RA-24-0033, Response to Request for Additional Information Regarding the Aging Management Program and Inspection Plan for the Shearon Harris Nuclear Power Plant, Unit 1, Reactor Vessel Internals

Response to Request for Additional Information Regarding the Aging Management Program and Inspection Plan for the Shearon Harris Nuclear Power Plant, Unit 1, Reactor Vessel Internals

ML24055A001
Person / Time
Site:	Harris
Issue date:	02/24/2024
From:	Haaf T Duke Energy Progress
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RA-24-0033
Download: ML24055A001 (1)

Text

J'_~ DUKE

~ ENERGY February 24, 2024 Serial: RA-24-0033 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 / RENEWED LICENSE NO. NPF-63 Thomas P. Haaf Site Vice President Harris Nuclear Plant 5413 Shearon Harris Rd New Hill, NC 27562-9300 984-229-2512

Subject:

Response to Request for Additional Information Regarding the Aging Management Program and Inspection Plan for the Shearon Harris Nuclear Power Plant, Unit 1, Reactor Vessel Internals Ladies and Gentlemen:

By letter dated September 21, 2023 (Agencywide Documents Access Management System (ADAMS) Accession No. ML23264A034), Duke Energy Progress, LLC (Duke Energy),

submitted for review the Aging Management Program (AMP) and Inspection Plan for the reactor vessel internals (RVI) for the Shearon Harris Nuclear Power Plant, Unit 1 (Harris), in accordance with Section 18.1 of the Harris Updated Final Safety Analysis Report (UFSAR).

The U.S. Nuclear Regulatory Commission (NRC) staff reviewed the submittal and determined that additional information is needed to complete their review. Duke Energy received the request for additional information (RAI) from the NRC through electronic mail on January 25, 2024 (ADAMS Accession No. ML24025A159).

The Enclosure to this letter provides Duke Energy's response to the RAI. There are no new regulatory commitments contained within this letter.

Should you have any questions concerning this letter, or require additional information, please contact Sarah McDaniel, Harris Regulatory Affairs, at 984-229-2002.

Sincerely,

~

Site Vice President

Enclosure:

Response to Request for Additional Information

U.S. Nuclear Regulatory Commission Page 2 of 2 Serial: RA-24-0033 cc:

(all with Enclosure)

L. Dudes, USNRC Region II - Regional Administrator P. Boguszewski, USNRC Senior Resident Inspector, HNP M. Mahoney, USNRC Project Manager, HNP

U.S. Nuclear Regulatory Commission Serial: RA-24-0033 Enclosure ENCLOSURE RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION

U.S. Nuclear Regulatory Commission Page 1 of 19 Serial: RA-24-0033 Enclosure NRC RAI INTRODUCTION By letter dated September 21, 2023 (Agencywide Documents Access Management System (ADAMS) Accession No. ML23264A034), Duke Energy Progress, LLC (Duke Energy),

submitted for review the Aging Management Program (AMP) and Inspection Plan for the reactor vessel internals (RVI) for the Shearon Harris Nuclear Power Plant, Unit 1 (Harris), in accordance with Section 18.1 of the Harris Updated Final Safety Analysis Report (UFSAR).

The AMP and Inspection Plan for Harris is based upon the Electric Power Research Institute (EPRI) Technical Report MRP-227, Revision 1-A, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (ADAMS Accession No. ML19339G350), which provides guidance for managing age-related material degradation in RVI components through the period of extended operation.

The U.S. Nuclear Regulatory Commission (NRC) staff reviewed the request and determined that additional information is needed to complete their review. Duke Energy received the request for additional information (RAI) from the NRC through electronic mail on January 25, 2024 (ADAMS Accession No. ML24025A159).

RAI-1

By letter dated May 19, 2023 (ML23290A020) the EPRI Materials Reliability Program (MRP) issued interim guidance in MRP 2023-005 to the members of the Pressurized Water Reactor (PWR) Materials Management Program (PMMP) Executive Committee Members and MRP Research Integration Committee Members. This interim guidance was based on recent operating experience at a United States PWR and was developed by the Core Barrel Focus Group to update the guidelines within MRP-227 regarding inspection of core barrel welds.

Section 4.2.5 of WCAP-18710-NP, Aging Management Program and Inspection Plan for Shearon Harris Unit 1 Reactor Vessel Internals, Revision 0 (Enclosure 1 to letter dated September 21, 2023), states that during a fall 2022 inspection, a 3-loop Westinghouse-designed plant identified crack-like surface indications on the inside diameter of the core barrel upper girth weld. The licensee stated it will participate in an industry focus group, evaluate the impact of this fall 2022 operating experience, and address any resulting industry guidance changes at Harris.

However, the NRC staff noted that the inspections for the core barrel in the licensees RVI AMP and Inspection Plan appear to be based on MRP-227, Revision 1-A and does not address this newly issued interim guidance; thus, it is not clear to the NRC staff whether the licensees RVI AMP and Inspection Plan addresses this recent operating experience.

Discuss the applicability of the interim guidance in MRP 2023-005 to the Harris AMP and Inspection Plan for RVIs:

If this interim guidance is applicable, provide any necessary revisions to the Harris AMP and Inspection Plan for RVIs.

If this interim guidance is not applicable, justify that this recent operating experience and the interim guidance in MRP 2023-05 is not relevant.

U.S. Nuclear Regulatory Commission Page 2 of 19 Serial: RA-24-0033 Enclosure Duke Energy Response to RAI-1 Upon issuance of MRP 2023-005, the impact on the Harris RVI Program was evaluated via the NEI 03-08 protocol. As stated in MRP 2023-005, Utility members who operate WEC/CE-design PWR units are required to implement the attached NEI 03-08 Needed interim guidance at the time of the next planned RV internals core barrel removal coinciding with MRP-227 examinations of the RV internals. Therefore, the guidance is applicable to the Harris AMP and Inspection Plan for RVIs. Harris intends to implement Tables 2, 4, and 6 in MRP 2023-005 at the time of the next planned RV internals core barrel removal coinciding with MRP-227 examinations of the RV internals. Tables 7-1, C-1, C-2, and C-4 from WCAP-18710-NP are revised as follows. Affected line items and notes are denoted as [Revised by MRP 2023-005].

Harris implements work products issued under the implementation protocol of NEI 03-08 in accordance with Duke Energy administrative procedures. These procedures also implement activities to update the RVI program with the issuance of industry documents that can impact the program. A failure to meet a Needed or a Mandatory requirement is a deviation from the guidelines and a written justification for the deviation must be prepared and approved as described in Appendix B of NEI 03-08 and Duke Energy administrative procedures, including notification to the NRC for information only.

U.S. Nuclear Regulatory Commission Page 3 of 19 Serial: RA-24-0033 Enclosure Table 7-1 Shearon Harris Unit 1 Primary Component Inspection Plan RVIComponent (Note1)

ExaminationMethod ExaminationFrequency HarrisDueDate (RefuelingOutage)

ProjectedEFPY (Note2)

W1.ControlRodGuide TubeAssembly Guideplate(cards)

PertheRequirements ofWCAP17451P PertherequirementsofWCAP17451P,including subsequentexaminations.(Notes3and4)

RFO27 Spring2027 35.9 W2.ControlRodGuide TubeAssembly Lowerflangewelds Enhancedvisual(EVT1) examination Nolaterthan2refuelingoutagesfromthebeginningofthe licenserenewalperiodandsubsequentexaminationona 10yearinterval.

RFO28 Fall2028 37.4 W3.CoreBarrelAssembly Upperflangeweld(UFW)

[RevisedbyMRP2023005]

Enhancedsurfacevisual (EVT1),eddycurrent (ET),orvolumetric(UT) examination Nolaterthan2refuelingoutagesfromthebeginningofthe licenserenewalperiodandsubsequentexaminationona 10yearinterval.

RFO28 Fall2028 37.4 W3a.CoreBarrel Assembly Uppergirthweld(UGW)

[RevisedbyMRP2023005]

Enhancedsurfacevisual (EVT1),eddycurrent (ET),orvolumetric(UT) examination Nolaterthan2refuelingoutagesfromthebeginningofthe licenserenewalperiodandsubsequentexaminationona 10yearinterval.

RFO28 Fall2028 37.4 W4.CoreBarrelAssembly Lowergirthweld(LGW)

[RevisedbyMRP2023005]

Enhancedsurfacevisual (EVT1),eddycurrent (ET),orvolumetric(UT) examination Nolaterthan2refuelingoutagesfromthebeginningofthe licenserenewalperiodandsubsequentexaminationona 10yearinterval.

RFO28 Fall2028 37.4 W6.BaffleFormer Assembly Baffleformerbolts Volumetric(UT) examination Dependentonplantdesign.Subsequentexaminationis dependentontheplantdesignandtheresultsofthe baselineinspection.(Notes5and6)

RFO26 Fall2025 34.4

U.S. Nuclear Regulatory Commission Page 4 of 19 Serial: RA-24-0033 Enclosure Table 7-1 Shearon Harris Unit 1 Primary Component Inspection Plan RVIComponent (Note1)

ExaminationMethod ExaminationFrequency HarrisDueDate (RefuelingOutage)

ProjectedEFPY (Note2)

W7.BaffleFormer Assembly Assembly Visual(VT3) examination Baselineexaminationbetween20and40EFPYand subsequentexaminationsona10yearinterval.

RFO29 Spring2030 38.9 Notes:

(1) MRP-227, Revision 1-A Primary components W5. Baffle-former assembly baffle-edge

bolts, W8.

Alignment and interfacing components internals hold down spring, and W9. Thermal shield assembly thermal shield flexures are not applicable to Harris. See Table C-1 of this document for further detail.

(2) Projected EFPY values were calculated based on the total EFPD at the end of cycle R23 (29.9 EFPY) with 1.5 EFPY per cycle (1 EFPY per year) conservatively assumed for the future.

(3) Due to the timing of the associated NRC reviews of industry documents and issuance of WCAP-17451-P, Revision 2 [30], MRP-227, Revision 1-A states utilities should Use WCAP-17451-P, Revision 1, including the modified requirements due to the interim guidance provided in EPRI letter MRP 2018-007 dated 3/7/2018 [47] and PWROG letter OG-18-46 dated 2/20/2018 [48]. However, Harris will follow the guidance in WCAP-17541-P, Revision 2 for management of guide card and lower guide tube continuous guidance wear at Shearon Harris Unit 1.

(4) Section 5.5.1 of WCAP-17451-P allows utilities to optionally perform video inspections of the guide cards to help screen when guide tubes will need to be measured. This screening may be used to estimate the wear condition of the guide cards and develop an alternate initial wear measurement schedule.

(5) Shearon Harris Unit 1 is a Tier 4 plant in NSAL-16-1 [56]. In accordance with MRP-227, Revision 1-A, Harris will perform baseline UT examination no later than 35 EFPY.

(6) Re-examination periods shall be determined by plant-specific evaluation per the MRP-227 Needed Requirement 7.5 as documented and dispositioned in the owners plant corrective action program. If atypical or aggressive baffle-former bolt degradation as defined in MRP 2017-009 [58] (i.e., 3% of baffle-former bolts with UT or visual indications or clustering* for downflow plants and 5% of baffle-former bolts with UT or visual indications or clustering* for upflow plants) is observed, the interim guidance (MRP 2016-021 [57] and MRP 2017-009) provides limitations to the permitted reinspection interval (not to exceed 6 years maximum) unless further evaluation is performed to justify a longer interval (See Applicant/Licensee Action Item 1 in the NRC SE for evaluation submittal requirements [19]). If evaluation justifies a longer reinspection interval, it is not permitted to exceed 10 years.

Clustering is defined per NSAL-16-1, Revision 1 [56] as three or more adjacent defective baffle-former bolts or more than 40%

defective baffle-former bolts on the same baffle plate. Untestable bolts should be reviewed on a plant-specific basis consistent with WCAP-17096-NP-A for determination if these should be considered when evaluating clustering.

U.S. Nuclear Regulatory Commission Page 5 of 19 Serial: RA-24-0033 Enclosure Table C-1: MRP-227, Revision 1-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Item Applicability Effect (Mechanism)

Expansion Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage W1.Control Rod Guide Tube Assembly Guide plates (cards)

All plants (See WEC NSAL 1)

Loss of Material (Wear)

None Per the requirements of WCAP-17451-P, including subsequent examinations.

(Note 5)

Examination coverage per the requirements of WCAP-17451-P, Revision 1.

(Note 5)

See Figure A-2.

W2.Control Rod Guide Tube Assembly Lower flange welds All plants Cracking (SCC, Fatigue)

Aging Management (IE and TE)

W2.1.Remaining CRGT assembly lower flange welds W2.2.BMI column bodies Enhanced visual (EVT-1) examination to determine the presence of crack-like surface flaws in flange welds no later than 2 refueling outages from the beginning of the license renewal period and subsequent examination on a 10-year interval.

100% of outer (accessible)

CRGT lower flange weld surfaces and 0.25-inch of the adjacent base metal on the individual periphery CRGT assemblies.

(Note 2)

See Figure A-3.

W3.Core Barrel Assembly Upper flange weld (UFW)

[Revised by MRP 2023-005]

All plants Cracking (SCC)

W3.2.Upper axial welds (UAW)

W3.3.Lower flange weld (LFW)

W3.4.Lower support forging or casting Enhanced surface visual (EVT-1),

eddy current (ET), or volumetric (UT) examination, no later than 2 refueling outages from the beginning of the license renewal period and subsequent examination on a ten-year interval.

100% of the accessible weld length of both surfaces (ID surface and OD surface) of the UFW and 3/4 of adjacent base metal shall be examined. If UT is performed, it need only be completed from one surface, either ID or OD.

(Note 11)

See Figure A-4.

U.S. Nuclear Regulatory Commission Page 6 of 19 Serial: RA-24-0033 Enclosure Table C-1: MRP-227, Revision 1-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Item Applicability Effect (Mechanism)

Expansion Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage W3a.Core Barrel Assembly Upper girth weld (UGW)

[Revised by MRP 2023-005]

All plants Cracking (SCC)

W3.2. Upper core barrel upper axial welds (UAW)

W3.3. Lower flange weld (LFW)

W3.4. Lower support forging or casting Enhanced visual (EVT-1),

volumetric (UT), or surface (ET) examination, no later than 2 refueling outages from the beginning of the license renewal period and subsequent examination on a ten-year interval.

100% of the accessible weld length of both surfaces (ID surface and OD surface) of the UGW and 3/4 of adjacent base metal shall be examined. If UT is performed, it need only be completed from one surface, either ID or OD.

(Note 11).

See Figure A-4.

W4.Core Barrel Assembly Lower girth weld (LGW)

[Revised by MRP 2023-005]

All plants Cracking (SCC, IASCC),

Aging Management (IE)

W4.1.Upper core plate W4.2.Middle axial welds (MAW)

W4.3.Lower axial welds (LAW)

W4.4.Lower support column bodies (cast, non-cast)

Enhanced visual (EVT-1),

volumetric (UT), or surface (ET) examination, no later than 2 refueling outages from the beginning of the license renewal period and subsequent examination on a ten-year interval.

100% of the accessible weld length of the OD surface of the LGW and 3/4 of adjacent base metal shall be examined (ID surface is inaccessible for visual/ET-based surface exams due to baffle-former assembly).

UT is performed from OD-surface.

(Note 6)

See Figure A-4.

U.S. Nuclear Regulatory Commission Page 7 of 19 Serial: RA-24-0033 Enclosure Table C-1: MRP-227, Revision 1-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Item Applicability Effect (Mechanism)

Expansion Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage W5.Baffle-Former Assembly Baffle-edge bolts All plants with baffle-edge bolts (Note 12)

Cracking (IASCC, Fatigue) that results in:

Lost or broken locking devices Failed or missing bolts Protrusion of bolt heads Aging Management (IE and ISR)

(Note 4)

None Visual (VT-3) examination, with baseline examination between 20 and 40 EFPY and subsequent examinations on a 10-year interval.

Bolts and locking devices on high-fluence seams. 100% of components accessible from core side.

See Figures A-5, A-6, and A-7.

W6.Baffle-Former Assembly Baffle-former bolts (Note 7)

All plants (See WEC NSAL 1)

Cracking (IASCC, Fatigue)

Aging Management (IE and ISR)

(Note 4)

W6.1.Barrel-former bolts W6.2.Lower support column bolts Baseline volumetric (UT) examination interval is dependent on the plant design (Note 8).

Subsequent examination is dependent on the plant design and the results of the baseline inspection (Note 9).

100% of accessible bolts.

(Note 3)

See Figures A-5, A-6, and A-7.

U.S. Nuclear Regulatory Commission Page 8 of 19 Serial: RA-24-0033 Enclosure Table C-1: MRP-227, Revision 1-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Item Applicability Effect (Mechanism)

Expansion Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage W7.Baffle-Former Assembly Assembly (Includes: Baffle plates, baffle edge bolts, corner bolts, and indirect effects of void swelling in former plates)

All plants Distortion (Void Swelling), or Cracking (IASCC) that results in:

Abnormal interaction with fuel assemblies Gaps between plates Vertical displacement of baffle plates Broken or damaged edge bolts None Visual (VT-3) examination to check for evidence of distortion, with baseline examination between 20 and 40 EFPY and subsequent examinations on a 10-year interval.

Core side surface:

High fluence baffle joints Top and bottom edge of baffle plates Bolts and locking devices See Figure A-6.

W8.Alignment and Interfacing Components Internals hold-down spring All plants with 304 stainless steel hold-down springs (Note 13)

Distortion (Loss of Load due to Stress Relaxation)

None Direct measurement of spring height within 3 cycles of the beginning of (before or after) the license renewal period. If the first set of measurements is not sufficient to assess remaining life, additional spring height measurements will be required.

Measurements should be taken at several points around the circumference of the spring, with a statistically adequate number of measurements at each point to minimize uncertainty.

W9.Thermal Shield Assembly Thermal shield flexures All plants with thermal shields (See WEC TB-19-5)

(Note 14)

Cracking (Fatigue) or Loss of Material (Wear) that results in thermal shield flexures excessive wear, fracture, or complete separation None Visual (VT-3) no later than 2 refueling outages from the beginning of the license renewal period. Subsequent examinations on a 10-year interval.

100% of accessible surfaces of 100% of thermal shield flexures.

(Notes 10 and 14)

U.S. Nuclear Regulatory Commission Page 9 of 19 Serial: RA-24-0033 Enclosure Notes:

(1) Examination acceptance criteria and expansion criteria for the Westinghouse components are listed in Table C-4.

(2) A minimum of 75% of the total identified sample population must be examined.

(3) A minimum of 75% of the total population (examined + unexamined),

including coverage consistent with the Expansion criteria in Table C-4 must be examined for inspection credit.

(4) Void swelling effects on this component are managed through management of void swelling on the entire baffle-former assembly.

(5) In WCAP-17451-P the baseline examination schedule has been adjusted for various CRGT designs, the extent of individual CRGT examination modified, and flexible subsequent examination regimens correlating to initial baseline sample size, accuracy of wear estimation and examination results. Initial inspection prior to the license renewal period may be required. Use WCAP-17451-P [30], Revision 1, including the modified requirements due to the interim guidance provided in EPRI letter MRP 2018-007 dated 3/7/2018 [47]

and PWROG letter OG-18-46 dated 2/20/2018 [48].

Harris will follow the inspection and evaluation guidance within WCAP-17451-P, Revision 2, which includes the modified requirements provided in MRP 2018-007 and OG-18-46.

(6) Examination coverage requires a minimum of 50% of the length of the OD of the weld being examined. [Revised by MRP 2023-005]

(7) Baffle-former bolt inspection includes inspection of the corner plate bolts when applicable.

(8) In accordance with MRP 2017-009 [58] and MRP 2017-010 [77], Tier 1 plants are to perform the baseline UT examination by 20 EFPY or during the next refueling outage after March 1, 2016. Per MRP 2017-009 [58], Tier 2 plants are to perform the baseline UT examination at no later than 30 EFPY (initial Tier 2 plant baseline UT exams performed prior to 1/1/2018 are acceptable). All other remaining plants are to perform the baseline UT examination at no later than 35 EFPY.

(9) Re-examination periods shall be determined by plant-specific evaluation per the MRP-227 Needed Requirement 7.5 as documented and dispositioned in the owners plant corrective action program. If atypical or aggressive baffle-former bolt degradation as defined in MRP 2017-009 [58] (i.e., 3% of baffle-former bolts with UT or visual indications or clustering* for downflow plants and 5%

of baffle-former bolts with UT or visual indications or clustering* for upflow plants) is observed, the interim guidance (MRP 2016-021 [57] and MRP 2017-009 [58]) provides limitations to the permitted reinspection interval (not to exceed 6 years maximum) unless further evaluation is performed to justify a longer interval (See Applicant/Licensee Action Item 1 in the NRC SE for evaluation submittal requirements [19]). If evaluation justifies a longer reinspection interval, it is not permitted to exceed 10 years.

• Clustering is defined per NSAL-16-1, Revision 1 [56] as three or more adjacent defective baffle-former bolts or more than 40% defective baffle-former bolts on the same baffle plate. Untestable bolts should be reviewed on a plant-specific basis consistent with WCAP-17096-NP-A for determination if these should be considered when evaluating clustering.

(10) See Westinghouse Technical Bulletin TB-19-5 dated 10/9/2019 and MRP 2019-017 dated 5/31/2019 for additional details on inspection recommendations.

(11) Examination coverage requires a minimum of 75% of the weld length of both the ID and the OD of the weld being examined. [Revised by MRP 2023-005]

(12) Harris does not have baffle-edge bolts [43]; therefore, this item is not applicable to Harris.

(13) The Harris hold-down spring is 403 SS [43]; therefore, this component is not applicable to Harris.

(14) Harris does not have a thermal shield [43]; therefore, this item is not applicable to Harris.

U.S. Nuclear Regulatory Commission Page 10 of 19 Serial: RA-24-0033 Enclosure Table C-2: MRP-227, Revision 1-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Expansion Item Applicability Effect (Mechanism)

Primary Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage Control Rod Guide Tube Assembly W2.1.Remaining CRGT lower flange welds All plants Cracking (SCC, Fatigue)

Aging Management (IE and TE)

W2.CRGT Lower Flange Welds Enhanced visual (EVT-1) examination to determine the presence of crack-like surface flaws in flange welds.

Subsequent examination on a 10-year interval.

A minimum of 75% of the CRGT assembly lower flange weld surfaces and 0.25-inch of the adjacent base metal for the flange welds not inspected under the primary link.

See Figure A-3.

Bottom Mounted Instrumentation System W2.2.Bottom-mounted instrumentation (BMI) column bodies All plants Cracking (Fatigue) including the detection of completely fractured column bodies Aging Management (IE)

W2.CRGT Lower Flange Welds Visual (VT-3) examination.

Reinspection every 10 years following initial inspection.

100% of BMI column bodies for which difficulty is detected during flux thimble insertion/withdrawal.

See Figure A-9 and A-11.

Core Barrel Assembly W3.2.Upper Axial Weld (UAW)

[Revised by MRP 2023-005]

All plants Cracking (SCC)

W3.Upper Core Barrel Flange Weld (UFW)

W3a. Upper girth weld (UGW)

Enhanced surface visual (EVT-1), eddy current (ET), or volumetric (UT) examination.

Reinspection every 10 years following initial inspection.

100% of the accessible weld length of both surfaces (ID surface and OD surface) of the UAW and 3/4 of adjacent base metal shall be examined. If UT is performed, it need only be completed from one surface, either ID or OD.

(Note 2).

See Figure A-4.

Core Barrel Assembly W3.3.Lower Flange Weld (LFW)

[Revised by MRP 2023-005]

All plants Cracking (SCC)

W3.Upper Core Barrel Flange Weld (UFW)

W3a. Upper girth weld (UGW)

Enhanced visual (EVT-1) examination. Reinspection every 10 years following initial inspection.

100% of the accessible weld length of the OD surface of the LFW and 3/4 of adjacent base metal shall be examined.

(Note 5)

See Figure A-4.

U.S. Nuclear Regulatory Commission Page 11 of 19 Serial: RA-24-0033 Enclosure Table C-2: MRP-227, Revision 1-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Expansion Item Applicability Effect (Mechanism)

Primary Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage Lower Internals Assembly W3.4.Lower support forging or castings

[Revised by MRP 2023-005]

All plants (Note 7)

Cracking (SCC)

Aging Management (TE in Casting)

W3.Upper Core Barrel Flange Weld (UFW)

W3a. Upper girth weld (UGW)

Visual (VT-3) examination.

Reinspection every 10 years following initial inspection.

Minimum of 25% of bottom (non-core side) surface.

(Note 3)

See Figure A-8 and A-9.

Upper Internals Assembly W4.1.Upper core plate All plants Cracking (Fatigue),

Wear, Aging Management (IE)

W4.Lower Girth Weld (LGW)

Visual (VT-3) examination.

Reinspection every 10 years following initial inspection.

Minimum of 25% of core side surfaces.

(Note 3)

See Figure A-12.

Core Barrel Assembly W4.2.Middle Axial Welds (MAW) and W4.3.Lower Axial Welds (LAW)

All plants Cracking (SCC, IASCC)

Aging Management (IE)

W4.Lower Girth Weld (LGW)

Enhanced visual (EVT-1) examination. Reinspection every 10 years following initial inspection.

100% of the accessible weld length of the OD of the MAW and LAW and 3/4of adjacent base metal shall be examined.

(Notes 5 and 6)

See Figure A-4.

Lower Support Assembly W4.4.Lower support column bodies (both cast and non-cast)

All plants (Note 8)

Cracking (IASCC)

Aging Management (IE)

W4.Lower Girth Weld (LGW)

Visual (VT-3) examination.

Reinspection every 10 years following initial inspection.

25% of the total number of column assemblies (both visible and non-visible from above the lower core plate) using a VT-3 examination from above the lower core plate. The inspection coverage must be evenly distributed across the population of column assemblies.

(Notes 3 and 4)

See Figures A-8, A-9, and A-10.

U.S. Nuclear Regulatory Commission Page 12 of 19 Serial: RA-24-0033 Enclosure Table C-2: MRP-227, Revision 1-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Expansion Item Applicability Effect (Mechanism)

Primary Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage Core Barrel Assembly W6.1.Barrel-former bolts All plants Cracking (IASCC, Fatigue)

Aging Management (IE, Void Swelling, and ISR)

W6.Baffle-former bolts (also refer to MRP 2018-002)

Volumetric (UT) examination.

Reinspection every 10 years following initial inspection.

100% of accessible barrel-former bolts (Minimum of 75% of the total population). Accessibility may be limited by presence of thermal shield or neutron pads.

See Figure A-7.

Lower Support Assembly W6.2.Lower support column bolts All plants Cracking (IASCC, Fatigue)

Aging Management (IE and ISR)

W6.Baffle-former bolts Volumetric (UT) examination.

Reinspection every 10 years following initial inspection.

100% of accessible lower support column (LSC) bolts (Minimum of 75% of the total population) or as supported by plant-specific justification.

See Figures A-8, A-9, and A-10.

Notes:

(1) Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table C-4.

(2) Examination coverage requires a minimum of 75% of the weld length of both the ID and the OD of the weld being examined. [Revised by MRP 2023-005]

(3) The stated minimum coverage requirement is the minimum if no significant indications are found. However, the Examination Acceptance criteria in Section 5 of MRP-227, Revision 1-A require that additional coverage must be achieved in the same outage if significant flaws are found. This contingency should be considered for inspection planning purposes.

(4) Justification that adequate distribution of the inspection coverage has been achieved can be based on geometric or layout arguments. Possible examples include, but are not limited to, inspection of all column assemblies in one quadrant of the lower core plate (based on the azimuthal symmetry of the plate) or inspecting every fourth column across the entire plate.

(5) A minimum coverage of 75% of the weld length on the surface being examined shall be achieved; however, for welds with limited access (Note 6), a minimum examination coverage of 50% of the weld length on the surface being examined shall be achieved.

(6) Accessibility to the MAW and LAW may be limited by the thermal shield or neutron panels - no disassembly to achieve higher weld length coverage is required.

(7) Harris has a lower support forging.

(8) The Harris lower support column bodies are non-cast.

U.S. Nuclear Regulatory Commission Page 13 of 19 Serial: RA-24-0033 Enclosure Table C-4: MRP-227, Revision 1-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals Item Applicability Examination Acceptance Criteria (Note 1)

Expansion Link(s)

Expansion Criteria Additional Examination Acceptance Criteria W1.Control Rod Guide Tube Assembly Guide plates (cards)

All plants Pertherequirementsof WCAP17451P.

The specific relevant condition is wear that could lead to loss of control rod alignment and impede control assembly insertion.

None N/A Per WCAP-17451-P.

W2.Control Rod Guide Tube Assembly Lower flange welds All plants Enhanced visual (EVT-1) examination.

The specific relevant condition is a detectable crack-like surface indication.

W2.1.Remaining accessible CRGT lower flange welds W2.2.Bottom-mounted instrumentation (BMI) column bodies Confirmation of surface-breaking indications in two or more CRGT lower flange welds shall require visual (EVT-1) examination of the remaining accessible CRGT lower flange welds and visual (VT-3) examination of BMI column bodies by the completion of the next refueling outage.

For BMI column bodies, the specific relevant condition for the VT-3 examination is completely fractured column bodies.

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Expansion Criteria Additional Examination Acceptance Criteria W3.Core Barrel Assembly Upper flange weld (UFW)

[Revised by MRP 2023-005]

All plants Enhanced visual (EVT-1),

volumetric (UT), or eddy current (ET) examination.

The specific relevant condition is a detectable crack-like surface indication.

W3.2.Upper axial welds (UAW)

W3.3.Lower flange weld (LFW)

W3.4.Lower support forging/casting

a. The confirmed detection and sizing of a surface-breaking indication with a length greater than 2 inches in the UFW shall require that the inspection be expanded to include the LFW by the completion of the next refueling outage.

b. The confirmed detection and sizing of a surface breaking indication with a length greater than 2 inches in the LFW shall require that the inspection be expanded to include the UAW by the completion of the next refueling outage.

c. The confirmed detection of a surface-breaking indication with a length greater than 2 inches in the LFW shall require the inspection of the lower support forging or casting (25% of the non-core side surface) within the next 3 refueling outages. If an indication is found in this inspection of the lower support forging or casting, the examination coverage shall be expanded to 100% of the accessible surface of the noncore side surface of the lower support forging or casting during the same refueling outage.

The specific relevant condition for the expansion core barrel welds (LFW, UAW) and lower support forging or casting examinations is a detectable crack-like surface indication.

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Expansion Link(s)

Expansion Criteria Additional Examination Acceptance Criteria W3a.Core Barrel Assembly Upper girth weld (UGW)

[Revised by MRP 2023-005]

All plants Enhanced visual (EVT-1),

volumetric (UT), or eddy current (ET) examination The specific relevant condition is a detectable crack-like surface indication.

W3.2. Upper axial welds (UAW)

W3.3. Lower flange weld (LFW) (Note 2)

W3.4. Lower support forging/casting

a. The confirmed detection and sizing of a surface-breaking indication with a length greater than two inches in the UGW shall require that the inspection be expanded to include the LFW by the completion of the next refueling outage.

b. The confirmed detection and sizing of a surface breaking indication with a length greater than two inches in the UGW shall require that the inspection be expanded to include the UAW by the completion of the next refueling outage.

c. The confirmed detection of a surface-breaking indication with a length greater than two inches in the LFW shall require the inspection of the lower support forging or casting (25% of the non-core side surface) within the next three refueling outages. If an indication is found in this inspection of the lower support forging or casting, the examination coverage shall be expanded to 100% of the accessible surface of the non-core side surface of the lower support forging or casting during the same refueling outage.

The specific relevant condition for the expansion core barrel welds (LFW, UAW) and lower support forging or casting examinations is a detectable crack-like surface indication.

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Expansion Link(s)

Expansion Criteria Additional Examination Acceptance Criteria W4.Core Barrel Assembly Lower girth weld (LGW)

All plants Periodic enhanced visual (EVT-1) examination.

The specific relevant condition is a detectable crack-like surface indication.

W4.1.Upper core Plate W4.2.Middle axial welds (MAW)

W4.3.Lower axial welds (LAW)

W4.4.Lower support column bodies (cast and non-cast)

a. The confirmed detection and sizing of a surface-breaking indication with a length greater than two inches in the LGW shall require inspection of the upper core plate (25%

of the core-side surface) within the next 3 refueling outages. If an indication is found in this inspection of the upper core plate, the examination coverage shall be expanded to 100% of the accessible surface of the core-side surface of the upper core plate during the same refueling outage.

b. The confirmed detection and sizing of a surface-breaking indication with a length greater than two inches in the LGW shall require inspection of the lower support column bodies (cast and non-cast) within the next 3 refueling outages.

The confirmed detection of fractured, misaligned, or missing lower support columns shall require examination of 100% of the accessible uninspected lower support column assemblies using a VT-3 examination from above the lower core plate (minimum of 75%

of the total population of lower support column assemblies) during the same outage.

c. The confirmed detection and sizing of a surface-breaking indication with a length greater than two inches in the LGW shall require that the inspections be expanded to include the lower core barrel cylinder axial welds by the completion of the next refueling outage.

a. The specific relevant conditions for the inspection of the upper core plate are broken or missing parts of the plate.

b. The specific relevant conditions for the inspection of the lower support column bodies (cast and non-cast) are fractured, misaligned, or missing columns.

c. The specific relevant condition for the expansion MAW and LAW inspections is a detectable crack-like surface indication.

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Expansion Link(s)

Expansion Criteria Additional Examination Acceptance Criteria W5.Baffle-Former Assembly Baffle-edge bolts All plants with baffle-edge bolts (Note 3)

Visual (VT-3) examination.

The specific relevant conditions are missing or broken locking devices, failed or missing bolts, and protrusion of bolt heads.

None N/A N/A W6.Baffle-Former Assembly Baffle-former bolts All plants Volumetric (UT) examination.

The examination acceptance criteria for the UT of the baffle-former bolts shall be established as part of the examination technical justification.

W6.1.Barrel-former bolts (Note

2)

W6.2.Lower support column bolts Confirmation that more than 5% of the baffle former bolts actually examined on the four baffle plates at the largest distance from the core (presumed to be the lowest-dose locations) contain unacceptable indications shall require UT examination of the lower support column bolts within the next 3 fuel cycles.

Confirmation that more than 5% of the lower support column bolts actually examined contain unacceptable indications shall require inspection of the barrel-former bolts within 3 refueling cycles.

The examination acceptance criteria for the UT of the lower support column bolts and the barrel-former bolts shall be established as part of the examination technical justification.

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Expansion Link(s)

Expansion Criteria Additional Examination Acceptance Criteria W7.Baffle-Former Assembly Assembly (Includes:

Baffle plates, baffle edge bolts, corner bolts, and indirect effects of void swelling in former plates)

All plants Visual (VT-3) examination.

The specific relevant conditions are evidence of abnormal interaction with fuel assemblies, gaps along high fluence baffle plate joints, vertical displacement of baffle plates near high fluence joints, or more than 2 broken or damaged edge bolt locking systems along high fluence baffle plate joints.

None N/A N/A W8.Alignment and Interfacing Components Internals hold-down spring All plants with 304 stainless steel hold-down springs (Note 4)

Direct physical measurement of spring height.

The examination acceptance criterion for this measurement is that the remaining compressible height of the spring shall provide hold-down forces within the plant-specific design tolerance.

None N/A N/A

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Expansion Criteria Additional Examination Acceptance Criteria W9.Thermal Shield Assembly Thermal shield flexures All plants with thermal shields (Note 5)

Visual (VT-3) examination.

The specific relevant conditions for thermal shield flexures are excessive wear, fracture, or complete separation.

None N/A N/A Notes:

(1) The examination acceptance criterion for visual examination is the absence of the specified relevance condition(s).

(2) If significant baffle-former bolt clustering (as defined in MRP 2017-009 [58]) is discovered, Harris will implement the Needed requirements of MRP 2018-002

[59] within 3 fuel cycles.

(3) Harris does not have baffle-edge bolts; therefore, this item is not applicable to Harris.

(4) The Harris hold-down spring is 403 SS; therefore, this component is not applicable to Harris.

(5) Harris does not have a thermal shield; therefore, this item is not applicable to Harris.