RA-19-0005, Response to NRC Request for Additional Information Regarding Review Request of the Aging Management Program and Inspection Plan for the Reactor Vessel Internals to Implement MRP-227-A

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Response to NRC Request for Additional Information Regarding Review Request of the Aging Management Program and Inspection Plan for the Reactor Vessel Internals to Implement MRP-227-A
ML19031C914
Person / Time
Site: McGuire, Mcguire  Duke Energy icon.png
Issue date: 01/30/2019
From: Teresa Ray
Duke Energy Carolinas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML19032A094 List:
References
MRP-227-A, RA-19-0005
Download: ML19031C914 (35)
v • d • e


Text

Thomas D. Ray Site Vice President McGuire Nuclear Station Duke Energy 12700 Hagers Ferry Rd Huntersville, NC 28078 980.875.4805 Tom.Ray@duke-energy.com PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 3 THIS LETTER IS UNCONTROLLED Serial: RA-19-0005 10 CFR 50.90 January 30, 2019 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 MCGUIRE NUCLEAR STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-369 AND 50-370 RENEWED LICENSE NOS. NPF-9 AND NPF-17

SUBJECT:

RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION (RAI)

REGARDING REVIEW REQUEST OF THE AGING MANAGEMENT PROGRAM AND INSPECTION PLAN FOR THE REACTOR VESSEL INTERNALS TO IMPLEMENT MRP-227-A

REFERENCES:

1. Duke Energy letter, Review Request for the Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A, dated December 13, 2017 (ADAMS Accession No. ML17356A184).
2. Duke Energy letter, Resolution of Commitments related to Review Request for the Aging Management Program and Inspection Plan, dated May 9, 2018 (ADAMS Accession No. ML18135A087).
3. NRC E-Mail, Request for Additional Information - McGuire Nuclear Station, Units 1 and 2 - MRP-227 Review (EPID L-2017-LLA-0414), dated December 18, 2018 (ADAMS Accession No. ML18352A805).

Ladies and Gentlemen:

In Reference 1, as supplemented by Reference 2, Duke Energy Carolinas, LLC (Duke Energy) submitted a Review Request for the Aging Management Program (AMP) and Inspection Plan for the McGuire Nuclear Station (MNS), Units 1 and 2 Reactor Vessel Internals (RVIs).

By correspondence dated December 18, 2018 (Reference 3), the Nuclear Regulatory Commission (NRC) staff requested additional information from Duke Energy that is needed to complete the review.

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 3 THIS LETTER IS UNCONTROLLED

U.S. Nuclear Regulatory Commission RA-19-0005 Page2 PROPRIETARY INFORMATIO N+ WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENlr 3 THIS LETTER IS UNCONTROLLED is a letter from Westinghouse that provides the responses to Reference 3, RAl 1 and RAI 3. Attachment 3 is a letter from Framatome to Duke Energy that provides the response to Reference 3, RAI 2. Attachment 3 contains information that is proprietary to Framatome. In accordance with 10 CFR 2.390, Duke Energy requests that Attachment 3 be withheld from public disclosure. An Affidavit is included (Attachment 2) attesting to the proprietary nature of . A non-proprietary version of Attachment 3 is Included in Attachment 4.

This submittal contains no regulatory commitments.

Should you have any questions concerning this letter, or require additional information, please contact Art Zaremba, Manager- Nuclear Fleet Licensing, at 980-373-2062.

I declare unde'J~nalty of perjury that the foregoing is true and correct. Executed on I /:1. s ~orj .

Sincerely, IVIJll.i"J j)* tJ _A_ A Thomas D. Ray ~

Site Vice President, McGuire Nuclear Station NDE Attachments:

1. Response to RAls 1 and 3: Westinghouse letter, ~Responses Supporting NRG Request for Additional Information on the Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals," dated January 21, 2019.
2. Framatome Affidavit
3. Response to RAI 2: Framatome letter, "Response to Requests for Additional Information for Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A, "dated January 2019 (Proprietary).
4. Response to RAI 2: Framatome letter, "Response to Requests for Additional Information far Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A, "dated January 2019 (Redacted).

PROPRIETARY INFORMATIO N-WITHHOL D UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 3 THIS LETTER IS UNCONTROLLED

U.S. Nuclear Regulatory Commission RA-19-0005 Page 3 PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 3 THIS LETTER IS UNCONTROLLED cc: (all with Attachments unless otherwise noted)

C. Haney, Regional Administrator USNRC Region II G.A. Hutto, USNRC Senior Resident Inspector M. Mahoney, NRR Project Manager W.L. Cox, III, Section Chief, NC DHSR PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 3 THIS LETTER IS UNCONTROLLED

U.S. Nuclear Regulatory Commission RA-19-0005, Attachment 1 Attachment 1 Response to RAIs 1 and 3: Westinghouse letter, Responses Supporting NRC Request for Additional Information on the Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals, dated January 21, 2019.

Westinghouse Non-Proprietary Class 3 Page 2 of 10 LTR-AMLR-19-1, Rev. 0 January 21, 2019 Responses to NRC Requests for Additional Information (RAIs) on the Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals 1.0 Introduction and Background Duke Energy submitted the Reactor Vessel Internals (RVI) Aging Management Program (AMP) and Inspection Plan for McGuire Unit 1 and Unit 2 [1] to the NRC on December 13, 2017 [2], as supplemented by [3]. The NRC reviewed this submittal and requested additional information from McGuire within [4].

The responses to RAI 1 and RAI 3 are within the following sections.

2.0 NRC RAI 1

RAI-1

The licensee stated, in part, in Section 6.2.2 in Attachment 1 (ADAMS Accession No. ML17356A178) of its December 13, 2017 letter, that a detailed tabulation of the McGuire RVI components was completed and compared to typical Westinghouse PWR RVI components in Table 4-4 of MRP-191, Screening, Categorization and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (ADAMS Accession No. ML091910130). This effort identified one McGuire RVI component (access plug assembly spring) that has no corresponding MRP-191 component.

Further assessment of Attachment 1 in the May 9, 2018 supplement (ADAMS Package No.

ML18135A087), indicated that the McGuire access plug assembly spring is made of Inconel X-750, and no degradation mechanisms were identified through the process of MRP-191. Attachment 1 of the May 9, 2018 supplement also indicated that the McGuire anti-vibration sleeves are made of 304 stainless steel (SS), and although several degradation mechanisms have been identified, the [McGuire-specific expert]

panel considered the likelihood of failure and damage to be low.

Please address the following:

In its consideration of the acceptability of MRP-191, as a basis for MRP-227, the NRC staff performed three tasks: 1) assessed the nature and qualifications of the expert panel used in developing MRP-191; 2) assessed the process used by the expert panel in screening, categorization, and ranking the RVI components, and 3) developed an independent assessment of the MRP-191 disposition for verification. The NRC staff intends to review those items applicable to McGuire that were not included in MRP-191 in the same manner as it reviewed MRP-191. To permit the NRC staff to perform its analysis, please compare the makeup and processes used by the McGuire expert panel to those used by the MRP-191 expert panel.

      • This record was final approved on 1/22/2019 8:25:30 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 3 of 10 LTR-AMLR-19-1, Rev. 0 January 21, 2019 The above conclusion on McGuire anti-vibration sleeves was made in Attachment 1 of the May 9, 2018 supplement, without justification. Please provide justification regarding this conclusion (e.g., negligible stresses, absence of hostile environment, or negligible neutron fluence).

Response

The McGuire Expert Panel applied the current Failure Modes, Effects, and Criticality Analysis (FMECA) approach for Westinghouse designed plants as described within Section 6.2 of MRP-191 Revision 0 [5].

By applying Section 6.2 of [5], the McGuire Expert Panel used the same process, definitions, and inputs as the MRP-191 Revision 0 Expert Panel. The McGuire Expert Panel was composed of experts filling the following FMECA Roles:

Component design, testing, and repair Structural modeling and analysis Thermal-hydraulics and systems analysis Neutron fluence and radiation analysis Materials degradation and failure experience Component inspection experience Risk assessment Inspection requirements System function and operating experience Licensing and regulatory interaction The McGuire Expert Panel required consistent definitions for some of the key terms and concepts applied throughout assigning likelihood and consequence levels to the RVI components. Therefore, the following definitions and categories from MRP-191 Revision 0 [5] were used throughout the FMECA:

Component Failure: Material degradation of a given component by one or more credible degradation mechanisms, as identified in the Screening Evaluation, causes the component to lose its ability to perform its intended design function either during normal operation or under accident conditions. Accident conditions include design basis earthquake or pipe break with no credit for the low likelihood of these accidents actually occurring. Cosmetic wear, craze cracking, and plastic deformation (excluding springs) were not considered failures.

Failure Likelihood: The likelihood that component failure(s) will occur during 60 years of operation.

The four categories of failure likelihood are defined in Table 6-2 of [5].

Core Damage: Physical damage to one or more fuel assemblies or other internals components - either through direct impact with the fuel, flow-jetting, loss of core support / fuel spring hold-down force, loose parts, blockage / diversion of coolant flow, or loss of insertion ability for more than one control rod that would impair the ability to safely shut down the reactor.

Damage Likelihood: The conditional likelihood that component failure(s) results in core damage given that the failure occurs irrespective of the actual failure likelihood. The four categories of conditional damage likelihood are defined in Table 6-3 of [5].

      • This record was final approved on 1/22/2019 8:25:30 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 4 of 10 LTR-AMLR-19-1, Rev. 0 January 21, 2019 Based on failure likelihood and damage likelihood, the access plug assembly spring and flux thimble anti-vibration sleeves were assigned a FMECA Group, as seen in Table 1, which is Table 6-4 in [5].

Table 1: Reactor Internals FMECA (Significance) Groups Used by the McGuire Expert Panel Consequence (Damage Likelihood)

Failure Likelihood Low Medium High High 2 3 3 Medium 1 2 3 Low 1 1 2 None 0 0 0 The McGuire Expert Panel discussed the component design, key parameters, and function of the access plug assembly spring and flux thimble anti-vibration sleeves, in addition to material information of each component. The McGuire Expert Panel then reviewed the geometry, location, and function of each component, identified screened-in degradation mechanisms for each component, and discussed relevant information from previous Westinghouse Pressurized Water Reactor Owners Group (PWROG) sponsored FMECAs and the MRP Issue Management Table [8]. A likelihood of failure was determined for each component based on the component review in addition to a review of degradation and failure experience.

A likelihood of damage was determined for each component based on a review of the effects and consequences of degradation and failure.

The access plug assembly spring was determined to have no screened-in degradation mechanisms since the component experiences negligible stresses. Therefore, the McGuire Expert Panel did not evaluate the access plug assembly spring for likelihood of failure or likelihood of damage and the access plug assembly spring was assigned a FMECA ranking of 0 and a Category of A.

The flux thimble anti-vibration sleeves were determined to screen in for irradiation assisted stress corrosion cracking (IASCC), wear, fatigue, irradiation embrittlement (IE), void swelling (VS), and irradiation-induced stress relaxation/irradiation creep (ISR/IC). The McGuire Expert Panel first considered failure of the anti-vibration sleeves and then consequences of failure. If the anti-vibration sleeves were to fail, wear would increase on the flux thimble tubes. This, however, does not compromise safety as McGuire periodically inspects the flux thimble tubes by eddy current examination and monitors wall thinning of the flux thimble tubes in order to predict when tubes should be repaired or replaced [1].

Leakage due to failure would be detectable at the flux thimble tube seal table. The McGuire Expert Panel considered loose parts as a concern; however the anti-vibration sleeves are not expected to become a loose part based on design.

The McGuire Expert Panel assigned a likelihood of failure ranking of Low for the flux thimble anti-vibration sleeves based on a consideration of the screened in degradation mechanisms. ISR/IC would

      • This record was final approved on 1/22/2019 8:25:30 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 5 of 10 LTR-AMLR-19-1, Rev. 0 January 21, 2019 reduce the anti-vibration sleeves preload, which may allow for more vibration. However, the anti-vibration sleeve would still be locked in place during operation and would still contribute to vibration reduction. SCC would also reduce preload, but the anti-vibration sleeve is not expected to become a loose part based on design, and would still be present during operation to reduce vibration. The effects of IE and VS are considered to be low due to the anti-vibration sleeves proximity to the core. IE and VS are not expected to prevent the anti-vibration sleeves function. The McGuire Expert Panel assigned a likelihood of damage ranking of Low to the flux thimble anti-vibration sleeves, since loose parts are not a concern and loss of vibration reduction would not cause core damage. Therefore, the flux thimble anti-vibration sleeves were assigned to FMECA Group 1 and Category A based on the likelihood of failure ranking of Low and the likelihood of damage ranking of Low.

3.0 NRC RAI 3

RAI-3

Regarding Applicant/Licensee Action Item 7 in the NRC safety evaluation of MRP-227-A, dated June 22, 2011 (ADAMS Accession No. ML111600498), there is new NRC staff guidance on the threshold limits for thermal embrittlement (TE) and IE of Cast Austenitic Stainless Steel (CASS). The bases for the NRC staff's new consensus on the threshold limits are described in NRC Position on Aging Management of CASS Reactor Vessel Internal Components (ADAMS Accession No. ML14163A112), as amended in the NRC staffs SE of the Boiling Water Reactor Vessel Internals Project (BWRVIP)-234, Thermal Aging and Neutron Embrittlement Evaluation of Cast Austenitic Stainless Steel for BWR Internals, June 22, 2016 (ADAMS Accession No. ML16096A002).

a. Please address any difference between the new guidance and the evaluation performed for McGuire, Units 1 and 2. In particular, please address, the new screening guidelines of CASS materials for loss of fracture toughness of highly irradiated components (i.e., components susceptible to IE), in addition to TE. If any changes to the evaluation are necessary, please submit the re-evaluation, if not, please explain why not. This evaluation could affect some of the CASS components that are listed as susceptible to TE in Tables 6-2 and 6-3 of Attachment 1 of the December 13, 2017, letter, for McGuire, Units 1 and 2, respectively.
b. The NRC staff notes that MRP-191, Rev. 1 has not been reviewed by the NRC. Therefore, placing the following CF8 components in Category A in accordance with MRP-191, Rev. 1, is not considered by the NRC staff as having sufficient support:

CF8 upper guide tube enclosures CF8 intermediate flanges CF8 Brackets, clamps, terminal blocks, and conduit straps Please demonstrate that the screening and the failure modes, effects and criticality analyses and ranking considerations in MRP-191, Rev. 1 are equivalent or better than MRP-191, Rev. 0, to support the Category A determination for these three RVI components.

      • This record was final approved on 1/22/2019 8:25:30 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 6 of 10 LTR-AMLR-19-1, Rev. 0 January 21, 2019

c. Section 6.2.7 states, The lower support column bodies are not CASS material for either McGuire Unit 1 or Unit 2. WCAP-17397-NP, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Salem Nuclear Generating Station, Salem 1, states that, The lower internals assembly column cap is a CASS piece welded onto the top of the core support column shaft. These two pieces together constitute the lower internals assembly - column body. For Salem 1, only the lower support column caps were identified as CASS. To ensure that McGuire units do not have the same lower support column bodies as Salem 1, please confirm that the lower support column caps are not CASS material. If this cannot be confirmed, please provide an explanation of how aging degradation due to TE and IE of the lower support column caps is being managed and will be managed during the period of extended operation because the NRC staffs initial review indicated that, in addition to TE, the lower support column caps are also susceptible to IE.

Response

Part a:

The McGuire Unit 1 and Unit 2 CASS evaluation within the McGuire Units 1 and 2 Aging Management Program Plan (AMP) documented in WCAP-18265 [1] relies on the scoping and screening evaluations conducted for MRP-191 Revision 0 [5] and Revision 1 [9]. MRP-191 Revision 0 and Revision 1 reference the screening criteria of MRP-175 Revision 0 [6] for irradiation embrittlement (IE) and thermal embrittlement (TE) of CASS material components. MRP-175 Revision 0 includes the degradation mechanism screening for IE and TE. The new guidance on screening criteria for IE and TE provided within the safety evaluation (SE) of BWRVIP-234 [7] is almost the same as those provided in MRP-175 Revision 0. The only relevant difference is the increase of the TE threshold for centrifugally cast materials with low molybdenum (Mo) content (<0.5 wt%) from 20% ferrite content (MRP-175, Revision 0 [6]) to 25% ferrite content (SE on BWRVIP-234 [7]). Since the new guidance is equivalent to or slightly relaxed from the guidance used in the McGuire Units 1 and 2 CASS evaluations, a re-evaluation of the McGuire CASS components is not necessary.

Part b:

The MRP-191 Revision 1 Expert Panel applied the current FMECA approach for Westinghouse designed plants as described within Section 6.2 of MRP-191 Revision 0 [5]. By applying Section 6.2 of [5], the MRP-191 Revision 1 Expert Panel used the same process, definitions, and inputs as the MRP-191 Revision 0 Expert Panel. The MRP-191 Revision 1 Expert Panel was composed of experts filling the following FMECA Roles:

Component design, testing, and repair Structural modeling and analysis Thermal-hydraulics and systems analysis Neutron fluence and radiation analysis Materials degradation and failure experience Component inspection experience Risk assessment Inspection requirements

      • This record was final approved on 1/22/2019 8:25:30 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 7 of 10 LTR-AMLR-19-1, Rev. 0 January 21, 2019 System function and operating experience Licensing and regulatory interaction The MRP-191 Revision 1 Expert Panel required consistent definitions for some of the key terms and concepts applied throughout assigning likelihood and consequence levels to the RVI components.

Therefore, the following definitions and categories from MRP-191 Revision 0 [5] were used throughout the FMECA:

Component Failure: Material degradation of a given component by one or more credible degradation mechanisms, as identified in the Screening Evaluation, causes the component to lose its ability to perform its intended design function either during normal operation or under accident conditions. Accident conditions include design basis earthquake or pipe break with no credit for the low likelihood of these accidents actually occurring. Cosmetic wear, craze cracking, and plastic deformation (excluding springs) were not considered failures.

Failure Likelihood: The likelihood that component failure(s) will occur during 60 years of operation.

The four categories of failure likelihood are defined in Table 6-2 of [5].

Core Damage: Physical damage to one or more fuel assemblies or other internals components - either through direct impact with the fuel, flow-jetting, loss of core support / fuel spring hold-down force, loose parts, blockage / diversion of coolant flow, or loss of insertion ability for more than one control rod that would impair the ability to safely shut down the reactor.

Damage Likelihood: The conditional likelihood that component failure(s) results in core damage given that the failure occurs irrespective of the actual failure likelihood. The four categories of conditional damage likelihood are defined in Table 6-3 of [5].

Based on failure likelihood and damage likelihood, the CF8 upper guide tube enclosures and CF8 brackets, clamps, terminal blocks, and conduit straps were assigned a FMECA Group, as seen in Table 2, which is Table 6-4 in [5]. The CF8 intermediate flanges were evaluated within MRP-191 Revision 0 [5]

and therefore, the likelihoods and FMECA Group were applied as identified in [5].

      • This record was final approved on 1/22/2019 8:25:30 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 8 of 10 LTR-AMLR-19-1, Rev. 0 January 21, 2019 Table 2: Reactor Internals FMECA (Significance) Groups Used by the MRP-191, Revision 0 and MRP-191, Revision 1 Expert Panels Consequence (Damage Likelihood)

Failure Likelihood Low Medium High High 2 3 3 Medium 1 2 3 Low 1 1 2 None 0 0 0 The MRP-191 Revision 1 Expert Panel discussed the component design, key parameters, and function of the upper guide tube enclosures and the brackets, clamps, terminal blocks, and conduit straps fabricated from CF8 material. The MRP-191 Revision 1 Expert Panel then reviewed the geometry, location, and function of each component, identified screened-in degradation mechanisms for each component, and discussed relevant information from previous Westinghouse PWROG sponsored FMECAs and the MRP Issue Management Table [8]. A likelihood of failure was determined for each component based on the component review in addition to a review of degradation and failure experience. A likelihood of damage was determined for each component based on a review of the effects and consequences of degradation and failure.

The CF8 upper guide tube enclosures were determined to screen in for stress corrosion cracking (SCC) of the weld, like their 304 stainless steel counterpart in MRP-191 Revision 0 [5], and additionally screen in for TE and IE due to the CF8 material1. The MRP-191 Revision 1 Expert Panel determined that the addition of potential embrittlement for the enclosures fabricated from CF8 will have no effect on the failure and damage likelihoods given to the 304 stainless steel upper guide tube enclosures within MRP-191 Revision 0 [5]. Therefore, the MRP-191 Revision 1 Expert Panel assigned a likelihood of failure ranking of Low and a likelihood of damage ranking of Medium to the CF8 upper guide tube enclosures. Based on these FMECA rankings, the CF8 upper guide tube enclosures were assigned to FMECA Group 1.

The CF8 intermediate flanges were evaluated within MRP-191 Revision 0 [5]. This component screened in for SCC of the weld, fatigue, and TE. The CF8 intermediate flanges were assigned a likelihood of 1

The upper guide tube enclosures were conservatively screened in for IE within MRP-191 Revision 1 [5].

However, this has separately been determined to be overly conservative due to the components distance from the core [10]. Despite this, the more conservative assumption was applied for the CASS evaluation supporting the McGuire AMP.

      • This record was final approved on 1/22/2019 8:25:30 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 9 of 10 LTR-AMLR-19-1, Rev. 0 January 21, 2019 failure ranking of Low, a likelihood of damage ranking of Medium, and were assigned to FMECA Group 1 [5].

The CF8 brackets, clamps, terminal blocks, and conduit straps - conduit positioners were determined to screen in for TE due to the CF8 material. The conduit positioners did not screen in for IE due to proximity from the core. The MRP-191 Revision 1 Expert Panel assigned a likelihood of failure ranking of Low to the CF8 conduit positioners, because TE alone would not cause a loss of thermocouple function, since a cracking mechanism would be required. The MRP-191 Revision 1 Expert Panel assigned a likelihood of damage ranking of Low to the CF8 conduit positioners since a thermocouple failure would be detected as part of normal plant operation. Based on these FMECA rankings, the CF8 brackets, clamps, terminal blocks, and conduit straps were assigned to FMECA Group 1.

Part c:

McGuire Unit 1 and Unit 2 do not have lower support column caps. The lower support columns at McGuire Unit 1 and Unit 2 are fabricated from one piece of Type 304 stainless steel. The lower support column design at Salem is different from the design at McGuire Unit 1 and Unit 2; therefore, the component, lower support column cap, does not apply to either McGuire unit.

      • This record was final approved on 1/22/2019 8:25:30 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 10 of 10 LTR-AMLR-19-1, Rev. 0 January 21, 2019 References

1. Westinghouse Report, WCAP-18265-NP, Rev. 0, Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals (Application to Implement MRP-227-A), December 2017.
2. Duke Energy Letter, MNS-17-050, Duke Energy Carolinas, LLC (Duke Energy) McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 Review Request for the Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A, December 13, 2017. (U. S. NRC ADAMS Accession No. ML17356A177)
3. Duke Energy Letter, MNS-18-029, Duke Energy Carolinas, LLC (Duke Energy) McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 Resolution of Commitments related to Review Request for the Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A, May 9, 2018. (U. S. NRC ADAMS Accession No. ML18135A071)
4. Email Correspondence from Michael Mahoney to Art Zaremba, Request for Additional Information - McGuire Nuclear Station, Units 1 and 2 - MRP-227 Review (EPID L-2017-LLA-0414), December 18, 2018. (U. S. NRC ADAMS Accession No. ML18352A805)
5. Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191). EPRI, Palo Alto, CA: 2006. 1013234.
6. Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175). EPRI, Palo Alto, CA: 2005. 1012081.
7. U.S. Nuclear Regulatory Commission Document, Final Safety Evaluation of the BWRVIP-234:

Thermal Aging and Neutron Embrittlement Evaluation of Cast Austenitic Stainless Steel for BWR Internals (TAC No. ME5060), June 22, 2016. (U. S. NRC ADAMS Accession No. ML16096A002)

8. Materials Reliability Program: Pressurized Water Reactor Issue Management Table, PWR-IMT Consequence of Failure (MRP-156). EPRI, Palo Alto, CA: 2005. 1012110.
9. Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191, Revision 1). EPRI, Palo Alto, CA: 2016. 3002007960.
10. Westinghouse Letter, LTR-RIAM-16-17, Rev. 1, Summary Letter for MRP-191, Revision 0 Update to MRP-191, Revision 1, August 30, 2016. (Westinghouse Proprietary)
11. Duke Energy License Renewal Application, Application to Renew the Operating Licenses of McGuire Nuclear Station, Units 1 & 2 and Catawba Nuclear Station, Units 1 & 2, June 2001.
      • This record was final approved on 1/22/2019 8:25:30 AM. (This statement was added by the PRIME system upon its validation)

LTR-AMLR-19-1 Revision 0 Proprietary Class 3

    • This page was added to the quality record by the PRIME system upon its validation and shall not be considered in the page numbering of this document.**

Approval Information Author Approval Terek Taylor Jan-21-2019 10:23:40 Verifier Approval Szweda Karli N Jan-21-2019 10:27:40 Reviewer Approval Mckinley Joshua K Jan-21-2019 11:13:32 Manager Approval Fici Melanie R Jan-21-2019 14:04:55 Hold to Release Approval Terek Taylor Jan-22-2019 08:25:30 Files approved on Jan-22-2019

      • This record was final approved on 1/22/2019 8:25:30 AM. (This statement was added by the PRIME system upon its validation)

U.S. Nuclear Regulatory Commission RA-19-0005, Attachment 2 Attachment 2 Framatome Affidavit

U.S. Nuclear Regulatory Commission RA-19-0005, Attachment 3 Attachment 3 Response to RAI 2: Framatome letter, Response to Requests for Additional Information for Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A, dated January 2019 (Proprietary).

Note: Text that is within bolded brackets is proprietary to Framatome.

U.S. Nuclear Regulatory Commission RA-19-0005, Attachment 4 Attachment 4 Response to RAI 2: Framatome letter, Response to Requests for Additional Information for Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A, dated January 2019 (Redacted).

Note: Text that is within bolded brackets is proprietary to Framatome and has been removed.

ANP-3750NP Revision 0 Copyright © 2019 Framatome Inc.

All Rights Reserved

Framatome Inc. ANP-3750NP Revision 0 Response to Requests for Additional Information for Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A Technical Report Page i Nature of Changes Section(s)

Item or Page(s) Description and Justification 1 All Initial Issue

Framatome Inc. ANP-3750NP Revision 0 Response to Requests for Additional Information for Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A Technical Report Page ii Contents Page

1.0 INTRODUCTION

AND

SUMMARY

................................................................... 1-1 2.0 FRAMATOME INC. RESPONSE TO NRC RAI 2 .............................................. 2-1 2.1 Statement of RAI 2 ................................................................................. 2-1 2.2 Response to RAI 2 .................................................................................. 2-1

3.0 REFERENCES

.................................................................................................. 3-1

Framatome Inc. ANP-3750NP Revision 0 Response to Requests for Additional Information for Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A Technical Report Page iii List of Figures Figure 2-1 Stress Ratio versus Dose .......................................................................... 2-4

Framatome Inc. ANP-3750NP Revision 0 Response to Requests for Additional Information for Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A Technical Report Page iv Nomenclature Acronym Definition AMP Aging Management Program BOL Beginning of Life CW Cold-worked dpa Displacements Per Atom EOL End of Life IC Irradiation-Enhanced Creep IE Irradiation Embrittlement ISR Irradiation-Enhanced Stress Relaxation MRP Materials Reliability Program NRC Nuclear Regulatory Commission PWR Pressurized Water Reactor RAI Request for Additional Information SS Stainless Steel US United States

Framatome Inc. ANP-3750NP Revision 0 Response to Requests for Additional Information for Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A Technical Report Page v ABSTRACT By submittals dated December 13, 2017 and May 9, 2018, Duke Energy submitted for U. S. Nuclear Regulatory Commission (US NRC) staff review the Aging Management Program (AMP) and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals (Application to Implement MRP-227-A). The US NRC has issued Requests for Additional Information (RAIs) on this submittal. This report provides the Framatome Inc. response to RAI 2.

Framatome Inc. ANP-3750NP Revision 0 Response to Requests for Additional Information for Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A Technical Report Page 1-1

1.0 INTRODUCTION

AND

SUMMARY

By submittals dated December 13, 2017 (References 1, 2) and May 9, 2018 (References 3, 4), Duke Energy submitted for US NRC staff review the Aging Management Program (AMP) and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals (Application to Implement MRP-227-A). The US NRC has issued RAIs on this submittal (Reference 5) and this report provides the Framatome Inc. response to RAI 2.

Information considered proprietary to Framatome Inc. is marked in square brackets.

Framatome Inc. ANP-3750NP Revision 0 Response to Requests for Additional Information for Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A Technical Report Page 2-1 2.0 FRAMATOME INC. RESPONSE TO NRC RAI 2 2.1 Statement of RAI 2 Regarding the core barrel and lower former plate plugs, Attachment 2 of the May 9, 2018 supplement states that the McGuire core barrel and lower former plate plugs is made of Type 316L SS [stainless steel]; and irradiation embrittlement (IE),

irradiation-enhanced stress relaxation and creep (ISR/IC), fatigue, and wear were identified as degradation mechanisms for them through the process of MRP-191. The NRC staff notes that the interface (contact) pressure between the plugs and the hosting components could be lost due to stress relaxation. To address this, the attachment states that the minimum required stress ratios were calculated for the core barrel plugs and the lower former plate plugs based on their respective loads and required interface pressure for 60 years. The licensee states in section 3.3.2 of Attachment 2 of the May 9, 2018, letter, these stresses ratios were less than the estimated stress ratios based on laboratory studies for these two types of plugs.

Please provide detailed information and data regarding these laboratory studies and explain why the data can be used in this application for the McGuire core barrel and lower former plate plugs.

2.2 Response to RAI 2 For evaluation of the stress ratio of end of life (EOL) interface pressure to beginning of life (BOL) interface pressure for the lower former plate and core barrel plugs at 60 years, recent data were reviewed. Appendix H of MRP-175, Revision 1 (Reference 6) was consulted for recent sources of irradiation-enhanced stress relaxation data. These sources were identified as:

Framatome Inc. ANP-3750NP Revision 0 Response to Requests for Additional Information for Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A Technical Report Page 2-2

1. Obata et al. (Reference 7)

This paper evaluates ISR of residual stress, which is formed in solution annealed Type 304 stainless steel by welding. Data were provided up to about 2E21 n/cm2 and reported to be about 3.6 dpa.

2. Takakura et al. (Reference 8)

This paper evaluates the irradiation-enhanced creep rate of cold-worked (CW)

Type 316 material. Data were provided up to about 1 dpa (about1 6.67E20 n/cm2).

3. Garnier et al. (Reference 9)

This paper evaluates the deformation under combined irradiation and constant applied stress of solution annealed Type 304L and CW Type 316 materials. Data were provided to 120 dpa (about1 8.0E22 n/cm2).

4. Foster et al. (Reference 10)

This paper evaluates the relationship between stress relaxation and irradiation creep and develops a method that may be used to calculate stress relaxation using the CW Type 316 irradiation creep data. Data were provided up to 1.9 dpa (about1 1.27E21 n/cm2).

5. Ishiyama et al. (Reference 11)

This paper obtains fundamental data of ISR of stainless steels and evaluates relaxation behavior by using bent beam and C-ring specimens. The specimens were fabricated from solution annealed Type 304, 316L, and XM-19 stainless steels and irradiated up to 4.43 dpa (about 2.5E21 n/cm2).

1 Using a conversion factor of 15 dpa = 1E22 n/cm2.

Framatome Inc. ANP-3750NP Revision 0 Response to Requests for Additional Information for Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A Technical Report Page 2-3 Data from Causey et al. (Reference 12) were also considered as this reference [

]

Three of these references (Foster et al., Ishiyama et al., and Causey et al.) provide values of stress ratio based on dose. From these three references, [

] Therefore, based on the dose and material type, the data [ ] were selected to calculate the stress ratio of EOL interface pressure to BOL interface pressure for the lower former plate and core barrel plugs at 60 years.

Framatome Inc. ANP-3750NP Revision 0 Response to Requests for Additional Information for Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A Technical Report Page 2-4 Figure 2-1 Stress Ratio versus Dose

[ ] the stress ratio of EOL interface pressure to BOL interface pressure for the lower former plate and core barrel plugs was calculated for each McGuire unit.

Framatome Inc. ANP-3750NP Revision 0 Response to Requests for Additional Information for Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A Technical Report Page 3-1

3.0 REFERENCES

1. Letter MNS-17-050, Duke Energy Carolinas, LLC (Duke Energy),

McGuire Nuclear Station, Units 1 and 2, Docket Nos. 50-369 and 50-370, Review Request for the Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A, December 13, 2017, NRC Accession Number ML17356A177.

2. Attachment 1 (to Letter MNS-17-050), Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals (Application to Implement MRP-227-A), WCAP-18265-NP, Rev. 0 (Non-Proprietary), NRC Accession Number ML17356A178.
3. Letter MNS-18-029, Duke Energy Carolinas, LLC (Duke Energy),

McGuire Nuclear Station, Units 1 and 2, Docket Nos. 50-369 and 50-370, Resolution of Commitments related to Review Request for the Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A, May 9, 2018, NRC Accession Number ML18135A071.

4. Attachment 2 (to Letter MNS-18-029), Framatome ANP-3658 Revision 0, Evaluation of the McGuire Units 1 and 2 Upflow Modification for 60 Years, Technical Report, NRC Accession Number ML18135A072.
5. Email from Michael Mahoney (US NRC) to Art Zaremba (Duke Energy),

Request for Additional Information - McGuire Nuclear Station, Units 1 and 2 - MRP-227 Review (EPID L-2017-LLA-0414), December 18, 2018, NRC Accession Number ML18352A805.

6. Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175, Revision 1).

EPRI, Palo Alto, CA: 2017. 3002010268.

Framatome Inc. ANP-3750NP Revision 0 Response to Requests for Additional Information for Aging Management Program and Inspection Plan for the McGuire Nuclear Station Units 1 and 2 Reactor Vessel Internals to Implement MRP-227-A Technical Report Page 3-2

7. Obata, M., et al., Radiation-Induced Stress Relaxation of Welded Type 304 Stainless Steel Evaluated by Neutron Diffraction, Effects of Radiation on Materials: 22nd Symposium, American Society for Testing and Materials, Vol. 1475 STP, 2006, p. 15.
8. Takakura, K., et al., In-Pile Creep Behavior of Type 316 Stainless Steel at Halden Reactor, International Symposium: Contributions of Materials Investigations to Improve the Safety and Performance of LWRs, Fontevraud 7, SFEN, 2010.
9. Garnier, J., et al., Irradiation Creep of SA 304L and CW316 Stainless Steels: Mechanical Behavior and Microstructural Aspects. Part I:

Experimental Results, Journal of Nuclear Materials, Vol. 413 (2011), pp.

63-69.

10. Foster, J.P., Gilbert, E.R., Bunde, K., and Porter, D.L., Relationship Between In-Reactor Stress Relaxation and Irradiation Creep, Journal of Nuclear Materials, 252 (1998), pp. 89-97.
11. Ishiyama, Y., et al., Stress Relaxation Caused by Neutron-Irradiation at 561 K in Austenitic Stainless Steels, 11th International Conference on Environmental Degradation of Materials in Nuclear Power Systems -

Water Reactors, ANS, 2003.

12. Causey, A. R., et al., In-Reactor Stress Relaxation of Selected Metals and Alloys at Low Temperatures, Journal of Nuclear Materials 90 (1980) pp. 216-223.
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