NRC 2014-0021, Reactor Vessel Internals Inspection Plan Response to Request for Addtional Information

Reactor Vessel Internals Inspection Plan Response to Request for Addtional Information

ML14111A050
Person / Time
Site:	Point Beach
Issue date:	04/21/2014
From:	Mccartney E Point Beach
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC 2014-0021
Download: ML14111A050 (15)

Text

NEXTeraM ENERGY~

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April 18, 2014 NRC 2014-0021 10 CFR 54 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 Reactor Vessel Internals Inspection Plan Response to Request for Additional Information

References:

(1) NextEra Energy Point Beach, LLC letter to NRC, dated December 19, 2011, License Renewal Commitment, Reactor Vessel Internals Program Submittal (ML113540301)

(2) NRC electronic mail to NextEra Energy Point Beach, LLC, dated June 7, 2012, Point Beach Units 1 and 2- Draft RAI on the Reactor Vessel Internals Inspection Plan (TAC ME8235 and ME8236)

(ML12159A113)

(3) NRC electronic mail to NextEra Energy Point Beach, LLC, dated July 10, 2012, Point Beach Nuclear Plant Units 1 and 2- Draft Request for Additional Information re: Reactor Vessel Internals Inspection Plan

{TAC Nos. ME8235 and ME8236) (ML12198A050)

(4) NextEra Energy Point Beach, LLC letter to NRC, dated August 16, 2012, Reactor Vessel Internals Inspection Plan, Response to Request for Additional Information (ML12229A580)

(5) NRC electronic mail to NextEra Energy Point Beach, LLC, dated January 31, 2013, Point Beach Nuclear Plant, Units 1 and 2- Draft Request for Additional Information (Second Round) re: Reactor Vessel Internals Inspection Plan Review (TAC Nos. ME8235 and ME8236)

(ML13036A300)

(6) NextEra Energy Point Beach, LLC letter to NRC, dated March 15, 2013, Reactor Vessel Internals Inspection Plan, Response to Request for Additional Information (ML13077A356)

NextEra Energy Point Beach, LLC (NextEra) submitted the Point Beach Nuclear Plant (PBNP) program NP 7.7.30, Reactor Vessel Internals Program, via Reference (1 ). The PBNP Reactor Vessel Internals Program is based on Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) Technical Report MRP-227, Pressurized Water Reactor (PWR)

Internals Inspection and Evaluation Guidelines, Revision 0.

NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241

Document Control Desk Page 2 Via References (2) and (3), the NRC determined additional information was required to enable the staff's continued review of the PBNP Reactor Vessel Internals Program. NextEra responded to the request via Reference (4).

In Reference (5), the NRC requested a second round of additional information in order to enable the staff's continued review of the PBNP Reactor Vessel Internals Program. In Reference (6),

NextEra partially answered the questions in Reference (5). NextEra committed to providing the remaining answers once the information was available from Westinghouse.

The enclosure to this letter provides the remaining answers to Reference (5).

Summary of Regulatory Commitments:

This letter completes the following regulatory commitment:

• NextEra Energy Point Beach, LLC will contract with Westinghouse to determine which components are not bounded by the assumptions on stress, temperature, and fluence contained in MRP-191, MRP-232, and MRP-227-A. Westinghouse will complete the plant-specific evaluation to demonstrate that the MRP-227 recommended inspections will ensure functionality of these components until the next scheduled inspection. This information is expected to be available by January 31, 2014. This information will be provided to the NRC within 45 days of receipt from Westinghouse.

This letter contains no additional regulatory commitments or any changes to existing commitments.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on April 18, 2014.

Very truly yours, NextEra Energy Point Beach, LLC Eric McCartney Site Vice President Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW

ENCLOSURE NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 REACTOR VESSEL INTERNALS INSPECTION PLAN RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION The NRC staff determined that additional information was required (Reference 11) to enable the continued review of the Point Beach Nuclear Plant (PBNP) Reactor Vessel Internals (RVI)

Program (Reference 12). The following information is provided by NextEra Energy Point Beach, LLC (NextEra) in response to the NRC staff's request.

RAI-1a In response to RAI-1, Item 4, the licensee stated:

The PBNP RVI components are bounded by the typical Westinghouse PWR internals components outlined in MRP-227-A and the applicable referenced documents, including MRP-191 and MRP-232. The PBNP reactor vessel internals inspection program was written to comply with MRP-227-A. No changes in the inspection requirements are being proposed at this time in that the PBNP inspection program complies with MRP-227-A as indicated above. "

The licensee did not mention that an extended power uprate (EPU) for each unit was approved by the NRC on May 5, 2011. The MRP has stated that the inspection recommendations are applicable to all U.S. PWR [pressurized water reactor] operating plants as of May 2007. In discussions with the MRP, the MRP indicated that EPUs after 2007 were not considered during the development of the MRP-227 recommendations.

The NRC staff reviewed the licensee's 2009 submittal for the EPU (ADAMS Accession No. ML091250566) and noted several instances for which the stress, fatigue usage, and neutron fluence values provided in the EPU evaluation appear to exceed the screening values of MRP-191 and, therefore, may not be bounded by the assumptions made in development of MRP-227-A. The following examples are noted:

1. Table 2.2.3-3 of the 2009 EPU submittal lists the stresses and fatigue usage factors (CUF) of different core support structures. For the upper core plate alignment pins, the document includes values of 51.48 ksi for stress and 0.30 for the CUF before the EPU, and 38.507 ksi for stress and 0. 583 for the CUF after the EPU with an ASME Code-allowable stress intensity of 34.44 ksi.

The licensee 's August 16, 2012, response to RAI-1, Item 2, lists that component as having an effective stress< 30 ksi, matching the stress for that component in MRP-191.

2. Section 2. 1.4. 2. 2 of the 2009 EPU submittal lists the maximum fast neutron fluence (E > 1. 0 MeV) for 54 EFPY as 8.83E+22 n/cm 2 for Unit 1 and 8. 77E+22 n!cm 2 for Unit 2, whereas MRP-191 states that the maximum fluence values for any internal component is 5E+22 n/cm2*

Requested Information The NRC staff requests the licensee review its current licensing basis (CLB), including the 2009 EPU submittal, and list all components that are not bounded by the assumptions on stress, temperature, and fluence contained in MRP-191, MRP-232, and MRP-227-A. Evaluate the need for Page 1 of 13

changes to the inspection requirements and/or inspection frequency for any components that are not bounded by the screening values of fluence, temperature, and stress. Describe the process used to perform these evaluations. Provide a technical justification for either changing or not changing the inspection requirements and/or inspection frequency.

NextEra Response In support of the extended power uprate (EPU) submittal process, a scoping study was performed on potential adverse impacts of the PBNP Units 1 and 2 (WEP/WIS) EPU program on the plant-specific implementation of the industry generic recommendations for reactor internals inspections to manage aging.

This study considered the PBNP Units 1 and 2 plant-specific and MRP-227 design and operational characteristics including:

• Physical characteristics o Geometry o Materials

• Systematic or operational parameters o Performance Capability Working Group (PCWG) parameters o Transients o Neutronics

• Plant and industry history

• MRP-227 applicability criteria o Less than 30 years of operation with high-leakage core loading patterns (fresh fuel assemblies loaded in peripheral locations) followed by implementation of a low-leakage fuel management strategy for the remaining 30 years of operation, o Base load operation, i.e., plant typically operates at fixed power levels and does not usually vary power on a calendar or load demand schedule, o No design changes beyond those identified in general industry guidance or as recommended by the original vendors design, o Applicable to all operating U.S. pressurized water reactors (PWRs) operating as of May 2007.

In all cases, the PBNP Units 1 and 2 EPU parameters were directly bounded by the original design basis, or shown to be acceptable for the period of life extension (60 years) under EPU conditions.

Demonstration of compliance relative to the plant-specific design criteria or governing regulatory requirements under EPU conditions was performed through detailed studies and reconciliatory design investigations. The study concluded that the information that forms the basis for the MRP-227 inspection requirements related to aging or design compliance remains reasonable and applicable for the EPU parameters.

The study concluded that there was no impact, adverse or otherwise, from the PBNP Units 1 and 2 EPU program on the plant-specific implementation of the MRP-227 requirements.

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With regards to the 2 examples noted:

1. For the upper core plate alignment pins, the stresses noted in Table 2.2.3-3 of the 2009 EPU submittal (Reference 13) are for the most severe, normal and upset conditions. Stresses during normal operation are< 30 ksi, matching the stress for the component in MRP-191.

2. MRP-191, Section 4.3.2, Heat Generation and Neutron Fluence, states "for the purposes of analysis, six distinct fluence regions were defined:"

Region 6: 5x1 022 n/cm 2 (75 dpa)  ::;; cpt Where, cpt (fluence) is for neutron energies with E > 1 MeV The Neutron Fluence Region provided in MRP-191, Table A-1, Results of Parameter Screening Input and Interviews with Analysts-Westinghouse Reactor Internals for Baffle and Former Assembly components is "Region 6". Region 6 fluence is consistent with the maximum neutron exposure received by the Reactor Internals Baffle Plates for 54 EFPY in Section 2.1.4.2.2 of the 2009 EPU submittal.

To address the NRC requested information, NextEra Energy Point Beach, LLC contracted with Westinghouse to determine which components are not bounded by the assumptions on stress, temperature, and fluence contained in MRP-191, MRP-232, and MRP-227-A. Westinghouse has completed the plant-specific evaluation to demonstrate that the MRP-227 recommended inspections will ensure functionality of these components until the next scheduled inspection.

Results of the plant-specific evaluations are provided in the following .

Applicability of FMECA and Functionality Analysis Assumptions The process used to verify that PBNP Unit 1 (WEP) and PBNP Unit 2 (WIS) are reasonably represented by the generic industry program assumptions with regard to neutron fluence, temperature, materials, and stress values used in the development of MRP-227-A is:

1. Identify typical Westinghouse pressurized water reactor (PWR) internals components (MRP-191, Table4-4).

2. Identify WEP/WIS PWR internals components.

3. Compare the typical Westinghouse PWR internals components to the WEP/WIS PWR internals components.

a. Confirm that no additional items were identified by this comparison (primarily supports Applicant/Licensee Action Item 2).

b. Confirm that the materials identified for WEP/WIS are consistent with those materials identified in MRP-191, Table 4-4.

c. Confirm that the WEP/WIS internals are the same as, or equivalent to, the typical Westinghouse PWR internals regarding design and fabrication.

4. Confirm that the WEP/WIS operating history is consistent with the assumptions in MRP-227-A regarding core loading patterns.

5. Confirm that WEP/WIS operates at base load .

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6. Confirm that the WEP/WIS RVI materials operated at temperatures within the original design basis parameters.

7. Determine stress values based on design basis documents.

8. Confirm that any changes to the WEP/WIS RVI components do not impact the application of the MRP-227-A generic aging management strategy.

PBNP Unit 1 (WEP) Compliance WEP reactor internals 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) (Reference 2) and the MRP-232 (Reference 3) functionality analyses based on the following:

1. WEP operating history is consistent with the assumptions in MRP-227-A with regard to neutron fluence.

a. The FMECA and functionality analyses for MRP-227 -A 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. In fuel cycle 8, WEP switched to use of a low-leakage core design. The core loading pattern was changed significantly prior to 30 years of operation. Therefore, WEP meets the fluence and fuel management assumptions in MRP-191 and requirements for MRP-227-A application.

b. Early in operating life, WEP followed load as needed depending on Wisconsin Electric Power Company system requirements. Load following operations diminished in the late 1970s as system requirements changed. Since this time, WEP has typically operated at a fixed power level. The resulting load follow cycles are well within the reactor vessel internals' design basis transient analysis. This existence of the load following history will not affect the component categorizations, nor will it affect recommended inspections since WEP has been operated within its design and licensing basis.

2. The WEP reactor vessel materials operate at temperatures between That and Tcald that have nominally been not less than 523°F for Tcald and not higher than 611 oF for That*

3. WEP internals components and materials are comparable to the typical Westinghouse PWR internals components, as summarized in MRP-191, Table 4-4.

a. No additional components are identified for W EP by this comparison (Reference 6).

b. Most of the materials identified for WEP are consistent with, or nearly equivalent to, those materials identified in MRP-191, Table 4-4 for Westinghouse-designed plants.

Material differences are listed in Table 1.

i. Where minor differences exist, there is no impact on the W EP RVI program.

ii. There are some WEP RVI components with material differences that require more rigor to determine any potential impacts. These components are: control rod guide tube assemblies and flow downcomers (guide cards/plates); control rod guide tube assemblies and flow downcomers (housing plates); upper instrumentation conduit and supports (brackets, clamps, terminal blocks, and conduit straps); and bottom-mounted instrumentation (BMI) column cruciforms. A FMECA expert panel (Reference 10) determined that the material differences had no effect on the recommended Materials Reliability Program (MRP) aging management inspection sampling strategy.

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c. WEP RVI component materials are the same as, or equivalent to, the typical Westinghouse PWR internals regarding design and fabrication.

d. A generic FMECA (Reference 8) was performed for use of cast austenitic stainless steel (CASS) materials for upper instrumentation conduit and supports - brackets, clamps, terminal blocks, and conduit straps. The FMECA concluded that the components could be classified as "No Additional Measures" based on a consideration of the likelihood of failure and the likelihood of damage. There is no change in MRP-227-A inspection requirements as a result of the inclusion of CF8 for these components (brackets, clamps, terminal blocks, and conduit straps).

4. WEP has made modifications to the reactor internals. These modifications were all performed with the involvement of Westinghouse, the RVI designer. MRP-227 states that the recommendations are applicable to all U.S. PWR operating plants as of May 2007 for the three designs considered. WEP has not made any modifications to reactor internals components since May 2007 other than replacement of split pins for WEP with an upgraded material in 2008. The split pin replacement was performed by Westinghouse. The modification will have no impact on the applicability of MRP-227, and is an example of the WEP proactive approach to managing aging reactor internals. Therefore, the WEP stress values are represented by the assumptions in MRP-191 , MRP-232, and MRP-227 -A, confirming the applicability of the generic FMECA.

5. A WEP extended power uprate (EPU) request has been approved . The EPU increases the fluence on various components. It does not significantly change the identified aging mechanisms or the ability of programs to adjust and manage aging effects (Reference 7).

Therefore, it is concluded that there is no impact, adverse or otherwise, from the WEP EPU program on the plant-specific implementation of the MRP-227 requirements.

Conclusion The assumptions regarding plant design and operating history made in the FMECA and functionality analyses for the Westinghouse design apply to WEP. There are no components at WEP not contained in the FMECA and functionality analysis . There are components with materials different than those assumed in the FMECA; however, evaluations have been completed to verify that these differences do not affect the current aging management strategy. WEP meets the requirements for application of MRP-227 -A as a strategy for managing age-related material degradation in reactor internals components. W EP will implement and apply the approved version of MRP-227 (MRP-227 -A) as a strategy for managing age-related material degradation in reactor internals components.

PWR Vessel Internal Components within the Scope of License Renewal - WEP Compliance This compliance assessment requires comparison of the RVI components that are within the scope of license renewal for WEP to those components contained in MRP-191, Table 4-4. A detailed tabulation of the WEP RVI components was completed, and compared favorably to typical Westinghouse PWR internals components in MRP-191.

Several components have different materials than those specified in the MRP-191 assessment; however, these have no effect on the recommended MRP aging. Therefore, no modifications to the program details in MRP-227-A need to be proposed.

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The upper internals assembly- control rod guide tube assemblies and flow downcomers -guide cards/plates were identified as potentially being stainless steel or CASS; conversely, MRP-191 identified these components as Type 304 SS. A FMECA expert panel [Reference 10) was conducted, aligned with the process specified in [Reference 2, Section 6], to address these components as CF8 rather than the 304 SS that was listed in MRP-191 . Based on consideration of the likelihood of failure, likelihood of damage, and FMECA group number, it was concluded with 100% consensus that there is no additional action necessary for WEP to conform to MRP-227 -A guidelines.

The upper internals assembly- control rod guide tube assemblies and flow downcomers- housing plates were identified as potentially being stainless steel or CASS; conversely, MRP-191 identified these components as Type 304 SS. A FMECA expert panel was conducted to address these components as CF8 rather than the 304 SS that was listed in MRP-191. Based on consideration of the likelihood of failure, likelihood of damage, and FMECA group number, it was concluded with 100% consensus that there is no additional action necessary for WEP to conform to MRP-227-A guidelines.

The upper internals assembly - upper instrumentation conduit and supports - brackets, clamps, terminal blocks, and conduit straps) were identified as potentially being stainless steel or CASS; conversely, MRP-191 identified these components as Type 304 SS. A generic FMECA (Reference 8) was performed for use of CASS materials for upper instrumentation conduit and supports (brackets, clamps, terminal blocks, and conduit straps). The FMECA concluded that the components could be classified as "No Additional Measures" based on consideration of the likelihood of failure and the likelihood of damage. There is no change in MRP-227-A inspection requirements as a result of the inclusion of CF8 for these components (brackets, clamps, terminal blocks, and conduit straps).

The lower internals assembly- BMI column assemblies- BMI column cruciforms were identified as being Type 304 SS in addition to the MRP-191 identification of these components as CF8. A FMECA expert panel will be conducted, aligned with the process specified in MRP-191, Section 6, to determine if there are any changes to the current MRP inspection of the BMI column cruciforms.

A FMECA expert panel was conducted to address these components as 304 SS rather than the CF8 that was listed in MRP-191. Based on consideration of the likelihood of failure, likelihood of damage, and FMECA group number, it was concluded with 100% consensus that there is no additional action necessary for WEP to conform to MRP-227-A guidelines.

This panel and the conclusions drawn from it will support the requirement that the Aging Management Program Plan shall provide assurance that the effects of aging on the WEP RVI components within the scope of license renewal, but not included in the generic Westinghouse-designed RVI components from Table 4-4 of MRP-191, will be managed for the period of extended operation.

Conclusion All components required to be included in the WEP program are consistent with those contained in MRP-191. Several components have materials different than those specified in MRP-191; however, evaluations have been completed to show that these differences have no effect on the MRP aging management strategy. WEP meets the requirements for application of MRP-227-A as a strategy for managing age-related material degradation in reactor internals components.

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Point Beach Unit 2 (WIS) Compliance WIS reactor internals 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) and the MRP-232 functionality analyses based on the following:

1. WIS operating history is consistent with the assumptions in MRP-227-A with regard to neutron fluence.

a. The FMECA and functionality analyses for MRP-227 -A 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. In fuel cycle 6, WIS switched to use of a low-leakage core design. The core loading pattern was changed significantly prior to 30 years of operation . Therefore, WIS meets the fluence and fuel management assumptions in MRP-191 and requirements for MRP-227-A application .

b. Early in operating life, WIS followed load as needed depending on Wisconsin Electric Power Company system requirements . Load following operations diminished in the late 1970s as system requirements changed. Since this tinie, WIS has typically operated at a fixed power level. The resulting load follow cycles are well within the reactor vessel internal's design basis transient analysis . This existence of the load following history will not affect the component categorizations, nor will it affect recommended inspections since WIS has been operated within its design and licensing basis.

2. The WIS reactor vessel materials operate at temperatures between That and Tcald that have nominally been not less than 523°F for Tcald and not higher than 611 oF for That*

3. WIS internals components and materials are comparable to the typical Westinghouse PWR internals components, as summarized in MRP-191, Table 4-4.

a. No additional components are identified for WISby this comparison .

b. Most of the materials identified for WIS are consistent with, or nearly equivalent to, those materials identified in MRP-191, Table 4-4 for Westinghouse-designed plants . Material differences are listed in Table 2.

i. Where minor differences exist, there is no impact on the WIS RVI program.

ii. There are some WIS RVI components with material differences that require more rigor to determine any potential impacts. These components are: control rod guide tube assemblies and flow downcomers (guide cards/plates); control rod guide tube assemblies and flow downcomers (housing plates); upper instrumentation conduit and supports (brackets, clamps , terminal blocks, and conduit straps); bottom-mounted instrumentation (BMI) column cruciforms; secondary core support (SCS) assembly (SCS guidepost); and SCS assembly (SCS housing). A FMECA expert panel determined that the material differences had no effect on the recommended MRP aging management inspection sampling strategy.

c. WIS RVI component materials are the same as, or equivalent to, the typical Westinghouse PWR internals regarding design and fabrication.

d. A generic FMECA was performed for use of cast austenitic stainless steel (CASS) materials for upper instrumentation conduit and supports (brackets, clamps , terminal blocks, and Page 7 of 13

conduit straps). The FMECA concluded that the components could be classified as "No Additional Measures" based on a consideration of the likelihood of failure and the likelihood of damage. There is no change in MRP-227-A inspection requirements as a result of the inclusion of CF8 for these components (brackets, clamps, terminal blocks , and conduit straps).

4. WIS has made modifications to the reactor internals. These modifications were all performed with the involvement of Westinghouse, the RVI designer. MRP-227 states that the recommendations are applicable to all U.S. PWR operating plants as of May 2007 for the three designs considered. WIS has not made any modifications to reactor internals components since May 2007 other than replacement of split pins for WEP with an upgraded material in 2005.

The split pin replacement was performed by Westinghouse. The modification will have no impact on the applicability of MRP-227, and is an example of the WIS proactive approach to managing aging reactor internal. Therefore, the WIS stress values are represented by the assumptions in MRP-191, MRP-232, and MRP-227-A, confirming the applicability of the generic FMECA.

5. A WIS extended power uprate (EPU) request has been approved. The EPU increases the fluence on various components. It does not significantly change the identified aging mechanisms or the ability of programs to adjust and manage aging effects. Therefore, it is concluded that there is no impact, adverse or otherwise, from the WIS EPU program on the plant-specific implementation of the MRP-227 requirements.

Conclusion The assumptions regarding plant design and operating history made in the FMECA and functionality analyses for the Westinghouse design apply to WIS. There are no components at WIS not contained in the FMECA and functionality analysis. There are components with materials different than those assumed in the FMECA; however, evaluations have been completed to verify that these differences do not affect the current aging management strategy. WIS meets the requirements for application of MRP-227 -A as a strategy for managing age-related material degradation in reactor internals components . WIS will implement and apply the approved version of MRP-227 (MRP-227-A) as a strategy for managing age-related material degradation in reactor internals components.

PWR Vessel Internal Components within the Scope of License Renewal - WIS Compliance This compliance assessment requires comparison of the RVI components that are within the scope of license renewal for WIS to those components contained in MRP-191, Table 4-4. A detailed tabulation of the WIS RVI components was completed , and compared favorably to the typical Westinghouse PWR internals components in MRP-191 .

Several components have different materials than those specified in the MRP-191 assessment; however, these have no effect on the recommended Materials Reliability Program (MRP) aging .

Therefore, no modifications to the program details in MRP-227-A need to be proposed.

The upper internals assembly- control rod guide tube assemblies and flow downcomers- guide cards/plates were identified as potentially being stainless steel or CASS; conversely, MRP-191 identified these components as Type 304 SS. A FMECA expert panel was conducted to address these ~omponents as CF8 rather than the 304 SS that was listed in MRP-191. Based on consideration of the likelihood of failure, likelihood of damage, and FMECA group number, it was Page 8 of 13

concluded with 100% consensus that there is no additional action necessary for WIS to conform to MRP-227-A guidelines.

The upper internals assembly- control rod guide tube assemblies and flow downcomers - housing plates were identified as potentially being stainless steel or CASS; conversely, MRP-191 identified these components as Type 304 SS. A FMECA expert panel was conducted to address these components as CF8 rather than the 304 SS that was listed in MRP-191. Based on consideration of the likelihood of failure, likelihood of damage, and FMECA group number, it was concluded with 100% consensus that there is no additional action necessary for WIS to conform to MRP-227-A guidelines.

The upper internals assembly- upper instrumentation conduit and supports - brackets, clamps, terminal blocks, and conduit straps) were identified as potentially being stainless steel or CASS; conversely, MRP-191 identified these components as Type 304 SS. A generic FMECA was performed for use of CASS materials for upper instrumentation conduit and supports (brackets, clamps, terminal blocks, and conduit straps). The FMECA concluded that the components could be classified as "No Additional Measures" based on a consideration of the likelihood of failure and the likelihood of damage. There is no change in MRP-227 -A inspection requirements as a result of the inclusion of CF8 for these components (brackets, clamps, terminal blocks, and conduit straps).

The lower internals assembly- BMI column assemblies- BMI column cruciforms were identified as being Type 304 SS in addition to the MRP-191 identification of these components as CF8. A FMECA expert panel was conducted to address these components as 304 SS rather than the CF8 that was listed in MRP-191. Based on consideration of the likelihood of failure, likelihood of damage, and FMECA group number, it was concluded with 100% consensus that there is no additional action necessary for WIS to conform to MRP-227-A guidelines.

The lower internals assembly- SCS assembly- SCS guide posts were confirmed by the utility as being CF8; conversely, MRP-191 identified this component as Type 304 SS. A FMECA expert panel was conducted to address these components as CF8 rather than the 304 SS that was listed in MRP-191. Based on consideration of the likelihood of failure, likelihood of damage, and FMECA group number, it was concluded with 100% consensus that there is no additional action necessary for WIS to conform to MRP-227-A guidelines.

The lower internals assembly- SCS assembly- SCS housing was confirmed by the utility as being CF8; conversely, MRP-191 identified this component as Type 304 SS. A FMECA expert panel was conducted to address these components as CF8 rather than the 304 SS that was listed in MRP-191. Based on consideration of the likelihood of failure, likelihood of damage, and FMECA group number, it was concluded with 100% consensus that there is no additional action necessary for WIS to conform to MRP-227-A guidelines.

This panel and the conclusions drawn from it will support the requirement that the Aging Management Program Plan shall provide assurance that the effects of aging on the WIS RVI components within the scope of license renewal, but not included in the generic Westinghouse-designed RVI components from Table 4-4 of MRP-191, will be managed for the period of extended operation.

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Conclusion All components required to be included in the WIS program are consistent with those contained in MRP-191. Several components have materials different than those specified in MRP-191; however, evaluations have been completed to show that these differences have no effect on the MRP aging management strategy. WIS meets the requirements for application of MRP-227 -A as a strategy for managing age-related material degradation in reactor internals components .

RAI-8

In the December 19, 2011, submittal, there is no discussion of Action Item 7 from MRP-227-A. In Section 3. 3. 7 of the SE for MRP-227, the NRC staff stated that the licensee shall develop a plant-specific analysis to demonstrate that components manufactured from CASS materials will maintain their functions during the period of extended operation. This requirement applies to all susceptible components for which the licensee has determined aging management is required, which includes components designated as expansion and existing as well as the primary category.

Requested Information The NRC staff requests that the licensee provide a list of all reactor vessel internal components manufactured from CASS materials for Point Beach, Units 1 and 2, along with the plant-specific analysis required by Action Item 7. Provide plant-specific aging management requirements for any components that are not already covered by the "Primary'~ "Expansion'~ or "Existing Programs '~

categories under MRP-227-A.

NextEra Response Table 1 and Table 2 list reactor vessel internal components manufactured from CASS materials for Point Beach, Units 1 and 2, respectively.

As stated in the response to RAI 1a, for some of the CASS materials, a FMECA expert panel was conducted to address these CASS components. Based on consideration of the likelihood of failure, likelihood of damage, and FMECA group number, it was concluded with 100% consensus that there is no additional action necessary for WEP/WIS to conform to MRP-227-A guidelines.

A generic FMECA was performed for use of cast austenitic stainless steel (CASS) materials for upper instrumentation conduit and supports- brackets, clamps, terminal blocks, and conduit straps.

The FMECA concluded that the components could be classified as "No Additional Measures" based on a consideration of the likelihood of failure and the likelihood of damage. There is no change in MRP-227 -A inspection requirements as a result of the inclusion of CF8 for these components (brackets, clamps, terminal blocks, and conduit straps).

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References

1. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A). EPRI, Palo Alto, CA: 2011. 1022863.

2. 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.

3. Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals (MRP-232). EPRI, Palo Alto , CA: 2008. 1016593.

4. Point Beach Nuclear Plant Procedures Manual, NP 7.7.30, Rev. 3, "Reactor Vessel Internals Program," October 3, 2012.

5. NextEra Energy Point Beach Letter, NRC 2012-0052, Reactor Vessel Internals Inspection Plan Response to Request for Additional Information," August 16, 2012.

6. Point Beach Nuclear Generating Station Application for Renewed Operating License, February 26, 2004.

7. Point Beach Nuclear Plant License Renewal Project, Reactor Vessel Internals Program Basis Document for License Renewal, Rev. 9, "LR-AMP-015-RVINT," May 2, 2013

8. Westinghouse Letter, LTR-RIAM-12-105, Rev. 0, "MRP-191 Failure Modes, Effects, and Criticality Analysis (FMECA) Impact Assessment as a Result of Alternate Materials for Reactor Vessel Internals Upper Instrumentation Conduit and Supports Brackets, Clamps, Terminal Blocks, and Conduit Straps.," September 13, 2012. (Westinghouse Proprietary Class 2)

9. NextEra Energy Point Beach Letter, NRC 2013-0029, "Reactor Vessel Internals Inspection Plan Response to Request for Additional Information," March 15, 2013.

10. Westinghouse Letter, LTR-RIAM-14-6, Rev. 0, "Summary of Point Beach Units 1 and 2 Expert Elicitation Panel Meeting Minutes for Reactor Internals Components and Materials," February 21, 2014. (Westinghouse Proprietary Class 2)

11. NRC electronic mail to NextEra Energy Point Beach, LLC dated January 31, 2013, Point Beach Nuclear Plant, Units 1 and 2 - Draft Request for Additional Information (Second Round) re:

Reactor Vessel Internals Inspection Plan Review (TAC ME8235 and ME8236)

12. NextEra Energy Point Beach, LLC letter to NRC, dated December 19, 2011 , License Renewal Commitment, Reactor Vessel Internals Prog ram Submittal (ML113540301)

13. NextEra Energy Point Beach, LLC letter to NRC, dated April 7, 2009, License Amendment 261, Extended Power Uprate (ML0191250566)

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Material Comparison Tables Table 1: Point Beach Unit 1 (WEP) RVI Material Differences from MRP-191 Assembly Subassembly Component Material Upper Control Rod Guide Guide plates/cards* MRP-191 304 ss Internals Tube Assemblies WEP CF8 Assembly and Flow Downcomers Housing plates* MRP-191 304 ss WEP CF8 Mixing Devices Mixing devices MRP-191 CF8 WEP 304 ss Upper Core Plate Fuel alignment pins MRP-191 316 ss and Fuel Alignment Pins WEP 304 ss Upper Bolting MRP-191 316 ss Instrumentation WEP 304 ss Conduit and Supports Brackets, clamps, MRP-191 304 ss terminal blocks, and conduit straps* WEP CF8 Locking caps MRP-191 304 ss WEP 304L SS Lower Bottom Mounted BMI column MRP-191 CF8 Internals Instrumentation cruciforms Assembly WEP 304 SS and (BMI) Column CF8 Assemblies Lower Core Plate Fuel alignment pins MRP-191 316 ss and Fuel Alignment WEP Pins 304 ss Lower Support Lower support MRP-191 304 ss Column column bolts WEP 316 ss Assemblies Neutron Panels/ Thermal shield MRP-191 316 ss Thermal Shield dowels WEP 304 ss Radial Support Radial support key MRP-191 304 ss Keys bolts WEP 316 ss

• - Some component drawings list ASTM A351, Type 304, Gr. CF8 as alternate material.

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Table 2: Point Beach Unit 2 (WIS) RVI Material Differences from MRP-191 Assembly Subassembly Component Material Upper Control Rod Guide Tube Guide plates/cards* MRP-191 304 ss Internals Assemblies and Flow WIS CF8 Assembly Down comers Housing plates* MRP-191 304 ss WIS CF8 Mixing Devices Mixing devices MRP-191 CF8 WIS 304 ss Upper Core Plate and Fuel Fuel alignment pins MRP-191 316 ss Alignment Pins WIS 304 ss Upper Instrumentation Bolting MRP-191 316 ss Conduit and Supports WIS 304 ss Brackets, clamps, MRP-191 304 ss terminal blocks, and conduit straps* WIS CF8 Locking caps MRP-191 304 ss WIS 304L SS Lower Bottom Mounted BMI column MRP-191 CF8 Internals Instrumentation (BMI) cruciforms WIS 304 SS and Assembly Column Assemblies CF8 Lower Core Plate and Fuel Fuel alignment pins MRP-191 316 ss Alignment Pins WIS 304SS Lower Support Column Lower support MRP-191 304 ss Assemblies column bolts WIS 316 ss Neutron Panels/ Therm al Therm al shield MRP-191 316 ss Shield dowels WIS 304 ss Radial Support Keys Radial support key MRP-191 304SS bolts WIS 316 ss Secondary Core Support MRP-1 91 304 ss (SCS) Assembly SCS guide post WIS CF8 MRP-191 304 ss SCS housing WIS CF8

• - Some component drawings list ASTM A351 , Type 304, Gr. CF8 as alternate material.

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