.2.5, Alternatives to Stroke Time Testing and Valve Relief Request VR 03, Use of Code Case OMN-1, Revision 1, on Various Motor Operated Valves. The valves will continue to be exercised (non-timed) during cold shutdowns.

Page 15 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 15 System: High Pressure Coolant Injection (HPC)

Reactor Core Isolation Cooling (RCI)

Valve(s): AO-23-18, AO-13-22 P & ID: NH-36250 (M-124), NH-36252 (M-126)

Code Class: 2 Category: C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: The HPCI and RCIC Injection Check Valves open to admit High Pressure Injection water to the reactor vessel via the main feedwater lines. The valves close to protect lower pressure rated pump suction piping when the injection motor-operated valves are open and pump discharge pressure drops.

Frequency Change Justification:

It is not practical to exercise open the HPCI and RCIC Injection check valves every 3 months per the requirements of ISTC-3510.

1. The HPCI and RCIC pumps are turbine driven using reactor steam and thus can only be operated when the plant is hot with sufficient steam pressure. However, during plant operation it is impractical to exercise these check valves open with flow per ISTC-5221(a)(1) because the injection of relatively cool water from the Condensate Storage Tanks and/or Suppression Chamber would produce reactivity excursions. Additionally, this relatively cool water could result in large, cyclic thermal stresses being experienced by various Class 1 piping systems.

2. These valves cannot be tested open per ISTC-5221(b) because the valves have no provision for torque measurement to determine a reference value or to trend breakaway torque. Further, the installed operator was not qualified or certified based on its capability to meet the requirements of ISTC-5221(b) since it does not have any design feature to allow measurement of torque or breakaway force.

3. The installed operator is only qualified for partial stroking testing, however, partial stroke testing of check valves is no longer a requirement of the OM Code.

Page 16 of 56

Monticello Nuclear Generating Plant Revision 0 It is not practical to exercise closed the HPCI and RCIC Injection check valves every 3 months per requirements of ISTC-3510.

1. Performing a close test on-line for RCIC AO-13-22 using a reverse flow test is impracticable because it would be an ALARA and personnel hazard to perform in an operating plant, as there is no isolation between it and an operating feedwater system.

2. Performing a close test on-line for HPCI AO-23-18 is impractical and would not be an increase in safety relative to ALARA. A reverse flow test would require isolating and venting the system and installing temporary bypasses (hose, gauges, etc.). Further this could pose a personnel hazard as the downstream valve MO-2068 is categorized as Category B and does not have a leakage limit consequential to its design function and it is the only boundary valve to the operating feedwater system.

The open test can only be performed by Disassembly and Examination (D & E) for these subject valves. Since both valves are considered Category C with no consequential leakage requirement, the D

& E test will fulfill the close exercise test as well and will be performed at the same time as the open test. Since each valve is a group of one , each valve will have its open and closed test performed during refueling, as required per ISTC-5221(c).

Alternate Test Frequency: Test exercise AO-23-18 and AO-3-22 open and closed via disassembly and examination during refueling in accordance with ISTC-5221(c). As an alternative to the disassembly and examination requirements of ISTC-5221(c), these check valves may be placed in the Check Valve Condition Monitoring (CVCM) Program and disassembled and examined per the requirements of Appendix II of the OM Code. If so, the frequency for disassembly and examination shall be specified in the CVCM Program Plan applicable to the valve Group.

Page 17 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 16 System: Residual Heat Removal (RHR)

Valve(s): MO-2029, MO-2030 P & ID: NH-36247 (M-121)

Code Class: 1 Category: A Code Frequency: ISTC-3510, Active Category A valves shall be exercised nominally every 3 months.

Function/Description: Shutdown Cooling (SDC) Isolation Valves open to allow RHR flow to the RHR pumps for SDC. These RHR valves are normally closed and have a closed safety position. They have interlocks in their open direction logic that prevents them from opening above a reactor pressure that is well below normal reactor operating pressure. This interlock protects the low pressure piping of the RHR system from the high-pressure reactor coolant on the other side of these valves. These valves auto-close to their safety position as primary containment isolation valves as necessary.

Frequency Change Justification: It is not practical to exercise the subject valves every 3 months per the requirements of ISTC-3510. They have interlocks in their open direction logic that prevents them from opening above a reactor pressure that is well below normal reactor operating pressure.

Alternate Test Frequency: Test exercise valves closed on a cold shutdown frequency in accordance with ISTC-3521(c).

Page 18 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 17 System: High Pressure Coolant Injection (HPC)

Valve(s): HPCI-9, HPCI-10, HPCI-14, HPCI-15, HPCI-31, HPCI-65, HPCI-71, HPCI-18 P & ID: NH-36249 (M-123), NH-36250 (M-124)

Code Class: 2 Category: C (HPCI-14, HPCI-15, HPCI-31, HPCI-65, HPCI-71, HPCI-18)

A/C (HPCI-9, HPCI-10)

Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: The Inboard/Outboard Steam Exhaust Line check valves (HPCI-9 and HPCI-10, respectively) open to pass HPCI turbine steam exhaust to the suppression pool during system operation. They close to provide a containment isolation function when the system is not operating.

The Steam Exhaust header drain check valves (HPCI-14 and HPCI-15) open to remove additional condensate to the suppression pool (below the waterline) during HPCI operation. They close to provide a containment isolation function when the system is not operating.

The Inboard/Outboard Exhaust Line Vacuum Breakers (HPCI-71 and HPCI-65) remain closed to ensure the suppression pool airspace does not pressurize during HPCI operation. They open to allow gas into the HPCI steam exhaust line from the suppression pool airspace in order to reduce the water level elevation in the line caused by the vacuum created from steam condensing in the exhaust line after shutdown of the HPCI turbine.

The Torus suction check valve (HPCI-31) opens to allow flow to the HPCI Booster pump when the selected water source is the torus. It closes to prevent water transfer from the Condensate Storage Tanks to the torus.

The HPCI Gland Seal Condenser Cooling Water Check valve (HPCI-18) opens to allow cooling water from the HPCI gland seal condenser and the lube oil cooler back to the suction of the booster pump. It closes to prevent backflow from the Gland seal condensate pump.

Frequency Change Justification: It is not practical to exercise closed the subject check valves every 3 months per the requirements of ISTC-3510. It is not practical to exercise open check valves HPCI-14, HPCI-15, HPCI-31, HPCI-65 and HPCI-71 every 3 months per the requirements of ISTC-3510. The only way to test the closed safety position of valves HPCI-9, HPCI-10, HPCI-14, HPCI-15, HPCI-65 and HPCI-71 is by using a reverse flow/leak test. The only way to test open HPCI-14, HPCI-Page 19 of 56

Monticello Nuclear Generating Plant Revision 0 15, HPCI-65 and HPCI-71 is by a similar test using a flow and/or pressure source. Testing HPCI-18 in the closed direction also requires a reverse flow test, with an extensive isolation sequence which would remove a single train of emergency high pressure injection from its standby-readiness status. HPCI-31 must be tested using a disassembly and examination process. Testing these valves at power requires isolating and venting the system, which includes manual valve realignments, opening motor operated valve breakers, and defeating auto start logic, which is a significant burden on plant resources without a compensating increasing in safety. HPCI is a single train safety system and exercising/testing these check valves as discussed here with the plant on-line requires a total loss of system function with in-plant manual actions required for restoration and corresponding substantial reduction in the level of safety.

Alternate Test Frequency: Test exercise closed HPCI-9, HPCI-10 and HPCI-18 at a cold shutdown frequency in accordance with ISTC-3522(b). Test exercise HPCI-14, 15, 65 and 71 open and closed at a cold shutdown frequency in accordance with ISTC-3522(b). Test exercise HPCI-31 open and closed at a refuel outage frequency by disassembly and examination in accordance with ISTC-5221(c).

As an alternative to reverse exercising check valves HPCI-9 and HPCI-10 during cold shutdowns, these Category A/C check valves may be placed in the Check Valve Condition Monitoring (CVCM)

Program and reverse exercised at the Appendix J, Option B frequency during the performance of Type C seat leakage testing. This extension of frequency for reverse exercising is per the guidelines provided in NUREG-1482, Revision 1, Section 4.4.7, Use of Appendix J, Option B, in Conjunction with ISTC Exercising Tests.

As an alternative to the disassembly and examination requirements of ISTC-5221(c) for HPCI-31, this check valve may be placed in the Check Valve Condition Monitoring (CVCM) Program and disassembled and examined per the requirements of Appendix II of the OM Code. If so, the frequency for disassembly and examination shall be specified in the CVCM Program Plan applicable to the valve Group.

Page 20 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 18 System: Reactor Core Isolation Cooling (RCI)

Valve(s): RCIC-9, RCIC-10, RCIC-16, RCIC-17, RCIC-31, RCIC-57, RCIC-59 P & ID: NH-36251 (M-125)

NH-36252 (M-126)

Code Class: 2 Category: C (RCIC-57, RCIC-59, RCIC-16, RCIC-17, RCIC-31)

A/C (RCIC-9, RCIC-10)

Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: The Inboard/Outboard Steam Exhaust Line check valves (RCIC-9 and RCIC-10, respectively) open to pass RCIC turbine steam exhaust to the suppression pool (below the water line) during system operation. They close to provide a containment isolation function when the system is not operating.

The Vacuum Pump Discharge check valves (RCIC-16 and RCIC-17) open to allow non-condensable gas removal from the RCIC Barometric Condenser to the Suppression Pool while the RCIC turbine is in operation. They close to provide a containment isolation function when the system is not operating.

.The Inboard/Outboard Exhaust Line Vacuum Breakers (RCIC-57 and RCIC-59) remain closed to ensure the suppression pool airspace does not pressurize during RCIC operation. They open to allow gas into the RCIC steam exhaust line from the suppression pool airspace in order to reduce the water level elevation in the line caused by the vacuum created from steam condensing in the exhaust line after shutdown of the RCIC turbine.

The Torus suction check valve (RCIC-31) opens to allow flow to the RCIC pump when the selected water source is the torus. It closes to prevent water transfer from the Condensate Storage Tanks to the torus.

Frequency Change Justification: It is not practical to exercise closed the subject check valves every 3 months per the requirements of ISTC-3510. It is not practical to exercise open check valves RCIC-31, RCIC-57 and RCIC-59 every 3 months per the requirements of ISTC-3510. The only way to test the closed safety position of valves RCIC-9, RCIC-10 RCIC-16, RCIC-17, RCIC-57 and RCIC-59 is by using a reverse flow/leak test. The only way to test open RCIC-57 and RCIC-59 is by a similar test using a flow and/or pressure source. RCIC-31 must be tested using a disassembly and examination process. Testing these valves at power requires isolating and venting the system, which includes Page 21 of 56

Monticello Nuclear Generating Plant Revision 0 manual valve realignments, opening motor operated valve breakers, and defeating auto start logic, which is a significant burden on plant resources without a compensating increase in safety. RCIC is a single train safety system and exercising/testing these check valves as discussed, with the plant on-line, requires a total loss of system function with in-plant manual actions required for restoration and corresponding substantial reduction in the level of safety.

Alternate Test Frequency: Test exercise RCIC-9 and RCIC-10 closed at a cold shutdown frequency in accordance with ISTC-3522(b). Test exercise RCIC-16 and 17, closed at a cold shutdown frequency in accordance with ISTC-3522(b). Test exercise RCIC-57 and RCIC-59 open and closed on a cold shutdown frequency in accordance with ISTC-3522(b). Test exercise RCIC-31 open and closed at a refuel outage frequency by disassembly and examination in accordance with ISTC-5221(c).

As an alternative to reverse exercising check valves RCIC-9 and RCIC-10 during cold shutdowns, these Category A/C check valves may be placed in the Check Valve Condition Monitoring (CVCM)

Program and reverse exercised at the Appendix J, Option B frequency during the performance of Type C seat leakage testing. This extension of frequency for reverse exercising is per the guidelines provided in NUREG-1482, Revision 1, Section 4.4.7, Use of Appendix J, Option B, in Conjunction with ISTC Exercising Tests.

As an alternative to the disassembly and examination requirements of ISTC-5221(c) for RCIC-31, this check valve may be placed in the Check Valve Condition Monitoring (CVCM) Program and disassembled and examined per the requirements of Appendix II of the OM Code. If so, the frequency for disassembly and examination shall be specified in the CVCM Program Plan applicable to the valve Group.

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Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 19 System: Alternate N2 (AN2)

Valve(s): AI-598, AI-708, AI-729, AI-730, SV-4235 P & ID: NH-36049-10 (M-131)

Code Class: 2 (AI-598, AI-708)

NC (AI-729, AI-730, SV-4235)

Category: A/C (AI-598, AI-708)

C (AI-729, AI-730)

B (SV-4235)

Code Frequency:

AI-598, AI-708, AI-729, AI-730 - ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

SV-4235 - ISTC-3510, Active Category B valves shall be exercised nominally every 3 months.

SV-4235 - Fail-Safe exercise test per ISTC-3560 at a frequency of every 3 months per ISTC-3510.

Function/Description: These valves must be open to supply safety grade nitrogen for critical component functions inside and outside the Drywell, among them holding the inboard MSIVs open (except for AI-729). AI-598 and AI-708 have a closed safety function as containment isolation valves.

Frequency Change Justification: It is not practical to exercise these valves every 3 months (exception: AI-729 Open Exercise Test) per the requirements of ISTC-3510 as it applies below. All of these valves, except AI-729, can only be exercised by isolating the safety related pneumatic supply to the inboard MSIVs. Since the inboard MSIVs fail closed, any drop in pneumatic supply pressure to them during power operation risks a plant transient from the MSIVs drifting closed. If there is a leak in the pneumatic supply inside containment, the accumulators there will start to lose their charge and as pressure drops the inboard MSIVs will drift close. Exercising AI-729 closed involves depressurizing and replacing at least one of the nitrogen cylinders for the A train of the alternate N2 safety related pneumatic supply, with an associated undesirable delay in restoring the train. AI-708 and AI-598 cannot be tested in a manner such that their required flow is passed. Therefore, these valves will be tested by disassembly and examination, which is not practical on a quarterly or cold shutdown frequency.

Alternate Test Frequency: Test exercise AI-729 and AI-730 closed at a cold shutdown frequency in accordance with ISTC-3522(b). Test exercise AI-730 open at a cold shutdown frequency in accordance with ISTC-3522(b). Test exercise AI-598 and AI-708 open and closed by disassembly and Page 23 of 56

Monticello Nuclear Generating Plant Revision 0 examination at a refueling outage frequency in accordance with ISTC-5221(c). Test exercise SV-4235 open and closed and perform Fail Safe open test at a cold shutdown frequency in accordance with ISTC-3521(c).

As an alternative to the disassembly and examination requirements of ISTC-5221(c) for AI-598 and AI-708, these check valves may be placed in the Check Valve Condition Monitoring (CVCM)

Program and disassembled and examined per the requirements of Appendix II of the OM Code. If so, the frequency for disassembly and examination shall be specified in the CVCM Program Plan applicable to the valve Group.

Page 24 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 20 System: Post Accident Sampling (PAS)

Valve(s): PAS-59-5, PAS-59-6 P & ID: NH-96042-1 Code Class: 2 Category: C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: These excess flow check valves are normally open and have a closed safety position to isolate the non-safety PASS system from the safety related RHR system on high sample line flow. These valves would limit discharge of potentially highly contaminated liquid into the Reactor Building environment such as might occur should there be a downstream break.

Frequency Change Justification: It is not practical to exercise closed these PASS system excess flow check valves every 3 months per the requirements of ISTC-3510. These valves are tested by isolating the system and removal of the valves from the system for bench testing). This test is impractical during power operations since it renders the non-redundant PAS system out of service and the additional resource burden does not result in a compensating increase in safety to perform at power operation.

Alternate Test Frequency: Test exercise valves closed at a cold shutdown frequency as allowed by ISTC-3522(b).

Page 25 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 21 System: Control Rod Drive Hydraulics (CRH)

Valve(s): BF-12, BF-14, BF-24, BF-26, BF-35, BF-37, BF-46, BF-48 P & ID: NH-36242-2 (M-116-2)

Code Class: 2 Category: A/C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: These Control Rod Drive to Reactor Vessel Instrumentation Check Valves are normally open and have a closed safety position to separate the non-safety control rod drive system from the safety related reactor coolant pressure boundary and minimize loss of reference leg liquid and resulting instrument error.

Frequency Change Justification: It is not practical to exercise closed the CRD backfill check valves every 3 months per the requirements of ISTC-3510. The lines containing these valves tie into the reference legs of multiple sensitive reactor vessel pressure and level instruments that are important during both power operation and cold shutdown operation. These check valves are exercised closed by isolating and venting the system, installing temporary bypasses (hoses, gages, etc.), and performing a reverse flow seat leakage test. Multiple instruments would have to be removed with outputs impaired/simulated before exercising these check valves closed. These actions could result in inadvertent reactor scrams and system actuations or failure to actuate.

Alternate Test Frequency: Test exercise valves closed at a refueling outage frequency as allowed by ISTC-3522(c).

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Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 22 System: Various Valve(s): XFV-1 through XFV-89 P & ID: Various Code Class: 1 and 2 Category: AC Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: These excess flow check valves are normally open and have a closed safety position to limit flow in the event of an instrument line break outside containment and release of potentially highly contaminated steam/water to the reactor building and containment. These valves are not considered check valves in the normal sense but are in reality a flow-limiting device. As such, bi-directional testing does not apply and the exercising test will be performed in the safety-related close (forward flow restricted) position.

Frequency Change Justification: It is not practical to exercise closed these excess flow check valves every 3 months per the requirements of ISTC-3510. The closed position of these valves is tested by installing temporary bypasses (hoses, gages, etc.) and performing steps similar to, but more extensive than, a check valve seat leakage test. The isolation of the lines containing these excess flow check valves takes instrumentation out of service that is important during both power operation and cold shutdown operation. Removing from and restoring to service many of these instruments risks plant transients, safety system actuations and/or blocking of safety system functions.

Alternate Test Frequency: Test exercise valves closed at a refuel outage frequency in accordance with ISTC-3522(c) (exception: as described above, bi-directional testing is not required to be performed) during Category AC leakage test.

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Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 23 System: Instrument and Service Air (AIR)

Valve(s): AI-571 P & ID: NH-36049-12 (M-131)

Code Class: 2 Category: A/C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: Instrument Air/Nitrogen pneumatic supply to drywell inboard containment isolation check valve, which is normally open to several components in the Drywell, including accumulator makeup/keepfill for the ADS and LLS function of two safety relief valves. It has a closed safety position as a primary containment isolation valve.

Frequency Change Justification: It is not practical to exercise open or closed the Instrument Air supply check valve every 3 months per the requirements of ISTC-3510. The valve closure exercise is performed by isolating and venting the system, installing temporary bypasses (hoses, gages, etc.) and performing a reverse flow, seat leakage type test. The valve open exercise would involve opening a downstream bleed line, which would take the pneumatic supply source away from safety-related components and increase the potential for a plant transient. Exercising this valve during power operation would compromise accumulator makeup/keepfill pneumatic supply for the ADS function of one SRV and the LLS function of another SRV. Removing the functional capability in order to exercise this check valve reduces plant safety as it increases the likelihood and challenge of a plant transient.

Alternate Test Frequency: Test exercise valve open and closed at a cold shutdown frequency in accordance with ISTC-3522(b). As an alternative to reverse exercising this check valve during cold shutdowns, this Category A/C check valve may be placed in the Check Valve Condition Monitoring (CVCM) Program and reverse exercised at the Appendix J, Option B frequency during the performance of Type C seat leakage testing. This extension of frequency for reverse exercising is per the guidelines provided in NUREG-1482, Revision 1, Section 4.4.7, Use of Appendix J, Option B, in Conjunction with ISTC Exercising Tests.

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Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 24 System: Instrument and Service Air (AIR)

Valve(s): AI-613 through 619, AI-663, AI-666, AI-669, AI-672, AI-675, AI-678, AI-681, AI-683, AI-685, AI-694, AI-695 P & ID: NH-36049-14 (M-131)

Code Class: NC Category: C (AI-663, AI-666, AI-669, AI-672, AI-675, AI-678, AI-681, AI-694, AI-695)

A/C (AI-613, AI-614, AI-615, AI-616, AI-617, AI-618, AI-619, AI-683, AI-685)

Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: These valves function to ensure a safety grade nitrogen supply is directed to the primary containment atmospheric control valves and not diverted from the safety grade nitrogen supply or component accumulators should there be a break or other failure in the non-safety pneumatic supply.

Frequency Change Justification: It is not practical to exercise closed the Instrument Air (normally nitrogen) check valves every 3 months per the requirements of ISTC-3510. The check valves are tested by isolating and venting the system, installing temporary bypasses (hoses, gages, etc.), and verifying full or no reverse flow across them. The inflatable soft seats (T-ring seals) of multiple primary containment isolation valves are non-functional (with a loss of Containment integrity) while these check valves are being exercised.

Alternate Test Frequency: Test exercise valves open (exception: AI-694 and AI-695) and closed at a refueling outage frequency in accordance with ISTC-3522(c).

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Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 25 System: Various Valve(s): AI-610-1, AI-610-2, AI-610-3, AI-610-4, AI-611, AI-612, AI-13-4, AI-13-7 P & ID: NH-36049-12 (M-131)

NH-36246 (M-120)

NH-36247 (M-121)

NH-36250 (M-124)

NH-36252 (M-126)

Code Class: NC Category: A/C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: These valves must close to ensure various safety related valve actuator pneumatic supply accumulator contents are not diverted to the non-safety portion of the pneumatic supply should a break or other failure occur there Frequency Change Justification: It is not practical to exercise closed these Instrument Air check valves every 3 months per the requirements of ISTC-3510. These check valves are exercised closed by isolating and venting the system, installing temporary bypasses (hoses, gages, etc), and verifying no reverse flow across them. The exercise open test would involve addition of temporary gauges, removing an ECCS pump, RCIC pump or SRV from service and blow-down of the associated air accumulator. The closure test is impractical to perform during normal power operations or during cold shutdown conditions. The open test (AI-610-1, AI-610-2, AI-610-3, AI-610-4, AI-611, AI-612) constitutes a hardship to perform during power operations or cold shutdown and would require in-plant manual operator actions to restore the system if an accident occurred during the test.

The open and closed exercise test for AI-13-4 and AI-13-7 requires de-inerting the drywell (primary containment) for access to these valves for connection of test equipment.

Alternate Test Frequency: Test exercise valves open and closed at a refueling outage frequency in accordance with ISTC-3522(c).

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Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 26 System: Residual Heat Removal (RHR)

Valve(s): AO-10-46A, AO-10-46B P & ID: NH-36246 (M-120)

NH-36247 (M-121)

Code Class: 1 Category: A/C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: The RHR Injection Check Valves permit injection flow into the reactor re-circulation loops from the Suppression Chamber when the applicable RHR pumps discharge pressure are above reactor pressure. These check valves also prevent reverse flow into the RHR System suction piping (lower design pressure rating) from the reactor when the motor operated containment isolation valves are open.

Frequency Change Justification: It is not practical to exercise open or closed the RHR LPCI Testable check valves every 3 months per the requirements of ISTC-3510. These check valves cannot be exercised during normal plant operation via a full flow test because the respective injection motor operated isolation valves can only be opened when reactor pressure has decreased below 460 psig (for protection of lower pressure rated piping from normal reactor pressure). Additionally, when the reactor is critical, or near critical, it is impractical to exercise these check valves open with flow due to reactivity excursions from the injection of relatively cool water from the Suppression Chamber.

Exercising the valves to the close position can only be performed via a seat leakage test, which requires access to the Drywell (de-inerting Primary Containment). The close exercise is considered impractical at cold shutdown due to the extensive isolation required for the test, radiation exposure during the test and rendering one division of RHR unavailable for shutdown cooling service if needed, constituting a reduction in safety.

Alternate Test Frequency: Test exercise valves open using the required flow at a cold shutdown frequency in accordance with ISTC-3522(b). Test exercise valves closed at a refueling outage frequency in accordance with ISTC-3522(c).

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Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 27 System: Core Spray (CSP)

Valve(s): AO-14-13A, AO-14-13B P & ID: NH-36248 (M-122)

Code Class: 1 Category: A/C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: The Core Spray Injection Check valves permit injection flow into the reactor vessel from the Suppression Chamber when the applicable Core Spray pump discharge pressure is above reactor pressure. These check valves also prevent reverse flow into the Core Spray System suction piping (lower design pressure rating) from the reactor when the motor operated containment isolation valves are open.

Frequency Change Justification: It is not practical to exercise open the Core Spray Injection Testable check valves every 3 months per the requirements of ISTC-3510. The CS system injection motor operated valves can only be opened when reactor pressure has decreased below 460 psig (for protection of lower pressure rated piping from normal reactor pressure). Additionally, when the reactor is critical, or near critical, it is impractical to full stroke exercise these check valves with flow due to reactivity excursions caused by the injection of relatively cool water from the Suppression Chamber or Condensate Storage Tanks.

Exercising the valves open is impractical on a Cold Shutdown frequency because unnecessary stress cycles, even from lesser temperature differentials, should be avoided, especially to that portion of the system within the reactor vessel. Full flow injection with the reactor vessel head in place involves an unnecessary personnel challenge without a compensating increase in the margin of plant safety, involves large, rapid water inventory transfers and potential reactor water chemistry perturbations.

Exercising the valves to the close position can only be performed via a seat leakage test, which requires access to the Drywell (de-inerting Primary Containment) during cold shutdown. This is considered impractical due to the extensive isolation required for the test, radiation exposure during the test and rendering one train of an ECCS unavailable for core coverage if needed, constituting a reduction in safety.

Alternate Test Frequency: Test exercise valves open and closed at a refueling outage frequency in accordance with ISTC-3522(c).

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Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 28 System: Primary Containment (PCT)

Valve(s): AO-2382A, AO-2382B, AO-2382C, AO-2382E, AO-2382F, AO-2382G, AO-2382H, AO-2382K P & ID: NH-36258 (M-143)

Code Class: NC Category: A/C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: The Torus to Drywell Vacuum Breakers open to permit gases to flow from the suppression chamber to the drywell. This prevents back flow of water and excessive hydrodynamic loads from clearing elevated water legs in the downcomers submerged in the suppression pool water and prevents the primary containment from exceeding the external design pressure. Additionally, these valves close to prevent suppression pool bypass in the event of a DBA LOCA.

Frequency Change Justification: It is not practical to exercise open and closed the Torus to Drywell Vacuum Breaker check valves every 3 months per the requirements of ISTC-3510. This testing cannot be conducted quarterly due to the necessity to enter the suppression chamber to perform testing. The actuators on the valves do not meet code requirements for actuator size and therefore cannot be used to test quarterly without a plant modification. Cold Shutdown testing is not practical due to the need to de-inert and enter the suppression chamber (Primary Containment), establish temporary lighting, perform leak rate tests on hatches, and establish confined space control. Therefore, these valves will be tested on a Refueling Outage frequency when all of the conditions for testing can be efficiently established.

These valves will be tested by measuring the breakaway torque and comparing this value to a reference value established when the valves are known to be in good condition in accordance with ISTC-5221(b) allowances. The return to closure will also be verified visually as the valve internals can be observed while stroking the valve.

Alternate Test Frequency: Test exercise valves open and closed at a refueling outage frequency in accordance with ISTC-3522(c) and ISTC-5221(b).

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Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 29 System: High Pressure Coolant Injection (HPC)

Reactor Core Isolation Cooling (RCI)

Valve(s): HPCI-32, RCIC-41 P & ID: NH-36250 (M-124)

NH-36252 (M-126)

Code Class: 2 Category: C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: These Condensate Storage (CST) Suction Check Valves allow flow to the HPCI and RCIC pumps from the CST. They also act as an isolation from back flow to the CSTs of potentially contaminated water from the suppression pool.

Frequency Change Justification: It is not practical to exercise these valves closed every 3 months per the requirements of ISTC-3510. Testing these valves for closure during plant operation on a quarterly basis is impractical due to the extensive isolation, removal of single train systems important to safety for an extended time period and the resultant Tech Spec Actions Completion Times and the in-plant manual actions needed to restore the systems to service if needed to mitigate the consequences of an accident or transient.

Alternate Test Frequency: Test exercise valves closed at a Cold Shutdown frequency in accordance with ISTC-3522(b).

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Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 30 System: High Pressure Coolant Injection (HPC)

Reactor Core Isolation Cooling (RCI)

Valve(s): HPCI-42, RCIC-37 P & ID: NH-36250 (M-124)

NH-35252 (M-126)

Code Class: 2 Category: C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: These pump minimum flow check valves open for pump protection purposes to the torus (suppression pool in Primary Containment) for the HPCI and RCIC pumps if needed during system operation. They close to prevent backflow of suppression pool/CSP water to the condensate storage tank, eliminate overpressurization concerns in RHR/CSP suction piping and prevent diversion of RHR in suppression pool cooling mode.

Frequency Change Justification: It is not practical to exercise open or closed the HPCI and RCIC minimum flow check valves every 3 months per the requirements of ISTC-3510. There is no means of measuring flow rate directly through these valves during quarterly pump testing, as instrumentation is not installed. Installation of instrumentation to perform this test constitutes a hardship, which does not result in a compensating increase in safety. Exercising the valves closed would require de-inerting containment, entering the torus and making multiple trains of ECCS systems unavailable. Therefore, disassembly and examination shall be used to meet Code exercising requirements. .

Alternate Test Frequency: Test exercise valves open and closed via disassembly and examination at a refueling outage frequency in accordance with ISTC-5221(c). As an alternative to the disassembly and examination requirements of ISTC-5221(c), these check valves may be placed in the Check Valve Condition Monitoring (CVCM) Program and disassembled and examined per the requirements of Appendix II of the OM Code. If so, the frequency for disassembly and examination shall be specified in the CVCM Program Plan applicable to the valve Group.

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Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 31 System: Control Rod Drive Hydraulics (CRH)

Valve(s): CRD-114, CRD-115, CRD-138, CV-126, CV-127 (typical - 121 valves total, one per HCU)

P & ID: NH-36245 (M-119)

Code Class: 2 Category: B (CV-126, CV-127)

C (CRD-114, CRD-115, CRD-138)

Code Frequency: CRD-114, CRD-115, CRD-138 - ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

CV-126, CV-127 - ISTC-3510, Active Category B valves shall be exercised nominally every 3 months. Fail-Safe testing of valves per ISTC-3560 at a frequency of every 3 months per ISTC-3510.

Function/Description: CRD-114 opens to allow exhaust water from the top of the drive piston to flow to the scram discharge volume during a control rod scram. CRD-115 closes to prevent diversion of scram accumulator water charge. CRD-138 closes to prevent control rod scram water from being diverted to the drive cooling water header when CV-126 opens for a rod scram. CV-126 opens to allow water from the scram accumulator to pressurize the underside of the drive piston. CV-127 opens to allow water to exhaust from the top of the drive piston during a rod scram.

Frequency Change Justification: For those control rod drive system valves where testing could result in the rapid insertion of one or more control rods, the rod scram test frequency identified in the Plant Technical Specifications may be used to minimize rapid reactivity transients and wear of the control rod drive mechanism. This position is that described in NUREG-1482, Revision 1, Section 4.4.6, Testing Individual Scram Valves for Control Rods in Boiling-Water Reactors.

CRD valves for which this position is applicable:

CRD-114 (for exercise to the open position)

CV-126* (for exercise to the open position, fail-open exercise)

CV-127* (for exercise to the open position, fail-open exercise)

CRD-115 (for exercise to the open position)

CRD-138 (for exercise to the closed position)

For the close test of CRD-138, it is recognized that not all Control Rod Drives (CRD) may be withdrawn (partially or fully) from the reactor core at any given time in the cycle. Therefore, the Page 36 of 56

Monticello Nuclear Generating Plant Revision 0 normal CRD exercise test may not exercise all CRD-138 valves to the closed position every quarter.

As discussed in NUREG-1482, Revision 1, Section 4.4.6 for this valve, normal control rod motion may verify the cooling water header check valve moving to its safety function position (closed). This valve will be tested quarterly if the respective control rod is in a withdrawn (partially or fully) position at the time of the normally scheduled quarterly surveillance. All CRD-138 valves will be tested to the closed position at least once per refueling cycle with a labeled frequency as TS. This position is in agreement with NUREG-1482, Revision 1, Section 4.4.6 since undue manipulation of CRD mechanisms and core reactivity changes is not warranted to perform the close test at a frequency beyond that required to meet Technical Specification test frequency.

In summary, MNGP Technical Specification 3.1.4 requires verifying CRD Scram Insertion Times are within specified limits after fuel movement within the affected core cell. During a routine refueling outage all control rods will be affected. As such, testing for the specified directions and tests up to the allowed refuel outage duration will be enveloped by testing the respective CRD during Scram Time Testing as specified by the Technical Specification requirement and during normal CRD exercise testing as discussed above.

Alternate Test Frequency: MNGP shall utilize a Technical Specification frequency, which does not exceed refueling outage frequency. Use of the frequency code TS identifies those valves for which this position is employed. This is described in NUREG-1482, Revision 1, Section 4.4.6..

NUREG-1482, Revision 1, Section 4.4.6. covers all required testing for CV-126 and CV-127.

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Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 32 System: Fuel Pool Cooling and Cleanup (FPC)

Valve(s): PC-20-1 and PC-20-2 P & ID: NH-36256 (M-135)

Code Class: NC Category: C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: The Spent Fuel Pool Return Line Anti-Siphon Check Valves close to prevent a large rapid water loss from the spent fuel pool should there be a break in the return line.

Frequency Change Justification: It is not practical to exercise closed the Spent Fuel Pool Return check valves every 3 months per the requirements of ISTC-3510. System design does not allow for a method to reverse flow through these check valves such that they could be exercised closed. The subject valves are augmented scope and are not Class 1, 2 or 3. Performing the open test in conjunction with the close test during disassembly and examination is appropriate and does not compromise degradation monitoring. Disassembly and examination has been the method of testing used in the 3rd IST interval with satisfactory results.

Alternate Test Frequency: The subject valves are augmented scope and are not Class 1, 2 or 3. As such, these valves will be grouped with at least one valve test exercised open and closed via disassembly and examination at a frequency of at least once each cycle, including refueling. This is satisfactory and essentially equivalent to the refueling outage frequency allowed by ISTC-5221(c)(3).

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

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Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 33 System: Instrument and Service Air (AIR)

Valve(s): AI-12-9, AI-12-10, AI-12-11, AI-12-12 P & ID: NH-36049-10 (M-131)

Code Class: NC Category: C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: The inboard MSIV pneumatic supply check valves open to supply safety grade nitrogen for operation of the inboard MSIVs. The pneumatic supply is required to assist the inboard MSIV actuator springs in closing the MSIVs during a DBA LOCA. These check valves close to allow the MSIVs to remain full open during momentary drops in pneumatic supply pressure. If more than 2 of the inboard MSIV drift close past the 90% open position, a reactor scam is initiated.

Frequency Change Justification: It is not practical to exercise these valves every 3 months per the requirements of ISTC-3510. These check valves are located inside the drywell (primary containment) which is inerted and not accessible with the plant on-line. These valves can only be exercised open by disassembly and examination because passing their required flow rate cannot be demonstrated by in-place testing. The subject valves are augmented scope and are not Class 1, 2 or 3. Performing the open test in conjunction with the close test during disassembly and examination is appropriate and does not compromise degradation monitoring.

Alternate Test Frequency: These subject valves are augmented scope and are not Class 1, 2 or 3.

As such, these valves will be grouped with at least one valve test exercised open and closed via disassembly and examination at a frequency of at least once each cycle, including refueling in accordance with ISTC-5221(c). This would be equivalent to the refueling outage frequency allowed by ISTC-5221(c)(3). As an alternative to the disassembly and examination requirements of ISTC-5221(c), these check valves may be placed in the Check Valve Condition Monitoring (CVCM)

Program and disassembled and examined per the requirements of Appendix II of the OM Code. If so, the frequency for disassembly and examination shall be specified in the CVCM Program Plan applicable to the valve Group.

Page 39 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 34 System: Residual Heat Removal (RHR)

Valve(s): RHR-81 P & ID: NH-36247 (M-121)

Code Class: 1 Category: A/C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: The RHR Shutdown Cooling (SDC) suction line containment penetration pressure equalizing check valve has a safety function to open to protect the piping through penetration X-12 from thermal overpressurization post accident when SDC valves MO-2029 and MO-2030 are closed. It also has a safety function to close as a Containment Isolation Valve.

Frequency Change Justification: It is not practical to exercise this valve every 3 months per the requirements of ISTC-3510. Establishing the system lineup needed to exercise this check valve open and closed during power operation requires defeating the overpressurization protective function for the respective containment penetration. It would require manual operator actions to restore the system if an accident occurred while the test is in progress in order to restore the overpressurization protection function of the valve.

Alternate Test Frequency: Test exercise valve open and closed at a cold shutdown frequency in accordance with ISTC-3522(b).

Page 40 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 35 System: Instrument and Service Air (AIR)

Valve(s): AI-243-1, AI-243-2, AI-244-1, AI-244-2 P & ID: NH-36246 (M-120) NH-36247 (M-121)

Code Class: NC Category: C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: These check valves allow non safety pneumatic supply to CV-1728 and CV-1729, the 11 and 12 RHR Heat Exchanger RHRSW Outlet pressure control valves, and to the CGCS A and B train containment isolation valves, and prevent loss of the safety related pneumatic supply when the RHR auxiliary air compressors are operating.

Frequency Change Justification: It is not practical to exercise these valves every 3 months per the requirements of IST-3510, nor cold shutdown per the requirements of ISTC-3522(b). Connection of test equipment and performance of a reverse flow or leak type test to exercise each of these series check valve pairs closed makes one train of RHRSW and one train of CGCS inoperable. An open test will require bleed-down through the respective check valves which would also require rendering one train of RHRSW and one train of CGCS inoperable. Functional loss of the respective train of RHRSW could occur, due to drain-down of the system process line piping if the respective RHRSW Control Valve would open upon loss of air. Further, it would require in-plant operator action to fill and vent the RHRSW piping, with delay to restore the systems to service if needed to mitigate the consequences of an accident or transient.

Alternate Test Frequency: Test exercise valves open and closed at a refueling outage frequency in accordance with ISTC-3522(c).

Page 41 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 36 System: Condensate Storage (CST)

Valve(s): CST-189 and CST-98 P & ID: NH-85509 (M-114-1)

Code Class: 2 Category: C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: These valves supply keep-fill water pressure to the B train of RHR. Their safety function is to close to ensure diversion of ECCS injection flow does not occur during ECCS System operation.

Frequency Change Justification: It is not practical to exercise these valves open every 3 months per the requirements of ISTC-3510. Isolation of the keep fill path to the respective system, connection of test equipment and required venting to perform the test may result in voiding in the RHR ECCS header. Additionally, it would be difficult to determine the length of venting time required to ensure any masking affect has been accounted for from potential pockets of compressed air or potential inter-system communication. Further, it would require in-plant operator action to fill and vent the RHR water piping, with delay to restore the system to service if needed to mitigate the consequences of an accident or transient Alternate Test Frequency: Test exercise valves open at a cold shutdown frequency in accordance with ISTC-3522(b).

Page 42 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 37 System: Standby Liquid Control (SLC)

Valve(s): XP-3-1, XP-3-2 P & ID: NH-36253 (M-127)

Code Class: 2 Category: C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: The SLC pump discharge check valves open to provide the operating pump an injection to the vessel.. They close to ensure there is no recirculation back-leakage from the operating SLC pump through the idle SLC pump, and to ensure the failure of the idle pumps relief valve does not divert injection flow back to the poison tank.

Frequency Change Justification: It is not practical to exercise the SLC pump discharge check valves closed every 3 months per the requirements of ISTC-3610. There are no upstream isolation valves and test connections to quantitatively or qualitatively perform a reverse flow test to verify closure of the check valves.

Installation of process line taps to perform this test is impractical without a compensating increase in safety. Therefore, disassembly and examination shall be used to meet Code exercising requirements.

Alternate Test Frequency: These valves will be grouped and tested in accordance with ISTC-5221(c). At least one valve from the group will be disassembled and examined each refueling outage for the open and close exercise test. As an alternative to the disassembly and examination requirements of ISTC-5221(c), these check valves may be placed in the Check Valve Condition Monitoring (CVCM) Program and disassembled and examined per the requirements of Appendix II of the OM Code. If so, the frequency for disassembly and examination shall be specified in the CVCM Program Plan applicable to the valve Group.

Page 43 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 38 System: Control Rod Drive Hydraulic (CRH)

Valve(s): CV-3-32A, CV-3-32B, CV-3-32C, CV-3-32D, CV-3-33A, CV-3-33B, CV-3-33C, CV 33D, SV-3-31A, SV-3-31B, SV-3-31C, and SV-3-31D.

P & ID: NH-36245 (M-119)

Code Class:

Class 2 - CV-3-32A, CV-3-32B, CV-3-32C, CV-3-32D, CV-3-33A, CV-3-33B, CV-3-33C, CV-3-33D Skid-mounted SV-3-31A, SV-3-31B, SV-3-31C and SV-3-31D.

Category: B Code Frequency: ISTC-3510, Active Category B valves shall be exercised nominally every 3 months.

Function/Description:

Scram Discharge Volume Vent (SDV) CV-3-32A, CV-3-32B, CV-3-32C, CV-3-32D SDV CV-3-33A, CV-3-33B, CV-3-33C, CV-3-33D These valves are held normally open by the application of air pressure to their piston operators. When open, the vent valves vent the high points of the scram discharge piping to atmosphere, and the drain valves drain the low point of the piping to the Reactor Building equipment drain tank (T-56) and the Reactor Building Floor drain tank (T-55).

During a scram, they close to their safety position by internal spring pressure upon the removal of control air pressure. This position contains leakage past the CRD drive seals to the SDV.

Scram Header Pilot Valves Control air pressure is supplied to the operators of the above SDV valves by two sets of two three-way solenoid-operated air valves. One set of solenoid valves SV-3-31A and SV-3-31B controls the inboard valves, while the other set SV-3-31C and SV-3-31D, controls the outboard valves. During normal reactor operation, each of the two logic channels of the Reactor Protection System energizes one of the two pilot valves in each set, so that control air pressure is supplied to the piston operators of the discharge volume vent and drain valves. Upon the initiation of a reactor scram, both Reactor Protection System logic channels are de-energized and both pilot valves (in both sets) isolate supply air and open a vent path, venting air from the vent and drain valve operators and permitting these valves to close.

Page 44 of 56

Monticello Nuclear Generating Plant Revision 0 Frequency Change Justification: It is not practicable to perform a close exercise test of the SDV vent and drain valves in combination with the Scram header pilot valves every 3 months per the requirements of ISTC-3510. The Scram header pilot valves are considered as skid-mounted per IST Technical Position TP 05, and it is not practicable to test these exclusive to the SDV vent and drain valves.

The plant does not have dedicated logic controls to perform this testing without putting in Reactor Protection System trips by lifting leads, installing multiple jumper blocks or manual scram trips.

Performing this activity on-line would unnecessarily risk plant transients or would actively shutdown an operating plant. Performing this test during cold shutdown conditions would unnecessarily wear the control rod drives. As such, it would create hardship that is not commensurate with a compensating increase in safety.

Technical Specification SR 3.1.8.3 requires testing of the SDV vent and drain valves on a 24 month frequency. More frequent testing would involve a rapid insertion of the control rods since dedicated logic controls do not exist, and the intention of NUREG-1482, Revision 1, Section 4.4.6 is to avoid this repeated situation, testing will be performed in accordance with the requirements of Technical Specification SR 3.1.8.3. In addition, Technical Specification 5.5.5.d states that nothing in the ASME OM Code shall be construed to supersede the requirements of the Technical Specifications.

Alternate Test Frequency: CV-3-32A, CV-3-32B, CV-3-32C, CV-3-32D, CV-3-33A, CV-3-33B, CV-3-33C, CV-3-33D in combination with SV-3-31A, SV-3-31B, SV-3-31C and SV-3-31D will be exercised on a Tech Spec required frequency.

Page 45 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 39 (DELETED)

Page 46 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 40 System: Main Steam Valve(s): AO-2-80A through D, AO-2-86A through D P & ID: NH-36241 (M-115)

Code Class: 1 Category: A Code Frequency: ISTC-3510, Active Category A valves shall be exercised nominally every 3 months.

ISTC-3560, Fail Safe Test for AO-2-86A through D at a frequency of every 3 months per ISTC-3510 (see DTJ 13 for AO-2-80A through D).

Function/Description:

These valves are normally open during power operations to deliver steam from the reactor vessel to the main turbine [USAR 1.3.7]. This function is not required for safe shutdown or accident mitigation.

These valves must close to prevent inventory loss following main steam line pipe breaks outside containment and to provide containment isolation. These valves receive a Group 1 Isolation Signal to close upon 1) Reactor low low water level; 2) Main steam line high flow; 3) Main steam line tunnel high temperature; 4) Main steam line low pressure (RUN mode only) [USAR Table 5.2-3b]. The maximum assumed closure time per the main steam line break accident analyses is 10.5 seconds

[USAR Table 14.7-18]; however, more restrictive stroke time limits of 3.3 to 9.9 seconds are imposed to assure that the analyzed limits are not exceeded [USAR Table 5.2.3b/TS SR 3.6.1.3.6]. These valves have a fail safe function in the closed position [USAR 5.2.2.5.3].

Valves AO-2-80A through D are spring and pneumatically closed [USAR 5.2.2.5.3.1]. Valves AO 86A through D are spring closed [USAR 5.2.2.5.3.2].

AO-2-80A is the inboard containment isolation valve for penetration X-7A [USAR Table 5.2-3a].

AO-2-80B is the inboard containment isolation valve for penetration X-7B [USAR Table 5.2-3a].

AO-2-80C is the inboard containment isolation valve for penetration X-7C [USAR Table 5.2-3a].

AO-2-80D is the inboard containment isolation valve for penetration X-7D [USAR Table 5.2-3a].

Page 47 of 56

Monticello Nuclear Generating Plant Revision 0 Frequency Change Justification: NUREG-1482 , Revision 1, "Guidelines for Inservice Testing at Nuclear Power Plants", Section 2.4.5, "Deferring Valve Testing to Cold Shutdown or Refueling Outages" identifies impractical conditions justifying test deferrals as those conditions that could result in unnecessary challenges to safety systems, place undue stress on components, cause unnecessary cycling of equipment, or unnecessarily reduce the life expectancy of the plant systems and components.

Full stroke testing of the MSIVs, (Fail-Safe testing of outboard MSIVs) even at a reduced power places the plant in an abnormal operating condition and introduces an unnecessary challenge to plant equipment and to the operators involved in the evolution. 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 which may not be related to the MSIVs themselves. This would introduce additional equipment cycling and plant thermal transients. Therefore, plant power reductions will not be performed specifically to perform full exercise stroke testing (Fail-Safe testing of outboard MSIVs) of the MSIVs.

Additionally, unnecessary cycling of the MSIVs by performing a full fast stroke on a quarterly basis (Fail-Safe testing of the outboard MSIVs) can contribute to MSIV degradation, thereby unnecessarily reducing the life expectancy of the component.

In summary, deferral of the above specified tests to cold shutdown does not conflict with NUREG 1482, Revision 1, guidance and the extended test frequency is consistent with other BWRs regarding MSIV testing. Continuance of the MSIV testing requirements of TRM TSR 3.6.1.3.2 will satisfy OM Code quarterly partial stroke exercise testing requirements.

Inboard valves AO-2-80A through AO-2-80D will not be included on the cold shutdown listing for a given shutdown if containment is not de-inerted, consistent with NUREG 1482, Revision 1, Section 3.1.1.3.

Alternate Test Frequency:

AO-2-80A through AO-2-80D and AO-2-86A through AO-2-86D quarterly full stroke exercise testing is deferred to cold shutdown frequency IAW ISTC-3521(c).

AO-2-86A through AO-2-86D fail-safe test is deferred to cold shutdown frequency IAW ISTC-3560.

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Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 41 System: Reactor Water Cleanup (RWCU)

Valve(s): MO-2397 and MO-2398 P & ID: NH-36254 (M-128)

Code Class: 1 Category: A Code Frequency: ISTC-3510, Active Category A valves shall be exercised nominally every 3 months.

Function/Description: These valves are normally open and are located in the RWCU supply line from the reactor recirculation loops and reactor bottom head drain line. This supply line feeds the regenerative and non-regenerative heat exchangers and provides a suction flowpath to the cleanup recirc pumps.

These valves must close to provide containment isolation during emergency and accident conditions.

These valves receive a Group 3 isolation signal to close. A Group 3 isolation signal is initiated by 1)

Reactor Low Low water level, 2) High drywell pressure, 3) High RWCU Flow, or 4)High RWCU Room Temperature [USAR Table 5.2-3b]. This valve is considered a containment isolation valve for penetration X-14 and receives an automatic signal to close within 40 seconds [USAR Tables 5.2-3a and 5.2-3b]. The valves also receive an auto close signal upon SBLC activation (NX-7823-4-10).

Frequency Change Justification: NUREG-1482, Revision 1, "Guidelines for Inservice Testing at Nuclear Power Plants", Section 2.4.5, "Deferring Valve Testing to Cold Shutdown or Refueling Outages" identifies impractical conditions justifying test deferrals as those conditions that could result in unnecessary challenges to safety systems, place undue stress on components, cause unnecessary cycling of equipment, or unnecessarily reduce the life expectancy of the plant systems and components.

The RWCU system is inservice during normal plant operations to maintain water purity. This system ensures that reactor coolant pH, chlorides, conductivity, and activity are maintained within specified limits. These limits are established to prevent the likelihood of exceeding 10CFR100 release limits.

Additionally, the RWCU system maintains water purity to minimize the occurrence of stress corrosion cracking of the vessel and attached stainless steel piping systems.

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Monticello Nuclear Generating Plant Revision 0 Basis for not performing Quarterly frequency full stroke exercising Exercising and stroke timing these valves closed quarterly requires the RWCU System to be taken out of service which could result in the degraded chemical makeup of reactor coolant and subjects the entire RWCU System to unnecessary transients. These unnecessary transients could lead to degradation and damage of related system components (e.g., pumps, valves, piping, demineralizers).

In addition, cycling the RWCU system in and out of service significantly increases the potential for water hammer and loss of demineralizer pre-coat.

Failure of these valves in the closed position would result in a complete loss of the RWCU system.

Failure of other components during the removal or restoration of the system to service could result in failure to maintain reactor water chemistry within the TRM 3.4.1 limits. This could lead to a reduction in power or a forced shutdown. Additionally, shutdown of the RWCU system can lead to hydraulic transients and crud bursts that will result in increases in radiation levels and higher dose.

Basis for not performing Quarterly frequency part stroke exercising Based on design, these valves cannot be partial stroked quarterly.

Basis for not performing Cold shutdown frequency full stroke exercising Cold shutdown close stroke testing is not practical since it is more critical during a forced shutdown to have RWCU in-service to mitigate the effects of a chemistry transient as a result of the shutdown.

Failure of these valves in the closed position (i.e. safety position) during a cold shutdown outage could result in loss of the RWCU system and could inhibit the ability to recover from any chemistry transient as a result of the shutdown. This could lead to a delay in the plant startup, which impacts unit availability. Shutdown to the RWCU system in a forced outage will also inhibit the ability to cleanup the reactor vessel and result in an increase in radiation levels and personnel dose. It is also impractical to de-inert for repair of MO-2397 if it fails during cold shutdown testing.

Basis for performing Refueling frequency full stroke exercising Refueling outages have sufficient duration to allow the RWCU system to adequately cleanup the primary coolant prior to being shutdown for testing. 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 availability and without adverse ALARA effects.

In summary, deferral of the full stroke exercise tests to refueling does not conflict with NUREG 1482, Revision 1, guidance, based upon:

1. Reduction of unnecessary cycling of equipment (i.e., RWCU pumps, demineralizer beds, related system valves)

2. Reduction of the life expectancy of the RWCU components from potential transients Page 50 of 56

Monticello Nuclear Generating Plant Revision 0

3. Could result in an unnecessary plant shutdown due to water chemistry transients if a valve does not reopen after tested to its close safety position.

4. Could contribute to additional personnel dose if an unplanned RWCU shutdown were to occur Alternate Test Frequency: MO-2397 and MO-2398 full stroke quarterly exercise testing is deferred to refueling frequency IAW ISTC-3521(e).

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Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 42 System: Residual Heat Removal (RHR)

Core Spray (CSP)

Valve(s): MO-2014 and MO-2015 (RHR)

MO-1753 and MO-1754 (CSP)

P & ID: NH-36247 (M-121): MO-2014 NH-36246 (M-120): MO-2015 NH-36248 (M-122): MO-1753 NH-36248 (M-122): MO-1754 Code Class: 1 Category: A Code Frequency: ISTC-3510, Active Category A valves shall be exercised nominally every 3 months.

Function/Description: These valves are normally closed and are located in the RHR and CSP injection lines to the respective reactor recirculation loops (RHR) and the in-vessel core spray spargers (CSP). These injection pathways supply low pressure Emergency Core Cooling System (ECCS) makeup during accident conditions, as well as providing shutdown cooling flow (RHR) and vessel floodup (CSP) capabilities during non-shutdown conditions.

The RHR valves receive an auto open signal as sent by as directed by LPCI-Loop selection logic (RHR). The ECCS System logic will open the selected loops LPCI valves and the CSP valves automatically when an initiation signal is present with reactor vessel pressure < 460 psig.

These valves must remain closed to provide containment isolation during emergency and accident conditions as required by System functional logic.

The RHR valves receive a Group 2 isolation signal to close when required. MO-2014 and MO-2015 are Automatic Containment Isolation valves for penetrations X-13B and X-13A, respectively (USAR Table 5.2-3a and 5.2-3b]. Core Spray valves MO-1753 and MO-1754 are considered Containment isolation valves for penetrations X-16B and X-16A, respectively (USAR Table 5.2-3a).

All four valves are specified as Pressure Isolation Valves (PIV) per MNGP commitments made to Generic Letter 87-06, Periodic Verification of Leak Tight Integrity of Pressure Isolation Valves.

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Monticello Nuclear Generating Plant Revision 0 Frequency Change Justification: NUREG-1482, Revision 1, "Guidelines for Inservice Testing at Nuclear Power Plants", Section 2.4.5, "Deferring Valve Testing to Cold Shutdown or Refueling Outages" identifies impractical conditions justifying test deferrals as those conditions that could result in unnecessary challenges to safety systems, place undue stress on components, cause unnecessary cycling of equipment, or unnecessarily reduce the life expectancy of the plant systems and components.

Basis for not performing Quarterly frequency full stroke exercising Each of these PIV MOVs are located in-series with PIV check valves located inside containment.

Quarterly exercise testing of these MOVs is performed at a nominal 1000 psig and results in a loss of one of the two in-series PIV valves designated to mitigate high-low pressure intersystem LOCAs.

Further, an undetected failure of one of the PIV check valves inside containment would result in a loss of both PIV barriers during Quarterly testing. In this condition, the probability of causing a high to low pressure inter-system LOCA would be increased with the following adverse consequences:

1. Placing undue stress on components.

2. Unnecessarily reducing the life expectancy of the system components due to over pressurization.

3. Setting up conditions for unnecessary cycling of system relief valves.

Therefore, the safest position for the PIV MOVs to be positioned in during reactor power conditions is their ECCS standby-readiness closed position.

Basis for not performing Quarterly frequency part stroke exercising Based on design, these valves cannot be partially stroked.

Alternate Test Frequency: MO-2014, MO-2015, MO-1753 and MO-1754 full stroke quarterly exercise testing is deferred to cold shutdown frequency IAW ISTC-3521(c). For MO-2014 and MO-2015, if Shutdown Cooling mode is not swapped during a given cold shutdown, only one of these valves would be required to be eligible for testing in that specific cold shutdown other than refueling.

As an alternative to stroke time testing MO-1753 and MO-1754 during cold shutdowns, the valves will be stroke timed at the frequency specified in the MOV Program for diagnostic testing. This proposed alternative is pursuant to the guidelines provided in NUREG-1482, Revision 1, Section 4.2.5, Alternatives to Stroke Time Testing and Valve Relief Request VR 03, Use of Code Case OMN-1, Revision 1, on Various Motor Operated Valves. The valves will continue to be exercised (non-timed) during cold shutdowns.

Page 53 of 56

Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 43 System: EDG Emergency Service Water (ESW)

Valve(s): ESW-1-1 and ESW-1-2 P & ID: NH-36665 (M-811)

Code Class: 3 Category: C Code Frequency: ISTC-3510, Active Category C check valves shall be exercised nominally every 3 months.

Function/Description: The EDG-ESW pump discharge check valves open to provide the operating pump a flow path to the emergency diesel generator heat exchanger. ESW-1-1 and ESW-1-2 will close to prevent flow diversion through the idle EDG-ESW pump during operation of the loop A/B crosstie and opposite loops EDG-ESW pump.

Frequency Change Justification: It is not practical to exercise the EDG-ESW pump discharge check valves closed every 3 months per the requirements of ISTC-3510. It requires opening normally closed isolation valves (SW-229/228), puts the plant in an abnormal configuration (crosstie open), and introduces the possibility of MIC into the associated piping.

There exists the concern that the EDG-ESW pump degradation will occur based on the potential for the EDG-ESW pumps to be deadheaded or at zero flow when operated with the crosstie open. Johnston Pump representatives provided an estimate that rubber seal degradation would be probable after a brief period of operation at zero flow (10 minutes or less). Therefore, disassembly and examination shall be used to meet Code exercising requirements.

Alternate Test Frequency: These valves will be grouped and tested in accordance with ISTC-5221(c). At least one valve from the group will be disassembled and examined each refueling outage for the open and close exercise test. As an alternative to the disassembly and examination requirements of ISTC-5221(c), these check valves may be placed in the Check Valve Condition Monitoring (CVCM) Program and disassembled and examined per the requirements of Appendix II of the OM Code. If so, the frequency for disassembly and examination shall be specified in the CVCM Program Plan applicable to the valve Group.

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Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 44 (DELETED)

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Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 45 (DELETED)

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Monticello Nuclear Generating Plant Revision 0 ATTACHMENT 7 Code Cases Invoked by IST Program Code Rev. Title OM Conditions Case No. Edition/Addenda Applied by Applicability Reg. Guide 1.192?

Use of a Pump Curve for Testing OM Code-1998 No OMN-16 0 (as stated in Pump Relief Request PR 06) through OMb-2006 (Not in R.G.)

Alternate Rules for Preservice and Inservice Testing of Active Electric OM Code 1995 OMN-1 1 Motor Operated Valve Assemblies in Yes1 through OMb-2000 Light-Water Reactor Power Plants (as stated in Valve Relief Request VR 03)

Alternative Rules for Testing ASME OM Code 1995 No OMN-17 0 Class 1 Pressure Relief/Safety Valves through OMb-2006 (Not in R.G.)

(as stated in Valve Relief Request VR 04)

Notes:

1. Code Case OMN-1, Revision 0 has been conditionally accepted by R.G. 1.192. Revision 1 of the Code Case is not specified in R.G. 1.192; however, the conditions of acceptance for Revision 0, as specified in the Reg. Guide, shall be considered applicable to Revision 1 of the Code Case.

Page 1 of 1

Monticello Nuclear Generating Plant Revision 0 ATTACHMENT 8 Pump Tables Page 1 of 5

Monticello Nuclear Generating Plant Inservice Testing Program - Pumps Component Description PID Class ASME Group Type Test Frequency RR P-109A 11 Residual Heat Removal Service NH-36665 3 A Vertical Q Q/Y2 PR 02 Water Pump M-811 DP Q/Y2 N/A V Q/Y2 N/A P-109B 12 Residual Heat Removal Service NH-36665 3 A Vertical Q Q/Y2 PR 02 Water Pump M-811 DP Q/Y2 N/A V Q/Y2 N/A P-109C 13 Residual Heat Removal Service NH-36665 3 A Vertical Q Q/Y2 PR 02 Water Pump M-811 DP Q/Y2 N/A V Q/Y2 N/A P-109D 14 Residual Heat Removal Service NH-36665 3 A Vertical Q Q/Y2 PR 02 Water Pump M-811 DP Q/Y2 N/A V Q/Y2 N/A Pump Table Page 2 of 5

Component Description PID Class ASME Group Type Test Frequency RR P-111A 11 Emergency Service Water Pump NH-36665 3 B Vertical Q Y2 N/A M-811 DP Q/Y2 N/A V Y2 N/A P-111B 12 Emergency Service Water Pump NH-36665 3 B Vertical Q Y2 N/A M-811 DP Q/Y2 N/A V Y2 N/A P-111C 13 Emergency Service Water Pump NH-36665 3 B Vertical Q Y2 N/A M-811 DP Q/Y2 N/A V Y2 N/A P-111D 14 Emergency Service Water Pump NH-36665 3 B Vertical Q Y2 N/A M-811 DP Q/Y2 N/A V Y2 N/A P-202A 11 Residual Heat Removal Pump NH-36247 2 A Centrifugal DP Q/Y2 N/A M-121 Q Q/Y2 PR 02 V Q/Y2 N/A Pump Table Page 3 of 5

Component Description PID Class ASME Group Type Test Frequency RR P-202B 12 Residual Heat Removal Pump NH-36246 2 A Centrifugal DP Q/Y2 N/A M-120 Q Q/Y2 PR 02 V Q/Y2 N/A P-202C 13 Residual Heat Removal Pump NH-36247 2 A Centrifugal DP Q/Y2 N/A M-121 Q Q/Y2 PR 02 V Q/Y2 N/A P-202D 14 Residual Heat Removal Pump NH-36246 2 A Centrifugal DP Q/Y2 N/A M-120 Q Q/Y2 PR 02 V Q/Y2 N/A P-203A 11 Standby Liquid Control Pump NH-36253 2 B Positive Q Q/Y2 PR 01 M-127 Displacement P Y2 N/A V Y2 PR 05 P-203B 12 Standby Liquid Control Pump NH-36253 2 B Positive Q Q/Y2 PR 01 M-127 Displacement P Y2 N/A V Y2 PR 05 Pump Table Page 4 of 5

Component Description PID Class ASME Group Type Test Frequency RR P-207 Reactor Core Isolation Cooling Pump NH-36252 2 B Centrifugal Q Y2 N/A M-126 DP Q/Y2 PR 04 V Y2 N/A N Q/Y2 N/A P-208A 11 Core Spray Pump NH-36248 2 B Centrifugal DP Q/Y2 N/A M-122 Q Y2 N/A V Y2 N/A P-208B 12 Core Spray Pump NH-36248 2 B Centrifugal DP Q/Y2 N/A M-122 Q Y2 N/A V Y2 N/A P-209 High Pressure Coolant Injection NH-36250 2 B Centrifugal Q Y2 N/A Pump M-124 DP Q/Y2 PR 04, PR 06 V Y2 PR 03 N Q/Y2 N/A Pump Table Page 5 of 5

Energy Monticello Nuclear Generating Plant Revision 0 I ATTACHMENT 9 Valve Tables Page 1 of 81 I

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Alternate N2 Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AI-598 MSIV/SRV Alternate N2 NH-36049-10 B-5 1.0 CK SA 2 AC O O/C A DI CM DTJ 19 Supply Inboard Containment M-131-10 Isolation Check Valve LT AJ N/A AI-599 SRV Pneumatic Supply NH-36049-10 D-5 1.0 CK SA 2 AC O O/C A CTO Q N/A Inboard Containment M-131-10 Isolation Check Valve CTC Q N/A LT AJ N/A AI-700 SRV Pneumatic Supply NH-36049-10 D-5 1.0 CK SA 2 AC O O/C A CTO Q N/A Outboard Containment M-131-10 Isolation Check Valve CTC Q N/A LT AJ N/A AI-705 Instrument Air/Nitrogen NH-36049-10 D-6 1.0 CK SA NC AC O C A CTC Q N/A Supply to SRV's Check Valve M-131-10 LT Y2 N/A CTO Q N/A AI-706 Instrument Air/Nitrogen NH-36049-10 D-6 1.0 CK SA NC AC O C A CTC Q N/A Supply to SRV's Check Valve M-131-10 LT Y2 N/A CTO Q N/A AI-708 MSIV/SRV Alternate N2 NH-36049-10 B-5 1.0 CK SA 2 AC O O/C A DI CM DTJ 19 Supply Outboard M-131-10 Containment Isolation Check LT AJ N/A Valve AI-713 Instrument Air/Nitrogen NH-36049-10 B-6 1.0 CK SA NC AC O C A CTC Q N/A Supply to MSIV's/SRV's M-131-10 Check Valve LT Y2 N/A CTO Q N/A AI-714 Instrument Air/Nitrogen NH-36049-10 B-6 1.0 CK SA NC AC O C A CTC Q N/A Supply to MSIV's/SRV's M-131-10 Check Valve LT Y2 N/A CTO Q N/A Valve Table Page 2 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Alternate N2 Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AI-729 Alternate Nitrogen Train A NH-36049-10 D-7 0.75 CK SA NC C O/C O A CTO Q N/A Check Valve M-131-10 CTC CS DTJ 19 AI-730 Alternate Nitrogen Train B NH-36049-10 B-7 0.75 CK SA NC C O/C O A CTO CS DTJ 19 Check Valve M-131-10 CTC CS DTJ 19 RV-4236 Train B Alternate Nitrogen NH-36049-10 B-7 1.5 RV SA NC C C O/C A RT SY10 N/A Relief Valve M-131-10 RV-4673 Train A Alternate Nitrogen NH-36049-10 D-7 1.5 RV SA NC C C O/C A RT SY10 N/A Relief Valve M-131-10 RV-4878 Train A Instrument NH-36049-10 D-5 1.5 RV SA NC C C O/C A RT SY10 N/A Air/Alternate Nitrogen Relief M-131-10 Valve RV-4880 Train B Instrument NH-36049-10 B-5 1.5 RV SA NC C C O/C A RT SY10 N/A Air/Alternate Nitrogen Relief M-131-10 Valve SV-4234 Train A Alternate Nitrogen NH-36049-10 D-6 1.0 GL SO NC B O O/C A STO Q N/A Supply to SRV's Isolation M-131-10 Valve PIT Y2 N/A STC Q N/A FO Q N/A Valve Table Page 3 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Alternate N2 Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ SV-4235 Train B Alternate Nitrogen NH-36049-10 B-6 1.0 GL SO NC B O O/C A STO CS DTJ 19 Supply to SRV's/MSIV's M-131-10 Isolation Valve PIT Y2 N/A STC CS DTJ 19 FO CS DTJ 19 Valve Table Page 4 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Auto Pressure Relief Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AI-13-4 Instrument Air Supply Check NH-36049-12 C-3 1.0 CK SA NC AC O/C C A CTC RF DTJ 25 Valve to RV 2-71D M-131-12 LT Y2 N/A CTO RF DTJ 25 AI-13-7 Instrument Air Supply Check NH-36049-12 B-3 1.0 CK SA NC AC O/C C A CTC RF DTJ 25 Valve to RV 2-71G M-131-12 LT Y2 N/A CTO RF DTJ 25 RV-2-71A Main Steam Safety/Relief NH-36241 B-4 6.00 RV AO/SA 1 BC C O/C A RT SY6 VR 04 Valve (ADS) M-115 STO SY6 VR 04 STC SY6 VR 04 RV-2-71B Main Steam Safety/Relief NH-36241 D-4 6.00 RV AO/SA 1 BC C O/C A RT SY6 VR 04 Valve M-115 STO SY6 VR 04 STC SY6 VR 04 RV-2-71C Main Steam Safety/Relief NH-36241 D-3 6.00 RV AO/SA 1 BC C O/C A RT SY6 VR 04 Valve (ADS) M-115 STO SY6 VR 04 STC SY6 VR 04 RV-2-71D Main Steam Safety/Relief NH-36241 B-3 6.00 RV AO/SA 1 BC C O/C A RT SY6 VR 04 Valve (ADS) M-115 STO SY6 VR 04 STC SY6 VR 04 RV-2-71E Main Steam Safety/Relief NH-36241 B-4 6.00 RV AO/SA 1 BC C O/C A RT SY6 VR 04 Valve (Low-Low Set) M-115 STO SY6 VR 04 STC SY6 VR 04 RV-2-71F Main Steam Safety/Relief NH-36241 B-3 6.00 RV AO/SA 1 BC C O/C A RT SY6 VR 04 Valve M-115 STO SY6 VR 04 STC SY6 VR 04 Valve Table Page 5 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Auto Pressure Relief Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ RV-2-71G Main Steam Safety/Relief NH-36241 D-4 6.00 RV AO/SA 1 BC C O/C A RT SY6 VR 04 Valve (Low-Low Set) M-115 STO SY6 VR 04 STC SY6 VR 04 RV-2-71H Main Steam Safety/Relief NH-36241 D-3 6.00 RV AO/SA 1 BC C O/C A RT SY6 VR 04 Valve (Low-Low Set) M-115 STO SY6 VR 04 STC SY6 VR 04 RV-3242A Safety Relief Valve RV-2 NH-36241-1 B-5 8.0 RV SA NC C C O/C A RT SY10 N/A A Discharge Vacuum M-115-1 Breaker RV-3243A Safety Relief Valve RV-2 NH-36241-1 C-6 8.0 RV SA NC C C O/C A RT SY10 N/A B Discharge Vacuum M-115-1 Breaker RV-3244A Safety Relief Valve RV-2 NH-36241-1 C-4 8.0 RV SA NC C C O/C A RT SY10 N/A C Discharge Vacuum M-115-1 Breaker RV-3245A Safety Relief Valve RV-2 NH-36241-1 A-4 8.0 RV SA NC C C O/C A RT SY10 N/A D Discharge Vacuum M-115-1 Breaker RV-7440A Safety Relief Valve RV-2 NH-36241-1 A-6 8.0 RV SA NC C C O/C A RT SY10 N/A E Discharge Vacuum M-115-1 Breaker RV-7441A Safety Relief Valve RV-2 NH-36241-1 A-4 8.0 RV SA NC C C O/C A RT SY10 N/A F Discharge Vacuum Breaker M-115-1 Valve Table Page 6 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Auto Pressure Relief Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ RV-7467A Safety Relief Valve RV-2 NH-36241-1 C-5 8.0 RV SA NC C C O/C A RT SY10 N/A G Discharge Vacuum M-115-1 Breaker RV-7468A Safety Relief Valve RV-2 NH-36241-1 C-4 8.0 RV SA NC C C O/C A RT SY10 N/A H Discharge Vacuum M-115-1 Breaker XFV-10 Safety Relief Valve NH-36241-1 C-7 1.00 FV SA 2 AC O C A XT RF DTJ 22 Discharge Line Excess Flow M-115-1 Check Valve XFV-11 Safety Relief Valve NH-36241-1 C-7 1.00 FV SA 2 AC O C A XT RF DTJ 22 Discharge Line Excess Flow M-115-1 Check Valve XFV-12 Safety Relief Valve NH-36241-1 A-7 1.00 FV SA 2 AC O C A XT RF DTJ 22 Discharge Line Excess Flow M-115-1 Check Valve XFV-13 Safety Relief Valve NH-36241-1 A-7 1.00 FV SA 2 AC O C A XT RF DTJ 22 Discharge Line Excess Flow M-115-1 Check Valve XFV-14 Safety Relief Valve NH-36241-1 C-3 1.00 FV SA 2 AC O C A XT RF DTJ 22 Discharge Line Excess Flow M-115-1 Check Valve XFV-15 Safety Relief Valve NH-36241-1 C-3 1.00 FV SA 2 AC O C A XT RF DTJ 22 Discharge Line Excess Flow M-115-1 Check Valve Valve Table Page 7 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Auto Pressure Relief Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ XFV-16 Safety Relief Valve NH-36241-1 A-3 1.00 FV SA 2 AC O C A XT RF DTJ 22 Discharge Line Excess Flow M-115-1 Check Valve XFV-17 Safety Relief Valve NH-36241-1 A-3 1.00 FV SA 2 AC O C A XT RF DTJ 22 Discharge Line Excess Flow M-115-1 Check Valve XFV-18 Safety Relief Valve NH-36241-1 A-3 1.00 FV SA 2 AC O C A XT RF DTJ 22 Discharge Line Excess Flow M-115-1 Check Valve XFV-19 Safety Relief Valve NH-36241-1 C-3 1.00 FV SA 2 AC O C A XT RF DTJ 22 Discharge Line Excess Flow M-115-1 Check Valve XFV-20 Safety Relief Valve NH-36241-1 A-7 1.00 FV SA 2 AC O C A XT RF DTJ 22 Discharge Line Excess Flow M-115-1 Check Valve XFV-21 Safety Relief Valve NH-36241-1 C-7 1.00 FV SA 2 AC O C A XT RF DTJ 22 Discharge Line Excess Flow M-115-1 Check Valve Valve Table Page 8 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Condensate and Feedwater Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ FW-91-1 Feedwater Line A Outboard NH-36241 A-3 14.00 CK SA 2 AC O C A CTC CS DTJ 12 Check Valve M-115 CTO Q N/A LT Y2 N/A FW-91-2 Feedwater Line B Outboard NH-36241 A-4 14.00 CK SA 2 AC O C A CTC CS DTJ 12 Check Valve M-115 CTO Q N/A LT Y2 N/A FW-94-1 Feedwater Line A Outboard NH-36241 A-3 14.00 CK SA 1 AC O O/C A CTO Q N/A Containment Isolation Check M-115 Valve CTC CS DTJ 12 LT AJ N/A FW-94-2 Feedwater Line B Outboard NH-36241 A-4 14.00 CK SA 1 AC O O/C A CTO Q N/A Containment Isolation Check M-115 Valve CTC CS DTJ 12 LT AJ N/A FW-97-1 Feedwater Line A Inboard NH-36241 A-3 14.00 CK SA 1 AC O O/C A CTO Q N/A Containment Isolation Check M-115 Valve CTC CS DTJ 12 LT AJ N/A FW-97-2 Feedwater Line B Inboard NH-36241 A-4 14.00 CK SA 1 AC O O/C A CTO Q N/A Containment Isolation Check M-115 Valve CTC CS DTJ 12 LT AJ N/A Valve Table Page 9 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Condensate Storage Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ CST-103-1 CST Check Valve to Core NH-36248 E-2 2.0 CK SA NC C O C A CTC Q N/A Spray Loop A M-122 CTO Q N/A CST-104-1 CST Check Valve to Core NH-36248 E-2 2.0 CK SA 2 C O C A CTC Q N/A Spray Loop A M-122 CTO Q N/A CST-189 CST Check Valve to RHR NH-85509 B-6 1.0 CK SA 2 C O C A CTO CS DTJ 36 System Flush to Reactor M-114-1 Vessel Head Spray CTC Q N/A CST-88 CST Check Valve to LPCI B NH-85509 B-3 2.0 CK SA 2 C O C A DI CM DTJ 02 Injection M-114-1 CST-90 CST Check Valve to RHR NH-85509 B-4 2.0 CK SA 2 C O C A DI CM DTJ 02 Suction Cross Tie Keepfill. M-114-1 CST-92 CST Check Valve to LPCI NH-85509 B-4 2.0 CK SA 2 C O C A DI CM DTJ 02 "A" Injection M-114-1 CST-94 CST Check Valve to NH-85509 B-4 2.0 CK SA 2 C O C A DI CM DTJ 02 Containment Spray Loop A M-114-1 CST-96 CST Check Valve to Core NH-85509 B-5 2.0 CK SA 2 C O C A CTC Q N/A Spray Loop B M-114-1 CTO Q N/A Valve Table Page 10 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Condensate Storage Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ CST-98 CST Check Valve to NH-85509 B-5 2.0 CK SA 2 C O C A CTC Q N/A Containment Spray Loop B M-114-1 CTO CS DTJ 36 Valve Table Page 11 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Control Rod Drive Hydraulics Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ CRD-114* CRD-Scram Discharge NH-36245 A-6 0.75 CK SA 2 C O/C O A CTO TS DTJ 31 Header Check Valve M-119 (Typical for 121 Drive Units) CTC N/A VR 01 CRD-115* CRD-Accumulator Charging NH-36245 B-4 0.50 CK SA 2 C O/C C A CTC RF DTJ 05 Water Header Check Valve M-119 (Typical for 121 Drive Units) CTO TS DTJ 31 CRD-138* CRD-Cooling Water Check NH-36245 D-4 0.50 CK SA 2 C O C A CTC TS DTJ 31 Valve (Typical for 121 Drive M-119 Units) CTO Q N/A CV-126* CRD-SCRAM Inlet Valve NH-36245 B-5 1.0 GL AO 2 B C O A STO TS DTJ 31 (Typical for 121 Drive Units) M-119 FO TS DTJ 31 CV-127* CRD-SCRAM Outlet Valve NH-36245 B-6 0.75 GL AO 2 B C O A STO TS DTJ 31 (Typical for 121 Drive Units) M-119 FO TS DTJ 31 CV-3-32A CRD-Scram Discharge NH-36245 D-4 1.0 GL AO 2 B O C A STC Q N/A Volume Vent Isolation Valve M-119 FC Q N/A PIT Y2 N/A STC TS DTJ 38 CV-3-32B CRD-Scram Discharge NH-36245 D-1 1.0 GL AO 2 B O C A STC Q N/A Volume Vent Isolation Valve M-119 FC Q N/A PIT Y2 N/A STC TS DTJ 38 Valve Table Page 12 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Control Rod Drive Hydraulics Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ CV-3-32C CRD-Scram Discharge NH-36245 D-4 1.0 GL AO 2 B O C A STC Q N/A Volume Vent Isolation Valve M-119 FC Q N/A PIT Y2 N/A STC TS DTJ 38 CV-3-32D CRD-Scram Discharge NH-36245 D-1 1.0 GL AO 2 B O C A STC Q N/A Volume Vent Isolation Valve M-119 FC Q N/A PIT Y2 N/A STC TS DTJ 38 CV-3-33A CRD-Scram Discharge NH-36245 C-3 2.0 GL AO 2 B O C A STC Q N/A Volume Drain Isolation Valve M-119 FC Q N/A PIT Y2 N/A STC TS DTJ-38 CV-3-33B CRD-Scram Discharge NH-36245 C-2 2.0 GL AO 2 B O C A STC Q N/A Volume Drain Isolation Valve M-119 FC Q N/A PIT Y2 N/A STC TS DTJ 38 CV-3-33C CRD-Scram Discharge NH-36245 C-3 2.0 GL AO 2 B O C A STC Q N/A Volume Drain Isolation Valve M-119 FC Q N/A PIT Y2 N/A STC TS DTJ 38 CV-3-33D CRD-Scram Discharge NH-36245 C-2 2.0 GL AO 2 B O C A STC Q N/A Volume Drain Isolation Valve M-119 FC Q N/A PIT Y2 N/A STC TS DTJ 38 Valve Table Page 13 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Core Spray Cooling Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AO-14-13A 11 CS Injection Air-Operated NH-36248 D-3 8.0 CK SA 1 AC C O/C A CTO RF DTJ 27 Testable Check Valve M-122 CTC RF DTJ 27 LT Y2 N/A PIT Y2 N/A AO-14-13B 12 CS Injection Air-Operated NH-36248 D-4 8.0 CK SA 1 AC C O/C A CTO RF DTJ 27 Testable Check Valve M-122 CTC RF DTJ 27 LT Y2 N/A PIT Y2 N/A CS-13-1 CS A Loop Manual Isolation NH-36248 D-3 8.0 GT MA 1 B O O P PIT Y2 N/A Valve to Reactor Vessel M-122 CS-13-2 CS B Loop Manual Isolation NH-36248 D-4 8.0 GT MA 1 B O O P PIT Y2 N/A Valve to Reactor Vessel M-122 CS-9-1 CS Pump P-208A Discharge NH-36248 B-2 10.0 CK SA 2 C C O/C A CTO Q N/A Check Valve M-122 CTC Q N/A CS-9-2 CS Pump P-208B Discharge NH-36248 B-5 10.0 CK SA 2 C C O/C A CTO Q N/A Check Valve M-122 CTC Q N/A MO-1741 CS Pump P-208A Motor NH-36248 A-3 12.0 GT MO 2 B O O P PIT Y2 N/A Operated Torus Suction M-122 Valve Valve Table Page 14 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Core Spray Cooling Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ MO-1742 CS Pump P-208B Motor NH-36248 A-4 12.0 GT MO 2 B O O P PIT Y2 N/A Operated Torus Suction M-122 Valve MO-1749 CS System A Loop Motor NH-36248 C-2 6.0 GL MO 2 B C C P PIT Y2 N/A Operated Test Return Valve M-122 MO-1750 CS System B Loop Motor NH-36248 C-5 6.0 GL MO 2 B C C P PIT Y2 N/A Operated Test Return Valve M-122 MO-1751 11 CS System Motor NH-36248 D-2 8.0 GT MO 2 A O O/C A LT AJ N/A Operated Outboard Injection M-122 Valve STO PV VR 03 STC PV VR 03 PIT Y2 N/A ET Q N/A MO-1752 12 CS System Motor NH-36248 B-5 8.0 GT MO 2 A O O/C A LT AJ N/A Operated Outboard Injection M-122 Valve STO PV VR 03 STC PV VR 03 PIT Y2 N/A ET Q N/A MO-1753 11 CS System Motor NH-36248 D-3 8.0 GT MO 1 A C O/C A LT AJ N/A Operated Inboard Injection M-122 Valve LT Y2 N/A STO PV VR 03 STC PV VR 03 PIT Y2 N/A Valve Table Page 15 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Core Spray Cooling Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ MO-1754 12 CS System Motor NH-36248 D-5 8.0 GT MO 1 A C O/C A LT AJ N/A Operated Inboard Injection M-122 Valve LT Y2 N/A STO PV VR 03 STC PV VR 03 PIT Y2 N/A RV-1745 11 Core Spray System Relief NH-36248 D-2 2.0 RV SA 2 C C O/C A RT SY10 N/A Valve M-122 RV-1746 12 Core Spray System Relief NH-36248 D-5 2.0 RV SA 2 C C O/C A RT SY10 N/A Valve M-122 XFV-82 11 Core Spray System NH-36248 C-3 1.0 FV SA 1 AC O C A XT RF DTJ 22 Excess Flow Check Valve M-122 XFV-83 12 Core Spray System NH-36248 C-4 1.0 FV SA 1 AC O C A XT RF DTJ 22 Excess Flow Check Valve M-122 Valve Table Page 16 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Demineralized Water System Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ DM-151 DM Outboard Containment NH-36039 E-1 1.0 GT MA 2 A C C P LT AJ N/A Isolation Valve M-108 DM-152 DM Inboard Containment NH-36039 E-1 1.0 GT MA 2 A C C P LT AJ N/A Isolation Valve M-108 Valve Table Page 17 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Diesel Generators Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ GSA-32-1 EDG Starting Air Dryer NH-36051 B-2 0.75 CK SA NC C O/C C A CTC Q N/A Discharge Check Valve M-133 CTO Q N/A GSA-32-2 EDG Starting Air Dryer NH-36051 A-2 0.75 CK SA NC C O/C C A CTC Q N/A Discharge Check Valve M-133 CTO Q N/A GSA-32-3 EDG Starting Air Dryer NH-36051 E-2 0.75 CK SA NC C O/C C A CTC Q N/A Discharge Check Valve M-133 CTO Q N/A GSA-32-4 EDG Starting Air Dryer NH-36051 D-2 0.75 CK SA NC C O/C C A CTC Q N/A Discharge Check Valve M-133 CTO Q N/A RV-3216 EDG Starting Air Receiver NH-36051 B-3 0.5 RV SA NC C C O/C A RT SY10 N/A Tank T-79A Relief Valve M-133 RV-3217 EDG Starting Air Receiver NH-36051 B-3 0.5 RV SA NC C C O/C A RT SY10 N/A Tank T-79B Relief Valve M-133 RV-3218 EDG Starting Air Receiver NH-36051 B-4 0.5 RV SA NC C C O/C A RT SY10 N/A Tank T-79C Relief Valve M-133 RV-3219 EDG Starting Air Receiver NH-36051 A-3 0.5 RV SA NC C C O/C A RT SY10 N/A Tank T-79D Relief Valve M-133 Valve Table Page 18 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Diesel Generators Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ RV-3220 EDG Starting Air Receiver NH-36051 A-3 0.5 RV SA NC C C O/C A RT SY10 N/A Tank T-79E Relief Valve M-133 RV-3221 EDG Starting Air Receiver NH-36051 A-4 0.5 RV SA NC C C O/C A RT SY10 N/A Tank T-79F Relief Valve M-133 RV-3224 EDG Starting Air Receiver NH-36051 E-3 0.5 RV SA NC C C O/C A RT SY10 N/A Tank T-80A Relief Valve M-133 RV-3225 EDG Starting Air Receiver NH-36051 E-3 0.5 RV SA NC C C O/C A RT SY10 N/A Tank T-80B Relief Valve M-133 RV-3226 EDG Starting Air Receiver NH-36051 E-4 0.5 RV SA NC C C O/C A RT SY10 N/A Tank T-80C Relief Valve M-133 RV-3227 EDG Starting Air Receiver NH-36051 D-3 0.5 RV SA NC C C O/C A RT SY10 N/A Tank T-80D Relief Valve M-133 RV-3228 EDG Starting Air Receiver NH-36051 D-3 0.5 RV SA NC C C O/C A RT SY10 N/A Tank T-80E Relief Valve M-133 RV-3229 EDG Starting Air Receiver NH-36051 D-4 0.5 RV SA NC C C O/C A RT SY10 N/A Tank T-80F Relief Valve M-133 Valve Table Page 19 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Diesel Oil Storage Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ RV-1523 Diesel Fuel Oil Transfer NH-36051 D-3 0.75 RV SA NC C C O/C A RT SY10 N/A Pump Relief Valve M-133 Valve Table Page 20 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System EDG Emergency Service Water Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AV-3155 Emergency Service Water NH-36665 B-4 2.0 AR SA 3 C O/C O/C A CTC Q N/A Pump P-111A Air Vent Valve M-811 CTO Q N/A AV-3156 Emergency Service Water NH-36665 B-6 2.0 AR SA 3 C O/C O/C A CTC Q N/A Pump P-111B Air Vent Valve M-811 CTO Q N/A ESW-1-1 Emergency Service Water NH-36665 B-4 4.0 CK SA 3 C O/C O A DI CM DTJ 43 Pump P-111A Discharge M-811 Check Valve ESW-1-2 Emergency Service Water NH-36665 B-8 4.0 CK SA 3 C O/C O A DI CM DTJ 43 Pump P-111B Discharge M-811 Check Valve ESW-3-1 Emergency Service Water NH-36665 C-4 4.0 GT MA 3 B O/C O/C A ET Y2 N/A Basket Strainer Bypass M-811 Isolation Valve ESW-3-2 Emergency Service Water NH-36665 C-6 4.0 GT MA 3 B O/C O/C A ET Y2 N/A Basket Strainer Bypass M-811 Isolation Valve ESW-71-1 11 ESW Biocide Injection NH-36665 B-4 0.5 CK SA 3 C O/C C A CTC Q N/A Check Valve M-811 CTO Q N/A ESW-71-2 12 ESW Biocide Injection NH-36665 B-6 0.5 CK SA 3 C O/C C A CTC Q N/A Check Valve M-811 CTO Q N/A Valve Table Page 21 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System EDG Emergency Service Water Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ ESW-73-1 11 ESW Dispersant Injection NH-36665 B-4 0.5 CK SA 3 C O/C C A CTC Q N/A Check Valve M-811 CTO Q N/A ESW-73-2 12 ESW Dispersant Injection NH-36665 B-6 0.5 CK SA 3 C O/C C A CTC Q N/A Check Valve M-811 CTO Q N/A Valve Table Page 22 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System EFT Emergency Service Water Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AV-4024 Emergency Service Water NH-36665 C-3 1.0 AR SA 3 C O/C O/C A CTC Q N/A Pump P-111C Air Vent Valve M-811 CTO Q N/A AV-4026 Emergency Service Water NH-36665 B-5 1.0 AR SA 3 C O/C O/C A CTC Q N/A Pump P-111D Air Vent Valve M-811 CTO Q N/A ESW-13 Service Water to Emergency NH-36665 C-5 4.0 CK SA NC C O C A CTC Q N/A Service Water Check Valve M-811 CTO Q N/A ESW-14 Service Water to Emergency NH-36665 C-6 4.0 CK SA 3 C O C A CTC Q N/A Service Water Check Valve M-811 CTO Q N/A ESW-15 Service Water to Emergency NH-36665 D-4 4.0 CK SA NC C O C A CTC Q N/A Service Water Check Valve M-811 CTO Q N/A ESW-16 Service Water to Emergency NH-36665 D-4 4.0 CK SA 3 C O C A CTC Q N/A Service Water Check Valve M-811 CTO Q N/A ESW-17 Emergency Service Water NH-36665 C-6 4.0 CK SA 3 C O/C O A CTO Q N/A Pump P-111D Discharge M-811 Check Valve CTC Q N/A ESW-18 Emergency Service Water NH-36665 C-4 4.0 CK SA 3 C O/C O A CTO Q N/A Pump P-111C Discharge M-811 Check Valve CTC Q N/A Valve Table Page 23 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System EFT Emergency Service Water Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ ESW-19 Manual Bypass Valve for NH-36665 C-6 4.0 GT MA 3 B O/C O/C A ET Y2 N/A Strainer YS-111D M-811 ESW-20 Manual Bypass Valve for NH-36665 C-4 3.0 GT MA 3 B O/C O/C A ET Y2 N/A Strainer YS-111C M-811 Valve Table Page 24 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Fuel Pool Cooling and Cleanup Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ PC-20-1 Spent Fuel Pool Return Line NH-36256 D-4 6.0 CK SA NC C O C A DI CM DTJ 32 Anti-Siphon Check Valve M-135 PC-20-2 Spent Fuel Pool Return Line NH-36256 D-4 6.0 CK SA NC C O C A DI CM DTJ 32 Anti-Siphon Check Valve M-135 Valve Table Page 25 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System High Pressure Coolant Injection Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AO-23-18 HPCI Injection Check Valve NH-36250 C-5 12.0 CK SA 2 C C O/C A DI CM DTJ 15 M-124 PIT Y2 N/A CV-2046A HPCI Steam Line Drain NH-36249 C-1 1.0 GL AO 2 B O C A STC Q N/A Isolation Valve M-123 FC Q N/A PIT Y2 N/A CV-2065 HPCI Pump Minimum Flow NH-36250 C-4 2.0 GL AO 2 B C O/C A STO Q N/A Control Valve M-124 STC Q N/A FO Q N/A PIT Y2 N/A HPCI-10 HPCI Turbine Exhaust NH-36249 C-5 16.0 CK SA 2 AC O/C O/C A CTO Q N/A Check Valve M-123 CTC CM DTJ 17 LT AJ N/A HPCI-103 HPCI Keep Fill Check Valve NH-36250 B-5 0.5 CK SA 2 C O C A CTO Q N/A M-124 CTC Q N/A HPCI-105 HPCI Keep Fill Check Valve NH-36250 B-5 0.5 CK SA 2 C O C A CTO Q N/A M-124 CTC Q N/A HPCI-14 HPCI Turbine Steam Trap to NH-36249 B-5 2.0 CK SA 2 C O/C C A CTC CS DTJ 17 Suppression Pool Check M-123 Valve CTO CS DTJ 17 HPCI-15 HPCI Turbine Steam Trap to NH-36249 B-5 2.0 CK SA 2 C O/C C A CTC CS DTJ 17 Suppression Pool Check M-123 Valve CTO CS DTJ 17 Valve Table Page 26 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System High Pressure Coolant Injection Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ HPCI-18 HPCI Gland Seal NH-36249 A-2 2.0 CK SA 2 C O/C O A CTO Q N/A Condenser/Lube Oil Cooler M-123 Cooling Water Return Check CTC CS DTJ 17 Valve HPCI-20 HPCI Gland Seal NH-36249 A-3 2.0 CK SA 2 C O/C C A CTO Q N/A Condenser/Condensate M-123 Pump Discharge Check CTC Q N/A Valve HPCI-31 HPCI Pump Suction from NH-36250 A-4 14.0 CK SA 2 C C O/C A DI CM DTJ 17 Suppression Pool Check M-124 Valve HPCI-32 HPCI Pump Suction from NH-36250 D-3 14.0 CK SA 2 C C O/C A CTC CS DTJ 29 Condensate Storage Tank M-124 Check Valve CTO Q N/A HPCI-42 HPCI Minimum Flow to NH-36250 B-4 4.0 CK SA 2 C C O/C A DI CM DTJ 30 Suppression Chamber M-124 Check Valve HPCI-65 HPCI Exhaust Line Vacuum NH-36249 B-6 2.0 CK SA 2 C C O/C A CTC CS DTJ 17 Breaker Check Valve M-123 CTO CS DTJ 17 HPCI-71 HPCI Exhaust Line Vacuum NH-36249 B-6 2.0 CK SA 2 C C O/C A CTC CS DTJ 17 Breaker Check Valve M-123 CTO CS DTJ 17 HPCI-9 HPCI Turbine Exhaust NH-36249 C-5 16.0 CK SA 2 AC O/C O/C A CTO Q N/A Check Valve M-123 CTC CM DTJ 17 LT AJ N/A Valve Table Page 27 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System High Pressure Coolant Injection Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ MO-2034 HPCI Steam Supply Inboard NH-36249 E-5 8.0 GT MO 1 A O O/C A/P LT AJ N/A Containment Isolation Valve M-123 STC Q N/A PIT Y2 N/A MO-2035 HPCI Steam Supply NH-36249 E-4 8.0 GT MO 1 A O O/C A/P LT AJ N/A Outboard Containment M-123 Isolation Valve STC Q N/A PIT Y2 N/A MO-2036 HPCI Turbine Steam Supply NH-36249 D-2 8.0 GT MO 2 B C O A STO PV VR 03 Isolation Valve M-123 PIT PV VR 03 ET Q N/A MO-2061 HPCI Suppression Pool NH-36250 A-5 14.0 GT MO 2 B C O/C A STO PV VR 03 Suction Isolation Valve M-124 STC PV VR 03 PIT Y2 N/A ET Q N/A MO-2062 HPCI Suppression Pool NH-36250 A-4 14.0 GT MO 2 B C O/C A STO PV VR 03 Suction Isolation Valve M-124 STC PV VR 03 PIT Y2 N/A ET Q N/A MO-2063 HPCI Condensate Storage NH-36250 E-3 14.0 GT MO 2 B O O/C A/P STC PV VR 03 Tank Suction Isolation Valve M-124 PIT PV VR 03 ET Q N/A MO-2067 HPCI Injection Isolation NH-36250 C-5 12.0 GT MO 2 B C O A STO PV VR 03 Valve M-124 PIT PV VR 03 ET Q N/A Valve Table Page 28 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System High Pressure Coolant Injection Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ MO-2068 HPCI Injection Isolation NH-36250 C-5 12.0 GT MO 2 B C O A STO PV VR 03 Valve M-124 PIT PV VR 03 ET Q N/A MO-2071 HPCI Test Return to NH-36250 D-5 8.0 GL MO 2 B O/C C A STC Q N/A Condensate Storage Tank M-124 Isolation Valve PIT Y2 N/A PSD-2038 HPCI Turbine Exhaust Line NH-36249 C-5 16.0 RD SA 2 D C O/C A DT Y5 N/A Rupture Disk M-123 PSD-2039 HPCI Turbine Exhaust Line NH-36249 C-5 16.0 RD SA NC D C O/C A DT Y5 N/A Rupture Disk M-123 RV-2056 HPCI Cooling Water Supply NH-36249 B-3 1.5 RV SA 3 C C O/C A RT SY10 N/A to Gland Seal Condenser M-123 Relief Valve RV-2064 HPCI Pump Suction Relief NH-36250 D-2 1.0 RV SA 2 C C O/C A RT Y10 N/A Valve M-124 XFV-84 HPCI Turbine Steam Supply NH-36249 D-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Line Excess Flow Check M-123 Valve XFV-85 HPCI Turbine Steam Supply NH-36249 D-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Line Excess Flow Check M-123 Valve Valve Table Page 29 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Hydrogen Oxygen Analyzer Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ SV-4001A Drywell to CAM Analyzer NH-91197 B-6 0.75 GL SO 2 A O/C C A STC Q N/A Panel Outboard Isolation NH-91197 Valve PIT Y2 N/A LT AJ N/A FC Q N/A SV-4001B Drywell to CAM Analyzer NH-91197 B-6 0.75 GL SO 2 A O/C C A STC Q N/A Panel Outboard Isolation NH-91197 Valve PIT Y2 N/A LT AJ N/A FC Q N/A SV-4002A Suppression Chamber to NH-91197 A-5 0.75 GL SO 2 A O/C C A STC Q N/A CAM Analyzer Panel Inboard NH-91197 Isolation Valve PIT Y2 N/A LT AJ N/A FC Q N/A SV-4002B Suppression Chamber to NH-91197 A-4 0.75 GL SO 2 A O/C C A STC Q N/A CAM Analyzer Panel Inboard NH-91197 Isolation Valve PIT Y2 N/A LT AJ N/A FC Q N/A SV-4003A Suppression Chamber to NH-91197 B-5 0.75 GL SO 2 A O/C C A STC Q N/A CAM Analyzer Panel NH-91197 Outboard Isolation Valve PIT Y2 N/A LT AJ N/A FC Q N/A SV-4003B Suppression Chamber to NH-91197 B-4 0.75 GL SO 2 A O/C C A STC Q N/A CAM Analyzer Panel NH-91197 Outboard Isolation Valve PIT Y2 N/A LT AJ N/A FC Q N/A Valve Table Page 30 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Hydrogen Oxygen Analyzer Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ SV-4004A Suppression Chamber to NH-91197 A-4 0.75 GL SO 2 A O/C C A STC Q N/A CAM Analyzer Panel Inboard NH-91197 Isolation Valve PIT Y2 N/A LT AJ N/A FC Q N/A SV-4004B Suppression Chamber to NH-91197 A-4 0.75 GL SO 2 A O/C C A STC Q N/A CAM Analyzer Panel Inboard NH-91197 Isolation Valve PIT Y2 N/A LT AJ N/A FC Q N/A SV-4005A Suppression Chamber to NH-91197 B-4 0.75 GL SO 2 A O/C C A STC Q N/A CAM Analyzer Panel NH-91197 Outboard Isolation Valve PIT Y2 N/A LT AJ N/A FC Q N/A SV-4005B Suppression Chamber to NH-91197 B-4 0.75 GL SO 2 A O/C C A STC Q N/A CAM Analyzer Panel NH-91197 Outboard Isolation Valve PIT Y2 N/A LT AJ N/A FC Q N/A SV-4020A Drywell to CAM Analyzer NH-91197 A-6 0.75 GL SO 2 A O/C C A STC Q N/A Panel Inboard Isolation Valve NH-91197 PIT Y2 N/A LT AJ N/A FC Q N/A SV-4020B Drywell to CAM Analyzer NH-91197 A-6 0.75 GL SO 2 A O/C C A STC Q N/A Panel Inboard Isolation Valve NH-91197 PIT Y2 N/A LT AJ N/A FC Q N/A Valve Table Page 31 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Instrument and Service Air Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AI-12-10 Inboard MSIV Pneumatic NH-36049-10 A-3 0.750 CK SA NC C O/C O A DI CM DTJ-33 Supply Check Valve M-131-10 AI-12-11 Inboard MSIV Pneumatic NH-36049-10 B-3 0.750 CK SA NC C O/C O A DI CM DTJ-33 Supply Check Valve M-131-10 AI-12-12 Inboard MSIV Pneumatic NH-36049-10 A-3 0.750 CK SA NC C O/C O A DI CM DTJ-33 Supply Check Valve M-131-10 AI-12-9 Inboard MSIV Pneumatic NH-36049-10 A-3 0.750 CK SA NC C O/C O A DI CM DTJ-33 Supply Check Valve M-131-10 AI-243-1 Instrument Air Supply Check NH-36247 A-3 0.75 CK SA NC C O/C C A CTC RF DTJ 35 Valve to CV-1728 M-121 CTO RF DTJ 35 AI-243-2 Instrument Air Supply Check NH-36246 A-5 0.75 CK SA NC C O/C C A CTC RF DTJ 35 Valve to CV-1729 M-120 CTO RF DTJ 35 AI-244-1 Instrument Air Supply Check NH-36247 A-3 0.75 CK SA NC C O/C C A CTC RF DTJ 35 Valve to CV-1728 M-121 CTO RF DTJ 35 AI-244-2 Instrument Air Supply Check NH-36246 A-5 0.75 CK SA NC C O/C C A CTC RF DTJ 35 Valve to CV-1729 M-120 CTO RF DTJ 35 Valve Table Page 32 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Instrument and Service Air Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AI-571 Inboard Instrument Air NH-36049-12 B-5 2.0 CK SA 2 AC O C A CTC CM DTJ 23 Supply Containment M-131-12 Isolation Valve LT AJ N/A CTO CS DTJ 23 AI-610-1 Instrument Air Supply to CV- NH-36247 A-4 0.75 CK SA NC AC O/C C A CTC RF DTJ 25 1994 Check Valve M-121 CTO RF DTJ 25 LT Y2 NA AI-610-2 Instrument Air Supply to CV- NH-36246 A-4 0.75 CK SA NC AC O/C C A CTC RF DTJ 25 1995 Check Valve M-120 CTO RF DTJ 25 LT Y2 N/A AI-610-3 Instrument Air Supply to CV- NH-36247 C-5 0.75 CK SA NC AC O/C C A CTC RF DTJ 25 1996 Check Valve M-121 CTO RF DTJ 25 LT Y2 N/A AI-610-4 Instrument Air Supply to CV- NH-36246 C-4 0.75 CK SA NC AC O/C C A CTC RF DTJ 25 1997 Check Valve M-120 CTO RF DTJ 25 LT Y2 N/A AI-611 Instrument Air Supply to NH-36250 D-4 0.75 CK SA NC AC O/C C A CTC RF DTJ 25 HPCI Min Flow Control Valve M-124 CV-2065 Check Valve CTO RF DTJ 25 LT Y2 N/A AI-612 RCIC Instrument Air NH-36252 C-4 0.75 CK SA NC AC O/C C A CTC RF DTJ 25 Isolation Check Valve (CV- M-126 2104) LT Y2 N/A CTO RF DTJ 25 AI-613 Instrument Air Supply to T- NH-36049-14 C-6 0.375 CK SA NC AC O/C C A CTC RF DTJ 24 Ring Seal AO-2381 M-131-14 LT Y2 N/A CTO RF DTJ 24 Valve Table Page 33 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Instrument and Service Air Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AI-614 Instrument Air Supply to T- NH-36049-14 C-6 0.375 CK SA NC AC O/C C A CTC RF DTJ 24 Ring Seal AO-2377 M-131-14 LT Y2 N/A CTO RF DTJ 24 AI-615 Instrument Air Supply to T- NH-36049-14 C-6 0.375 CK SA NC AC O/C C A CTC RF DTJ 24 Ring Seal AO-2378 M-131-14 LT Y2 N/A CTO RF DTJ 24 AI-616 Instrument Air Supply to T- NH-36049-14 B-6 0.375 CK SA NC AC O/C C A CTC RF DTJ 24 Ring Seal AO-2896 M-131-14 LT Y2 N/A CTO RF DTJ 24 AI-617 Instrument Air Supply to T- NH-36049-14 A-6 0.375 CK SA NC AC O/C C A CTC RF DTJ 24 Ring Seal AO-2383 M-131-14 LT Y2 N/A CTO RF DTJ 24 AI-618 Instrument Air Supply to T- NH-36049-14 B-2 0.375 CK SA NC AC O/C C A CTC RF DTJ 24 Ring Seal AO-2386 M-131-14 LT Y2 N/A CTO RF DTJ 24 AI-619 Instrument Air Supply to T- NH-36049-14 A-2 0.375 CK SA NC AC O/C C A CTC RF DTJ 24 Ring Seal AO-2387 M-131-14 LT Y2 N/A CTO RF DTJ 24 AI-625 Outboard Tip Purge NH-161004 B-5 1.0 CK SA 2 AC O C A CTC CM DTJ 09 Containment Isolation Valve M-2510 LT AJ N/A CTO Q N/A AI-626-1 Inboard Tip Purge NH-161004 B-5 1.0 CK SA 2 AC O C A CTC CM DTJ 09 Containment Isolation Valve M-2510 LT AJ N/A CTO Q N/A Valve Table Page 34 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Instrument and Service Air Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AI-629 Instrument Air Supply to NH-36049-14 B-5 0.75 CK SA 2 AC O/C C A CTC Q N/A Torus Ring Header Inboard M-131-14 Containment Isolation Valve LT AJ N/A CTO Q N/A AI-663 Alternate Nitrogen Supply to NH-36049-14 C-7 0.375 CK SA NC C O/C O A CTO RF DTJ 24 T-Ring Seal AO-2377 M-131-14 CTC RF DTJ 24 AI-666 Alternate Nitrogen Supply to NH-36049-14 C-5 0.375 CK SA NC C O/C O A CTO RF DTJ 24 T-Ring Seal AO-2378 M-131-14 CTC RF DTJ 24 AI-669 Alternate Nitrogen Supply to NH-36049-14 C-6 0.375 CK SA NC C O/C O A CTO RF DTJ 24 T-Ring Seal AO-2381 M-131-14 CTC RF DTJ 24 AI-672 Alternate Nitrogen Supply to NH-36049-14 A-6 0.375 CK SA NC C O/C O A CTO RF DTJ 24 T-Ring Seal AO-2383 M-131-14 CTC RF DTJ 24 AI-675 Alternate Nitrogen Supply to NH-36049-14 B-1 0.375 CK SA NC C O/C O A CTO RF DTJ 24 T-Ring Seal AO-2386 M-131-14 CTC RF DTJ 24 AI-678 Alternate Nitrogen Supply to NH-36049-14 B-1 0.375 CK SA NC C O/C O A CTO RF DTJ 24 T-Ring Seal AO-2387 M-131-14 CTC RF DTJ 24 AI-681 Alternate Nitrogen Supply to NH-36049-14 B-6 0.375 CK SA NC C O/C O A CTO RF DTJ 24 T-Ring Seal AO-2896 M-131-14 CTC RF DTJ 24 Valve Table Page 35 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Instrument and Service Air Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AI-683 Instrument Air Supply to NH-36049-14 B-5 0.375 CK SA NC AC O/C C A CTC RF DTJ 24 Actuator and T-Ring Seal M-131-14 AO-2379 LT Y2 N/A CTO RF DTJ 24 AI-685 Instrument Air Supply to NH-36049-14 C-4 0.375 CK SA NC AC O/C C A CTC RF DTJ 24 Actuator and T-Ring Seal M-131-14 AO-2380 LT Y2 N/A CTO RF DTJ 24 AI-694 Alternate Nitrogen Supply to NH-36049-14 B-5 0.375 CK SA NC C O/C O A CTO Q N/A Actuator and T-Ring Seal M-131-14 AO-2379 CTC RF DTJ-24 AI-695 Alternate Nitrogen Supply to NH-36049-14 C-4 0.375 CK SA NC C O/C O A CTO Q N/A Actuator and T-Ring Seal M-131-14 AO-2380 CTC RF DTJ-24 AS-78 Service Air Outboard NH-36049-4 D-7 0.75 GT MA 2 A LC C P LT AJ N/A Containment Isolation Valve M-131-4 AS-79 Service Air Inboard NH-36049-4 D-7 0.75 GT MA 2 A LC C P LT AJ N/A Containment Isolation Valve M-131-4 CV-1478 Outboard Instrument Air NH-36049-12 B-6 2.0 GL AO 2 A O C A STC Q N/A Supply Containment M-131-12 Isolation Valve PIT Y2 N/A LT AJ N/A FC Q N/A Valve Table Page 36 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Instrument and Service Air Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ CV-7956 Instrument Air Supply to NH-36049-14 B-6 0.75 PL AO 2 A O/C C A STC Q N/A Torus Ring Header M-131-14 Containment Isolation Valve PIT Y2 N/A LT AJ N/A FC Q N/A Valve Table Page 37 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Liquid Radwaste Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AO-2541A DW Floor Drain Sump (S-38) NH-36043 E-2 2.0 GT AO 2 A O C A STC Q N/A Inboard Isol Valve M-137 PIT Y2 N/A LT AJ N/A FC Q N/A AO-2541B DW Floor Drain Sump (S-38) NH-36043 E-1 2.0 GT AO 2 A O C A STC Q N/A Outboard Isol Valve M-137 PIT Y2 N/A LT AJ N/A FC Q N/A AO-2561A Drywell Equip Drain Sump NH-36044 E-2 2.0 GT AO 2 A O C A STC Q N/A (S-43) Inbd Isol Valve M-138 PIT Y2 N/A LT AJ N/A FC Q N/A AO-2561B Drywell Equip Drain Sump NH-36044 E-1 2.0 GT AO 2 A O C A STC Q N/A (S-43) Otbd Isol Valve M-138 PIT Y2 N/A LT AJ N/A FC Q N/A PSD-6047 DW Floor Drain Sump NH-36043 E-2 0.5 RD SA NC D C O A DT Y5 N/A Rupture Disk (X-18) M-137 PSD-6048 DW Equip Drain Sump NH-36044 E-2 0.5 RD SA NC D C O A DT Y5 N/A Rupture Disk (X-19) M-138 Valve Table Page 38 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Main Condenser Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AO-1825A SJAE Mechanical Vacuum NH-36035-2 A-3 6.0 BF AO NC B O/C C A STC CS DTJ 10 Pump Inlet Valve M-104-1 FC CS DTJ 10 AO-1825B SJAE Mechanical Vacuum NH-36035-2 A-3 6.0 BF AO NC B O/C C A STC CS DTJ 10 Pump Inlet Valve M-104-1 FC CS DTJ 10 Valve Table Page 39 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Main Steam Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AO-2-80A Main Steam Isolation Valve NH-36241 C-5 18.00 GL AO 1 A O C A STC CS DTJ 40 (MSIV) Inboard M-115 FC CS DTJ 13 LT AJ N/A PIT Y2 N/A PS Q N/A AO-2-80B Main Steam Isolation Valve NH-36241 D-5 18.00 GL AO 1 A O C A STC CS DTJ 40 (MSIV) Inboard M-115 FC CS DTJ 13 LT AJ N/A PIT Y2 N/A PS Q N/A AO-2-80C Main Steam Isolation Valve NH-36241 D-2 18.00 GL AO 1 A O C A STC CS DTJ 40 (MSIV) Inboard M-115 FC CS DTJ 13 LT AJ N/A PIT Y2 N/A PS Q N/A AO-2-80D Main Steam Isolation Valve NH-36241 C-2 18.00 GL AO 1 A O C A STC CS DTJ 40 (MSIV) Inboard M-115 FC CS DTJ 13 LT AJ N/A PIT Y2 N/A PS Q N/A AO-2-86A Main Steam Isolation Valve NH-36241 C-5 18.00 GT AO 1 A O C A STC CS DTJ 40 (MSIV) Outboard M-115 LT AJ N/A PIT Y2 N/A FC CS DTJ 40 PS Q N/A Valve Table Page 40 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Main Steam Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AO-2-86B Main Steam Isolation Valve NH-36241 D-5 18.00 GT AO 1 A O C A STC CS DTJ 40 (MSIV) Outboard M-115 LT AJ N/A PIT Y2 N/A FC CS DTJ 40 PS Q N/A AO-2-86C Main Steam Isolation Valve NH-36241 D-2 18.00 GT AO 1 A O C A STC CS DTJ 40 (MSIV) Outboard M-115 LT AJ N/A PIT Y2 N/A FC CS DTJ 40 PS Q N/A AO-2-86D Main Steam Isolation Valve NH-36241 C-2 18.00 GL AO 1 A O C A STC CS DTJ 40 (MSIV) Outboard M-115 LT AJ N/A PIT Y2 N/A FC CS DTJ 40 PS Q N/A MO-2373 Main Steam Line Drain NH-36241 B-5 3.00 GL MO 1 A O/C C A STC Q N/A Isolation Valve (Inboard) M-115 LT AJ N/A PIT Y2 N/A MO-2374 Main Steam Line Drain NH-36241 B-6 3.00 GL MO 1 A O/C C A STC Q N/A Isolation Valve (Outboard) M-115 LT AJ N/A PIT Y2 N/A XFV-1 Flange Leakoff to Clean NH-36241 E-1 1.00 FV SA 2 AC O C A XT RF DTJ 22 Radwaste Excess Flow M-115 Check Valve Valve Table Page 41 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Main Steam Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ XFV-2 Main Steam Line Excess NH-36241 D-2 1.00 FV SA 1 AC O C A XT RF DTJ 22 Flow Check Valve M-115 XFV-3 Main Steam Line Excess NH-36241 D-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Flow Check Valve M-115 XFV-4 Main Steam Line Excess NH-36241 D-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Flow Check Valve M-115 XFV-5 Main Steam Line Excess NH-36241 D-2 1.00 FV SA 1 AC O C A XT RF DTJ 22 Flow Check Valve M-115 XFV-6 Main Steam Line Excess NH-36241 B-2 1.00 FV SA 1 AC O C A XT RF DTJ 22 Flow Check Valve M-115 XFV-7 Main Steam Line Excess NH-36241 B-2 1.00 FV SA 1 AC O C A XT RF DTJ 22 Flow Check Valve M-115 XFV-8 Main Steam Line Excess NH-36241 B-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Flow Check Valve M-115 XFV-9 Main Steam Line Excess NH-36241 B-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Flow Check Valve M-115 Valve Table Page 42 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Post Accident Sampling Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ PAS-58-1 Post Accident Sampling NH-96042-1 A-6 0.75 CK SA 2 C O/C C A CTC Q N/A System to RHR System NH-96042-Check Valve CTO Q N/A PAS-58-2 Post Accident Sampling NH-96042-1 A-6 0.75 CK SA 2 C O/C C A CTC Q N/A System to RHR System NH-96042-Check Valve CTO Q N/A PAS-59-5 Post Accident Sampling NH-96042-1 B-6 0.75 FV SA 2 C O C A XT CS DTJ 20 System From RHR System NH-96042-Excess Flow Check Valve PAS-59-6 Post Accident Sampling NH-96042-1 B-6 0.75 FV SA 2 C O C A XT CS DTJ 20 System From RHR System NH-96042-Excess Flow Check Valve SV-4081 PASS Sample Line Inboard NH-96042-1 C-7 0.75 GL SO 2 A O/C C A STC Q N/A Isolation Valve NH-96042-PIT Y2 N/A LT AJ N/A FC Q N/A SV-4082 PASS Sample Line NH-96042-1 D-7 0.75 GL SO 2 A O/C C A STC Q N/A Outboard Isolation Valve NH-96042-PIT Y2 N/A LT AJ N/A FC Q N/A Valve Table Page 43 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Primary Containment Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AO-2377 Drywell/Torus Purge NH-36258 C-2 18.0 BF AO 2 A O/C C A STC Q N/A Outboard Containment M-143 Isolation Valve FC Q N/A PIT Y2 N/A LT AJ N/A AO-2378 Suppression Chamber Purge NH-36258 C-3 18.0 BF AO 2 A O/C C A STC Q N/A Inboard Containment M-143 Isolation Valve FC Q N/A PIT Y2 N/A LT AJ N/A AO-2379 Suppression Chamber NH-36258 C-2 20.0 BF AO 2 A C O/C A STO Q N/A Vacuum Relief Air Operated M-143 Valve STC Q N/A FO Q N/A PIT Y2 N/A LT AJ N/A AO-2380 Suppression Chamber NH-36258 B-2 20.0 BF AO 2 A C O/C A STO Q N/A Vacuum Relief Air Operated M-143 Valve STC Q N/A FO Q N/A PIT Y2 N/A LT AJ N/A AO-2381 Drywell Purge Supply NH-36258 C-3 18.0 BF AO 2 A O/C C A STC Q N/A Inboard Containment M-143 Isolation Valve FC Q N/A PIT Y2 N/A LT AJ N/A AO-2382A Suppression Chamber to NH-36258 B-4 18.0 CK SA NC AC C O/C A CTO RF DTJ 28 Drywell Vacuum Breaker M-143 Check Valve CTC RF DTJ 28 LT Y2 N/A PIT Y2 N/A Valve Table Page 44 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Primary Containment Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AO-2382B Suppression Chamber to NH-36258 B-4 18.0 CK SA NC AC C O/C A CTO RF DTJ 28 Drywell Vacuum Breaker M-143 Check Valve CTC RF DTJ 28 LT Y2 N/A PIT Y2 N/A AO-2382C Suppression Chamber to NH-36258 B-4 18.0 CK SA NC AC C O/C A CTO RF DTJ 28 Drywell Vacuum Breaker M-143 Check Valve CTC RF DTJ 28 LT Y2 N/A PIT Y2 N/A AO-2382E Suppression Chamber to NH-36258 B-4 18.0 CK SA NC AC C O/C A CTO RF DTJ 28 Drywell Vacuum Breaker M-143 Check Valve CTC RF DTJ 28 LT Y2 N/A PIT Y2 N/A AO-2382F Suppression Chamber to NH-36258 B-4 18.0 CK SA NC AC C O/C A CTO RF DTJ 28 Drywell Vacuum Breaker M-143 Check Valve CTC RF DTJ 28 LT Y2 N/A PIT Y2 N/A AO-2382G Suppression Chamber to NH-36258 B-4 18.0 CK SA NC AC C O/C A CTO RF DTJ 28 Drywell Vacuum Breaker M-143 Check Valve CTC RF DTJ 28 LT Y2 N/A PIT Y2 N/A AO-2382H Suppression Chamber to NH-36258 B-4 18.0 CK SA NC AC C O/C A CTO RF DTJ 28 Drywell Vacuum Breaker M-143 Check Valve CTC RF DTJ 28 LT Y2 N/A PIT Y2 N/A Valve Table Page 45 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Primary Containment Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AO-2382K Suppression Chamber to NH-36258 B-4 18.0 CK SA NC AC C O/C A CTO RF DTJ 28 Drywell Vacuum Breaker M-143 Check Valve CTC RF DTJ 28 LT Y2 N/A PIT Y2 N/A AO-2383 Suppression Chamber Vent NH-36258 B-6 18.0 BF AO 2 A O/C C A STC Q N/A to Standby Gas Containment M-143 Isolation Valve FC Q N/A PIT Y2 N/A LT AJ N/A AO-2386 Drywell Vent to Standby Gas NH-36258 C-6 18.0 BF AO 2 A O/C C A STC Q N/A Containment Isolation Valve M-143 FC Q N/A PIT Y2 N/A LT AJ N/A AO-2387 Drywell Otbd Vent to NH-36258 D-6 18.0 BF AO 2 A O/C C A STC Q N/A Standby Gas Containment M-143 Isolation Valve FC Q N/A PIT Y2 N/A LT AJ N/A AO-2896 Torus Main Exhaust to NH-36258 B-6 18.0 BF AO 2 A O/C C A STC Q N/A Standby Gas Containment M-143 Isolation Valve FC Q N/A PIT Y2 N/A LT AJ N/A AO-4539 Hard Pipe Vent Isolation NH-116629 C-3 8.0 BF AO 2 A C C P LT AJ N/A Valve NH-116629 PIT Y2 N/A AO-4540 Hard Pipe Vent Isolation NH-116629 C-4 8.0 BF AO 2 A C C P LT AJ N/A Valve NH-116629 PIT Y2 N/A Valve Table Page 46 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Primary Containment Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ CV-2384 Suppression Chamber Vent NH-36258 B-6 2.0 GL AO 2 A O/C C A STC Q N/A to Standby Gas Containment M-143 Isolation Valve FC Q N/A PIT Y2 N/A LT AJ N/A CV-2385 Suppression Chamber Vent NH-36258 C-6 2.0 GL AO 2 A O/C C A STC Q N/A to Standby Gas Containment M-143 Isolation Valve FC Q N/A PIT Y2 N/A LT AJ N/A CV-3267 Torus Nitrogen Make-up NH-46162 C-4 1.0 GL AO 2 A O/C C A STC Q N/A Inboard Isolation Valve M-130 PIT Y2 N/A LT AJ N/A FC Q N/A CV-3268 Drywell Nitrogen Make-up NH-46162 C-4 1.0 GL AO 2 A O/C C A STC Q N/A Inboard Isolation Valve M-130 PIT Y2 N/A LT AJ N/A FC Q N/A CV-3269 Drywell/Torus Nitrogen Make- NH-46162 D-4 1.0 GL AO 2 A O/C C A STC Q N/A up Outboard Isolation Valve M-130 PIT Y2 N/A LT AJ N/A FC Q N/A CV-3311 Torus Outboard Isolation to NH-46162 C-5 1.0 GL AO 2 A O/C C A STC Q N/A Oxygen Analyzer M-130 PIT Y2 N/A LT AJ N/A FC Q N/A Valve Table Page 47 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Primary Containment Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ CV-3312 Torus Inboard Isolation to NH-46162 B-5 1.0 GL AO 2 A O/C C A STC Q N/A Oxygen Analyzer M-130 PIT Y2 N/A LT AJ N/A FC Q N/A CV-3313 Oxygen Analyzer to Torus NH-46162 C-4 1.0 GL AO 2 A O/C C A STC Q N/A Return Line Outboard M-130 Isolation Valve PIT Y2 N/A LT AJ N/A FC Q N/A CV-3314 Oxygen Analyzer to Torus NH-46162 B-4 1.0 GL SO 2 A O/C C A STC Q N/A Return Line Inboard Isolation M-130 Valve PIT Y2 N/A LT AJ N/A FC Q N/A DWV-8-1 Suppression Chamber NH-36258 B-1 20.0 CK SA 2 AC C O/C A CTO Q N/A Vacuum Relief Check Valve M-143 CTC Q N/A LT AJ N/A DWV-8-2 Suppression Chamber NH-36258 C-1 20.0 CK SA 2 AC C O/C A CTO Q N/A Vacuum Relief Check Valve M-143 CTC Q N/A LT AJ N/A PSD-4543 Hard Pipe Vent Line Rupture NH-116629 C-6 10.0 RD SA NC D C C P DT Y5 N/A Disk NH-116629 SV-3307 Drywell to CAM Sample Line NH-46162 C-5 0.75 GL SO 2 A O/C C A STC Q N/A Outboard Isolation Valve M-130 PIT Y2 N/A LT AJ N/A FC Q N/A Valve Table Page 48 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Primary Containment Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ SV-3308 Drywell to CAM Sample Line NH-46162 C-5 0.75 GL SO 2 A O/C C A STC Q N/A Inboard Isolation Valve M-130 PIT Y2 N/A LT AJ N/A FC Q N/A Valve Table Page 49 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor and Vessel Assembly Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ BF-12 Control Rod Drive to Reactor NH-36242-2 D-3 0.38 CK SA 2 AC O C A LT Y2 N/A Vessel Reference Leg M-116-2 Backfill Check Valve CTC RF DTJ 21 CTO Q N/A BF-14 Control Rod Drive to Reactor NH-36242-2 D-3 0.38 CK SA 2 AC O C A LT Y2 N/A Vessel Reference Leg M-116-2 Backfill Check Valve CTC RF DTJ 21 CTO Q N/A BF-24 Control Rod Drive to Reactor NH-36242-2 D-3 0.38 CK SA 2 AC O C A LT Y2 N/A Vessel Reference Leg M-116-2 Backfill Check Valve CTC RF DTJ 21 CTO Q N/A BF-26 Control Rod Drive to Reactor NH-36242-2 D-3 0.38 CK SA 2 AC O C A LT Y2 N/A Vessel Reference Leg M-116-2 Backfill Check Valve CTC RF DTJ 21 CTO Q N/A BF-35 Control Rod Drive to Reactor NH-36242-2 C-3 0.38 CK SA 2 AC O C A LT Y2 N/A Vessel Reference Leg M-116-2 Backfill Check Valve CTC RF DTJ 21 CTO Q N/A BF-37 Control Rod Drive to Reactor NH-36242-2 C-3 0.38 CK SA 2 AC O C A LT Y2 N/A Vessel Reference Leg M-116-2 Backfill Check Valve CTC RF DTJ 21 CTO Q N/A BF-46 Control Rod Drive to Reactor NH-36242-2 C-3 0.38 CK SA 2 AC O C A LT Y2 N/A Vessel Reference Leg M-116-2 Backfill Check Valve CTC RF DTJ 21 CTO Q N/A BF-48 Control Rod Drive to Reactor NH-36242-2 C-3 0.38 CK SA 2 AC O C A LT Y2 N/A Vessel Reference Leg M-116-2 Backfill Check Valve CTC RF DTJ 21 CTO Q N/A Valve Table Page 50 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor and Vessel Assembly Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ CV-2371 Reactor Pressure Vessel NH-36241 E-5 0.75 DI AO 2 B C C P PIT Y2 N/A Head Vent Isolation Valve M-115 CV-2372 Reactor Pressure Vessel NH-36241 E-5 0.75 DI AO 2 B C C P PIT Y2 N/A Head Vent Isolation Valve M-115 XFV-22 Reactor Vessel Above Core NH-36242-1 C-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Plate Pressure M-116-1 Instrumentation Excess Flow Check Valve XFV-23 Reactor Vessel Below Core NH-36242-1 C-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Plate Pressure M-116-1 Instrumentation Excess Flow Check Valve XFV-24 Reactor Vessel Reference NH-36242 B-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Leg Instrumentation Excess M-116 Flow Check Valve XFV-25 Reactor Vessel Reference NH-36242 D-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Leg Instrumentation Excess M-116 Flow Check Valve XFV-26 Reactor Vessel Reference NH-36242 C-3 1.00 FV SA 1 AC O C A XT RF DTJ 22 Leg Instrumentation Excess M-116 Flow Check Valve XFV-27 Reactor Vessel Jet Pump #4 NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Instrumentation Excess Flow M-116-1 Check Valve Valve Table Page 51 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor and Vessel Assembly Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ XFV-28 Reactor Vessel Jet Pump #3 NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Instrumentation Excess Flow M-116-1 Check Valve XFV-29 Reactor Vessel Jet Pump #2 NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Instrumentation Excess Flow M-116-1 Check Valve XFV-30 Reactor Vessel Jet Pump #1 NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Instrumentation Excess Flow M-116-1 Check Valve XFV-31 Reactor Vessel Jet Pump #1 NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Instrumentation Excess Flow M-116-1 Check Valve XFV-32 Reactor Vessel Jet Pump NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22

10 Instrumentation Excess M-116-1 Flow Check Valve XFV-33 Reactor Vessel Jet Pump #5 NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Instrumentation Excess Flow M-116-1 Check Valve XFV-34 Reactor Vessel Jet Pump #6 NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Instrumentation Excess Flow M-116-1 Check Valve XFV-35 Reactor Vessel Jet Pump #7 NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Instrumentation Excess Flow M-116-1 Check Valve Valve Table Page 52 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor and Vessel Assembly Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ XFV-36 Reactor Vessel Jet Pump #8 NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Instrumentation Excess Flow M-116-1 Check Valve XFV-37 Reactor Vessel Jet Pump #9 NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Instrumentation Excess Flow M-116-1 Check Valve XFV-38 Reactor Vessel Jet Pump #6 NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Instrumentation Excess Flow M-116-1 Check Valve XFV-39 Reactor Vessel Jet Pump NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22

20 Instrumentation Excess M-116-1 Flow Check Valve XFV-40 Reactor Vessel Jet Pump NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22

14 Instrumentation Excess M-116-1 Flow Check Valve XFV-41 Reactor Vessel Jet Pump NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22

12 Instrumentation Excess M-116-1 Flow Check Valve XFV-42 Reactor Vessel Jet Pump NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22

13 Instrumentation Excess M-116-1 Flow Check Valve XFV-43 Reactor Vessel Jet Pump NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22

11 Instrumentation Excess M-116-1 Flow Check Valve Valve Table Page 53 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor and Vessel Assembly Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ XFV-44 Reactor Vessel Jet Pump NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22

11 Instrumentation Excess M-116-1 Flow Check Valve XFV-45 Reactor Vessel Jet Pump NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22

15 Instrumentation Excess M-116-1 Flow Check Valve XFV-46 Reactor Vessel Jet Pump NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22

19 Instrumentation Excess M-116-1 Flow Check Valve XFV-47 Reactor Vessel Jet Pump NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22

17 Instrumentation Excess M-116-1 Flow Check Valve XFV-48 Reactor Vessel Jet Pump NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22

18 Instrumentation Excess M-116-1 Flow Check Valve XFV-49 Reactor Vessel Jet Pump NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22

16 Instrumentation Excess M-116-1 Flow Check Valve XFV-50 Reactor Vessel Jet Pump NH-36242-1 A-5 1.00 FV SA 1 AC O C A XT RF DTJ 22

16 Instrumentation Excess M-116-1 Flow Check Valve XFV-51 Reactor Vessel Below Core NH-36242-1 C-3 1.00 FV SA 1 AC O C A XT RF DTJ 22 Plate Flow Instrumentation M-116-1 Excess Flow Check Valve Valve Table Page 54 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor and Vessel Assembly Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ XFV-52 Reactor Vessel Reference NH-36242 D-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Leg Instrumentation Excess M-116 Flow Check Valve XFV-53 Reactor Vessel Reference NH-36242 D-3 1.00 FV SA 1 AC O C A XT RF DTJ 22 Leg Instrumentation Excess M-116 Flow Check Valve XFV-54 Reactor Vessel Reference NH-36242 C-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Leg Instrumentation Excess M-116 Flow Check Valve XFV-55 Reactor Vessel Reference NH-36242 C-3 1.00 FV SA 1 AC O C A XT RF DTJ 22 Leg Instrumentation Excess M-116 Flow Check Valve XFV-56 Reactor Vessel Reference NH-36242 C-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Leg Instrumentation Excess M-116 Flow Check Valve XFV-57 Reactor Vessel Reference NH-36242 B-3 1.00 FV SA 1 AC O C A XT RF DTJ 22 Leg Instrumentation Excess M-116 Flow Check Valve XFV-88 Reactor Vessel Reference NH-36242 D-5 1.00 FV SA 1 AC O C A XT RF DTJ 22 Leg Instrumentation Excess M-116 Flow Check Valve XFV-89 Reactor Vessel Reference NH-36242 D-3 1.00 FV SA 1 AC O C A XT RF DTJ 22 Leg Instrumentation Excess M-116 Flow Check Valve Valve Table Page 55 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor Building Closed Cooling Water Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ MO-1426 RBCCW Return Line Inboard NH-36042-2 C-7 8.0 GT MO 2 A O C A STC PV VR 03 Containment Isolation Valve M-111-1 PIT Y2 N/A LT AJ N/A ET CS DTJ 11 MO-4229 RBCCW Supply Line NH-36042-2 D-3 8.0 GT MO 2 A O C A STC PV VR 03 Outboard Containment M-111-1 Isolation Valve PIT Y2 N/A LT AJ N/A ET CS DTJ 11 MO-4230 RBCCW Return Line NH-36042-2 D-7 8.0 GT MO 2 A O C A STC PV VR 03 Outboard Containment M-111-1 Isolation Valve PIT Y2 N/A LT AJ N/A ET CS DTJ 11 RBCC-15 RBCCW Supply Line NH-36042-2 C-3 8.0 CK SA 2 AC O C A CTC CM DTJ 01 Inboard Containment M-111-1 Isolation Valve LT AJ N/A CTO Q N/A Valve Table Page 56 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor Core Isolation Cooling Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AO-13-22 RCIC Injection Check Valve NH-36252 B-5 4.0 CK AO 2 C C O/C A DI CM DTJ 15 M-126 PIT Y2 N/A CV-2082A RCIC Steam Line Drain NH-36251 B-1 1.0 GL AO 2 B O C A STC Q N/A Isolation Valve M-125 FC Q N/A PIT Y2 N/A CV-2104 RCIC Pump Minimum Flow NH-36252 A-2 2.0 GL AO 2 B C O/C A STO Q N/A Control Valve M-126 PIT Y2 N/A FO Q N/A STC Q N/A MO-2075 RCIC Steam Supply Inboard NH-36251 E-5 3.0 GT MO 1 A O O/C A/P LT AJ N/A Containment Isolation Valve M-125 STC Q N/A PIT Y2 N/A MO-2076 RCIC Steam Supply NH-36251 E-4 3.0 GT MO 1 A O O/C A/P LT AJ N/A Outboard Containment M-125 Isolation Valve STC Q N/A PIT Y2 N/A MO-2078 RCIC Turbine Steam Supply NH-36251 D-2 3.0 GL MO 2 B C O/C A STO PV VR 03 Isolation Valve M-125 STC PV VR 03 PIT PV VR 03 ET Q N/A Valve Table Page 57 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor Core Isolation Cooling Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ MO-2096 RCIC Lube Oil to Barometric NH-36251 A-4 2.0 GL MO 2 B C O/C A STO PV VR 03 Condenser Cooling Isolation M-125 Valve STC PV VR 03 PIT PV VR 03 ET Q N/A MO-2100 RCIC Suppression Pool NH-36252 A-5 6.0 GT MO 2 B C O A STO PV VR 03 Suction Isolation Valve M-126 PIT Y2 N/A ET Q N/A MO-2101 RCIC Suppression Pool NH-36252 D-4 6.0 GT MO 2 B C O A STO PV VR 03 Suction Isolation Valve M-126 PIT PV VR 03 ET Q N/A MO-2102 RCIC Condensate Storage NH-36252 D-3 6.0 GT MO 2 B O O/C A/P STC PV VR 03 Tank Suction Valve M-126 PIT PV VR 03 ET Q N/A MO-2106 RCIC Pump Discharge Valve NH-36252 B-5 4.0 GT MO 2 B C O A STO PV VR 03 M-126 PIT PV VR 03 ET Q N/A MO-2107 RCIC Pump Discharge NH-36252 B-5 4.0 GT MO 2 B C O A STO PV VR 03 Valve - Injection Line to M-126 Feedwater Isolation PIT PV VR 03 ET Q N/A MO-3502 RCIC Test Return Line NH-36252 D-5 4.0 GT MO 2 B O/C C A STC Q N/A Isolation Valve M-126 PIT Y2 N/A Valve Table Page 58 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor Core Isolation Cooling Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ PSD-2089 RCIC Turbine Exhaust Line NH-36251 C-5 8.0 RD SA 2 D C O/C A DT Y5 N/A Rupture Disc M-125 PSD-2090 RCIC Turbine Exhaust Line NH-36251 C-5 8.0 RD SA NC D C O/C A DT Y5 N/A Rupture Disc M-125 RCIC-10 RCIC Turbine Exhaust NH-36251 C-6 8.0 CK SA 2 AC C O/C A CTO Q N/A Check Valve M-125 CTC CM DTJ 18 LT AJ N/A RCIC-14 RCIC Condensate Return NH-36251 C-5 2.0 CK SA 2 C C O/C A CTO Q N/A Check Valve M-125 CTC Q N/A RCIC-16 RCIC Barometric Condenser NH-36251 A-5 2.0 CK SA 2 C O/C C A CTC CS DTJ 18 Vacuum Pump Discharge to M-125 Suppression Pool Check CTO Q N/A Valve RCIC-17 RCIC Barometric Condenser NH-36251 A-5 2.0 CK SA 2 C O/C C A CTC CS DTJ 18 Vacuum Pump Discharge to M-125 Suppression Pool Check CTO Q N/A Valve RCIC-31 RCIC Torus Suction Supply NH-36252 A-4 6.0 CK SA 2 C C O/C A DI CM DTJ 18 Line Check Valve M-126 RCIC-37 RCIC Minimum Flow Line to NH-36252 A-5 2.0 CK SA 2 C C O/C A DI CM DTJ 30 Suppression Pool Check M-126 Valve Valve Table Page 59 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor Core Isolation Cooling Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ RCIC-41 RCIC Condensate Storage NH-36252 D-3 6.0 CK SA 2 C O/C O/C A CTO Q N/A Tank Suction Check Valve M-126 CTC CS DTJ 29 RCIC-57 RCIC Exhaust Line Vacuum NH-36251 B-6 1.5 CK SA 2 C C O/C A CTC CS DTJ 18 Breaker Check Valve M-125 CTO CS DTJ 18 RCIC-59 RCIC Exhaust Line Vacuum NH-36251 B-6 1.5 CK SA 2 C C O/C A CTC CS DTJ 18 Breaker Check Valve M-125 CTO CS DTJ 18 RCIC-9 RCIC Turbine Exhaust NH-36251 C-5 8.0 CK SA 2 AC C O/C A CTO Q N/A Check Valve M-125 CTC CM DTJ 18 LT AJ N/A RV-2097 RCIC Cooling Water Supply NH-36251 B-3 1.0 RV SA 3 C C O/C A RT SY10 N/A to Barometric Condenser M-125 Relief Valve RV-2103 RCIC Suction Supply Line NH-36252 D-3 1.0 RV SA 2 C C O/C A RT Y10 N/A Relief Valve M-126 XFV-86 RCIC Turbine Steam Supply NH-36251 D-5 1.0 FV SA 1 AC O C A XT RF DTJ 22 Line Excess Flow Check M-125 Valve XFV-87 RCIC Turbine Steam Supply NH-36251 D-5 1.0 FV SA 1 AC O C A XT RF DTJ 22 Line Excess Flow Check M-125 Valve Valve Table Page 60 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor Recirculation Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ CV-2790 Reactor Water Sample NH-36243 D-5 0.75 GL AO 2 A O/C C A STC Q N/A Inboard Containment M-117 Isolation Valve PIT Y2 N/A LT AJ N/A FC Q N/A CV-2791 Reactor Water Sample NH-36243 D-6 0.75 GL AO 2 A O/C C A STC Q N/A Outboard Containment M-117 Isolation Valve PIT Y2 N/A LT AJ N/A FC Q N/A MO-2-53A Reactor Recirculation Pump NH-36243 B-2 28.0 GT MO 1 B O C A STC PV VR 03 P-200B Discharge Valve M-117 PIT PV VR 03 ET CS DTJ 14 MO-2-53B Reactor Recirculation Pump NH-36243 B-6 28.0 GT MO 1 B O C A STC PV VR 03 P-200A Discharge Valve M-117 PIT PV VR 03 ET CS DTJ 14 XFV-58 Reactor Recirculation NH-36243 B-2 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-59 Reactor Recirculation NH-36243 B-2 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-60 Reactor Recirculation NH-36243 A-3 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve Valve Table Page 61 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor Recirculation Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ XFV-61 Reactor Recirculation NH-36243 B-6 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-62 Reactor Recirculation NH-36243 B-6 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-63 Reactor Recirculation NH-36243 A-5 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-64 Reactor Recirculation NH-36243 D-5 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-65 Reactor Recirculation NH-36243 D-5 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-66 Reactor Recirculation NH-36243 E-2 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-67 Reactor Recirculation NH-36243 C-5 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-68 Reactor Recirculation NH-36243 C-5 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve Valve Table Page 62 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor Recirculation Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ XFV-69 Reactor Recirculation NH-36243 E-6 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-70 Reactor Recirculation NH-36243 E-6 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-71 Reactor Recirculation NH-36243 E-2 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-72 Reactor Recirculation NH-36243 E-6 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-73 Reactor Recirculation NH-36243 E-6 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-74 Reactor Recirculation NH-36243 E-2 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-75 Reactor Recirculation NH-36243 E-2 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-76 Reactor Recirculation NH-36243 D-3 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve Valve Table Page 63 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor Recirculation Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ XFV-77 Reactor Recirculation NH-36243 D-3 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-78 Reactor Recirculation NH-36243-1 C-3 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-79 Reactor Recirculation NH-36243-1 C-3 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-80 Reactor Recirculation NH-36243-1 C-4 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XFV-81 Reactor Recirculation NH-36243-1 C-4 1.0 FV SA 1 AC O C A XT RF DTJ 22 System Excess Flow Check M-117 Valve XR-25-1 Recirc Pump (11) Seal NH-36244 A-3 1.0 CK SA 2 AC O C A CTC CM DTJ 03 Outboard Seal Injection M-118 Check Valve LT AJ N/A CTO Q N/A XR-25-2 Recirc Pump (12) Seal NH-36244 A-3 1.0 CK SA 2 AC O C A CTC CM DTJ 03 Outboard Seal Injection M-118 Check Valve LT AJ N/A CTO Q N/A XR-27-1 Reactor Recirculation NH-36243-1 D-3 1.0 CK SA 2 AC O C A LT AJ N/A System Loop 11 Inboard M-117 Seal Injection Check Valve CTC CM DTJ 03 CTO Q N/A Valve Table Page 64 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor Recirculation Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ XR-27-2 Reactor Recirculation NH-36243-1 D-5 1.0 CK SA 2 AC O C A LT AJ N/A System Loop 12 Inboard M-117 Seal Injection Check Valve CTC CM DTJ 03 CTO Q N/A Valve Table Page 65 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor Water Cleanup Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ MO-2397 RWCU Supply Line Inboard NH-36254 C-8 4.0 GT MO 1 A O C A STC RF DTJ 41 Containment Isolation Valve M-128 PIT Y2 N/A LT AJ N/A MO-2398 RWCU Supply Line NH-36254 C-7 4.0 GT MO 1 A O C A STC RF DTJ 41 Outboard Containment M-128 Isolation Valve PIT Y2 N/A LT AJ N/A RC-6-1 RWCU Return Line Check NH-36254 D-6 1.5 CK SA 2 AC O/C C A CTC RF DTJ 08 Valve to B FW Line M-128 CTO RF DTJ 08 LT Y2 N/A RC-6-2 RWCU Return Line Check NH-36254 D-7 1.5 CK SA 2 AC O/C C A CTC RF DTJ 08 Valve to A FW Line M-128 CTO RF DTJ 08 LT Y2 N/A Valve Table Page 66 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Residual Heat Removal Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AO-10-46A Residual Heat Removal NH-36247 C-5 16.0 CK SA 1 AC O/C O/C A CTO CS DTJ 26 Injection Check Valve M-121 CTC RF DTJ 26 LT Y2 N/A PIT Y2 N/A AO-10-46B Residual Heat Removal NH-36246 D-2 16.0 CK SA 1 AC C O/C A CTO CS DTJ 26 Injection Check Valve M-120 CTC RF DTJ 26 LT Y2 N/A PIT Y2 N/A CV-1994 Residual Heat Removal NH-36247 B-4 2.0 GL AO 2 B O/C O/C A STO Q N/A Pump A Minimum Flow M-121 Isolation Valve STC Y10 N/A FO Q N/A PIT Y2 N/A CV-1995 Residual Heat Removal NH-36246 B-4 2.0 GL AO 2 B O/C O/C A STO Q N/A Pump B Minimum Flow M-120 Isolation Valve STC Q N/A FO Q N/A PIT Y2 N/A CV-1996 Residual Heat Removal NH-36247 B-5 2.0 GL AO 2 B O/C O/C A STO Q N/A Pump C Minimum Flow M-121 Isolation Valve STC Q N/A FO Q N/A PIT Y2 N/A CV-1997 Residual Heat Removal NH-36246 C-4 2.0 GL AO 2 B O/C O/C A STO Q N/A Pump D Minimum Flow M-120 Isolation Valve STC Q N/A FO Q N/A PIT Y2 N/A Valve Table Page 67 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Residual Heat Removal Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ MO-1986 RHR Pump Torus Suction NH-36247 B-6 20.0 GT MO 2 B O O P PIT Y2 N/A Isolation Valve M-121 MO-1987 RHR Pump Torus Suction NH-36246 B-2 20.0 GT MO 2 B O O P PIT Y2 N/A Isolation Valve M-120 MO-1988 RHR Shutdown Cooling NH-36247 B-6 18.0 GT MO 2 B C C P PIT Y2 N/A Suction Isolation Valve M-121 MO-1989 RHR Shutdown Cooling NH-36246 A-2 18.0 GT MO 2 B C C P PIT Y2 N/A Suction Isolation Valve M-120 MO-2002 Residual Heat Removal Heat NH-36247 B-3 14.0 GL MO 2 B O O/C A STO PV VR 03 Exchanger Bypass Isolation M-121 Valve STC PV VR 03 PIT PV VR 03 ET Q N/A MO-2003 Residual Heat Removal Heat NH-36246 B-5 14.0 GL MO 2 B O O/C A STO PV VR 03 Exchanger Bypass Isolation M-120 Valve STC PV VR 03 PIT PV VR 03 ET Q N/A MO-2006 Residual Heat Removal to NH-36247 D-3 12.0 GT MO 2 B C O/C A STO PV VR 03 Torus Discharge Isolation M-121 Valve STC PV VR 03 PIT Y2 N/A ET Q N/A Valve Table Page 68 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Residual Heat Removal Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ MO-2007 Residual Heat Removal to NH-36246 D-5 12.0 GT MO 2 B C O/C A STO PV VR 03 Torus Discharge Isolation M-120 Valve STC PV VR 03 PIT Y2 N/A ET Q N/A MO-2008 Residual Heat Removal to NH-36247 C-3 10.0 GL MO 2 B C O/C A STO PV VR 03 Torus Cooling Isolation Valve M-121 STC PV VR 03 PIT Y2 N/A ET Q N/A MO-2009 Residual Heat Removal to NH-36246 C-6 10.0 GL MO 2 B C O/C A STO PV VR 03 Torus Cooling Isolation Valve M-120 STC PV VR 03 PIT Y2 N/A ET Q N/A MO-2010 Residual Heat Removal to NH-36247 C-3 4.0 GL MO 2 A C O/C A STO PV VR 03 Torus Spray Isolation Valve M-121 STC PV VR 03 PIT Y2 N/A LT AJ N/A ET Q N/A MO-2011 Residual Heat Removal to NH-36246 C-5 4.0 GL MO 2 A C O/C A STO PV VR 03 Torus Spray Isolation Valve M-120 STC PV VR 03 PIT Y2 N/A LT AJ N/A ET Q N/A MO-2012 Residual Heat Removal NH-36247 C-4 16.0 GL MO 2 A O O/C A STO PV VR 03 LPCI Outboard Isolation M-121 Valve STC PV VR 03 PIT Y2 N/A LT AJ N/A ET Q N/A Valve Table Page 69 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Residual Heat Removal Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ MO-2013 Residual Heat Removal NH-36246 C-3 16.0 GL MO 2 A O O/C A STO PV VR 03 LPCI Outboard Isolation M-120 Valve STC PV VR 03 PIT Y2 N/A LT AJ N/A ET Q N/A MO-2014 Residual Heat Removal NH-36247 C-5 16.0 GT MO 1 A C O/C A STO CS DTJ 42 LPCI Inboard Isolation Valve M-121 STC CS DTJ 42 PIT Y2 N/A LT AJ N/A LT Y2 N/A MO-2015 Residual Heat Removal NH-36246 C-2 16.0 GT MO 1 A C O/C A STO CS DTJ 42 LPCI Inboard Isolation Valve M-120 STC CS DTJ 42 PIT Y2 N/A LT AJ N/A LT Y2 N/A MO-2020 Residual Heat Removal NH-36247 E-5 10.0 GT MO 2 A C O/C A STO PV VR 03 Containment Spray M-121 Outboard Isolation Valve STC PV VR 03 PIT Y2 N/A LT AJ N/A ET Q N/A MO-2021 Residual Heat Removal NH-36246 E-2 10.0 GT MO 2 A C O/C A STO PV VR 03 Containment Spray M-120 Outboard Isolation Valve STC PV VR 03 PIT Y2 N/A LT AJ N/A ET Q N/A Valve Table Page 70 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Residual Heat Removal Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ MO-2022 Residual Heat Removal NH-36247 E-5 10.0 GT MO 2 A C O/C A STO PV VR 03 Containment Spray Inboard M-121 Isolation Valve STC PV VR 03 PIT Y2 N/A LT AJ N/A ET Q N/A MO-2023 Residual Heat Removal NH-36246 E-2 10.0 GT MO 2 A C O/C A STO PV VR 03 Containment Spray Inboard M-120 Isolation Valve STC PV VR 03 PIT Y2 N/A LT AJ N/A ET Q N/A MO-2029 Reactor Coolant to RHR NH-36247 D-6 18.0 GT MO 1 A C C A STC CS DTJ 16 Shutdown Cooling Supply M-121 Isolation Valve PIT Y2 N/A LT AJ N/A LT Y2 N/A MO-2030 Reactor Coolant to RHR NH-36247 C-6 18.0 GT MO 1 A C C A STC CS DTJ 16 Shutdown Cooling Supply M-121 Isolation Valve PIT Y2 N/A LT AJ N/A LT Y2 N/A MO-2032 RHR Discharge to Waste NH-36247 C-3 4.0 GT MO 2 B C C A STC Q N/A Surge Tank Isolation Valve M-121 PIT Y2 N/A MO-2033 Residual Heat Removal NH-36246 C-6 14.0 GT MO 2 B O O P PIT Y2 N/A Cross Tie Isolation Valve M-120 Valve Table Page 71 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Residual Heat Removal Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ MO-4085A Residual Heat Removal NH-36247 C-5 4.0 GT MO 1 B C C A STC Q N/A Intertie Isolation Valve M-121 PIT Y2 N/A MO-4085B Residual Heat Removal NH-36246 C-1 4.0 GT MO 1 B C C A STC Q N/A Intertie Isolation Valve M-120 PIT Y2 N/A MO-4086 Residual Heat Removal NH-36247 C-6 4.0 GT MO 1 B O O P PIT Y2 N/A Intertie Isolation Valve M-121 RHR-2-1 Residual Heat Removal NH-36247 A-4 10.0 CK SA 2 C C O/C A CTO Q N/A Pump A Discharge Check M-121 Valve CTC Q N/A RHR-2-2 Residual Heat Removal NH-36246 B-5 10.0 CK SA 2 C C O/C A CTO Q N/A Pump B Discharge Check M-120 Valve CTC Q N/A RHR-2-3 Residual Heat Removal NH-36247 B-4 10.0 CK SA 2 C C O/C A CTO Q N/A Pump C Discharge Check M-121 Valve CTC Q N/A RHR-2-4 Residual Heat Removal NH-36246 B-5 10.0 CK SA 2 C C O/C A CTO Q N/A Pump D Discharge Check M-120 Valve CTC Q N/A RHR-6-1 Residual Heat Removal NH-36247 C-5 16.0 GT MA 1 B LO O P PIT Y2 N/A Injection Line Manual M-121 Isolation Valve Valve Table Page 72 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Residual Heat Removal Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ RHR-6-2 Residual Heat Removal NH-36246 D-1 16.0 GT MA 1 B LO O P PIT Y2 N/A Injection Line Manual M-120 Isolation Valve RHR-81 Thermal Overpressurization NH-36247 D-4 1 CK SA 1 AC C O/C A LT Y2 N/A Relief Check Valve for M-121 Penetration X-12 CTO CS DTJ-34 CTC CS DTJ-34 LT AJ N/A RHR-8-1 Residual Heat Removal NH-36247 C-5 3.0 CK SA 2 C C O/C A DI CM DTJ 06 Minimum Flow Check Valve M-121 RHR-8-2 Residual Heat Removal NH-36246 C-3 3.0 CK SA 2 C C O/C A DI CM DTJ 06 Minimum Flow Check Valve M-120 RV-1990 Residual Heat Removal NH-36247 A-5 1.0 RV SA 2 C C O/C A RT SY10 N/A Suction Relief Valve M-121 RV-1991 Residual Heat Removal NH-36246 B-3 1.0 RV SA 2 C C O/C A RT SY10 N/A Suction Relief Valve M-120 RV-1992 Residual Heat Removal NH-36247 B-5 1.0 RV SA 2 C C O/C A RT SY10 N/A Suction Relief Valve M-121 RV-1993 Residual Heat Removal NH-36246 C-3 1.0 RV SA 2 C C O/C A RT SY10 N/A Suction Relief Valve M-120 Valve Table Page 73 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Residual Heat Removal Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ RV-2004 Residual Heat Removal NH-36247 D-2 1.0 RV SA 2 C C O/C A RT SY10 N/A Discharge Relief Valve M-121 RV-2005 Residual Heat Removal NH-36246 D-6 1.0 RV SA 2 C C O/C A RT SY10 N/A Discharge Relief Valve M-120 RV-2031 RHR Shutdown Cooling NH-36247 B-6 1.0 RV SA 2 C C O/C A RT SY10 N/A Suction Relief Valve M-121 RV-4281 Residual Heat Removal Heat NH-36247 B-2 2.5 RV SA 2 C C O/C A RT Y10 N/A Exchanger Shell Side Relief M-121 Valve RV-4282 Residual Heat Removal Heat NH-36246 B-6 2.5 RV SA 2 C C O/C A RT Y10 N/A Exchanger Shell Side Relief M-120 Valve Valve Table Page 74 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System RHR Service Water Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AV-3147 RHR Service Water Pump P- NH-36665 B-3 3.0 AR SA 3 C O/C O/C A CTC Q N/A 109A Air Vent Valve M-811 CTO Q N/A AV-3148 RHR Service Water Pump P- NH-36665 B-8 3.0 AR SA 3 C O/C O/C A CTC Q N/A 109D Air Vent Valve M-811 CTO Q N/A AV-3149 RHR Service Water Pump P- NH-36665 B-3 3.0 AR SA 3 C O/C O/C A CTC Q N/A 109C Air Vent Valve M-811 CTO Q N/A AV-3150 RHR Service Water Pump P- NH-36665 B-8 3.0 AR SA 3 C O/C O/C A CTC Q N/A 109B Air Vent Valve M-811 CTO Q N/A CV-1728 RHR Service Water to RHR NH-36664 A-5 12.0 GL AO 3 B O/C O A ET Q VR 02 Heat Exchanger Flow M-112 Control Valve PIT Y2 N/A CV-1729 RHR Service Water to RHR NH-36664 A-4 12.0 GL AO 3 B O/C O A ET Q VR 02 Heat Exchanger Flow M-112 Control Valve PIT Y2 N/A RHRSW-1-1 RHR Service Water Pump P- NH-36665 B-3 12.0 CK SA 3 C O/C O/C A CTO Q N/A 109A Discharge Check Valve M-811 CTC Q N/A RHRSW-1-2 RHR Service Water Pump P- NH-36665 B-8 12.0 CK SA 3 C O/C O/C A CTO Q N/A 109B Discharge Check Valve M-811 CTC Q N/A Valve Table Page 75 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System RHR Service Water Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ RHRSW-1-3 RHR Service Water Pump P- NH-36665 B-3 12.0 CK SA 3 C O/C O/C A CTO Q N/A 109C Discharge Check Valve M-811 CTC Q N/A RHRSW-1-4 RHR Service Water Pump P- NH-36665 B-8 12.0 CK SA 3 C O/C O/C A CTO Q N/A 109D Discharge Check Valve M-811 CTC Q N/A RHRSW-3-1 RHR Service Water Basket NH-36665 C-3 18.0 GT MA 3 B LC O/C A ET Y2 N/A Strainer Bypass Isolation M-811 Valve RHRSW-3-2 RHR Service Water Basket NH-36665 C-7 18.0 GT MA 3 B LC O/C A ET Y2 N/A Strainer Bypass Isolation M-811 Valve RHRSW-48-1 A RHR Service Water NH-36665 B-3 0.5 CK SA 3 C O/C C A CTC Q N/A Biocide Injection Check M-811 Valve CTO Q N/A RHRSW-48-2 B RHR Service Water NH-36665 B-7 0.5 CK SA 3 C O/C C A CTC Q N/A Biocide Injection Check M-811 Valve CTO Q N/A RHRSW-50-1 A RHR Service Water NH-36665 B-3 0.5 CK SA 3 C O/C C A CTC Q N/A Dispersent Check Valve M-811 CTO Q N/A RHRSW-50-2 B RHR Service Water NH-36665 B-7 0.5 CK SA 3 C O/C C A CTC Q N/A Dispersent Check Valve M-811 CTO Q N/A Valve Table Page 76 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System RHR Service Water Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ RV-3202 RHR Heat Exchanger E- NH-36664 C-5 2.5 RV SA 3 C C O/C A RT Y10 N/A 200A Relief Valve M-112 RV-3203 RHR Heat Exchanger E- NH-36664 C-4 2.5 RV SA 3 C C O/C A RT Y10 N/A 200B Relief Valve M-112 RV-4908A Residual Heat Removal NH-36247 A-4 0.25 RV SA NC C C O/C A RT SY10 N/A Auxiliary Compressor Relief M-121 Valve RV-4908B Residual Heat Removal NH-36246 A-5 0.25 RV SA NC C C O/C A RT SY10 N/A Auxiliary Compressor Relief M-120 Valve SW-21-1 Service Water to RHR NH-36665 C-2 1.0 CK SA 3 C O/C C A CTC Q N/A Service Water Check Valve M-811 CTO Q N/A SW-21-2 Service Water to RHR NH-36665 C-7 1.0 CK SA 3 C O/C C A CTC Q N/A Service Water Check Valve M-811 CTO Q N/A SW-22-1 Service Water to RHR NH-36665 C-2 1.0 CK SA NC C O/C C A CTC Q N/A Service Water Check Valve M-811 CTO Q N/A SW-22-2 Service Water to RHR NH-36665 C-7 1.0 CK SA NC C O/C C A CTC Q N/A Service Water Check Valve M-811 CTO Q N/A Valve Table Page 77 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Service Water Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ SW-101 Service Water to Emergency NH-36664 D-2 3.0 CK SA NC C O C A CTC Q N/A Service Water Check Valve M-112 CTO Q N/A SW-102 Service Water to Emergency NH-36664 D-1 3.0 CK SA 3 C O C A CTC Q N/A Service Water Check Valve M-112 CTO Q N/A SW-103 Service Water to Emergency NH-36664 D-3 3.0 CK SA NC C O C A CTC Q N/A Service Water Check Valve M-112 CTO Q N/A SW-104 Service Water to Emergency NH-36664 D-3 3.0 CK SA 3 C O C A CTC Q N/A Service Water Check Valve M-112 CTO Q N/A Valve Table Page 78 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Standby Liquid Control Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ RV-11-39A Standby Liquid Control NH-36253 C-4 2.0 RV SA 2 C C O/C A RT SY10 N/A Pump Discharge Relief Valve M-127 RV-11-39B Standby Liquid Control NH-36253 B-4 2.0 RV SA 2 C C O/C A RT SY10 N/A Pump Discharge Relief Valve M-127 XP-11-14A SBLC A Explosive Actuated NH-36253 C-5 1.50 XP SA 2 D LC O A DT SY10 N/A Squibb Valve M-127 XP-11-14B SBLC B Explosive Actuated NH-36253 D-5 1.50 XP SA 2 D LC O A DT SY10 N/A Squibb Valve M-127 XP-3-1 SBLC Pump P-203A NH-36253 C-4 1.50 CK SA 2 C O/C O/C A DI CM DTJ37 Discharge Check Valve M-127 CTO Q N/A XP-3-2 SBLC Pump P-203B NH-36253 B-4 1.50 CK SA 2 C O/C O/C A DI CM DTJ37 Discharge Check Valve M-127 CTO Q N/A XP-6 SBLC Outboard NH-36253 D-6 1.50 CK SA 1 AC C O/C A LT AJ N/A Containment Isolation Check M-127 Valve CTO RF DTJ 07 CTC CM DTJ 07 XP-7 SBLC Inboard Containment NH-36253 C-6 1.50 CK SA 1 AC C O/C A LT AJ N/A Isolation Check Valve M-127 CTO RF DTJ 07 CTC CM DTJ 07 Valve Table Page 79 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Standby Liquid Control Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ XP-8 Standby Liquid Control NH-36253 C-6 1.50 GT MA 1 B O O P PIT Y2 N/A Injection Line Manual M-127 Isolation Valve Valve Table Page 80 of 81

Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Traversing In-Core Probe Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ TIP-1-1 TIP Inboard Containment GE0719E520 D-5 0.25 BA SO 2 A O/C C A LT AJ N/A Isolation Valve n/a PIT Y2 N/A STC Q N/A TIP-1-2 TIP Shear Valve GE0719E520 D-5 0.25 XP XP 2 D O C A DT SY10 N/A n/a TIP-2-1 TIP Inboard Containment GE0719E520 D-5 0.25 BA SO 2 A O/C C A LT AJ N/A Isolation Valve n/a PIT Y2 N/A STC Q N/A TIP-2-2 TIP Shear Valve GE0719E520 D-5 0.25 XP XP 2 D O C A DT SY10 N/A n/a TIP-3-1 TIP Inboard Containment GE0719E520 D-5 0.25 BA SO 2 A O/C C A LT AJ N/A Isolation Valve n/a PIT Y2 N/A STC Q N/A TIP-3-2 TIP Shear Valve GE0719E520 D-5 0.25 XP XP 2 D O C A DT SY10 N/A n/a Valve Table Page 81 of 81

Monticello Nuclear Generating Plant Revision 0 ATTACHMENT 10 NRC Safety Evaluation Reviews (SER) for IST Program Relief Requests Pending Issuance by the NRC for the Fifth 10-Year Inservice Test Interval

L-MT-12-020, Fifth Ten-Year Lnservice Testing Plan

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L-MT-12-020, Fifth Ten-Year Lnservice Testing Plan

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