ML070570319

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AMRM-31, Rev. 0, Aging Management Review of the Reactor Pressure Vessel
ML070570319
Person / Time
Site: Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date: 01/25/2006
From: Finnin R
Entergy Nuclear Operations
To:
Office of Nuclear Reactor Regulation
O'Hara T, RI/DRS/PSB2, (610) 337-5043
References
AMRM-31, Rev 0
Download: ML070570319 (43)
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VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 2 of 43 REVISION DESCRIPTION SHEET Revision Number Description Pages and/or Sections Revised 0

Initial Issue

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 3 of 43 TABLE OF CONTENTS 1.0 Introduction.....................................................................................................................4 1.1 Purpose........................................................................................................................4 1.2 System Description.......................................................................................................4 1.3 System and Component Intended Functions...............................................................8 2.0 Screening.......................................................................................................................10 2.1 Component Evaluation Boundaries............................................................................10 2.2 Materials for Subcomponents Subject to Aging Management Review (SAMR).........12 2.3 Environments for the Reactor Vessel Subcomponents..............................................15 3.0 Aging Effects Requiring Management........................................................................17 3.1 Low Alloy Steel, Including Carbon Steel, Exposed to Treated Water.........................18 3.2 Low Alloy Steel, including Carbon Steel, exposed to Neutron Fluence......................19 3.3 Low Alloy Steel, Including Carbon Steel, Exposed to Air-indoor................................20 3.4 Stainless Steel Cladding.............................................................................................20 3.5 Stainless Steel and Nickel-Based Alloys Exposed to Treated Water.........................21 3.6 Stainless Steel and Nickel-Based Alloys Exposed to Neutron Fluence.....................22 3.7 Stainless Steel and Nickel-Based Alloys Exposed to Air-indoor................................22 3.8 Bolting.........................................................................................................................23 3.9 Operating Experience.................................................................................................24 4.0 Demonstration That Aging Effects will be Managed..................................................25 4.1 Aging Management Programs....................................................................................25 4.2 Time-Limited Aging Analyses.....................................................................................28 5.0 Summary and Conclusions..........................................................................................29 6.0 References.....................................................................................................................30 : Aging Management Review Results.............................................................33

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 4 of 43 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 pressure vessel (RPV) passive subcomponents 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 License Renewal Project Plan. (Ref. 12)

The purpose of this report is to demonstrate that aging effects for passive mechanical subcomponents of the VYNPS reactor pressure vessel will be adequately managed for the period of extended operation associated with license renewal. This aging management review report (AMRR) includes the reactor pressure vessel and associated interior and exterior integral attachments. 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 aging effects.

Applicable aging effects were determined using EPRI report 1003056 Non-Class 1 Mechanical Implementation Guideline and Mechanical Tools (Ref. 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 pressure vessel subcomponents covered in this AMR include materials and environments evaluated in the Mechanical Tools.

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

Report (Ref. 1) and BWR Vessel and Internals Project Reports BWRVIP-05 (Ref. 6), 74 (Ref.

5), 86 (Ref. 28), and 116 (Ref. 29), 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 pressure vessel.

1.2

System Description

As described in UFSAR Section 4.2, the reactor vessel is designed and fabricated in accordance with ASME Boiler and Pressure Vessel Code,Section III (1965 Edition with Summer 1966 Addenda), its interpretations, and applicable requirements for Class A vessels as defined therein. (Table 4.1-1 of Ref. 4) The vessel contains the following subcomponents.

1. Reactor vessel shell and heads

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 5 of 43

2. Reactor vessel cladding
3. Reactor vessel nozzles, safe ends and thermal sleeves
4. Control rod drive penetrations
5. Incore flux detector penetrations
6. Reactor vessel internal attachments
7. Reactor vessel supports
8. Reactor vessel exterior attachments
9. Reactor vessel pressure boundary bolting
10. Reactor vessel insulation 1.2.1 Reactor Vessel Shell and Heads The reactor vessel shell is a welded vertical cylinder with hemispherical heads. The cylindrical shell and hemispherical heads are fabricated of low alloy steel plate. The vessel bottom head is welded directly to the vessel shell. Full penetration welds are used at all joints including nozzles throughout the vessel, except for some nozzles and penetrations of less than 3-inch nominal size. (Section 4.2.4.1 of Ref. 4)

The flanged reactor vessel upper head is secured to the vessel shell by studs and nuts (See Section 1.2.9). The head and vessel flanges are low alloy steel forgings. The flanged joint between the vessel and vessel head is sealed by two concentric stainless steel seal rings designed for no detectable leakage. Taps are provided between the two rings and outside the outer ring to indicate seal leakage. (Section 4.2.4.1 of Ref. 4) 1.2.2 Reactor Vessel Cladding The cladding is not part of the pressure boundary; rather it provides a protective barrier to minimize corrosion of the low alloy steel and to minimize the release of corrosion products. The reactor vessel shell and heads are clad on the interior with stainless steel weld overlay (0.125 minimum). Some of the low alloy steel nozzles are fully clad, some are partially clad, and some are unclad as identified in Section 2.2.3. The sealing surfaces of the reactor vessel head and shell flanges are weld overlay clad with austenitic stainless steel.

1.2.3 Reactor Vessel Nozzles, Safe Ends, and Thermal Sleeves The vessel nozzles are low alloy steel forgings made in accordance with ASTM A508 as modified by ASME Code Case 1332-3, Paragraph 5. Nozzles of 3-inch nominal size or larger are full penetration welded to the vessel. Nozzles of less than 3-inch nominal size may be partial penetration welded as permitted by ASME Code,Section III. Nozzles which are partial penetration welded are nickel-chromium-iron forgings made in accordance with ASME SB-166 as modified by Code Case 1336. The vessel top head nozzles are provided with flanges with small groove facing. The drain nozzle is of the full penetration weld design and extends 14 inches below the bottom outside surface of the vessel. Feedwater inlet nozzles have thermal sleeves similar to those shown in the detail of Figure 4.2-2. The nozzle provided for the control rod drive hydraulic return line is capped. The cap is connected to the safe end with a full penetration weld. Nozzles connecting to stainless steel piping have "safe ends" of stainless

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 6 of 43 steel of types which are compatible with the material of the mating pipe. Nozzles for connecting carbon steel piping are clad through at least the thickness of the vessel wall or one-half the diameter of the nozzle bore, whichever is less. (Section 4.2.4.1 and Figure 4.2-2 of the UFSAR)

Table 1.2-1 gives a complete listing of all vessel nozzles, safe ends and thermal sleeves.

Table 1.2-1 Reactor Vessel Nozzles, Safe Ends and Thermal Sleeves Nozzle #

Description Qty Size (inches)

Nozzle Type Safe End Thermal sleeve N11 Reactor recirculation outlets1 21 281 Forged, full penetration2 Yes4 No N21 Reactor recirculation inlets1 101 121 Forged, full penetration2 Yes4 Yes5 N31 Main steam outlets1 41 181 Forged, full penetration3 Yes4 No N41 Feedwater inlets1 41 101 Forged, full penetration2 Yes4 Yes5 N51 Core spray inlets1 21 81 Forged, full penetration2 Yes4 Yes5 N6A1 Head spray1 11 61 Forged, full penetration3 No -

flanged10 No N6B1 Head instrument1 11 61 Forged, full penetration3 No -

flanged10 No N71 Head vent1 11 41 Forged, full penetration3 No -

flanged10 No N81 Jet pump instruments1 21 41 Forged, full penetration3 Yes4 No N91 CRD return1 11 31 Forged, full penetration2 No -

capped7 No N101 Core DP/SLC1 11 21 Forged, full penetration2 No3 No8 N11 &

N121 Instrumentation1 41 21 Insert, partial penetration3 Yes4 No N13 &

N141 RPV flange leakoff1 21 11 Drilled penetrations1 No6 No N151 Drain1 11 21 Full penetration9 Yes4 No

1.

The nozzle numbers, description, quantity, and size came from drawing 919D294 in Ref.

19 and Table 4.2-2 of the UFSAR [Ref. 4]

2.

Nozzle configurations taken from drawing 919D294 in Ref. 19.

3.

Nozzle configurations taken from drawing 104R940 in Ref. 19.

4.

Nozzles connected to stainless steel piping have safe ends of stainless steel. Nozzles connected to carbon steel pipe have safe ends made of low allow steel.

5.

The recirculation inlet nozzles, feedwater inlet nozzles, and core spray inlet nozzles have thermal sleeves. (Figure 4.2-2 of Ref. 4)

6.

The flange leakoff nozzles (N13 and N14) consist of drilled penetrations in the vessel flange (between the two seal rings and outside the outer ring) with nickel-based alloy piping inserted in the holes.

7.

The control rod drive hydraulic return line nozzle was capped and the associated piping removed due to stress corrosion cracking. (Section 4.2.4.1 and Figure 4.2-2 of Ref. 4).

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 7 of 43 The cap is connected directly to the nozzle with a full penetration weld. The safe end and thermal sleeve were removed.

8.

The nozzle for the core differential pressure and standby liquid control pipe is designed with a transition that provides an annular region between the nozzle and the inner standby liquid control line to minimize thermal shock effects on the reactor vessel in the event that use of the Standby Liquid Control System is required. (Section 4.2.4.1 of the UFSAR)

9.

The drain nozzle is of the full penetration weld design and extends 14 inches below the bottom outside surface of the vessel. [Section 4.2.4.1 of the UFSAR]

10.

The vessel top head nozzles are provided with flanges with small groove facing. [Section 4.2.4.1 of the UFSAR]

1.2.4 Control Rod Drive Penetrations There are 89 control rod drive penetrations (6 inch) in the reactor vessel bottom head. A control rod drive stub tube is inserted into each penetration from inside the reactor vessel and is secured by a partial penetration weld. The control rod drive (CRD) housings are then inserted through the stub tube and welded to the end of the stub tube inside the reactor vessel. Each CRD housing transmits loads to the bottom head of the reactor vessel. (Section 4.2.4.6, table 4.2.1, and figure 4.2-2 of Ref. 4) The housings extend below the reactor vessel and terminate in flanges to which the drive mechanisms are bolted.

1.2.5 Incore Flux Detector Penetrations There are 30 incore flux detector penetrations (2) in the bottom head of the reactor vessel. An instrument housing is inserted in each penetration and partial-penetration welded to the inside of the reactor vessel head. The instrument housing terminates in a flange at the lower end.

Either a dry tube (for source and intermediate range detectors) or a local power range monitor (LPRM) assembly is inserted into the instrument housing. The dry tube or LPRM bolts to the flange at the bottom of the incore housing to complete the pressure boundary. Traveling incore probes (TIPs) travel in guide tubes inside the local power range monitors. (Sections 4.2.4.8, 7.5, and Table 4.2.1 of Ref. 4) 1.2.6 Reactor Vessel Internal Attachments There are multiple attachments to the reactor pressure vessel, for supporting various internal components. These internal attachments include the following.

Internal Attachment Quantity Reference Shroud support ring pad 1

Figure 2.9.2.4 of Ref. 39 Shroud support feet 14 Drawing 5920-252 Ref. 41 Jet pump riser support pads 20 Table 2.4-3 of the UFSAR Guide rod brackets 2

Table 2.4-3 of the UFSAR Steam dryer brackets 4

Table 2.4-3 of the UFSAR Dryer holddown brackets 4

Table 2.4-3 of the UFSAR Feedwater sparger brackets 8

Table 2.4-3 of the UFSAR Core spray brackets 4

Table 2.4-3 of the UFSAR Surveillance specimen holder brackets 6

Table 2.4-3 of the UFSAR

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 8 of 43 1.2.7 Reactor Vessel Supports The reactor vessel is supported by a low-alloy steel skirt. The top of the skirt is welded to the bottom of the vessel. The skirt is then supported by a concrete and steel pedestal which carries the load through the drywell to the reactor building foundation slab.

The reactor vessel is laterally and vertically supported and braced to make it as rigid as possible without impairing movement required for thermal expansion. (Section 4.2.4.3 of Ref. 4) Vessel stabilizers are connected between the reactor vessel stabilizer brackets and the top of the shield wall surrounding the vessel. (Section 4.2.4.4 of Ref. 4) 1.2.8 Reactor Vessel Exterior Attachments There are multiple external attachments to the reactor pressure vessel. (The support skirt and stabilizer brackets were discussed in Section 1.2.7) The external attachments include the following.

External Attachment Quantity Reference Head lifting lugs 4

Table 4.2.3 of the UFSAR Insulation supports 2

Table 4.2.3 of the UFSAR Insulation support brackets 24 Table 4.2.3 of the UFSAR Thermocouple pads 32 Table 4.2.3 of the UFSAR 1.2.9 Reactor Vessel Pressure Boundary Bolting The reactor vessel upper head is secured to the reactor vessel shell by studs and nuts. The studs pass through the head closure flange and are threaded into the vessel closure flange.

The vessel flanges are sealed as discussed in Section 1.2.1 above. [Section 4.2.4.1 of the UFSAR]

Other bolting reviewed in this AMR are the CRD mechanism to CRD housing bolts, the incore dry tube to incore housing bolts, the LPRM assembly to incore housing bolts, and the vessel nozzle flange to connecting flange bolts for nozzles N6 and N7.

1.2.10 Reactor Vessel Insulation The lower head and cylindrical shell insulation is permanently installed. The insulation panels for the cylindrical shell of the vessel are held in place by vessel insulation supports located at two elevations on the vessel. The support brackets for each support are full-penetration welded to the vessel at 12 evenly spaced locations around the circumference. (Section 4.2.4.9 of the UFSAR) 1.3 System and Component Intended Functions As described in UFSAR Section 4.2.1, the power generation objectives of the reactor vessel are:

(1) to contain the reactor core, reactor internals, and the reactor coolant moderator, and (2) to serve as a high integrity barrier against leakage of radioactive materials to the drywell.

There are no safety objectives for the reactor vessel in the UFSAR; however, VY Vermont Yankee Site Specific Guidance and System Safety Function Sheets, ENN-MS-S-009-VY, give

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 9 of 43 the safety functions for each system at VY. The safety functions of the reactor coolant system (nuclear boiler system) are the following. (Ref. 3)

1. Provide and maintain a high integrity reactor coolant pressure boundary inside and out to the first isolation outside primary containment to prevent leakage of radioactive materials.
2. Provide for primary containment isolation/boundary.
3. Provide flow paths for ECCS system injection into the vessel.
4. Contain/structurally support the reactor core, reactor internals, reactor coolant moderator and reactivity control portions of the system.
5. Relieve any overpressure that occurs during abnormal operational transients and over pressurization of the nuclear system via four (4) SRVs and three (3) SVs.
6. Provide pressure relief (in conjunction with ADS) via four (4) SRVs to allow for core cooling.
7. Provide means for emergency and alternate cooling (i.e., feed and bleed) via the core spray spargers and nozzles.
8. Provide valid signals to interfacing plant systems necessary for reactivity control/RPS, SBGT initiation, containment isolation, ARI/RPT and emergency core cooling initiation to prevent the onset and mitigate the consequences of an accident.
9. Provide indication for operators to initiate and control systems used during and following accident and abnormal conditions or to monitor the status of safety systems during design basis accident events (e.g., R.G. 1.97 Category 1 variables).
10. Provide steam quenching capability and primary containment integrity following a LOCA (by directing SRV discharges below the water level of the suppression pool via the discharge lines and actuation of vacuum breakers).
11. Provide structural integrity for safety-related portions of safety-related systems.

The intended functions for the reactor pressure vessel subcomponents subject to aging management review are the following.

reactor coolant system pressure boundary support for Criterion (a)(1) equipment corrosion protection (for cladding)

Refer to VYNPS Report LRPD-01, System and Structure Scoping Results, for additional information on scoping and intended functions of systems and structures for license renewal.

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 10 of 43 2.0 Screening 2.1 Component Evaluation Boundaries The major components of the reactor pressure vessel include the reactor pressure vessel shell, lower head, upper closure head, cladding, flanges, studs, nuts, nozzles and safe ends. The component evaluation boundaries for this AMRR are the welds between the safe ends and attached piping and the interface flanges for bolted connections. Thermal sleeves that are welded to vessel nozzles or safe ends are reviewed in this AMRR. The control rod drive stub tubes, control rod drive housings, and incore housings are also included in this AMRR. The vessel support skirt, vessel interior welded attachments, and vessel exterior attachments are also considered in this AMRR. Each reactor pressure vessel subcomponent is discussed below. All reactor pressure vessel subcomponents subject to aging management review in this AMRR, along with their materials of construction, are listed in Attachment 1.

2.1.1 Reactor Vessel Shell and Heads The reactor pressure vessel shell, lower head and upper closure head are subject to aging management review and are reviewed in this AMRR. The reactor internals are evaluated in AMRM-32, Aging Management Review of the Reactor Vessel Internals.

The upper head and vessel flanges, including the studs and nuts, are subject to aging management review and are reviewed in this AMRR. See Section 2.1.9 for discussion of the studs and nuts. The RPV o-rings are periodically replaced. Therefore, these o-rings are not subject to aging management review per 10 CFR 54, section 54.21(a)(1)(ii). The flange leak-off lines are evaluated in AMRM-33, Aging Management Review of the Reactor Coolant System Pressure Boundary.

2.1.2 Reactor Vessel Cladding The reactor vessel cladding is subject to aging management review and is reviewed in this AMRR.

2.1.3 Reactor Vessel Nozzles, Safe Ends, and Thermal Sleeves All vessel nozzles and welds that attach the nozzles to the vessel are subject to aging management review and are reviewed in this AMRR. All vessel nozzle safe ends, including the welds that attach the nozzles to safe ends are subject to aging management review and are reviewed in this AMR. The welds that attach safe ends to attached piping and the attached Class 1 piping and valves out to the ASME Section XI IWB inspection boundary are evaluated in AMRM-33, Aging Management Review of the Reactor Coolant System Pressure Boundary.

Thermal sleeves attached to the recirculation inlet nozzles, feedwater inlet nozzles, and core spray inlet nozzles are subject to aging management review and are reviewed in this AMRR.

The internal piping attached to these nozzles/thermal sleeves is reviewed in AMRM-32, Aging Management Review of the Reactor Vessel Internals.

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 11 of 43 2.1.4 Control Rod Drive Penetrations The control rod drive stub tubes, control rod drive housings and the bolts for the housing flange are subject to aging management review and are reviewed in this AMRR. The mating flange on the control rod drive mechanism and the remainder of the control rod drive hydraulic system are evaluated in AMRM-33, Aging Management Review of the Reactor Coolant System Pressure Boundary. The control rod guide tubes and the thermal sleeve that locks the guide tube to the housing are reviewed in AMRM-32, Aging Management Review of the Reactor Vessel Internals.

2.1.5 Incore Detector Penetrations The incore housings are subject to aging management review and are reviewed in this AMRR.

The incore guide tubes, dry tubes and LPRM assemblies are evaluated in AMRM-32, Aging Management Review of the Reactor Vessel Internals.

2.1.6 Reactor Vessel Internal Attachments Vessel interior welded attachments (as described in Section 1.2.6) are subject to aging management review and are reviewed in this AMRR. The subcomponents attached to these attachments (not welded to the vessel) are evaluated in AMRM-32, Aging Management Review of the Reactor Vessel Internals.

2.1.7 Reactor Vessel Supports The support skirt and vessel stabilizer brackets are subject to aging management review and are reviewed in this AMRR. The reactor pressure vessel stabilizers are evaluated in AMRC-06, Aging Management Review of Bulk Commodities.

2.1.8 Reactor Vessel Exterior Attachments External attachments are subject to aging management review if they are load bearing attachments connected to pressure retaining portions of the vessel. The lifting lugs do not bear significant weight during power operation and are not subject to aging management review.

The thermocouple pads and insulation support brackets bear insignificant weight and are not subject to aging management review. The refueling bellows support is connected to the outer surface of the vessel flange, beyond the pressure boundary, and is not subject to aging management review.

2.1.9 Reactor Vessel Pressure Boundary Bolting The pressure boundary bolting discussed in section 1.2.9 is subject to aging management review and is reviewed in this AMRR. Bolting in this AMR includes the reactor pressure vessel closure flange bolting, the CRD closure bolting, the incore detector closure bolting, and the three nozzles on the upper head flange bolting.

2.1.10 Reactor Vessel Insulation The vessel insulation is reviewed in AMRC-06, Aging Management Review of Bulk Commodities. (Ref. 47)

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 12 of 43 2.2 Materials for Subcomponents Subject to Aging Management Review (SAMR)

This section lists the materials of construction for those subcomponents that were identified in Section 2.1 as subject to aging management review.

2.2.1 Reactor Vessel Shell, Heads, and Flanges The reactor vessel shell and heads are made of low alloy steel, A533 Grade B cc1339-2. The closure flanges are forged of low alloy steel, SA508 Class 2 cc1332. (UFSAR table 4.2-1 and Refs. 17, 31, and 32) 2.2.2 Reactor Vessel Cladding The vessel shell and heads are weld overlay clad with austenitic stainless steel which consists of a minimum of two layers and a minimum of 0.125 inch total thickness after all machining.

The first layer is deposited with a composition equivalent to ASTM A371, Type ER309, and the second layer has a composition equivalent to ASTM A371, Type ER308, except that the carbon content does not exceed 0.08%.

The sealing surfaces of the reactor vessel head and shell flanges are weld overlay clad similar to the vessel shell and heads except with a minimum of 0.25-inch total thickness after all machining, including the area under the seal grooves. (Section 4.2.4.1 of the UFSAR),

(NE8067 Appendix A, Section 19.1, Ref. 22), (GE specification 21A1115, Ref. 31) 2.2.3 Reactor Vessel Nozzles, Safe Ends, and Thermal Sleeves The materials for the vessel nozzles, safe ends, and thermal sleeves are given in table 2.2-1.

Table 2.2-1 Nozzle

  1. 1 Description1 Nozzle Material Clad Safe End Material Thermal Sleeve or Cap or Blank Flange Material N1 Reactor recirculation outlets LAS, A508 Cl2 cc13322 Yes10 SS, A182 F3164 NA N2 Reactor recirculation inlets LAS, A508 Cl22 Yes10 SS, A182 F3164 Thermal sleeve:

Type 304 Austenitic Stainless Steel2 N3 Main steam outlets LAS, A508 Cl22 Partial9 CS, SA516 Grade 705 NA N4 Feedwater inlets LAS, A508 Cl22 Partial9 CS, SA508 Class 26 Thermal Sleeve:

SS, SA312 Grade 304 NBA, Alloy 6002 N5 Core spray inlets LAS, A508 Cl22 Yes10 NBA, MS-16 (Inconel)7 Thermal sleeve:

Type 304 Austenitic Stainless Steel2

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 13 of 43 Nozzle

  1. 1 Description1 Nozzle Material Clad Safe End Material Thermal Sleeve or Cap or Blank Flange Material N6A Head spray LAS, A508 Cl22 Yes11 NA11 Blank Flange:

A182 Grade F3048 Blank Flange -

A182 Grade F3048 N6B Head instrument LAS, A508 Cl22 Yes11 NA11 Blank Flange:

A182 Grade F3048 N7 Head vent LAS, A508 Cl22 Yes12 NA12 Blank Flange:

A182 Grade F3048 N8 Jet pump instruments LAS, A508 Cl22 Yes SS, A336 class F84 NA N9 CRD return LAS, A508 Cl22 Yes10 NA Cap:

SA182 Grade 316L2 N10 Core DP/SLC LAS, A508 Cl22 Yes10 A336 Class 1 F84 NA N11 &

N12 Instrumentation NBA, SB 166 cc 13362 No A479 Type 316(SE)4 NA N13 &

N14 RPV flange leakoff NBA, SB-1663 No13 NA13 NA N15 Drain LAS, A508-1, Nuclear3 Partial12 LAS, A508-112 NA

1.

The nozzle numbers and description repeat the table in section 1.2.3 and came from drawing 919D294 in Reference 19 and Table 4.2-2 of the UFSAR (Ref. 4)

2.

Nozzle and thermal sleeve material from Table 4.2.1 of the UFSAR and the certified test reports, Ref. 32.

3.

Nozzle material taken from Section 3 of the certified test reports, Ref. 32

4.

Material taken from the PP7015 (Ref. 17) Appendix C, Table 5

5.

Drawing 5920-483, Ref. 40

6.

Drawing 5920-241, Ref. 50

7.

Drawing 5920-624, Ref. 36

8.

Section 8.12 of GE specification 21A1115, Ref. 31

9.

Nozzles for connecting carbon steel piping are clad through at least the thickness of the vessel wall or one-half the diameter of the nozzle bore, whichever is less. (Section 4.2.4.1 of the UFSAR)

10.

Cladding is shown on drawing 919D294 of Ref 19.

11.

Nozzles 6A and 6B are shown on drawing 5920-243 (Ref. 42)

12.

Nozzles 7 and 15 are shown on drawing 5920-244 (Ref. 43)

13.

RPV leakoff, nozzles N13 and N14, are shown on drawing 5920-324 (Ref. 44) 2.2.4 Control Rod Drive Penetrations The control rod drive stub tubes are made of nickel-based alloy, SB167 cc 1336 and the control rod housings are Type 304 austenitic stainless steel. (Drawing 919D294 of Ref. 19 and Table 4.2.1 and figure 4.2-2 of the UFSAR)

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 14 of 43 2.2.5 Incore Detector Penetrations The incore housings are type 304 or type 316 stainless steel. (Drawing 729E946 of Ref. 19 and Table 4.2.1 of the UFSAR) 2.2.6 Reactor Vessel Internal Attachments Internal Attachment Material Reference Shroud Support ring pad NBA, Alloy 182 Sec. 4.2 of Ref. 22 Shroud support feet NBA, Alloy 600 Sec. 4.6 of Ref. 22 Jet pump riser support pads SS, E308L Sec. 9.6 of Ref. 22 Guide rod brackets SS, E308L Sec. 5.5 of Ref. 22 Steam dryer brackets SS, SA240 Type 304 Drawing 5920-329 (Ref. 45)

Dryer holddown brackets SS, Type 304 Section 3.3.4 of the UFSAR Feedwater sparger brackets SS, SA240 Type 304 Drawing 5920-330 (Ref. 46)

Core spray brackets SS, Type 304 Section 3.3.4 of the UFSAR Surveillance specimen holder brackets SS, Type 304 Section 3.3.4 of the UFSAR 2.2.7 Reactor Vessel Supports The support skirt is made of low alloy steel, A533 Grade B. (Appendix C, Table 5 of Ref. 17)

The vessel stabilizer brackets are also made of A533 Grade B per Section 8.11.2 of the vessel specification, Ref. 31.

2.2.8 Reactor Vessel Exterior Attachments The reactor pressure vessel exterior attachments are not subject to aging management review as discussed in section 2.1.8.

2.2.9 Reactor Vessel Pressure Boundary Bolting The reactor vessel closure studs, nuts and washers/bushings are made of low alloy steel, A-540 Grade B. (Appendix C, Table 5 of Ref. 17)

The CRD flange closure bolting (capscrews and washers) is low alloy steel, SA193 Grade B7.

(page 3.4-27 of the UFSAR)

The bolts for the three flanges on the upper head (N6A, N6B, N7) are low alloy steel, A193 Grade B7. (Section 8.12.3 of Ref. 31) The nuts are low alloy steel SA194 Grade 2H. (Section 8.12.3 of Ref. 31) Material is consistent with the piping specification, Ref. 23.

The incore detector closure bolting consists of a stainless steel nut and washer to connect the stainless steel flange to the dry tube. The flange bolts that hold the dry tube or LPRM assembly to the incore housing are stainless steel, SA182 F304 or F316. (Ref. 49) 2.2.10 Reactor Vessel Insulation The reactor pressure vessel insulation is not reviewed in this AMRR as discussed in section 2.1.10.

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 15 of 43 2.3 Environments for the Reactor Vessel Subcomponents The reactor pressure vessel operating environments are treated water and neutron fluence on internal surfaces and air-indoor (i.e., drywell environment) on external surfaces.

2.3.1 Treated Water The majority of 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. 2).

Treated water greater than 270 ºF. Above this threshold, stainless steel is susceptible to fatigue (Appendix H of Ref. 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. 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. (Ref. 8) Refer to Section 4.1.10 for more information regarding the VYNPS Water Chemistry Program.

The vessel support skirt is the one subcomponent that is normally below 212 °F.

2.3.2 Neutron Fluence The region of the reactor vessel immediately around the core, the beltline, is exposed to neutron radiation in excess of 1x1017 n/cm2. Section 3.2.1 provides further discussion of the radiation effect on the beltline region.

2.3.3 Air-indoor (External)

Mechanical subcomponents of the reactor pressure vessel are located in the primary containment (drywell), which is a sheltered and controlled environment. The atmosphere is inerted with nitrogen to a maximum oxygen level of 4% (Section 5.2.6.3 of the UFSAR), making the atmosphere less corrosive than air. Normal drywell temperature during plant operation is between 135°F and 165°F. (Section 5.2.3.2 of the UFSAR). For purposes of this AMR, no credit is taken for the nitrogen, and all subcomponents of the reactor pressure vessel that are exposed to containment atmosphere are conservatively assumed to be exposed to air-indoor.

External surfaces of most of the reactor vessel subcomponents normally exceed 212 °F and thus do not have moisture present on those surfaces.

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 16 of 43 Subcomponents completely within the reactor pressure vessel (internal vessel attachments and thermal sleeves) have the internal environment of treated water and no external environment.

External vessel attachments have the external environment of air-indoor with no internal environment.

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 17 of 43 3.0 Aging Effects Requiring Management Industry reports (BWRVIPs), NUREG-1801, and EPRI report 1003056 are used in this section to identify and evaluate aging effects for the reactor pressure vessel. The following aging effects, and the associated aging mechanisms, were identified for the material/environmental combinations present in the reactor pressure vessel.

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. 2)

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

Galvanic corrosion is not applicable due to the materials chosen for class 1 components.

Erosion and flow-accelerated corrosion are not applicable to the reactor pressure vessel as the components are made of non-susceptible materials.

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

Thermal embrittlement is not applicable to the reactor vessel as none of the VYNPS reactor vessel subcomponents are made of the susceptible material - duplex ferritic-austenitic stainless steel castings (CASS).

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 the specific subcomponent materials and environments. The review was performed for groups of subcomponents with similar operating environments and materials of construction.

The AMR results are tabulated in Attachment 1.

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 18 of 43 3.1 Low Alloy Steel, Including Carbon Steel, Exposed to Treated Water The reactor pressure vessel subcomponents of carbon steel or low alloy steel clad with stainless steel and exposed to treated water are the upper head, bottom head, shell, closure flange on the vessel, closure flange on the upper head, and all or part of various nozzles as identified in Table 2.2-1. The base metal is discussed in this section and the cladding is discussed in Section 3.3.

The reactor pressure vessel subcomponents of unclad carbon steel or low alloy steel exposed to treated water are all or part of various nozzles and safe ends as identified in Table 2.2-1, and the flanges on the upper head nozzles.

All of these subcomponents are exposed to treated water on the inside and air - primary containment on the outside.

3.1.1 Loss of Material Unclad low-alloy and carbon steel internal surfaces exposed to treated water are susceptible to general corrosion. These surfaces are also susceptible to loss of material due to pitting and crevice corrosion in the presence of high oxygen levels and contaminants. Therefore, loss of material (general corrosion, pitting corrosion and crevice corrosion) is an aging effect requiring management for RPV unclad low alloy steel items exposed internally to treated water.

The carbon steel or low alloy steel base metal of those components clad with stainless steel is not exposed to the treated water, hence loss of material is not a concern for the base metal.

3.1.2 Cracking - Fatigue Carbon steel and low alloy steel items, clad or unclad, exposed to treated water are susceptible to cracking by fatigue whenever the temperature exceeds 220 degrees °F. ASME Section III, Subsection NB requires calculation of cumulative usage factors (CUF), and the usage factors must be less than one for the period of extended operation. Cumulative usage factor assessment is a time-limited aging analysis (TLAA). For more information on TLAA, see Section 4.12. Cracking due to fatigue is discussed in VYNPS Report LRPD-04, TLAA -

Mechanical Fatigue.

3.1.3 Cracking - Other Than Fatigue Service loads may result in the growth of pre-service flaws (Ref. 9) or initiation and growth of service-induced flaws. 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 within the joint. Therefore, cracking (flaw initiation and growth) within the welded joints is an aging effect requiring management for carbon steel and low alloy steel components, clad or unclad, 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.

Cracking has occurred in cladding of BWR vessels, including the VYNPS reactor vessel. See Section 3.9, Operating Experience for more details. Cracking of the cladding is not expected to

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 19 of 43 propagate into the low-alloy steel base metal. This is summarized in the NRC SER for Hatch Plant (Ref. 11) as follows, As for SCC of the low-alloy steel vessel shells, BWRVIP-05, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations, (Ref. 6) and BWRVIP-60, Evaluation of Crack Growth in BWR Low Alloy Steel RPV Internals, (Ref. 30) indicate that even if cracks were to emanate from the vessel cladding, they are not expected to propagate into the low-alloy steel of the reactor vessel. BWRVIP-05 and BWRVIP-60 have been reviewed and approved by the staff. Cracking of the base material due to cracking in the cladding is not an aging effect requiring management for low alloy steel clad with stainless steel.

Stress corrosion cracking and intergranular attack are not significant aging mechanisms for low alloy steel in treated water.

3.2 Low Alloy Steel, including Carbon Steel, exposed to Neutron Fluence 3.2.1 Reduction of Fracture Toughness Low-alloy steels subjected to high levels of high-energy neutrons are susceptible to increase in material strength and resultant decreased low cycle fatigue resistance known as reduction of fracture toughness. Reduction of fracture toughness is typically measured in terms of Charpy transition temperature shift and Charpy upper-shelf energy decrease. Based on the materials and projected fluence levels, only reactor pressure vessel shell subcomponents in the beltline region are susceptible to reduction of fracture toughness.

The beltline is defined by 10 CFR 50 Appendix G, Fracture Toughness Requirements (Ref. 10) to be the region of the reactor pressure vessel that directly surrounds the effective height of the active core and adjacent regions of the reactor pressure vessel that are predicted to experience sufficient neutron irradiation damage to be considered in the selection of the most limiting material with regard to radiation damage. In addition, 10 CFR 50 Appendix H (Ref. 35) does not require material surveillance testing for ferritic materials unless the peak neutron fluence at the end of the design life exceeds 1.0x1017 n/cm2. The beltline may thus be alternatively defined as reactor pressure vessel ferritic materials with an end-of-life fluence that exceeds 1.0x1017 n/cm2.

At VYNPS, the beltline for 40-years consists of four plates (1-14, 1-15, 1-16 and 1-17) and their connecting welds, all adjacent to the active fuel zone. There are no nozzles in the beltline region for the current term of operation (Ref. 5). The beltline has been re-evaluated for 60 years (see VYNPS report LRPD-03, TLAA and Exemption Evaluation Results) by extrapolating the data found in Refs. 33 and 34. No nozzles will be added to the beltline region due to additional fluence incurred during the period of extended operation at the uprated power. The plate and weld material currently in the beltline for 40-years remain the only materials in the beltline for the period of extended operation. Reduction of fracture toughness applies equally to clad or unclad low alloy steel; however, there are no unclad components in the beltline region.

Reduction of fracture toughness due to radiation embrittlement is an aging effect requiring management for the base metals and weld metals within the beltline region. Analysis of reduction of fracture toughness is a TLAA addressed in VYNPS report LRPD-03, TLAA and Exemption Evaluation Results.

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 20 of 43 3.3 Low Alloy Steel, Including Carbon Steel, Exposed to Air-indoor Low alloy steel components exposed only to air are the vessel OD attachments; including the stabilizer attachment brackets and the vessel support skirt and its attachment weld. The reactor pressure vessels external surfaces (shell, heads, and nozzles) are exposed to air-indoor.

3.3.1 Loss of Material External ferritic steel surfaces exposed to air-indoor are susceptible to loss of material only if the surface is exposed to moisture. (Appendix E of Ref. 2). The reactor vessel is over 212 °F during plant operation and consequently the material is not exposed to moisture. The vessel support skirt is typically less than 212 °F.

Therefore, loss of material due to general corrosion is considered an aging effect requiring management only for the reactor vessel support skirt.

3.3.2 Cracking - Fatigue Cracking due to fatigue is discussed in Section 3.1.2 above.

3.3.3 Cracking - Other Than Fatigue External exposure to air-indoor does not contribute to cracking of low alloy steel or carbon steel.

Therefore, cracking is not an aging effect requiring management for carbon steel and low alloy steel external surfaces exposed to air-indoor.

3.3.4 Reduction of Fracture Toughness Reduction of fracture toughness is discussed in Section 3.2.1 above for the neutron fluence environment.

3.4 Stainless Steel Cladding The reactor vessel internal cladding is considered an independent subcomponent at VYNPS.

The internal environment of the cladding is treated water. Since the cladding is welded to the base metal, there is no external environment to address. Other stainless steel subcomponents are addressed in Sections 3.4 and 3.5.

3.4.1 Loss of Material Due to physical configuration or small surface defects, system fluid contaminants could concentrate in crevices within the reactor pressure vessel. With a high enough concentration of contaminants in the treated water, the cladding is susceptible to loss of material by pitting and crevice corrosion (Ref. 2). Therefore, loss of material (crevice and pitting corrosion) is an aging effect requiring management for stainless steel cladding.

3.4.2 Cracking - Fatigue At or near attachment welds, cracking of cladding due to fatigue is an aging effect requiring management.

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 21 of 43 3.4.3 Cracking - Other Than Fatigue A high concentration of contaminants in treated water may cause stainless steel cladding and attachment welds to the cladding to crack due to stress corrosion cracking/intergranular attack.

Stress corrosion cracking of cladding at VYNPS has occurred as discussed in Section 3.9, Operating Experience. While the cladding does not directly contribute to the primary pressure boundary, clad cracking may expose the underlying ferritic steel to treated water, which may lead to corrosion of the base metal. However, as discussed in Section 3.1.3, the cladding cracks will not propagate into the underlying ferritic steel Cracking (SCC) is an aging effect requiring management for stainless steel cladding.

3.4.4 Reduction of Fracture Toughness Reduction in fracture toughness due to radiation embrittlement is an aging effect that can affect the strength of the pressure boundary in the reactor pressure vessel beltline region. However, the stainless steel cladding is not credited for adding strength to the reactor vessel, only for protecting the low alloy steel from treated water. Therefore, reduction of fracture toughness due to radiation embrittlement is not an aging effect requiring management for the reactor vessels stainless steel cladding.

3.5 Stainless Steel and Nickel-Based Alloys Exposed to Treated Water Reactor pressure vessel stainless steel and nickel-based alloy subcomponents exposed to treated water include the control rod drive housings, control rod drive stub tubes, incore housings, thermal sleeves, instrumentation nozzles, nozzle safe ends, and various internal attachments. The stainless steel safe ends were buttered with nickel-based alloy weld material prior to welding to the associated nozzle (Ref. 4). Internal attachments are fabricated of stainless steel, nickel-based alloy, or ferritic steel clad with stainless steel. (Ref. 5)

The control rod drive hydraulic return line nozzle was capped and the associated piping removed due to stress corrosion cracking. (Section 4.2.4.1 and Figure 4.2-2 of Ref. 4) The cap for this line is forged austenitic stainless steel.

3.5.1 Loss of Material Stainless steel and nickel-based alloys are inherently immune to general corrosion. Due to physical configuration or small surface defects, system fluid contaminants could concentrate in crevices or pits within the reactor pressure vessel. With a high enough concentration of contaminants in the treated water, the stainless steel and nickel-based alloy internal surfaces may be susceptible to loss of material by pitting and crevice corrosion. Therefore, loss of material (pitting corrosion and crevice corrosion) are aging effects requiring management for stainless steel and nickel-based alloy steel items exposed to treated water.

3.5.2 Cracking - Fatigue Cracking by fatigue is an aging effect that applies to stainless steel and nickel-based alloy items designed in accordance with ASME Section III, Subsection NB. The ASME Design Code requires the calculation of cumulative usage factors (CUF) and the usage factors must be less than one for the period extended operation. Cracking due to thermal fatigue is an aging effect

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 22 of 43 requiring management for the stainless steel and nickel-based alloy Class 1 components since the operating temperature of the valves and the piping exceeds 270 ºF.

Therefore, cracking due to fatigue is an aging effect requiring management for stainless steel and nickel base alloys in a treated water environment. Fatigue (cumulative usage factor assessment) is a time-limited aging analysis (TLAA). For more information on TLAA, see Section 4.2.

3.5.3 Cracking - Other Than Fatigue Cracking at welded joints by growth of fabrication flaws (Ref. 9) due to service loads is an aging effect requiring management for stainless steel and nickel-based alloy subcomponents exposed to treated water.

Cracking from stress corrosion and intergranular attack is an aging effect requiring management for stainless steel and nickel-based alloys. Stainless steel thermal sleeves are susceptible to intergranular stress corrosion cracking (IGSCC). In particular there is industry operating experience of recirculation inlet thermal sleeve cracking.

Therefore, cracking (flaw growth, SCC/IGSCC) is an aging effect requiring management for stainless steel and nickel-based alloy steel items. 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.

3.6 Stainless Steel and Nickel-Based Alloys Exposed to Neutron Fluence 3.6.1 Reduction of Fracture Toughness Reduction in fracture toughness due to radiation embrittlement is an applicable aging effect in the reactor pressure vessel beltline region. However, there are no stainless steel (other than the vessel clad discussed in section 3.3 above) or nickel-based alloy items in the beltline region.

Therefore, reduction of fracture toughness is not an aging effect requiring management for reactor pressure vessel stainless steel and nickel-based alloy subcomponents.

3.7 Stainless Steel and Nickel-Based Alloys Exposed to Air-indoor Reactor pressure vessel stainless steel and nickel-based alloy subcomponents exposed externally to air-indoor include the control rod drive housings, incore housings, instrumentation nozzles, and nozzle safe ends.

3.7.1 Loss of Material Stainless steel and nickel-based alloys exposed to air (primary containment atmosphere) environment are not susceptible to loss of material.

3.7.2 Cracking - Fatigue External exposure to air-indoor does not promote fatigue as the temperature/pressure changes caused by the external environment are small. The temperature/pressure changes associated with the treated water contribute to the fatigue of the vessel subcomponents and are discussed

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 23 of 43 in Section 3.4.2 above. Therefore, cracking due to fatigue is not an aging effect requiring management for stainless steel and nickel-based alloy subcomponents due to external exposure to air-indoor.

3.7.3 Cracking - Other Than Fatigue External exposure to air-indoor does not contribute to cracking of stainless steel or nickel-based alloys. Therefore, cracking is not an aging effect requiring management for stainless steel and nickel-based alloy subcomponents externally exposed to air-indoor.

3.7.4 Reduction of Fracture Toughness External exposure to air-indoor does not contribute to reduction of fracture toughness.

Therefore, reduction of fracture toughness is not an aging effect requiring management for stainless steel and nickel-based alloy subcomponents externally exposed to air-indoor.

3.8 Bolting The reactor pressure vessel closure bolting for the vessel head to shell, the closure bolting for the three upper head nozzle flanges, and the closure bolting for the CRD housing flanges are all constructed of low alloy steel. The closure bolting (bolts, nut, washer, and flange) on the incore detector housings are all stainless steel.

3.8.1 Loss of Material Loss of material is an aging effect requiring management for the reactor vessel closure bolting.

This bolting is exposed to water and steam during refueling operations and subsequent plant heatup.

Loss of material for other pressure boundary bolting is associated with boric acid wastage of low alloy steel bolting. As VYNPS does not normally use boric acid in the reactor coolant system, this is not an aging effect requiring management for the other closure bolting in this AMR.

Stainless steel bolting is not susceptible to loss of material.

External low-alloy steel surfaces exposed to air-indoor are susceptible to general corrosion when the temperature is below 212 °F. This bolting is exposed to these conditions only during plant shutdown, and general corrosion during shutdown is accounted for in the metal corrosion allowance. Metal corrosion allowance is a TLAA, see section 4.2 for more information on TLAA.

3.8.2 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.

3.8.3 Cracking - Other Than Fatigue All of the identified bolting is susceptible to cracking by stress corrosion cracking (SCC) as identified in Appendix F of the Mechanical Tools (Ref. 2).

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 24 of 43 3.8.4 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.

3.9 Operating Experience The review of site-specific and recent industry operating experience, documented in LRPD-05, Operating Experience Review Results (Ref. 7), did not identify any aging effects not addressed in this aging management review report. Cracking in the VYNPS reactor vessel cladding and an indication in one of the reactor vessel shell welds have been found by inservice inspection.

Each is discussed below.

3.9.1 Vessel Cladding A visual inspection of the reactor vessel head, conducted in the 1992 outage as prompted by GE SIL 539, found linear rust indications in the head cladding. Independently, the visual inspection of the vessel found additional linear rust indications. (Section 12 of Appendix A of Ref. 22, Ref. 24) The visual indications prompted additional inspections by ultrasonic testing.

Evaluations concluded that the cracks were due to intergranular stress corrosion of the cladding and that there was no penetration into the base metal. Details can be found in Ref. 24 and 25.

The cracks are being tracked for measurement of growth and are periodically re-inspected as part of the Reactor Vessel Internals Management Program (Ref. 14). These results are consistent with known industry operating experience and the position that clad cracking does not propagate into base metal as discussed in section 3.1.1 of this report.

3.9.2 Reactor Vessel Weld Indication Reactor vessel plate 1-15 had one indication in the 1995 inspection. The indication is located in the plate below weld EF which joins plates 1-12 and 1-15. The weld is outside the core region.

The indication is acceptable for continued service per calculation package YAEC-25Q-301, which was submitted to the NRC by VYNPS letter BVY 96-119 (Ref. 37), and approved by the NRC in their letter of 11 Oct 96 (Ref. 38). For further discussion of this flaw analysis, see LRPD-04, TLAA - Mechanical Fatigue.

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 25 of 43 4.0 Demonstration That Aging Effects will be Managed Section 2.0 described the subcomponents within the reactor pressure vessel that are subject to aging management review. For those subcomponents, 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 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 the reactor pressure vessel subcomponents subject to aging management review and identifies the aging effects requiring management for the material and environment combinations. The following programs, in combination will manage the effects of aging, thereby precluding loss of the intended functions of the reactor pressure vessel. Section 4.1 discusses these programs in detail and provides the clear relationship between the reactor vessel subcomponent, the aging effect and the aging management program actions that 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 aging effects, see VYNPS Report LRPD-02, Aging Management Program Evaluation Results.

Time-limited aging analyses (TLAA) that have been identified for the reactor pressure vessel are described in section 4.2.

4.1 Aging Management Programs 4.1.1 BWR CRD Return Line Nozzle Program The BWR CRD Return Line Nozzle Program is credited with managing cracking for the CRD return line nozzle. This program includes system modification (the line has been removed and the nozzle capped), inservice inspection, and water chemistry control. Monitoring and controlling reactor coolant water chemistry is in accordance with the Water Chemistry Control -

BWR Program. For additional information on the BWR CRD Return Line Nozzle Program, see VYNPS Report LRPD-02, Aging Management Program Evaluation Results.

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 26 of 43 4.1.2 BWR Feedwater Nozzle Program The BWR Feedwater Nozzle Program is credited with managing cracking of the feedwater nozzles. This program augments the ISI program to incorporate the additional inspections recommended in GE-NE-523-A71-0594-A, Revision 1, May 2000, Alternate BWR Feedwater Nozzle Inspection Requirements, as recommended by NUREG-1801 Program XI.M5.

Monitoring and controlling reactor coolant water chemistry is in accordance with the Water Chemistry Control - BWR Program. For additional information on the BWR Feedwater Nozzle Program, see VYNPS Report LRPD-02, Aging Management Program Evaluation Results.

4.1.3 BWR Penetrations Program The BWR Penetrations Program is credited with managing cracking for the SLC/DP and instrumentation nozzles. The program includes inspection and flaw evaluation in accordance with BWRVIP-27-A, BWR Standby Liquid Control System/Core Plate Delta-P Inspection and Flaw Evaluation Guidelines and BWRVIP-49-A, Instrument Penetration Inspection and Flaw Evaluation Guidelines. Monitoring and controlling reactor coolant water chemistry is in accordance with the Water Chemistry Control - BWR Program. For additional information on the BWR Penetrations Program, see VYNPS Report LRPD-02, Aging Management Program Evaluation Results.

4.1.4 BWR Stress Corrosion Cracking Program The BWR Stress Corrosion Cracking Program is credited with managing cracking (SCC) of BWR stainless steel piping (safe ends) of greater than or equal to 4 inches nominal diameter.

This program includes the inspection recommendations of BWRVIP-75-A as accepted by the NRC (Ref. 48). Monitoring and controlling reactor coolant water chemistry is 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.

4.1.5 BWR Vessel ID Attachment Weld Program The BWR Vessel ID Attachment Weld Program is credited with managing cracking for the pressure vessel internal attachment welds. This program includes inspection and flaw evaluation per the guidelines of BWRVIP-48-A (approved by the NRC), Vessel ID Attachment Weld Inspection and Evaluation Guidelines and per ASME Section XI. Monitoring and controlling reactor coolant water chemistry is in accordance with the Water Chemistry Control -

BWR Program. For additional information on the BWR Vessel ID Attachment Welds Program, see VYNPS Report LRPD-02, Aging Management Program Evaluation Results.

4.1.6 BWR Vessel Internals Program The BWR Vessel Internals Program is designed to manage cracking of the reactor vessel internals and includes inspection and flaw evaluation in accordance with the BWRVIP documents. The following subcomponents of the reactor pressure vessel are managed by the BWR Vessel Internals Program: the CRD housings, the CRD stub tubes, core spray nozzle thermal sleeves, recirculation inlet nozzle thermal sleeves, and the SLC nozzle to safe end weld. The BWR Vessel Internals Program is credited with managing cracking for these stainless steel and nickel-based alloy components. (Section 19 of Appendix A of Ref. 14)

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 27 of 43 For more details on the reactor vessel internals, see VYNPS Report AMRM-32, Aging Management Review of the Reactor Vessel Internals. For additional information on the BWR Vessel Internals Program, see VYNPS Report LRPD-02, Aging Management Program Evaluation Results.

4.1.7 Inservice Inspection Program The reactor pressure vessel is included in the VYNPS ASME Section XI, Subsections IWB, IWC, and IWD, Inservice Inspection (ISI) Program (Ref. 17). The VYNPS ISI Program is a nondestructive examination program that provides for the implementation of ASME Code,Section XI, 1998 Edition, 2000 Addenda Subsection IWB, in accordance with the provisions of 10 CFR 50.55a. The VYNPS ISI program has been augmented to include recommended inspections from various BWRVIP documents and other industry documents that are reflected in Section XI of NUREG-1801.

The Inservice Inspection Program is credited for managing cracking and loss of material for numerous subcomponents of the reactor pressure vessel. All components crediting ISI are listed in Attachment 1.

For additional information on the Inservice Inspection Program, see VYNPS Report LRPD-02, Aging Management Program Evaluation Results.

4.1.8 Reactor Head Closure Studs Program The Reactor Head Closure Stud Program manages cracking and loss of material of the reactor head closure studs. This program includes a combination of nondestructive examination and vessel bolting/unbolting procedures. The program includes preventive actions and inspection techniques for BWRs. For additional information on the Reactor Head Closure Studs Program, see VYNPS Report LRPD-02, Aging Management Program Evaluation Results.

4.1.9 Reactor Vessel Surveillance Program This program monitors changes in the fracture toughness of ferritic materials in the reactor pressure vessel beltline region caused by exposure to neutron radiation. This monitoring is accomplished through surveillance material testing in accordance with BWRVIP guidelines and in accordance with Table 4.2.4 of the VYNPS UFSAR (Ref. 4). Fracture toughness test data are obtained from material specimens exposed in surveillance capsules, which are withdrawn periodically from the reactor pressure vessel. This program is based on programs documented in the following industry topical reports BWRVIP-86A, BWR Integrated Surveillance Program (Ref. 28)

BWRVIP-116, Integrated Surveillance Program Implementation for License Renewal (Ref. 29)

The Reactor Vessel Surveillance Program is credited with managing reduction of fracture toughness (radiation embrittlement) of the vessel beltline shell and welds.

For additional information on the Reactor Vessel Surveillance Program, see VYNPS Report LRPD-02, Aging Management Program Evaluation Results.

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 28 of 43 4.1.10 Water Chemistry Control - BWR Program Class 1 subcomponents exposed to treated water that are subject to aging management review include carbon steel subcomponents, wrought and forged stainless steel subcomponents, nickel-based alloy subcomponents, and welds. The Water Chemistry Control - BWR Program will manage the following aging effects.

Loss of material due to general corrosion (carbon and low alloy steel)

Loss of material due to crevice corrosion and pitting corrosion (all materials)

Cracking by SCC and intergranular attack (IGA) (stainless steel and nickel-based alloys)

The VYNPS Water Chemistry Control - BWR Program optimizes the primary water chemistry to minimize the potential for loss of material and cracking. For additional information on the Water Chemistry Control - BWR program, see VYNPS Report LRPD-02, Aging Management Program Evaluation Results.

4.2 Time-Limited Aging Analyses The following time-limited aging analyses (TLAA) have been identified for the reactor pressure vessel.

The reactor pressure vessel is exposed to elevated temperatures, thermal cycling, and the associated metal thermal fatigue. The evaluation of metal fatigue is a TLAA.

Corrosion allowances are provided for all exposed surfaces of carbon and low alloy steels. The evaluation of the corrosion allowance for the period of extended operation is a TLAA.

The reactor pressure vessel beltline receives neutron radiation in excess of 1x1017 n/cm2 and as such is subject to reduction in fracture toughness. The evaluation of reduction of fracture toughness due to radiation embrittlement is a TLAA.

For additional information, refer to VYNPS Report LRPD-03, TLAA and Exemption Evaluation Results for the evaluation of TLAA and to LRPD-04, TLAA - Mechanical Fatigue for the evaluation of the metal fatigue for the period of extended operation.

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 29 of 43 5.0 Summary and Conclusions The following aging management programs address the aging effects requiring management for the reactor pressure vessel.

VYNPS Program ASME Section XI Inservice Inspection Water Chemistry Control - BWR Reactor Head Closure Studs BWR Vessel ID Attachment Welds BWR Feedwater Nozzles BWR Control Rod Drive Return Line Nozzle BWR Stress Corrosion Cracking BWR Penetrations BWR Vessel Internals Reactor Vessel Surveillance For additional review of the programs credited for the 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 pressure vessel.

In conclusion, programs described in Section 4.0 will provide reasonable assurance that the effects of aging on the VYNPS reactor pressure vessel 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 Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 30 of 43 6.0 References

1.

NUREG 1801, Vol. 1, Generic Aging Lessons Learned (GALL) Report, U.S.

Nuclear Regulatory Commission, April 2001

2.

EPRI report 1003056, Non-Class 1 Mechanical Implementation Guideline and Mechanical Tools, Revision 3, November 2001 (a.k.a. the Mechanical Tools)

3.

ENN-MS-S-009-VY, Vermont Yankee Site Specific Guidance and System Safety Function Sheets, Revision 0, 03/22/2005

4.

VYNPS Updated Final Safety Analysis Report, Revision 17

5.

BWRVIP-74, (EPRI report TR-113596), Final Report, September, 1999, BWR Vessel and Internals Project BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines (BWRVIP-74)

6.

BWRVIP-05, (EPRI report TR-105697), September, 1995, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)

7.

LRPD-05, Operating Experience Review Results

8.

BWRVIP-79, BWR Water Chemistry Guidelines2000 Revision, EPRI Report TR-103515-R2, Final Report, February 2000

9.

Nondestructive Examination Standards, Technical Basis and Development of Boiler and Pressure Vessel Code, ASME Section XI, Division 1, EPRI-NP-1406-SR, May 1980

10. Code of Federal Regulations, Title 10, Part 50, Appendix G: Fracture Toughness Requirements, December 19, 1995
11. NUREG 1803, Safety Evaluation Report Related to the License Renewal of the Edwin I Hatch Nuclear Plant, Units 1 and 2, December, 2001
12. LRPG-01, License Renewal Project Plan
13. Licensed Operator Training Program Student Handout, LOT-00-299H, Rev. 0, September, 2001
14. VY Procedure PP7027, Rev 2, 6 April 2004, Reactor Vessel Internals Management Program
15. VY Procedure OP4612, Sampling and Treatment of the Reactor Water System, Rev 23, September, 2000
16. VY Procedure OP 2199, Hydrogen Water Chemistry System, Rev 1, September 2003
17. VY Procedure PP 7015, Rev. 3, September 2003, Vermont Yankee Inservice Inspection Program
18. BWRVIP-47 (EPRI Report, TR-108727), BWR Lower Plenum Inspection and Flaw Evaluation Guidelines December, 1997
19. GEK-9608, General Electric Report Operation and Maintenance Instructions:

Reactor Assembly for Vermont Yankee Nuclear Power Station, December 1970

20. AMRM-32, Aging Management Review of the Reactor Vessel Internals

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 31 of 43

21. AMRM-33, Aging Management Review of the Reactor Coolant System Pressure Boundary
22. VY Procedure NE 8067, Revision 2, 4 April 2004, Reactor Vessel Internals Inspection Details
23. Piping Specification, VYNP-V1-III-P-1, November 13, 1967; Ebasco Specification 62-65T, General Power Piping
24. Corrective Action Report CAR 92-015, 5/1/92, Reactor Vessel Crack Adjacent to Dryer Support Bracket
25. BVY 92-56, 4/10/92, L.A. Tremblay, Jr. to USNRC Document Control Desk, Supplemental Information Regarding Proposed Alternatives for Compliance with 10CFR50.55a Regarding RPV Cladding Indications
26. NVY-88-080, Core Spray Safe-end Inspection (TAC No.67522), May 9, 1988, V.L. Rooney, NRC to R. W. Capstick, VYNPC
27. FVY 88-19, March 1, 1988, W. P. Murphy to United States Nuclear Regulatory Commission Document Control Desk, Long-Term Operation with Core Spray Safe End Nozzle Weld Overlays
28. BWRVIP-86, (EPRI report 1000888), Final Report, December 2000, BWRVOP-86: BWR Vessel and Internals Project, BWR Integrated Surveillance Program Implementation Plan
29. BWRVIP-116, (EPRI report 1007824), Final Report, July 2003, BWRVIP-116:

BWR Vessel and Internals Project Integrated Surveillance Program (ISP)

Implementation For License Renewal

30. BWR Vessel and Internals Project, Evaluation of Stress Corrosion Crack Growth in Low Alloy Steel Vessel Materials in the BWR Environment (BWRVIP-60), EPRI Report TR-108709, March 1999.
31. General Electric Specification 21A1115, Reactor Pressure Vessel, 7/14/69
32. CURATOR entry CB&I STRESS CONTRACT #9-6201-I, titled CB&I Stress Contract Pressure Vessel Record Exhibit D Certified Test Reports, date written 11-25-68 Boiling Water Nuclear Reactor Vessel 17.167 x 63.167 Ins. Hds.

Manufacturers Serial No. B-4698 Vermont Yankee Project, Vernon, VT.

G.E. Co. P.O. 205-55565-I

33. GE-NE-0000-0007-2342-R1-NP, Rev. 1, July 2003, Entergy Northeast Vermont Yankee Neutron Flux Evaluation
34. BVY 03-80, September 10, 2003, J.K. Thayer to USNRC Document Control Desk, Technical Specification Proposed Change No 263, Extended Power Uprate - Attachment 4, NEDC-33090P, Rev 0, September 2003, Safety Analysis Report for the Vermont Yankee Nuclear Power Station Constant Pressure Power Uprate
35. Code of Federal Regulations, Title 10, Part 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, December 31, 2003

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 32 of 43

36. Drawing 5920-624, 8 Q nozzles MK N5A/B, updated 5/7/1969
37. BVY 96-119, 9 October 1996, J. J. Duffy to USNRC Document Control Desk, Reactor Pressure Vessel Inspection at Vermont Yankee
38. NRC Letter dated 11 October 1996, C. C. Harbuck to D. A. Reid, Evaluation of Flaw Indication found during reactor pressure vessel inspections at Vermont Yankee Nuclear Power Station (TAC No. M96670)
39. BWRVIP-15, EPRI Report TR-106368, Final Report, March 1996, Configurations of Safety-Related BWR Reactor Internals
40. Drawing 5920-483
41. Drawing 5920-252, Revision 7, Shroud Support
42. Drawing 5920-243 Revision 5, 6 in dia 1500 Nozzles Mk N6A & N6B
43. Drawing 5920-244, Revision 6 4 dia Vent Nozzle Mk N7 2 dia Drain Nozzle MK N15
44. Drawing 5920-324, Revision 3, Shell flange
45. Drawing 5920-329, Revision 2, Steam Dryer Support Bracket
46. Drawing 5920-330, Revision 2, Feedwater Sparger Bracket
47. AMRC-06, Aging Management Review of Bulk Commodities, Revision 0
48. BWRVIP-75, (EPRI Report TR-113932, Feb 29, 2000) Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules and NUREG-0313, Initial Safety Evaluation Report by the Office of Nuclear Reactor Regulation for BWRVIP-75,September 15, 2000
49. Drawing 5920-0495, Revision 9, July, 1999, SRM/IRM Unit (GE dwg 729E946)
50. Drawing 5920-241, Sheet 1, Rev. 4, Detail of stub, N4A thru N4B, October, 2001

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 33 of 43

Aging Management Review Results Reactor Pressure Vessel Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Programs Bolting Closure flange studs, nuts, washers and bushings Pressure boundary Low alloy steel Air-indoor (ext)

Loss of material Reactor head closure studs A540 Grade B3 Cracking - fatigue TLAA - metal fatigue Cracking Reactor head closure studs Incore housing bolting Pressure boundary Stainless steel Air-indoor (ext)

Cracking - fatigue TLAA - metal fatigue flange bolts A182 Grade F3042 flange Cracking Inservice Inspection nut and washer Other pressure boundary bolting Pressure boundary Low alloy steel Air-indoor (ext)

Cracking - fatigue TLAA - metal fatigue flange bolts and nuts (N6A, N6B, N7)

Bolts: SA193 Grade B710 Nuts: SA194 Grade 2H2 Cracking Inservice Inspection CRD flange capscrews and washers SA193 Grade B711

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 34 of 43 ATTACHMENT 1: Aging Management Review Results Reactor Pressure Vessel Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Programs Heads and Shell Dome - bottom head Pressure boundary Low alloy steel with SS cladding Treated water

>220ºF (int)

Loss of material Water chemistry control - BWR A533 Grade B2,3 Cracking - fatigue TLAA - metal fatigue Cracking Inservice inspection Water chemistry control - BWR Air-indoor (ext)

None None Dome - upper head Pressure boundary Low alloy steel with SS cladding Treated water

>220ºF (int)

Loss of material Water chemistry control - BWR A533 Grade B2,3 Cracking - fatigue TLAA - metal fatigue Cracking Inservice inspection Water chemistry control - BWR Air-indoor (ext)

None None Flanges (closure)

Pressure boundary Low alloy steel with SS cladding Treated water

>220ºF (int)

Loss of material Water chemistry control - BWR upper head SA508 Class 22,3,4 Cracking - fatigue TLAA - metal fatigue vessel shell Cracking Inservice inspection Water chemistry control - BWR Air-indoor (ext)

None None Reactor vessel shell Pressure boundary Low alloy steel with SS cladding Treated water

>220ºF (int)

Loss of material Water chemistry control - BWR Upper Shell Cracking - fatigue TLAA - metal fatigue Intermediate nozzle shell A533 Grade B2,3 Cracking Inservice inspection Water chemistry control - BWR Lower shell Air-indoor (ext)

None None

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 35 of 43 ATTACHMENT 1: Aging Management Review Results Reactor Pressure Vessel Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Programs Reactor vessel shell Pressure boundary Low alloy steel with SS cladding Treated water

>220ºF (int)

Loss of material Water chemistry control - BWR Intermediate beltline shell A533 Grade B2,3 Cracking - fatigue TLAA - metal fatigue Cracking Inservice inspection Water chemistry control Neutron fluence Reduction in fracture toughness Reactor vessel surveillance TLAA - neutron fluence Air-indoor (ext)

None None Nozzles and Penetrations CRD housings Pressure boundary Stainless steel Treated water

>270ºF (int)

Loss of material Inservice inspection Water chemistry control - BWR TP3049 Cracking - fatigue TLAA - metal fatigue Cracking BWR Vessel Internals Water chemistry control - BWR Air-indoor (ext)

None None CRD stub tubes Pressure boundary Nickel-based alloy Treated water

>270ºF (int)

Loss of material Inservice inspection Water chemistry control - BWR SB167 cc 13369 Cracking - fatigue TLAA - metal fatigue Cracking BWR Vessel Internals Water chemistry control - BWR

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 36 of 43 ATTACHMENT 1: Aging Management Review Results Reactor Pressure Vessel Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Programs In-core housings Pressure boundary Stainless steel Treated water

>270ºF (int)

Loss of material Inservice inspection Water chemistry control - BWR TP304, TP3169 Cracking - fatigue TLAA - metal fatigue Cracking Inservice inspection Water chemistry control - BWR Air-indoor (ext)

None None Nozzles Pressure boundary Low alloy steel with SS cladding Treated water

>220ºF (int)

Loss of material Water chemistry control - BWR Recirc outlets (N1)

SA508 Class 24 Cracking - fatigue TLAA - metal fatigue Recirc inlets (N2)

Cracking Inservice inspection Water chemistry control - BWR Air-indoor (ext)

None None Nozzles Pressure boundary Low alloy steel with partial SS cladding Treated water

>220ºF (int)

Loss of material Inservice inspection Water chemistry control - BWR Main steam (N3)

SA508 Class 24 Cracking - fatigue TLAA - metal fatigue Cracking Inservice inspection Water chemistry control - BWR Air-indoor (ext)

None None Nozzles Pressure boundary Low alloy steel with partial SS cladding Treated water

>220ºF (int)

Loss of material Inservice inspection Water chemistry control - BWR Feedwater (N4)

SA508 Class 24 Cracking - fatigue TLAA - metal fatigue Cracking BWR feedwater nozzle Air-indoor (ext)

None None

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 37 of 43 ATTACHMENT 1: Aging Management Review Results Reactor Pressure Vessel Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Programs Nozzles Pressure boundary Low alloy steel with SS cladding Treated water

>220ºF (int)

Loss of material Water chemistry control - BWR Core Spray (N5)

SA508 Class 24 Cracking - fatigue TLAA - metal fatigue Head Spray (N6A)

Cracking Inservice inspection Head Instr. (N6B)

Water chemistry control - BWR Head vent (N7)

Jet pump inst. (N8)

Air-indoor (ext)

None None Nozzle Pressure boundary Low alloy steel with SS cladding Treated water

>220ºF (int)

Loss of material Water chemistry control - BWR CRD return (N9)

SA508 Class 24 Cracking - fatigue TLAA - metal fatigue Cracking BWR CRD return line nozzle Air-indoor (ext)

None None Nozzle Pressure boundary Low alloy steel with SS cladding Treated water

>220ºF (int)

Loss of material Water chemistry control - BWR Core DP/SLC (N10)

SA508 Class 24 Cracking - fatigue TLAA - metal fatigue Cracking BWR penetrations Water chemistry control - BWR Air-indoor (ext)

None None Nozzles Pressure boundary Nickel-based alloy Treated water

>270ºF (int)

Loss of material Inservice inspection Water chemistry control - BWR

-Instrumentation (N11, N12)

SB-1664 Cracking - fatigue TLAA - metal fatigue Cracking BWR penetrations Water chemistry control - BWR Air-indoor (ext)

None None

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 38 of 43 ATTACHMENT 1: Aging Management Review Results Reactor Pressure Vessel Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Programs Nozzles Pressure boundary Nickel-based alloy Treated water

>270ºF (int)

Loss of material Inservice inspection Water chemistry control - BWR Flange leakoff (N13, N14)

SB1664 Cracking - fatigue TLAA Cracking Inservice inspection Water chemistry control - BWR Air-indoor (ext)

None None Nozzles Pressure boundary Low alloy steel with partial SS cladding Treated water

>220ºF (int)

Loss of material Inservice inspection Water chemistry control - BWR Drain (N15)

SA508 Class 14 Cracking - fatigue TLAA - metal fatigue Cracking Water chemistry control - BWR Inservice inspection Air-indoor (ext)

None None Safe Ends, Thermal Sleeves, Flanges, Caps CAP Pressure boundary Stainless steel Treated water

>270ºF (int)

Loss of material Inservice inspection Water chemistry control - BWR CRD return line (N9)

SA182 Grade 316L7 Cracking - fatigue TLAA - metal fatigue Cracking BWR CRD return line nozzle Water chemistry control - BWR Air-indoor (ext)

None None

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 39 of 43 ATTACHMENT 1: Aging Management Review Results Reactor Pressure Vessel Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Programs Flanges Pressure boundary Stainless steel A182 Grade F3042 Treated water

>270ºF (int)

Loss of material Inservice inspection Water chemistry control - BWR head nozzle flanges (N6, N7)

Cracking - fatigue TLAA - metal fatigue blank flanges (N6)

Cracking Inservice inspection Water chemistry control - BWR Air-indoor (ext)

None NA Safe ends >=4 Pressure boundary Stainless steel Treated water

>270ºF (int)

Loss of material Inservice Inspection Water chemistry control - BWR recirc outlet (N1)

A182 F3163 Cracking - fatigue TLAA - metal fatigue recirc inlet (N2)

Cracking BWR stress corrosion cracking Water chemistry control - BWR Air-indoor (ext)

None None Safe ends >=4 Pressure boundary Carbon steel Treated water

>220ºF (int)

Loss of material Inservice Inspection Water chemistry control - BWR Main steam (N3)

SA516 Grade 706 Cracking - fatigue TLAA - metal fatigue Feedwater (N4)

ASTM A508 Class 112 Air-indoor (ext)

None None Safe ends >=4 Pressure boundary Nickel-based alloy Treated water

>270ºF (int)

Loss of material Inservice Inspection Water chemistry control - BWR Core spray (N5)

MS16 (Inconel)5 Cracking - fatigue TLAA - metal fatigue Cracking BWR stress corrosion cracking Water chemistry control - BWR Air-indoor (ext)

None None

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 40 of 43 ATTACHMENT 1: Aging Management Review Results Reactor Pressure Vessel Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Programs Safe ends >=4 Pressure boundary Stainless steel Treated water

>270ºF (int)

Loss of material Inservice Inspection Water chemistry control - BWR Jet pump instrument (N8)

A336 Cl F83 Cracking - fatigue TLAA - metal fatigue Cracking BWR stress corrosion cracking Water chemistry control - BWR Air-indoor (ext)

None None Safe ends <4 Pressure boundary Stainless steel Treated water

>270ºF (int)

Loss of material Inservice Inspection Water chemistry control - BWR Core DP / SLC (N10)

A336 Cl F82 Cracking - fatigue TLAA - metal fatigue Instrumentation (N11, N12)

A479 TP316(SE) 2 Cracking Inservice inspection Water chemistry control - BWR Air-indoor (ext)

None None Safe ends <=4 Pressure boundary Low alloy steel Treated water

>220ºF (int)

Loss of material Inservice Inspection Water chemistry control - BWR Drain (N15)

ASTM A508-113 Cracking - fatigue TLAA - metal fatigue Air-indoor (ext)

None None Thermal sleeves Pressure boundary Stainless steel Treated water

>270ºF (int)

Loss of material Water chemistry control - BWR Recirc inlet (N2)

Type 3047,8 Cracking - fatigue TLAA - metal fatigue Core spray (N5)

Cracking BWR vessel internals Water chemistry control - BWR

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 41 of 43 ATTACHMENT 1: Aging Management Review Results Reactor Pressure Vessel Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Programs Thermal sleeves Pressure boundary Stainless steel and Nickel-based alloy Treated water

>270ºF (int)

Loss of material Water chemistry control - BWR Feedwater (N4)

SA 312 Gr 304 Alloy 6007 Cracking - fatigue TLAA - metal fatigue Cracking Inservice inspection Water chemistry control - BWR Weld Pressure boundary Nickel-based alloy3 Treated water

>270ºF (int)

Loss of material BWR vessel internals Water chemistry control - BWR N10 - SLC nozzle to safe end weld Cracking - fatigue TLAA - metal fatigue Cracking BWR Penetrations Water chemistry control - BWR Air-indoor (ext)

None None

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 42 of 43 ATTACHMENT 1: Aging Management Review Results Reactor Pressure Vessel Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Programs Vessel Attachments and Supports Internal attachments Pressure boundary Stainless steel &

Nickel-based alloy Treated water

>270ºF (int)

Loss of material Inservice inspection Water chemistry control - BWR Shroud support ring pad (1)

Alloy 1821 Cracking - fatigue TLAA - metal fatigue Shroud support feet (14)

Alloy 6001 Cracking BWR vessel ID attachment welds Jet pump riser pads (20)

E308L1 Water chemistry control - BWR Core spray brackets (4)

E308L1 Guide rod brackets (2)

E308L1 Steam dryer brackets (4)

SA240 Type 304 Dryer holddown brackets (4)

SS, Type 304 Surveillance specimen holder brackets SS, Type 304 Feedwater sparger brackets (8)

SA240 Type 304 Supports Stabilizer pads Support skirt Support for Criterion (a)(1) equipment Low alloy steel A533 Grade B2 A533 Grade B3 Air-indoor (ext)

Loss of material Inservice inspection Cracking - fatigue TLAA - metal fatigue

VYNPS License Renewal Project Aging Management Review of the Reactor Pressure Vessel AMRM-31 Revision 0 Page 43 of 43 ATTACHMENT 1: Aging Management Review Results Reactor Pressure Vessel Footnotes to Attachment 1:

1 NE 8067, App A, Sec. 19.1 (Ref. 22) 2 GE Spec 21A1115, (Ref. 31) 3 PP7015, Appendix C, table 5 (Ref. 17) 4 Certified Test Reports (Ref. 32) 5 Drawing 5920-624 (Ref. 36) 6 Drawing 5920-483 (Ref. 40) 7 UFSAR Table 4.2-1 8

BWRVIP-75 (Ref. 5) 9 GEK-9608 (Ref. 32) 10 VYNP-V1-III-P-1, (Ref. 23) 11 UFSAR, page 3.4-27 12 Customer Review.

13 Drawing 5920-244 (Ref. 43)

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