ML023370107
| ML023370107 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 11/22/2002 |
| From: | Forbes J Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML023370107 (170) | |
Text
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Operated by: NUCLEAR MANAGEMENT COMPANY 700 1st Street HUDSON, WISCONSIN 54106 Owned by: XCEL ENERGY 414 Nicollet Mall MINNEAPOLIS, MN 55401-1993 MONTICELLO NUCLEAR GENERATING PLANT (Provisional License Granted: September 8 th, (Unit 1) 1970) 2807 WEST HIGHWAY 75 MONTICELLO, MINNESOTA 55362 PUMP & VALVE IN-SERVICE TESTING PROGRAM PLAN Prepared I Reviewed Reviewed Approved FOURTH INTERVAL - REVISION 0 MARCH 9th, 2003 THROUGH MAY 31st, 2012 By:
O9-2c - 02_c2-A CX'Z6'2fj SDean M. SPk t Date STProgram etoa r
Jam jK. Hayward Dt MotcldProgram E }ineeroncel B:Jerome
- 0. Bettle o9Date Mniel Program Engineer Gary C. Park atI Supv, Material Inspection and Repair
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 Table of Contents Section Page
1.0 INTRODUCTION
1.1 Purpose 1
1.2 Scope 2
1.3 Program Bases 5
1.4 References 5
2.0 IN-SERVICE TESTING (IST) PLAN FOR PUMPS 2.1 Pump IST Plan Description 6
2.2 Pump Plan Table Description 6
2.3 -
Pump In-Service Test Requirements 7
3.0 IST PLAN FOR VALVES 3.1 Valve IST Plan Description 8
3.2 Valve Plan Table Description 8
ATTACHMENTS 1
IST Plan System Listing and Acronyms 2
IST Plan Notes 3
Relief Requests and Deferred Testing Justifications (DTJs) Index 4
Pump Relief Requests (PR) 5 Valve Relief Requests (VR) 6 DTJs 7
Pump Tables 8
Valve Tables 9
MNGP Code Class Color Classifications and Drawings (P & IDs)
Program text superscript signifies reference to Attachment 2 IST Plan Notes i
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0
1.0 INTRODUCTION
-1.1 Purpose The Monticello Nuclear Generating Plant (MNGP).Technical Specifications (Section 6.8.G) requires the performance of IST of ASME Code Class 1, 2, and 3 pumps and valves in accordance with the Code of Federal Regulations. 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. However, due to the diversity in component and plant designs and significant differences in the licensing bases among facilities, the OM Code requirements are very general in nature. This leaves significant room for interpretation regarding the implementation of the requirements.
Since the rules for IST were first incorporated into the Code of Federal Regulations in 1976, the ASME Section XI Code (which originally defined IST requirements) has been a work in progress with regard to IST requirements. The ASME Section XI Code requirements changed significantly since the early editions and now continue to change within the OM Code (which has been designated by 10 CFR50.55a as the code now applicable to IST). Additionally, many NRC staff positions regarding IST have changed over time, especially within the last ten years. Since 1976, a myriad of regulatory documents, ASME Code Cases, and ASME Code Interpretations have been issued which impact the scope and implementation of the IST Program. Considering the need for a uniform approach for the development of the IST program scope,-an IST Program Basis Document was developed at MNGP to:
establish consistent guidelines for determining IST Program scope, provide evaluations of plant systems and related components in a consistent manner to determine which components should be included in the IST Program, to determine the extent of testing, and to provide justifications for these determinations, provide evaluations of plant systems and related components in a consistent manner to determine which components may be excluded from the IST Program, and to provide justifications for these determinations.
Page 1 of 12
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 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.
1.2 Scope 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 Design Basis Documents (DBDs). The MNGP IST Program Plan will be in effect through the fourth ten-year interval at MNGP (March 9, 2003 through May 31, 2012). This plan will be updated as required in accordance with 10CFR50.55a(f). The program plan meets the requirements of 10 CFR 50.55a(b) 12 months prior to the fourth interval start date of March 9, 2003, in that it complies with the 1995 Edition, 1996 Addenda of ASME OM Code.
Requirements found in sections of later editions of the OM Code, that have been approved by the NRC in the Code of Federal Regulations, have also been incorporated into the program, as allowed by 10 CFR 50.55(a)(f)(4)(iv) 2 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 the 1995 edition, 1996 addendum OM Code.
The following general guidelines are set forth for evaluating pumps and valves with respect to their inclusion in, or exclusion from, the MNGP IST Program.
The MNGP USAR, regulatory commitments, and related licensing basis or design basis documents (such as docketed design and testing commitments), are the primary references for determining which components perform functions within the scope of the OM Code.
Technical Specifications and several other plant source documents (DBDs, design guides, emergency and abnormal operating procedures, etc.)
identify components that may be important to safe operation of the facility, an enhancement to system reliability, or are operated in conjunction with accident recovery. However, unless specific credit is taken for a component or system in design or licensing basis documents for achieving safe shutdown or mitigating the consequences of an accident, the component need not be included in the IST Program.
Page 2 of 12
Nuclear Management Company, LLC 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.
The basic scope of the IST Program is defined in Subsection ISTA (general requirements) of the OM Code. Per paragraph ISTA 1.1, 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 condition], 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.
USAS B31.1, 1967 edition, was used as the piping construction code for MNGP. Since MNGP piping was not constructed to Section III of the ASME Boiler and Pressure Vessel Code, components were originally neither designed to ASME Code Class 1, 2, and 3 requirements, nor classified as such. The ASME Code classifications of systems and components at MNGP were established only to define components subject to In-Service Inspection and IST requirements. The NRC staff issued Reg. Guide 1.26 to provide guidance on ASME Code classification of components for non-Section III plants. NRC Standard Review Plan, NUREG-0800, Revision 1, dated July 1981, Section 3.2.2, directed licensees to use either Reg. Guide 1.26 or the ANS standards (ANSI/ANS 58.14) for establishing component classifications. Both documents classify components according to the safety functions that they perform during a design bases accident. However, Reg. Guide 1.26 does not cover many components and systems that ma'y perform safety-related functions Page 3 of 12
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 such as emergency diesel Support systems, HVAC systems, and instrument air/nitrogen systems. Also, Reg. Guide 1.26 does not define classification requirements for primary containment penetration piping and containment isolation valves. Therefore, many components that perform safety functions would not be classified as ASME Class 1, 2, or 3 by Reg. Guide 1.26.
NUREG-1482, "Guidelines for In-Service Testing at Nuclear Power Plants," dated April 1995, lists typical safety-related, ASME Code Class systems in boiling water reactors and lists typical components included in IST Programs. However, this NRC guidance is generic in nature. The requirements for classifications of systems and component testing may vary significantly, even between plants of similar design. This is because the licensing bases differ from plant to plant and the lack of standardized plant designs. Additionally, the accident analysis input assumptions and the components credited with active safety functions may differ.
Further, NUREG-1482, "Guidelines for In-Service Testing at Nuclear Power Plants", stipulates the IST Program must state the licensed safe shutdown condition' to which the program has been formulated.
The term "accident" in the scope statement of the OM Code is interpreted at MNGP as to apply to only those postulated design basis accidents identified in Chapter 14 of the USAR. The NRC staff has not provided definitive guidance regarding the events that should be considered "accidents" within the scope of the OM Code.
Consistent with industry practice, components required solely to mitigate the consequences of 1 OCFR50 Appendix R fires and station blackout events are outside the scope of the IST Program since these events are beyond the facility design basis. Beyond design basis events are initiated by multiple (and sometimes complete) failures of safety-related components and systems. The facility design is based on the requirement that each safety system be capable of performing its safety-related functions given a failure of the most limiting active component. Although regulations have been imposed to have the capability to cope with or mitigate these events, they are outside the scope of the facility design basis accident analyses. Components whose safety functions are solely to mitigate these events are not required by regulations to be classified as safety-related. These components are considered QA scope per the MNGP Q list, are classified as Augmented Quality (AQ) components and are outside the scope of the IST Program. AQ componients which are not classified as Code Class 1, 2 or 3, but do perform a function important to safety are within the scope of the program.
Page4 of 12
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 Per OMa-1996, skid-mounted 3 pumps and valves are excluded provided they are tested as part of the major component (valve assembly,.turbine, engine, etc.) and are justified by the Owner to be adequately tested.
1.3 Program Basis 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 1ST. All ASME Code Class 1, 2, and 3 components were evaluated for inclusion in the IST Program. In some cases, the IST Program also identifies MNGP defined augmented testing requirements for NC components that perform functions determined to be important to 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 and database 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 ISTC Section 1.4. Pumps included in the IST Program were identified as either centrifugal, vertical line shaft or positive displacement ini accordance with ISTB Section 1.1 and then grouped as either Group A or Group B in accordance with ISTB Section 1.3.
Subsequent to determining component safety function, classification and categorization, ISTB, ISTC and Appendix I 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.
1.4 References MNGP Technical Specifications MNGP Updated Safety Analysis Report 10CFR50.55a(f) 10CFR50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Plants ASME OM Code, 1995 Edition and 1996 Addenda, "Code for Operation and Maintenance of Nuclear Power Plants" NRC Generic Letter No. 89-04, "Guidance on Developing Acceptable In-Service Testing Programs" Page 5 of 12
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 NRC NUREG-1482, "Guidelines for In-Service Testing at Nuclear Power Plants" MNGP Piping and Instrument Diagrams NRC Safety Evaluation Report, "SafetyEvaluation (SE) of Relief Requests and Action Item Responses for the Pump and Valve In-Service Testing Program (TAC Nos. M88972 And M82638)," dated September 9, 1994 NRC Safety Evaluation Report, "Safety Evaluation (SE) of Relief Request for the Pump and Valve In-Service Testing Program (TAC No. M903410)," dated December 8, 1994 Nuclear Management Corporation, In-Service Testing Standard, Rev. 0 2.0 IST PLAN FOR PUMPS 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 brinfing the reactor to the safe shutdown' condition, maintaining the safe shutdown condition, or mitigating the consequences of an accident.
I This program plan meets the requirements of the 1995 Edition with 1996 Addenda OM Code Subsection ISTB, with the exception of specific relief requests contained in Attachment 4 and incorporated sections from later OM Code editions 2 2.2 Pump Plan Table Description The pumps included in the MNGP Nuclear Generating Station IST Plan are listed in Attachment 7. 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 Instrumentation Diagram on which the pump is depicted Page 6 of 12
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 2.2.4 Class - The ASME Code classification ofthe pump (1, 2, 3, or NC for ASME non-Code Class) 2.2.5 Group - The ISTB pump group A - Group A (those pumps ini standby systems that are operated continuously or routinely during normal operation, safe shutdown', or refueling operations)
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 QIY2 -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 that 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 5.4.
2.3.2 Test Parameters Note: Test Parameters are stated per TableISTB 4.1-1.
Speed (N) - Pump speed is a reference variable for variable speed pumps Page 7 or 12
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 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 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.
Vibration (V) - Pump bearing vibration is a trended variable measured at a given reference value during pump testing (in/sec peak velocity) 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 shutdowni condition, in maintaining the safe shutdown' 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 and Appendix I of-the 1995 Edition, 1996 Addenda OM Code with the exception of the specific relief requests contained in Attachment 5 or specified sections adopted from later OM Code editions2 Where the frequency requirements for valve testing have been determined to be impracticable, cold shutdown or refueling outage 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 8. 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 1.2 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.
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Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 3.2.3 Description - Functional/noun name for the valve.
3.2.4 P&ID - The Piping and Instrumentation 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:
' cK Check BF 3
Butterfly iGT Gate
'GL Globe RV Relief RD Rupture Disk PL Plug XP Explosive AR Air Relief FV Excess Flow Check 3.2.8 Act - The valve actuator type is indicated by the following abbreviations:
[Actuaeto <6 Desefiptibif
-I 1,MO I Motor Operated IAO
!Air.Operated
!SO Solenoid Operated IMA IManual 1SA
'SelfActuated 3.2.9 Class - The ASME Code non-Code Class).
classification of the valve (1, 2, 3, or NC for ASME Page 9 of 12 WAIV67W66 1Desb7ff6tio'_n'
.'I
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 3.2.10 Cat - The category assigned to the valve per the definitions of ISTC, 1.4.
3.2.11 Norm Pos - The position(s) of the valve during normal power operations is indicated as follows:
P 6sifi 'o
- Des&iifi-o'ti 0
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:
10 1Opeh 1
C IClosed O/C Open and Closed 3.2.13 A/P - Active or Passive valve function as indicated below:
A/P D6&.WciptftV A
Active P
Pasiv Page 10 of 12 4
Descrintion
A Manual or power-operated valves for which seat leakage is limited to a specific amount in the closed position for fulfillment of their required safety function.
B Manual or power-operated valves 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)
BC Valves that are self-actuating but are also equipped with power actuators that are credited with remote operating capability.
AC Check valves for which seat leakage is limited to a specific amount in the closed position for fulfillment of their required safety function.
D Actuated by an energy source capable of only one operation (e.g., rupture disks and explosive actuated valves).
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 3.2.14 Test - Th'e tests performed to fulfill the requirements of ISTC, Section 4 as indicated below.
ILT Category A Seat Leakage Test Category A Containment, Pressure Isolation Valves(CIV, PIV), & Own7er Specified STO Category A/B Exercise Test Category A/B Power Operated & Manual Open Valves STC Category A/B Exercise Test Category A/B Power Operated & Manual Closed Valves 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 4.7 EVs (TIP & SBLC): ISTC 4.6 PIT Position Indication Test Passive/Active Valves with Remote Position Indication RT Relief Valve Test ASME Class 1: Appendix I, Section 1 3.3.1 ASME Class 2, 3 & Augmented: Appendix I, Section 1 3.3.5 ASME Class 2 & 3 Vacuum Relief Valves:
Appendix I, Section 1 3.3.7 FO Fail Safe Open Test Category A/B Power Operated Valves with Fail-Safe Function 1FC Fail Safe Closed Test Category AJB Power Operated Valves with Fail-Safe Function DI Disassemble/Inspect Check Valves (Examination) Test XT Excess Flow Check Valve Test Excess-Flow Check Valves Page 11 of 12
Nuclear Management Company, LLC 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 Section 4. The following abbreviations are used for test frequencies:
Q Quarterly
- CS
- Cold Shutdown............
S.
- GLCS
- RF
.efueling Y2 Every Two Years
'k'"....Ev
- T o er.....
Y5 Every Five Years AJ.Per Appendix J Program SY5 Class 1 Safety and 'R'elief Valves:
- "Once per 5 yrs minimum & Group Sampling every 24 months
.Sy10 Class 2, 3 and Augmented Safety and Relief Valves:
.Once per 10 yrs mni & Group Sampling every 48 months
- 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.
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".
Page 12 of 12
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 ATTACHMENT 1 In-Service Plan System Listing & Acronyms SYSTEM ACRONYM COMMON USE (Site 3-Letter 11D)
ACRONYM Alternate N2 AN2 Auto Pressure Relief APR SRV Combustible Gas Control CGC CGCS 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 I
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 ATTACHMENT 2 In-Service Testing Plan Notes (Scripted as superscript in IST Plan Document as applicable)
- 1. MNGP is licensed to meet the Hot Shutdown Plant Condition for Safe Shutdown.
- 2. As desired, MNGP may use selected portions of the latest OM Code when published in the Code of Federal Regulations. Relief is not required to adopt portions of later Code editions and addenda provided all related requirements as they pertain are met. Technical Positions will be developed to document these additions in the MNGP IST Program (and/or Administrative related documents).
"i.
The 1998 OM CODE w/ 1999 Addenda is adopted for manual valve exercising in accordance with ISTC 3540 with the required "modification to test exercise frequency as published in the Volume 67 No. 187 of the Federal Register on September 26, 2002. Section 50.55a(b)(3)(vi) is revised to clarify that the interval for exercising manual valves may not exceed 2 years. Adverse conditions have not been observed which would require more frequent test exercising.
MNGP adopts ISTC 3540 wholly for test exercising IST Program manual valves at a frequency not to exceed two years.
- 3. Skid-Mounted Components will be detailed in the MNGP IST Program (and/or Administrative related documents) by Technical Position.
Page 1 of1
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 ATTACHMENT 3 In-Service Testing Plan Relief Requests and Deferred Testing (DTJs) Justification Index Pump Relief Requests PR 01 Test Methods PR 02 Instrument Range PR 03 HPCI Pump Vibration PR 04 Instrument Range PR 05 SBLC Pump Vibration Frequency Response Range Valve Relief Requests VR 01 CRD-114 Closure Testing VR 02 RHRSW Flow Control Valve Exercising Deferred Testing Justifications 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 DJT 04 RHRSW-57-1,-RHRSW-57-2, RHRSW-57-3, RHRSW-57-4 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 DTJ 14 MO-2-53A, MO-2-53B DTJ 15 AO-23-18, AO-13-22 DTJ 16 MO-2026, MO-2027, MO-2029, MO-2030 DTJ 17 HPCI-9, HPCI-10, HPCI-14, HPCI-15, HPCI-31, HPCI-65, HPCI-71 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 DTJ 21 BF-12, BF-14, BF-24, BF-26, BF-35, BF-37, BF-46, BF-48 Page 1 of 2
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 Deferred Testing Justifications (cont)
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-610-1 through AI-610-4, AI-611, AI-612, AI-13-4, AI-13-7 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 I
DTJ 35 AI-243-1, AI-243-2, AI-244-1, AI-244-2 DTJ 36 CST-189, CST-96, CST-98 DTJ 37 XP-3-1, XP-3-2
-Page 2 of 2
Nuclear Management Company, LLC Monticello Nuclear Generating Plant ATTACHMENT 4 Pump Relief Requests (PR).
Page 1 of 11 Revision 0
Nuclear Management Company, LLC Monticello Nuclear Generating Plant PUMP RELIEF REQUEST
- NUMBER: PR 01 System: Standby Liquid Control (SLC)
Pumps: P-203A and P-203B P&ID: NH-36253 (M-127)
Class: 2 Pump Group: B Function: To inject liquid poison into the reactor.
Impractical Test Requirement:
ISTB 4.7.5; Flow Rate; When measuring flow rate, a rate or quantity meter shall be installed in the pump test circuit.,
ISTB 5.6.3; Comprehensive Test; After pump conditions are as stable as the system permits, each pump shall be run at least 2 min. At the end of this time at least one measurement or determine of each of the quantities required by Table ISTB 4.1-1 shall be made and recorded.
Basis for Relief: The positive displacement SBLC pumps are designed to pump a constant flow rate regardless of system resistance. The SBLC 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.
Relief is requested in accordance with 10 CFR 50.55a(a)(3)(ii). The Code requirements to use flow rate instrumentation and a two-minute test duration are considered a burden, which would result in a hardship without a corresponding increase in the level of quality or safety.
Alternative Testing: Determine pump flow rate by measuring changes in tank level over time. The pump will be started with suction from the condensate storage system and will discharge to the test tank. The test tank level is approximately the same at the beginning of each test to ensure repeatability. After approximately 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) = 'AL (In)/At (Sec)
Page 2 of11 Revision 0
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Where T is a constant which reflects tank dimensions and unit conversions. 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.
Status: Approval of similar relief was previously granted to Monticello by the NRC on July 6, 1993.
Page 3 ofl1 Revision 0
Nuclear Management Company, LLC Monticello Nuclear Generating Plant PUMP RELIEF REQUEST NUMBER: PR 02 System: Residual Heat Removal (RHR)
Pumps: P-202A/B/C/D, 11/12/13/14 RHR Pumps P-109A/B/C/D, 11/12/13/14 RHRSW Pumps P&ID: NH-36246 (M-120), NH-36247 (M-121)
Class: 2 and 3 Pump Group: A Function: Provides a flow signal to an indicating device.
Impractical Test Requirement:
ISTB 4.7. l(b)(1); Full scale range of each analog instrument shall not be greater than three times the reference value.
Basis for Relief: Flow transmitters FT-10-11 1A, FT-10-1111B, 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 instrument's 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 as follows:
Instruments FT-10-111IA FT-10-111B FT-10-97A FT-10-97B Pumps
- P-202A P-202C P-202B P-202D P-109A P-109C P-I09B P-109D Instrument Span (Range) 4-20 mA 10-50 mA Equivalent Reference Value 6.4 mA 18.37 mA 18.57 mA Range to Reference Value Ratio (16/2.4)
= 6.67 (40/8.37)
= 4.78 (40/8.57)
= 4.67 Page 4 of 11 Revision 0
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 NOTE 1:
The transmitters FT-10-111A, FT-10-111iB, and FT-10-97A output signal is 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 4-2OmA 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.
NOTE 2:
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.
Relief is requested in accordance with 10 CFR 50.55a(a)(3)(i). The alternative testing described below provides an acceptable level of quality and safety because the variance of flow instrumentation is within the maximum variance allowed in Table ISTB 4.7.1-1, Required Instrument Accuracy, and will ensure the intent of the Code is met.
Alternative Testing: 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 4.7.1-1, 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.
Status: Approval of similar relief was previously granted to Monticello by the NRC on July 6, 1993.
Page 5 of 11
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 PUMP RELIEF REQUEST NUMBER: PR 03 System: High Pressure Coolant Injection (HPC)
Pump: P-209 P&LD: NH-36250 (M-124)
Class: 2 Group: B Function: Injects coolant into the reactor vessel independent of AC power.
Impractical Test Requirement:
ISTB Table 5.2.1-1; Comprehensive pump test vibration alert limit of 0.325 in/Sec ISTB 6.2.1; Acceptance Alert Range for the horizontal vibration data points and the resulting increased pump test frequency.
Basis for Relief: The 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. 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 0.415 in/Sec. Application of a 0.325 in/Sec alert limit would require MNGP to enter accelerated test frequency each time the pump was tested because one or more of these points measured would exceed this limit.
Prior to the third ten-year interval, the alert limit of 0.325 in/Sec was not a code requirement at Monticello. MNGP has many years of In-Service test data showing that baseline vibrations at 0.415 in/Sec relresent acceptable pump operation and that vibration levels have not trended up. MNGP has had these vibration levels analyzed by an Engineering Consultant that specializes in vibration analysis. Their analysis shows that this pump can operate at vibration levels up to 0.700 in/Sec.
EPIX and NPRDS (Jan 1, 1991 to present) component 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 MNGP 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 would be operating at its normal vibration Page 6 ofll"
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 range and no change would be seen. The additional annual test would require a significant amount of time and resources and only create additional maintenance due to normal wear of the system., Modifications to try and 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.
Relief is requested in accordance with 10 CFR 50.55a(a)(3)(ii). Compliance with the specified Code requirements of the listed sections would result in a hardship without a corresponding increase in the level of quality or safety.
Alternative Testing: A vibration alert limit of 0.500 in/Sec will be used for the pump horizontal vibration data points. The ISTB Table 5.2.1-1 required action limit of 0.700 in/Sec will be adhered to.
Status: Approval of similar relief was previously granted to Monticello by the NRC on September 9, 1994.
Page 7 of 11
Nuclear Management Company, LLC Monticello Nuclear Generating Plant PUMP RELIEF REQUEST NUMBER: PR 04 System: High-Pressure Coolant Injection (HPC)
Reactor Core Isolation Cooling (RCI)
Pumps: P-209 HPCI and P-207 RCIC P&mD: NH-36250 (M-124), NH-36252 (M-126)
Class: 2 Pump Group: B Function: To inject coolant into the reactor independent of AC power.
Impractical Test Requirement:
ISTB 4.7. 1(b)(1), analog instrument range shall not exceed 3 times the reference value.
Basis for Relief: 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 mAmp 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 as follows:
Instrument PI-23-116 (See NOTE 1)
PT-23-100 (See NOTE 2)
PI-13-66 (See NOTE 1)
Pump Range P-209 30" Hg-100 PSI P-209 l0to50mAmps P-207 30" Hg - 100 PSI Reference Ratio 33.7 PSI 114.7133.7 = 3.4 11.8 mAmps 40/11.8 = 3.4 33.7 PSI 114.7/33.7 = 3.4 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 rafnge = 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 mAmp range, or a span of 40 mAmps.
The ratio for this instrument must be determined by reducing the reference value to its value on the 40 mAmp span.
Page 8 of 11 Revision 0
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 The code requires the instrument range to be less than 3 times the test parameter reference value. The 2% code allowable instrument tolerance is then taken from this range requirement. The same instrument calibration tolerance can be applied to these instruments by simply calculating the code-required tolerance from the code equivalent range as follows:
Instrument Reference Value Code Equivalent Range 2% of Code Equivalent Range PI-23-116 33.7 psi 3 x 33.7 = 101 psi
+/- 2 psi PT-23-100" 21.8 mAmps 3 x 11.8 = 35.4 mAmps
+/-0.7 mAmps PI-13-66 33.7 psi 3x33.7=101psi
+/-2 psi
- 21.8 mAmps equates to 11.8 mAmps on the 40 mAmp span The current instrument calibration tolerances are +/- 2 psi for the pressure indicators and
+/- 0.7 mAmps 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.
Relief is requested in accordance with 10 CFR 50.55a(a)(3)(i). The alternative testing described below provides an acceptable level of quality and safety.
Alternative Testing: The instruments identified above will be calibrated to 2% of a code equivalent range. The code equivalent range will be calculated by multiplying the current test parameter reference value by three. For pressure indicators PI-23-116 and PI-13-66 this currently results in a allowable tolerance equal to +/- 2 psi on the output or +/- 0.7 mAmps on the input (Note: PI-23-'116 may be calibrated by determining the mAmps input signal required to establish a set PSI output).I Pressure transmitter PT-23-1 00 currently is calibrated to +/- 0. 7 mAmps for the reasons discussed above.
Status: Approval of similar relief was previously granted to Monticello by the NRC on December 8, 1994.
Page 9 of 11
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 PUMP RELIEF REQUEST NUMBER: PR 05 System: Standby Liquid Control (SLC)
Pumps: P-203A and P-203B P&ID: NH-36253 (M-127)
Class: 2 Pump Group: B Function: To inject liquid poison into the reactor.
Impractical Test Requirement:
ISTB 4.7. 1(f); General, Frequency Response Range; 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 H-z.
Basis for Relief: The nominal shaft rotational speed of these pumps is 280 RPM, which is equivalent to approximately 4.7 Hz. Based on this frequency and ISTB 4.7. l(f), the required frequency response range of instruments used for measuring pump vibration are 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 equipment cost.
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 are 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 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 frequencynor 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 4.7.1 (f) would serve no significant purpose.
The significant modes of vibration, with respect to equipment monitoring, are as follows:
Page 10 of11
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 a
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.
a 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.
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.
Relief is requested in accordance with 10 CFR 50.55a(a)(3)(ii). Compliance with the specified Code requir6ments of the listed section would result in a hardship without a corresponding increase in the level of quality or safety, Alternative Testing: Vibration levels of the Standby Liquid Control Pumps will be measuied in accordance with the applicable portions of Subsection ISTB with the exception of the lower frequency response limit for the instrumentation listed in Paragraph 4.7.1 (f). The frequency response range for vibration measurement for the Standby Liquid Control Pumps shall be 4 to 1000 Hz.
Status: New Relief Request submitted for 4 th Ten Year Code Interval starting 3/9/03.
This Monticello Relief Request (PR'05) is similar to a Relief Request that was approved by the NRC for the Duane Arnold Energy Center on March 4, 1994. Monticello's pumps are similar to DAEC's pumps, with only small differences in rotational speed and flow rate.
Page 11 of11
Nuclear Management Company, LLC Monticello Nuclear Generating Plant ATTACHMENT 5 Valve Relief Requests (VR)
Page 1 of 6 Revision 0
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 VALVE RELIEF REQUEST NUMBER: VR 01 System: Control Rod Drive Hydraulics (CRH)
Valves: CRD-1 14 (121 valves total, one per HCU)
P&LD: NH-3 6245 (M-119)
Category: C Class: 2 Function: To open during a reactor scram, providing a flow path for the exhaust water from the CRDs to the Scram Discharge Volume.
Impractical Test Requirement: ISTC 4.5.1 Exercising Test Frequency requirement to exercise nominally every 3 months.
Basis for Relief: ISTC 4.5.2 requires that each check valve tested include an open and close test. In order to complete the open and close tests, both test directions shall be K>.
completed for each valve at a frequency no greater than each refueling outage.
The subject check valves are a simple ball-check design. There are no internal parts in the check valves that are susceptible to rapid degradation and sudden failure. 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. Individually testing these valves would require the associated control rod to be fully inserted for the duration of the test. Additionally, performing the close test on all 121 valves each refueling outage is a hardship that is not offset by a compensating increase in the level of quality and safety.
The volume of testing results in a substantial burden on plant resources via the expenditure of person-hours and person-REM to perform the associated system filling, venting and testing tasks.
Furthermore, the valves are welded into the line and it is not practical to perform a disassembly and inspection of each valve in accordance with ISTC 4.5.4(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 aids such as a boroscope.
Relief is requested in accordance with 10 CFR 50.55a(a)(3)(ii). Compliance with the specified Code requirements of the listed section would result in a hardship without a J jcompensating increase in the level of quality or safety.
Page2 of 6
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 Alternative Testing: ISTC 4.5.4(c) provides a means of demonstrating the necessary valve obturator movement by the grouping of valves and then performing a disassembly and inspection for some number (sample) of valves each refueling outage.
All 121 of these check valves are identical in all essential iespects and will be considered a single group for a similar sampling approach for exercise to the closed position.
Similar to the requirement of ISTC 4.5.4(c)(3), at least one of these check valves will be exercised closed each refueling outage. This approach is consistent with valves that must be disassembled and inspected to verify proper valve function when testing is not practicable. Testing of the valves in this manner adequately demonstrates valve performance capability and provides a means to monitor for valve degradation in a like manner to disassembly and examination.
Status: New Relief Request submitted for 4e Ten Year Code Interval starting 3/9/03.
Page 3 of 6
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 VALVE RELIEF REQUEST NUMBER: VR 02 System: RHR Service Water (RSW)
Valves: CV-1728 and CV-1729 P&mD: NH-36247 (M-121), NH-36664 (M-1 12)
Category: B Class: 3 Function: To open, providing a flow path for RHRSW through the RHR heat exchanger.
Impractical Test Requirement:
ISTC 4.2.4, Power Operated Valve Stroke Testing.
ISTC 4.2.8, Stroke Time Acceptance Criteria ISTC.4.2.9, Corrective Action Basis for Relief:
IOCFR Part 50, Section 50.55a(f) (5) and (6) states, (in part):
(5)(iii) If the licensee has determined that conformance with certain code requirements is impractical for its facility, the licensee SHALL notify the commission....
(6)(i)... The commission may grant relief and may impose alternative requirements...
giving due consideration to the burden upon the licensee....
In addition, IOCFR Part 50, Section 50.55a(a)(3) states (in part):
"Proposed alternatives to the requirements of paragraphs... may be used when... The proposed alternatives would provide an acceptable level of quality and safety..."
ISTC 4.2.4 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 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 1.3 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 Page 4 of 6
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 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, 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 by manual adjustment of the DPIC setpoint. The valve positioner positions the valve and modulates the valve position as necessary to maintain this control point. Stroke time testing of these valves on quarterly basis is not consistent with the design of the valve's control scheme and is not in the interest of plant safety.
These valves are interlocked to receive a closed signal when the R-RSW pumps are de energized. This interlock is provided to ensure that system water inventory is not lost during system shutdown. Stroke time testing of valves CV-1728 and CV-1729 when the R-IRSW 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 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 an undesirable burden on plant resources via the expenditure of person-hours and person-REM to perform system filling and venting.
Stroke time testing 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 during pump operation would require the valve 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 valve to perform stroke time testing 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.
Proper stroke time testing would require the plant to modify the control logic of the valves. This hardship 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 4.2.8 for establishing stroke time acceptance criteria is impractical. Similarly, if there are no stroke time limits applicable, then the requirement of ISTC 4.2.9 for corrective action when stroke time limits are exceeded is not applicable.
Page 5 of 6
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 Alternative Testing: ISTC 4.2.3 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 R{RSW 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. Testing of the valves in this manner demonstrates valve performance capability and provides a means to monitor for valve degradation.
Status: Approval of similar relief was previously grantedto Monticello by the NRC on August 25, 1995.
Page 6 of 6
Nuclear Management Company, LLC Monticello Nuclear Generating Plant ATTACHMENT 6
- Deferred Testing Justifications (DTJ)
Page 1 of 42 Revision 0
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 01 System: Reactor Building Closed Cooling Water (RBC)
Valve: RBCC-15 P & ID: NH-36042-2 (M-1 11-1)
Code Class: 2 Category: A/C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1 Function/Description: Inboard (located outside of containment) Containment Isolation Valve for RBCCW supply to the drywell which is 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 Priniary 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 4.5.1. 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 4.5.2(c).
Page 2 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION "NUMBER: DTJ 02 System: Condensate Storage (CST)
Valve: CST-88, CST-90, CST-92, CST-94 P & ID: NH-85509 (M-114-1)
Code Class: 2 Category: C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1 Function/Descriptio'n: 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 4.5. 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 a hardship without a compensating increase in safety. 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 4.5.4(c). At least one valve from the group will be disassembled and examined each refueling outage for the open and close exercise test.
Page 3 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 03 System: Reactor Recirculation (REC)
Valve: XR-27-1, XR-27-2, XR-25-1, XR-25-2 P & ID: NH-36243-1 (M-1 17), NH-36244 (M-1 18)
Code Class: 2 Category: A/C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1.,
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 4.5.1. These valves are the inboard and outboard Primary Containment isolation valves for the CRD injection flow to the Recire 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 4.5.2(c),
Page 4 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 04 System: RHR Service Water (RSW)
Valve: RHRSW-57-1, RHRSW-57-2, RHRSW-57-3, RHRSW-57-4 P & ID: NH-36665 (M-811)
Code Class: 3 Category: C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5,1 Function/Description: The RHRSW Pump Motor Cooling Check Valves'open to allow cooling water flow to the respective RHR SW Pump Motor and close to minimize drain down of the piping.. They close to ensure cooling water is available immediately to the motor heat exchangers when flow is restored.
Frequency Change Justification: It is not practical to exercise close these valves every 3 months per the requirements of ISTC 4.5.1. These valves are tested by performing equipment isolation, isolating and removing one RHRSW Pump from service and installing temporary bypasses (hoses, gages, etc.) to accomplish the required testing.
Connection of test equipmeht and performance of a reverse flow or leak type test to exercise each of these check valves closed makes one of the RHRSW pumps inoperable.
Additionally, manual operator actions are required to restore the pump to a functional status if an accident occurred while the test is in progress Alternate Test Frequency: Test exercise the valves closed at a cold shutdown frequency in accordance with ISTC 4.5.2(b).
Page 5 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 05 System: Control Rod Hydraulics (CRH)
Valve: CRD-1 15 (121 valves total, one per HCU)
P & ID: NH-36245 (M-119)
Code Class: 2 Category: C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1.
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 K>*
control unit charging water inlet check valves every 3 months per the requirements of ISTC 4.5.1. 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 4.5.2(c).
Page6 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 06 System: Residual Heat Removal (RHR)
Valve: RHR-8-1, RHR-8-2 P & ID: NH-36247 (M-121), NH-36246 (M-120))
Code Class: 2 Category: C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1 Function/Description: The RHR minimum flow check valves open to the torus to provide minimum flow recirculation from the RHR pumps. They close to isolate the minimum flow line from the test flow return line.
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 4.5.1. There is
",K-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.
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 4.5.4(c). At least one valve from the group will be disassembled and examined'each refueling outage for the open and close exercise test.
Page 7 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 07 System: Standby Liquid Control (SLC)
Valve: XP-6, XP-7 P & ID: NH-36253 (M-127)
Code Class: 1 Category: A/C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1 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 4.5.1. 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 4.5.2(c).
Page 8 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant
-Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 08 System: Reactor Water Clean-up (RWC)
Valve: RC-6-1, RC-6-2 P & ID: NH-3625A (M-128)
Code Class: 2 Category: C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1 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 HPCIiRCIC injection.
Frequency Change Justification: It is not practical to exercise closed the RWCU return "check valves every 3 months per the requirements of ISTC 4.5.1. 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 4.5.2(c)'
Page 9 of 42
Nuclear Management Company, LLC Monticello Nuclear Geneiating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 09 System: Instrument and Service Air (AIR)
Valve: AI-626-1 and AI-625 P & ID: NH-36049-14 (M-131)
Code Class: 2 Category: A/C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1 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 4.5.1. Check valves Al 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 4.5.2(c).
Page I0 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 10 System: Main Condenser (CDR)
Valves: AO-1825A, AO-1825B P & ID: NH-36035-2 (M-104-1)
Code Class: NC Category: B Code Frequency: Category B exercise test frequency of every 3 months per ISTC. 4.2.1.
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 4.2.1.
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 4.2.2(c). Fail-safe test valves closed at a cold shutdown frequency in accordance with ISTC 4.2.6.
Page II of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 11 System: Reactor Building Closed Cooling Water (RBC)
Valves: MO-1426, MO-4229, MO-4230 P & ID: NH-36042-2 (M-1 11-1)
Code Class: 2 Category: A Code Frequency: Category A exercise test frequency of every 3 months per ISTC 4.2.1.
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 4.2.1.
Interrupting RBCCW flow to Primary Containment and the heat loads there (including Reciiculation 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 4.2.2(c).
Page 12 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 12 System: Condensate and Feedwater (CFW)
Valves: 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: C (FW-91-1, FW-91-2)
A/C (FW-94-1, FW-94-2, FW-97-1, FW-97-2)
Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1 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 4.5.1. 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 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 4.5.2(b).
Page 13 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 13 System: Main Steam (MST)
Valves: AO-2-80A through D P & ID: NH-36241 (M-115)
Code Class: 1 Category: A Code Frequency: Fail Safe Test per ISTC 4.2.6.
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 4.2.6. 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 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 4.2.6.
Page 14 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 14 System: Reactor Recirculation (REC)
Valves: MO-2-53A, MO-2-53B P & ID: NH-36243 (M-1 17)
Code Class: I Category: B Code Frequency: Category A/B exercise test frequency of every 3 months per ISTC 4.2.1 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 4.2.1. 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 4.2.2(c).
Page 15 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 15 System: High Pressure Coolant Injection (HPC)
Reactor Core Isolation Cooling (RCI)
Valves: AO-23-18, AO-13-22 P & ID: NH-36250 (M-124),NI-I-36252 (M-126)
Code Class: 2 Category: C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1 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 and closed the HPCI and RCIC Injection check valves every 3 months per the requirements of ISTC 4.5.1. 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 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 I piping systems.
Since the open test can only be performed by disassembly and examination for these subject valves, ISTC 4.5.4(c) allows this test per refueling outage.
Alternate Test Frequency: Test exercise AO-23-18 and AO-3-22 open and closed via disassembly and examination at a refueling outage frequency in accordance with ISTC 4.5.4(c). These valves are each in a singular test group.
Page 16 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 16 System: Residual Heat Removal (RHR)
Valves: MO-2026, MO-2027, MO-2029, MO-2030 P & ID: NH-36247 (M-121)
Code Class: 1 Category: AK Code Frequency: Category A exercise test frequency of every 3 months per ISTC 4.2.1 Function/Description: Reactor Head Spray and Shutdown Cooling (SDC) Isolation Valves open to allow RHR flow for reactor vessel head cooling and to allow flow to the RIR pumps for SDC, respectively.. These RHR valves are normally closed and have a closed safety position (exception: MO-2027 has open safety position as well, see below).
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 4.2.1. 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.
MO-2027 has an open safety position to limit he differential pressure created due to the thermal expansion of the liquid volume between MO-2027 and MO-2026 and the reactor vessel. This open position is actually a small deflection of the outboard disk (MO-2027 is a double disk gate valve) from the body seat to allow enough fluid to pass to avoid exceeding allowed stresses in the penetration piping. Performing the disk deflection exercise requires connection of a hydrostatic pump test rig and performing a test similar to that done for seat leakage.
Alternate Test Frequency: Test exercise valves closed on a cold shutdown frequency in accordance with ISTC 4.2.2(c). Test exercise MO-2027 open at a cold shutdown frequency for GL-96-06 commitments.
Page 17 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 17 System: High Pressure Coolant Injection (HPC)
Valves: HPCI-9, HPCI-10, HPCI-14, HPCI-15, HPCI-31, HPCI-65, HPCI-71 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)
A/C (HPCI-9, HPCI-10)
,Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1 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 funmcion 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-3 1) 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.
Frequency Change Justification: It is not practical to exercise closed the subject check valves every 3 months per the requirements of ISTC 4.5.1. 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 4.5.1. 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-15, HPCI-65 and HPCI-71 is by a similar test using a flow and/or pressure source. HPCI-31 must be tested Page 18 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 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 comp&nsating 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 and HPCI-10 at a cold shutdown frequency in accordance with ISTC 4.5.2(b). Test exercise HPCI-14, 15, 65 and 71 open and closed at a cold shutdown frequency in accordance with ISTC 4.5.2(b).
Test exercise HPCI-31 open and closed at a refuel outage frequency by disassembly and examination in accordance with ISTC 4.5.4.(c).
Page 19 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 18 System: Reactor Core Isolation Cooling (RCI)
Valves: 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: Category C exercise test frequency of every 3 months per ISTC 4.5.1 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 4.5.1. It is not practical to exercise open check valves RCIC-3 1, RCIC-57 and RCIC-59 every 3 months per the requirements of ISTC 4.5.1. 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 6nly way to test open RCIC-57 and RCIC-59 is by a similar test using a flow and/or Page 20 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 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 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 4.5.2(b). Test exercise RCIC-16 and 17, closed at a cold shutdown frequency in accordance with ISTC 4.5.2(b). Test exercise RCIC-57 and RCIC-59 open and closed on a cold shutdown frequency in accordance with ISTC 4.5.2(b). Test exercise RCIC-31 open and closed at a refuel outage frequency by disassembly and examination in accordance with ISTC 4.5.4(c).,
Page 21 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 19 System: Alternate N2 (AN2)
Valves: 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 - Category C exercise test frequency of every 3 months per ISTC 4.5.1 SV-4234 - Category A/B exercise test frequency of every 3 months per ISTC 4.2.1 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 MSIV's 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 4.5.1 or ISTC 4.2.1 (SV-4234) 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 MSIV's fail closed, any drop in pneumatic supply pressure to them during power operation risks a plant transient from the MSIV's 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 4.5.2(b). Test exercise AI-730 open at a cold Page 22 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 shutdown frequency in accordance with ISTC 4.5.2(b). Test exercise AI-598 and AI-708 open and closed by disassembly and examination at a refueling outage frequency in accordance with ISTC 4.5.4(c). Test exercise SV-4235 open and closed at a cold shutdown frequency in accordance with ISTC 4.2.2(c).
Page 23 of 42
Nuclear Management Company, LLC "Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 20 System: Post Accident Sampling (PAS)
Valves: PAS-59-5, PAS-59-6 P & ID: NH-96042-1 Code Class: 2 Category: C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1 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 4.5.1.
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 4.5.2(b).
Page 24 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 21 System: Control Rod Drive Hydraulics (CRH)
Valves: 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: Category C exercise test frequency of every 3 months per ISTC 4.5.1.
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. 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 4.5.2(c).
Page 25 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 22 System: Various Valves: XFV-1 through 89 P & ID: Various Code Class: 1 and 2 Category: C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1.
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. 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 4.5.2(c) (exception: as described above, bi-directional testing is not required to be performed).
Page 26 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 23 System: Instrument and Service Air (AIR)
Valves: AI-571 P & ID: NH-36049-12 (M-131)
Code Class: 2 Category: A/C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1.
Function/Description:: Instrument Air/Nitrogen pneumatic supply to dryweli 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. 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 4.5.2(b).
Page 27 of 42
Nuclear Management Company, LLC Monticello -Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 24 System: Instrument and Service Air (AIR)
Valves: 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: Category C exercise test frequency of every 3 months per ISTC 4.5.1.
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 nitiogen) check'valves every 3 months per the requirements of ISTC 4.5.1. 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 4.5.2(c).
Page 28 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 25 System: Various Valves: AI-610-1, AI-610-2, AI-610-3, AI-610-4, AI-61 1, 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)
NIH-36252 (M-126)
Code Class: NC Category: A/C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1.
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 4.5.1. 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-61 1, 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 4.5.2(c).
Page 29 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 26 System: Residual Heat Removal (RHR)
Valves: AO-10-46A, AO-10-46B P & ID: NH-36246 (M-120)
NH-36247 (M-121)
Code Class: 1 Category: A/C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1.
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 4.5.1. 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 a hardship 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 4.5.2(b). Test exercise valves closed at a refueling outage frequency in accordance with ISTC 4.5.2(c).
Page 30 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 27 System: Core Spray (CSP)
Valves: AO-14-13A, AO-14-13B P &-ID: NH-36248 (M-122)
Code Class: 1 Category: A/C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1.
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 4.5.1. 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 a hardship 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 4.5.2 Page 31 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 28 System: Primary Containment (PCT)
Valves: 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: Category C exercise test frequency of every 3 months per ISTC 4.5.1.
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 4.5.1. 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 and constitutes a hardship 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 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 4.5.2 (c) and 4.5.4(b).
Page 32 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 29 System: High Pressure Coolant Injection (HPC)
Reactor Core Isolation Cooling (RCI)
Valves: HPCI-32 RCIC-41 P & ID: NH-36250 (M-124)
NH-36252 (M-126)
Code Class: 2 Category: C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1.
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 4.5.1. 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 LCO 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 4.5.2(b).
Page 33 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 30 System: High Pressure Coolant Injection (HPC)
Reactor Core Isolation Cooling (RCI)
Valves: HPCI-42, RCIC-37 P & ID: NH-36250 (M-124)
NH-35252 (M-126)
Code Class: 2 Category: C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1.
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.
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 4.5.1. 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 4.5.4 (c).
Page 34 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 31 System: Control Rod Drive Hydraulics (CRH)
Valve: CRD-1 14, 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 - Category C exercise test frequency of every 3 months per ISTC 4.5.1.
CV-126, CV-127 - Category A and B exercise test frequency of every 3 months, cold shutdown frequency or refueling outages per ISTC 4.2.1 and ISTC 4.2.2. Fail-Safe testing of valves per ISTC 4.2.6.
Function/Description: CRD-1 14 opens to allow exhaust water from the topof the drive piston to flow to the scram discharge volume during a control rod scram. CRD-1 15 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, Guidelines for Inservice Testing at Nuclear Power Plants, Appendix A, NRC Staff Position 7.
CRD valves for which this position is applicable:
CRD-1 14 (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-1 15 (for exercise to the open position)
CRD-138 (for exercise to the closed position)
Page 35 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 MNGP Technical Specification 4.3.C requires Control Rod Drive Scram insertion time testing once per operating cycle for each control rod. As such, testing for the specified directions and tests up to the allowed refueling outage duration will be enveloped by testing the respective control rod drive during Scram time testing as specified by the cyclic Technical Specification requirement..
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.
Page 36 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 32 System: Fuel Pool Cooling and Cleanup (FPC)
Valves: PC-20-1 and PC-20-2 P & ID: NH-36256 (M-135)
Code Class: NC Category: C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1.
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 4.5.1.
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 4.5.4 (c)(3).
Page 37 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 33 System: Alternate N2 (AN2)
Valves: AI-12-9, AI-12-10, AI-12-1 1, AI-12-12 P & ID: NH-36049-10 (M-131)
Code Class: NC Category: C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1.
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 4.5.1. 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 4.5.4 (c). This would be e-quivalent to the refueling outage frequency allowed by ISTC 4.5.4 (c).
Page 38 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 34 System: Residual Heat Removal (RHR)
Valve: RHR-81 P & ID: NH-36247 (M-121)
Code Class: 1 Category: A/C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1 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 4.5.1. Establishing the system lineup needed to exercise this check valve open and closed during power operation requires defeating the overpressurization protdctive 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 4.5.2(b).
Page 39 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 35 System: Instrument and Service Air (AIR)
Valve: 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: Category C exercise test frequency of every 3 months per ISTC 4.5.1 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 ISTC 4.5.1, nor cold shutdown per the requirements of ISTC 4.5.2(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 4.5.2(c).
Page 40 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 36 System: Condensate Storage (CST)
Valve: CST-189, CST-96, and CST-98 P & ID: NH-85509 (M-1 14-1)
Code Class: 2 Category: C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1 Function/Description: These valves supply keep-fill water pressure to the B Loop of RHR and B train of Core Spray. 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 4.5.1. 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 respective 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 or Core Spray water 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 at a cold shutdown frequency in accordance with ISTC 4.5.2(b).
Page 41 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant Revision 0 DEFERRED TESTING JUSTIFICATION NUMBER: DTJ 37 System: Standby Liquid Control (SLC)
Valve: XP-3-1, XP-3-2 P & ID: NH-36253 (M-127)
Code Class: 2 Category: C Code Frequency: Category C exercise test frequency of every 3 months per ISTC 4.5.1 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 pump's 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 4.5.1. 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 constitutes a hardship, which does not result in 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 4.5.4(c). At least one valve from the group will be disassembled and examined each refueling outage for the open and close exercise test.
Page 42 of 42
Nuclear Management Company, LLC Monticello Nuclear Generating Plant ATTACHMENT 7 Pump Tables Pump Valve Page I of 6 Revision 0
K>
Monticello Nuclear Generating Plant Inservice Testing Program - Pumps Component Description PID Class ASME Group Type Test Frequency RR P-I 09A 11 Residual Heat Removal Service NH-36665 3
A Vertical Q
Q/Y2 PR 02 Water Pump M-811 DP QN2 N/A V
Q/Y2 N/A P-1 09B 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-1 09C 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
QN2 N/A P-1 09D 14 Residual Heat Removal Service NH-36665 3
A Vertical Q
Q/Y2 PR 02 Water Pump M-81 1 DP Q/Y2 N/A V
QY2 N/A PunRp Table Page 2 of 6
P-1I Diesel Oil Transfer Pump NH-36051 NC B
Reciprocating Q
Q/Y2 N/A M-133 P
Y2 NIA V
Y2 N/A P-11i A
II Emergency Service Water Pump NH-36665 3
B Vertical Q
Y2 N/A M-81l DP QN2 N/A V
Y2 N/A P-111 B 12 Emergency Service Water Pump NH-36665 3
B Vertical Q
Y2 N/A M-81 1 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-I 11D 14 Emergency Service Water Pump NH-36665 3
B Vertical Q
Y2 N/A M-81 I DP Q/Y2 N/A V
Y2 NIA Pumnp Table Page 3 of 6 Component Description PID Class ASME Group
-Type Test Frequency RR
Description Combustible Gas Control System Booster Pump A PID NH-94896 NH-94896 Class ASME Group 3
B P-1 BICGC Combustible Gas Control System NH-94897 3
B Centrifugal DP QIY2 N/A Booster Pump B NH-94897 Q
Y2 N/A V
Y2 N/A P-202A 11 Residual Heat Removal Pump NH-36247 2
A Centnfugal DP Q/Y2 N/A M-121 Q
Q/Y2 PR02 V
Q/Y2 N/A P-202B 12 Residual Heat Removal Pump NH-36246 2
A Centrifugal DP QN2 N/A M-120 Q
QN'2 PR02 V
QY2 N/A P-202C 13 Residual Heat Removal Pump NH-36247 2
A Centrifugal DP Q/Y2 N/A M-121 Q
Q/Y2 PR02 V
Q/Y2 N/A Pkintnp TablekPage 4 of 6 Component P-iA/CGC Type Centrifugal Test DP Q
V Frequency Q/Y2 Y2 Y2 RR N/A NIA N/A
Class ASME Group Type Test Frequency RR P-202D 14 Residual Heat Removal Pu'mp NH-36246 2
A Centrifugal DP Q/Y2 N/A M-120 Q
Q/Y2 PR02 V
QN2 N/A P-203A 11 Standby Liquid Control Pump NH-36253 2
B Reciprocating Q
Q/Y2 PR 01
'M-127 P
Y2 N/A V
Y2 PR05 P-203B 12 Standby Liquid Control Pump NH-36253 2
B Reciprocating Q
QY2 PR 01 M-127 P
Y2 N/A V
Y2 PRO5 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 QIY2 N/A M-122 Q
Y2 N/A V
Y2 N/A Puitp Table Page 5 of 6 Component Description PID K)ý
KJ Description 12 Core Spray Pump PID NH-36248 M-122 Class ASME Group 2
B P-209 High Pressure Coolant Injection Pump NH-36250 2
B Centnfugal 0
Y2 N/A M-124 DP QIY2 PR 04 V
Y2 PR 03 N
Q/Y2 N/A Puip Table Page 6 of 6 Component P-208B Type Centrifugal Test DP Q
V Frequency QNY2 Y2 Y2 RR N/A NIA N/A
Nuclear Management Company, LLC Monticello Nuclear Generating Plant ATTACHMENT 8 Valve Tables Page I of 87 Revision 0
Syst em Alternate N2 Component Description PID Coord Size Typ7e Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ MSIV/SRV Pneumatic Supply Inboard Containment Isolation Check Valve NH-36049-10 M-131-10 B-5 10 CK SA 2
AC 0
O/C A
DI RF DTJ 19 LT AJ N/A SRV Pneumatic Supply NH-36049-10 D-5 1 0 CK SA 2
AC 0
O/C A
CTO Q
N/A Inboard Containment Isolation M-131-10 Check Valve CTC Q
N/A LT AJ N/A SRV Pneumatic Supply NH-36049-10 D-5 1.0 CK SA 2
AC 0
0/C A
CTO Q
N/A Outboard Containment M-131-10 Isolation Check Valve CTC Q
N/A LT AJ N/A Instrument Air/Nitrogen NH-36049-10 D-6 1 0 CK SA NC AC 0
C A
CTC Q
N/A Supply to SRV's Check Valve M-131-10 LT Y2 N/A CTO Q
N/A Instrument Air/Nitrogen NH-36049-10 D-6 1 0 CK SA NC AC 0
C A
CTC Q
N/A Supply to SRV's Check Valve M-131-10 LT Y2 N/A CTO 0
N/A MSIV/SRV Pneumatic Supply NH-36049-10 B-5 1 0 CK SA 2
AC 0
0/C A
DI RF DTJ 19 Outboard Containment M-131-10 Isolation Check Valve LT AJ N/A Instrument Air/Nitrogen NH-36049-10 B-6 1 0 CK SA NC AC 0
C A
CTC Q
N/A Supply to MSIVs/SRVs M-131-10 Check Valve LT Y2 N/A CTO Q
N/A AI-599 AI-700 AI-705 AI-706 AI-708 AI-713 AI-714 NH-36049-1 0 M-1 31-10 B-6 1.0 CK SA NC AC 0
C A
CTC Q
N/A LT Y2 N/A CTO 0
N/A Valive Table Page 2 of 8 7 Instrument Air/Nitrogen Supply to MSIVs/SRVs Check Valve Monticello Nuclear Generating Plant Inservice Testing Program - Valves Al-598
I Monticeflo 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-7 075 CK SA NC C
O/C 0
A CTO Q
N/A Check Valve M-131-10 CTC CS DTJ 19 Alternate Nitrogen Train B NH-36049-10 B-7 0.75 CK SA NC C
O/C 0
A CTO CS DTJ 19 Check Valve M-131-10 CTC CS DTJ 19 Train B Alternate Nitrogen NH-36049-10 B-7 1.5 RV SA NC C
C O/C A
RT SYO N/A Relief Valve M-131-10 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-1 31 -10 Train A Instrument NH-36049-10 D-5 1 5 RV SA NC C
C O/C A
RT SY1O N/A Air/Alternate Nitrogen Relief M-131-10 Valve 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 Train A Alternate Nitrogen NH-36049-10 D-6 1.0 GL SO NC B
0 O/C A
STO Q
N/A Supply to SRV's Isolation M-131-10 Valve PIT Y2 N/A STC Q
N/A FC Q
N/A Valve Table Page 3 of 87 AI-730 RV-4236 RV-4673 RV-4878 RV-4880 SV-4234
Monticello Nuclear Generating Plant Inservice Testing Programn-Valves System Alternate N2 Description Train B Alternate Nitrogen Supply to SRV's Isolation Valve PID Coord NH-36049-1 0 B-6 M-131-10 Size 10 Type Act GL SO Class Cat Norm Pos Safe Pos A/P Test NC B
0 0/c A
STO PIT STC FC Valve Table Page 4 of[87 1
Component SV-4235 Freq cs Y2 CS Q
RR/DTJ DTJ 19 N/A DTJ 19 N/A
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 Instrument Air Supply Check NH-36049-12 Valve to RV 2-71D M-131-12 C-3 1.0 CK SA NC AC O/C C
A CTC RF DTJ 25 Al-13-4 Al-13-7 RV-2-71A RV-2-71 B RV-2-710C RV-2-71 D RV-2-71E RV-2-71 F Valve Table Page S of 8 7 LT Y2 N/A CTO RF DTJ 25 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 Main Steam Safety/Relief NH-36241 B-4 6.00 RV AO/SA 1
BC C
O/C A
M-1 15 Main Steam Safety/Relief NH-36241 D-4 6.00 RV AO/SA 1
BC C
O/C A
RT SY5 N/A Valve M-115 Main Steam Safety/Relief NH-36241 D-3 6.00 RV AO/SA 1
BC C
0/C A
M-115 Main Steam Safety/Relief NH-36241 8:3 600 RV AO/SA 1
BC C
0/C A
M-1 15 Main Steam Safety/Relief NH-36241 B-4 6.00
, RV AO/SA 1
BC C
O/C A
RT SY5 N/A Valve (Low-Low Set)
M-1 15 Main Steam Safety/Relief NH-36241 B-3 6.00 RV AO/SA 1
BC C
O/C A
RT SY5 N/A Valve M-115
S
-K.
- w.
- .K 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 Main Steam Safety/Relief Valve (Low-Low Set)
NH-36241
,M-115 D-4 600 RV AO/SA I
BC C
O/C A
RT SY5 N/A Main Steam Safety/Relief NH-36241 D-3 6.00 RV AO/SA 1
BC C
O/C A
RT SY5 N/A Valve (Low-Low Set)
M-115 Safety Relief Valve RV-2 NH-36241-1 B-5 80 RV SA NC C
C O/C A
RT SYIO N/A A Discharge Vacuum Breaker M-115-1 Safety Relief Valve RV-2 NH-36241-1 C-6 80 RV SA NC C
C O/C 4A RT SYI0 N/A B Discharge Vacuum Breaker M-1 15-1 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 Breaker M-115-1 Safety Relief Valve RV-2 NH-36241-1 A-4 8.0 RV SA NC C
C O/C A
RT SY10 NIA D Discharge Vacuum Breaker M-1 15-1 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 Breaker M-1 15-1 Safety Relief Valve RV-2-71-F NH-36241-1 A-4 8.0 RV SA NC C
C O/C A
RT SY1O N/A Discharge Vacuum Breaker M-1 15-1 Valh Table Page 6 of 8 7 RV-2-71 G RV-2-71 H RV-3242A RV-3243A RV-3244A RV-3245A RV-7440A RV-7441A
I Monticello Nuclear Generating Plant Inservice Testing Programi-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 80 RV SA NC C
C O/C A
RT SYID N/A G Discharge Vacuum Breaker M-1 15-1 RV-7468A Safety Relief Valve RV-2 NH-36241-1 C-4 8.0 RV SA NC C
C O/C A
RT SYIO N/A H Discharge Vacuum Breaker M-1 15-1 XFV-10 Safety Relief Valve Discharge NH-36241-1 C-7 1.00 FV SA 2
C 0
C A
XT RF DTJ 22 Line Excess Flow Check Valve M-1 15-1 XFV-1 1 Safety Relief Valve Discharge NH-36241-1 C-7 1.00 FV SA 2
C 0
C A
XT RF DTJ 22 Une Excess Flow Check Valve M-1 15-1 XFV-12 Safety Relief Valve Discharge NH-36241-1 A-7 1.00 FV SA 2
C 0
C A
XT RF DTJ 22 Line Excess Flow Check Valve M-1 15-1 XFV-13 Safety Relief Valve Discharge NH-36241-1 A-7 1.00 FV SA 2
C 0
C A
XT RF DTJ 22 Line Excess Flow Check Valve M-1 15-1 XFV-1 4 Safety Relief Valve Discharge NH-36241-1 C-3 1.00 FV SA 2
C 0
C A
XT RF DTJ 22 Line Excess Flow Check Valve M-1 15-1 XFV-15 Safety Relief Valve Discharge NH-36241-1 C-3 1.00 FV SA 2
C 0
C A
XT RF DTJ 22 Line Excess Flow Check Valve M-115-1 Valie Table Page 7of87
Monticello Nuclear Generating Plant Inservice Testing Program - Valves N
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 Discharge NH-36241-1 A-3 1.00 FV SA 2
C 0
C A
XT RF DTJ 22 Line Excess Flow Check Valve M-1 15-1 XFV-17 Safety Relief Valve Discharge NH-36241-1 A-3 1.00 FV SA 2
C 0
C A
XT RF DTJ 22 Line Excess Flow Check Valve M-11 5-1 XFV-18 Safety Relief Valve Discharge NH-36241-1 A-3 1.00 FV SA 2
C 0
C A
XT RF DTJ 22 Line Excess Flow Check Valve M-1 15-1 1
XFV-19 Safety Relief Valve Discharge NH-36241-1 C-3 1.00 FV SA 2
C 0
C A
XT RF DTJ 22 Une Excess Flow Check Valve M-1 15-1 XFV-20 Safety Relief Valve Discharge NH-36241-1 A-7 1.00 FV SA 2
C 0
C A
XT RF DTJ 22 Line Excess Flow Check Valve M-1 15-1 XFV-21 Safety Relief Valve Discharge NH-36241 -1 C-7 1.00 FV SA 2
C 0
C A
XT RF DTJ 22 Line Excess Flow Check Valve M-1 15-1 Valtv Table Page 8 of 87
Monticello Nuclear Generathig Plant Inservice Testing Program - Valves System Combustible Gas Control Conmponent Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos AP. Test Freq RR/DTJ AO-7422A AO-7422B AO-7423A AO-7423B AO-7424A Drywell to Recombiner Supply NH-94896 D-7 40 GL AO 2
A C
O/C A
STO Q
N/A Inboard Isolation Valve NH-94896 STC Q
N/A FC Q
N/A PIT Y2 N/A LT AJ N/A Drywell to Recombiner Supply NH-94897 D-7 40 GL AO 2
A C
O/C A
STO Q
NIA Inboard Isolation Valve NH-94897 STC Q
N/A FC Q
N/A PIT Y2 N/A LT AJ NIA Drywell to Recombiner Supply NH-94896 D-7 40 GL AC 2
A C
O/C A
STO Q
N/A Outboard Isolation Valve NH-94896 STC C
N/A FC C
N/A PIT Y2 N/A LT AJ N/A Drywellto Recombiner Supply NH-94897 D-7 40 GL AO 2
A C
0/C A
STO Q
N/A Outboard Isolation Valve NH-94897 STC Q
N/A FC Q
N/A PIT Y2 N/A LT AJ N/A Recombiner to Suppression NH-94896 B-7 6.0 GL AO 2
A C
0/C A
STO Q
N/A Chamber Inboard Isolation NH-94896 Valve STC 0
N/A FC Q
N/A PIT Y2 N/A LT AJ N/A Valve Table Page 9 of 8 7
Monticello Nuclear Generating Plant Inservice Testing Programn-Valves System Combustible Gas Control Component Description PID Coord Size Type Act Class Cat NormPos SafePos A/P Test Freq RR1DTJ Recombiner to Suppression NH-94897 B-7 6 0 GL AO Chamber Inboard Isolation NH-94897 Valve 2
A C
O/C A
STO Q
N/A STC Q
N/A FC 0
N/A PIT Y2 N/A LT AJ N/A Recombiner to Suppression NH-94896 B-7 6.0 GL AO 2*A C
O/C A
STO Q
N/A Chamber Outboard Isolation NH-94896 Valve STC Q
N/A FC Q
N/A PIT Y2 N/A ILT AJ N/A Recombiner to Suppression NH-94897 B-7 6 0 GL AO 2
A C
O/C A
STO Q
N/A Chamber Outboard Isolation NH-94897 Valve STC Q
N/A FC Q
N/A PIT Y2 N/A LT AJ N/A CGC System Cooling Water NH-94896 A-5 1 5 CK SA 3
C C
O/C A
CTO Q
N/A Booster Pump Bypass Check NH-94896 Valve CTC Q
N/A CGC System Cooling Water NH-94897 A-5 1 5 CK SA 3
C C
O/C A
CTO Q
N/A Booster Pump Bypass Check NH-94897 Valve CTC Q
N/A CGCS Recombiner Inlet Valve NH-94896 D-5 3 0 GL MO 3
B C
0 A
STO Q
N/A NH-94896 PIT Y2 N/A Valm Table Page 10 of 87 AO-7424B AO-7425A AO-7425B CGC-12-1 CGC-12-2 MO-4043A
- .%.i..*........
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- /
Monticello Nuclear Generating Plant Inservice Testing Program - Valves Systenm Combustible Gas Control Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq Rt/DTJ Drywellto Recombiner Supply NH-94897 Isolation Valve NH-94897 D-5 30 GL MO 3
B C
0 A
STO Q
N/A PIT Y2 N/A CGC Water Separator to NH-94896 C-5 30 GL MO 3
B C
0 A
STO Q
NIA Recombiner Suction Isolation NH-94896 Valve PIT Y2 N/A MO-4044B CGC Water Separator to NH-94897 Recombiner Suction Isolation NH-94897 Valve C-5 30 GL MO 3
B C
0 A
STO Q
N/A PIT Y2 N/A CGC Cooling Water to NH-94896 B-4 0.75 GL MO 3
B C
0 A
STO Q
N/A Recombiner Supply Isolation NH-94896 Valve PIT Y2 N/A CGC Cooling Water to NH-94897 B-4 075 GL MO 3
B C
0 A
STO Q
NIA Recombiner Supply Isolation NH-94897 ValvePIT Y2 N/A CGC Cooling Water Relief NH-94896 B-5 1.0 RV SA 3
C C
O/C A
RT SY1 0 N/A Valve NH-94896 CGC Cooling Water Relief NH-94897 B-5 1 0 RV SA 3
C C
O/C A
RT SYIO N/A Valve NH-94897 CGC Cooling Water Booster NH-94896 A-6 20 GL SO 2
B C
0 A
STO Q
N/A Pump Suction Isolation Valve NH-94896 PIT Y2 N/A FO Q
N/A MO-4043B MO-4044A MO-4047A MO-4047B FV-4032A RV-4032B SV-4033A Valtv Table Page 11 of.87
Monticello Nuclear Generathig Plant Inservice Testing Program - Valves Sy'stem Combustible Gas Control Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ CGC Cooling Water Booster NH-94897 A-6 20 GL SO 2
B C
0 A
STO Q
N/A Pump Suction Isolation Valve NH-94897 PIT Y2 N/A FO Q
N/A CGC Cooling Water Booster NH-94896 A-6 2 0 GL SO 2
B C
0 A
STO Q
N/A Pump Bypass Isolation Valve NH-94896 PIT Y2 N/A FO 0
N/A CGC Cooling Water Booster NH-94897 A-6 20 GL SO 2
B C
0 A
STO Q
N/A Pump Bypass Isolation Valve NH-94897 PIT Y2 N/A FO Q
N/A CGC Cooring Water Strainer NH-94896 A-5 075 GL SO 3
B C
0 A
STO Q
N/A to Suppression Chamber NH-94896 Isolation Valve PIT Y2 N/A FO Q
NIA CGC Cooling Water Strainer NH-94897 A-5 075 GL SO 3
B C
0 A
STO 0
N/A to Suppression Chamber NH-94897 Isolation Valve PIT Y2 N/A FO Q
N/A Valve Table Page 12 of 8 7 SV-4033B SV-4034A SV-4034B SV-4054A SV-4054B
Monticello Nuclear Generating Plant Inservice Testing Programn-Valves Spystem Condensate and Feedwater Component Description PID Coord Size Tjpe Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ Feedwater Line A Outboard Check Valve NH-36241 M-115 A-3 1400 CK SA 2
C 0
C A
CTC CS DTJ12 CTO Q
N/A Feedwater Line B Outboard NH-36241 A-4 1400 CK SA 2
C 0
C A
CTC CS DTJ 12 Check Valve' M-115 CTO Q
N/A Feedwater Line A Outboard NH-36241 A-3 1400 CK SA 1
AC 0
0/C A
CTO Q
N/A Containment Isolation Check M-1 15 Valve CTC CS DTJ 12 LT AJ N/A Feedwater Line B Outboard NH-36241 A-4 1400 CK SA 1
AC 0
O/C A
CTO Q
N/A Containment Isolation Check M-115 Valve CTC CS DTJ 12 LT AJ N/A Feedwater Line A Inboard NH-36241 A-3 14.00 CK SA 1
AC 0
O/C A
CTO Q
N/A Containment Isolation Check M-115 Valve CTC CS DTJ 12 LT AJ N/A Feedwater Line B Inboard NH-36241 A-4 1400 CK SA 1
AC 0
O/C A
CTO Q
N/A Containment Isolation Check M-115 Valve CTC CS DTJ 12 LT AJ N/A FW-91-1 FW-91-2 FW-94-1 FW-94-2 FW-97-1 FW-97-2 ValVe Table Page 13 of 8 7
MonticellO Nuclear Generathig Plant Inservice Testing Program - Valves System Condensate Storage Component Description PID Coord Size Tjpe Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ Condensate Service Water NH-36248 Isolation Check Valve M-122 E-2 20 CK SA NC C
0 C
A CTC Q
N/A CTO Q
N/A Condensate Service Water NH-36248 E-2 20 CK SA 2
C 0
C A
CTC Q
N/A Isolation Check Valve M-1 22 CTO 0
N/A CST Check Valve to RHR NH-85509 B-6 1.0 CK SA 2
C 0
C A
CTO CS DTJ 36 System Flush to Reactor M-1 14-1 Vessel Head Spray CTC Q
N/A CST Check Valve to LPCI B NH-85509 B-3 2.0 CK SA 2
C 0
C A
DI SY8 DTJ 02 Injection M-1 14-1 CST Check Valve to RHR NH-85509 B-4.
20 CK SA 2
C 0
C A
DI SY8 DTJ 02 Suction Cross Tie Keepfill M-1 14-1 CST Check Valve to LPCI "A" NH-85509 B-4 20 CK SA 2
C 0
C A
DI SY8 DTJ 02 Injection M-1 14-1 CST Check Valve to NH-85509 B-4 2.0 CK SA 2
C 0
C A
DI SY8 DTJ 02 Containment Spray Loop A M-1 14-1 CST Check Valve to Core NH-85509 B-5 20 CK SA 2
C 0
C A
CTC Q
N/A Spray Loop B M-1 14-1 CTO CS DTJ 36 Valve Table Page 14 of 87 CST-103-1 CST-104-1 CST-189 CST-88 CST-90 CST-92 CST-94 CST-96
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Condensate Storage Component Description CST-98 CST Check Valve to Containment Spray Loop B PID Coord NH-85509 M-114-1 B-5 Size 20 Type Act Class Cat Norm Pos Safe Pos A/P Test CK SA 2
C 0
C A
CTC CTO Freq PIJDTJ Q
N/A CS DTJ 36 Val*e Table Page 15 of 87
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Control Rod Drive Hydraulics Conmonent Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos AiP Test Freq RR/DTJ Control Rod Drive to Reactor NH-36242-2 D-3 0 38 CK SA Vessel Instrumentation Check M-1 16-2 Valve 2
AC 0
c A
LT Y2 N/A CTC RF DTJ 21 CTO Q
N/A Control Rod Drive to Reactor NH-36242-2 D-3 038 CK SA 2
AC 0
C A
LT Y2 N/A Vessel Instrumentation Check M-1 16-2 Valve CTC RF DTJ 21 CTO Q
N/A Control Rod Drive to Reactor NH-36242-2 D-3 038 CK SA 2
AC 0C A
LT Y2 N/A Vessel Instrumentation Check M-116-2 Valve CTC RF DTJ 21 CTO Q
N/A Control Rod Drive to Reactor NH-36242-2 D-3 038 CK SA 2
AC 0
C A
LT Y2 N/A Vessel Instrumentation Check M-1 16-2 Valve CTC RF DTJ 21 CTO Q
N/A Control Rod Drive to Reactor NH-36242-2 C-3 0 38 CK SA 2
AC 0
C A
LT Y2 N/A Vessel Instrumentation Check M-1 16-2 Valve CTC RF DTJ 21 CTO Q
N/A Control Rod Drive to Reactor NH-36242-2 C-3 0.38 CK SA 2
AC 0
C A
LT Y2 N/A Vessel Instrumentation Check M-1 16-2 Valve CTC RF DTJ 21 CTO Q
N/A Control Rod Drive to Reactor NH-36242-2 C-3 0 38 CK SA 2
AC 0
C A
LT Y2 N/A Vessel Instrumentation Check M-1 16-2 Valve CTC RF DTJ 21 CTO 0
N/A Control Rod Drive to Reactor NH-36242-2 C-3 038 CK SA 2
AC 0
C A
LT Y2 N/A Vessel Instrumentation Check M-1 16-2 Valve CTC RF DTJ 21 CTO Q
N/A Valve Table Page 160of87 BF-12 BF-14 BF-24 BF-26 BF-35 BF-37 BF-46 BF-48
Monticello Nuclear Generating Plant Inservice Testing Program - Valves S;ystemn Control Rod Drive Hydraulics Conponent Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos AIP Test Freq RR/DTJ CRD-Scram Discharge Header Check Valve (Typical for 121 Drive Units)
NH-36245 M-119 A-6 0 75 CK SA 2
C 0/C 0
A CTO TS DTJ 31 CTC SY8 VR02 CRD-Accumulator Charging NH-36245 B-4 050 CK SA 2
C O/C C
A CTC RF DTJ 05 Water Header Check Valve M-1 19 (Typical for 121 Drive Units)
CTO TS DTJ 31 CRD-Cooling Water Check NH-36245 D-4 050 CK SA 2
C 0
C A
CTC TS DTJ 31 Valve (Typical for 121 Drive M-119 Units)
CTO Q
N/A CRD-SCRAM Inlet Valve NH-36245 B-5 1.0 GL AO 2
B C
0 A
STO TS DTJ 31 (Typical for 121 Drive Units)
M-1 19 FO TS DTJ31 PIT Y2 N/A CRD-SCRM Outlet Valve NH-36245 B-6 0.75 GL AO 2
B C
0 A
STO TS DTJ 31 (Typical for 121 Drive Units)
M-119 FO TS DTJ 31 PIT Y2 N/A CRD-Scram Discharge NH-36245 D-4 1.0 GL AO 2
B 0
C A
STC Q
N/A Volume Vent Isolation Valve M-1 19 FC Q
N/A PIT Y2 N/A CRD-Scram Discharge NH-36245 D-1 1.0 GL AO 2
B 0
C A
STC Q
N/A Volume Vent Isolation Valve M-1 19 FC 0
N/A PIT Y2 N/A CRD-Scram Discharge NH-36245 D-4 1.0 GL AO 2
B 0
C A
STC Q
N/A Volume Vent Isolation Valve M-1 19 FC Q
N/A PIT Y2 N/A Jaliv Table Page 17of 877 CRD-1i14*
CRD-115" CRD-138" CV-126*
CV-127*
CV-3-32A CV-3-32B CV-3-32C
L L...................
~
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-Scram Discharge NH-36245 D-1 1.0 GL AO 2
B 0
C A
STC Q
N/A Volume Vent Isolation Valve M-1 19 FC 0
N/A PIT Y2 N/A CRD-Scram Discharge NH-36245 C-3 20 GL AO 2
B 0
C A
STC Q
N/A Volume Drain Isolation Valve M-1 19 FC Q
N/A PIT Y2 N/A CRD-Scram Discharge NH-36245 C-2 20 GL AO 2
B 0
C A
STC 0
N/A Volume Drain Isolation Valve M-119 FC Q
NIA PIT Y2 N/A CRD-Scram Discharge NH-36245 C-3 20 GL AO 2
B 0
C A
STC Q
N/A Volume Drain Isolation Valve M-119 FC Q
N/A PIT Y2 NIA CRD-Scram Discharge NH-36245 C-2 20 GL AO 2
B 0
C A
STC Q
NIA Volume Drain Isolation Valve M-1 19 FC Q
N/A PIT Y2 N/A Valve Table Page 18 of 87 CV-3-32D CV-3-33A CV-3-33B CV-3-33C CV-3-33D
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 CS Injection Air-Operated NH-36248 Testable Check Valve M-1 22 D-3 80 CK SA 1
AC C
O/C A
CTO RF DTJ 27 CTC RF DTJ 27 LT Y2 N/A 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 CS A Loop Manual Isolation NH-36248 D-3 8.0 GT MA I
B 0
0 P
PIT Y2 N/A Valve to Reactor Vessel M-1 22 CS B Loop Manual Isolation NH-36248 D-4 8 0 GT MA 1
B 0
0 P
PIT Y2 N/A Valve to Reactor Vessel M-1 22 CS Pump P-208A Discharge NH-36248 B-2 100 CK SA 2
C C
O/C A
CTO Q
N/A Check Valve M-1 22 CTC Q
N/A CS Pump P-208B Discharge NH-36248 B-5 100 CK SA 2
C C
O/C A
CTO Q
N/A Check Valve M-122 CTC Q
N/A CS Pump P-208A Motor NH-36248 A-3 120 GT MO 2
B 0
0 P
PIT Y2 NIA Operated Suction Valve M-122 CS Pump P-208B Motor NH-36248 A-4 120 GT MO 2
B "0
0 P
PIT Y2 NIA Operated Suction Valve M-122 AO-14-13B CS-1i3-1 CS-1i3-2 CS-9-1 CS-9-2 MO-1741 MO-1742 Valve Table Page 19 of 87
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 CS System A Loop Motor NH-36248 Operated Test Return Valve M-1 22 C-2 60 GL MO 2
B C
C P
PIT Y2 N/A CS System 8 Loop Motor NH-36248 C-5 6 0 GL MO 2
B C
C P
PIT Y2 N/A Operated Test Return Valve M-1l22 CS System Motor Operated NH-36248 D-2 80 GT MO 2
A 0
O/C A
LT AJ N/A Outboard Injection Valve M-122 STC Q
N/A PIT Y2 N/A CS System Motor Operated NH-36248 B-5 80 GT MO 2
A 0
O/C A
LT AJ N/A Outboard Injection Valve M-122 STO Q
N/A STO Q
N/A PIT Y2 NIA CS System Motor Operated NH-36248 D-3 80 GT MO 1
A C
O/C A
LT AJ N/A Inboard Injection Valve M-122 LT Y2 N/A STO 0
N/A STC Q
N/A PIT Y2 NIA CS System Motor Operated NH-36248 D-5 80 GT MO 1
A C
O/C A
LT AJ N/A Inboard Injection Valve M-122 IT Y2 N/A STO Q
N/A STC Q
N/A PIT Y2 N/A MO-1749 MO-1750 MO-1i751 MO-1752 MO-1753 MO-1i754 Valm Table Page 20 of 87
Monticello Nuclear Generating Plant Inservice Testing Program - Valves Sj'stem Core Spray Cooling Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RRIDTJ Core Spray System Relief Valve NH-36248 M-122 D-2 20 RV SA 2
C C
O/C A
RT SYl 0 N/A Core Spray System Relief NH-36248 D-5 20 RV SA 2
C C
O/C A
RT SY10 N/A Valve M-122 Core Spray System Excess NH-36248 C-3 1.0 FV SA 1
C 0
C A
XT RF DTJ 22 Flow Check Valve M-122 Core Spray System Excess NH-36248 C-4 1 0 FV SA 1
C 0
C A
XT RF DTJ 22 Flow Check Valve M-1 22 Valhe Tab& Page 21 of 8 7 RV-1745 RV-1 746 XFV-82 XFV-83
...'...............
Monticello Nuclear Generating Plant Inservice Testing Programi-Valves System Demineralized Water System Conponent Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RRI/DTJ DM Outboard Containment Isolation Valve NH-36039 M-108 E-1 1 0 GT MA
.2 A
C C
P LT AJ N/A DM Inboard Containment NH-36039 E-1 1.0 GT MA 2
A C
C P
LT AJ NIA Isolation Valve M-108 DM-151 DM-152 Valve Table Page 22 of 87
Monticello Nuclear Generating Plant Inservice Testing Program - Valves fesel Generators Description PID Coord' Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ System Di Component GSA-32-1 GSA-32-2 GSA-32-3 GSA-32-4 RV-3216 RV-3217 RV-3218 RV-3219 1"
Valm Table Page 23 of 87 EDG Starling Air Dryer NH-36051 B-2 075 CK SA NC C
O/C C
A CTC Q
N/A Discharge Check Valve M-133 CTO 0
N/A EDG Starting Air Dryer NH-36051 A-2 075 CK SA NC C
O/C C
A CTC Q
N/A Discharge Check Valve M-133 CTO 0
N/A 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 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-1 33 CT0 Q
N/A EDG Starting Air Receiver NH-36051 B-3 05 RV SA NC C
C O/C A
RT SY10
- N/A Tank T-79A Relief Valve M-133 EDG Starting Air Receiver NH-36051 B-3 05 RV SA NC C
C O/C A
RT SYl0 N/A Tank T-79B Relief Valve M-133 EDG Starting Air Receiver NH-36051 B-4 05 RV SA NC C
C 0/C A
RT SY10 N/A Tank T-79C Relief Valve M-133 EDG Starting Air Receiver NH-36051 A-3 05 RV SA NC C
C O/C A
RT SYIO N/A Tank T-79D Relief Valve M-133
Monticello Nuclear Generating Plant Inservice Testing Progran - Valves S;ystem Diesel Generators Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ RV-3220 RV-3221 RV-3224 RV-3225 RV-3226 RV-3227 RV-3228 RV-3229 Valve Table Page 24 of 87 EDG Starbng Air Receiver' NH-36051 A-3 05 RV SA NC C
C O/C A
RT SY10 N/A Tank T-79E Relief Valve M-1 33 EDG Starting Air Receiver NH-36051 A-4 05 RV SA NC C
C O/C A
RT SYlO N/A Tank T-79F Relief Valve M-1 33
~~
EDG Starting Air Receiver NH-36051 E-3 05 RV SA NC C
C O/C A
RT SYI0 N/A Tank T-80A Relief Valve M-1 33 EDG Starting Air Receiver NH-36051 E-3 05 RV SA NC C
C 0/C A
RT SYl0 N/A Tank T-80B Relief Valve M-133 EDG Starting Air Receiver NH-36051 E-4 05 RV SA NC C
C O/C A
RT SY 0 N/A Tank T-80C Relief Valve M-1 33 EDG Starting Air Receiver NH-36051 D-3 05 RV SA NC C
C 0/C A
RT SYIO N/A Tank T-80D Relief Valve M-1 33 EDG Starting Air Receiver NH-36051 D-3 05 RV SA NC C
C O/C A
RT SYIO N/A Tank T-80E Relief Valve M-133 EDG Starting Air Receiver NH-36051 D-4 05 RV SA NC C
C 0/C A
RT SYl0 N/A Tank T-80F Relief Valve M-133
-J
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 Diesel Oil Service Pump Discharge Check Valve NH-36051 M-133 C-3 1.0 CK SA NC C
0 C
A CTC Q
N/A CTO Q
NIA Diesel Oil Storage Tank to NH-36051 B-2 1 5 CK SA NC C
C O/C A
CTO Q
N/A Fuel Oil Pumps Check Valve M-133 CTC Q
N/A Diesel Oil Storage Tank to NH-36051 B-2 1 5 CK SA NC C
C O/C A
CTO 0
N/A Fuel Oil Pumps Check Valve M-133 CTC 0
N/A Diesel Oil Transfer Pump NH-36051 C-3 20 CK SA NC C
C 0
A CTO Q
N/A Discharge Check Valve M-1 33 CTC Q
N/A Diesel Fuel Oil Transfer NH-36051 D-3 075-RV SA NC C
C O/C A
RT SYI0 N/A Pump Relief Valve M-133 FO-43 FO-44 FO-5 RV-1523 Valive Table Page 25 of87 FO-2
Inservice Testing Program - Valves System EDG Emergency Service Water Component Description PID Coord Size Tjpe Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ Emergency Service Water Pump P-1i11A Air Vent Valve NH-36665 M-81 1 B-4 20 AR SA 3
C O/C O/C A
CTC Q
N/A CTO 0
N/A AV-3155 AV-3156 ESW-1-1 ESW-1-2 ESW-3-1 ESW-3-2 ESW-71 -1 ESW-71-2 Val*v Table Page 26 of 87 Emergenc'y Service Water NH-36665 B-6 20 AR SA 3
C O/C O/C A
CTC Q
N/A Pump P-i 11B Air Vent Valve M-811 CTO Q
N/A Emergency Service Water NH-36665 B-4 40 CK SA 3
C O/C 0
A CTO Q
NIA Pump P-1IIA Discharge M-811 Check Valve CTC Q
N/A Emergency Service Water NH-36665 B-8 40 CK SA NC C
O/C 0
A CTO Q
N/A Pump P-1I11B Discharge M-811 Check Valve Emergency Service Water NH-36665 C-4 40 GT MA 3
B O/C 0/C A
STO Y2 N/A Basket Strainer Bypass M-811 Isolation Valve STC Y2 N/A Emergency Service Water NH-36665 C-5 40 GT MA 3
B O/C O/C A
STO Y2 N/A Basket Strainer Bypass M-811 Isolation Valve STC Y2 N/A Emergency Service Water NH-36665 B-4 05 CK SA 3
C O/C C
A CTC Q
N/A Check Valve M-811 CTO Q
N/A Emergency Service Water NH-36665 B-6 05 CK SA 3
C O/C C
A CTC Q
N/A Check Valve M-811 CTO 0
N/A Monticello Nuclear Generating Plant
Monticello Nuclear Generating Plant Inservice Testing Program - Valves S;rstemn EDG Emergency Service Water Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos AlP Test Freq RR/DTJ ESW-73-1 Emergency Service Water NH-36665 B-4 05 CK SA 3
C 0/C C
A CTC Q
N/A Check Valve M-811 CTO Q
N/A Emergency Service Water NH-36665 B-6 05 CK SA 3
C O/C C
A CTC Q
N/A Check Valve M-811I CTO Q
N/A Service Water to Emergency NH-36665 D-7 40 CK SA NC C
O/C C
A CTC Q
N/A Service Water Check Valve M-811 CTO 0
N/A Service Water to Emergency NH-36665 D-7 40 CK SA 3
C O/C C
A CTC 0
N/A Service Water Check Valve M-811 CTO 0
N/A Service Water to Emergency NH-36665 D-7 40 CK SA NC C
O/C C
A CTC Q
N/A Service Water Check Valve M-81 1 CTO 0
N/A Service Water to Emergency NH-36665 D-7 40 CK SA 3
C O/C C
A CTC Q
N/A Service Water Check Valve M-811 CTO Q
N/A Valve Table Page27 of 87 ESW-73-2 SW-15 SW-16 SW-17 SW-18
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System EFT Emergency Service Water Description PID Emergency Service Water NH-36665 Pump P-1i11C Air Vent Valve M-811 Coord Size Type Act Class Cat Norm Pos SafePos A/P Test Freq RR/DTJ C-3 10 AR SA 3
C O/C O/C A
CTC Q
N/A Conponent AV-4024 AV-4026 ESW-13 ESW-14 ESW-15 ESW-16 ESW-17 ESW-18 CTO Q
N/A Emergency Service Water NH-36665 B-5 1.0 AR SA 3
C O/C O/C A
CTC Q
N/A Pump P-1IllD Air Vent Valve M-811 CTO Q
N/A Service Water to Emergency NH-36665 C-5 40 CK SA NC C
O/C C
A CTC Q
N/A Service Water Check Valve M-81 1 CTO Q
N/A Service Water to Emergency NH-36665 C-6 40 CK SA 3
C O/C C
A CTC Q
N/A Service Water Check Valve M-811N CTO Q
N/A Service Water to Emergency NH-36665 D-4 40 CK SA NC C
O/C C
,A CTC Q
N/A Service Water Check Valve M-811 CTO 0
N/A Service Water to Emergency NH-36665 D-4 40 CK SA 3
C O/C C
A CTC Q
N/A Service Water Check Valve M-81 1 CTO 0
N/A Emergency Service Water NH-36665 C-6 40 CK SA 3
C O/C 0
A CTO Q
N/A Pump P-i 110D Discharge M-811 Check Valve CTC CS DTJ 37 Emergency Service Water NH-36665 C-4 40 CK SA 3
C O/C 0
A CTO Q
N/A Pump P-1IllC Discharge M-81I1 Check Valve CTC CS DTJ 37 Valve Table Page 28 of 87
Monticello Nuclear Generating Plant Inservice Testing Program - Valves S ystem EFT Emergency Service Water Co*ponent Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A.P Test Freq RR/DTJ Manual Bypass Valve for Strainer YS-111D NH-36665 M-811 C-6 40 GT MA 3
B 0/C O/C A
STO STC Y2 Y2 N/A N/A Manual Bypass Valve for NH-36665 C-4 30 GT MA 3
B 0/C O/C A
STO Y2 N/A Strainer YS-1I11C M-811 STC Y2 N/A TCV-4030B Manual Bypass NH-36041 E-2 30 GIL MA 3
B C
0 A
STO Y2 N/A Valve M-110 TCV-4030A Manual Bypass NH-36041 D-2 3 0 GL MA 3
B C
0 A
STO Y2 N/A Valve M-110 Valve Table Page 29 of 87 ESW-19 ESW-20 ESW-33-1 ESW-34-1
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 Spent Fuel Pool Return Line NH-36256 Anti-Siphon Check Valve M-135 D-4 60 CK SA NC C
0 C
A DI SY8 DTJ 32 Spent Fuel Pool Return Line NH-36256 D-4 60 CK SA NC C
0 C
A DI SY8 DTJ 32 Anti-Siphon Check Valve M-135 PC-20-1 PC-20-2 Valm Table Page 30 of 87
Monticello Nuclear Generating Plant Inservice Testing Programn-Valves Siystem High Pressure Coolant Injection Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos AIP Test Freq RR/DTJ HPCI Injection Check Valve NH-36250 M-124 C-5 120 CK SA 2
C C
O/C A
DI RF DTJ 15 HPCI Steam Line Drain NH-36249 C-1 1.0 GL AO 2
B 0
C A
STC Q
NIA Isolation Valve M-123 FC Q
N/A PIT Y2 N/A HPCI Pump Minimum Flow NH-36250 C-4 20 GL AO 2
B C
O/C A
STO Q
NIA Control Valve M-1 24 STC Q
N/A FO Q
N/A PIT Y2 N/A HPCI Turbine Exhaust Check NH-36249 C-5 160 CK SA 2
AC O/C O/C A
CTO Q
N/A Valve M-1 23 CTC CS DTJ 17 LT AJ N/A 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 Valve M-1 23 CTO CS DTJ 17 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 Valve M-123 CTO CS DTJ 17 HPCI Gland Seal NH-36249 A-2 20 CK SA 2
C O/C 0
A "CTO Q
N/A Condenser/Lube Oil Cooler M-123 Cooling Water Return Check CTC Q
N/A Valve HPCI Gland Seal NH-36249 A-3 20 CK SA 2
C O/C C
A CTO Q
N/A Condenser/Condensate Pump M-123 Discharge Check Valve T
a 0 Q
N/A Valve Table Page 31 ofg 7 AO-23-18 CV-2046A CV-2065 HPCI-10 HPCI-14 HPCI-15 HPCI-18 HPCI-20
Monticello Nuclear Generating Plant hiservice 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 Pump Suction from NH-36250 Suppression Pool Check Valve M-124 A-4 140 CK SA 2
C C
O/C A
DI RF DTJ 17 HPCI Pump Suction from NH-36250 D-3 140 CK SA 2
C C
O/C A
CTC CS DTJ 29 Condensate Storage Tank M-124 Check Valve CTO Q
NIA HPCI Minimum Flowto NH-36250 BA 40 CK SA 2
C C
0 A
DI RF DTJ 30 Suppression Chamber Check M-124 Valve 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 Exhaust Line Vacuum NH-36249 B-6 20 CK SA 2
C C
O/C A
CTC CS DTJ 17 Breaker Check Valve M-1 23 CTO CS DTJ 17 HPCI Turbine Exhaust Check NH-36249 C-5 160 CK SA 2
AC 0/C O/C A
CTO Q
N/A Valve M-1 23 CTC CS DTJ 17 LT AJ N/A HPCI Steam Supply Inboard NH-36249 E-5 80 GT MO 1
A 0
O/C A
STO Q
N/A Containment Isolation Valve M-1 23 STC Q
N/A PIT
, Y2 N/A LT AJ N/A HPCI-31 HPCI-32 HPCI-42 HPCI-65 HPCI-71 HPCI-9 MO-2034 Valve Table Page 32 of87
Monticellb Nuclear Generating Plant Inservice Testing Program - Valves S)ysteni High Pressure Coolant Injection Conponent Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ MO-2035 HPCI Steam Supply Outboard NH-36249 E-4 8.0 GT MO 1
A 0
O/C A
STO Q
N/A Containment Isolation Valve M-1 23 STC Q
N/A PIT Y2 N/A LT AJ N/A HPCI Turbine Steam Supply NH-36249 D-2 80 GT MO 2
B C
0 A
STO Q
N/A Isolation Valve M-123 PIT Y2 N/A HPCI Suppression Pool NH-36250 A-5 140 GT MO 2
B C
0/C A
STO Q
N/A Suction Isolation Valve M-124 STC Q
N/A PIT Y2 N/A HPCI Suppression Pool NH-36250 A-4 14.0 GT MO 2
B C
O/C A
STO Q
N/A Suction Isolation Valve M-1 24 STC Q
N/A PIT Y2 N/A HPCI Condensate Storage NH-36250 E-3 140 GT MO 2
B 0
O/C A
STC Q
N/A Tank Suction Isolation Valve M-124 STO Q
NIA PIT Y2 N/A HPCI Injection Isolation Valve NH-36250 C-5 120 GT MO 2
B C
0 A
STO Q
N/A M-124 PIT Y2 N/A HPCI Injection Isolation Valve NH-36250 C-5 120 GT MO 2
B C
0 A
STO Q
N/A M-124 PIT Y2 N/A HPCI Test Return to NH-36250 D-5 80 GL MO 2
B 0/C C
A STC Q
NIA Condensate Storage Tank M-1 24 Isolation Valve PIT Y2 NIA Ta................
Valve Table Pag-e 33 of 87 MO-2036 MO-2061 MO-2062 MO-2063 MO-2067 MO-2068 MO-2071
Monticello Nuclear Generating Plant Inservice Testing.Program - Valves Sjystem High Pressure Coolant Injection Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ HPCI Turbine Exhaust Line Rupture Disk NH-36249 M-123 C-5 160 RD SA 2
D C
O/C A
DT Y5 N/A HPCI Turbine Exhaust Line NH-36249 C-5 160 RD SA NC D
C O/C A
DT Y5 N/A Rupture Disk M-123 HPCI Cooling Water Supply NH-36249 B-3 1 5 RV SA 3
C C
0/C A
RT SYIO NIA to Gland Seal Condenser M-123 Relief Valve HPCI Pump Suction Relief NH-36250 D-2 1 0 RV SA 2
C C
OIC A
RT SYIO N/A Valve M-124 HPCI Turbine Steam Supply NH-36249 D-5 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Line Excess Flow Check Valve M-123 HPCI Turbine Steam Supply NH-36249 D-5 1.00 FV SA 1
C 0
C
. A XT RF DTJ 22 Line Excess Flow Check Valve M-123 Valve Table Page 34 of 8 7 PSD-2038 PSD-2039 RV-2056 RV-2064 XFV-84 XFV-85
J 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 PR/DTJ SV-4001A Drywell to CAM Analyzer NH-91197 B-6 075 GL SO 2
A O/C C
A STC O
N/A Panel Outboard Isolation Valve NH-91197 PIT Y2 N/A LT AJ N/A FC Q
N/A SV-4001B Drywell to CAM Analyzer NH-91197 B-6 075 GL SO 2
A O/C C
A STC Q
N/A Panel Outboard Isolation Valve NH-91197 PIT Y2 N/A LT AJ NIA FC Q
NIA SV-4002A Suppression Chamber to NH-91197 A-5 075 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 0
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-91 197 Outboard Isolation Valve PIT Y2 N/A LT AJ N/A FC Q
N/A Valsv Table Page 35 of 87
-\\
i
.Monticello Nuclear Generating Plant Inservice Testing Pr-ogram - 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 SV-4004B SV-4005A SV-4005B SV-4020A SV-4020B Suppression Chamber to NH-91197 A-4 0.75 GL SO 2
A O/C C
A STC 0
N/A CAM Analyzer Panel Inboard NH-91 197 Isolation Valve PIT Y2 N/A LT AJ N/A FC Q
N/A 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 Suppression Chamber to NH-91197 B-4 075 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 0
N/A Suppression Chamber to NH-91197 B-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 0
N/A 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-91 197 PIT Y2 N/A LT AJ N/A FC Q
N/A Drywell to CAM Analyzer NH-91197 A-6 075 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 36 of 87 1
Monticello Nuclear Generating Plant Inservice Testing Program - Valves Srstem Instrument and Service Air Component Description PID Coord Size TjTe 'Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AM-12-10 Vahle Table Page 37 of 87 Inboard MSIV Pneumatic NH-36049-10 A-3 1.0 CK SA NC C
0/C 0
A DI SY8 DTJ-33 Supply Check Valve M-131-10 Inboard MSIV Pneumatic NH-36049-10 B-3 1 0 CK SA NC C
0/C 0
A DI SY8 DTJ-33 Supply Check Valve M-131-10 Inboard MSIV Pneumatic NH-36049-10 A-3 1.0 CK SA NC C
O/C 0
A DI SY8 DTJ-33 Supply Check Valve M-1 31-10 Inboard MSIV Pneumatic NH-36049-10 A-3 1.0 CK SA NC C
0/C 0
A DI SY8 DTJ-33 Supply Check Valve M-131-10 Instrument Air Supply Check NH-36247
. A-3 075 CK SA NC C
O/C C
A CTC RF DTJ 35 Valve to CV-1728 M-1 21 CTO RE DTJ 35 Instrument Air Supply Check NH-36246 A-5 075 CK SA NC C
O/C C
A CTC RF DTJ 35 Valve to CV-1729 M-120 CTO RE DTJ35 Instrument Air Supply Check NH-36247 A-3 075 CK SA NC C
O/C C
A CTC RF DTJ 35 Instrument Air Supply Check.
NH-36246 A-5 075 CK SA' NC
-~ C 01C C
A CTC RF DTJ 35 Valve to CV-1729 M-120 CTO RF DTJ 35 Al-12-11 AM-2-12 Al-12-9 AI-243-1 AI-243-2 AI-244-1 AI-244-2
.. ~.
-,ý,
Monticello Nuclear Generating Plant Inservice Testing Program - Valves Systenm Instrument and Service Air Component Description PID Coord Size Tjpe Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AI-571 Inboard Instrument Air Supply NH-36049-12 B-5 20 CK SA 2
AC 0
C A
CTC CS DTJ 23 Containment Isolation Valve M-131-12 LT AJ N/A CTO CS DTJ 23 AI-610-1 Instrument Air Supply to CV-NH-36247 A-4 075 CK SA NC AC O/C C
A CTC RF DTJ 25 1994 Check Valve M-1 21 CTO RF DTJ 25 LT Y2 NA AI-610-2 Instrument Air Supply to CV-NH-36246 A-4 075 CK SA NC AC 0/C C
A CTC RF DTJ 25 1995 Check Valve M-120 CTO RF DTJ 25 LT Y2 N/A Al-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-1 21 CTO RF DTJ 25 LT Y2 N/A AI-610-4 Instrument Air Supply to CV-NH-36246 C-4 075 CK SA NC AC O/C C
A CTC RF DTJ 25 1997 Check Valve M-1 20 CTO RF DTJ 25 LT Y2 NIA AI-611 Instrument Air Supply to HPCI NH-36250 D-4 075 CK SA NC AC O/C C
A CTC RF DTJ 25 Mm Flow Control Valve CV-M-124 2065 Check Valve CTO RF DTJ 25 LT Y2 N/A AI-612 RCIC Instrument Air Isolation NH-36252 C-4 075 CK SA NC AC O/C C
A CTC RF DTJ 25 Check Valve (CV-2104)
M-126 LT Y2 N/A CTO RF DTJ'25 AI-613 Instrument Air Supply to T-NH-36049-14 C-6 025 CK SA NC AC 0/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 38 of87
Monticello Nuclear Generating Plant Inservice Testing Program - Valves Systemn 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 AI-615 AI-616 AI-617 AI-618 AI-619 AI-625 AI-626-1 Instrument Air Supply to T-NH-36049-14 C-6 025 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 Instrument Air Supply to T-NH-36049-14 C-6 025 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 Instrument Air Supply to T-NH-36049-14 B-6 025 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 Instrument Air Supply to T-NH-36049-14 A-6 025 CK SA NC
, AC 0/C C
A CTC RF DTJ 24 Ring Seal AO-2383 M-131-14 LT Y2 N/A CTO RF DTJ 24 Instrument Air Supply to T-NH-36049-14 B-2 025 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 Instrument Air Supply to T-NH-36049-14 A-2 025 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 Outboard Tip Purge NH-36049-14 C-4 075 CK SA 2
AC 0
C A
CTC RF DTJ 09 Containment Isolation Valve M-131-14 LT AJ N/A CTO Q
N/A Inboard Tip Purge NH-36049-14 C-4 075 CK SA 2
AC 0
C A
CTC RF DTJ 09 Containment Isolation Valve M-131-14 IT AJ N/A CTO Q
N/A Valve Table Page 39 of 87
'Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Instrument and Service Air Component Description PID Coord Size TyTe Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ AI-629 AI-663 AI-666 AI-669 AI-672 AI-675 AI-678 Al-681 Instrument Air Supply to NH-36049-14 B-5 075 CK SA 2
AC 0/C C
A CTC Q
N/A Torus Ring Header Inboard M-131-14 Containment Isolation Valve LT AJ NIA CTO Q
N/A Alternate Nitrogen Supply to T-NH-36049-14 C-7 0 375 CK SA NC C
O/C 0
A CTO RF DTJ 24 Ring Seal AO-2377 -
M-131-14 CTC RE 0TJ24 Alternate Nitrogen Supply to T-NH-36049-14 C-5 0375 CK SA NC C
0/C 0
A CTO RF DTJ 24 Ring Seal AO-2378 M-1 31-14 CTC RE DTJ24 Alternate Nitrogen Supply to T-NH-36049-14 C-6 0375 CK SA NC C
O/C 0
A CTO RF DTJ 24 Ring Seal AO-2381 M-131-14 CTC RE DTJ24 Alternate Nitrogen Supply to T-NH-36049-14 A-6 0375 CK SA NC C
O/C 0
A CTO RF DTJ 24 Ring Seal AO-2383 M-131-14 CTC RE DTJ24 Alternate Nitrogen Supply to T-NH-36049-14 B-i 0375 CK SA NC C
0/C 0
A CTO RE DTJ 24 Ring Seal AO-2386 M-1 31-14 T
RF DJ2 CTC RE DTJ 24 Alternate Nitrogen Supply to T-NH-36049-14 B-1 0.375 CK SA NC C
O/C 0
A CTO RF DTJ 24 Ring Seal AO-2387 M-131-14 CTC RF DTJ 24 Alternate Nitrogen Supply to T-NH-36049-14 B-6 0375 CK SA NC C
0/C 0
A CTO RF DTJ 24 Ring Seal AO-2896 M-131-14 CTC REF DTJ24 Valve TablekPage 40 of 87
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Instrument and Service Air Conmponent Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ Instrument Air Supply to Actuator and T-Ring Seal AO 2379 NH-36049-14 M-131-14 B-5 0375 CK SA NC AC O/C C
A CTC RF DTJ 24 LT Y2 N/A CTO RF DTJ 24 Instrument Air Supply to NH-36049-14 C-4 0375 CK SA NC AC O/C C
A CTC RF DTJ 24 Actuator and T-Ring Seal AO-M-131-14 2380 LT Y2 NIA CTO RF DTJ 24 Alternate Nitrogen Supply to NH-36049-14 B-5 0375 CK SA NC C
O/C 0
A CTO Q
N/A Actuator and T-Ring Seal AO-M-131-14 2379 CTC RF DTJ-24 Alternate Nitrogen Supply to NH-36049-14 C-4 0375 CK SA NC C
O/C 0
A CTO Q
N/A Actuator and T-Ring Seal AO-M-131-14 2380 CTC RF DTJ-24 Service Air Inboard NH-36049-4 D-7 075 GT' MA 2
A LC C
P LT AJ N/A Containment Isolation Valve M-131-4 Service Air Inboard NH-36049-4 D-7 075 GT MA 2
A LC C
P LT AJ N/A Containment Isolation Valve M-131-4 Outboard Instrument Air NH-36049-12 B-6 20 GL AO 2
A 0
C A
STC Q
N/A Supply Containment Isolation M-131-12 Valve PIT Y2 N/A LT AJ N/A FC Q
N/A Valve Table Page 41 of 8 7 AI-683 AI-685 AI-694 AI-695 AS-78 AS-79 CV-1 478
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Instrument and Service Air Component Description CV-7956 Instrument Air Supply to Torus Ring Header Containment Isolation Valve PID Coord NH-36049-14 B-6 M-131-14 Size 075 Tjpe Act GL AO Class Cat Norm Pos Safe Pos A/P 2
A O/C c
A Valve Table Page 42 of 8 7 Test STC PIT LT FC Freq Q
Y2 AJ Q
RR/DTJ NIA NIA N/A N/A
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 Open (Dirty) Radwaste NH-36043 E-2 20 GT AO Drywell Floor Drain Sump M-137 Inboard Containment Isolation Valve 2
A 0
C A
STC Q
N/A PIT Y2 N/A LT AJ N/A FC 0
N/A Open (Dirty) Radwaste NH-36043 C-2 2.0 GT AO 2
A 0
C A
STC Q
N/A Drywell Floor Drain Sump M-137 PIT Y2 N/A Outboard Containment Isolation Valve LT AJ N/A FC Q
N/A Closed Radwaste Drywell NH-36044 E-2 20 GT AO 2
A 0
C A
STC Q
N/A Equipment Drain Sump M-138 PIT Y2 N/A Inboard Containment Isolation Valve LT AJ N/A FC 0
N/A Closed Radwaste Drywell NH-36044 E-1 20 GT AO 2
A 0
C A
STC Q
N/A Equipment Drain Sump M-138 Outboard ContainmentPIT Y2 N/A Isolation Valve LT AJ N/A FC Q
N/A
~
Open (Dirty) Radwaste NH-36043 E-2 0.5 RD SA NC D
C 0
A DT Y5 N/A Drywell Floor Drain Sump M-137 Pump Discharge Rupture Disk Closed Radwaste Drywell NH-36044 E-2 05 RD SA NC D
C 0
A DT Y5 N/A Equipment Sump Pump M-138 Discharge Rupture Disk (Clean)
Valve TablePage 43 of87 AO-2541A AO-2541 B AO-2561A AO-2561 B PSD-6047 PSD-6048
S..........................*L.
Monticello Nuclear Generating Plant Inservice Testing Program - Valves Description SJAE Mechanical Vacuum Pump Inlet Valve PID Coord NH-36035-2 A-3 M-104-1 Size 60 Type Act BF AO Class NC Cat Norm Pos Safe Pos B
O/C C
A/P Test A
STC FC Freq RR/DTJ CS DTJ10 Cs DTJ 10 SJAE Mechanical Vacuum NH-36035-2 A-3 6.0 BF AO NC B
01C C
A STC CS DTJ 10 Pump Inlet Valve M-1i04-1 FC CS DTJ10 System Main Condenser Component AO-1 825A
'
AO-1 825B Valve Table Page 44 of987
ýlf -
Monticello Nuclear Generating Plant Iiservice Testing Program - Valves System Main Steam Component Description PID Coord Size. Tjpe Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ Main Steam Isolation Valve NH-36241 (MSIV) Inboard M-115 C-5 1800 GL' AO I
A C
A STC Q
N/A FC CS DTJ 13 LT AJ N/A PIT Y2 N/A Main Steam Isolation Valve NH-36241 D-5 1800 GL AO 1
A 0
c A
STC Q
N/A (MSIV) Inboard M-115 FC CS DTJ 13 LT AJ N/A PIT Y2 N/A Main Steam Isolation Valve NH-36241 D-2 1800 GL AO 1
A 0
C-A STC Q
N/A (MSIV) Inboard M-1 15 FC CS DTJ 13 LT AJ N/A PIT Y2 N/A Main Steam Isolation Valve NH-36241 C-2 1800 GL AO 1
A Oc A
STC Q
N/A (MSIV) Inboard M-115 FC CS DTJ 13 LT AJ N/A PIT Y2 N/A Main Steam Isolation Valve NH-36241 C-5 1800 GT AO 1
A Oc A
STC Q
N/A (MSIV).butboard M-115 LT AJ NIA PIT Y2 N/A FC Q
N/A Main Steam Isolation Valve NH-36241 D-5 18 00 GT AO 1
A 0
c A
STC Q
N/A (MSIV) Outboard M-1 15 LT AJ N/A PIT Y2 N/A FC Q
N/A AO-2-80A AO-2-80B AO-2-80C AO-2-80D AO-2-86A AO-2-86B Valve Table Page 45 of 87
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Main Steam Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos AIP Test Freq RR/DTJ Main Steam Isolation Valve (MSlV) Outboard NH-36241 M-1 15 D-2 1800 GT AO I
A 0
C A
STC Q
N/A LT AJ N/A PIT Y2 N/A FC Q
N/A Main Steam Isolation Valve NH-36241 C-2 1800 GL AO I
A 0
C A
STC Q
N/A (MSIV) Outboard M-115 LT AJ N/A PIT Y2 N/A FC Q
N/A Main Steam Line Drain NH-36241 B-5 300 GT MO 1
A O/C C
A STC Q
N/A Isolation Valve (Inboard)
M-115 LT AJ N/A PIT Y2 N/A Main Steam Line Drain NH-36241 B-6 3.00 GT MO 1
A O/C C
A STC Q
N/A Isolation Valve (Outboard)
M-115 LT AJ N/A PIT Y2 N/A Flange Leakoff to Clean NH-36241 E-1 1 00 FV SA 2
C 0
C A
XT RF DTJ 22 Radwaste Excess Flow M-1 15 Check Valve Main Steam Line Excess Flow NH-36241 D-2 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Check Valve M-115 Main Steam Line Excess Flow NH-36241 D-5, 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Check Valve M-115 Valve Table Page 46 of 87 AO-2-86C AO-2-86D MO-2373 MO-2374 XFV-1 XFV-2 XFV-3
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 XMain Steam Line Excess Flow Check Valve NH-36241 M-115 D-5 100 FV SA I
C 0
C A
XT RF DTJ 22 Main Steam Line Excess Flow NH-36241 D-2 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Check Valve M-115 Main Steam Line Excess Flow NH-36241 B-2 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Check Valve M-1 15 Main Steam Line Excess Flow NH-36241 B-2 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Check Valve M-115 Main Steam Line Excess Flow NH-36241 B-5 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Check Valve M-1 15 Main Steam Line Excess Flow NH-36241 B-5 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Check Valve M-115 XFV-5 XFV-6 XFV-7 XFV-8 XFV-9 Valhe Table Page 47 of 87 XFV-4
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Post Accident Sampling Conmonent Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq JR/DTJ Post Accident Sampling Systerm to RHR System Check Valve NH-96042-1 NH-96042-1 A-6 075 CK SA 2
C O/C C
A CTC Q
N/A CTO Q
N/A Post Accident Samplri~g NH-96042-1 A-6 075 CK SA 2
C O/C C
A CTC Q
N/A System to RHR System NH-96042-1 Check Valve CTO Q
N/A Post Accident Sampling NH-96042-1 B-6 0.75 FV SA 2
C 0
C A
XT CS DTJ 20 System From RHR System NH-96042-1 Excess Flow Check Valve Post Accident Sampling NH-96042-1 B-6 075 FV SA 2
C 0
C A
XT CS DTJ 20 System From RHR System NH-96042-1 Excess Flow Check Valve PASS Sample Line Inboard NH-96042-1 C-7 075 GL SO 2
A O/C C
A STC Q
N/A Isolation Valve NH-96042-1 PIT Y2 N/A LT AJ N/A FC 0
N/A PASS Sample Line Outboard NH-96042-1 D-7 0.75 GL SO 2
A O/C C
A STC Q
N/A Isolation Valve NH-96042-1 PIT Y2 N/A LT AJ N/A FC 0
N/A Valm Table Page 48 of 87 PAS-58-1 PAS-58-2 PAS-59-5 PAS-59-6 SV-4081 SV-4082
Monticello Nuckar 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 AO-2378 AO-2379 AO-2380 AO-2381 AO-2382A Drywell Purge Supply NH-36258 G-2 180 BF AO 2
A O/C C
A STC 0
N/A Outboard Containment M-143 Isolation Valve FC Q
NIA PIT Y2 N/A LT AJ N/A Suppression Chamber Purge NH-36258 C-3 18.0 BF AO 2
A O/C C
A STC Q
N/A Supply Inboard Containment M-1 43 Isolation Valve FC Q
N/A PIT Y2 N/A LT AJ N/A 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 0
N/A FO Q
NIA PIT Y2 N/A LT AJ N/A 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-1 43 Valve STC Q
N/A FO 0
N/A PIT Y2 N/A LT AJ N/A Drywell Purge Supply Inboard NH-36258 C-3 180 BF AO 2
A 0C C
A STC 0
N/A Containment Isolation Valve M-1 43 FC Q
N/A PIT Y2 N/A LT AJ N/A Suppression Chamber to NH-36258 B-4 180 OK 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 Valme Table Page 49"of87
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,
Suppression Chamber to Drywell Vacuum Breaker Check Valve NH-36258 M-1 43 B-4 180 CK SA NC AC C
O/C A
CTO RF DTJ 28 CTC RF DTJ 28 LT Y2 N/A PIT Y2 N/A Suppression Chamber to NH-36258 B-4 180 CK SA NC AC C
O/C A
CTO RF DTJ 28 Drywell Vacuum Breaker M-1 43 Check Valve CTC RF DTJ 28 LT Y2 N/A PIT Y2 N/A Suppression Chamber to NH-36258 B-4 180 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 Suppression Chamber to NH-36258 B-4 180 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 Suppression Chamber to NH-36258 B-4 180 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 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-1 43 Check Valve CTC RF DTJ 28 LT Y2 N/A PIT Y2 N/A Vale Table Page 50 of 87 AO-2382B AO-2382C AO-2382E AO-2382F AO-2382G AO-2382H
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Primary Containment Component Description PID Coord Size Type Act Class Cat NormPos Safe Pos A/P Test Freq RR/DTJ AO-2382K AO-2383 AO-2386 AO-2387 AO-2896
-AO-4539 AO-4540 2
A C
c P
LT AJ N/A PIT -
Y2 N/A V al..---...
T able.P age.....f..
7...........
Valiv Table Page 51 of 8 7 Suppression Chamber to NH-36258 B-4 180 CK SA NC AC C
0/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 Suppression Chamber Vent to NH-36258 B-6 18.0 BF AO 2
A O/C C
A STC 0
N/A Standby Gas Containment M-143 Isolation Valve FC Q
N/A PIT Y2 N/A
" LT AJ N/A Drywell Vent to Standby Gas NH-36258 C-6 18.0 BF AO 2
A O/C C
A STC Q
NIA Containment Isolation Valve M-1 43 FC Q
N/A PIT
,Y2 N/A LT AJ N/A Drywell Vent to Standby Gas NH-36258 D-6 180 BF AO 2
A O/C C
A STC Q
N/A Containment Isolation Valve M-1 43 FC Q
N/A PIT Y2 N/A LT AJ N/A Suppression Chamber Vent to NH-36258 B-6 18 0 BF AO 2
A 0/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 Hard Pipe Vent Isolation Valve NH-116629 C-3 8.0 BF AO 2
A C
C.
P LT AJ N/A NH-116629 PIT Y2 N/A Hard Pipe Vent Isolation Valve NH-116629 C-4 8.0 BF AO N H-1i16629
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 Suppression Chamber Vent to NH-36258 Standby Gas Containment M-1 43 Isolation Valve B-6 20 GL AO 2
A O/C A
STC Q
N/A FC.
Q N/A PIT Y2 N/A LT AJ N/A Suppression Chamber Vent to NH-36258 C-6 20 GL AO 2
A 0/C C
A STC Q
N/A Standby Gas Containment M-1 43 Isolation Valve FC Q
NIA PIT Y2 N/A LT AJ N/A Torus Nitrogen Purge Inboard NH-46162 C-4 1.0 GL AO 2
A 0/C C
A STC Q
N/A Isolation Valve M-130 PIT Y2 N/A LT AJ N/A FC Q
NIA Drywell Nitrogen Purge 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 0
N/A Drywell/Torus Nitrogen Purge NH-46162 D-4 1 0 GL AO 2
A O/C c
A STC Q
N/A Outboard Isolation Valve M-1 30 PIT Y2 N/A LT AJ N/A FC Q
N/A Torus Outboard Isolation to NH-46162 C-5 1.0 GL AO 2
A O/C c
A STC Q
N/A CAM Analyzer B Supply Valve M-1 30 PIT Y2 N/A LT AJ N/A FC Q
N/A Vale Table Page 52 of 8 7 CV-2384 CV-2385 CV-3267 CV-3268 CV-3269 CV-3311
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Primary Containment Component Description PID Coord Size Type Act Class Cat Norm Pos SafePos A/P Test Freq RR/DTJ Torus Inboard Isolation Supply Valve to CAM Analyzer B NH-46162 M-130 B-5 1.0 GL AO 2
A O/C C
A STC Q
N/A PIT Y2 N/A LT AJ N/A FC Q
N/A 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 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 NIA LT AJ N/A FC Q
N/A Suppression Chamber NH-36258 B-1 20.0 CK SA 2
AC C
O/C A
CTO 0
N/A Vacuum Relief Check Valve M-143 CTC Q
N/A LT AJ N/A 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 Hard Pipe Vent Une Rupture NH-116629 C-6 100 RD SA NC D
C 0
A DT Y5 N/A Disk NH-116629 Drywellito CAM Sample Line NH-46162 C-5 075 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 53 of.87 CV-3312 CV-3313 CV-3314 DWV-8-1 DWV-8-2 PSD-4543 SV-3307
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Primary Containment Component Description SV-3308 Drywell to CAM Sample Line Inboard Isolation Valve PID Coord NH-46162 M-130 C-5 Size 0.75 Type Act GL SO Class Cat Norm Pos Safe Pos A/P Test 2
A 0/C C
A STC PIT LT FC Valve Table Page 54 of 87 Freq Q
Y2 AJ Q
RR/DTJ NIA N/A N/A N/A
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor and Vessel Assembly Component Description PID Coord Size Tjpe Act Class Cat Norm Pos Safe Pos A/P Test Freq RRIDTJ Reactor Pressure Vessel Head Vent Isolation Valve NH-36241 M-115 E-5 075 GL AO 2
B C
C P
PIT Y2 NIA Reactor Pressure Vessel NH-36241 E-5 075 GL AO 2
B C
C P
PIT Y2 N/A Head Vent Isolation Valve M-1 15 Reactor Vessel Above Core NH-36242-1 C-5 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Plate Pressure M-1 16-1 Instrumentation Excess Flow Check Valve Reactor Vessel Below Core NH-36242-1 C-5 1.00 FV SA 1
C 0
C A
XT RF DTJ 22 Plate Pressure M-1 16-1 Instrumentation Excess Flow Check Valve Reactor Vessel Reference NH-36242 B-5 1.00 FV SA 1
C 0
C A
XT RF DTJ 22 Leg Instrumentation Excess M-116 Flow Check Valve Reactor Vessel Reference NH-36242 D-5 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Leg Instrumentation Excess M-1 16 Flow Check Valve Reactor Vessel Reference NH-36242 C-3 1.00 FV SA 1
C 0
C A
XT RF DTJ 22 Leg Instrumentation Excess M-1 16 Flow Check Valve Reactor Vessel Jet Pump #4 NH-36242-1 A-5 1.00 FV SA 1
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve CV-2371 CV-2372 XFV-22 XFV-23 XFV-24 XFV-25 XFV-26 XFV-27 Valve Table Page 55 of 8 7
S.*.....*
- ......'....................
.*.*
Monticello Nuclear Generating Plant hIservice Testing Prograimn-Valves System Reactor and Vessel Assembly Conmponent 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
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow Check Valve M-116-1 Reactor Vessel Jet Pump #2 NH-36242-1 A-5 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve Reactor Vessel Jet Pump #1 NH-36242-1 A-5 1.00 FV SA 1
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve Reactor Vessel Jet Pump #1 NH-36242-1 A-5 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve Reactor Vessel Jet Pump #10 NH-36242-1 A-5 1.00 FV SA 1
C.0 C
A XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve Reactor Vessel Jet Pump #5 NH-36242-1 A-5 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve Reactor Vessel Jet Pump #6 NH-36242-1 A-5 1 00 FV SA 1
C 0
C.
A XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve Reactor Vessel Jet Pump #7 NH-36242-1 A-5 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve Valme Table Page 56 of 87 XFV-29 XFV-30 XFV-31 XFV-32 XFV-33 XFV-34 XFV-35
Monticello Nuclear Generathig Plant Inservice Testing Program - Valves Siystem Reactor and Vessel Assembly Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ Reactor Vessel Jet Pump #8 Instrumentation Excess Flow Check Valve NH-36242-1 M-1 16-1 A-5 1.00 FV SA 1
C 0
C A
XT RF DTJ 22 Reactor Vessel Jet Pump #9 NH-36242-1 A-5 1.00 FV SA 1
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve Reactor Vessel Jet Pump #6 NH-36242-1 A-5 1.00 FV SA 1
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve Reactor Vessel Jet Pump #20 NH-36242-1 A-5 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve Reactor Vessel Jet Pump #14 NH-36242-1 A-5 100 FV SA 1
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve Reactor Vessel Jet Pump #12 NH-36242-1 A-5 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve Reactor Vessel Jet Pump #13 NH-36242-1 A-5 1.00 FV SA I
-C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve Reactor Vessel Jet Pump #11 NH-36242-1 A-5 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve Vale Table Page 57 of 87 XFV-36 XFV-37 XFV-38 XFV-39 XFV-40 XFV-41 XFV-42 XFV-43
Monticello Nuclear Generating Plant Inservice Testing Prograin - Valves System Reactor and Vessel Assembly Component Description PID Coord Size TIpe Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ
.rC11 AA
I,,,
- n.
- Reactor vessel Jet P'ump #11 Instrumentation Excess Flow Check Valve NH-36242-1 M-116-1 A-5 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Reactor Vessel Jet Pump #15 NH-36242-1 A-5 1.00 FV SA 1
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve Reactor Vessel Jet Pump #19 NH-36242-1 A-5 1.00 FV SA 1
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1, Check Valve Reactor Vessel Jet Pump #17 NH-36242-1 A-5 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve Reactor Vessel Jet Pump #18 NH-36242-1 A-5 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve Reactor Vessel Jet Pump #16 NH-36242-1 A-5 1.00 FV SA 1
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-116-1 Check Valve Reactor Vessel Jet Pump #16 NH-36242-1 A-5 1.00 FV SA 1
C 0
C A
XT RF DTJ 22 Instrumentation Excess Flow M-1 16-1 Check Valve Reactor Vessel Below Core NH-36242-1 C-3 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Plate Flow Instrumentation M-1 16-1 Excess Flow Check Valve XFV-44 XFV-45 XFV-46 XFV-47 XFV-48 XFV-49 XFV-50 XFV-51 Valve Table Page 58 of 87
Monticello Nuclear Generaiing Plant Inservice Testing Program - Valves Systenz 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
C 0
C A
XT RF DTJ 22 Leg Instrumentation Excess Flow Check Valve M-1 16 Reactor Vessel Reference NH-36242 D-3 1.00 FV SA 1
C 0
C A
XT RF DTJ 22 Leg Instrumentation Excess M-1 16 Flow Check Valve Reactor Vessel Reference NH-36242 C-5 1 00 FV SA I
C 0
C A
XT RF DTJ 22 Leg Instrumentation Excess M-1 16 Flow Check Valve Reactor Vessel Reference NH-36242 C-3 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Leg Instrumentation Excess M-116 Flow Check Valve Reactor Vessel Reference NH-36242 C-5 1.00 FV SA 1
C 0
C A
XT RF DTJ 22 Leg Instrumentation Excess M-1 16 Flow Check Valve Reactor Vessel Reference NH-36242 B-3 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Leg Instrumentation Excess M-1 16 Flow Check Valve Reactor Vessel Reference NH-36242 D-5 1.00 FV SA 1
C 0
C A
XT RF DTJ 22 Leg Instrumentation Excess M-1 16 Flow Check Valve Reactor Vessel Reference NH-36242 D-3 1 00 FV SA 1
C 0
C A
XT RF DTJ 22 Leg Instrumentation Excess M-1 16 Flow Check Valve Valve Table Page 59 of 8 7 XFV-53 XFV-54 XFV-55 XFV-56 XFV-57 XFV-88 XFV-89
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor Building Closed Cooling Water Conponent - Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ RBCCW Return Line Inboard Containment Isolation Valve NH-36042-2 M-111-1 C-7 80 GT MO 2
A 0
C A
STC CS DTJ 11 PIT Y2 N/A LT AJ N/A RBCCW Supply Line.
NH-36042-2 D-3 8.0 GT MO 2
A 0
C A
STC CS DTJ 11 Outboard Containment M-111-1 Isolation Valve PIT Y2 N/A LT AJ N/A RBCCW Return Line NH-36042-2 D-7 80 GT MO 2
A 0
C A
STC CS DTJ 11 Outboard Containment M-1 11-1 Isolation Valve PIT Y2 N/A LT AJ N/A RBCCW Supply Line Inboard NH-36042-2 C-3 8 0 CK SA 2
AC 0
C A
CTC RF DTJ 01 Containment Isolation Valve M-111-1 LT AJ N/A CTO 0
NIA Malre Table Page 60 of 87 MU-1426i MO-4229 MO-4230 RBCC-15
Monticello Nuclear Generating Plant hIservice Testing Program - Valves System Reactor Core Isolation Cooling Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos AIP Test Freq RR/DTJ RCIC Injection Check Valve NH-36252 B-5 40 CK AO M-126 2
C C
0/C A
DI RF DTJ 15 RCIC Steam Line Drain NH-36251 B-1 1 0 GL AO 2
B 0
C A
STC Q
N/A Isolation Valve M-125 FC Q
NIA PIT Y2 N/A RCIC Pump Minimum Flow NH-36252 A-2 20 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 RCIC Steam Supply Inboard NH-36251 E-5 30 GT MO 1
A 0
O/C A
LT AJ N/A Containment Isolation Valve M-125 STC.
Q N/A PIT Y2 N/A STO Q
N/A RCIC Steam Supply Outboard NH-36251 E-4 30 GT MO 1
A 0
0/C A
LT AJ N/A Containment Isolation Valve M-125 STC 0
N/A PIT Y2 N/A STO Q
N/A RCIC Turbine Steam Supply NH-36251 D-2 3 0 GL
'MO 2
B C
O/C A
STO Q
N/A Isolation Valve M-125 STC 0
N/A PIT Y2 N/A RCIC Lube Oil to Barometric NH-36251 A-4 20 GL MO 2
B C
O/C A
STO Q
N/A Condenser Cooling Isolation M-125 Valve STC Q
N/A PIT Y2 N/A Valve Table Page 61 of 87 AO-13-22 CV-2082A CV-2104 MO-2075 MO-2076 MO-2078 MO-2096
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 RRIDTJ RCIC Suppression Pool Suction Isolation Valve NH-36252 M-126 A-5 60 GT MO 2
B C
0 A
STO Q
N/A PIT Y2 N/A RCIC Suppression Pool NH-36252 D-4 60 GT MO 2
B C
0 A
STO 0
N/A Suction Isolation Valve M-126 PIT Y2 N/A RCIC Condensate Storage NH-36252 D-3 60 GT MO 2
B 0
O/C A
STO Q
N/A Tank Suction Valve M-126 STC 0
N/A PIT Y2 N/A RCIC Pump Discharge Valve NH-36252 B-5 40 GT MO 2
B O/C 0
A STO Q
N/A M-126 PIT Y2 N/A RCIC Pump Discharge NH-36252 B-5 40 GT MO 2
B O/C 0
A STO 0
N/A Valve - Injection Line to M-126 Feedwater Isolation PIT Y2 N/A RCIC Test Return Line NH-36252 D-5 40 GT MO 2
B O/C C
A STC Q
N/A Isolation Valve M-126 PIT Y2 N/A RCIC Turbine Exhaust Line NH-36251 C-5 80 RD SA 2
D C
O/C A
DT Y5 N/A Rupture Disc M-125 RCIC Turbine Exhaust Line NH-36251 C-5 80 RD SA NC D
C O/C A
DT Y5 N/A Rupture Disc M-125 Valtve Table Page 62 of 87 MO-2100 MO-2101 MO-2102 MO-2106 MO-2107 MO-3502 PSD-2089 PSD-2090
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 Turbine Exhaust Check NH-36251 Valve M-125 C-6 8.0 CK SA 2
AC C
O/C A
CTO Q-N/A CTC CS DTJ 18 LT AJ' N/A RCIC Condensate Return NH-36251 C-5 20 CK SA 2
C C
O/C A
CTO Q
N/A Check Valve M-125 CTC 0
N/A RCIC Barometric Condenser NH-36251 A-5 20 CK SA 2
AC O/C C
A CTC CS DTJ 18 Vacuum Pump Discharge to M-125 Suppression Pool Check Valve CTO Q
N/A RCIC Barometric Condenser NH-36251 A-5 20 CK SA 2
AC O/C C
A CTC CS DTJ 18 Vacuum Pump Discharge to M-125 Suppression Pool Check Valve CTO Q
N/A RCIC Torus Suction Supply NH-36252 A-4 60 CK SA 2
C C
O/C A
DI RF DTJ 18 Line Check Valve M-1 26 RCIC Minimum Flow Line to NH-36252 A-5 20 CK SA 2
C O/C 0
A DI RF DTJ 30 Suppression Pool Check Valve M-1 26 RCIC Condensate Storage NH-36252 D-3 60 CK SA 2
C O/C O/C A
CTO Q
N/A Tank Suction Check Valve M-1 26 CTC CS DTJ 29 RCIC Exhaust Line Vacuum NH-36251 B-6 1 5 CK SA 2
AC C
O/C A
CTC CS DTJ 18 Breaker Check Valve M-1 25 CTO CS DTJ 18 Val*v Table Page 63 of 8 7 RCIC-10 RCIC-14 RCIC-16 RCIC-17 RCIC-31 RCIC-37 RCIC-41 RCIC-57
Monticello Nuclear Generating Plant Inservice Testing Programn-Valves Siystem Reactor Core Isolation Cooling Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ RCIC Exhaust Line Vacuum Breaker Check Valve NH-36251 M-125 B-6 1.5 CK SA 2
AC C
O/C A
CTC CS DTJ 18 CTO CS DTJ 18 RCIC Turbine Exhaust Check NH-36251 C-5 80 CK SA 2
AC C
OIC A
CTO Q
N/A Valve M-125 CTC CS DTJ 18 LT AJ N/A 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 RCIC Suction Supply Line NH-36252 D-3 1.0 RV SA 2
C C
O/C A
RT SY10 NIA Relief Valve M-1 26 RCIC Turbine Steam Supply NH-36251 D-5 1.0 FV SA I
C 0
C A
XT RF DTJ 22 Line Excess Flow Check Valve M-1 25 RCIC Turbine Steam Supply NH-36251 D-5 1.0 FV SA 1
C 0
C A
XT RF DTJ 22 Line Excess Flow Check Valve M-125 Valve Table Page 64 of 87 RCIC-59 RCIC-9 RV-2097 RV-2103 XFV-86 XFV-87
S..........................
Monticello Nuclear Generating Plant Iiservice Testing Program - Valves System Reactor Recirculation Conponent Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RRIDTJ Reactor Water Sample Inboard Containment Isolation Valve NH-36243 M-1 17 D-5 0.75 GL AO 2
A 0/C C
A STC Q
N/A PIT Y2 N/A LT AJ N/A FC Q
N/A Reactor Water Sample NH-36243 D-6 075 GL AO 2
A O/C C
A STC Q
N/A Outboard Containment M-1 17 Isolation Valve PIT Y2 N/A LT AJ N/A FC Q
N/A Reactor Recirculation Pump NH-36243 B-2 280 GT MO 1
B 0
C A
STC CS DTJ 14 P-200B Discharge Valve M-117 PIT Y2
.N/A Reactor Recirculation Pump NH-36243 B-6 280 GT MO 1
B 0
C A
STC CS DTJ 14 P-200A Discharge Valve M-1 17 PIT Y2 N/A Reactor Recirculation System NH-36243 B-2 1 0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-1 17 Reactor Recirculation System NH-36243 B-2 1 0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-117 Reactor Recirculation System NH-36243 A-3 1 0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-1 17 Valre Table Page 65 of 87 CV-2790 CV-2791 MO-2-53A MO-2-53B XFV-58 XFV-59 XFV-60
K......
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor Recirculation Contponent XFV-61 Val*e Table Page 66 of.87 Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A"P Test Freq RR/DTJ Reactor Recirculation System NH-36243 B-6 1 0 FV SA I
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-1 17 Reactor Recirculation System NH-36243 B-6 1.0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-117 Reactor Recirculation System NH-36243 A-5 1.0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve, M-1 17 Reactor Recirculation System NH-36243 D-5 1 0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-1 17 Reactor Recirculation System NH-36243 D-5 1.0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-1 17 Reactor Recirculation System NH-36243 E-2 1.0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-1 17 Reactor Recirculation System NH-36243 C-5 1.0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-1 17 Reactor Recirculation System NH-36243 C-5 1.0 FV SA I
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-1 17 XFV-62 XFV-63 XFV-64 XFV-65 XFV-66 XFV-67 XFV-68
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor Recirculation Component Description PID Coord Size Tjpe Act Class Cat NormPos SafePos A/P Test Freq RR/DTJ XFV-69 Reactor Recirculation System NH-36243 E-6 1 0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-1 17 Reactor Recirculation System NH-36243 E-6 1 0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-1 17 Reactor Recirculation System NH-36243 E-2 1 0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-1 17 Reactor Recirculation System NH-36243 E-6 1.0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-117 Reactor Recirculation System NH-36243 E-6 1 0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-1 17 Reactor Recirculation System NH-36243 E-2 1.0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-117 Reactor Recirculation System NH-36243 E-2 1 0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-1 17 Reactor Recirculation System NH-36243 D-3 1.0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-1 17 Valve Table Page 67 of'87 XFV-70 XFV-71 XFV-72 XFV-73 XFV-74 XFV-75 XFV-76
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor Recirculation Conponent Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos AIP Test Freq RR/DTJ XFV-77 Reactor Recirculation System NH-36243 D-3 1.0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-1 17 Reactor Recirculation System NH-36243-1 C-3 1 0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-1 17 Reactor Recirculation System NH-36243-1 C-3 1.0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-117 Reactor Recirculation System NH-36243-1 C-4 1 0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-1 17 Reactor Recirculation System NH-36243-1 C-4 1.0 FV SA 1
C 0
C A
XT RF DTJ 22 Excess Flow Check Valve M-1 17 Recirc Pump (11) Seal NH-36244 A-3 1.0 CK SA 2
AC O/C C
A CTC RF DTJ03 Outboard Seal Injection M-1 18 Check Valve LT AJ N/A CTO 0
N/A Recirc Pump (12) Seal NH-36244 A-3 1.0 CK SA 2
AC O/C C
A CTC RF DTJ 03 Outboard Seal Injection M-118 Check Valve LT AJ NIA CTO Q
N/A Reactor Recirculation NH-36243-1 D-3 1.0 CK SA 2
AC O/C C
A LT AJ N/A System Loop 11 Inboard Seal M-1 17 CTC RE DTJ03 Injection Check Valve CTO Q
N/A I1%*
.%~w*Im*~
- o%*.**~
o*
- t
- 4****..
Valve Table Page 68 of 87 XFV-78 XFV-79 XFV-80 XFV-81 XR-25-1 XR-25-2 XR-27-1
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Reactor Recirculation Component Description PID Coord Size Tjpe Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ XR-27-2 Reactor Recirculation System NH-36243-1 D-5 1 0 CK SA 2
AC O/C C
A LT AJ N/A Loop 12 Inboard Sea! Injection M-117 Check Valve Val*e Table Page 69 of987 CTC CTO RF Q
DTJ 03 N/A
Monticello Nuclear Generating Plant Inservice Testing Program - Valves Siystem Reactor Water Cleanup Component Description PID Coord Size Tjpe Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ RWCU Supply Line Inboard NH-36254 Containment Isolation Valve M-1 28 C-8 40 GT MO I
A 0
c A
Y2 AJ N/A N/A N/A RWCU Supply Line Outboard NH-36254 C-7 40 GT MO I
A 0
C A
STC Q
N/A Containment Isolation Valve M-1 28 PIT Y2 N/A LT AJ N/A RWCU Return Line Check NH-36254 D-6 1.5 CK SA 2
C O/C C
A CTC RF DTJ 08 Valve M-128 CTO RF DTJ08 RWCU Return Line Check NH-36254 D-7 1.5 CK SA 2
C O/C C
A CTC RF DTJ 08 Valve M-128 CTO RF DTJ08 Valve Tabk Page 70 of 87 MO-2397 MO-2398 RC-6-1
- RC-6-2
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 RRIDTJ AO-10-46A Residual Heat Removal Injection Check Valve NH-36247 C-5 160 CK SA M-121 I
AC O/C O/C A
CTO CS DTJ 26 CTC RF DTJ 26 LT Y2 N/A Residual Heat Removal NH-36246 D-2 160 CK SA 1
AC C
O/C A
CTO CS DTJ 26 Injection Check Valve M-1 20 CTC RF DTJ26 LT Y2 N/A Residual Heat Removal Pump NH-36247 B-4 20 GL AO 2
B 0/C O/C A
STO Q
N/A A Minimum Flow Isolation M-1 21 Valve STC Q
N/A FO 0
N/A PIT Y2 N/A Residual Heat Removal Pump NH-36246 B-4 2.0 GL AO 2
B O/C 0/C A
STO Q
N/A B Minimum Flow Isolation M-120 Valve STC Q
N/A FO Q
N/A PIT Y2 N/A Residual Heat Removal Pump NH-36247 B-5 2.0 GL AO 2
B O/C -
O/C A
STO Q
N/A C Minimum Flow Isolation M-121 Valve STC 0
N/A FO 0
N/A PIT Y2 N/A Residual Heat Removal Pump NH-36246 C-4 20 GL AO 2
B 0/C O/C A
STO 0
N/A D Minimum Flow Isolation M-120 Valve STC Q
N/A FO Q
NIA PIT Y2 N/A RHR Pump Torus Suction NH-36247 B-6 200 GT MO 2
B 0
0 P
PIT Y2 Isolation Valve M-121 AO-10-46B CV-1994 CV-1995 CV-1996 CV-1997 MO-1986 Valtve Table Page 71 of 87
/
.
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 RRIDTJ RHR Pump Tows Suction NH-36246 Isolation Valve M-120 B-2 200 GT MO 2
B O
RHR Shutdown Cooling NH-36247 B-6 18.0 GT MO 2
B C
C P
PIT Y2 Suction Isolation Valve M-121 RHR Shutdown Cooling NH-36246 A-2 180 GT MO 2
B C
C P
PIT Y2 Suction Isolation Valve M-1 20 Residual Heat Removal Heat NH-36247 B-3 140 GL MO 2
B 0
O/C A
STO Q
N/A Exchanger Bypass Isolation M-1 21 Valve STC Q
N/A PIT Y2 N/A Residual Heat Removal Heat NH-36246 B-5 14.0 GL MO 2
B 0
O/C A
STO Q
N/A Exchanger Bypass Isolation M-120 Valve STC 0
N/A PIT Y2 N/A Residual Heat Removal to NH-36247 D-3 120 GT MO 2
B C
O/C A
STO Q
N/A Torus Discharge Isolation M-121 Valve STC Q
N/A PIT Y2 N/A Residual Heat Removal to NH-36246 D-5 120 GT MO 2
B C
O/Ce A
STO Q
N/A Tows Discharge Isolation M-120 Valve STC Q
N/A PIT Y2 N/A Residual Heat Removal to NH-36247 C-3 10.0 GL MO 2
B C
O/C A
STO Q
N/A Torus Cooling Isolation Valve M-121 STC Q
N/A PIT Y2 N/A s.s.s..v.
Y*..*° v.*.
y.. %.%*y~p J*.*¢ Valve Table Page 72 of8 7 MO-1i987 O
P PIT Y2 MO-1988 MO-1989 MO-2002 MO-2003 MO-2006 MO-2007 MO-2008
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 Residual Heat Removal to NH-36246 C-6 100 GL MO Torus Cooling Isolation Valve M-1 20 2
B C
O/C A
STO Q
N/A STC 0
N/A PIT Y2 N/A MO-2009 MO-2010 MO-2011 MO-2012 MO-2013 MO-2014 Valve Table Page 73 of.87 Residual Heat Removal to NH-36247 C-3 40 GL MO 2
A C
01C A
STO Q
N/A Torus Spray Isolation Valve M-1 21 STC Q
N/A PIT Y2 N/A LT AJ N/A Residual Heat Removal to NH-36246 C-5 40 GL MO 2
A C
O/C A
STO Q
N/A Torus Spray Isolation Valve M-1 20 STC Q
NIA PIT Y2 N/A LT AJ N/A Residual Heat Removal LPCI NH-36247 C-4 160 GL MO 2
A 0
O/C A
STO Q
N/A Outboard Isolation Valve M-121 STC 0
N/A PIT
- Y2 N/A LT AJ N/A Residual Heat Removal LPC[
NH-36246 C-3 160 GL MO 2
A 0
O/C A
STO Q
N/A Outboard Isolation Valve M-120 N
STC O
N/A PIT Y2 N/A LT AJ N/A Residual Heat Removal LPCI NH-36247 C-5 16.0 GT MO 1
A C
O/C A
STO Q
N/A Inboard Isolation Valve M-1 21 STC 0
N/A PIT Y2 N/A LT AJ" N/A LT Y2 N/A
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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 Residual Heat Removal LPCI NH-36246 C-2 16.0 GT MO Inboard Isolation Valve M-120 1
A C
O/C A
STO Q
N/A STC Q
N/A PIT Y2 N/A LT AJ N/A IT V2 I tA MO-2015 MO-2020 MO-2021 MO-2022 MO-2023 MO-2026 Residual Heat Removal NH-36247 E-5 10.0 GT MO 2
A C
O/C A
STO Q
N/A Containment Spray Outboard M-121 Isolation Valve STC Q
NIA PIT Y2 N/A LT AJ N/A NIA Residual Heat Removal NH-36246 E-2 10.0 GT MO 2
A C
O/C A
STO Q
N/A Containment Spray Outboard M-120 Isolation Valve STC Q,
N/A PIT Y2 NIA LT AJ N/A Residual Heat Removal NH-36247 E-5 10.0 GT MO 2
A C
O/C A
STO Q
NIA Containment Spray Inboard M-1 21 Isolation Valve STC Q
N/A PIT Y2 N/A LT AJ N/A Residual Heat Removal NH-36246 E-2 10.0 GT MO 2
A C
O/C A
STO Q
N/A Containment Spray Inboard M-120 Isolation Valve STC
,Q N/A PIT Y2 NIA LT AJ N/A Reactor Vessel Head Spray NH-36247 D-6 40 GT MO 1
A O/C C
A STC CS DTJ 16 Outboard Isolation Valve M-121 PIT Y2 N/A LT AJ N/A LT Y2 N/A Valhv Table Page 74 of 87
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Monticello Nuclear Generating Plant Inservice Testing Programn-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-2027 Reactor Vessel Head Spray NH-36247 D-6 40 GT MO I
A O/C O/C A
STC CS DTJ 16 Inboard Isolation Valve M-121 PIT Y2 N/A LT AJ N/A LT Y2 N/A LT GLCS DTJ16 Reactor Coolant to RHR NH-36247 D-6 18.0 GT MO 1
A O/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 Reactor Coolant to RHR NH-36247 C-6 180 GT MO 1
A O/C C
A STC CS DTJ16 Shutdown Cooling Supply M-121 Isolation Valve PIT Y2 N/A LT AJ N/A LT Y2 N/A RHR Discharge to Waste NH-36247 G-3 40 GT MO 2
B O/C C
A STC Q
N/A Gas Surge Tank Isolation M-121 Valve PIT Y2 N/A Residual Heat Removal Cross NH-36246 C-6 14.0 GT MO 2
B 0
0 P
PIT Y2 N/A Tie Isolation Valve M-120 Residual Heat Removal NH-36247 C-6 4.0 GT MO I
B 0/C C
A STC Q
N/A Intertie Isolation Valve M-1 21 PIT Y2 N/A Residual Heat Removal NH-36246 C-1 4.0 GT MO 1
B O/C C
A STC Q
N/A Intertie Isolation Valve M-120 PIT Y2 N/A MO-2029 MO-2030 MO-2032 MO-2033 MO-4085A MO-4085B Valve Tabk Page 75 of 87 I -
Monticello Nuclear Generating Plant Inservice Testing Programh-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-4086 Residual Heat Removal NH-36247 C-6 40 GT MO 1
B 0
0 P
PIT Y2 N/A Intertie Isolation Valve M-121 RHR-2-1 Residual Heat Removal Pump NH-36247 A-4 10.0 CK SA 2
C O/C O/C A
CTO Q
N/A A Discharge Check Valve M-121 CTC Q
N/A RHR-2-2 Residual Heat Removal Pump NH-36246 B-5 10.0 CK SA 2
C O/C O/C A
CTO Q
NIA B Discharge Check Valve M-120 CTC O
N/A RHR-2-3 Residual Heat Removal Pump NH-36247 B-4 10.0 CK SA 2
C O/C O/C A
CTO Q
N/A C Discharge Check Valve M-1 21 CTC 0
N/A RHR-2-4 Residual Heat Removal Pump NH-36246 B-5' 10.0 CK SA 2
C O/C O/C A
CTO Q
N/A D Discharge Check Valve M-1 20 CTC Q
N/A RHR-6-1 Residual Heat Removal NH-36247 C-5 16 0 GT MA 1
B LO 0
P PIT Y2 N/A Injection Line Manual Isolation M-121 Valve RHR-6-2 Residual Heat Removal NH-36246 D-1 160 GT MA 1
B LO 0
P PIT Y2 N/A Injection Line Manual Isolation M-120 Valve RHR-81 Thermal Overpressurization NH-36247 D-4 1
CK SA I
AC C
0/C A
LT Y2 N/A Relief Check Valve for M-1 21 Penetration X-12 CTO CS DTJ-34 CTC CS DTJ-34 LT AJ N/A Vale Table Page 76 of 8 7
System Residual Heat Removal Component RHR-8-1 RHR-8-2 RV-1990 RV-1991 RV-1 992 RV-1993 RV-2004 RV-2005 4
Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq,RR/DTJ Residual Heat Removal NH-36247 C-5 3.0 CK SA 2
C O/C 0
A DI SY8 DTJ 06 Minimum Flow Check Valve M-1 21 Residual Heat Removal NH-36246 C-3 3.0 CK SA 2
C O/C 0
A DI SY8 DTJ 06 Minimum Flow Check Valve M-120 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 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 Residual Heat Removal NH-36247 B-5 1.0 RV SA 2
C C
O/C A
RT SYI0 N/A Suction Relief Valve M-1 21 Residual Heat Removal NH-36246 C-3 1 0 RV SA 2
C C
O/C A
RT SYIO N/A Suction Relief Valve M-120 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 Residual Heat Removal NH-36246 D-6 1.0 RV SA 2
C C
O/C A
RT SYI0 N/A Discharge Relief Valve M-120 Valv* Table Page 77 of 87 Monticello Nuclear Generating Plant Inservice Testing Program - Valves
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Residual Heat Removal Description PID Reactor Vessel Head Spray NH-36247 Line Relief Valve M-1 21 Coord E-4 Size 1.0 Type Act RV SA Class 2
Cat Norm Pos Safe Pos A/P Test C
C O/C A
RT Freq SY10 RR/DTJ N/A 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 Residual Heat Removal Heat NH-36247 B-2 2.5 RV SA 2
C C
O/C A
RT SY1O N/A Exchanger Shell Side Relief M-121 Valve Residual Heat Removal Heat NH-36246 B-6 2.5 RV SA 2
C C
O/C A
RT SY1O N/A Exchanger Shell Side Relief M-120 Valve Valie Table Page 78 of 87 Component RV-2025 RV-2031 RV-4281 RV-4282
Monticello Nuclear Generatilg 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 RHR Service Water Pump P-NH-36665 109A Air Vent Valve M-811 B-3 30 AR SA 3
C O/C O/C A
CTC Q
N/A CTO Q
N/A RHR Service Water Pump P-NH-36665 B-8 30 AR SA 3
C O/C 0/C A
CTC Q
N/A 109D Air Vent Valve M-811 CTO 0
N/A RHR Service Water Pump P-NH-36665 B-3 30 AR SA 3
C O/C 0/C A
CTC Q
N/A 109C Air Vent Valve M-81 1 CTO Q
N/A RHR Service Water Pump P-NH-36665 B-8 30 AR SA 3
C 0/C O/C A
CTC Q
N/A 109B Air Vent Valve M-811 CTO Q
N/A RHR Serce Water to RHR NH-36664 A-5 120 GL AO 3
B O/C 0
A STO Q
VR03 Heat Exchanger Flow Control M-1 12 Valve PIT Y2 N/A RHR Service Water to RHR NH-36664 A-4 12.0 GL AO 3
B 0/C 0
A STO Q
VR 03 Heat Exchanger Flow Control M-112 Valve PIT Y2 NIA RHR Service Water Pump P-NH-36665 B-3 120 CK SA 3
C 0/C 0/C A
CTO Q
N/A 109A Discharge Check Valve M-81 1 CTC Q
NIA S~~~-
~~------------------------.
RHR Service Water Pump P-NH-36665 B-8 120 CK SA 3
C 0/C O/C A
CTO Q
N/A 109B Discharge Check Valve M-811I CTC Q
N/A Valtv Table Page 79 of 8 7 AV-3147 AV-3148 AV-3149 AV-3150 CV-1 728 CV-1729 RHRSW-1-1 RHRSW-1-2
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System RHR Service Water Component Description PID Coord Size Tjpe Act Class Cat Norm Pos Safe Pbs A/P Test Freq RR/DTJ RHRSW-1-3 RHR Service Water Pump P 109C Discharge Check Valve RHRSW-1-4 RHRSW-3-1 RHRSW-3-2 RHRSW-48-1 RHRSW-48-2 RHRSW-50-1 RHRSW-50-2 NH-36665 M-811 B-3 120 CK SA 3
C O/C O/C A
CTO Q
N/A CTC Q
N/A RHR Service Water Pump P-NH-36665 B-8 12.0 CK SA 3
C OlC O/C A
CTO Q
N/A 109D Discharge Check Valve M-811N CTC 0
N/A RHR Service Water Basket NH-36665 C-3 150 GT MA 3
B LC O/C A
STO Y2 NIA Strainer Bypass Isolation M-811 Valve STC Y2 N/A RHR Service Watet Basket NH-36665 C-7 180 GT MA 3
B LC O/C A
STO Y2 N/A Strainer Bypass Isolation M-811 Valve STC Y2 N/A A RHR Service Water Biocide NH-36665 B-3 05 CK SA 3
C O/C C
A CTC Q
N/A Injection Check Valve M-811 CTO 0
N/A B RHR Service Water Biocide NH-36665 B-7 05 CK SA 3
C O/C C
A CTC Q
N/A Injection Check Valve M-81I1 CTO Q
N/A 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 0
N/A A RHR Service Water NH-36665 B-7 05 CK SA 3
C O/C C
A CTC Q
N/A Dispersent Check Valve M-811 CTO 0
N/A Valve Table Page 80 of 87
Monticello Nuclear Generating Plant Inservice Testing Program Valves System RHR Service Water Component Description PID Coord Size Tjpe Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ RHRSW-57-1 RHR Service Water Pump P-NH-36665 109A Motor Cooling Water M-811 Supply Check Valve B-2 10 CK SA 3
C O/C 0
A CTO Q
N/A CTC CS DTJ 04 RHR Service Water Pump P-NH-36665 B-7 1 0 CK SA 3
C O/C 0
A CTO Q
N/A 109B Motor Cooling Water M-811 Supply Check Valve CTC CS DTJ 04 RHR Service Water Pump P-NH-36665 B-2 1.0 CK SA 3
C O/C 0
A CTO Q
N/A 109C Motor Cooling Water M-81 1 Supply Check Valve CTC CS DTJ 04 RHR Service Water Pump P-NH-36665 B-7 1.0 CK SA 3
C O/C 0
A CTO Q
N/A 109D Motor Cooling Water M-811 Supply Check Valve CTC CS DTJ 04 RHR Service Water Pump NH-36665 C-2 1 0 RV SA 3
C C
O/C A
RT SYIO N/A Cooling Water Supply Relief M-811 Valve RHR Service Water Pump NH-36665 B-7 10 RV SA 3
C C
0/C A
RT SYl0 N/A Cooling Water Supply Relief M-811 Valve RHR Heat Exchanger E-200A NH-36664 C-5 25 RV SA 3
C C
O/C A
RT SYl0 N/A Relief Valve M-1 12 RHR Heat Exchanger E-200B NH-36664 C-4 25 RV SA 3
C C
O/C A
RT SYIO N/A Relief Valve M-1 12 RHRSW-57-2 RH RSW-57-3 RHRSW-57-4 RV-3038 RV-3039 RV-3202 RV-3203 Valve Table Page 81 of87
Monticello Nuclear Generating Plant Inservice Testing Program - Valves Si'stem RHR Service Water Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ Pv-AflflA P.
A..IW I
PIJ-A'A c
M 1-Reidr1uall Meat R*emoval Auxiliary Compressor Relief Valve NH M-3D6247 M-121 A-4 u02 5
'V SA NC C
C O/C A
RT SYl0 NIA Residual Heat Removal NH-36246 A-5 025 RV SA NC C
C O/C A
RT SY1 0 N/A Auxiliary Compressor Relief M-120 Valve RHR Service Water Pump P-NH-36665 A-2 1 0 GT SO 3
B C
0 A
STO Q
N/A 109A Motor Cooling Water M-811 Supply Solenoid Valve FO Q
N/A RHR Service Water Pump P-NH-36665 B-7 1.0 GT SO 3
B C
0 A
STO 0
N/A 109B Motor Cooling Water M-81 1 Supply Solenoid Valve FO Q
N/A RHR Service Water Pump P-NH-36665 B-2 1 0 GT SO 3
B C
0 A
STO Q
N/A I09C Motor Cooling Water M-811 Supply Solenoid Valve FO Q
N/A RHR Service Water Pump P-NH-36665 B-7 1 0 GT SO 3
B C
0 A
STO Q
N/A 109D Motor Cooling Water M-811 Supply Solenoid Valve FO Q
NIA 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 0
N/A Service Water to RH R NH-36665 C-7 10 CK SA 3
C O/C C
A CTC Q
N/A Service Water Check Valve M-811 CTO Q
N/A Valve Table Page 82 of 87 K~V-49UOP RV-4908B SV-4937A SV-4937B SV-4937C SV-4937D SW-21 -1 SW-21-2
Monticello Nuclear Generating Plant Inservice Testing Program - Valves S-ystem RHR Service Water Description Service Water to RHR Service Water Check Valve PID Coord NH-36665 M-811 C-2 Size 10 Type Act CK SA Class Cat Norm Pos Safe Pos A/P Test NC C
O/C C
A CTC CTO Freq RR/DTJ Q
N/A Q
N/A 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 0
N/A Component SW-22-1 SW-22-2 Vahw Table Page 83 of 8 7
Monticello Nuclear Generating Plant Inservice Testing Prograimn-Valves System Service Water Component Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RR/DTJ Service Water to Emergency Service Water Check Valve NH-36664 M-112 D-2 30 CK SA NC C
0 C
A CTC Q
N/A CTO Q
N/A Service Water to Emergency NH-36664 0-1 3.0 CK SA 3
C 0
C A
CTC Q
N/A Service Water Check Valve M-1 12 OTO 0
N/A Service Water to Emergency NH-36664 D-3 30 CK SA NC C
0 C
A CTC Q
N/A Service Water Check Valve M-112 CTO Q
N/A Service Water to Emergency NH-36664 D-3 30 CK SA 3
C 0
C A
CTC Q
N/A Service Water Check Valve M-1 12 CTO 0
N/A SW-101 SW-102 SW-103 SW-104 Valve Table Page 84 of 87
K.-~
"P"J Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Standby Liquid Control Conm onent Description PID Coord Size Type Act Class Cat Norm Pos Safe Pos A/P Test Freq RRIDTJ RV-11-39A Standby Liquid Control Pump NH-36253 C-4 2.0 RV SA 2
C C
0/C A
RT SY10 N/A Discharge Relief Valve M-127 RV-11-39B Standby Uquid Control Pump, NH-36253 B4 2.0 RV SA 2
C C
O/C A
RT SY1O N/A Discharge Relief Valve M-127 XP-11-14A SBLC A Explosive Actuated NH-36253 C-5 1.50 XP SA 2
D C
0 A
DT SY10 N/A Squibb Valve M-1 27 XP-111-148 SBLC B Explosive Actuated NH-36253 D-5 1.50 XP SA 2
D C
0 A
DT SY10 N/A Squibb Valve M-1 27 XP-3-1 SLBC Pump P-302A NH-36253 C-4 1.50 CK SA 2
C 0/C O/C A
DI SY8 DTJ37 Discharge Check Valve M-127 XP-3-2 SLBC Pump P-302B NH-36253 B-4 1.50 CK SA 2
C O/C O/C A
DI SY8 DTJ37 Discharge Check Valve M-127 XP-6 SLBC Outboard Containment NH-36253 D-6 1.50 CK SA 1
AC C
0/C A
LT AJ N/A Isolation Check Valve M-127 CTO RF DTJ07 CTC RF DTJ07 XP-7 SLBC Inboard Containment NH-36253 C-6 1.50 CK SA 1
AC C
O/C A
LT AJ N/A Isolation Check Valve M-1 27 CTO RF DTJ07 CTC RF DTJO7 Valve Table Page $S of 87
Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Standby Liq'uid Control Component Description PID XP-8 Standby Liquid Control NH-36253 Injection Line Manual Isolation M-127 Valve Coor C-6 Size Type Act 150 GT MA Class 1
Cat Norm Pos Safe Pos A/P Test B
0 0
P PIT Freq RR/DTJ Y2 N/A Valve Table Page 86 of 87
k Monticello Nuclear Generating Plant Inservice Testing Program - Valves System Traversing In-Core Probe Component Description PID Coord Size Type Act Class Cal. Norm Pos Safe Pos A/P Test Freq RR/DTJ TIP Inboard Containment Isolation Valve GE0719E520 nla D-5 025 BA SO 2
A O/C C
N/A TIP Shear Valve GE0719E520 D-5 025 XP XP 2
D 0
C A
DT SY10 N/A n/a 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 0
N/A TIP Shear Valve GE0719E520 D-5 0.25 XP XP 2
D 0
C A
DT SY10 N/A n/a 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
NIA TIP Shear Valve GE0719E520 D-5 0.25 XP XP 2
D 0
C A
DT SY10 N/A n/a TIP-1-1 TIP-1-2 TIP-2-1 TIP-2-2 TIP-3-1 TIP-3-2 Valve Table Page 87 of 8;