Text

PWROG-14083-NP, Revision 1, NextEra Energy, Seabrook Unit 1, Summary Report for Applicant/Licensee Action Items 1, 2, and 7.
	ML15149A280
Person / Time
Site:	Seabrook
Issue date:	05/31/2015
From:	Carpenter B, Molkenthin J, Paesano P, Mary Walsh; PWR Owners Group, Westinghouse
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15149A278	List: ML15149A280; ML15149A282; ML15149A283; SBK-L-15073, Responses to Applicant/Licensee Action Items for the Inspection and Evaluation Guidelines for Pressurized Water Reactor Vessel Internals (MRP-227-A);
References
	MRP-227-A, SBK-L-15073, TAC ME4028 PWROG-14083-NP, Rev. 1
	Download: ML15149A280 (16)
	v • d • e

Enclosure 6 to SBK-L-15073 Non-Proprietary Class 3 Westinghouse Report (PWROG-14083-NP, Revision 1)

NextEra Energy, Seabrook Unit 1 Summary Report for Applicant/Licensee Action Items 1, 2, and 7 (Non-Proprietary)

PRRES S URI ZED WAT ER R. E A C TO R OW-N EýR S,,-G R O U P WPWROG-14083-NP PWROG- Revision 1 Qners WESTINGHOUSE NON-PROPRIETARY CLASS 3 NextEra Energy, Seabrook Unit 1 Summary Report for Applicant/Licensee Action Items 1, 2, and 7 Materials Committee PA-MSC-0983 Revision 0 Tasks 3, 4, and 5:

May 2015

(*) Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 PWROG-14083-NP Revision I NextEra Energy, Seabrook Unit 1 Summary Report for Applicant/Licensee Action Items 1, 2, and 7 PA-MSC-0983 Revision 0 Tasks 3, 4, and 5 Mary Ann T. Walsh*

Reactor Internals Aging Management May 2015 Verifier: Bradley T. Carpenter*

Reactor Internals Aging Management Approved: Patricia C. Paesano*, Manager Reactor Internals Aging Management Approved: James P. Molkenthin*, Program Director PWR Owners Group PMO

Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii LEGAL NOTICE This report was prepared as an account of work performed by Westinghouse Electric Company LLC. Neither Westinghouse Electric Company LLC, nor any person acting on its behalf:

1. Makes any warranty or representation, express or implied including the warranties of fitness for a particular purpose or merchantability, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or

2. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this report.

COPYRIGHT NOTICE This report has been prepared by Westinghouse Electric Company LLC and bears a Westinghouse Electric Company copyright notice. Information in this report is the property of and contains copyright material owned by Westinghouse Electric Company LLC and /or its subcontractors and suppliers. It is transmitted to you in confidence and trust, and you agree to treat this document and the material contained therein in strict accordance with the terms and conditions of the agreement under which it was provided to you.

DISTRIBUTION NOTICE This report was prepared for the PWR Owners Group. This Distribution Notice is intended to establish guidance for access to this information. This report (including proprietary and non-proprietary versions) is not to be provided to any individual or organization outside of the PWR Owners Group program participants without prior written approval of the PWR Owners Group Program Management Office. However, prior written approval is not required for program participants to provide copies of Class 3 Non-Proprietary reports to third parties that are supporting implementation at their plant, and for submittals to the NRC.

PWROG-14083-NP May 2015 Revison 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iv TABLE OF CONTENTS 1 PURPOSE AND BACKGROUND ................................................................................. 1-1 2 SUPPORT FOR ADDRESSING MRP-227-A SAFETY EVALUATION A/LAI 1,2, AND 7 FO R S EA BR O O K UNIT 1............................................................................................. 2-1 2.1 AILAI 1: APPLICABILITY OF FMECAAND FUNCTIONALITY ANALYSIS A S S UMPT IO NS ................................................................................................ 2-1 2.2 A/LAI 2: PWR VESSEL INTERNAL COMPONENTS WITHIN THE SCOPE OF LIC E NS E R E NEWA L........................................................................................ 2-4 2.3 A/LAI 7: PLANT-SPECIFIC EVALUATION OF CASS MATERIALS ................... 2-5 3 R E F E R E NC E S ............................................................................................................. 3-1 May 2015 PWROG-1 4083-NP PWROG-14083-NP May 2015 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 V LIST OF TABLES Table 2-1 Summary of Seabrook Unit 1 CASS Components and Their Susceptibility to Therm al Em brittlem ent ..................................................................................... 2-6 PWROG-14083-NP May 2015 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-1 I PURPOSE AND BACKGROUND The scope of this effort was authorized by NextEra Energy through (Pressurized Water Reactor Owner's Group [PWROG]) Project Authorization PA-MSC-0983, Revision 0, Tasks 3, 4, and 5.

NextEra Energy Seabrook has committed to develop a Reactor Vessel Internals Inspection Program for the Seabrook Unit 1 reactor vessel internals (RVI) listed in the License Renewal Application (LRA) [1]. As a result of the U.S. Nuclear Regulatory Commission (NRC) review of MRP-227, Revision 0 [2] and subsequent Safety Evaluation (SE), Revision 1, additional requirements have been specified in MRP-227-A [3].

These responses supporting demonstration of Seabrook Unit 1 compliance with the new requirements, specifically the Applicant/Licensee Action Items (A/LAI) 1, 2, and 7, per PA-MSC-0983, Tasks 3, 4, and 5 are in subsections 2.1, 2.2, and 2.3, respectively.

The NextEra Energy Seabrook response to requests for additional information (RAI) includes revised AMR items for reactor vessel internals (revised LRA Table 3.1.2-3) [12, Enclosure 3].

The table lists baffle-edge bolts, cast lower core support column bodies and a thermal shield; however, the Westinghouse records show that these components are not present in Seabrook Unit 1. Seabrook Unit I design does not include baffle-edge bolts; Seabrook Unit 1 design has wrought lower core support column bodies and neutron panels.

The response to A/LAI 7 addresses the susceptibility to thermal embrittlement of the Seabrook Unit 1 cast austenitic stainless steel (CASS) component, the lower internals assembly - bottom-mounted instrumentation (BMI) column assemblies, BMI column cruciforms. Based on data from certified material test reports and guidance for the application of Hull's formula [10], the ferrite content for each of the 26 cruciform castings was calculated. For customer information, the BMI column cruciforms have a minimum ferrite content of 5.5 percent, a maximum of 15.9 percent, and a mean ferrite content of 10.2 percent, where the mean is calculated using the 26 calculated ferrite content values as unique data points. Thus, the known ferrite content values for Seabrook Unit 1, BMI column cruciforms are well below the 20 percent threshold listed in [9, Table 2].

In response to customer comment and request, a material comparison table is in Attachment 1.

Revision I is to resolve customer comments on Revision 0.

PWROG-14083-NP May 2015 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1 2 SUPPORT FOR ADDRESSING MRP-227-A SAFETY EVALUATION A/LAI 1, 2, AND 7 FOR SEAfROOK UNIT 1 2.1 A/LAI 1: APPLICABILITY OF FMECA AND FUNCTIONALITY ANALYSIS ASSUMPTIONS The SE A/LAI 1: Applicability of FMECA and Functionality Analysis Assumptions text from MRP-227-A states:

"As addressed in Section 3.2.5.1 of this SE, each applicant/licenseeis responsible for assessing its plant's design and operatinghistory and demonstrating that the approvedversion of MRP-227 is applicable to the facility Each applicant/licenseeshall refer, in particular,to the assumptions regardingplant design and operatinghistory made in the FMECA and functionality analyses for reactorsof their design (i.e., Westinghouse, CE, or B&V) which supportMRP-227 and describe the process used for determining plant-specificdifferences in the design of their RVI components or plant operatingconditions, which result in different component inspection categories.The applicant/licenseeshall submit this evaluation for NRC review and approval as part of its application to implement the approved version of MRP-227. This is Applicant/Licensee Action Item 1"[3].

Seabrook Unit I Response to AILAI I The process used to provide reasonable assurance that the RVI components at Seabrook Unit 1 are reasonably represented by the generic industry program assumptions (with regard to neutron fluence, temperature, stress values, and materials used in the development of MRP-227-A) [3] is:

1. Identification of typical Westinghouse-designed pressurized water reactor (PWR)

RVI components (MRP-191 [4, Table 4-4]).

2. Identification of Seabrook Unit 1 RVI components.

3. Comparison of the typical Westinghouse designed PWR RVI components to the Seabrook Unit 1 RVI components:

a. Confirmation that no additional items were identified by this comparison (primarily supports A/LAI 2).

b. Confirmation that the materials from MRP-1 91 [4, Table 4-4] are consistent with Seabrook Unit 1 RVI component materials.

c. Confirmation that the design and fabrication of Seabrook Unit 1 RVI components are the same as, or equivalent to, the typical Westinghouse-designed PWR RVI components.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2

4. Confirmation that the Seabrook Unit 1 operating history is consistent with the assumptions in MRP-227-A [3] regarding core loading patterns and base load operation.

5. Confirmation that the Seabrook Unit 1 RVI materials operated at temperatures within the original design basis parameters.

6. Determination of stress values based on design basis documents.

7. Confirmation that any changes to the Seabrook Unit 1 RVI components do not impact the application of the MRP-227-A [3] generic aging management strategy.

The Seabrook Unit 1 RVI components are reasonably represented by the design and operating history assumptions regarding neutron fluence, temperature, materials, and stress values in the MRP-191 generic failure modes, effects, and criticality analysis (FMECA) [4], and in the MRP-232 [7] functionality analysis based on the following:

1. Seabrook Unit 1 operating history is consistent with the assumptions in MRP-227-A [3]

with regard to neutron fluence and fuel management.

a. The FMECA and functionality analysis for MRP-227-A [3] were based on the assumption of 30 years of operation with high-leakage core loading patterns, followed by 30 years of low-leakage core fuel management strategy. As stated in [6], Seabrook Unit 1 fuel management program changed from a high to a low-leakage core loading pattern prior to 30 years of operation. By operating with a low-leakage core design prior to 30 years, Seabrook Unit 1 meets the fluence and fuel management assumptions in MRP-191 and requirements for MRP-227-A application.

b. Seabrook Unit 1 has always operated as a base load unit. Therefore, Seabrook Unit 1 satisfies the assumptions in MRP documents regarding operational parameters affecting fluence.

2. The Seabrook Unit 1 reactor coolant system operates between Tdd and Th, [5]. T.o1d is no lower than 537.7 0 F and Thor is no higher than 621.4 0F [5, Table 5.1-1]. The design temperature for the vessel is 650°F [5, Table 5.3-1]. Therefore, Seabrook Unit 1 operating history is within original design basis parameters and is consistent with the assumptions used to develop the MRP-227-A [3] aging management strategy with regard to temperature operational parameters.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3

3. As discussed below, the Seabrook Unit 1 RVI components and materials are comparable to the typical Westinghouse-designed PWR RVI components (MRP-191 [4, Table 4-4]).

a. Components required to be in the Seabrook Unit 1 aging management program are consistent with those contained in MRP-191, Table 4-4 [4]. No additional components were identified for NextEra Energy Seabrook by this comparison.

b. Seabrook Unit 1 RVI component materials are consistent with, or equivalent to, those materials identified in MRP-1 91 [4, Table 4-4] for Westinghouse-designed plants. Where differences exist, there is no impact on the Seabrook Unit 1 RVI program or the component is already credited as being managed under an alternate Seabrook Unit 1 aging management program.

c. Design and fabrication of Seabrook Unit 1 RVI components are the same as, or equivalent to, the typical Westinghouse-designed PWR RVI components, with the exception that Seabrook Unit 1 utilizes a double-concentric thimble tube design fabricated from wear resistant, seamless nickel alloy material (INCONEL 600).

4. Seabrook Station is a single unit, 1,295 net megawatts electric Westinghouse 4-loop pressurized water reactor with a turbine generator built by General Electric. Two power uprates have been implemented since initial commercial operation. In Cycle 11 (2004),

the rated thermal power was increased to 3587 MWt, and in Cycle 12 (2006), the rated thermal power was increased to 3648 MWt [5].

The original Alloy X-750 guide tube assembly split pins were replaced with cold worked 316 stainless steel (SS) in 2006. Therefore, modifications to the Seabrook Unit 1 RVI made over the lifetime of the plant are those specifically directed by the original equipment manufacturer (OEM). The OEM has developed or evaluated design changes and satisfied assumptions for A/LAI 1. The design has been maintained over the lifetime of the plant as specified by the OEM, operational parameters are compliant with MRP-227-A [3] requirements with regard to fluence and temperature, and the components are consistent with those considered in MRP-191 [4]. The materials for the components are consistent with those considered in MRP-191 [4]. Therefore, the Seabrook Unit 1 RVI stress values are represented by the assumptions in MRP-191 [4],

MRP-227-A [3], and MRP-232 [7], confirming the applicability of the generic FMECA.

Conclusion Seabrook Unit 1 complies with A/LAI 1 of the NRC SE regarding MRP-227, Revision 0.

Therefore, the requirement is met for application of MRP-227-A as a strategy for managing age-related material degradation in the RVI components.

PWROG-14083-NP May 2015 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-4 2.2 A/LAI 2: PWR VESSEL INTERNAL COMPONENTS WITHIN THE SCOPE OF LICENSE RENEWAL NRC Applicant/Licensee Action Item 2: PWR Vessel Internal Components within the Scope of License Renewal text in MRP-227-A states:

"As discussed in Section 3.2.5.2 of this SE, consistent with the requirements addressedin 10 CFR 54.4, each applicant/licenseeis responsible for identifying which RVI components are within the scope of LR for its facility. Applicants/licensees shall review the information in Tables 4-1 and 4-2 in MRP-189, Revision 1, and Tables 4-4 and 4-5 in MRP-191 and identify whether these tables contain all of the RVI components that are within the scope of LR for their facilities in accordance with 10 CFR 54.4. If the tables do not identify all the RVI components that are within the scope of LR for its facility, the applicantor licensee shall identify the missing component(s) andpropose any necessarymodifications to the program defined in MRP-227, as modified by this SE, when submitting its plant-specific NRC-AMP The NRC-AMP shall provide assurancethat the effects of aging on the missing component(s) will be managed for the period of extended operation. This issue is Applicant/LicenseeAction Item 2." [3]

Seabrook Unit I Response to A/LAI 2 This action item requires comparison of the Seabrook Unit 1 RVI components that are within the scope of license renewal for Seabrook Unit 1 to those components contained in MRP-191 [4, Table 4-4], as Seabrook Unit 1 has a Westinghouse plant design. The components required to be in the Seabrook Unit 1 aging management program, as described in the PWR Vessel Internals Inspection Plan [12], are consistent with those contained in MRP-191, Table 4-4 [4].

No additional components were identified in the NextEra Energy Seabrook design.

Several components have different materials than that specified in MRP-191 [4, Table 4-4], but the differences have no effect on the recommended MRP aging strategy, or aging is already managed by an alternate Seabrook Unit 1 program. Seabrook Unit 1 follows the program strategy in MRP-227-A [3], with the exception of the flux thimble tubes. Therefore, as stated in the LRA [1, Appendix B, subsection B.2.0], and in the responses to RAI 3.1.1-60-01 and 3.1.1-60-02 (dated February 3, 2011), RAI 3.1.1-60-01/02 (dated April 22, 2011), and follow-up RAI 3.1.1-60-02 (dated November 2, 2011) [8, subsection 3.1.2.1.1], Seabrook does not credit the flux thimble tube inspection program for aging management. Additionally, the Safety Evaluation Report (SER) to the LRA states that the NRC staff found that Seabrook has provided an acceptable basis such that a Flux Thimble Tube Inspection Program is not needed to manage wear or cracking of the Seabrook Unit 1 flux thimble tubes [8, subsection 3.1.2.1.1].

This supports the requirement that the NRC-AMP shall provide assurance that the effects of aging on the Seabrook Unit I RVI components within the scope of license renewal will be managed for the period of extended operation.

The generic scoping and screening of the RVI, as summarized in MRP-191 [4] and MRP-232 [7], to support the inspection sampling approach for aging management of the RVI specified in MRP-227-A [3], are applicable to Seabrook Unit 1 with the exception of the aging management strategy of the Seabrook Unit 1 flux thimble tubes:

PWROG-14083-NP May 2015 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-5 Conclusion Seabrook Unit 1 complies with A/LAI 2 of the NRC SE on MRP-227, Revision 0, and therefore meets the requirement for application of MRP-227-A as a strategy for managing age-related material degradation in reactor internals components.

2.3 AILAI 7: PLANT-SPECIFIC EVALUATION OF CASS MATERIALS NRC Applicant/Licensee Action Item 7: Plant-specific Evaluation of Cast Austenitic Stainless Steel (CASS) materials text in MRP-227-A states:

"As discussed in Section 3.3.7 of this SE, the applicants/licenseesof B&W, CE, and Westinghouse reactorsare requiredto develop plant-specific analysesto be applied for their facilities to demonstrate that B&W IMI guide tube assembly spiders and CRGT spacercastings, CE lower support columns, and Westinghouse lower supportcolumn bodies will maintain their functionality during the period of extended operation or for additionalRVI components that may be fabricatedfrom CASS, martensiticstainlesssteel or precipitationhardenedstainlesssteel materials. These analyses shall also consider the possible loss of fracture toughness in these components due to thermal and irradiationembrittlement,and may also need to consider limitationson accessibilityfor inspection and the resolution/sensitivityof the inspection techniques. The requirementmay not apply to components that were previously evaluated as not requiring aging managementduring development of MRP-227. That is, the requirement would apply to components fabricatedfrom susceptible materialsfor which an individual licensee has determinedaging managementis required,for example during their review performed in accordance with Applicant/LicenseeAction Item 2. The plant-specificanalysis shall be consistent with the plant's licensing basis and the need to maintain the functionality of the components being evaluated under all licensing basis conditions of operation. The applicant/licenseeshall include the plant-specificanalysis as part of their submittal to apply the approved version of MRP-227. This is Applicant/LicenseeAction Item 7." [3]

Seabrook Unit I Response to A/LAI 7 A/LAI 7, from the NRC's final SE on MRP-227, Revision 0, notes that, for assessment of CASS materials, the applicant/licensee for license renewal may apply the criteria in the NRC letter of May 19, 2000, License Renewal Issue No. 98-0030, "Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components" [9] as the basis for determining whether the CASS materials are susceptible to the thermal aging mechanism. Ifthe application of the applicable screening criteria for the components' material demonstrates that the components are not susceptible to either thermal embrittlement (TE) or irradiation embrittlement (IE), or to the synergistic effects of TE and IE combined, then no other evaluation would be necessary. The Seabrook Unit 1 CASS reactor vessel internals components and the assessment of their susceptibility to TE are summarized in Table 2-1.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-6 The Seabrook Unit 1 lower internals assembly - bottom-mounted instrumentation (BMI) column cruciforms are ASME SA-351, Grade CF8 material. The elemental percentages, from the certified material test reports (CMTRs) for the CASS BMI column cruciform, are input into Hull's formula (per guidance of NUREG/CR-4513 [10]) to calculate the delta ferrite content of the CASS material. The CMTRs do not list the element percentage for nitrogen; thus, per the guidance of NUREG/CR-4513, nitrogen is assumed to be 0.04 percent. The CMTRs do not list an elemental percentage for molybdenum. SA-351, Grade CF8 did not have a requirement for percent molybdenum in 1971. The 2013 Edition of the American Society of Mechanical Engineers (ASME) Code has SA-351, Grade CF8 chemistry requirements that specify a maximum of 0.5 percent molybdenum [11]; thus, this maximum value is input into Hull's formula.

There are a total of 26 BMI column cruciforms - 25 standard cruciforms and 1 special cruciform.

As summarized in Table 2-1, the ferrite content is less than or equal to 20 percent; thus, based on the NRC criteria [9], the 26 Seabrook CASS BMI column cruciforms are not susceptible to TE.

The Seabrook Unit 1 reactor internals hold-down spring was manufactured from 403 SS, a martensitic stainless steel.

No martensitic precipitation hardened stainless steel (PH-SS) components were identified for the Seabrook Unit 1 reactor vessel internals.

Table 2-1 Summary of Seabrook Unit I CASS Components and Their Susceptibility to Thermal Embrittlement Calculated Susceptibility to TE Molybdenum Casting Ferrite (Based on CASS Component Content Method Content NRC letter [9])

Lower Internals Assembly, 0.5% Static < 20%(1) 26 of 26 BMI column cruciforms Maximum Not Susceptible (Standard and Special)

Note:

1. Ferrite content is based on CMTR chemistry data, nitrogen 0.04 percent, and molybdenum 0.5 percent.

Conclusion It is concluded that continued application of the MRP-227-A strategy will meet the requirement for managing age-related degradation of the Seabrook Unit 1 CASS reactor vessel internals components.

PWROG-14083-NP May 2015 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-1 3 REFERENCES

1. NextEra Energy Seabrook, LLC, "License Renewal Application, NextEra Energy Seabrook, LLC, Et Al., Docket No. 50-443, Seabrook Station, Unit No. 1, Facility Operating License No. NPF-86."

2. MaterialsReliability Program:PressurizedWater Reactor Intemals Inspection and Evaluation Guidelines (MRP-227-Rev. 0). EPRI, Palo Alto, CA: 2008. 1016596.

3. MaterialsReliability Program:PressurizedWater Reactor InternalsInspection and Evaluation Guidelines (MRP-227-A). EPRI, Palo Alto, CA: 2011. 1022863.

4. MaterialsReliability Program:Screening, Categorization,and Ranking of Reactor Internals Components for Westinghouse and Combustion EngineeringPWR Design (MRP-191). EPRI, Palo Alto, CA: 2006. 1013234.

5. NextEra Energy Seabrook, LLC, Seabrook Station, Table 5.1-1, Rev. 12 and Table 5.3-1, Rev. 8, "Updated Final Safety Analysis Report."

6. NextEra Energy Seabrook, LLC, Seabrook Station Report, "Updated Final Safety Analysis Report," Section 4.3 "Nuclear Design," Rev. 8.

7. MaterialsReliability Program:Aging Management Strategies for Westinghouse and Combustion EngineeringPWR Internals (MRP-232, Revision 1). EPRI, Palo Alto, CA:

2012. 1021029.

8. U.S. Nuclear Regulatory Commission Report, "Safety Evaluation Report WNith Open Items Related to the License Renewal of Seabrook Station," Docket Number 50-443, June 2012 (NRC ADAMS Accession No. ML12160A374).

9. NRC Letter, "License Renewal Issue No. 98-0030, "Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components"," May 19, 2000 (NRC ADAMS Accession No. ML003717179).

10. U.S. Nuclear Regulatory Commission, NUREG/CR-4513, Rev. 1, "Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems,"

August 1994 (NRC ADAMS Accession No. ML052360554).

11. ASME Boiler and Pressure Vessel Code,Section II, 2013 Edition, American Society of Mechanical Engineers.

12. NextEra Energy Seabrook, LLC Letter SBK-L-14089, "Response to Request for Additional Information Related to the Review of The Seabrook Station License Renewal Application - Set 21 (Tag No. ME4028)," June 24, 2014.

PWROG-14083-NP May 2015 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Attachment 1 Page 1 of 2 Attachment 1: Material Comparison Table PWROG-14083-NP May 2015 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Attachment 1 Page 2 of 2 Attachment 1: Seabrook Unit I RVI Material Differences from MRP-191 Assembly Subassembly Component Material MRP-191 CF8 Mixing Devices Mixing Devices Seabrook Unit 1 304 SS Upper Internals Assembly Upper Support MRP-191 CF8 Column Column Bases Assemblies Seabrook Unit 1 304 SS MRP-191 304SS Flux Tlub Thimble Philu Seabrook Unit 1 316 SS Tubes Plugs Flux Thimbles Seabrook Unit 1 A 600 (Tubes)

Flux Thimbles MRP-191 316 SS (Tubes) Seabrook Unit 1 A 600 Lower Internals Lower Support Lower Support MRP-191 304SS Assembly Column Column Bolts Assemblies Seabrook Unit 1 316 SS Neutron Neutron Panel MRP-191 304 SS Panels/Thermal Lock Caps Shield Seabrook Unit 1 304L SS MRP-191 304SS Radial Support Radial Support Keys Key Bolts Seabrook Unit 1 316 SS PWROG-14083-NP May 2015 Revision 1

v t e Letter Sequence Other
TAC:ME4028, Steam Generator Tube Integrity (Approved, Closed)
Initiation Request Acceptance... Supplement, Supplement Administration Withholding Request Acceptance, Withholding Request Acceptance Meeting, Meeting, Meeting, Meeting, Meeting Questions 9 June 2011 18 November 2010 18 November 2010 18 November 2010 14 December 2010 14 December 2010 5 January 2011 5 January 2011 5 January 2011 21 January 2011 24 February 2011 17 March 2011 7 March 2011 30 March 2011 23 May 2011 15 June 2011 14 October 2011 4 November 2011 4 November 2011 28 October 2010 26 March 2012 22 June 2012 22 June 2012 21 June 2012 18 September 2012 15 January 2014 25 April 2014 28 August 2015 2 October 2015 3 November 2015 4 December 2015 Answers 22 December 2010 17 December 2010 13 January 2011 13 January 2011 3 February 2011 18 February 2011 5 April 2011 22 April 2011 2 June 2011 24 June 2011 15 May 2014 24 June 2014 9 March 2015 23 February 2015 30 June 2015 9 October 2015 3 December 2015 23 November 2015 Results Approval, Approval, Approval, Approval, Approval, Approval, Approval Other: ML103620318, ML110270026, ML110280424, ML110970260, ML11280A045, ML11304A002, ML11304A109, ML12009A007, ML12164A756, ML12164A759, ML12214A607, ML12242A198, ML12264A648, ML13261A145, ML15149A280, ML15149A282, ML15183A020, ML15287A017, ML15307A022, ML15310A102, ML15337A047, ML15343A186, ML15349A918, ML19205A362, ML19205A365, ML19205A389, SBK-L-15073, Responses to Applicant/Licensee Action Items for the Inspection and Evaluation Guidelines for Pressurized Water Reactor Vessel Internals (MRP-227-A)
MONTHYEAR2011-02-04 [Table View]

ML15149A280

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