PNP 2013-022, Response to Request for Additional Information - Revised Program Plan for Aging Management of Reactor Vessel Internals

Response to Request for Additional Information - Revised Program Plan for Aging Management of Reactor Vessel Internals

ML13094A414
Person / Time
Site:	Palisades
Issue date:	04/04/2013
From:	Entergy Nuclear Operations
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
PNP 2013-022, TAC ME9569
Download: ML13094A414 (15)

Text

Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043 Tel 269 764 2000 Otto W Gustafson Licensing Manager PNP 2013-022 April 04, 2013 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Response to Request for Additional Information - Revised Program Plan for Aging Management of Reactor Vessel Internals Palisades Nuclear Plant Docket 50-255 License No. DPR-20

References:

1. Entergy Nuclear Operations, Inc. letter to NRC, "Revised Program Plan for Aging Management of Reactor Vessel Internals," September 13, 2012 (ADAMS Accession Number ML12257A352)

2. NRC e-mail, "Palisades - Request for Additional Information - ME9569

- Revised Program Plan for Aging Management of Reactor Vessel Internals," March 4, 2013 (ADAMS Accession Number ML13063A318)

Dear Sir or Madam:

Entergy Nuclear Operations, Inc. (ENO) submitted Reference 1 to the Nuclear Regulatory Commission (NRC) which provided the revised program plan for aging management of reactor vessel internals. ENO received an electronic request for additional information (RAI) from the NRC in Reference 2. ENO and the NRC held a conference call, on March 20, 2013, to clarify the RAI.

Attached is the ENO response to the RAI.

This letter contains no new or revised commitments.

Sincerely, owg/jpm

PNP 2013-022 Page 2 of2 : RAI Response on Revised Program Plan for Aging Management of Reactor Vessel Internals cc: Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC

Attachment 1 RAI Response on Revised Program Plan for Aging Management of Reactor Vessel Internals Request for additional information received by electronic mail dated March 4. 2013 Nuclear Regulatory Commission (NRC) Request

1. Table IV of NUREG-1801, "Generic Aging Lessons Learned (GALL) Report,"

Volume 2, Revision 1, identifies aging effects for some of the RVI components that were designed by Combustion Engineering. Aging effects that are pertinent to some of the RVI components that are considered part of RVI AMP are addressed in GALL Table IV 83. The following table provides information related to the aging effects which were not included in the AMP for the RVI components listed in GALL Table IV 83.

Aging Effect RVI Component GALL Report-Table ID number Loss of fracture Core support IV.B3-16, 17 toughness/ neutron barrel upper irradiation embrittlement, flange void swelling, loss of material Loss of fracture Core support IV.83-18 toughness/thermal aging column and neutron irradiation embrittlement Provide a supplement to your AMP that addresses these aging effects for these components.

Entergy Nuclear Operations, Inc (ENO) Response

1. In accordance with the guidance provided in MRP-227-A, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection And Evaluation Guidelines," the upper core barrel flange (core support barrel upper flange) is listed in Table 3 of the Palisades Reactor Vessel Internals Aging Management Program as an expansion component. The nuclear regulatory commission (NRC) approved version of MRP-227-A identifies the aging mechanisms associated with the core support barrel upper flange to be stress-corrosion-cracking (SCC) of the welds and wear for the flange.

NUREG-1801, Volume 2, Revision 1, Item IV.B3-16 identifies the core support barrel upper flange as a component of interest, but this was later screened out during the scoping and screening process for MRP-227-A. Specifically, Page 1 of 13

Attachment 1 RAI Response on Revised Program Plan for Aging Management of Reactor Vessel Internals MRP-191, "Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design," used a screening criterion for loss of fracture toughness due to neutron irradiation in austenitic stainless steels of 1x1 021 n/cm 2 . For void swelling in stainless steels, a screening criterion of 1.3x1022 n/cm 2 was used. In MRP-191 the fluence in the upper core barrel flange region for Combustion Engineering (C-E) designed plants was assumed to be less than 1x1020 n/cm 2 and therefore below the screening criteria for susceptibility. PNP projected fluence for the core support barrel upper flange will remain below 1x1020 n/cm 2 at the end of the license extension (EOLE) period. This is based on the Westinghouse letter, LTR-RIAM-13-25, "Palisades Core Barrel Upper Flange Fluence," statin~ the fluence in this region is less than the MRP-191 screening criteria of 1x1 0 2 n/cm 2 [Reference 5]. This validates that this component will remain below the fluence limits for susceptibility due to the effects of loss of fracture toughness/neutron irradiation embrittlement and void swelling.

Therefore, this is correctly identified to be an expansion component and no further changes are needed to the AMP or the inspection plan.

Similarly, based on the guidance provided in MRP-227-A, core support column welds have been identified as a primary component in Table 2, "C-E Plants Primary Category Components from MRP-227-A [3b]," of the Palisades Reactor Vessel Internals Aging Management Program with the relevant degradation mechanisms identified as cracking (stress corrosion cracking (SCC),

irradiation-assisted stress corrosion cracking (lASCC), and fatigue) including damage or fractured material with aging management required for irradiation embrittlement and thermal embrittlement. NUREG-1801, Volume 2, Revision 1, identifies the aging management program for item IV.B3-18 (lower internal assembly-core support column) as Chapter XI.M13, 'Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS)." This is not applicable to PNP as CASS materials are not present as part of the reactor vessel internals, or specifically, the support columns [Reference 2].

NRC Requesf

2. The inspection plan contains numerous examples where the licensee intends to implement visual testing (VT-3) examinations to identify cracking in some pressurized water reactor (PWR) RVI components. Historically, enhanced visual testing (EVT-1) or ultrasonic testing (UT) methods are used to effectively identify cracks. Explain why the use of a VT-3 inspection method should be considered acceptable for identifying cracking in some PWR RVI components.

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Attachment 1 RAI Response on Revised Program Plan for Aging Management of Reactor Vessel Internals ENO Response

2. The VT-3 examination was specified to identify general degradation in the aged components. In none of the cases where VT-3 is specified is the examination objective considered to be the detection of the onset of cracking with accompanying tight crack opening displacements (COOs). In the cases where cracking associated with IASCC/SCC is the anticipated mechanism or one of the mechanisms postulated and VT-3 is specified, the objective of the examination is detection of:

• Broken or missing bolt locking devices and welds

• Protruding bolts

• Broken or missing pieces (e.g., supports, spider arms, dowels)

In these cases, VT-3 as currently defined in Section XI, Table IWA-2211-1, "Visual Examinations," of the American Society of Mechanical Engineering (ASME) Code is quite capable of this level of detection (0.105" character height resolution). This practice was deemed appropriate for the existing examination requirements listed in MRP-227-A, Tables 4-8 and 4-9.

As used in MRP-227-A, VT-3 is consistent with Section XI, Subsection IWA-2213, VT-3 Examination, rules for component supports looking for:

• Missing bolts

• Gross degradation

• Misalignment

• General structural condition Table RAI-2-2 in MRP-227-A (Table 1 below), [Reference 9], provides a summary of primary and expansion components for the C-E design where VT-3 is specified for cracking. There are only two C-E components in this category:

• Instrument guide tubes attached to the upper control element assembly (CEA) structure and the welds in the lower support structure. The inspection for this component specifies that the concern is cracking that results in missing supports or separation at the welded joint between the tubes and the supports

• Core support column welds. The inspection for this component specifies the concern as' damaged or fractured material.

VT-3 examinations for cracking were not specified in cases where the data would potentially be used in a fracture mechanics analysis to demonstrate the structural Page 3 of 13

Attachment 1 RAI Response on Revised Program Plan for Aging Management of Reactor Vessel Internals integrity of the vessel internals. The more sensitive crack detection capabilities of the UT and EVT-1 examinations have been specified for other components where it was determined that early detection and protection against fracture was critical. However, even in these cases, the example calculations performed in MRP-210, "Materials Reliability Program: Fracture Toughness Evaluation of Highly Irradiated PWR Stainless Steel Internal Components," demonstrate that the internals structures are extremely flaw-tolerant. An example is the critical flaw sizes for the postulated stresses for a through-wall edge crack in a flat plate (Figure 3-18 and Table 3-3 in MRP-210). Thus, a VT-3 examination is capable of identifying subcritical crack growth well before a crack would become critical. For justification, see for example the final paragraph of Section 3.2.3.1 of MRP-231, Revision 2, "Materials Reliability Program: Aging Management Strategies for B&W Pressurized Water Reactor Internals," [Reference 7].

The MRP-227-A inspection recommendations for VT-3 to monitor the effects of cracking are appropriate because they are limited to cases where the intent of the examination is to monitor the general condition of the component. These recommendations are consistent with the approach used in the ASME Section XI examinations, which require VT-3 inspections for accessible core support structures. The practice is also consistent with the approach used in boiling water reactor applications.

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Attachment 1 RAI Response on Revised Program Plan for Aging Management of Reactor Vessel Internals Table 1: Primary and Expansion Components from C-E Designed Plants with Cracking Identified as an Effect and VT-3 Examinations Specified

[Reference 9]

Item Effect (Mechanism) Examinatlo n Examinatlo n Coverage Method/Frequenc y Primary Cracking (SCC, fatigue) that Visual (VT-3) examination, no later 100% oftubes in peripheral CEA results in missing supports or than 2 refueli ng outages from the shroud assemblies (i.e. , those separation at the welded joint beginning of the license renewal adjacent to the perimeter of the Control Element between the tubes and period. Subsequent examination fuel alignment plate).

Assembly supports. on a ten-year Interval.

Instrument guide tubes See Figure 4-18 of MRP-227-A.

Plant-specific component integrity assessments may be required if degradation is detected and remedial action is needed.

Primary Cracking (SCC, IASCC, Visual (VT-3) examination no later 100% of the accessible surfaces of fatigue Including damaged or than 2 refueling outages from the the core support column welds.

fractured material) beginning of the license renewal Lower Support Structure period.

Core support column See Rgure 11 of Aging welds Management [Reference 1)

Subsequent examinations on a ten-year interval.

ExpanSion Cracking (SCC, fatigue) that Visual (VT-3) examination with 100% of tubes in CEA shroud results in missing supports or initial and subsequent assemblies.

separation at the welded jOint examinations dependent on the Control Element between the tubes and results of the instrument guide Assembly supports. tube examinations. See Figure 4-18 of MRP-227-A.

Remaining instrument guide tubes Page 5 of 13

Attachment 1 RAI Response on Revised Program Plan for Aging Management of Reactor Vessel Internals The following table outlines the specific situations where the Palisades AMP identifies visual testing (VT-3) of reactor internal components under MRP-227-A.

Table 2: Reactor Internal Components with VT-3 Examination from Palisades Reactor Vessel Internals Aging Management Program

[Reference 1]

Item, Description Reference [6] Reference [6] Relevant Cond ition Location Core Shroud Assembly (Bolted The specific relevant conditions Page 73, Table 2. C-E Plants 1 Assembly) are evidence of abnormal Primary Category Components interaction with fuel assemblies, from MRP-227-A gaps along high fluence shroud plate joints, and vertical displacement of shroud plates near high fluence jOints.

lower Support Structure Core The specific relevant condition Page 74, Table 2. C-E Plants 2 Support Column Welds is missing or separated welds . Primary Category Components from MRP-227-A Control Element Assembly, The specific relevant conditions Page 76, Table 2. C-E Plants 3 Instrument Guide Tubes are missing supports and Primary Category Components separation at the welded joint from MRP-227-A between the tubes and the Sllpports.

Control Element Assembly, The specific relevant conditions Page 79, Table 3. C-E Plants 4 Remaining Instrument Guide are missing supports and Expansion Category Tubes separation at the welded joint Components from MRP-227-A between the tubes and the supports.

Core Shroud Assembly Guide The specific relevant condition Page 80, Table 4. C-E Plants 5 lug Inserts and Bolts is excessive or asymmetric Existing Program Components wear. Credited in MRP-227-A Core Barrel Assembly Upper The specific relevant condition Page 80, Table 4. C-E Plants 6 Flange is loss of material (wear). Existing Program Components Credited in MRP-227-A lower Support Structure Fuel The specific relevant conditions Page 80, Table 4. C-E Plants 7 Alignment Pins are severed fuel alignment pins, Existing Program Components Plants with full-height shroud missing locking tabs, or Credited in MRP-227-A plates excessive wear on the fuel alignment pin nose or flange.

(Not Applicable to Palisades) lower Support Structure Fuel The specific relevant condition Page 80, Table 4. C-E Plants 8 Alignment Pins is loss of material (wear). Existing Program Components Plants with Shroud assembled Credited In MRP-227-A in two vertical sections (Not Applicable to Palisades)

Upper Guide Structure Spacer The specific relevant condition Page 121, Attachment C, 9 Shim is loss of material (wear). "Planned Scope of EXisting And (Palisades Plant Specific Item) Augmented Reactor Vessel Internals examinations" Thus, the examination acceptance criteria for components in the PNP reactor internals requiring visual (VT-3) examination under MRP-227-A is the absence of the relevant condition(s) specified in MRP-227-A, Table 5-2, liCE plants examination acceptance and expansion criteria." From the above table, it is evident that the objective of the required MRP-227-A VT-3 examinations at PNP is for detection of loss of material, distortion, and missing or severed items versus early detection of flaws that could grow to be a structural concern. A Page 6 of 13

Attachment 1 RAI Response on Revised Program Plan for Aging Management of Reactor Vessel Internals complete list of relevant conditions for the MRP-227-A required visual (VT-3) examinations includes:

1. Structural distortion or displacement of parts to the extent that component function may be impaired;

2. Loose, missing, cracked, or fractured parts, bolting, or fasteners;

3. Corrosion or erosion that reduces the nominal section thickness by more than 5%;

4. Wear of mating surface that may lead to loss of functionality; and

5. Structural degradation of interior attachments such that the original cross-sectional area is reduced more than 5%.

As noted above, other examination techniques such as EVT-1 or ultrasonic testing are used for detection and monitoring of cracking at PNP when needed for aging management under MRP-227-A.

NRC Request

3. During the extended period of operation, some PWR RVI components are subject to high levels neutron radiation that may lead to irradiation embrittlement and a loss of fracture toughness and the potential for irradiation assisted stress corrosion cracking (IASCC). In combination, these effects may lead to the potential for component failure under some design basis loading conditions.

Explain how the Palisades AMP will account for potential reduction in fracture toughness when evaluating cracks that are detected during the required inspections, in particular when establishing the frequency of subsequent inspections after cracking is identified.

ENO Response

3. Cracking detected during examinations will be evaluated using the evaluation acceptance criteria and methodologies currently being developed by the Pressurized Water Reactor Owners Group (PWROG) Materials Subcommittee.

The flaw evaluation methodology for assessing examinations that do not meet the acceptance criteria in MRP-227-A is outlined in WCAP-17096-NP, "Reactor Internals Acceptance Criteria Methodology and Data Requirements," Revision 2

[Reference 8], or latest NRC-approved revision. WCAP-17096-NP outlines the approach that will be used for establishing the frequency of subsequent inspections after cracking is identified. The procedures outlined in WCAP-17096-NP are similar to the steps included in Section 6 of MRP-227-A.

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Attachment 1 RAI Response on Revised Program Plan for Aging Management of Reactor Vessel Internals Section 6 of MRP-227-A describes potential evaluation steps that can be followed when cracking is detected during the specified examination, whether the cracking is due to loss of fracture toughness or from irradiation assisted stress corrosion cracking. These steps include the assessment of cracking through limit load and/or fracture mechanics evaluations, depending upon the extent of the loss of fracture toughness. The steps also include the potential for increasing the frequency of subsequent examinations as determined from the evaluations using crack growth rates described in Section 6.2.4 of MRP-227-A. The descriptions in Section 6 of MRP-227-A are "For Information Only." Additional information on evaluation of cracking and changes in toughness properties is available from other sources, such as Reference 3.

NRC Request

4. Loose parts could be generated due to deterioration of some PWR RVI components during the extended period of operation. Provide information that addresses how the following consequences of loose parts generation were considered in development of the inspection program given in the proposed AMP.

(a) potential for fuel bundle flow blockage and consequential fuel damage, (b) potential for interference with control rod operation, and (c) potential for impact damage on reactor internals.

ENO Response

4. In general, loose parts are included in existing plant-specific monitoring and evaluation procedures and they remain a plant specific issue. As the intent of the inspection program developed in MRP-227 -A is to monitor for potential age-related degradation, all reactor vessel (RV) internals items were evaluated for aging degradation, and the consequences of the generation of loose parts were considered. However, the results and the consequences of the loose parts generation were previously considered, evaluated, and documented in Section 11 of MRP-156, "Materials Reliability Program: Pressurized Water Reactor Issue Management Table, PWR-IMT Consequence of Failure," and Section 6 of MRP-157, "Materials Reliability Program: Updated B&W Design Information for the Issue Management Tables," and noted throughout the failure modes, effects, and criticality analysis (FMECA) efforts summarized in MRP-190, "Materials Page 8 of 13

Attachment 1 RAI Response on Revised Program Plan for Aging Management of Reactor Vessel Internals Reliability Program: Failure Modes, Effects, and Criticality Analysis of 8&W-Designed PWR Internals," and MRP-191, "Materials Reliability Program:

Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design." Specific information for items (a) - (c) above is available in these reports.

As an example, below the core, the RV internals consist of the support structure.

This component is very substantial and would not be expected to be damaged by loose parts. The primary purpose of this component is to position and support the core and direct the flow to the fuel assemblies. Large loose parts that are capable of being lifted by the flow will be filtered by this component and likely to be lodged in or pinned against this structure. Smaller parts or fragments that can pass through the flow holes are typically trapped in the lower end of the fuel assembly. The fuel assembly bottom plates have a fuel guard debris filter for added protection against loose parts as prevention against fuel failures. Flow area blockage associated with loose parts in the lower internals will have an insignificant effect on core performance since the flow will be redistributed downstream of the blockage and in the lower span of the fuel assemblies.

For situations where loose parts are detected the PNP corrective action program is used to resolve loose parts concerns.

NRC Request

5. Historically, the following materials used in the PWR RVI components were known to be susceptible to some of the aging degradation mechanisms that are identified in the MRP-227-A report. In this context, the NRC staff requests that the licensee confirm that these materials are not currently used in the RVI components at Palisades.

(1) Nickel base alloys-lnconeI600; Weld Metals-Alloy 82 and 182 and Alloy X-750 (excluding control rod guide tube split pins)

(2) Alloy A-286 ASTM A 453 Grade 660, Condition A or B (3) Stainless steel (SS) type 347 material (excluding baffle-former bolts)

(4) Precipitation hardened (PH) stainless steel materials-17-4 and 1fr5 (5) Type 431 stainless steel material Page 9 of 13

Attachment 1 RAI Response on Revised Program Plan for Aging Management of Reactor Vessel Internals ENO Response

5. PNP has confirmed through review of RV internals components the following:

There are no nickel base alloys, including Inconel 600, Alloy 82 and 182 welds, or Alloy X-7S0 materials, in RV internals components. Furthermore, there are no type 347 stainless steel materials, or type 431 stainless steel materials.

A-286 material is used in the plunger assemblies connected to the hold down ring that bears on the top of the flange of the upper guide structure and provides the spring hold down force. Plunger assembly components fabricated from A-286 material include the plungers, plunger shims and plunger washers. In performing their function, the plunger components are in compression and would not be susceptible to aging degradation mechanisms that could cause cracking or otherwise degrade the ability to maintain compressive force.

A-286 material is also used in jack screws that provide lateral force to draw the hold down ring against the outer diameter of the upper guide structure flange.

The concern for aging of the A-286 material is sec under high tensile stresses.

The mechanical advantage of the jack screw is that it provides a force by insertion of the screw into the threaded hole, and when serving its function there is no tension on the screws since they have no heads but are merely inserts (acts like a set screw). Once in place, the threads of the jack screws are loaded in bending (causing a shear stress) and the functionality of the screws are not challenged by susceptibility to see. Therefore, there is no need for aging management of these jack screw components.

17-4 PH material is used in grid ring bushings in the upper guide structure.

These bushings provide the threaded attachment pOints for the upper guide structure lifting rig that is used for lifting and insertion of the upper guide structure. Aging effects such as thermal aging of the 17-4 PH material would not lead to degradation that could cause the bushings to lose their ability to function when needed during lifting and insertion of the upper guide structure. This is because strength of the bushings is the primary concern, and studies and plant experience have shown type 17-4 PH material is prone to secondary thermal aging resulting in an increase in strength and hardness after long term exposure to temperature above approximately 500°F, [Reference 4]. In addition, the grid ring bushings are inspected for thread damage prior to use per PNP refueling procedure RFL-O-17, "Upper Guide Structure Lift Rig Installation." For these reasons, management of thermal aging in the 17-4 PH martensitic stainless steel (MSS) RV internals components is not needed for PNP.

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Attachment 1 RAI Response on Revised Program Plan for Aging Management of Reactor Vessel Internals Therefore, the confirmatory review of known susceptible materials has been completed.

NRC Request

6. With respect to the management of cast austenitic stainless steel (CASS) aging and embrittlement, the licensee states that no CASS is present in the Palisades lower core support structures. No statement is provided for other possible locations. Either confirm that no CASS is present in any of the locations covered by the AMP, or provide a discussion of how the AMP adequately addresses the requirements specified in GALL AMP, XI.M12, "Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS), " and GALL AMP XI.M13, "Thermal Aging and Neutron Embrittlement of Cast Austenitic Stainless Steel (CASS), " for CASS materials used in PWR RVI components.

ENO Response

6. A confirmatory review was conducted of PNP drawings and the aging management evaluation of the reactor vessel internals developed for license renewal [Reference 2]. The review confirmed cast austenitic stainless steel (CASS) is not present in any of the reactor internal locations covered by the AMP.

NRC Request

7. When exposed to a light-water reactor temperature of approximately 500 OF or higher, the 17-4 precipitation hardened (PH) martensitic stainless steel (MSS) undergoes embrittlement and an increase in hardness (i.e., a reduction in Charpy "V" notch fracture toughness value). Operating experience from Oconee Nuclear Station shows that thermally embrittled 17-4 PH MSS is susceptible to failure when exposed to unexpected loading conditions. On March 7, 2007, the staff issued Information Notice (IN)-2007-02 (Agencywide Document Access and Management System Accession Number ML0701004590), in which the staff recommends that the licensees can prevent the deleterious effects of thermal embrittlement in the 17-4 PH MSS components by identifying aging degradation (i.e., cracks), implementing early corrective actions, and monitoring and trending age-related degradation. The licensee did identify that none exists in the internals lower support structures but did not clarify if any exist elsewhere.

Therefore, the staff requests that the subject AMP should include thermal Page 11 of 13

Attachment 1 RAI Response on Revised Program Plan for Aging Management of Reactor Vessel Internals embrittlement as an aging effect for any 17-4 PH MSS RVI components at Palisades if any exist.

ENO Response

7. As noted in the response to RAI-5, 17-4 PH material is only used in grid ring bushings in the upper guide structure of the PNP RV internals. These bushings provide the threaded attachment points for the upper guide structure lifting rig that is used for lifting and insertion of the upper guide structure. Aging effects such as thermal aging of the 17-4 PH material would not lead to degradation that could cause the bushings to lose their ability to function when needed during lifting and insertion of the upper guide structure. This is because strength of the bushings is the primary concern, and studies and plant experience have shown type 17-4PH is prone to secondary thermal aging resulting in an increase in strength and hardness after long term exposure to temperature above approximately 500°F [Reference 4]. Therefore, management of thermal aging in the 17-4 PH MSS RV internals components is not needed for PNP.

NRC Request

8. Note 5 of Table 2 lacks an actual minimum. Clarify the intent of Note 5.

ENO Response

8. ENO agrees that existing Note 5 is ambiguous. The note is clarified below based on Note 5 from MRP-227-A, Table 4-2, "CE Plants Primary Components:"

"A minimum of 75% of the total population of core support column welds."

References:

[1] Palisades Reactor Vessel Internals Aging Management Program, CEP-RVI-PNP, Revision 1, July 26,2012 [ADAMS Accession Number ML12257A352].

[2] ENO document, LR-AMR-RVI, Revision 4, "Aging Management Review, Reactor Vessel Internals, Palisades Nuclear Plant, License Renewal Project," October 20, 2005.

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Attachment 1 RAI Response on Revised Program Plan for Aging Management of Reactor Vessel Internals

[3] "Fracture Toughness of Irradiated Stainless Steel in Nuclear Power Systems," S.

Fyfitch, et. aI., 14th Internal Conference on Environmental Degradation of Materials in Nuclear Power Systems, Virginia Beach, Virginia, August 23 - 27,2009.

[4] H. Xu and S. Fyfitch, "Aging Embrittlement Modeling of 17-4PH at LWR Temperatures," 10th International Conference on Environmental Degradation in Nuclear Power Systems - Water Reactors, National Association of Corrosion Engineers, Lake Tahoe, CA, August 3 - 9, 2001.

[5] Westinghouse letter, LTR-RIAM-13-25, "Palisades Core Barrel Upper Flange Fluence", March 27,2013.

[6] END Letter PNP 2012-080, "Revised Program Plan for Aging Management of Reactor Vessel Internals," Attachment 1, "Palisades Reactor Vessel Internals Aging Management Program," Attachment C, "Planned Scope of Existing and Augmented Reactor Vessel Internals Examinations" [ADAMS Accession Number ML12257A352].

[7] Electric Power Research Institute (EPRI) document, "Materials Reliability Program:

Aging Management Strategies for B&W Pressurized Water Reactor Internals, MRP-231, Revision 2.

[8] WCAP-17096-NP, "Reactor Internals Acceptance Criteria Methodology and Data Requirements," Revision 2, December 2009 [ADAMS Accession Number ML101460157].

[9] NRC transmittal, "PWR Reactor Internals & Evaluation Guidelines (MRP-227-A),

Part 4 of 7, RAI Set #2 RAI-2-2 through RAI Set #3", January 9,2012 [ADAMS Accession Number ML12017A191].

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