ML070570335

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AMRM-33, Rev. 0, Aging Management Review of the Reactor Coolant System Pressure Boundary.
ML070570335
Person / Time
Site: Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date: 01/25/2006
From:
Entergy Nuclear Operations
To:
Office of Nuclear Reactor Regulation
O'Hara T, RI/DRS/PSB2, (610) 337-5043
References
AMRM-33, Rev 0
Download: ML070570335 (57)
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VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 2 of 57 REVISION DESCRIPTION SHEET Revision Number Description Pages and/or Sections Revised 0 Initial Issue

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 3 of 57 TABLE OF CONTENTS 1.0 Introduction ....................................................................................................................... 4 1.1 Purpose............................................................................................................................. 4 1.2 System Description ........................................................................................................... 4 1.3 System and Component Intended Functions .................................................................... 9 2.0 Screening ........................................................................................................................ 10 2.1 Component Evaluation Boundaries................................................................................. 10 2.2 Materials.......................................................................................................................... 15 2.3 Environments .................................................................................................................. 19 3.0 Aging Effects Requiring Management............................................................................. 21 3.1 Carbon Steel Components Exposed to Treated Water ................................................... 22 3.2 Carbon Steel Components Exposed to Air-indoor .......................................................... 23 3.3 Stainless Steel Components Exposed to Treated Water ................................................ 23 3.4 Carbon Steel and Stainless Steel Components Exposed to Nitrogen ............................ 24 3.5 Stainless Steel Components Exposed to Air-indoor ....................................................... 25 3.6 Bolting ............................................................................................................................. 25 3.7 Operating Experience ..................................................................................................... 26 4.0 Demonstration That Aging Effects Will Be Managed ...................................................... 27 4.1 Aging Management Programs ........................................................................................ 27 4.2 Time-Limited Aging Analyses.......................................................................................... 30 5.0 Summary and Conclusions ............................................................................................. 31 6.0 References...................................................................................................................... 32 Attachments: - Components Subject to Aging Management Review.........................................36 - Aging Management Review Results...................................................................50

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 4 of 57 1.0 Introduction 1.1 Purpose This report is part of the aging management review (AMR) of the integrated plant assessment (IPA) performed to extend the operating license of Vermont Yankee Nuclear Power Station (VYNPS). This report demonstrates the effects of aging on reactor coolant system (RCS) passive components will be adequately managed so that the intended functions will be maintained consistent with the current licensing basis (CLB) as required by 10 CFR 54.21(a)(3).

For additional information on the license renewal project and associated documentation, refer to the License Renewal Project Plan (Ref. 6.3.1).

The purpose of this report is to demonstrate that aging effects for passive mechanical components will be adequately managed for the period of extended operation associated with license renewal. The approach for demonstrating management of aging effects is to first identify the components that are subject to aging management review in Section 2.0. The next step is to define the aging effects requiring management for the system components in Section 3.0. Section 4.0 then evaluates if existing programs and commitments adequately manage those effects.

Applicable aging effects were determined using EPRI report 1003056, Non-Class 1 Mechanical Implementation Guideline and Mechanical Tools (Ref. 6.2.2); herein after referred to as the Mechanical Tools. This EPRI report provides the bases for identification of aging effects based on specific materials and environments and documents confirmation of the validity of the aging effects through review of industry experience. The Mechanical Tools were not written to specifically address environments and materials in Class 1 systems; however, the Mechanical Tools are applicable where the materials and environments are the same as the non-Class 1 materials and environments. The reactor coolant system pressure boundary subcomponents covered in this AMRR include materials and environments evaluated in the Mechanical Tools.

Other industry references, including NUREG-1801, Generic Aging Lessons Learned (GALL)

Report, were used to address material and environment combinations not addressed in the Mechanical Tools.

This aging management review report (AMRR), in conjunction with EPRI report 1003056, documents the identification and evaluation of aging effects requiring management for mechanical components in the reactor coolant system pressure boundary.

1.2 System Description The reactor coolant system (RCS) is described in Chapter 4 of the Updated Final Safety Analysis Report (UFSAR) (Ref. 6.1.1). The RCS includes the reactor vessel, a limited number of components of the reactor vessel internals, supporting systems (e.g. control rod drive and reactor recirculation), and those attached systems and components that form portions of the nuclear system process barrier. These systems and components contain or transport the fluids coming from or going to the reactor. Certain portions of the RCS are described in other UFSAR sections as referenced below.

VYNPS has no isolation condenser (Ref. 6.1.11).

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 5 of 57 The following systems, in whole or in part, compose the reactor coolant system pressure boundary. These do not necessarily correspond to the EMPAC system codes (see Section 1.3).

1. Control rod drive CRD
2. Core spray system CS
3. Feedwater FW
4. High pressure coolant injection system HPCI
5. Main steam MS
6. Nuclear boiler NB
7. Nuclear boiler vessel instrumentation NBVI
8. Reactor core isolation cooling system RCIC
9. Reactor recirculation RR
10. Reactor water cleanup system RWCU
11. Residual heat removal system RHR
12. Standby liquid control system SLC 1.2.1 Control Rod Drive System (CRD)

As described in UFSAR Section 3.4, the CRD hydraulic system operates the CRD mechanisms using water from the condensate system as hydraulic fluid. Each drive has an associated hydraulic control unit (HCU) that controls the flow to and from a drive. The water discharged from the drives during a scram flows through the HCUs to the scram discharge volume. The water discharged from a drive, during a normal control rod positioning operation, flows through its HCU and the exhaust header to the RWCU system discharge line. An adequate volume of pressurized water is maintained in each scram accumulator to ensure rod insertion following a reactor trip.

The system is normally in operation, pressurized with treated water and controlling rod position.

The CRD mechanisms (drives) are located under the reactor vessel, inside the primary containment, while the hydraulic system and scram headers are located in the reactor building outside primary containment.

For additional description of the system and its components, see the CRD system design basis document (Ref. 6.1.2.1).

1.2.2 Core Spray System (CS)

As described in Section 6 of the UFSAR, the core spray system delivers cooling water spray to the core for accident mitigation. The class 1 section of the core spray system includes piping, valves, and instrumentation from the primary containment penetration exterior isolation valves into the reactor vessel.

1.2.3 Feedwater System (FW) (Class 1 components)

As described in Section 4.1 of the UFSAR, feedwater lines provide water to the reactor vessel entering near the top of the vessel downcomer annulus. This function is part of normal operation. Two feedwater lines divide and enter the vessel through four nozzles.

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 6 of 57 For additional description of the system and its components, see the condensate and reactor feedwater system design basis document (Ref. 6.1.2.3).

1.2.4 High Pressure Coolant Injection System (HPCI)

As described in Section 6.4.1 of the UFSAR, HPCI assures that the reactor core is adequately cooled in the event of a small break in the nuclear system and loss of coolant which does not result in rapid depressurization of the reactor vessel. The HPCI system turbine is driven by steam from the reactor which is generated by decay heat and residual heat. The steam is extracted from a main steam header upstream of the main steam line isolation valves. The two HPCI system isolation valves in the steam line to the HPCI system turbine are normally open to keep the piping to the turbine at elevated temperatures to permit rapid startup of the HPCI system. A HPCI system steam supply valve, just upstream of the turbine stop valve, automatically opens on receipt of a HPCI system initiation signal.

For additional description of the system and its components, see the high pressure coolant injection system design basis document (Ref. 6.1.2.4).

1.2.5 Main Steam System (MS) (Class 1 Components)

As described in Chapter 4 of the UFSAR, four steam lines are utilized between the reactor and the turbine. Two main steam line isolation valves are installed on each main steam line. One valve in each line is inside the primary containment, the other outside primary containment.

These valves automatically close off the nuclear system process barrier in the event a pipe break occurs downstream of the valves. This action limits the loss of coolant and the release of radioactive materials from the nuclear system. In the event that a main steam line break occurs inside primary containment, closure of the isolation valve outside the containment acts to seal primary containment.

Three safety valves (SVs) and four safety/relief valves (SRVs), all of which are located on the main steam lines within the drywell between the reactor vessel and the first isolation valve, are designed to protect the nuclear system process barrier from damage due to overpressure. The pressure operated safety valves are provided to discharge steam from the nuclear system to the primary containment. The SRVs actuate when inlet pressure exceeds the popping pressure (pilot stage) set point. They can also be actuated by energizing solenoids which allow the pneumatic actuator to open the valve. Each SRV has its own discharge line which terminates below the suppression pool minimum water level. These valves automatically depressurize the nuclear system in the event of a loss of coolant accident in which the high pressure coolant injection (HPCI) system fails to deliver rated flow or where break flow exceeds HPCI capacity (intermediate break). The depressurization of the nuclear system allows low pressure standby cooling systems to supply enough cooling water to adequately cool the fuel. The safety and safety/relief valves are distributed among the main steam lines (one SRV on each steam line and one SV on the A, C and D steam lines) so that a single accident cannot completely disable a safety, relief, or automatic depressurization function.

Each steam line contains a venturi-type flow restrictor, installed near the reactor vessel, but downstream of the pressure relief and safety valves. The restrictors are designed to limit the loss of coolant resulting from a main steam line break outside primary containment. The coolant loss is limited so that reactor vessel water level remains above the top of the core during the

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 7 of 57 time required for the main steam line isolation valves to close. This action protects the fuel barrier. These flow restrictors also provide the delta-pressure for main steam flow indication.

The main steam system is normally in operation. For additional description of the system and its components, see the MS system design basis document. (Ref. 6.1.2.5) 1.2.6 Nuclear Boiler (NB)

As discussed in Chapter 4.0 of the UFSAR, the Reactor Coolant System (nuclear boiler system) includes almost all the class 1 components covered by this AMR. However, for clarity this AMR has divided the components into functional systems similar to the UFSAR discussion. Only a few miscellaneous components not included in any of the functional systems remain to be discussed in the NB system. These components include the reactor vessel vent line off nozzle N7, the reactor vessel drain line off nozzle N15, and the flange leak detection lines off nozzles N13 and N14, with their associated valves and instrumentation.

These components are normally not in service. The vent and drain lines are used only during outages, and the leak detection lines contain water only if an RV flange o-ring is leaking, or during refueling operations. There is no design basis document for this system.

For additional information on the nuclear boiler system, see UFSAR Section 4, Reactor Coolant System; there is no design basis document for the nuclear boiler system.

1.2.7 Nuclear Boiler Vessel Instrumentation (NBVI)

As discussed in Section 7.8 of the UFSAR, the reactor vessel instrumentation monitors reactor vessel parameter information.

The nuclear boiler vessel instrumentation system is normally in operation. For additional description of the system and its components, see the NBVI system design basis document (Ref 6.1.2.6).

1.2.8 Reactor Core Isolation Cooling (RCIC)

As described in Section 4.7 of the UFSAR, the reactor core isolation cooling system provides makeup water to the reactor vessel during shutdown and isolation to supplement or replace the normal makeup sources and operates automatically in time to obviate any requirement for the core standby cooling systems. The RCIC system turbine is driven by steam from the reactor.

The steam is extracted from a main steam header upstream of the main steam line isolation valves.

For additional description of the system and its components, see the RCIC system design basis document (Ref 6.1.2.7).

1.2.9 Reactor Recirculation (RR)

As described in Section 4.3 of the UFSAR, the reactor recirculation system pumps coolant through the core. Adjustment of the core coolant flow rate changes reactor power output, thus providing a means of following plant load demand without adjusting control rods. The recirculation system is designed with sufficient fluid and pump inertia that fuel thermal limits

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 8 of 57 cannot be exceeded as a result of recirculation system malfunctions. The arrangement of the recirculation system is designed so that a piping failure cannot compromise the integrity of the floodable inner volume of the reactor vessel.

The reactor recirculation pumps are Byron-Jackson pumps. The portions of the reactor recirculation pumps that perform a pressure boundary function are the pump casing, the cover assembly, the driver mount, the seals, and the pressure retaining closure bolting. Each reactor recirculation pump consists of a cast austenitic stainless steel (CASS) casing with a CASS cover assembly. The cover includes a thermal barrier subassembly with integral heat exchanger. The driver mount is bolted to the casing, clamping the cover assembly in position.

(VYEM 0132, Ref. 6.1.6)

The RR system flow is completely within primary containment; however, there are instrumentation lines that penetrate containment, with tubing, valves, and transmitters outside the primary containment. The RR system is normally in operation.

For additional information on the reactor recirculation system, see UFSAR Section 4.3, Reactor Recirculation System; there is no design basis document for the RR system.

1.2.10 Reactor Water Cleanup System (RWCU)

As described in Section 4.9 of the UFSAR, the reactor water cleanup system functions to maintain the required purity of reactor coolant by circulating coolant through a system of filters and demineralizers. As such, the system is normally in operation. There is no design basis document for the RWCU system.

For additional information on the reactor recirculation system, see UFSAR Section 4.9, Reactor Water Cleanup System; there is no design basis document for the RWCU system.

1.2.11 Residual Heat Removal System (RHR)

As discussed in Section 4.8 of the UFSAR, the residual heat removal system removes residual and decay heat from the reactor core under a variety of situations. The class 1 portion of the system includes piping and valves from the connections (one suction and two discharge) to the recirculation system to the primary containment penetration exterior isolation valves. Low pressure coolant injection (LPCI) is an operating mode of the residual heat removal system (see UFSAR Chapter 6.)

For additional description of the system and its components, see the RHR system design basis document (Ref 6.1.2.8).

1.2.12 Standby Liquid Control System (SLC)

As discussed in Section 3.8 of the UFSAR, the standby liquid control system pumps a boron neutron absorber solution into the reactor vessel during accident conditions. The class 1 portion of the system includes piping, valves, and instrumentation from the primary containment exterior isolation valve (check valve) into the reactor vessel.

For additional description of the system and its components, see the SLC system design basis document (Ref 6.1.2.9).

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 9 of 57 1.3 System and Component Intended Functions The license renewal intended functions of each system that forms part of the RCS pressure boundary are given in the VYNPS scoping report LRPD-01 (Ref. 6.3.3) Refer to the scoping report for detailed function information.

For license renewal, the primary intended function of the reactor coolant pressure boundary components is to maintain system pressure boundary. Many of the systems attached to the nuclear boiler penetrate the primary containment, and as such they also have the function of maintaining containment integrity. The main steam flow restrictors have the additional function of restricting flow.

The reactor coolant system pressure boundary is credited with maintaining its pressure boundary in response to regulated events (Fire, Anticipated Transient without Scram, and Station Blackout) such that the reactor core can be shutdown, cooled, and isolated following the event. These functions are included in the functions of maintaining pressure boundary and maintaining containment integrity. No additional reactor coolant system pressure boundary components are subject to aging management review based on being needed for response to regulated events.

The control rod drive system (CRD and HCU) has the function of shutting down the reactor core both in normal and emergency conditions, including in response to regulated events (Fire, Anticipated Transient without Scram, and Station Blackout). This function is included in the CRD system function of maintaining pressure boundary of the system, including the pressure boundary of the scram accumulators and scram discharge headers. No additional control rod drive components are subject to aging management review based on being needed for response to regulated events.

The SRVs in the NB system are relied on in safety analyses to perform a function that demonstrates compliance with the Commission's regulations for station blackout (10CFR50.63).

As this AMR includes class 1 piping that forms the reactor coolant system pressure boundary, it does not include any non-safety related components. Consequently, there are no non-safety related components whose failure could affect safety related components [10 CFR 54.4(a)(2)].

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 10 of 57 2.0 Screening 2.1 Component Evaluation Boundaries The major components of the reactor coolant pressure boundary include the reactor vessel (nuclear boiler), reactor vessel internals (incore dry tubes and local power range monitors only),

and portions of various systems connected to the reactor vessel. The reactor vessel internals are reviewed in AMRM-32, Aging Management Review of the Reactor Vessel Internals. The reactor vessel, including vessel nozzles and safe ends, are reviewed in AMRM-31, Aging Management Review of the Reactor Pressure Vessel.

All class 1 piping attached to the vessel nozzles or safe ends, including the welded joints, class 1 pumps, and class 1 boundary isolation valves are included in this AMR. In addition, this AMR reviews connected class 2 piping that is not part of another AMR as far as needed to complete the RCS pressure boundary; this includes vents, drains, leakoff, sample lines, and instrumentation lines up to the transmitters. Class 1 and class 2 as used herein are consistent with the definitions of safety classifications (i.e., SC-1 and SC-2) for VYNPS mechanical components as specified in Section 5.1 of Reference 6.1.3. The evaluation boundaries of this AMR extend to any or all of the following.

1 This AMR extends to the outboard containment isolation valve on system piping which penetrates primary reactor containment, consistent with the class 1 boundary.

2 For piping which does not penetrate the containment, this AMR extends to the first normally closed isolation valve. For instrumentation that does not have a normally closed isolation valve, this AMR extends to the instrument housing.

3 This AMR includes the reactor coolant system safety and safety/relief valves up to the valve seat, i.e. normally pressurized components.

4 This AMR includes instrumentation root valves and associated instrumentation lines up to the instruments.

5 Most of the systems reviewed here include containment penetrations. In this AMR, only the piping and connecting welds to the containment penetrations are reviewed. The rest of the penetration is reviewed in AMRC-01, Primary Containment.

A listing of reactor coolant pressure boundary passive mechanical components subject to aging management review and associated materials of construction are included as Attachment 1.

Applicable P&IDs associated with the RCS pressure boundary and highlighted to reflect those portions of each system reviewed in this AMRR, are listed in the subsections below.

2.1.1 Control Rod Drive System (CRD)

The control rod drive mechanisms (i.e. the outer shell pressure boundary of the drive mechanisms) form part of the RCS pressure boundary and as such are subject to aging management review. The CRD mechanisms are reviewed in this AMR.

Portions of the (89) HCUs form part of the RCS pressure boundary, and those portions are subject to aging management review. The scram accumulator tanks (water and nitrogen) are

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 11 of 57 subject to aging management review with the intended function of maintaining their pressure boundary so there will be adequate pressure to scram the reactor. The Hydraulic Control Units have their own system code (HCU) in EMPAC.

The scram discharge headers and associated piping are also subject to aging management review as part of the post-trip reactor coolant pressure boundary. The scram valves (CV-126 and CV-127) are air operated valves which open on loss of air, resulting in a control rod scram.

Loss of air system pressure boundary integrity will not prevent a scram. Therefore, the scram valve operators, their solenoid valves and associated air tubing do not require aging management review.

The water supply portion of the CRD hydraulic system, including the pumps, headers, valves, and water supply lines, are used only for routine (non-scram) rod movement and are not part of the RCS pressure boundary. They are not required for the system to perform its license renewal intended function. Therefore, aging management review is not required for these components (Ref. 6.2.1).

The control rod drive mechanisms are located inside the primary containment while the insert/withdraw piping to those mechanisms penetrate the containment. The rest of the CRD system, including the scram discharge headers and the hydraulic control units are located outside the primary containment in the reactor building.

The components of the CRD system subject to aging management review are highlighted on drawing LRA-G-191170-0 and are listed in Attachment 1 for the CRD system.

The CRD housings, stub tubes, and the mechanism to housing bolts are reviewed in AMRM-31, Aging Management Review of the Reactor Pressure Vessel. The CRD guide tubes are reviewed in AMRM-32, Aging Management Review of the Reactor Vessel Internals. There is no separate AMR for the CRD system.

2.1.2 Core Spray System (CS)

The class 1 portion of the core spray system attached to the reactor vessel forms part of the reactor coolant pressure boundary and provides containment integrity. The class 1 portion of the core spray system is subject to aging management review and is reviewed in this AMR. The components of the core spray system reviewed in this AMR include valves and piping from the outer containment isolation valve to the vessel safe ends, including instrumentation piping, valves, and orifices. The components of the CS system subject to aging management review are highlighted on drawing LRA-G-191168-0 and are listed in Attachment 1 for the CS system.

The remainder of the core spray system is also subject to aging management review, and is reviewed in several AMRS. The piping, headers and spargers inside the reactor vessel are reviewed in AMRM-32, Aging Management Review of the Reactor Vessel Internals. The vessel nozzles, safe ends, and thermal sleeves are reviewed in AMRM-31, Aging Management Review of the Reactor Pressure Vessel. The class 2 section of the core spray system is reviewed in AMRM-03, Aging Management Review of the Core Spray System.

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 12 of 57 2.1.3 Feedwater (FW)

The class 1 portion of the feedwater system attached to the reactor vessel forms part of the reactor coolant pressure boundary and provides containment integrity. The class 1 portion of the feedwater system is subject to aging management review and is reviewed in this AMR. The components of the feedwater system reviewed in this AMR include piping and valves from the primary containment isolation valves (including valves V-27A and V-96A) to the vessel. These components are subject to aging management review and are reviewed in this AMR.

The components of the FW system subject to aging management review are highlighted on drawing LRA-G-191167-0 and are listed in Attachment 1 for the FW system.

The feedwater headers inside the reactor vessel are reviewed in AMRM-32, Aging Management of the Reactor Vessel Internals. The nozzles, safe ends, and thermal sleeves are reviewed in AMRM-31, Aging Management Review of the Reactor Pressure Vessel. The class 2 feedwater components needed to deliver HPCI are reviewed in AMRM-05, Aging Management Review for the High Pressure Coolant Injection System. The class 2 feedwater components needed to deliver RCIC are reviewed in AMRM-06, Aging Management Review for the Reactor Core Isolation Cooling. The remainder of the feedwater system is not subject to aging management review and there is no FW system AMR.

2.1.4 High Pressure Coolant Injection (HPCI)

The class 1 portions of the HPCI system attached to main steam line D forms part of the reactor coolant pressure boundary and provides containment integrity. The class 1 portion of the HPCI system is subject to aging management review and is reviewed in this AMR. The components of the HPCI system reviewed in this AMR include The class 1 section of the HPCI system includes the steam supply from main steam line D to the outboard containment isolation valve (V23-16).

The components of the HPCI system subject to aging management review are highlighted on drawing LRA-G-191167-0 and are listed in Attachment 1 for the main steam system.

The non-class 1 portion of the HPCI system, including the section of feedwater piping containing the HPCI return, up to the class 1 boundary (see drawing LRA-G-191167-0), is reviewed in AMRM-05, Aging Management Review of the High Pressure Coolant Injection System.

2.1.5 Main Steam (MS)

The class 1 portion of the main steam system attached to the reactor vessel forms part of the reactor coolant pressure boundary and provides containment integrity. The main steam flow restrictors have the function of restricting flow. The class 1 portion of the main steam system is subject to aging management review and is reviewed in this AMR. The components of the main steam system reviewed in this AMR include piping from the vessel to the MSIVs, the main steam flow restrictors, the safety/relief valves (SRVs), safety valves (SVs) and the MSIVs. The supply lines to HPCI and RCIC, main steam drain lines, and instrumentation lines are reviewed in this AMR. This AMR also reviews components from the Automatic Depressurization System (ADS), which includes the safety valves (SVs) and safety/relief valves (SRVs) along with the connecting piping that completes the RCS pressure boundary.

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 13 of 57 The components of the MS system subject to aging management review are highlighted on drawings LRA-G-191167-0, LRA-G-191169-0 and LRA-G-191174-0 and are listed in for the MS system.

Sections 3.3.1 and 3.4.1 of the design basis document for the main steam system (Ref. 6.1.2.5) state that the safety valves and safety relief valves are removed and replaced with reconditioned valves during each refueling outage. As such, these valves are not long-lived components and are not subject to aging management review.

The reactor vessel nozzles and safe ends are reviewed in AMRM-31, Aging Management Review of the Reactor Pressure Vessel. The non-class 1 portion of main steam (EMPAC system code MS) is reviewed in AMRM-26, Main Condenser and MSIV Leakage Pathway. The non-class 1 portion of the HPCI steam supply is reviewed in AMRM-05, Aging Management Review for the High Pressure Coolant Injection System. The non-class 1 portion of the RCIC steam supply is reviewed in AMRM-06, Aging Management Review for the Reactor Core Isolation Cooling. The rest of the ADS, including the SRV discharge piping, is reviewed in AMRM-04, Aging Management Review for the Automatic Depressurization System.

2.1.6 Nuclear Boiler (NB)

The entire nuclear boiler system forms part of the reactor coolant pressure boundary. The nuclear boiler system is subject to aging management review and is reviewed in this AMR. The components of the NB system not included in other sections of this AMR (see section 1.2.1 description of NB) and subject to aging management review include the reactor vessel vent line off nozzle N7, the reactor vessel drain line off nozzle N15, and the flange leak detection lines off nozzles N13 and N14. The bolting for the second flange off nozzle N7 is included in this AMR.

The components of the NB system subject to aging management review are highlighted on drawings LRA-G-191167-0 and LRA-G-191267 and are listed in Attachment 1 for the NB system.

The associated vessel nozzles and safe ends, and the vessel flange bolts for nozzle N7, are reviewed in AMRM-31, Aging Management Review of the Reactor Pressure Vessel. There is no other AMR for the NB system.

2.1.7 Nuclear Boiler Vessel Instrumentation (NBVI)

The entire NBVI system forms part of the reactor coolant pressure boundary and provides containment integrity. The NBVI system is subject to aging management review and is reviewed in this AMR. The components of the NBVI reviewed in this AMR include instrument piping, restricting orifices, condensing chambers, isolation valves, and flow limiting valves. The flange bolts for the second flange off nozzle N6B are included in this AMR. The components of the NBVI system subject to aging management review are listed in Attachment 1 for the NBVI system.

The components of the NBVI system subject to aging management review are highlighted on drawings LRA-G-191267-0, sheets 1 and 2, LRA-G-191168, and LRA-G-191170-0. The components are also listed in Attachment 1 for the NBVI system.

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 14 of 57 The reactor vessel instrument nozzles and the first flange bolts for nozzle N6B are reviewed in AMRM-31, Aging Management Review of the Reactor Pressure Vessel. There is no separate AMR for the NBVI system.

2.1.8 Reactor Core Isolation Cooling (RCIC)

The class 1 portion of the reactor core isolation cooling system, attached to main steam line C, forms part of the reactor coolant pressure boundary and provides containment integrity. The class 1 portion of the reactor core isolation cooling system is subject to aging management review and is reviewed in this AMR. The components of the reactor core isolation cooling system reviewed in this AMR include valves and piping from steam line C to the outboard containment isolation valve. The components of the RCIC system subject to aging management review are highlighted on drawing LRA-G-191174-0 and are listed in Attachment 1 for the main steam system.

The RCIC system, and the section of feedwater piping up to the class 1 boundary (see drawing LRA-G-191174-0), is reviewed in AMRM-06, Aging Management Review of the Reactor Core Isolation Cooling System. This includes the section of feedwater piping containing the RCIC return as highlighted on drawing LRA-G-191167-0, 2.1.9 Reactor Recirculation (RR)

The entire reactor recirculation system forms part of the reactor coolant system pressure boundary. The entire system is subject to aging management review and is reviewed in this AMR. The components of the RR system subject to aging management review include piping, valves, pumps, reactor vessel nozzles, safe ends, and thermal sleeves, vents, drains, instrumentation lines, and instrumentation valves. Portions of the reactor recirculation pumps reviewed in this AMRR include the pump casing, the cover assembly (including the thermal barrier), the driver mount, and the pressure retaining closure bolting. The pump seals are not subject to aging management review as they are periodically monitored and replaced as needed. (Section 4.3.4 of the UFSAR, and Ref. 6.1.6).

The components of the RR system subject to aging management review are highlighted on drawing LRA-G-191167-0 and are listed in Attachment 1 for the RR system.

The vessel nozzles, safe ends, and thermal sleeves are reviewed in AMRM-31, Aging Management Review of the Reactor Pressure Vessel. The connecting piping inside the vessel, including the jet pumps, is reviewed in AMRM-32, Aging Management Review of the Reactor Vessel Internals. There is no separate AMR for the RR system.

2.1.10 Reactor Water Cleanup System (RWCU)

The class 1 portion of the reactor water cleanup system forms part of the reactor coolant pressure boundary and provides containment integrity. The class 1 portion of the RWCU system is subject to aging management review and is reviewed in this AMR. The components of the reactor water cleanup system reviewed in this AMR include valves and piping in the suction line from the residual heat removal suction line and the reactor vessel drain line to the class 1 outboard containment isolation valve for the combined line (V12-18).

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 15 of 57 The components of the RWCU system subject to aging management review and included in this AMR are highlighted on drawings LRA-G-191167-0 and LRA-G-191178-0 and are listed in for the RWCU system.

A portion of the RWCU system is shared by RCIC and is evaluated in AMRM-06, Aging Management Review of the Reactor Core Isolation Cooling System. The remainder of the RWCU system is not subject to aging management review (unless portions meet the requirements of 10CFR54.4(a)(2)), and there is no separate AMR for the RWCU system.

2.1.11 Residual Heat Removal System (RHR)

The class 1 portions of the residual heat removal system form part of the reactor coolant system pressure boundary and provide containment integrity. The class 1 portions of the residual heat removal system are subject to aging management review, and are reviewed in this AMR. The components of the residual heat removal system reviewed in this AMR include suction piping and valves from the reactor recirculation suction line to the outboard containment isolation valve and the two discharge lines from their outboard containment isolation valves to the reactor recirculation return lines. Also included are vent, drain and test connections; there is no instrumentation associated with the class 1 portions of the RHR system.

The components of the RHR system subject to aging management review in this AMR are highlighted on drawing LRA-G-191172-0 and are listed in Attachment 1 for the RHR system.

The class 2 section of the RHR system is reviewed in AMRM-02, Aging Management Review of the Residual Heat Removal System.

2.1.12 Standby Liquid Control System (SLC)

The class 1 portion of the standby liquid control system forms part of the reactor coolant system pressure boundary and provides containment integrity. The class 1 portion of the SLC system is subject to aging management review and is reviewed in this AMR. The components of the standby liquid control system reviewed in this AMR include valves and piping from the outer containment isolation valve (check valve 16) to the vessel nozzle safe end, including one test connection.

The components of the SLC system subject to aging management review in this AMR are highlighted on drawing LRA-G-191171-0 and are listed in Attachment 1 for the SLC system.

The piping and sparger inside the reactor vessel are reviewed in AMRM-32, Aging Management Review of the Reactor Vessel Internals. The vessel nozzle, safe end, and thermal sleeve are reviewed in AMRM-31, Aging Management Review of the Reactor Pressure Vessel. The class 2 section of the SLC system is reviewed in AMRM-01, Aging Management Review of the Standby Liquid Control System.

2.2 Materials This section lists the material of construction for those components identified in Section 2.1 as subject to aging management review.

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 16 of 57 Piping in the scope of this AMR is designed in accordance with USAS B31.1.0, Power Piping, or appropriate parts of ASME Sections I, III and VIII. Valves in the scope of this AMR are designed in accordance with USAS B31.1, USAS B16.5 and ASME Sections I, III, and VIII. The recirculation pumps were designed in accordance with ASME Section III, Class C. See Table 4.1.1 of the UFSAR for additional details of the applicable codes for RCS components.

Information regarding the materials of construction of piping, valves, and other components were obtained using the pipe class specified on the LRA drawing, and the definition of that pipe class in Appendix A to the Piping Specification (Ref. 6.1.10). Orifices and condensing chambers are considered to be fittings for the purposes of determining material of construction from the appropriate piping specification.

2.2.1 Control Rod Drive System (CRD)

The control rod drive mechanism housings are stainless steel. The insert/withdraw tubing and the hydraulic control units are class SS-5 per LRA-G191170. Per the piping specification, class SS-5 is ASTM A376 or A312, Grade TP304 or TP316 stainless steel.

The scram accumulators were originally carbon steel. They are gradually being replaced by stainless steel units. At present they are a mix of carbon steel and stainless steel accumulators.

(Ref. 6.1.23)

The scram discharge header is class CS-4 per LRA-G191170. Per the piping specification class CS-4 is ASTM A106 Grade B carbon steel.

2.2.2 Core Spray System (CS)

The class 1 core spray piping from the reactor vessel to isolation valves V-14-14A/B is SA312 Type 316L stainless steel per Note 6 to LRA-G191168. The core spray piping from isolation valves V-14-14A/B back to isolation valves V-14-12A/B is mostly piping is class SS-6, ASTM A376 or A312, Grade TP304 or TP316. However, the final 6 inches is piping class CS-5, ASTM A106 Grade B carbon steel, as shown on drawing LRA-G-191168. Valves V-14-12A/B are also carbon steel.

The instrumentation lines connected to the class 1 core spray line are class SS-6. Per the piping specification, class SS-6 is ASTM A376 or A312, Grade TP304 or TP316.

2.2.3 Feedwater System (FW)

The class 1 feedwater piping is pipe class CS-5 per LRA-G191167. CS-5 is ASTM A106 Grade B carbon steel.

There are small test connections that are also carbon steel. There is no instrumentation on the class 1 section of the feedwater piping.

2.2.4 High Pressure Coolant Injection System (HPCI)

The class 1 steam lines to the HPCI system are pipe class CS-5 per LRA-G191169. CS-5 is ASTM A106 Grade B carbon steel.

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 17 of 57 The instrument lines off the HPCI steam line are pipe class SS-6. SS-6 is ASTM A376 or A312, Grade TP304 or TP316 stainless steel.

2.2.5 Main Steam System (MS) (Class 1 Components)

The class 1 main steam piping is provided by General Electric and is not covered by the piping specification. Main steam piping connected to nozzle N3 is ASTM A106 Grade B carbon steel (Ref. 6.1.26).

The instrument lines off the main steam lines are pipe class SS-6. SS-6 is ASTM A376 or A312, Grade TP304 or TP316 stainless steel. Condensing chambers and restricting orifices are stainless steel based on their location between the stainless steel flow element and the stainless steel instrument piping.

The class 1 steam drains are pipe class CS-5. CS-5 is ASTM A106 Grade B carbon steel.

Section 3.2 of the Design Basis Document for the Main Steam System (Ref. 6.1.2.5) states that the main steam isolation valves (V2-80A, B, C, & D and V2-86A, B, C, & D) are 18x16x18 cast carbon steel, air operated valves.

Section 4.5.3 of the UFSAR states that the main steam flow restrictor assembly consists of a venturi-type nozzle insert welded into a carbon steel pipe. The venturi-type nozzle insert is constructed utilizing all austenitic stainless steel and is held in place with a full circumferential fillet weld. This is also stated in Section 3.5 of the main steam system design basis document (Ref. 6.1.2.5)

As discussed in Section 3.1.5, the main steam safety valves and safety relief valves are not subject to aging management review.

Pressure containing parts of the relief valve body are fabricated of ASTM A216, Grade WCB stainless steel. The instrumentation off the relief valves (up to valve V2-90) is assumed to be stainless steel as it is also a pressure retaining part.

2.2.6 Nuclear Boiler (NB)

The vessel vent line off nozzle N7 does not have a pipe class on LRA-G-191167. It connects with the carbon steel main steam line on drawing LRA-G-191167 and with stainless steel instrumentation on drawing LRA-G-191267. This line was assumed to be stainless steel.

The flange leak off lines off nozzles N13 and N14 are stainless steel.

The vessel drain line off nozzle N15 is a mixture of stainless steel and carbon steel as shown on LRA-G-191167. The stainless steel is pipe class SS-6, ASTM A376 or A312, Grade TP304 or TP316. Per drawing 919D294 in Ref. 6.1.9 the carbon steel is SA-106 Grade B, which corresponds to CS-5.

2.2.7 Nuclear Boiler Vessel Instrumentation (NBVI)

Per drawing LRA-G-192167 the nuclear boiler instrumentation piping is pipe class SS-6. SS-6 is ASTM A376 or A312, Grade TP304 or TP316 stainless steel.

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 18 of 57 2.2.8 Reactor Core Isolation Cooling (RCIC)

The class 1 steam lines to the RCIC system are pipe class CS-5 per LRA-G191174. CS-5 is ASTM A106 Grade B carbon steel.

The instrument lines off the RCIC steam line are pipe class SS-6. SS-6 is ASTM A376 or A312, Grade TP304 or TP316 stainless steel.

2.2.9 Reactor Recirculation (RR)

The original reactor recirculation piping was replaced per EDCR 85-01. All the new recirculation piping and fittings are now Type 316 stainless steel (ASME SA-376, SA-182, or SA-403) per the piping specification included as Enclosure C to EDCR 85-01 (Ref. 6.1.25).

The instrumentation associated with the reactor recirculation system is pipe class SS-6, ASTM A376 or A312, Grade TP304 or TP316 stainless steel.

The reactor recirculation pump casings and covers are ASTM A351 Grade CF8M cast austenitic stainless steel. The driver mount is ASTM A216 Grade WCB carbon steel. The cover to case bolts are ASTM A193 Grade 7. (Ref. 6.1.6) 2.2.10 Reactor Water Cleanup System (RWCU)

The 4 inch RWCU return to the RHR system, including the 2 inch connection to the reactor vessel drain, is pipe class SS-6. SS-6 is ASTM A376 or A312, Grade TP304 or TP316 stainless steel.

2.2.11 Residual Heat Removal System (RHR)

Pipe class information is from drawing LRA-G-191172. The 20 inch RHR suction line from the reactor recirculation system is pipe class SS-6 from the RR system to the second isolation valve. From the second isolation valve to the class 1 boundary this 20 inch line is pipe class CS-5. The 24 inch return line is pipe class CS-5 from the class 1 boundary to the check valve inside primary containment, and then class SS-6 from the check valve to the RR system.

Smaller connecting pipes are also CS-5 or SS-6 matching the piping to which they connect.

CS-5 is ASTM A106 Grade B carbon steel and SS-6 is ASTM A376 or A312, Grade TP304 or TP316 stainless steel.

There is no instrumentation in the class 1 portion of the RHR system.

2.2.12 Standby Liquid Control System (SLC)

The class 1 portion of the SLC 1.5 inch return line to the vessel is pipe class SS-6 per LRA-G191171. The 3/4 inch test connection is also SS-6. SS-6 is ASTM A376 or A312, Grade TP304 or TP316 stainless steel.

There is no instrumentation off the class 1 portion of SLC.

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 19 of 57 2.3 Environments The bulk of the systems reviewed in this AMR are located inside primary containment. Most of the systems reviewed have instrumentation tubing and instrument valves outside primary containment but inside the reactor building. In addition, most of the systems reviewed include containment isolation valves outside primary containment, and the short run of piping from the containment penetration to those valves. The scram accumulators have an internal environment of nitrogen with an external environment of air-indoor.

The operating environments experienced by the reactor coolant system pressure boundary components are treated water or nitrogen on internal surfaces and air-indoor (i.e., containment environment) on external surfaces.

2.3.1 Treated Water The components reviewed in this report have the internal environment of treated water. The reactor coolant system water varies in temperature from less than 212 degrees in small, no flow areas to greater than 500 degrees in the vessel interior. There are four environments based on temperature for treated water.

Treated water. This implies cold (<212 °F) treated water. At this low temperature, moisture may be present on the outside surface of the material.

Treated water greater than 220 ºF. Above this threshold, carbon steel is susceptible to fatigue (Appendix H of Ref. 6.2.2).

Treated water greater than 270 ºF. Above this threshold, stainless steel is susceptible to fatigue (Appendix H of Ref. 6.2.2).

Treated water greater than 482 ºF. Above this threshold, cast austenitic stainless steel (CASS) is susceptible to reduction of fracture toughness due to thermal embrittlement (Section 3.3.1 of Ref. 6.2.2).

For purposes of this report, steam is considered treated water. VYNPS water chemistry requirements are specified in the Updated Final Safety Analysis Report (UFSAR). Treated reactor water is described in Section 4.3 of the EPRI BWR Water Chemistry guidelines (BWRVIP-79) for normal water chemistry. Refer to Section 4.1.6 for more information regarding the VYNPS Water Chemistry Program.

Most of the components are normally in service. Systems such as reactor recirculation and main steam contain treated water greater than 482 ºF; other systems such as feedwater contain water greater than 270 ºF. Components in subsystems that are normally in standby may be less than 212 ºF; however, this is a function of the distance from the operating portion of the system and the size of the line. Systems normally below 212 ºF include the standby liquid control system, CRD hydraulic system, NBVI system, and PASS. Instrumentation tubing and valves, especially after condensing chambers or containment penetrations, are normally less that 212 ºF.

2.3.2 Nitrogen The scram accumulators have nitrogen overpressure. Each hydraulic control unit has one accumulator that is entirely nitrogen on the inside and one that is approximately half nitrogen

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 20 of 57 and half water. These accumulators are in the reactor building and are normally less than 212

°F.

2.3.3 Air-indoor (External)

The bulk of the systems reviewed in this AMR are located inside the primary containment where ambient temperature averages 150°F (Section 5.2.3.7 of Ref. 6.1.1). The primary containment is inerted with nitrogen to maintain oxygen concentration less than 4% (Section 5.2.6.2 of the UFSAR). Some portions of the systems (instrumentation and primary containment isolation valves, and the tubing or piping from the containment penetration to these components) are outside primary containment but inside the reactor building. Reactor building ambient air temperature is normally controlled between 55 and 100°F (Section 3.8.3 of Ref. 6.1.1) and is not inerted. For the purposes of this AMRR, primary containment ambient air is conservatively considered equivalent to reactor building ambient air and both are referred to as air-indoor.

External surfaces of some reactor coolant pressure boundary components normally exceed 212

°F and thus do not have moisture present on these surfaces; other subcomponents such as those normally in standby, are less than 212 °F and may have moisture on their external surfaces.

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 21 of 57 3.0 Aging Effects Requiring Management EPRI report 1003056 and other industry documents are used in this section to identify and evaluate aging effects. The following aging effects and associated mechanisms were identified for the material / environment combinations present in the reactor coolant system.

loss of material general corrosion, galvanic corrosion, erosion, flow accelerated corrosion, crevice corrosion, selective leaching and pitting corrosion, cracking fatigue, flaw growth, stress corrosion cracking (SCC) and intergranular attack (IGA),

reduction of fracture toughness thermal and radiation embrittlement and loss of preload various mechanisms (for bolting)

For additional information on aging effects, refer to EPRI report 1003056 (Ref. 6.2.2).

Several aging mechanisms can be eliminated based on the material/environment combinations in the reactor coolant system pressure boundary. These mechanisms are discussed here, and not addressed under each material/environment combination.

Erosion is not applicable to the reactor coolant system due to the purity of the water within the system.

Selective leaching is not applicable to the reactor coolant system since the susceptible materials (zinc-copper alloys, aluminum alloys, gray cast iron) are not present.

Radiation embrittlement is not applicable for RCS components within the scope of this AMRR as there are no components within the beltline region (see LRPD-03 for a definition of the beltline region for license renewal).

Loss of pre-load for bolting, in agreement with the Mechanical Tools, is a design driven effect. Loss of pre-load leads to gasketed closure leakage but does not defeat the function of the joint to maintain the pressure boundary. Consequently loss of preload is not an aging effect requiring management.

Cracking due to flaw growth is managed by the inspection requirements for Class 1 components in accordance with ASME Section XI, Subsection IWB. Because inservice inspection per ASME Section XI is required in accordance with 10 CFR 50.55a, cracking due to flaw growth is not identified on the tables in Attachment 1.

The following sections document the determination of aging effects requiring management based on specific component materials and environments. The review was performed for groups of components with similar operating environments and materials of construction. The AMR results are tabulated in Attachment 2.

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 22 of 57 3.1 Carbon Steel Components Exposed to Treated Water The systems comprising the RCS which have carbon steel components exposed to treated water include control rod drive, core spray, feedwater, main steam (including HPCI and RCIC steam supplies), nuclear boiler, and residual heat removal. See Attachment 1 for a list of the carbon steel components. These components are exposed to treated water on internal surfaces and air on external surfaces. The treated water ranges from low temperature water in the CRD hydraulics, NB vessel drains and RHR systems through high temperature water in the feedwater to steam in main steam and NB vessel vents.

3.1.1 Loss of Material Carbon steel surfaces exposed to treated water are susceptible to loss of material due to general corrosion. These surfaces are also susceptible to pitting and crevice corrosion in the presence of high oxygen levels and concentrated impurities. Carbon steel in contact with stainless steel in the presence of an electrolyte is also susceptible to galvanic corrosion.

Loss of material from flow-accelerated corrosion is possible at high velocity locations. Only the carbon steel feedwater lines and main steam lines see the continuous velocities necessary for flow accelerated corrosion. Therefore, loss of material due to flow-accelerated corrosion is an aging effect requiring management for the feedwater and main steam systems.

3.1.2 Cracking - fatigue Cracking due to thermal fatigue is an aging effect for carbon steel above 220°F. Many of the RCS subsystems operate above this temperature. Portions of idle systems will exceed this temperature due to their proximity to hotter systems. The ASME Design Code ASME Section III, Subsection NB requires the calculation of cumulative usage factors (CUF) and the usage factors must be less than one for the period of extended operation. Fatigue analysis (usage factor assessment) is a time-limited aging analysis (TLAA). For more information on TLAA, see Section 4.5 and VYNPS Report LRPD-04, TLAA - Mechanical Fatigue.

3.1.3 Cracking - non-fatigue Service loads may result in the growth of pre-service flaws or initiation and growth of service-induced flaws. (Ref. 6.2.4) The most susceptible locations for flaw initiation and growth are welded joints. Susceptibility is due to the variations in residual stresses and mechanical properties resulting from the various constituent zones (e.g., composite, unmixed, and heat-affected) within the joint. Therefore, cracking (initiation and growth) is considered an aging effect that requires management for the period of extended operation. However, because inservice inspection per ASME Section XI is required in accordance with 10 CFR 50.55a, cracking due to flaw growth is not identified on the tables in Attachment 1.

Stress corrosion cracking and intergranular attack are not significant aging mechanisms for carbon steel exposed to treated water.

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 23 of 57 3.1.4 Reduction of Fracture Toughness Reduction of fracture toughness due to thermal embrittlement is not an aging effect requiring management for carbon steel components, since carbon steel is not susceptible to thermal aging.

3.2 Carbon Steel Components Exposed to Air-indoor The systems comprising the RCS which have carbon steel components exposed externally to air-indoor include control rod drive, core spray, feedwater, main steam (including HPCI and RCIC steam supplies), nuclear boiler, and residual heat removal. The reactor recirculation pump driver mounts are carbon steel components exposed only to air-indoor (no internal environment). See Attachment 1 for a list of the carbon steel components.

External ferritic steel surfaces exposed to air-indoor are susceptible to loss of material due to general corrosion only if the exterior of the component is exposed to moisture (Appendix E of Ref. 6.2.2). Most of the RCS system operating temperatures remain above 212 °F, and, as such, those portions of the system do not have moisture on the external surface. Carbon steel components routinely below 212 °F includes the scram accumulators, scram discharge headers, portions of the RHR system, and the RR pump driver mounts. Loss of material (external) is an aging effect requiring management for those portions of the RCS where normal service temperatures remain below 212 °F.

3.3 Stainless Steel Components Exposed to Treated Water The systems comprising the RCS pressure boundary which contain stainless steel components exposed to treated water include control rod drive, core spray, main steam, nuclear boiler, nuclear boiler vessel instrumentation, reactor recirculation, residual heat removal, reactor water clean up, and standby liquid control. Selected RCS valves (bodies and bonnets) and the recirculation pump casing and cover are constructed of cast austenitic stainless steel (CASS).

The recirculation pump cover is also CASS and is exposed to treated water both as reactor coolant and as closed cycle cooling water in the thermal barrier subassembly. See Attachment 1 for a list of the stainless steel components. The treated water includes low temperature water in the CRD hydraulics, core spray, NB vessel drains, RHR, and SLC; high temperature water in the NBVI; and steam in the MS system.

3.3.1 Loss of Material Stainless steel is inherently resistant to general corrosion and erosion. However, stainless steel internal surfaces are susceptible to loss of material due to pitting and crevice corrosion in the presence of high oxygen levels and contaminants. Therefore, loss of material is an aging effect requiring management from internal surfaces of stainless steel components.

3.3.2 Cracking - fatigue Cracking due to thermal fatigue is an aging effect for stainless steel above 270°F. Many of the RCS subsystems operate above this temperature. Portions of idle systems will exceed this temperature due to their proximity to hotter systems. The ASME Design Code ASME Section III, Subsection NB requires the calculation of cumulative usage factors (CUF) and the usage factors must be less than one for the period of extended operation. Fatigue analysis (usage

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 24 of 57 factor assessment) is a time-limited aging analysis (TLAA). For more information on TLAA, see Section 4.2. Cracking due to fatigue is discussed in VYNPS Report LRPD-04, TLAA -

Mechanical Fatigue.

3.3.3 Cracking - non-fatigue Service loads may result in the growth of pre-service flaws or initiation and growth of service-induced flaws. (Ref. 6.2.4) The most susceptible locations for flaw initiation and growth are the welded joints. Susceptibility is due to the variations in residual stresses and mechanical properties resulting from the various constituent zones (e.g., composite, unmixed, and heat-affected) within the joint. Therefore, cracking (initiation and growth) is considered an aging effect that requires management for the period of extended operation. However, because inservice inspection per ASME Section XI is required in accordance with 10 CFR 50.55a, cracking due to flaw growth is not identified on the tables in Attachment 1.

Cracking due to stress corrosion and intergranular attack is an aging effect requiring management for stainless steel exposed to treated water.

The reactor vessel flange leak off lines are periodically wetted (during floodup for refueling) and dried, resulting in a concentration of impurities and an increased susceptibility to stress corrosion cracking. These lines will be managed by a component-specific one time inspection in addition to water chemistry control. (See Section 4.1.4) 3.3.4 Reduction of Fracture Toughness Reduction of fracture toughness due to thermal embrittlement is an aging effect requiring management for cast austenitic stainless steel components operating above 482 °F. (Appendix A, section 3.3.1 of Ref. 6.2.2). The reactor recirculation pump casings and main steam flow restrictors are cast austenitic components operating in this temperature range and hence are subject to reduction of fracture toughness due to thermal embrittlement.

3.4 Carbon Steel and Stainless Steel Components Exposed to Nitrogen The only components in the RCS exposed to nitrogen are the scram accumulators in the CRD system and the piping and valves used to recharge the nitrogen accumulators. This is an ambient temperature system located in the reactor building. Some of the scram accumulators are carbon steel and some are stainless steel.

Industry experience shows that commercial grade nitrogen is provided as a high quality product with little if any external contaminants (Section 2.2.2 of Appendix D to Ref. 6.2.2). Appendix D to the Mechanical Tools (Ref. 6.2.2) gives no aging effects for carbon steel or stainless steel exposed to nitrogen.

The internal surfaces are not susceptible to general corrosion or to pitting/crevice corrosion.

Therefore, loss of material from internal surfaces of the scram accumulators is not an aging effect requiring management.

Cracking due to stress corrosion and intergranular attack is not an aging effect requiring management since the system temperature remains below the 140°F threshold for these mechanisms in stainless steel.

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 25 of 57 Cracking due to thermal fatigue is not an aging effect requiring management since the system temperature remains below the 220/270°F threshold for carbon/stainless steel thermal fatigue.

Reduction of fracture toughness due to thermal embrittlement is likewise not an aging effect requiring management as these accumulators are in the reactor building and not exposed to significant radiation. (Section 3.3.1 of Ref. 6.2.2)

In summary, there are no aging effects requiring management for the nitrogen filled scram accumulators and connecting piping.

3.5 Stainless Steel Components Exposed to Air-indoor The systems comprising the RCS pressure boundary which contain stainless steel components exposed externally to air-indoor include control rod drive, core spray, main steam, nuclear boiler, nuclear boiler vessel instrumentation, reactor recirculation, residual heat removal, reactor water clean up, and standby liquid control. See Attachment 1 for a list of the stainless steel components.

Insulation on class 1 piping is free of contaminants that could cause cracking of stainless steel in air-indoor. (Sections 11.2 and 12.2 of Ref. 6.1.21)

Stainless steel components exposed to air-indoor are not susceptible to any aging effects requiring management.

3.6 Bolting The reactor coolant system pressure boundary includes the following bolting:

NB the bolting for the second flange off nozzle N7.

NBVI the bolting for the second flange off nozzle N6B.

RR the reactor recirculation pump casing studs.

Valves the body to bonnet bolts for numerous valves are in this AMR.

All the bolting for blank flanges on flush connections, cross connects, test connections, etc. are included in this AMR.

Pressure retaining bolting in these systems is low alloy steel or stainless steel and is exposed to air.

3.6.1 Cracking - Fatigue All of the identified bolting is susceptible to cracking by fatigue. Therefore, cracking due to fatigue is an aging effect requiring management for pressure boundary bolting. Fatigue analysis is a time-limited aging analysis (TLAA). For more information on TLAA, see Section 4.2.

3.6.2 Cracking - Stress Corrosion Cracking All of the identified bolting is susceptible to cracking by stress corrosion cracking (SCC) as identified in Appendix F of the Mechanical Tools (Ref. 6.2.2).

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 26 of 57 3.6.3 Loss of Material External ferritic steel surfaces exposed to dry indoor ambient conditions are susceptible to loss of material due to general corrosion only if the exterior of the component is less than 212 °F (Appendix E of Ref. 6.2.2). The reactor coolant system operating temperature is well over 212

°F, but some of the connecting systems may not be over 212 °F. Consequently loss of material due to general corrosion is an aging effect requiring management for low alloy steel RCS bolting in systems less than 212 °F.

3.6.4 Reduction of Fracture Toughness Reduction of fracture toughness is only applicable to materials in the reactor vessel beltline region. As there is no bolting in the beltline region, reduction of fracture toughness is not an aging effect requiring management for closure bolting.

3.7 Operating Experience The review of site-specific operating experience and recent industry operating experience, documented in VYNPS Report LRPD-05, Operating Experience Review Results (Ref. 6.3.7) did not identify any aging effects not addressed in this aging management review report.

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 27 of 57 4.0 Demonstration That Aging Effects Will Be Managed Section 2.0 described the components in the RCS and associated auxiliary systems that are subject to aging management review. For those components, Section 3.0 documented the determination of aging effects requiring management. The aging management review is completed by demonstrating that existing programs, when continued into the period of extended operation, can manage the aging effects identified in Section 3.0. No further action is required for license renewal when the evaluation of an existing program demonstrates that it is adequate to manage the aging effect such that corrective action may be taken prior to loss of the system intended functions. Alternately, if existing programs cannot be shown to manage the aging effects for the period of extended operation, then action(s) will be proposed to augment existing program(s) or create new programs to manage the identified effects of aging.

Demonstration for the purposes of this license renewal technical evaluation is accomplished by establishing a clear relationship among:

1. the components under review,
2. the aging effects on these items caused by the material-environment-stress combinations which, if undetected, could result in the loss of the intended function such that the system could not perform its function(s) within the scope of license renewal in the period of extended operation, and
3. the credited aging management programs (AMP) whose actions serve to preserve the system intended function(s) for the period of extended operation. lists each of the component groups subject to aging management review and identifies the aging effects requiring management for the material and environment combination.

The following programs in combination will manage the effects of aging thereby precluding the loss of the intended functions of the system. Section 4.1 provides the clear relationship between the component, the aging effect and the aging management program actions which preserve the intended functions for the period of extended operation.

For a comprehensive review of the programs credited for license renewal of VYNPS and a demonstration of how these programs will manage the aging effects, see VYNPS report LRPD-02, Aging Management Program Evaluation Results.

Potential time-limited aging analyses (TLAA) that have been identified for the reactor coolant system are described in Section 4.2.

4.1 Aging Management Programs 4.1.1 Boiling Water Reactor Stress Corrosion Cracking Program The BWR Stress Corrosion Cracking (SCC) Program consists of inspection and flaw evaluation to monitor the effects of SCC of BWR piping of 4 inch or greater nominal diameter. Monitoring and controlling reactor coolant water chemistry in accordance with the Water Chemistry Control

- BWR Program. For additional information on the BWR stress corrosion cracking program, see VYNPS Report LRPD-02, Aging Management Program Evaluation Results.

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 28 of 57 The BWR stress corrosion cracking program is credited with managing cracking of stainless steel piping, pumps, valves, and fittings of greater than 4 inches nominal diameter.

4.1.2 Flow-Accelerated Corrosion Program The Flow-Accelerated Corrosion (FAC) Program is controlled by VYNPS procedure PP 7028 (Ref. 6.1.17). VYNPS commitments to GL 89-01 manage FAC, and these commitments will be carried forward through the period of extended operation. This program includes (a) an analysis to determine critical locations, (b) initial inspections to determine the extent of thinning at these locations, and (c) follow-up inspections to confirm predictions, or repair or replace components as necessary. For additional information on the Flow Accelerated Corrosion Program, see VYNPS report LRPD-02, Aging Management Program Evaluation Results.

The Flow-Accelerated Corrosion Program is credited with managing flow accelerated corrosion of selected RCS piping and valves fabricated from carbon steel.

4.1.3 Inservice Inspection Program The VYNPS Inservice Inspection (ISI) Program is a nondestructive examination program that provides for the implementation of ASME Code,Section XI, Subsections IWB, IWC, and IWD (1998 Edition, 2000 Addenda) in accordance with the provisions of 10 CFR 50.55a. (Ref.

6.1.16). For additional information on the inservice inspection program, see VYNPS Report LRPD-02, Aging Management Program Evaluation Results.

The ISI Program is credited with managing cracking and loss of material for carbon steel, stainless steel, and cast austenitic stainless steel components within the RCS. It is credited with managing reduction of fracture toughness for cast austenitic stainless steel (reactor recirculation pump casings). It is also credited with managing cracking and loss of material for bolting.

4.1.4 One Time Inspection Program The One Time Inspection Program includes measures to verify effectiveness of an aging management program (AMP) and confirm the absence of an aging effect. For example, this program verifies effectiveness of the Water Chemistry Control - BWR program by confirming that unacceptable degradation is not occurring inside components exposed to treated water.

The One Time Inspection program will inspect a sample of the carbon steel scram accumulators, if any carbon steel accumulators remain for the period of extended operation (See section 2.2.1).

The One Time Inspection program will inspect the main steam flow restrictors for loss of material and cracking, including crack growth that may occur due to reduction of fracture toughness.

The One Time Inspection program will inspect the reactor vessel flange leak off lines for cracking because their increased susceptibility to this aging effect.

The One Time Inspection Program will inspect small bore class 1 piping (< 4 inch NPS),

including pipe, valves, fittings, thermocouples, condensing chambers, and branch connections,

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 29 of 57 in the reactor coolant system. These components do not receive volumetric inspection in accordance with ASME Section XI, Examination Category B-J or B-F. The One Time Inspection Program will use combinations of NDE, including visual, ultrasonic, and surface techniques, performed by qualified personnel following procedures consistent with the ASME Code and 10 CFR 50, Appendix B., for representative samples of the small bore piping population. One time inspection is listed for managing reduction of fracture toughness for small valves made of CASS in that the one time inspection program will inspect for crack growth due to reduction of fracture toughness. This is a conservative approach, as NUREG-1801,Section XI.M12 states that Inservice Inspection according to ASME Section XI is adequate to manage ROFT of small bore valves based on an NRC performed bounding integrity analysis.

This program will be initiated prior to the end of the initial operating period for VYNPS. The one time inspection will occur at or near the end of the initial operating period. For additional information on the One Time Inspection Program, see VYNPS report LRPD-02, Aging Management Program Evaluation Results.

The One Time Inspection Program is credited with managing cracking and loss of material of carbon steel and stainless steel pipe, valves, thermowells, restrictors and fittings.

4.1.5 System Walkdown Program Under the System Walkdown Program, walkdowns are conducted to manage aging effects on components (Ref. 6.1.19). For additional information on the Systems Walkdown Program, see VYNPS report LRPD-02, Aging Management Program Evaluation Results.

The System Walkdown Program is credited with managing loss of material for external surfaces of carbon steel components by requiring periodic visual inspections. It is also credited with managing loss of material for pressure boundary bolting.

4.1.6 Water Chemistry Control - BWR Program The VYNPS Water Chemistry Control - BWR program optimizes the reactor water chemistry to minimize the potential for loss of material and cracking by limiting the levels of contaminants in the RCS. Additionally, hydrogen water chemistry (HWC) is incorporated as part of the Water Chemistry Control - BWR Program. HWC limits the potential for intergranular SCC by reducing the concentration of the dissolved oxygen in the treated water. For additional information on the Water Chemistry Control - BWR program, see VYNPS Report LRPD-02, Aging Management Program Evaluation Results.

The Water Chemistry Control - BWR Program will manage the following aging effects.

  • Loss of material due to general corrosion (carbon steel)
  • Loss of material due to galvanic corrosion (carbon steel)
  • Loss of material due to crevice corrosion and pitting corrosion (all materials)
  • Cracking by SCC and intergranular attack (IGA) (stainless steel)

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 30 of 57 4.1.7 Water Chemistry Control - Closed Cooling Water Program The VYNPS Water Chemistry Control - Closed Cooling Water Program optimizes closed cooling water chemistry to minimize the potential for loss of material and cracking. For additional information on the water chemistry control - closed cooling water program, see VYNPS Report LRPD-02, Aging Management Program Evaluation Results.

The Water Chemistry Control - Closed Cooling Water Program is credited with managing cracking and loss of material of the reactor recirculation pump thermal barrier.

4.2 Time-Limited Aging Analyses The following potential time-limited aging analyses (TLAA) were identified during preparation of this AMR as applicable to the reactor coolant system components.

  • Corrosion allowances are provided for all exposed surfaces of carbon and low alloy steels. The evaluation of wall thinning due to corrosion is a potential TLAA that must be evaluated for the period of extended operation.

For additional information, refer to VYNPS report LRPD-03, TLAA and Exemption Evaluation Results for the evaluation of these potential TLAA and to VYNPS report LRPD-04, TLAA -

Mechanical Fatigue for the evaluation of metal fatigue for the period of extended operation.

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 31 of 57 5.0 Summary and Conclusions The following aging management programs address the aging effects requiring management for the reactor coolant system.

  • One-Time Inspection Program (One-time inspection)
  • System Walkdown Program (System walkdowns)
  • Water Chemistry Control - BWR Program (Water chemistry control - BWR)
  • Water Chemistry Control - Closed Cooling Water Program (Water chemistry control - CCW)

The parenthetical expressions above are used in Attachment 2 to identify these programs. For additional review of the programs credited for license renewal of Vermont Yankee Nuclear Power Station, see VYNPS report LRPD-02, Aging Management Program Evaluation Results. contains the aging management review results for the reactor coolant system components included in this AMR.

In conclusion, the programs described in Section 4.0 will provide reasonable assurance that the effects of aging on the VYNPS reactor coolant system will be managed such that the intended functions will be maintained consistent with the current licensing basis throughout the period of extended operation.

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 32 of 57 6.0 References 6.1 VYNPS Documents 6.1.1. VYNPS Updated Final Safety Analysis Report (UFSAR), Revision 18.

6.1.2. VYNPS Design Basis Documents 6.1.2.1. Document CRD, Design Basis Document for Control Rod Drive System, Revision 0, 9/25/98 (GE-NE-C1100337-01) 6.1.2.2. Document CS, Design Basis Document for Core Spray System, Revision 0, 12/30/98 (GE-NE-E2100123-01) 6.1.2.3. Document CFW, Design Basis Document for Condensate and Reactor Feedwater System, Revision 0, 8/14/98 6.1.2.4. Document HPCI, Design Basis Document for High Pressure Coolant Injection System, Revision 0, 10/30/98 (GE-NE-E41-00113-01) 6.1.2.5. Document MS, Design Basis Document for the Main Steam System (MS),

Revision 0, 1/22/9) (GE-NE-B2200067-01) 6.1.2.6. Document NBVI, Design Basis Document for Nuclear Boiler Vessel Instrumentation System, Revision 0, 1/18/99, (GE-NE-B2200068-01) 6.1.2.7. Document RCIC, Design Basis Document for Reactor Core Isolation Cooling System, Revision 0, 12/31/98, (GE-NE-E5100191-01) 6.1.2.8. Document RHR, Design Basis Document for Residual Heat Removal System, Revision 1, 1/20/99 6.1.2.9. Document SLC, Design Basis Document for Standby Liquid Control System, Revision 0, 1/27/99 (GE-NE-C4100169-01) 6.1.3. ENN-MS-S-009-VY, Vermont Yankee Site Specific Guidance and System Safety Function Sheets, Revision 0, 03/22/2005 6.1.4. Document SADBD, Topical Design Basis Document for Safety Analysis, Revision 2 6.1.5. VYNPS Appendix R Safe Shutdown Capability Analysis (SSCA), Revision 6, change number 3, 03/26/2001 6.1.6. VYEM 0132, Reactor Recirculation Pumps - Installation and Operation Instructions, BW/IP manual 8020/1F5936. 09/26/89 6.1.7. VYNPS Flow Diagrams 6.1.7.1. G191155, Piping and Instrument Symbols

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 33 of 57 6.1.7.2. G191167, Nuclear Boiler 6.1.7.3. G191168, Core Spray System 6.1.7.4. G191169, High Pressure Coolant Injection System 6.1.7.5. G191170, Control Rod Drive Hydraulic System 6.1.7.6. G191171, Standby Liquid Control System 6.1.7.7. G191172, Residual Heat Removal System 6.1.7.8. G191174, Reactor Core Isolation Cooling System 6.1.7.9. G191178, Reactor Water Cleanup System 6.1.7.10. G191267, Nuclear Boiler Vessel Instrumentation 6.1.8. 3, Drawing B-191261, Instrument Installation Details 6.1.9. GEK-9608, Operation and Maintenance Instruction, Reactor Assembly for Vermont Yankee Nuclear Power Station, 12/70 6.1.10. Piping Specification, VYNP-V1-III-P-1; Ebasco Specification 62-65T, General Power Piping, November 13, 1967 6.1.11. GE Service Information Letter (SIL-185), Isolation Condenser Tubing Inspection, August 31, 1976 6.1.12. Licensed Operator Training Program Student Handout, LOT-00-299H, Revision 0, September, 2001 6.1.13. VYNPS Procedure OP4612, Sampling and Treatment of the Reactor Water System, Revision 23, September, 2000 6.1.14. VYNPS Procedure OP 2199, Hydrogen Water Chemistry System, Revision 1, September 2003 6.1.15. VYNPS Procedure RP 4626, Sampling and Treatment of the Condensate and Feedwater Systems, Revision 14, March 2000 6.1.16. VYNPS Procedure PP 7015, Vermont Yankee Inservice Inspection Program, Revision September, 2003 6.1.17. VYNPS Procedure PP 7028, Piping Flow Accelerated Corrosion Inspection Program, Revision 0, May 2001 6.1.18. VYNPS Procedure PP 7006, Primary Containment Leakage Rate Testing Program, Revision 7, November 2003

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 34 of 57 6.1.19. VYNPS Procedure PP 7014, System Engineering Program, Revision 1, January 2004 and System Engineering Department Implementation Guide #2, Performing System Walkdowns/Inspections, Revision 4, December 2000 6.1.20. VYNPS Procedure OP 0212, General Bolting Requirements, Revision 2, February 2003 6.1.21. VYNP-III-I-1, Ebasco Specification - Insulation, Revision 2, 12/23/69 6.1.22. GE Specification 21A1058, 11/12/1968, Steam Flow Element 6.1.23. Email, Paul Rainey to R. Finnin, J. Hoffman, A. Cardine; LR Open Item 125 on CRD HCUs, 20 Jan 2005 6.1.24. VYNPS Procedure OP 4101, Revision 26, RPV Operational System Leakage Test, 02/24/2004 6.1.25. EDCR 85-01, Recirculation Piping Replacement, 1985; and Enclosure C:

VYNPS Specification VY-EDCR-85-01, Specification for Procurement of Nuclear Grade Stainless Steel Piping for Replacement of Vermont Yankee Nuclear Power Station Recirculation System 6.1.26. VYNPS Drawing 5920-00570, Primary Steam Piping - Nuc Boiler 2 3, Revision 5, 1991 (GE drawing 729E792 Sheet 2) 6.2. Industry Documents 6.2.1. NUREG-1801, Generic Aging Lessons Learned (GALL) Report, April 2001 6.2.2. EPRI report 1003056, Revision 3, Non-Class 1 Mechanical Implementation Guideline and Mechanical Tools, November 2001 (a.k.a. the Mechanical Tools) 6.2.3. EPRI report TR-103515-R2, BWR Water Chemistry Guidelines2000 Revision, Final Report, February 2000 6.2.4. Nondestructive Examination Standards, Technical Basis and Development of Boiler and Pressure Vessel Code, ASME Section XI, Division 1, EPRI-NP-1406-SR, May 1980 6.2.5. License Renewal Issue 98-0012, Consumables, April 12, 1999 6.3. License Renewal Documents 6.3.1. LRPG-01, License Renewal Project Plan 6.3.2. LRPG-04, Mechanical System Screening and Aging Management Reviews 6.3.3. LRPD-01, System and Structure Scoping Results 6.3.4. LRPD-02, Aging Management Program Evaluation Results

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Revision 0 Pressure Boundary Page 35 of 57 6.3.5. LRPD-03, TLAA and Exemption Evaluation Results 6.3.6. LRPD-04, TLAA - Mechanical Fatigue 6.3.7. LRPD-05, Operating Experience Review Results 6.3.8. AMRM-01, Aging Management Review of the Standby Liquid Control System 6.3.9. AMRM-02, Aging Management Review of the Residual Heat Removal System 6.3.10. AMRM-03, Aging Management Review of the Core Spray System 6.3.11. AMRM-04, Aging Management Review of the Automatic Depressurization System 6.3.12. AMRM-05, Aging Management Review of the High Pressure Coolant Injection System 6.3.13. AMRM-06, Aging Management Review of the Reactor Core Isolation Cooling System 6.3.14. AMRM-31, Aging Management Review of the Reactor Pressure Vessel 6.3.15. AMRM-32, Aging Management Review of the Reactor Vessel Internals 6.3.16. AMRC-01, Aging Management Review of the Primary Containment

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant System Pressure Revision 0 Boundary Page 36 of 57 Attachment 1 - Components Subject to Aging Management Review Comp ID Component Type Component Name Material Control Rod Drive System Components (Components marked -XX-YY have 89 components 02-19 thru 42-27)

CRD-XX-YY DRIVE CRD MECHANISMs Stainless Steel CRD CS FLANGES FLANGE CRD cleanouts (2" & 6") Carbon Steel 3/4" CRD CS PIPING, 2" or less 3/4" CRD hydraulics Carbon Steel 1" CRD CS PIPING, 2" or less 1" CRD hydraulics Carbon Steel 1" (80, CS 1, 1.7) PIPING, 2" or less 1" CRD hydraulics Carbon Steel 1" (160, CS 4, 1.1) PIPING, 2" or less 1" CRD hydraulics Carbon Steel 2" CRD CS PIPING, 2" or less 2" CRD hydraulics Carbon Steel 2" (160, CS 4, 1.1) PIPING, 2" or less 1" CRD hydraulics Carbon Steel 1/2" CRD Piping PIPING, 2" or less 1/2" SS CRD PIPING Stainless Steel 3/4" (80,SS 5, 1.1) PIPING, 2" or less 3/4" SS CRD PIPING Stainless Steel 1" (80, SS 5, 1.1) PIPING, 2" or less 1" SS CRD PIPING Stainless Steel 10" CRD-4 PIPING, 6" or greater 10" CRD hydraulics Carbon Steel 10" CRD-4A PIPING, 6" or greater 10" CRD hydraulics Carbon Steel 6" (80, CS 4, 1.1) PIPING, 6" or greater 6" CRD hydraulics Carbon Steel 6" CRD-5 PIPING, 6" or greater 6" CRD hydraulics Carbon Steel 6" CRD-5A PIPING, 6" or greater 6" CRD hydraulics Carbon Steel 6" CRD-6 PIPING, 6" or greater 6" CRD hydraulics Carbon Steel 6" CRD-6A PIPING, 6" or greater 6" CRD hydraulics Carbon Steel 6" CRD-7 PIPING, 6" or greater 6" CRD hydraulics Carbon Steel 6" CRD-8 PIPING, 6" or greater 6" CRD hydraulics Carbon Steel PCV-3-32A VALVE BALL VALVE Carbon Steel PCV-3-32B VALVE BALL VALVE Carbon Steel LCV-3-33A VALVE BALL VALVE Carbon Steel LCV-3-33B VALVE BALL VALVE Carbon Steel LCV-3-33C VALVE BALL VALVE Carbon Steel LCV-3-33D VALVE BALL VALVE Carbon Steel V3-118A VALVE ISOLATION VALVE Carbon Steel V3-118B VALVE ISOLATION VALVE Carbon Steel V3-118C VALVE ISOLATION VALVE Carbon Steel V3-118D VALVE ISOLATION VALVE Carbon Steel V3-118E VALVE ISOLATION VALVE Carbon Steel V3-118F VALVE ISOLATION VALVE Carbon Steel V3-118G VALVE ISOLATION VALVE Carbon Steel V3-118H VALVE ISOLATION VALVE Carbon Steel V3-119A VALVE ISOLATION VALVE Carbon Steel V3-119B VALVE ISOLATION VALVE Carbon Steel V3-119C VALVE ISOLATION VALVE Carbon Steel V3-119D VALVE ISOLATION VALVE Carbon Steel V3-119E VALVE ISOLATION VALVE Carbon Steel V3-119F VALVE ISOLATION VALVE Carbon Steel V3-119G VALVE ISOLATION VALVE Carbon Steel V3-119H VALVE ISOLATION VALVE Carbon Steel

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant Revision 0 System Pressure Boundary Page 37 of 57 Attachment 1 - Components Subject to Aging Management Review Comp ID Component Type Component Name Material V3-120A VALVE Inst vent/drain on North Scram disch hdr Carbon Steel V3-120B VALVE Inst vent/drain on North Scram disch hdr Carbon Steel V3-120C VALVE Inst vent/drain on North Scram disch hdr Carbon Steel V3-120D VALVE Inst vent/drain on North Scram disch hdr Carbon Steel V3-120E VALVE Inst vent/drain on South Scram disch hdr Carbon Steel V3-120F VALVE Inst vent/drain on South Scram disch hdr Carbon Steel V3-120G VALVE Inst vent/drain on South Scram disch hdr Carbon Steel V3-120H VALVE Inst vent/drain on South Scram disch hdr Carbon Steel V3-154A VALVE GLOBE VALVE Carbon Steel V3-154B VALVE GLOBE VALVE Carbon Steel V3-155A VALVE GLOBE VALVE Carbon Steel V3-155B VALVE GLOBE VALVE Carbon Steel V3-162A VALVE SDV VENT CHECK VALVE Carbon Steel V3-162B VALVE SDV VENT CHECK VALVE Carbon Steel V3-165A VALVE CRD ISOLATION VALVE Carbon Steel V3-165B VALVE CRD ISOLATION VALVE Carbon Steel V3-166A VALVE CRD ISOLATION VALVE Carbon Steel V3-167A VALVE CRD ISOLATION VALVE Carbon Steel V3-169A VALVE CRD ISOLATION VALVE Carbon Steel V3-170A VALVE CRD ISOLATION VALVE Carbon Steel V3-101-XX-YY VALVE GLOBE VALVE Stainless Steel V3-102-XX-YY VALVE GLOBE VALVE Stainless Steel V3-107-XX-YY VALVE GLOBE VALVE Stainless Steel V3-111-XX-YY VALVE GLOBE VALVE Stainless Steel V3-112-XX-YY VALVE GLOBE VALVE Stainless Steel V3-114-XX-YY VALVE CHECK VALVE Stainless Steel V3-115-XX-YY VALVE CHECK VALVE Stainless Steel SO-3-120-XX-YY VALVE GLOBE VALVE Stainless Steel SO-3-121-XX-YY VALVE GLOBE VALVE Stainless Steel SO-3-122-XX-YY VALVE GLOBE VALVE Stainless Steel SO-3-123-XX-YY VALVE GLOBE VALVE Stainless Steel TK-3-125-XX-YY TANK CRD scram accumulator, water/nitrogen Stainless Steel TK-3-125-XX-YY TANK CRD scram accumulator, water/nitrogen Carbon Steel CV-3-126-XX-YY VALVE AIR OPERATED VALVE Stainless Steel CV-3-127-XX-YY VALVE AIR OPERATED VALVE Stainless Steel TK-3-128-XX-YY TANK CRD scram accumulator, nitrogen Stainless Steel TK-3-128-XX-YY TANK CRD scram accumulator, nitrogen Carbon Steel V3-130-XX-YY VALVE GLOBE VALVE Stainless Steel V3-132-XX-YY VALVE GLOBE VALVE Stainless Steel S-3-132-XX-YY RUPTURE DISK Scram Accumulator Rupture Disk Stainless Steel FE-3-135-XX-YY FILTER CRD HCU filter element Stainless Steel FE-3-136-XX-YY FILTER CRD HCU filter element Stainless Steel V3-137-XX-YY VALVE CHECK VALVE Stainless Steel V3-138-XX-YY VALVE CHECK VALVE Stainless Steel

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant Revision 0 System Pressure Boundary Page 38 of 57 Attachment 1 - Components Subject to Aging Management Review Comp ID Component Type Component Name Material Core Spray System Components RO-14-31A ORIFICE Instrument Orifice Stainless Steel RO-14-31B ORIFICE Instrument Orifice Stainless Steel 1" CS-12A PIPING, 1" or less 1" Instrument Line Stainless Steel 1" CS-12B PIPING, 1" or less 1" Instrument Line Stainless Steel 3/4" PIPING, 1" or less 3/4" vent lines near V-14A/B & V-12A/B Stainless Steel 8" CS-4A PIPING, 8" Core Spray V14-12A to vessel N5B Stainless Steel 8" CS-4A PIPING, 8" Core Spray 6 off V14-12A Carbon Steel 8" CS-4B PIPING, 8" Core Spray V14-12B to vessel N5A Stainless Steel 8" CS-4B PIPING, 8" Core Spray 6 off V23-12B Carbon Steel SL-14-31A VALVE NB FLOW LIMITING VALVE Stainless Steel SL-14-31B VALVE NB FLOW LIMITING VALVE Stainless Steel V14-12A VALVE MOTOR OPERATED VALVE Carbon Steel V14-12B VALVE MOTOR OPERATED VALVE Carbon Steel V14-13A VALVE CHECK VALVE Stainless Steel V14-13B VALVE CHECK VALVE Stainless Steel V14-14A VALVE GLOBE VALVE Stainless Steel V14-14B VALVE GLOBE VALVE Stainless Steel V14-30A VALVE GLOBE VALVE Stainless Steel V14-30B VALVE GLOBE VALVE Stainless Steel V14-61A VALVE GLOBE VALVE Stainless Steel V14-61B VALVE GLOBE VALVE Stainless Steel V14-840A VALVE INSTRUMENT ISOLATION Stainless Steel V14-840B VALVE INSTRUMENT ISOLATION Stainless Steel V14-843A VALVE GLOBE VALVE Stainless Steel V14-843B VALVE GLOBE VALVE Stainless Steel Feedwater System Components 10" FDW-18 PIPING, 10" or greater 10" FW piping from V29B to N4C Carbon Steel 10" FDW-19 PIPING, 10" or greater 10" FW piping from V29A to N4A Carbon Steel 10" FDW-20 PIPING, 10" or greater 10" FW piping from V29B to N4D Carbon Steel 10" FDW-21 PIPING, 10" or greater 10" FW piping from V29A to N4B Carbon Steel 16" FDW-16 PIPING, 10" or greater 16" FW piping from V27A to V29A Carbon Steel 16" FDW-17 PIPING, 10" or greater 16" FW piping from V96A to V29B Carbon Steel 3/4" PIPING, 3/4" 3/4" FW test connections Carbon Steel V2-26A VALVE ISOLATION VALVE Carbon Steel V2-26B VALVE ISOLATION VALVE Carbon Steel V2-27A VALVE NBS SWING CHECK VALVE Carbon Steel V2-28A VALVE NBS SWING CHECK VALVE Carbon Steel V2-28B VALVE NBS SWING CHECK VALVE Carbon Steel V2-29A VALVE FEEDWATER STOP Carbon Steel V2-29B VALVE FEEDWATER STOP Carbon Steel V2-94A VALVE NBS GLOBE VALVE Carbon Steel V2-94B VALVE NBS GLOBE VALVE Carbon Steel V2-96A VALVE NBS SWING CHECK VALVE Carbon Steel

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant Revision 0 System Pressure Boundary Page 39 of 57 Attachment 1 - Components Subject to Aging Management Review Comp ID Component Type Component Name Material Main Steam System Components CC-2-115A CHAMBER MS FLOW A INST. COND. CHAM. Stainless Steel CC-2-115B CHAMBER MS FLOW A INST. COND. CHAM. Stainless Steel CC-2-115C CHAMBER MS FLOW B INST. COND. CHAM. Stainless Steel CC-2-115D CHAMBER MS FLOW B INST. COND. CHAM. Stainless Steel CC-2-115E CHAMBER MS FLOW C INST. COND. CHAM. Stainless Steel CC-2-115F CHAMBER MS FLOW C INST. COND. CHAM. Stainless Steel CC-2-115G CHAMBER MS FLOW D INST. COND. CHAM. Stainless Steel CC-2-115H CHAMBER MS FLOW D INST. COND. CHAM. Stainless Steel RO-2-73A ORIFICE INSTRUMENT ORIFICE Carbon Steel RO-2-73B ORIFICE INSTRUMENT ORIFICE Carbon Steel RO-2-73C ORIFICE INSTRUMENT ORIFICE Carbon Steel RO-2-73D ORIFICE INSTRUMENT ORIFICE Carbon Steel RO-2-73E ORIFICE INSTRUMENT ORIFICE Carbon Steel RO-2-73F ORIFICE INSTRUMENT ORIFICE Carbon Steel RO-2-73G ORIFICE INSTRUMENT ORIFICE Carbon Steel RO-2-73H ORIFICE INSTRUMENT ORIFICE Carbon Steel RO-23-37A ORIFICE Instrument Orifice, MS flow to HPCI Stainless Steel RO-23-37B ORIFICE Instrument Orifice, MS flow to HPCI Stainless Steel RO-13-55A ORIFICE Instrument Orifice, Flow to RCIC Stainless Steel RO-13-55B ORIFICE Instrument Orifice, Flow to RCIC Stainless Steel RO-23-80A ORIFICE Instrument Orifice, MS flow to HPCI Stainless Steel RO-23-80B ORIFICE Instrument Orifice, MS flow to HPCI Stainless Steel RO-13-137A ORIFICE Instrument Orifice, Flow to RCIC Stainless Steel RO-13-137B ORIFICE Instrument Orifice, Flow to RCIC Stainless Steel 3/4" MS PIPING, 3" or less 3/4" test connection piping Carbon Steel 1" MSD CS-5 PIPING, 3" or less MS Drain Piping Carbon Steel 11/2 MSD CS-5 PIPING, 3" or less MS Drain Piping Carbon Steel 2" MS PIPING, 3" or less 2" MSD lines Carbon Steel 3" MS-5A PIPING, 3" or less MS piping to RCIC Carbon Steel 3" MSD-2 PIPING, 3" or less 3" msd lines Carbon Steel 3/4" MS SS-6 PIPING, 3" or less Instrumentation Piping 3/4" ms ss-6 Stainless Steel 1" MS PIPING, 3" or less MS Inst lines 1" Stainless Steel 3" MS PIPING, 3" or less Inst. Off RV2-71A/B/C/D Stainless Steel 6" SRV PIPING, 6" or greater 6" to SRVs on main steam Carbon Steel 10" MS-4A PIPING, 6" or greater MS to HPCI piping Carbon Steel 18" MS-7A PIPING, 6" or greater MS line A, N3A to V86A Carbon Steel 18" MS-7B PIPING, 6" or greater MS line B, N3B to V86B Carbon Steel 18" MS-7C PIPING, 6" or greater MS line C, N3C to V86C Carbon Steel 18" MS-7D PIPING, 6" or greater MS line D, N3D to V86D Carbon Steel FE-2-114A RESTRICTOR STEAM FLOW ELEMENT, Line A CASS FE-2-114B RESTRICTOR STEAM FLOW ELEMENT, Line B CASS FE-2-114C RESTRICTOR STEAM FLOW ELEMENT, Line C CASS FE-2-114D RESTRICTOR STEAM FLOW ELEMENT, Line D CASS

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant Revision 0 System Pressure Boundary Page 40 of 57 Attachment 1 - Components Subject to Aging Management Review Comp ID Component Type Component Name Material V-2-115A VALVE GLOBE VALVE Stainless Steel V-2-115B VALVE GLOBE VALVE Stainless Steel V-2-115C VALVE GLOBE VALVE Stainless Steel V-2-115D VALVE GLOBE VALVE Stainless Steel V-2-115E VALVE GLOBE VALVE Stainless Steel V-2-115F VALVE GLOBE VALVE Stainless Steel V-2-115G VALVE GLOBE VALVE Stainless Steel V-2-115H VALVE GLOBE VALVE Stainless Steel V13-15 VALVE MS VALVE, MS to RCIC Carbon Steel V13-150A VALVE GLOBE VALVE, MSD on MS to RCIC Carbon Steel V13-16 VALVE MS VALVE, MS to RCIC Carbon Steel V13-46A VALVE MS test conn isolation, on MS to RCIC Carbon Steel V23-15 VALVE GLOBE VALVE, MS to HPCI Carbon Steel V23-16 VALVE GLOBE VALVE, MS to HPCI Carbon Steel V23-160A VALVE MS to HPCI drain valve Carbon Steel V23-27A VALVE MS drain off MS to HPCI Carbon Steel V2-74 VALVE MOV Carbon Steel V2-74A VALVE LIFT CHECK VALVE Carbon Steel V2-74A1 VALVE ISOLATION VALVE Carbon Steel V2-74A2 VALVE ISOLATION VALVE Carbon Steel V2-74A3 VALVE ISOLATION VALVE Carbon Steel V2-74A4 VALVE ISOLATION VALVE Carbon Steel V2-75 VALVE NBS GLOBE VALVE Carbon Steel V2-77 VALVE MOV Carbon Steel V2-80A VALVE MS ISOLATION VALVE Carbon Steel V2-80B VALVE MS ISOLATION VALVE Carbon Steel V2-80C VALVE MS ISOLATION VALVE Carbon Steel V2-80D VALVE MS ISOLATION VALVE Carbon Steel V2-83A VALVE GLOBE VALVE Carbon Steel V2-83B VALVE GLOBE VALVE Carbon Steel V2-83C VALVE GLOBE VALVE Carbon Steel V2-83D VALVE GLOBE VALVE Carbon Steel V2-86A VALVE MS ISOLATION VALVE Carbon Steel V2-86B VALVE MS ISOLATION VALVE Carbon Steel V2-86C VALVE MS ISOLATION VALVE Carbon Steel V2-86D VALVE MS ISOLATION VALVE Carbon Steel SL-13-55A VALVE NB FLOW LIM. Valve, MS to RCIC inst. Stainless Steel SL-13-55B VALVE NB FLOW LIM. Valve, MS to RCIC inst. Stainless Steel SL-13-55C VALVE NB FLOW LIM. Valve, MS to RCIC inst. Stainless Steel SL-13-55D VALVE NB FLOW LIM. Valve, MS to RCIC inst. Stainless Steel SL-23-37A VALVE NB FLOW LIM Valve, MS to HPCI inst. Stainless Steel SL-23-37B VALVE NB FLOW LIM Valve, MS to HPCI inst. Stainless Steel SL-23-37C VALVE NB FLOW LIM Valve, MS to HPCI inst. Stainless Steel SL-23-37D VALVE NB FLOW LIM Valve, MS to HPCI inst. Stainless Steel SL-2-73A VALVE NB FLOW LIMITING VALVE Stainless Steel

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant Revision 0 System Pressure Boundary Page 41 of 57 Attachment 1 - Components Subject to Aging Management Review Comp ID Component Type Component Name Material SL-2-73B VALVE NB FLOW LIMITING VALVE Stainless Steel SL-2-73C VALVE NB FLOW LIMITING VALVE Stainless Steel SL-2-73D VALVE NB FLOW LIMITING VALVE Stainless Steel SL-2-73E VALVE NB FLOW LIMITING VALVE Stainless Steel SL-2-73F VALVE NB FLOW LIMITING VALVE Stainless Steel SL-2-73G VALVE NB FLOW LIMITING VALVE Stainless Steel SL-2-73H VALVE NB FLOW LIMITING VALVE Stainless Steel V13-54A VALVE GLOBE VALVE, MS to RCIC instrum. Stainless Steel V13-54B VALVE GLOBE VALVE, MS to RCIC instrum. Stainless Steel V13-54C VALVE GLOBE VALVE, MS to RCIC instrum. Stainless Steel V13-54D VALVE GLOBE VALVE, MS to RCIC instrum. Stainless Steel V23-35A VALVE ISOLATION VALVE, MS to HPCI inst. Stainless Steel V23-35B VALVE ISOLATION VALVE, MS to HPCI inst. Stainless Steel V23-35C VALVE ISOLATION VALVE, MS to HPCI inst. Stainless Steel V23-35D VALVE ISOLATION VALVE, MS to HPCI inst. Stainless Steel V2-72A VALVE NBS GLOBE VALVE Stainless Steel V2-72B VALVE NBS GLOBE VALVE Stainless Steel V2-72C VALVE NBS GLOBE VALVE Stainless Steel V2-72D VALVE NBS GLOBE VALVE Stainless Steel V2-72E VALVE NBS GLOBE VALVE Stainless Steel V2-72F VALVE NBS GLOBE VALVE Stainless Steel V2-72G VALVE NBS GLOBE VALVE Stainless Steel V2-72H VALVE NBS GLOBE VALVE Stainless Steel V2-90A VALVE Inst. Isolation off RV-71A Stainless Steel V2-90B VALVE Inst. Isolation off RV-71B Stainless Steel V2-90C VALVE Inst. Isolation off RV-71C Stainless Steel V2-90D VALVE Inst. Isolation off RV-71D Stainless Steel V2-90E VALVE Inst. Isolation off RV-71A Stainless Steel V2-90F VALVE Inst. Isolation off RV-71B Stainless Steel V2-90G VALVE Inst. Isolation off RV-71C Stainless Steel V2-90H VALVE Inst. Isolation off RV-71D Stainless Steel

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant Revision 0 System Pressure Boundary Page 42 of 57 Attachment 1 - Components Subject to Aging Management Review Comp ID Component Type Component Name Material Nuclear Boiler Components FLANGE BOLTS BOLTING RPV VENT FLANGE BOLTS off N7 Low Alloy Steel FLANGE FLANGE RPV VENT FLANGE off N7 Carbon Steel RO-2-23 ORIFICE Instrument Orifice off N13 leak ind. Carbon Steel 1/2 NB PIPING, 2" or less Piping off N13 leak ind. Carbon Steel 1 PSE PIPING, 2" or less 1 PSE (SS-6, 1.1) line off N13 leak ind. Stainless Steel 1" NB PIPING, 2" or less 1" leak detection off N13 & N14 Carbon Steel 2" NB PIPING, 2" or less 2" drain lines off N15 & vent lines off N7 Carbon Steel 3/4" NB PIPING, 2" or less 3/4" drain lines off N15 Stainless Steel 2" NB PIPING, 2" or less 2" drain lines off N15 Stainless Steel FCV-2-20 VALVE AIR OPERATED VALVE off N13 Carbon Steel SL-2-23 VALVE NB FLOW LIMITING VALVE off N13 Carbon Steel V2-15 VALVE GLOBE VALVE off N7 Carbon Steel V2-46 VALVE GLOBE VALVE off N7 Carbon Steel FCV-2-17 VALVE GLOBE VALVE off N7 Carbon Steel FCV-2-18 VALVE GLOBE VALVE off N7 Carbon Steel V2-19 VALVE ISOLATION VALVE, N7 to MS Carbon Steel FCV-2-21 VALVE AIR OPERATED VALVE off N13 Carbon Steel V2-22 VALVE NBS GLOBE VALVE off N13 Carbon Steel V2-49 VALVE RV drain valve off N15 Stainless Steel V2-50 VALVE RV drain valve off N15 Carbon Steel V2-99 VALVE GLOBE VALVE off N15 Stainless Steel Nuclear Boiler Vessel Instrumentation Components FLANGE BOLTS BOLTING INST. LINE FLANGE BOLTS, off N7 Stainless Steel 2-3-3A CHAMBER INST. CONDENSING CHAM., off N11A Stainless Steel 2-3-3B CHAMBER INST. CONDENSING CHAM., off N11B Stainless Steel 2-3-1 CHAMBER INST. CONDENSING CHAM., off N7 Stainless Steel FLANGE FLANGE INST. LINE FLANGE, off N7 Stainless Steel RO-2-3-25 ORIFICE INSTRUMENT ORIFICE off N10 Stainless Steel RO-2-3-27 ORIFICE INSTRUMENT ORIFICE off N10 Stainless Steel RO-2-3-11 ORIFICE INSTRUMENT ORIFICE off N7 Stainless Steel RO-2-3-13A ORIFICE INSTRUMENT ORIFICE, off N11A Stainless Steel RO-2-3-13B ORIFICE INSTRUMENT ORIFICE, off N11B Stainless Steel RO-2-3-15A ORIFICE INSTRUMENT ORIFICE off N12A Stainless Steel RO-2-3-15B ORIFICE INSTRUMENT ORIFICE off N12B Stainless Steel RO-2-3-19A ORIFICE INSTRUMENT ORIFICE, off N11A Stainless Steel RO-2-3-19B ORIFICE INSTRUMENT ORIFICE, off N11B Stainless Steel PIPING PIPING, 1" or less 1.5" instrument lines Stainless Steel PIPING PIPING, 1" or less 1" INSTRUMENT PIPING Stainless Steel TE-80A TEMP ELEMENT Temp Element off N11A Stainless Steel TE-80B TEMP ELEMENT Temp Element off N11B Stainless Steel TE-80C TEMP ELEMENT Temp Element off N11A Stainless Steel TE-80D TEMP ELEMENT Temp Element off N11B Stainless Steel V2-3-10 VALVE NBVI GLOBE VALVE off N7 Stainless Steel

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant Revision 0 System Pressure Boundary Page 43 of 57 Attachment 1 - Components Subject to Aging Management Review Comp ID Component Type Component Name Material SL-2-3-11 VALVE NBVI FLOW LIMIT VALVE off N7 Stainless Steel V2-3-12A VALVE NBVI GLOBE VALVE, off N11A Stainless Steel V2-3-12B VALVE NBVI GLOBE VALVE, off N11B Stainless Steel SL-2-3-13A VALVE NBVI FLOW LIMIT VALVE, off N11A Stainless Steel SL-2-3-13B VALVE NBVI FLOW LIMIT VALVE, off N11B Stainless Steel V2-3-14A VALVE NBVI GLOBE VALVE, off N12A Stainless Steel V2-3-14B VALVE NBVI GLOBE VALVE, off N12B Stainless Steel SL-2-3-15A VALVE NBVI FLOW LIMIT VALVE, off N12A Stainless Steel SL-2-3-15B VALVE NBVI FLOW LIMIT VALVE, off N12B Stainless Steel V2-3-16A VALVE NBVI GLOBE VALVE, off N12A Stainless Steel V2-3-16B VALVE NBVI GLOBE VALVE, off N12B Stainless Steel SL-2-3-17A VALVE NBVI FLOW LIMIT VALVE, off N12A Stainless Steel SL-2-3-17B VALVE NBVI FLOW LIMIT VALVE, off N12B Stainless Steel V2-3-18A VALVE NBVI GLOBE VALVE, off N11A Stainless Steel V2-3-18B VALVE NBVI GLOBE VALVE, off N11B Stainless Steel SL-2-3-19A VALVE NBVI FLOW LIMIT VALVE, off N11A Stainless Steel SL-2-3-19B VALVE CHECK VALVE, off N11B Stainless Steel V2-3-20A VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-20B VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-20C VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-20D VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel SL-2-3-21A VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-21B VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-21C VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-21D VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel V2-3-22A VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-22B VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-22C VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-22D VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel SL-2-3-23A VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-23B VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-23C VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-23D VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel V2-3-24 VALVE NBVI GLOBE VALVE off N10 Stainless Steel SL-2-3-25 VALVE NBVI FLOW LIMIT VALVE off N10 Stainless Steel V2-3-26 VALVE NBVI GLOBE VALVE off N10 Stainless Steel SL-2-3-27 VALVE NBVI FLOW LIMIT VALVE off N10 Stainless Steel V2-3-28A VALVE NBVI GLOBE VALVE, off N11A Stainless Steel V2-3-28B VALVE NBVI GLOBE VALVE, off N11B Stainless Steel SL-2-3-29A VALVE NBVI FLOW LIMIT VALVE, off N11A Stainless Steel SL-2-3-29B VALVE NBVI FLOW LIMIT VALVE, off N11B Stainless Steel V2-3-30A VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-30B VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-30C VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-30D VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant Revision 0 System Pressure Boundary Page 44 of 57 Attachment 1 - Components Subject to Aging Management Review Comp ID Component Type Component Name Material V2-3-30E VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-30F VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-30G VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-30H VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-30I VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-30J VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-30K VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-30L VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-30M VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-30N VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-30O VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-30P VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel V2-3-30Q VALVE NBVI GLOBE VALVE, jet pump inst Stainless Steel SL-2-3-31A VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-31B VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-31C VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-31D VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-31E VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-31F VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-31G VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-31H VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-31I VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-31J VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-31K VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-31L VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-31M VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-31N VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-31P VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel SL-2-3-31Q VALVE NBVI FLOW LIMIT VALVE, jet pump inst Stainless Steel V2-3-32 VALVE NBVI GLOBE VALVE off N10 Stainless Steel SL-2-3-33 VALVE CHECK VALVE off N10 Stainless Steel V2-3-34 VALVE NBVI GLOBE VALVE off N10 Stainless Steel SL-2-3-35 VALVE CHECK VALVE off N10 Stainless Steel SN-3-78 VALVE NBVI GLOVE VALVE, off N10 Stainless Steel SN-3-80 VALVE NBVI GLOVE VALVE, off N10 Stainless Steel SN-3-81 VALVE NBVI GLOVE VALVE, off N10 Stainless Steel SN-3-85 VALVE NBVI GLOVE VALVE, off N10 Stainless Steel SN-3-86 VALVE NBVI GLOVE VALVE, off N10 Stainless Steel SN-3-88 VALVE NBVI GLOVE VALVE, off N10 Stainless Steel SN-2-3-102A VALVE NBVI GLOVE VALVE, off N11A Stainless Steel SN-2-3-102B VALVE NBVI GLOVE VALVE, off N11A Stainless Steel SN-2-3-103A VALVE NBVI GLOVE VALVE, off N11B Stainless Steel SN-2-3-103B VALVE NBVI GLOVE VALVE, off N11B Stainless Steel V2-3-428A VALVE NBVI isolation valve off N11A Stainless Steel V2-3-428B VALVE NBVI isolation valve off N11A Stainless Steel

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant Revision 0 System Pressure Boundary Page 45 of 57 Attachment 1 - Components Subject to Aging Management Review Comp ID Component Type Component Name Material V2-3-429A VALVE NBVI isolation valve off N11B Stainless Steel V2-3-429B VALVE NBVI isolation valve off N11B Stainless Steel V2-3-430A VALVE NBVI check valve off N11A Stainless Steel V2-3-430B VALVE NBVI check valve off N11A Stainless Steel V2-3-431A VALVE NBVI isolation valve off N11B Stainless Steel V2-3-431B VALVE NBVI isolation valve off N11B Stainless Steel V2-3-432A VALVE NBVI check valve off N11A Stainless Steel V2-3-432B VALVE NBVI check valve off N11A Stainless Steel V2-3-433A VALVE NBVI check valve off N11B Stainless Steel V2-3-433B VALVE NBVI check valve off N11B Stainless Steel V2-3-434A VALVE NBVI isolation valve off N11A Stainless Steel V2-3-434B VALVE NBVI isolation valve off N11A Stainless Steel V2-3-435A VALVE NBVI check valve off N11B Stainless Steel V2-3-435B VALVE NBVI check valve off N11B Stainless Steel V2-3-437A VALVE NBVI isolation valve off N11B Stainless Steel V2-3-437B VALVE NBVI isolation valve off N11B Stainless Steel V2-3-440A VALVE NBVI vent/drain valve off N11A Stainless Steel V2-3-440B VALVE NBVI vent/drain valve off N11A Stainless Steel V2-3-442A VALVE NBVI vent/drain valve off N11B Stainless Steel V2-3-442B VALVE NBVI vent/drain valve off N11B Stainless Steel V2-3-443A VALVE NBVI vent/drain valve off N11B Stainless Steel V2-3-443B VALVE NBVI vent/drain valve off N11B Stainless Steel V2-3-444A VALVE NBVI vent/drain valve off N11A Stainless Steel V2-3-444B VALVE NBVI vent/drain valve off N11A Stainless Steel V2-3-445A VALVE NBVI vent/drain valve off N11B Stainless Steel V2-3-445B VALVE NBVI vent/drain valve off N11B Stainless Steel V2-3-447A VALVE NBVI vent/drain valve off N11B Stainless Steel V2-3-447B VALVE NBVI vent/drain valve off N11B Stainless Steel V2-3-448A VALVE NBVI vent/drain valve off N11A Stainless Steel V2-3-448B VALVE NBVI vent/drain valve off N11A Stainless Steel V2-3-451A VALVE NBVI vent/drain valve off N11B Stainless Steel V2-3-451B VALVE NBVI vent/drain valve off N11B Stainless Steel Residual Heat Removal Components 3/4" RHR PIPING, 2" or less RHR vent/drain PIPING Carbon Steel 3/4" RHR PIPING, 2" or less Test and Drain Piping Stainless Steel 20" RHR-33 PIPING, 4" or greater RHR PIPING Carbon Steel 24" RHR Piping PIPING, 4" or greater RHR PIPING Carbon Steel 20" RHR-32 PIPING, 4" or greater RHR PIPING Stainless Steel 24" RHR Piping PIPING, 4" or greater RHR PIPING Stainless Steel V10-17 VALVE GLOBE VALVE, RHR suction from RR Carbon Steel V10-17A1 VALVE GLOBE VALVE, vent/drain near V10-17 Carbon Steel V10-18A VALVE CHECK VALVE, RHR relief piping Carbon Steel V10-18A2 VALVE ANGLE VALVE, RHR relief piping Carbon Steel V10-18A4 VALVE GLOBE VALVE, RHR relief piping Carbon Steel

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant Revision 0 System Pressure Boundary Page 46 of 57 Attachment 1 - Components Subject to Aging Management Review Comp ID Component Type Component Name Material V10-25A VALVE GLOBE VALVE, RHR return A Carbon Steel V10-25B VALVE GLOBE VALVE, RHR return B Carbon Steel V10-27A VALVE GLOBE VALVE, MOV, RHR return A Carbon Steel V10-27B VALVE GLOBE VALVE, MOV, RHR return B Carbon Steel V10-78A VALVE GLOBE VALVE, RHR return A, 3/4" Carbon Steel V10-78B VALVE GLOBE VALVE, RHR return B, 3/4" Carbon Steel V10-84 VALVE GLOBE VALVE, RHR suction vent/drain Carbon Steel V10-91A VALVE GLOBE VALVE, RHR return A, drain/vent Carbon Steel V10-91B VALVE GLOBE VALVE, RHA return B, drain/vent Carbon Steel V10-18 VALVE GLOBE VALVE, RHR suction from RR Stainless Steel V10-18A1 VALVE ANGLE VALVE, RHR relief piping Stainless Steel V10-18A3 VALVE GLOBE VALVE, RHR relief piping Stainless Steel V10-18A5 VALVE CHECK VALVE, RHR relief piping Stainless Steel V10-18A6 VALVE ANGLE VALVE, RHR relief piping Stainless Steel V10-18A7 VALVE GLOBE VALVE, RHR relief piping Stainless Steel V10-46A VALVE CHECK VALVE, RHR return A Stainless Steel V10-46B VALVE CHECK VALVE, RHR return B Stainless Steel V10-81A VALVE GLOBE VALVE, RHR return A Stainless Steel V10-81B VALVE GLOBE VALVE, RHR return B Stainless Steel V10-88 VALVE GLOBE VALVE, RHR suction from RR Stainless Steel V10-88A VALVE GLOBE VALVE, vent/drain near V10-88 Stainless Steel V10-92A VALVE GLOBE VALVE, RHR return A, drain/vent Stainless Steel V10-92B VALVE GLOBE VALVE, RHA return B, drain/vent Stainless Steel V10-93A VALVE GLOBE VALVE, RHR return A, drain/vent Stainless Steel V10-93B VALVE GLOBE VALVE, RHA return B, drain/vent Stainless Steel V10-193A VALVE GLOBE VALVE, V10-46A leakoff Stainless Steel V10-193B VALVE GLOBE VALVE, V10-46B leakoff Stainless Steel V10-194 VALVE GLOBE VALVE, MO V18 leakoff Stainless Steel Reactor Recirculation System Components P18-1A-S/N/W-1 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1B-S/N/W-1 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1A-S/N/W-2 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1B-S/N/W-2 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1A-S/N/W-3 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1B-S/N/W-3 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1A-S/N/W-4 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1B-S/N/W-4 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1A-S/N/W-5 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1B-S/N/W-5 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1A-S/N/W-6 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1B-S/N/W-6 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1A-S/N/W-7 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1B-S/N/W-7 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1A-S/N/W-8 BOLTING STUD, NUT, WASHER Low Alloy Steel

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant Revision 0 System Pressure Boundary Page 47 of 57 Attachment 1 - Components Subject to Aging Management Review Comp ID Component Type Component Name Material P18-1B-S/N/W-8 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1A-S/N/W-9 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1B-S/N/W-9 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1A-S/N/W-10 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1B-S/N/W-10 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1A-S/N/W-11 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1B-S/N/W-11 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1A-S/N/W-12 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1B-S/N/W-12 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1A-S/N/W-13 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1B-S/N/W-13 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1A-S/N/W-14 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1B-S/N/W-14 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1A-S/N/W-15 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1B-S/N/W-15 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1A-S/N/W-16 BOLTING STUD, NUT, WASHER Low Alloy Steel P18-1B-S/N/W-16 BOLTING STUD, NUT, WASHER Low Alloy Steel FLANGE FLANGE FLANGE on P-18-1A suction xconn Stainless Steel FLANGE FLANGE FLANGE on P-18-1A discharge xconn Stainless Steel FLANGE FLANGE FLANGE on P-18-1B suction xconn Stainless Steel FLANGE FLANGE FLANGE on P-18-1B discharge xconn Stainless Steel FE-2-109A FLOW ELEMENT FLOW ELEMENT Stainless Steel FE-2-109B FLOW ELEMENT FLOW ELEMENT Stainless Steel RO-2-062A ORIFICE INSTRUMENT ORIFICE, RR pump A DP Stainless Steel RO-2-062B ORIFICE INSTRUMENT ORIFICE, RR pump A DP Stainless Steel RO-2-062C ORIFICE INSTRUMENT ORIFICE, RR pump B DP Stainless Steel RO-2-062D ORIFICE INSTRUMENT ORIFICE, RR pump B cap Stainless Steel RO-2-064A ORIFICE INSTRUMENT ORIFICE, RR loop A flow Stainless Steel RO-2-064B ORIFICE INSTRUMENT ORIFICE, RR loop A flow Stainless Steel RO-2-064C ORIFICE INSTRUMENT ORIFICE, RR loop B flow Stainless Steel RO-2-064D ORIFICE INSTRUMENT ORIFICE, RR loop B flow Stainless Steel RO-2-97A ORIFICE INSTRUMENT ORIFICE, RR pump A cap Stainless Steel RO-2-97B ORIFICE INSTRUMENT ORIFICE, RR pump B DP Stainless Steel RO-2-305A ORIFICE INSTRUMENT ORIFICE, RR B suct pres Stainless Steel RO-2-305B ORIFICE INSTRUMENT ORIFICE, RR B suct pres Stainless Steel 3/4" RR PIPING, 1" or less 3/4" PSA Recirc Sample lines Stainless Steel 1" RR PIPING, 1" or less 1" PSA RR instrument lines Stainless Steel 4" PLR PIPING, 4" or greater 4" Recirc Xconnect Stainless Steel 12" PLR PIPING, 4" or greater 12" Recirc lines Stainless Steel 22" PLR PIPING, 4" or greater 22" Recirc lines Stainless Steel 28" PLR PIPING, 4" or greater 28" Recirc lines Stainless Steel P-18-1A PUMP CASING RX RECIRC PUMP A CASS P-18-1B PUMP CASING RX RECIRC PUMP B CASS P-18-1A PUMP COVER ASSY RX RECIRC PUMP A CASS P-18-1B PUMP COVER ASSY RX RECIRC PUMP B CASS

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant Revision 0 System Pressure Boundary Page 48 of 57 Attachment 1 - Components Subject to Aging Management Review Comp ID Component Type Component Name Material P-18-1A DRIVER MOUNT RX RECIRC PUMP A Carbon Steel P-18-1B DRIVER MOUNT RX RECIRC PUMP B Carbon Steel TX-2-107A THERMOWELL P18-1A SUCTION TEMP Stainless Steel TX-2-107B THERMOWELL P18-1B SUCTION TEMP Stainless Steel TX-2-108A THERMOWELL P18-1A DISCH TEMP Stainless Steel TX-2-108B THERMOWELL P18-1B DISCH TEMP Stainless Steel V2-38A VALVE GLOBE VALVE, MOV-54A leakoff Stainless Steel V2-38B VALVE GLOBE VALVE, MOV-54B leakoff Stainless Steel V2-39A VALVE GLOBE VALVE, MOV-54A leakoff Stainless Steel V2-39B VALVE GLOBE VALVE, MOV-54B leakoff Stainless Steel V2-39AA VALVE CHECK VALVE, RR sample Stainless Steel V2-39AA1 VALVE ANGLE VALVE, RR sample Stainless Steel V2-39AA2 VALVE ANGLE VALVE, RR sample Stainless Steel V2-39AA3 VALVE GLOBE VALVE, RR sample Stainless Steel V2-39AA4 VALVE GLOBE VALVE, RR sample Stainless Steel FCV-2-39 VALVE GLOBE VALVE, RR sample Stainless Steel FCV-2-40 VALVE GLOBE VALVE, RR sample Stainless Steel V2-41 VALVE GLOBE VALVE, RR sample Stainless Steel V2-42 VALVE GLOBE VALVE, RR sample Stainless Steel V2-43A VALVE GLOBE VALVE, RR pump A suct MOV Stainless Steel V2-43B VALVE GLOBE VALVE, RR pump B suct MOV Stainless Steel V2-45A VALVE GLOBE VALVE, V2-43A leakoff Stainless Steel V2-45B VALVE GLOBE VALVE, V2-45B leakoff Stainless Steel V2-46A VALVE GLOBE VALVE, V2-43A leakoff Stainless Steel V2-46B VALVE GLOBE VALVE, V2-45B leakoff Stainless Steel V2-53A VALVE GLOBE VALVE, RR pump A disc MOV Stainless Steel V2-53B VALVE GLOBE VALVE, RR pump B disc MOV Stainless Steel V2-54A VALVE MOV, RR pump A disc xconn Stainless Steel V2-54B VALVE MOV, RR pump B disc xconn Stainless Steel V2-56A VALVE GLOBE VALVE, V2-53A leakoff Stainless Steel V2-56B VALVE GLOBE VALVE, V2-53B leakoff Stainless Steel V2-57A VALVE GLOBE VALVE, V2-53A leakoff Stainless Steel V2-57B VALVE GLOBE VALVE, V2-53B leakoff Stainless Steel V2-61A VALVE NBS GLOBE VALVE off RO62A Stainless Steel V2-61B VALVE NBS GLOBE VALVE off RO62B Stainless Steel V2-61C VALVE NBS GLOBE VALVE off RO62C Stainless Steel V2-61D VALVE NBS GLOBE VALVE off RO97B Stainless Steel SL-2-62A VALVE NB FLOW LIMITING VALVE off RO62A Stainless Steel SL-2-62B VALVE NB FLOW LIMITING VALVE off RO62B Stainless Steel SL-2-62C VALVE NB FLOW LIMITING VALVE off RO62C Stainless Steel SL-2-62D VALVE NB FLOW LIMITING VALVE off RO97B Stainless Steel V2-63A VALVE NBS GLOBE VALVE off RO64A Stainless Steel V2-63B VALVE NBS GLOBE VALVE off RO64B Stainless Steel V2-63C VALVE NBS GLOBE VALVE off RO64C Stainless Steel V2-63D VALVE NBS GLOBE VALVE off RO64D Stainless Steel

VYNPS License Renewal Project AMRM-33 Aging Management Review of the Reactor Coolant Revision 0 System Pressure Boundary Page 49 of 57 Attachment 1 - Components Subject to Aging Management Review Comp ID Component Type Component Name Material SL-2-64A VALVE NB FLOW LIMITING VALVE off RO64A Stainless Steel SL-2-64B VALVE NB FLOW LIMITING VALVE off RO64B Stainless Steel SL-2-64C VALVE NB FLOW LIMITING VALVE off RO64C Stainless Steel SL-2-64D VALVE NB FLOW LIMITING VALVE off RO64D Stainless Steel V2-91 VALVE SAMPLE LINE off RR A loop return hdr Stainless Steel V2-92A VALVE GLOBE VALVE, RR drain A pump suct Stainless Steel V2-92B VALVE GLOBE VALVE, RR drain B pump suct Stainless Steel V2-93A VALVE GLOBE VALVE, RR drain A pump suct Stainless Steel V2-93B VALVE GLOBE VALVE, RR drain B pump suct Stainless Steel SL-2-97A VALVE NB FLOW LIMITING VALVE, off RO97A Stainless Steel SL-2-97B VALVE NB FLOW LIMITING VALVE, off RO62D Stainless Steel V2-98A VALVE NBS GLOBE VALVE off RO97A, capped Stainless Steel V2-98B VALVE NBS GLOBE VALVE off RO62D, capped Stainless Steel V2-304A VALVE NBS GLOBE VALVE off RO305A Stainless Steel V2-304B VALVE NBS GLOBE VALVE off RO305B Stainless Steel SL-2-305A VALVE NB FLOW LIMITING VALVE off RO305A Stainless Steel SL-2-305B VALVE NB FLOW LIMITING VALVE off RO305B Stainless Steel Reactor Water Cleanup System Components 3/4" PIPING, 2" or less drain line Stainless Steel 2" CUW(5S-6, 1.1) PIPING, 2" or less RV N15 TO RWCU Stainless Steel 2 CUW-19 Piping From RPV drain Stainless Steel 4" CUW-18 PIPING, 4" CUW-18, from RHR Stainless Steel V12-15 VALVE GLOBE VALVE, RHR to RWCU Stainless Steel V12-16 VALVE GLOBE VALVE, test/drain RWCU Stainless Steel V12-18 VALVE GLOBE VALVE, RHR to RWCU Stainless Steel V12-46 VALVE GLOBE VALVE, RHR to RWCU Stainless Steel V12-117A VALVE Test line isolation on RV drain Stainless Steel Standby Liquid Control System Components FSH-11-54 FLOW ELEMENT Flow Element Stainless Steel 3/4" SLC Piping, 1 1/2" or less SLC test conn piping Stainless Steel 1 1/2" SLC-11 Piping, 1 1/2" or less SLC Piping Stainless Steel V11-16 VALVE CHECK VALVE Stainless Steel V11-17 VALVE CHECK VALVE Stainless Steel V11-18 VALVE GLOBE VALVE Stainless Steel V11-36 VALVE GLOBE VALVE Stainless Steel

AMRM-33 VYNPS License Renewal Project Revision 0 Aging Management Review of the Reactor Coolant System Pressure Boundary Page 50 of 57 Attachment 2 - Aging Management Review Results Intended Component type Material Environment Aging effect Program function Bolting Pressure boundary Stainless steel Air-indoor (ext) Cracking - fatigue TLAA - metal fatigue (flanges, valves, etc.) Cracking Inservice Inspection Low alloy steel Air-indoor (ext) Loss of Material System walkdowns Inservice inspection Cracking - fatigue TLAA - metal fatigue Cracking Inservice inspection Condensing chambers Pressure boundary Stainless steel Treated water Loss of material Water chemistry control - BWR

>270 °F (int)

Cracking Water chemistry control - BWR One time inspection Cracking - fatigue TLAA - metal fatigue Air-indoor (ext) None NA Detector (CRD) Pressure boundary Stainless steel Treated water Loss of material Water chemistry control - BWR

>270 °F (int)

Cracking Water chemistry control - BWR One time inspection Cracking - fatigue TLAA - metal fatigue Air-indoor (ext) None NA Drive (CRD) Pressure boundary Stainless steel Treated water Loss of material Water chemistry control - BWR

>270 °F (int)

Cracking Water chemistry control - BWR Inservice Inspection Cracking - fatigue TLAA - metal fatigue Air-indoor (ext) None NA Driver mount (RR) Pressure boundary Carbon steel Air-indoor (ext) Loss of material System walkdowns

AMRM-33 VYNPS License Renewal Project Revision 0 Aging Management Review of the Reactor Coolant System Pressure Boundary Page 51 of 57 Attachment 2 - Aging Management Review Results Intended Component type Material Environment Aging effect Program function Filter housing Pressure boundary Stainless steel Treated water Water chemistry control - BWR Loss of material (CRD) >270 °F (int)

Cracking Water chemistry control - BWR One time inspection Cracking - fatigue TLAA - metal fatigue Air-indoor (ext) None NA Flow elements Pressure boundary Stainless steel Treated water Water chemistry control - BWR Loss of material (RR, SLC) >270 °F (int) Inservice inspection Cracking Water chemistry control - BWR Inservice inspection BWR stress corrosion cracking Cracking - fatigue TLAA - metal fatigue Air-indoor (ext) None NA Orifices Pressure boundary Stainless steel Treated water Water chemistry control - BWR Loss of material (Instrumentation) >270 °F (int)

Cracking Water chemistry control - BWR One time inspection Cracking - fatigue TLAA - metal fatigue Air-indoor (ext) None NA Piping & fittings Pressure boundary Stainless steel Treated water Water chemistry control - BWR Loss of material

<4 NPS >270 °F (int)

(includes flange leak Water chemistry control - BWR Cracking off lines) One time inspection Cracking - fatigue TLAA - metal fatigue Air-indoor (ext) None NA Carbon steel Treated water Loss of material Water chemistry control - BWR

>220 °F (int)

Cracking - fatigue TLAA - metal fatigue Air-indoor (ext) None NA

AMRM-33 VYNPS License Renewal Project Revision 0 Aging Management Review of the Reactor Coolant System Pressure Boundary Page 52 of 57 Attachment 2 - Aging Management Review Results Intended Component type Material Environment Aging effect Program function Piping & fittings Pressure boundary Stainless steel Treated water (int) Water chemistry control - BWR Loss of material

<4 NPS (instrumentation, vent, Water chemistry control - BWR Cracking and drains) One time inspection Air-indoor (ext) None NA Carbon steel Treated water °F Loss of material Water chemistry control - BWR (int) One time inspection Air-indoor (ext) Loss of material System walkdowns Piping & fittings Pressure boundary Stainless steel Nitrogen None NA

<4 NPS Air-indoor (ext) None NA (CRD)

Piping & fittings Pressure boundary Stainless steel Treated water Loss of material Water chemistry control - BWR

>= 4 NPS >270 °F (int) Inservice inspection Cracking Water chemistry control - BWR Inservice inspection BWR stress corrosion cracking Cracking - fatigue TLAA - metal fatigue Air-indoor (ext) None NA Carbon steel Treated water Loss of material Water chemistry control - BWR

>220 °F (int) Inservice inspection Flow accelerated corrosion Cracking - fatigue TLAA - metal fatigue Air-indoor (ext) None NA

AMRM-33 VYNPS License Renewal Project Revision 0 Aging Management Review of the Reactor Coolant System Pressure Boundary Page 53 of 57 Attachment 2 - Aging Management Review Results Intended Component type Material Environment Aging effect Program function Piping & fittings Pressure boundary Stainless steel Treated water (int) Loss of material Water chemistry control - BWR

>= 4 NPS Inservice inspection Cracking Water chemistry control - BWR Inservice inspection BWR stress corrosion cracking Air-indoor (ext) None NA Carbon steel Treated water (int) Loss of material Water chemistry control - BWR Inservice inspection Air-indoor (ext) Loss of material System walkdowns Pump casing and Pressure boundary CASS Treated water Loss of material Water chemistry control - BWR cover (RR) >482 °F (int) Inservice inspection Cracking Water chemistry control - BWR BWR stress corrosion cracking Inservice inspection Cracking - fatigue TLAA - metal fatigue Reduction of fracture Inservice inspection toughness Air-indoor (ext) None NA Pump cover thermal Pressure boundary CASS Treated water (int) Loss of material Water chemistry control - CCW barrier(RR) Inservice inspection Cracking Water chemistry control - CCW Inservice inspection Air-indoor (ext) None NA Restrictors Flow control CASS Treated water Loss of material Water chemistry control - BWR (MS) >482 °F (int) One time inspection Cracking Water chemistry control - BWR One time inspection Cracking - fatigue TLAA - metal fatigue Reduction of One time inspection fracture toughness

AMRM-33 VYNPS License Renewal Project Revision 0 Aging Management Review of the Reactor Coolant System Pressure Boundary Page 54 of 57 Attachment 2 - Aging Management Review Results Intended Component type Material Environment Aging effect Program function Rupture disc Pressure boundary Stainless steel Nitrogen (int) None NA (CRD) Air-indoor (ext) None NA Tank (CRD) Pressure boundary Carbon steel Treated water (int) Loss of material Water chemistry control - BWR Nitrogen (int) None NA Air-indoor (ext) Loss of Material System Walkdowns Stainless steel Treated water (int) Loss of material Water chemistry control - BWR Water chemistry control - BWR Cracking One time inspection Nitrogen (int) None NA Air-indoor (ext) None NA Thermowells Pressure boundary Stainless steel Treated water Water chemistry control - BWR

<4 NPS >270 °F (int) Loss of material (All systems)

Cracking Water chemistry control - BWR One time inspection Cracking - fatigue TLAA - metal fatigue Air-indoor (ext) None NA Thermowells Pressure boundary Stainless steel Treated water Water chemistry control - BWR

>=4 NPS >270 °F (int) Loss of material (All systems)

Cracking Water chemistry control - BWR Inservice inspection BWR stress corrosion cracking Cracking - fatigue TLAA - metal fatigue Air-indoor (ext) None NA

AMRM-33 VYNPS License Renewal Project Revision 0 Aging Management Review of the Reactor Coolant System Pressure Boundary Page 55 of 57 Attachment 2 - Aging Management Review Results Intended Component type Material Environment Aging effect Program function Valve bodies Pressure boundary Carbon steel Treated water Loss of material Water chemistry control - BWR

<4 NPS >220 °F (int)

Cracking - fatigue TLAA - metal fatigue Air-indoor (ext) None NA CASS Treated water Loss of material Water chemistry control - BWR

>482 °F (int)

Cracking Water chemistry control - BWR One time inspection Cracking - fatigue TLAA - metal fatigue Reduction of fracture One time inspection toughness Air-indoor (ext) None NA Stainless steel Treated water Loss of material Water chemistry control - BWR

>270 °F (int)

Cracking Water chemistry control - BWR One time inspection Cracking - fatigue TLAA - metal fatigue Air-indoor (ext) None NA Valve bodies Pressure boundary Carbon steel Treated water (int) Loss of material Water chemistry control - BWR

<4 NPS Air-indoor (ext) Loss of material System walkdowns CASS Treated water (int) Loss of material Water chemistry control - BWR Cracking Water chemistry control - BWR One time inspection Air-indoor (ext) None NA Stainless steel Treated water (int) Loss of material Water chemistry control - BWR Cracking Water chemistry control - BWR One time inspection Air-indoor (ext) None NA

AMRM-33 VYNPS License Renewal Project Revision 0 Aging Management Review of the Reactor Coolant System Pressure Boundary Page 56 of 57 Attachment 2 - Aging Management Review Results Intended Component type Material Environment Aging effect Program function Valve bodies Pressure boundary Stainless steel Nitrogen (int) None NA

<4 NPS Air-indoor (ext) None NA Valve bodies Pressure boundary Stainless steel Treated water Water chemistry control - BWR Loss of material

>4 NPS >270 °F (int) Inservice Inspection Cracking Water chemistry control - BWR Inservice inspection BWR stress corrosion cracking Cracking - fatigue TLAA - metal fatigue Air-indoor (ext) None NA Carbon steel Treated water Loss of material Water chemistry control - BWR

>220 °F (int) Inservice inspection Flow accelerated corrosion Cracking - fatigue TLAA - metal fatigue Air-indoor (ext) None NA CASS Treated water Loss of material Water chemistry control - BWR

>482 °F (int) Inservice inspection Cracking Water chemistry control - BWR Inservice inspection BWR stress corrosion cracking Cracking - fatigue TLAA - metal fatigue Reduction of fracture Inservice inspection toughness Air-indoor (ext) None NA

AMRM-33 VYNPS License Renewal Project Revision 0 Aging Management Review of the Reactor Coolant System Pressure Boundary Page 57 of 57 Attachment 2 - Aging Management Review Results Intended Component type Material Environment Aging effect Program function Valve bodies Pressure boundary Stainless steel Treated water (int) Loss of material Water chemistry control - BWR

>4 NPS Inservice Inspection Cracking Water chemistry control - BWR Inservice inspection BWR stress corrosion cracking Air-indoor (ext) None NA Carbon steel Treated water (int) Loss of material Water chemistry control - BWR Inservice inspection Air-indoor (ext) Loss of material System Walkdowns CASS Treated water (int) Loss of material Water chemistry control - BWR Inservice inspection Cracking Water chemistry control - BWR Inservice inspection BWR stress corrosion cracking Air-indoor (ext) None NA

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