ML13261A145

NextEra Energy Seabrook License Renewal Application Alkali-Silica Reaction (ASR) Monitoring Program

ML13261A145
Person / Time
Site:	Seabrook
Issue date:	09/13/2013
From:	Walsh K NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML13261A145 (35)

Text

NEXTera ENERGY September 13, 2013 SBK-L- 13162 Docket No. 50-443 U.S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852 Seabrook Station NextEra Energy Seabrook License Renewal Application Alkali-Silica Reaction (ASR) Monitoring Program

References:

1. NextEra Energy Seabrook, LLC letter SBK-L-10077, "Seabrook Station Application for Renewed Operating License," May 25, 2010. (Accession Number ML101590099)

2. NextEra Energy Seabrook, LLC letter SBK-L- 12101, "Seabrook Station NextEra Energy Seabrook License Renewal Application Structures Monitoring Program Supplement-Alkali-Silica Reaction (ASR) Monitoring," May 16, 2012 (Accession Number ML12142A323).

3. NextEra Energy Seabrook, LLC letter SBK-L-12217, "Seabrook Station Response to Request for

- Additional Information, NextEra Energy Seabrook License Renewal Application - Request for Additional Information," November 2, 2012 (Accession Number ML12312A017).

4. NextEra Energy Seabrook, LLC letter SBK-L-12247, "Clarification for Response to Follow up RAI B.2.1.31-1 Item (b)(2) provided in SBK-L-12217," November 20, 2012 (Accession Number NML12333A237).

5. NRC Letter, "Summary of Meeting Held on February 21, 2013, Between the US Nuclear Regulatory Commission and NextEra Energy Seabrook, LLC., Regarding the Seabrook Nuclear Power Station License Renewal Application (TAC NO. ME4028)," March 21, 2013 (Accession Number ML13066A488).

6. NextEra Energy Seabrook, LLC letter SBK-L-13055, "Response to Confirmatory Action Letter,"

May 15, 2013.

7. NextEra Energy Seabrook, LLC letter SBK-L-13080, "Response to Confirmatory Action Letter,"

May 1, 2013.

In Reference 1, NextEra Energy Seabrook, LLC (NextEra) submitted an application for a renewed facility operating license for Seabrook Station Unit 1 in accordance with the Code of Federal Regulations, Title 10, Parts 50, 51, and 54.

NextEra Energy Seabrook, LLC.

626 Lafayette Rd, Seabrook, NH 03874 , N.

United States Nuclear Regulatory Commission SBK-L-13162/Page 2 In Reference 2, NextEra provided changes to the License Renewal Application (LRA) associated with managing the effects of Alkali-Silica Reaction. A plant specific Alkali-Silica Reaction (ASR) Monitoring Program, B.2.1.31A which augments the existing Structures Monitoring Program, B.2.1.31 was contained in this correspondence.

In References 3 and 4, NextEra provided supplemental information related to staff RAIs regarding the Alkali-Silica Reaction Monitoring Program.

On February 21, 2013, NRC Staff met with NextEra to discuss additional information that was needed in order to complete their review of the Alkali-Silica Reaction Monitoring Program.

Notes of that meeting are provided in Reference 5.

In References 6 and 7, NextEra provided details of the Large Scale Test and Anchor Test Programs for ASR affected concrete.

Additional information based on the February 21, 2013 meeting with NRC Staff has been incorporated in this supplement to the License Renewal Application (LRA). contains changes to LRA Appendix A - Updated UFSAR Supplement, and Appendix B - Aging Management Programs associated with the Alkali-Silica Reaction Monitoring Program.

The changes are explained, and where appropriate to facilitate understanding, portions of the LRA are repeated with the change highlighted by strikethroughs for deleted text and bolded italics for inserted text. In some instances the entire text of a section has been replaced or added.

In these cases a note is included in the introduction indicating the replacement of the entire text of the section.

Commitment number 71 has been revised. There are no other new or revised regulatory commitments contained in this letter. Enclosure 2 provides a revised LRA Appendix A - Final Safety Report Supplement Table A.3, License Renewal Commitment List, updated to reflect the license renewal commitment changes made in NextEra Energy Seabrook correspondence to date.

If there are any questions or additional information is needed, please contact Mr. Richard R.

Cliche, License Renewal Project Manager, at (603) 773-7003.

If you have any questions regarding this correspondence, please contact Mr. Michael Ossing Licensing Manager, at (603) 773-7512.

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

Executed on September 13, 2013 Sincerely, Kevin T. Walsh Site Vice President NextEra Energy Seabrook, LLC

United States Nuclear Regulatory Commission SBK-L-13162/Page 3

Enclosures:

Enclosure 1-Changes to the Seabrook Station License Renewal Application Associated with Appendix A - Updated UFSAR Supplement, and Appendix B - Aging Management Programs Enclosure 2- LRA Appendix A - Final Safety Report Supplement Table A.3, License Renewal Commitment List cc:

W.M. Dean, NRC Region I Administrator J. G. Lamb, NRC Project Manager, Project Directorate 1-2 P. Cataldo NRC Resident Inspector R. A. Plasse Jr., NRC Project Manager, License Renewal L. M. James, NRC Project Manager, License Renewal Director Homeland Security and Emergency Management New Hampshire Department of Safety Division of Homeland Security and Emergency Management Bureau of Emergency Management 33 Hazen Drive Concord, NH 03305 John Giarrusso, Jr., Nuclear Preparedness Manager The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399

Enclosure 1 to SBK-L-13162 Changes to the Seabrook Station License Renewal Application Associated with Appendix A - Updated UFSAR Supplement and Appendix B - Aging Management Programs

United States Nuclear Regulatory Commission Page 2 of 18 SBK-L- 13162 / Enclosure 1 Introduction contains an update to the NextEra Energy Seabrook License Renewal Application (LRA), Appendix A and Appendix B. Included in this update are changes to information provided in References 2 and 3 relative to the plant specific Alkali-Silica Reaction (ASR) Monitoring Program, B.2.1.3 IA.

CHANGES TO LRA APPENDIX A The following changes have been made to Appendix A of the Seabrook License Renewal Application (LRA). For clarity, entire sentences or paragraphs from the LRA are provided with deleted text highlighted by strikethroughs and inserted text highlighted by bolded italics:

a) Revise section A.2.1.31A (Reference 2), as follows:

A.2.1.31A ALKALI-SILICA REACTION (ASR) MONITORING The Alkali-Silica Reaction (ASR) Monitoring Program manages cracking due to expansion and reaction with aggregates of concrete structures within the scope of license renewal. The potential impact of ASR on the structural strength and anchorage capacity of concrete is a consequence of strains resulting from the expansive gel. These strains produce the associated cracking. The program is consistent with the ten elements of an acceptable aging management program as described in NUREG- 1800 Appendix A. 1, Section A. 1.2.3 and Table A. I-1.

The Structural Monitoring Program performs visual inspections of the concrete structures at Seabrook for indications of the presence of alkali-silica reaction (ASR). ASR is detected by visual observation of cracking on the surface of the concrete. The cracking is typically accompanied by the presence of moisture and efflorescence. Concrete affected by expansive ASR is typically characterized by a network or "pattern" of cracks. ASR involves the formation of an alkali-silica gel which expands when exposed to water. Microcracking due to ASR is generated through forces applied by the expanding aggregate particles and/or swelling of the alkali-silica gel within and around the boundaries of reacting aggregate particles.

The ASR gel may exude from the crack forming white secondary deposits at the concrete surface. The gel also often causes a dark discoloration of the cement paste surrounding the crack at the concrete surface. If pattern or map cracking typical of concrete affected by ASR is identified, an evaluation will be performed to determine further actions. Monitoring of crack growth is used to assess the long term implications of ASR and specify monitoring intervals.

To manage the aging effects of ASR cracking due to expa..sion and r.eaeti.n with aggregates in concrete structures, the existing Structures Monitoring Program, B....31., has been augmented by this plant specific Alkali-Silica Reaction (ASR)

Monitoring Program, 1.2.!.3!A. The ASR Monitoring Program is structured according to the guidelines in ACI 349.3R, "Stuc.tural Condition Asess.ment of Buildings." "Evaluation of Existing Nuclear Safety-Related Concrete Structures."

United States Nuclear Regulatory Commission Page 3 of 18 SBK-L- 13162 / Enclosure 1 ASR is detected by visual observation of cracking on the surface of the concrete.

The cracking is typically accompanied by the presence of moisture and efflorescence. Monitoring of. crack growth is used to assess the long term implications of ASR and specify monitoring intervals.

A Combined Cracking Index (CCI) and Individual Crack Width criteria are established as thresholds at which structural evaluation is necessary. The Cracking Index is the summation of the crack widths on the horizontal or vertical sides of a 0.5m (20-inch) by 0.5m (20-inch) square on the ASR-affected concrete surface.

The horizontaland vertical Cracking Indices are averaged to obtain a Combined CrackingIndex (CCI)for each area of interest. The CC! represents the expan.ion along the entire perimeter- of the 20 ineh by 20 inch square. A CCI of less than the 1.0 mm/m and Individual Crack Width of less than 1.0 mm can be deemed acceptable with deficiencies. Deficiencies determined to be acceptable with further review are trended for evidence of further degradation. A CCI of 1.0 mm/m or greater, or an Individual Crack Width of 1.0 mm or greater requires structural evaluation.

The Alkali-Silica Reaction (ASR) Monitoring Program will continue to monitor surface cracking (CCI) of ASR impacted areas on the frequencies establishedin the program. Follow-up inspection of at least 20 locations that represent the highest CCI values recorded during the baseline inspections, these locatins will be performed at six month intervals.

Large scale destructive testing of concrete beams with acceleratedASR confirms parameters being monitored are appropriateto manage the effects of ASR and that acceptance criteria used provides sufficient margin. Anchor bolt testing quantifies the impact of ASR on anchor capacity as a function of the severity of ASR degradation.

b) Revise Commitment #71, Table A.3, "License Renewal Commitment List",

(Reference 2), as follows:

A.3 LICENSE RENEWAL COMMITMENT LIST No. PROGRAM or TOPIC COMMITMENT SCHEDULE Implement the Alkali-Silica Reaction (ASR) Monitoring Program. Testing Alkali-Silica Reaction will be performed to confirm that Prior to entering the 71 (ASR) Monitoring parametersbeing monitoredand period of extended Program acceptance criteria used are operation.

appropriateto manage the effects of ASR.

United States Nuclear Regulatory Commission Page 4 of 18 SBK-L-13 162 / Enclosure 1 CHANGES TO LRA APPENDIX B The following changes have been made to Appendix B of the Seabrook License Renewal Application (LRA). For clarity, revised sections (Program Description and Elements 3 through 6 and 10) are included in their entirety. Changes are shown with deleted text highlighted by strikethroughs and inserted text highlighted by bolded italics:

a) Revise Section B.2.1.31A, Alkali-Silica Reaction (ASR) Monitoring Program (Reference 2), as follows:

B.2.1.31A ALKALI-SILICA REACTION MONITORING PROGRAM Pro2ram Description NextEra Energy Seabrook Operating Experience (OE) indicates that Alkali-Silica Reaction (ASR) is present in concrete structures and will require monitoring through the Period of Extended Operation.

Alkali Siliea Reaction (ASR) is a reaction that occurs ever- time in concret-e bet-wen alkalinte ceetpaste and r-eactive, non cr-ystalline silica in aggregates.

An expansive gel is formed within aggregates resulting. inEmirocacks in.

aggr.egates and in the adjacent cement paste. Alkali-Silica Reaction (ASR) is an aging mechanism that may occur in concrete under certain circumstances.It is a reaction between the alkaline cement and reactiveforms of silicate material (if present) in the aggregate. The reaction, which requires moisture to proceed, produces an expansive gel materiaL This expansion results in strains in the material that can produce micro-cracking in the aggregate and in the cement paste. The potential impact of ASR on the structural strength and anchorage capacity of concrete is a consequence of strains resultingfrom the expansive geL These strains produce the associated cracking. Because the ASR mechanism requires the presence of moisture in the concrete, ASR has been predominantly detected in groundwater impacted portions of below grade structures, with limited impact to exterior surfaces of above grade structures.

The ASR developed at Seabrook because the concrete mix designs utilized an aggregate that was susceptible to Alkali-Silica Reaction, which was not known at the time. Although the testing was conducted in accordance with the ASTM C289 standard, this standard was subsequently identified as limited in its ability to predict long term ASR.

The plant specific Alkali-Silica Reaction (ASR) Monitoring Program manages cracking due to expansion and reaction with aggregates of concrete structures within the scope of license renewal The potential impact of ASR on the structural strength and anchorage capacity of concrete is a consequence of strains resultingfrom the expansive gel These strains produce the associated cracking. To manage the these aging effects of cracking due to expansion and r.eactin with aggregates in eonc.r.ete stfu*tur-es, the existing Structures Monitoring Program, B.2.1.31, has been augmented by this plant specific Alkali-Silica Reaction (ASR) Monitoring Program, B.2.1.31 A. The ASR Monitoring Program will be structured according to the guidelines in ACI 349.3R, "SructurIa

United States Nuclear Regulatory Commission Page 5 of 18 SBK-L- 13162 / Enclosure 1 Cndi*ien A4s-sessment of Buidings." "Evaluation of Existing Nuclear Safety-Related Concrete Structures."

Evaluations of a structure's condition are completed according to the guidelines set forth in the Structural Monitoring Program (the Seabrook Station Maintenance Rule program that implements the Structures Monitoring Program). The acceptance guidelines in the Structural Monitoring Program are a three-tier hierarchy similar to that described in ACI 349.3R-96, which provides quantitative degradation limits. Under this system, structures are evaluated for ASR as detailed in the following Table. ASR affected areas classified to be Unacceptable (requires further evaluation) or Acceptable with Deficiencies are monitored in accordance with the Alkali-Silica Reaction (ASR) Monitoring Program.

Unacceptable (requires further Structural Evaluation 1.0 mm/m or greater 1.0 mm or greater evaluation)

Quantitative Monitoring 0.5 mm/m or greater 0.2 mm or greater Acceptable with and Trending Deficiencies sQualitative Monitoring Any area with indications of pattern cracking or water ingress Routine inspection as Area has no indications of pattern cracking or Acceptable prescribed by Structures water ingress - No visual presence of ASR Monitoring Program Industry Expert Engagement Alkali-Silica Reaction is a newly identified aging mechanism to Seabrook and little or no experience exists in management of its aging effects in the US nuclear industry. NextEra has engaged several subject matter experts with knowledge of ASR and its impact on concrete structures. Based on this expertise, NextEra has concluded that the mechanical properties of the in-situ structures, that are highly reinforced with rebar steel, are expected to be higher than the results from core bore testing (4" concrete bore with no rebar).

Principle references used in development of this aging management program consist of the following input of subject matter experts and both nuclear and non-nuclear sources:

- "Report on the Diagnosis, Prognosis, and Mitigation of Alkali-Silica Reaction in Transportation Structures," U.S. Dept. of Transportation, Federal Highway Administration, January 2010, Report Number FHWA-HIF-09-004.

- "Structural Effects of Alkali-Silica Reaction: Technical Guidance on the Appraisal of Existing Structures," Institution of Structural Engineers, July 1992.

United States Nuclear Regulatory Commission Page 6 of 18 SBK-L- 13162 / Enclosure I

- ORNL/NRC/LTR-95/14, "In-Service Inspection Guidelines for Concrete Structures in Nuclear Power Plants," December 1995."

- "Seabrook Station: Impact of Alkali-Silica Reaction on Concrete Structures and Attachments", MPR-3727, Revision 0, April 2012.

Monitoring Monitoring of ASR associated cracks is an effective method for determining ASR progression. Monitoring of ASR associated cracks at Seabrook is implemented with two measurements. One is Cracking Index (CI) and the other is Individual Crack Width.

The CI is a quantitative assessment of cracking present in the cover concrete of affected structures. A CI measurement is taken on accessible surfaces exhibiting ASR pattern cracking. The process for determining the Cracking Index (CI) is described in the Federal Highway Administration (FHWA) document FHWA-HIF-09-004, "Report on the Diagnosis, Prognosis, and Mitigation of Alkali-Silica Reaction (ASR) in TransportationStructures." The FHWA Cracking Index is the summation of the crack widths on the horizontal or vertical sides of a 0.5m (20-inch) by 0.5m (20-inch) square on the ASR-affected concrete surface. Since each side of the square is 0.5 m, the Cracking Index in a given direction is reported in units of mm/m.

The horizontal and vertical Cracking Indices are averaged to obtain a Combined Cracking Index (CCI) for each area of interest. The CCI represents the expan.ion along the entire p.r.im.eter the

. 20

.f in.h by 20 in. h square. Criteria used in assessment of expansion is expressed in terms of CCI and based on recommendations provided in MPR-3727, Revision 0, Seabrook Station: Impact of Alkali-Silica Reaction on Concrete Structures and Attachments. The CCI and individual crack width are compared to ASR screening criteria and categorized for Qualitative or Quantitative Monitoring and Trending and Structural Evaluation.

The progression of ASR degradation of the concrete is an important consideration for assessing the long term implications of ASR and specifying monitoring intervals. The most reliable means for establishing the progression of ASR degradation is to monitor expansion of the in situ concrete in-sif.

Individual Crack Width measurement is also an effective method for assessment of ASR affected areas. Screening criteria used in the assessment of Individual Crack Width, expressed in terms of mm, are based on recommendations provided in MPR-3727, Revision 0, Seabrook Station: Impact of Alkali-Silica Reaction on Concrete Structuresand Attachments.

NextEra has performed a baseline inspection and ASR associated cracks have been evaluated and categorized. NextEra has assessed 131 accessible areas to date in this manner. The areas affected by ASR have been identified and assessed for apparent degradation from ASR, including estimation of in situ expansion. The results are presented in MPR-3727, Revision 0, Seabrook Station: Impact of Alkali-Silica Reaction on Concrete Structures and Attachments. Monitoring of Cracking Index and Individual Crack Width of at least 20 areas identified in the

United States Nuclear Regulatory Commission Page 7 of 18 SBK-L- 13162 / Enclosure 1 baseline inspection as having the largest CCI will be performed at six month intervals.. Trend data from these follow-up inspections will be used in determining the progression of ASR degradation and a basis for any change to the frequency of the inspection of ASR affected areas. Documentation and trend data will be maintained in accordance with the Structures Monitoring Program and established guidelines of ACI 349.3R, "Structural C.nditien A4ssesstnnt of Buildn.gs."

"Evaluationof Existing NuclearSafety-Related Concrete Structures."

Basisfor use of the crack index methodology is asfollows:

ASR produces a gel that expands as it absorbs moisture. This expansion exerts a tensile stress on the surrounding concrete which strains the concrete and eventually results in cracking.

The engineering strain in a structural member at the time of crack initiation (ecr is equivalent to the tensile strength of the concrete divided by the elastic modulus (Ecr = at / E). The Cracking Index quantifies the extent of the surface cracking. The total strain in the concrete can be approximatedas the sum of the strain at crack initiation plus the cracking index (. = .6, + CI). Figure A-]

depicts a concrete specimen with rebar being put in tension resulting in cracking.

Concrete has little strain capacity;therefore, in ASR-affected concrete, the crack widths comprise most of the expansion ('AL). As a result, even though the Cracking Index does not accountfor strain in the un-cracked concrete between cracks (Ebd, the Cracking Index provides a reasonable approximation of the total strain appliedto the concrete after crack initiation.

Rebar in Tension -

AL = L 2 - L1 L2 LI Cracks FigureA-i. Concrete Specimen put in Tension

United States Nuclear Regulatory Commission Page 8 of 18 SBK-L-13162/ Enclosure 1 For surfaces where horizontal and vertical cracking indices are similar (e.g.,

where there is equivalent reinforcement in both directions), a Combined Cracking Index (CCI) that averages the horizontal and vertical Cracking Indices can consolidatethe expansion assessment to a singleparameter.

Large Scale Testing While the monitoringprogram action levels are currently based on the generic information available in the literature, a full scale testing program has been undertaken to refine the impactsfor structuressimilar to those impacted by ASR at Seabrook Station. The purpose of the testing is to quantify the impact of ASR on structural performance for varied degrees of ASR-affected reinforced concrete that can be correlated to the ASR-affected concrete structures at Seabrook Station. In particular, test programs will focus on shear and reinforcementanchorage.Specific objectives of the testing are asfollows:

" Determine the extent to which the shearperformance, development length of reinforcement, andflexural stiffness of reinforced concrete beams are affected as a function of ASR-related expansion.

" Determine the effectiveness of retrofit techniques in the enhancement of shearperformance and on the development length in reinforced concrete beams.

Large scale destructive testing of concrete beams with acceleratedASR will be conducted to determine actualstructuralimpact of ASR. Structuralperformance will be establishedbased on correlationbetween the structuraltesting results and observed expansion levels/crack mapping. Large scale tests will confirm that parameters being monitored are appropriateto manage the effects of ASR and that acceptance criteriaused provides sufficient margin.

The potential impact of ASR on the structural strength of concrete is a consequence of strains resulting from the expansive gel. A direct method of monitoring ASR impact to a structural element would be to measure the accumulated strain. Because strain measuring devices were not installed during original construction, the accumulated strain to date cannot be directly measured. However, cracking can be used as an indication of accumulated strain. Monitoring of surface cracking and specifically crack mapping is the most effective way to correlatethe accumulatedexpansion in the structures.

NextEra Energy is currently using the CCI method to monitor and trend ASR expansion at Seabrook Station. The CCI is well-suited as a correlating parameter between the test specimens and reinforced concrete surfaces at Seabrook Station.

Core bores will be obtained from the test specimens for petrographic examination to confirm the presence of ASR. Qualitative assessments of the severity of ASR will also be performed, including a comparison of the severity of ASR through the depth of the test specimen. These assessments will use both the

United States Nuclear Regulatory Commission Page 9 of 18 SBK-L-13 162 / Enclosure 1 visual assessment ratingand the damage ratingindex, both of which have been used in evaluation of cores from Seabrook Station. This petrographic examination will further validate the CCI correlation between test and in situ measurements.

In addition to large scale destructive testing of concrete beams, anchor testing will be performed to quantify the impact of ASR on anchor capacity as a function of the severity of ASR degradation.

In the event these test results indicate a need to amend the monitoringprogram, NextEra will take such action.

Acceptance Criteria Several published studies describe screening methods to determine when structural evaluations of ASR affected concrete are appropriate and how to prioritize such evaluations. In MPR-3 727, Revision 0, Seabrook Station. Impact of Alkali-Silica Reaction on Concrete Structures and Attachments, these studies were reviewed for applicability. The report concludes: "while these screening methods are based on lightly or unreinforced concrete structures, they are useful in the absence of criteria directly relevant to the highly-reinforced concrete structures used in nuclear generatingfacilities," and, "Confinement provided by reinforcing steel and other restraints is a key .factor regarding the impact of ASR on reinforced concrete structures. Confinement limits ASR expansion of the in situ structure, which reduces the extent of deleterious cracking and the resultant reduction in concrete properties.

NextEra has performed a baseline inspection and ASR associated cracks have been evaluated and categorized. NextEra has assessed 131 accessible areas to date in this manner. The areas affected by ASR have been identified and assessed for apparent degradation from ASR, including estimation of in situ expansion. The results are presented in MPR-3727, Revision 0, Seabrook Station. hnpact of Alkali-Silica Reaction on Concrete Structures and Attachments. This report provides recommendations and the basis for thresholds used in the Alkali-Silica Reaction (ASR) Monitoring Program. The ASR Monitoring Program establishes the screening criteria for ASR affected areas. ASR affected areas are screened and categorized for Qualitative or Quantitative Monitoring and Trending and Structural Evaluation.

A Combined Cracking Index (CCI) of less than the 1.0 mmn/m and Individual Crack Width of less than 1.0 mm can be deemed Acceptable with Deficiencies. Areas with deficiencies determined to be acceptable with further review are trended for evidence of further degradation. Documentation and trend data will be maintained in accordance with the Structures Monitoring Program and established guidelines of ACI 349.3R, "Structural Cendifien Assessment .. f.

iding&.. ,-,"Evaluation of Existing Nuclear Safety-Related Concrete Structures."

Areas with a CCI of 1.0 mmn/m or greater, or an Individual Crack Width of 1.0 mm or greater are deemed unacceptable and require a structural evaluation. This

United States Nuclear Regulatory Commission Page 10 of 18 SBK-L- 13162 / Enclosure 1 evaluation is performed to ensure impacted structures are in compliance with the Current Licensing Basis and is documented in the Corrective Action Program. The engineering evaluation will be typical to those structural evaluations that have already been performed for Alkali-Silica Reaction. The Engineering Evaluation will consider the need to perform a detailed appraisal to determine potential capacity reductions, or the need to perform special studies, testing, and monitoring.

Additionally, Corrective Actions requiring repair are entered in the Work Control Program for implementation.

Evaluation Evaluations are performed to ensure impacted structures are in compliance with the Current Licensing Basis. These evaluations are documented in the Corrective Action Program. Additionally, Corrective Actions requiring repair are entered in the Work Control Program for implementation. Deficiencies determined to be acceptable by engineering review are trended for evidence of further degradation.

Conclusion To manage the aging effects of cracking due to expansion and reaction with aggregates in concrete structures, the existing Structures Monitoring Program, B.2.1.31, has been augmented by this plant specific Alkali-Silica Reaction (ASR)

Monitoring Program, B.2.1.31 A. The ASR Monitoring Program will be structured according to the guidelines in ACI 349.3R, "Structural Condition A4sses-sment B.i.i.... "-. "Evaluation of Existing Nuclear Safety-Related Concrete Structures."

A Combined Cracking Index (CCI) of less than the 1.0 mm/m and Individual Crack Width of less than 1.0 mm can be deemed Acceptable with Deficiencies. Areas with deficiencies determined to be acceptable with further review are trended for evidence of further degradation. A CCI of 1.0 mm/m or greater, or an Individual Crack Width of 1.0 mm or greater are deemed Unacceptable and require further evaluation.

NextEra has performed a baseline inspection and ASR associated cracks have been evaluated and categorized. NextEra has assessed 131 accessible areas to date in this manner. The areas affected by ASR have been identified and assessed for apparent degradation from ASR, including estimation of in situ expansion.

Monitoring of CI and Individual Crack Width of at least 20 areas identified in the baseline inspection as having the CCI will be performed at six month intervals.

Measurement of Cracking Index and Individual Crack Width will be performed in the same areas as the baseline. Trend data from these follow-up inspections will be used in determining the progression of ASR and a basis for any change to the frequency of the inspection.

United States Nuclear Regulatory Commission Page 11 of 18 SBK-L-13162/ Enclosure 1 Program Elements ELEMENT 3 - PARAMETERS MONITORED/[NSPECTED The Alkali-Silica Reaction (ASR) Monitoring Program manages the effects of cracking due to expansion and reaction with aggregates. The potential impact of ASR on the structural strength and anchorage capacity of concrete is a consequence of strains resultingfrom the expansive gel. These strains produce the associatedcracking.

The program focuses on identifying evidence of ASR, either past or present, which could lead to expansion due to reaction with aggregates. The program is consistent with published guidance for condition assessment of structures affected by ASR.

ASR is detected by visual observation of cracking on the surface of the concrete.

The cracking is typically accompanied by the presence of moisture and efflorescence. Concrete affected by expansive ASR is typically characterized by a network or "pattern" of cracks. ASR involves the formation of an alkali-silica gel which expands when exposed to water. Microcracking due to ASR is generated through forces applied by the expanding aggregate particles and/or swelling of the alkali-silica gel within and around the boundaries of reacting aggregate particles.

The ASR gel may exude from the crack forming white secondary deposits at the concrete surface. The gel also often causes a dark discoloration of the cement paste surrounding the crack at the concrete surface. Visual observation of the conditions described above is used to identify the presence of ASR.

Cracking Index A Cracking Index is determined for accessible surfaces exhibiting ASR pattern cracking. The process for determining the Cracking Index (CI) is described in the Federal Highway Administration (FHWA) document FHWA-HIF-09-004, "Report on the Diagnosis, Prognosis, and Mitigation of Alkali-Silica Reaction (ASR) in TransportationStructures." The FHWA Cracking Index is the summation of the crack widths on the horizontal or vertical sides of a 0.5m (20-inch) by 0.5m (20-inch) square on the ASR-affected concrete surface. Since each side of the square is 0.5 m, the Cracking Index in a given direction is reported in units of mm/m.

The horizontal and vertical Cracking Indices are averaged to obtain a Combined Cracking Index (CCI) for each area of interest. The CC! represents the , pni along the entire perimeter-o,f the 20 in.h by 20 inch square . Screening criteria used in the assessment of expansion is expressed in terms of CCI and based on recommendations provided in MPR-3 727, Revision 0, Seabrook Station: Impact of Alkali-Silica Reaction on Concrete Structures andAttachments.

The progression of ASR degradation of the concrete is an important consideration for assessing the long term implications of ASR and specifying monitoring intervals. The most reliable means for establishing the progression of ASR is to monitor expansion of the in situ concrete in-s-it .

United States Nuclear Regulatory Commission Page 12 of 18 SBK-L- 13162 / Enclosure I Crack Width Crack Width is a measurement of an individual crack width and is reported in units of mm. Progression of ASR is monitored by periodically measuring the individual crack width. Screening criteria used in the assessment of individual crack width are based on recommendations provided in MPR-3727, Revision 0, Seabrook Station:

Impact ofAlkali-Silica Reaction on Concrete Structuresand Attachments.

ELEMENT 4 - DETECTION OF AGING EFFECTS ASR is detected by visual inspections performed by qualified individuals. These individuals must either be a licensed Professional Engineer experienced in this area, or will work under the direction of a licensed Professional Engineer.

The Seabrook Station Alkali-Silica Reaction (ASR) Monitoring Program provides for management of aging effects due to the presence of ASR. Program scope includes concrete structures within the scope of License Renewal.

ASR is a reaction that occurs within the structure of the concrete due to the reaction between silica from the aggregate and alkali constituents in the cement.

The reaction produces a gel that expands as it absorbs moisture. Expansion of the gel exerts tensile stress on the concrete resulting in cracking. Surface cracking is a directphysical manifestation of the expansion induced by ASR within the core of the structuralmember.

The cracking propagates on the surface of the concrete where it is visually identifiable. The degree of cracking is most severe at the surface of the concrete due to severalfactors. The surface or cover concrete extends beyond the steel reinforcing bars. Because this surface is not within the steel reinforcedpart of the wall, the concrete is free to expand as the ASR gel is formed and ultimately expands. Additionally, the surface of the wall is subject to wetting and drying which can increase the flow of alkalis in this area. Consequently, the exposed surface will experience the largest and most visible cracking. This makes monitoring of the surface cracks an appropriateand reliable diagnostictoolfor monitoringthe progression of ASR.

Typical cracking resulting from ASR is described as "pattern" or "map" cracking and is usually accompanied by a dark staining adjacent to the cracks at the surface of the structure. The ASR gel may exude from the crack forming white secondary deposits at the concrete surface. Visual indications of pattern cracking, which are often accompanied by staining or residual gel deposits, are documented and evaluated as deficiencies. To identify and verify the presence of ASR, the maximum crack width, a cracking index, and a description of the cracking including any visible surface discoloration are documented. Documentation and trend data will be maintained in accordance with the Structural Monitoring Program and established guidelines of ACI 349.3R, "Structural Condition A.ss.ssn..nt of Buildinfgs-.." "Evaluation of Existing Nuclear Safety-Related Concrete Structures."

United States Nuclear Regulatory Commission Page 13 of 18 SBK-L- 13162 / Enclosure 1 ELEMENT 5 - MONITORING AND TRENDING Monitoring of ASR associated cracks is an effective method for determining ASR progression. Monitoring of ASR associated cracks at Seabrook is implemented with two measurements. One is Cracking Index (CI) and the other is Individual Crack Width.

The CI is a quantitative assessment of cracking present in the cover concrete of affected structures. A CI measurement is taken on accessible surfaces exhibiting ASR pattern cracking. The process for determining the Cracking Index (CI) is described in the Federal Highway Administration (FHWA) document FHWA-HIF-09-004, "Report on the Diagnosis, Prognosis, and Mitigation of Alkali-Silica Reaction (ASR) in TransportationStructures." The FHWA Cracking Index is the summation of the crack widths on the horizontal or vertical sides of a 0.5m (20-inch) by 0.5m (20-inch) square on the ASR-affected concrete surface. Since each side of the square is 0.5 m, the Cracking Index in a given direction is reported in units of mmn/m.

The horizontal and vertical Cracking Indices are averaged to obtain a Combined Cracking Index (CCI) for each area of interest. The CC! represents the expan.ion alng the. ,ntir,-. pem-, r-of the 20 inch by 20 inch square. Screening criteria used in the assessment of expansion is expressed in terms of CCI and based on recommendations provided in MPR-3727, Revision 0, Seabrook Station: Impact of Alkali-Silica Reaction on Concrete Structures and Attachments. The CCI is compared to screening criteria and trended to establish the progression of ASR degradation.

The progression of ASR degradation of the concrete is an important consideration for assessing the long term implications of ASR and specifying monitoring intervals. The most reliable means for establishing the progression of degradation is to monitor expansion of the concrete in situ.

Individual Crack Width measurement is also an effective method for assessment of ASR affected areas. Screening criteria used in the assessment of Individual Crack Width, expressed in terms of mm, are based on recommendations provided in MPR-3727, Revision 0, "Seabrook Station: Impact of Alkali-Silica Reaction on Concrete Structuresand Attachments."

NextEra has performed a baseline inspection and ASR associated cracks have been evaluated and categorized. NextEra has assessed 131 accessible areas to date in this manner. The areas affected by ASR have been identified and assessed for apparent degradation from ASR, including estimation of in situ expansion. The results are presented in MPR-3727, Revision 0, "Seabrook Station: Impact of Alkali-Silica Reaction on Concrete Structures and Attachments." Monitoring of Cracking Index and Individual Crack Width of at least 20 areas identified in the baseline inspection as having the largest CCI will be performed at six month intervals. Measurement of Cracking Index and Individual Crack Width will be performed in the same areas as the baseline. Trend data from these follow-up

United States Nuclear Regulatory Commission Page 14 of 18 SBK-L-13162 / Enclosure 1 inspections will-be is used in determining the progression of expansion and a basis for any change to the frequency of the inspection. Documentation and trend data will be maintained in accordance with the Structural Monitoring Program and established guidelines of ACI 349.3R, "Stuctural coidi*ien As.sess.men.t,]

B....ings. "Evaluation of Existing Nuclear Safety-Related Concrete Structures."

Deficiencies being repaired or trended are subject to follow-up inspections of increased frequency. Newly discovered areas exhibiting visual signs of ASR are identified during routinely performed Structural Monitoring Program inspections and documented as deficiencies. Deficiencies are reviewed in accordance with the Structural Monitoring Program and established guidelines of ACI 349.3R, "Strc.tural ondition As.sessnt. of Buildings." "Evaluationof Existing Nuclear Safety-Related Concrete Structures."

ELEMENT 6 - ACCEPTANCE CRITERIA Several published studies describe screening methods to determine when structural evaluations of ASR affected concrete are appropriate and how to prioritize such evaluations. In MPR-3 727, Revision 0, "Seabrook Station: Impact of Alkali-Silica Reaction on Concrete Structures and Attachments," these studies were reviewed for applicability. The report concludes: "while these screening methods are based on lightly or unreinforced concrete structures, they are useful in the absence of criteria directly relevant to the highly-reinforced concrete structures used in nuclear generatingfacilities," and, "Confinement provided by reinforcing steel and other restraints is a key factor regarding the impact of ASR on reinforced concrete structures. Confinement limits ASR expansion of the in situ structure, which reduces the extent of deleterious cracking and the resultant reduction in concrete properties."

NextEra has performed a baseline inspection and ASR associated cracks have been evaluated and categorized. NextEra has assessed 131 accessible areas to date in this manner. The areas affected by ASR have been identified and assessed for apparent degradation from ASR, including estimation of in situ expansion. The results are presented in MPR-3727, Revision 0, "Seabrook Station. Impact of Alkali-Silica Reaction on Concrete Structures and Attachments." Based on site specific assessment and review of industry source documentation this report provides recommendations for screening thresholds used in the Alkali-Silica Reaction (ASR) Monitoring Program. Using these thresholds, ASR affected areas are screened and categorized for Qualitative or Quantitative Monitoring and Trending and Structural Evaluation.

A Combined Cracking Index (CCI) of less than the 1.0 mm/m and Individual Crack Width of less than 1.0 mm can be deemed Acceptable with Deficiencies. Areas with deficiencies determined to be acceptable with further review are trended for evidence of further degradation.

A CCI of 1.0 mmn/m or greater, or an Individual Crack Width of 1.0 mm or greater are deemed Unacceptable and require a structural evaluation. Structural evaluations are performed to ensure impacted structures are in compliance with the

United States Nuclear Regulatory Commission Page 15 of 18 SBK-L-13162 / Enclosure 1 Current Licensing Basis are documented in the Corrective Action Program. The engineering evaluation will be typical to those structural evaluations that have already been performed for Alkali-Silica Reaction. The Engineering Evaluation will consider the need to perform a detailed appraisal to determine potential capacity reductions, or the need to perform special studies, testing, and monitoring.

Additionally, Corrective Actions requiring repair are entered in the Work Control Program for implementation.

ELEMENT 10 - OPERATING EXPERIENCE The primary source of OE, both industry and plant specific, was the Seabrook Station Corrective Action Program documentation. The Seabrook Station Corrective Action Program is used to document review of relevant external OE including INPO documents, NRC communications and Westinghouse documents, and plant specific OE including corrective actions, maintenance work orders generated in response to a structure, system or component deficiencies, system and program health reports, self-assessment reports and NRC and INPO inspection reports.

The Seabrook Station Corrective Action Program is used to track, trend and evaluate plant issues and events. Those issues and events, whether external or plant specific, that are potentially significant to the ASR Monitoring Program are evaluated. The ASR Monitoring Program is revised, as appropriate, if these evaluations show that program changes will enhance program effectiveness.

Historically, NextEra Energy Seabrook (NextEra) has experienced groundwater infiltration through cracks, capillaries, pore spaces, seismic isolation joints, and construction joints in the below grade walls of concrete structures. Some of these areas have shown signs of leaching, cracking, and efflorescence on the concrete due to the infiltration. During the early 1990's an evaluation was conducted to assess the effect of the groundwater infiltration on the serviceability of the concrete walls. That evaluation concluded that there would be no deleterious effect, based on the design and placement of the concrete and on the non-aggressive nature of the groundwater.

In 2009, NextEra tested seasonal groundwater samples to support the development of a License Renewal Application. The results showed that the groundwater had become aggressive and NextEra initiated a comprehensive review of possible effects to in-scope structures.

A qualitative walkdown of plant structures was performed and the "B" Electrical Tunnel was identified as showing the most severe indications of groundwater infiltration. Concrete core samples from this area were removed, tested for strength and elasticity values, and subjected to petrographic examinations. While the results showed that both strength and elasticity values had declined, they remained within the design margin. The results of the petrographic examinations also showed that the samples had experienced Alkali-Silica Reaction (ASR).

United States Nuclear Regulatory Commission Page 16 of 18 SBK-L-13162 / Enclosure 1 NextEra initiated an extent of condition evaluation and concrete core samples were taken from five additional areas of the plant - areas that showed characteristics with the greatest similarity to the "B" Electrical Tunnel. Additional concrete core samples were also taken from an expanded area around the original concrete core samples in the "B" Electrical Tunnel.

Tests on these core samples confirmed that the original "B" Electrical Tunnel core samples show the most significant ASR. For the five additional areas under investigation, final results of compressive strength and modulus testing indicate that the compressive strength in all areas is greater than the strength required by the design of the structures. Modulus of elasticity was in the range of the expected value except for the Diesel Generator, Containment Enclosure Buildings, Emergency Feedwater Pumphouse, and the Equipment Vaults which were less than the expected value in localized areas.

Evaluation of the test results shows that the affected structures to be fully capable of performing their safety function but margin had been reduced. The areas are potentially subject to further degradation of material properties due to the effects of ASR.

A review of industry related operating experience related to ASR was performed.

The review includes NRC generic communications issued such as Generic Letters, Bulletins, and Information Notices. Industry operating experience is discussed below:

1. In 1994, ASTM Standard C289 was clarified to caution that the tests described may not accurately predict aggregate reactivity when dealing with late- or slow-expanding aggregates containing strained quartz or microcrystalline quartz.

2. On August 4, 2010, NRC issued Information Notice (IN) 2010-14 "Containment Concrete Surface Condition Examination Frequency and Acceptance Criteria." Seabrook's assessment resulted in updating the IWL program with inspection guidelines typical to ACI 349.3R.

3. OE 34348 Operating Experience report submitted by NextEra: Preliminary -

Reduction in Concrete Properties Due to Distress from Alkali-Silica Reaction (ASR) was issued September 30, 2011.

4. On November 11, 2011, NRC issued Information Notice (IN) 2011-20, "Concrete Degradation by Alkali-Silica Reaction" to notify the industry based on the Seabrook issue.

The Seabrook Station plant specific operating experience identified the following:

1. As part of the Seabrook License Renewal process, the aggressiveness of the groundwater chemistry on concrete structures in contact with groundwater/soil must be determined. The first two samples collected in June 2009 from well locations indicate Chloride levels >500 PPM. Since the chloride levels exceed the acceptable limits, the groundwater was considered aggressive. Groundwater chemistry is now being performed

United States Nuclear Regulatory Commission Page 17 of 18 SBK-L- 13162 / Enclosure 1 every five years via the Structures Monitoring Program and awareness of the aggressive water chemistry is considered in the inspection of concrete.

2. Concrete core samples were removed from the Control Building B-Electrical Tunnel in April 2010 and Penetration Resistance Tests (PRT) performed. The results of the PRT indicated an average concrete compressive strength of 5340 psi and the concrete core testing indicated an average compressive strength of 4790 psi. An engineering evaluation was performed to review the results of core samples. The results show the structure to be fully capable of performing its safety function but margin had been reduced.

3. In September 2010, Testing was completed on concrete core samples removed from the walls at elevation (-) 20' in the Control Building B-Electrical Tunnel. The results of the modulus testing indicate that the average measured elastic modulus to be less than the expected elastic.

modulus. An engineering evaluation was performed to review the results of core samples. The results showed the structure to be fully capable of performing its safety function but margin had been reduced. Results of the petrographic examination completed in September 2010 showed that the samples had experienced Alkali-Silica Reaction (ASR).

4. Inspection performed April 2011 on elevation (-) 30' in the Containment CEVA identified a craze cracking pattern on a localized area of the Containment concrete shell. The observed craze cracks are tight, less than 40 mils width and contain fine dry white deposits. There were no other indications of degradation or distress in the concrete in this area. Craze cracking with white deposits suggests the presence of alkali-silica reaction.

Additional inspections of the exterior face of the Containment Structure were performed in September 2011. These two locations have been included in the second-tier evaluation criteria of the program due to the past groundwater in-leakage and follow-up inspections will be performed. Any identified crack growth will require additional evaluation. NextEra has previously committed to maintaining the exterior surface of the Containment Structure in a dewatered state (LRA Commitment #52).

5. In June 2011, as an extent of condition evaluation, five (5) building structures were investigated to determine if the building concrete was affected by alkali-silica reaction (ASR). The five (5) building structures were selected based on visual indications on the inside concrete surfaces and exposure to groundwater infiltration. An engineering evaluation was performed to review the results of concrete core samples. The results show these structures to be fully capable of performing their safety function but margin had been reduced. A root cause evaluation was performed to determine how the ASR developed and why its presence was not identified until 2010. The following root causes were identified:

- The ASR developed because the concrete mix designs utilized an aggregate that was susceptible to Alkali-Silica Reaction, which was not

United States Nuclear Regulatory Commission Page 18 of 18 SBK-L- 13162 / Enclosure 1 known at the time. Although the testing was conducted in accordance with ASTM standards, those testing standards were subsequently identified as limited in their ability to predict long term ASR.

- The health monitoring program for systems and structures does not contain a process for periodic reassessment of failure modes that were excluded from the monitoring criteria to ensure that the monitoring/mitigating strategies remain applicable and effective.

6. In April 2012, results were published of accessible area inspections performed to identify and categorize locations of ASR distress. The areas affected by ASR have been identified and assessed for apparent degradationfrom ASR, including estimation of in situ expansion. The results are presentedin MPR-3727, Revision 0, Seabrook Station: Impact ofAlkali-Silica Reaction on Concrete Structures and Attachments.

There were approximately 131 localized accessible areas with ASR distress. At least 20 areas having a Combined Cracking Index of 1.0 minnm or greater were categorized as requiring "FurtherEvaluation" and follow-up inspections are performed every six months. Follow-up inspection of these areas performed in June 2012, December 2012 and June 2013 show no notable expansion of concrete due to ASR distress.

Areas that were designated ASR "Acceptable with deficiencies" have a follow-up inspectionfrequency of 2 2years.

Trend datafrom these follow-up inspections will be used in determining the progression of ASR degradation and a basis for any change to the frequency of the inspection of ASR affected areas.

7. In October 2012, a structuralevaluation was performed to assess observed cracking at Azimuth 2700 of the containment building. The observed cracking meets the Tier 3 criteria which necessitates a structural evaluation. Although pattern cracking was noted at this location, the cracked area does not exhibit other indications of ASR (e.g., staining, deposits, etc.). Nevertheless, it was conservatively treated as if it were ASR and a structural evaluation was performed which concluded that containment isfully capable of meeting all its designfunctions.

Enclosure 2 to SBK-L-13162 LRA Appendix A - Final Safety Report Supplement Table A.3, License Renewal Commitment List Updated to Reflect Changes

United States Nuclear Regulatory Commission Page 2 of 14 SBK-L-13162 / Enclosure 2 A.3 LICENSE RENEWAL COMMITMENT LIST UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Program to be implemented prior to the period of extended operation.

Inspection plan to be submitted to An inspection plan for Reactor Vessel Internals will be NRC not later than 2 years after submitted for NRC review and approval. receipt of the renewed license or not less than 24 months prior to the period of extended operation, whichever comes first.

Enhance the program to include visual inspection for cracking, Prior to the period of extended

2. Closed-Cycle Cooling Water loss of material and fouling when the in-scope systems are A.2.1.12 operation opened for maintenance.

Inspection of Overhead Enhance the program to monitor eneral corrosion on the crane Heavy Load and Light Load nancthey rogram tomponito gnderal fc s of n the Ane Prior to the period of extended (Related to Refueling) and trolley structural components and the effects of wear on the A.2.1.13 operation Handling Systems rails in the rail system.

Inspection of Overhead Heavy Load and Light Load Prior to the period of extended

4. Heatedto Lightling) Enhance the program to list additional cranes for monitoring. A.2.1.13 operation (Related to Refueling)oprtn Handling Systems Enhance the program to include an annual air quality test Prior to the period of extended

5. Compressed Air Monitoring requirement for the Diesel Generator compressed air sub A.2.1.14 operation system.

6. Fire Protection Enhance the program to perform visual inspection of Prior to the period of extended penetration seals by a fire protection qualified inspector. operation.

United States Nuclear Regulatory Commission Page 3 of 14 SBK-L- 13162 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Enhance the program to add inspection requirements such as Prior to the period of extended

7. Fire Protection spalling, and loss of material caused by freeze-thaw, chemical A.2.1.15 operation.

attack, and reaction with aggregates by qualified inspector.

Enhance the program to include the performance of visual Prior to the period of extended

8. Fire Protection inspection of fire-rated doors by a fire protection qualified A.2.1.15 operation.

inspector.

Enhance, the program to include NFPA 25 guidance for "where sprinklers have been in place for 50 years, they shall be Prior to the period of extended

9. Fire Water System replaced or representative samples from one or more sample A.2.1.16 operation.

areas shall be submitted to a recognized testing laboratory for field service testing".

Enhance the program to include the performance of periodic Prior to the period of extended

10. Fire Water System flow testing of the fire water system in accordance with the A.2.1.16 operation.

guidance of NFPA 25.

Enhance the program to include the performance of periodic visual or volumetric inspection of the internal surface of the fire protection system upon each entry to the system for routine or corrective maintenance. These inspections will be documented and trended to determine if a representative number of Within ten years prior to the period

11. Fire Water System inspections have been performed prior to the period of extended A.2.1.16 of extended operation.

operation. If a representative number of inspections have not been performed prior to the period of extended operation, focused inspections will be conducted. These inspections will be performed within ten years prior to the period of extended operation.

Enhance the program to include components and aging effects A.2.1.17 Prior to the period of extended

12. Aboveground Steel Tanks required by the Aboveground Steel Tanks. operation.

United States Nuclear Regulatory Commission Page 4 of 14 SBK-L- 13162 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Enhance the program to include an ultrasonic inspection and Within ten years prior to the period

13. Aboveground Steel Tanks evaluation of the internal bottom surface of the two Fire A.2.1.17 of extended operation.

Protection Water Storage Tanks.

Enhance program to add requirements to 1) sample and analyze

14. Fuel Oil Chemistry new fuel deliveries for biodiesel prior to offloading to the A.2.1.18 Prior to the period of extended Auxiliary Boiler fuel oil storage tank and 2) periodically operation.

sample stored fuel in the Auxiliary Boiler fuel oil storage tank.

Enhance the program to add requirements to check for the Prior to the period of extended

15. Fuel Oil Chemistry presence of water in the Auxiliary Boiler fuel oil storage tank at A.2.1.18 operation.

least once per quarter and to remove water as necessary.

Enhance the program to require draining, cleaning and Prior to the period of extended

16. Fuel Oil Chemistry inspection of the diesel fire pump fuel oil day tanks on a A.2.1.18 operation.

frequency of at least once every ten years.

Enhance the program to require ultrasonic thickness measurement of the tank bottom during the 10-year draining, Prior to the period of extended

17. Fuel Oil Chemistry cleaning and inspection of the Diesel Generator fuel oil storage A.2.1.18 operation.

tanks, Diesel Generator fuel oil day tanks, diesel fire pump fuel oil day tanks and auxiliary boiler fuel oil storage tank.

I all pulled and tested Enhance the program to specify that Prior to the period of extended

18. Reactor Vessel Surveillance capsules, unless discarded before August 31, 2000, are placed A.2.1.19 operation.

in storage.

United States Nuclear Regulatory Commission Page 5 of 14 SBK-L-13162 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Enhance the program to specify that if plant operations exceed the limitations or bounds defined by the Reactor Vessel

19. Reactor Vessel Surveillance Surveillance Program, such as operating at a lower cold leg Prior to the period of extended temperature or higher fluence, the impact of plant operation operation.

changes on the extent of Reactor Vessel embrittlement will be evaluated and the NRC will be notified.

Enhance the program as necessary to ensure the appropriate withdrawal schedule for capsules remaining in the vessel such that one capsule will be withdrawn at an outage in which the

20. Reactor Vessel Surveillance capsule receives a neutron fluence that meets the schedule A.2.1.19 Prior to the period of extended requirements of 10 CFR 50 Appendix H and ASTM E185-82 operation.

and that bounds the 60-year fluence, and the remaining capsule(s) will be removed from the vessel unless determined to provide meaningful metallurgical data.

Enhance the program to ensure that any capsule removed,

21. Reactor Vessel Surveillance without the intent to test it, is stored in a manner which A.2.1.19 Prior to the period of extended maintains it in a condition which would permit its future use, operation.

including during the period of extended operation.

A.2.1.20 Within ten years prior to the period

22. One-Time Inspection Implement the One Time Inspection Program. A.2.1.20 extended prio of extended operation.

Implement the Selective Leaching of Materials Program. The Selective Leaching of program will include a one-time inspection of selected Within five years prior to the period

23. Leals components where selective leaching has not been identified A.2.1.21 of extended operation.

.Materials and periodic inspections of selected components where selective leaching has been identified.

24. Buried Piping And Tanks Implement the Buried Piping And Tanks Inspection Program. A.2.1.22 Within ten years prior to entering Inspection the period of extended operation

United States Nuclear Regulatory Commission Page 6 of 14 SBK-L-13 162 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION One-Time Inspection of Implement the One-Time Inspection of ASME Code Class I Within ten years prior to the period

25. ASME Code Class I Small Small Bore-Piping Program. A.2.1.23 of extended operation.

Bore-Piping Enhance the program to specifically address the scope of the program, relevant degradation mechanisms and effects of

26. External Surfaces Monitoring interest, the refueling outage inspection frequency, the A.2.1.24 Prior to the period of extended inspections of opportunity for possible corrosion under operation.

insulation, the training requirements for inspectors and the required periodic reviews to determine program effectiveness.

Inspection of Internal

27. Surfaces in Miscellaneous Implement the Inspection of Internal Surfaces in Miscellaneous A.2.1.25 Prior to the period of extended Piping and Ducting Piping and Ducting Components Program. operation.

Components Enhance the program to add required equipment, lube oil

28. Lubricating Oil Analysis analysis required, sampling frequency, and periodic oil A.2.1.26 prratote changes. operation.

29. Lubricating Oil Analysis Enhance the program to sample the oil for the Reactor Coolant A.2.1.26 Prior to the period of extended pump oil collection tanks. operation.

Enhance the program to require the performance of a one-time

30. Lubricating Oil Analysis ultrasonic thickness measurement of the lower portion of the A.2.1.26 Prior to the period of extended Reactor Coolant pump oil collection tanks prior to the period of operation.

extended operation.

31. ASME Section XI, Enhance procedure to include the definition of "Responsible A.2.1.28 Prior to the period of extended Subsection IWL Engineer". operation.

United States Nuclear Regulatory Commission Page 7 of 14 SBK-L-13162 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Structures Monitoring Enhance procedure to add the aging effects, additional Prior to the period of extended

32. Plocations, inspection frequency and ultrasonic test A.2.1.31 operation.

Program requirements.

Structures Monitoring Enhance procedure to include inspection of opportunity when Prior to the period of extended

33. Program planning excavation work that would expose inaccessible A.2.1.31 operation.

concrete.

Electrical Cables and Connections Not Subject to Implement the Electrical Cables and Connections Not Subject CFR 50.49 10 CFR 50.49 Environmental10 to 10CR Environmental Qualification Requirements A.2.1.32 Prior to the period of extended 5 .49Envionmetalprogam.operation.

Qualification Requirements program.

Electrical Cables and Connections Not Subject to Implement the Electrical Cables and Connections Not Subject

3. 10 CFR 50.49 Environmental peetteEetiaCalsadoncinsNtSbctPrior to the period of extended

35. 1 504Eniom to 10 CFR 50.49 Environmental Qualification Requirements A.2.1.33 operaton.

Qualification Requirements Used in Instrumentation Circuits program.operation.

Used in Instrumentation Circuits Inaccessible Power Cables 3notcesuibje tow10 CFR 5049 Implement the Inaccessible Power Cables Not Subject to 10 3.Not36.Envronentl 10 Subject to CFR 50.49 Qaliicaion CR5.9EvrnetlQaictoneqrmnsA2134 CFR 50.49 Environmental Qualification Requirements A.2.1.34 Prior oeain to the period of extended Environmental Qualification program. operation.

Requirements Prior to the period of extended

37. Metal Enclosed Bus Implement the Metal Enclosed Bus program. A.2.1.35 operaton.

operation.

Prior to the period of extended

38. Fuse Holders Implement the Fuse Holders program. A.2.1.36 operation.

United States Nuclear Regulatory Commission Page 8 of 14 SBK-L- 13162 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Electrical Cable NotraS bjetto 10Connections CFRne50io49Implement the Electrical Cable Connections Not Subject to 10

39. Not Subject to 10 CFR 50.'4 0.4 F 9EvrnetlQaictonRqrmnsA.137 Prior to the period of extended Environmental Qualification CFR 50.49 Environmental Qualification Requirements A.2.1.37 operation.

Requirements program.

Prior to the period of extended

40. 345 KV SF 6 Bus Implement the 345 KV SF 6 Bus program. A.2.2.1 oraton.

operation.

41. Metal Fatigue of Reactor Enhance the program to include additional transients beyond A.2.3.1 Prior to the period of extended Coolant Pressure Boundary those defined in the Technical Specifications and UFSAR. operation.

Metal Fatigue of Reactor Enhance the program to implement a software program, to Prior to the period of extended

42. Coolant Pressure Boundary count transients to monitor cumulative usage on selected A.2.3.1 operation.

components.

PThe updated analyses will be Pressure -Temperature Seabrook Station will submit updates to the P-T curves and submitted at the appropriate time to

43. Limits, including Low LTOP limits to the NRC at the appropriate time to comply with A.2.4.1.4 comply with 10 CFR 50 Appendix Temperature Overpressure 10 CFR 50 Appendix G. G, Fracture Toughness Protection Limits Requirements.

United States Nuclear Regulatory Commission Page 9 of 14 SBK-L-13162 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION NextEra Seabrook will perform a review of design basis ASME Class 1 component fatigue evaluations to determine whether the NUREG/CR-6260-based components that have been evaluated for the effects of the reactor coolant environment on fatigue usage are the limiting components for the Seabrook plant configuration. If more limiting components are identified, the most limiting component will be evaluated for the effects of the reactor coolant environment on fatigue usage. If the limiting location identified consists of nickel alloy, the environmentally-assisted fatigue calculation for nickel alloy will be performed using the rules of NUREG/CR-6909.

(I) Consistent with the Metal Fatigue of Reactor Coolant Pressure Boundary Program Seabrook Station will update the fatigue usage calculations using refined fatigue analyses, if necessary, to determine acceptable CUFs (i.e., less than 1.0) when accounting for the effects of the reactor water Environmentally-Assisted environment. This includes applying the appropriate Fen At least two years prior to entering

44. A.2.4.2.3 Fatigue Analyses (TLAA) factors to valid CUFs determined from an existing fatigue the period of extended operation.

analysis valid for the period of extended operation or from an analysis using an NRC-approved version of the ASME code or NRC-approved alternative (e.g., NRC-approved code case).

(2) If acceptable CUFs cannot be demonstrated for all the selected locations, then additional plant-specific locations will be evaluated. For the additional plant-specific locations, if CUF, including environmental effects is greater than 1.0, then Corrective Actions will be initiated, in accordance with the Metal Fatigue of Reactor Coolant Pressure Boundary Program, B.2.3.1. Corrective Actions will include inspection, repair, or replacement of the affected locations before exceeding a CUF of 1.0 or the effects of fatigue will be managed by an inspection program that has been reviewed and approved by the NRC (e.g., periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method accepted by the NRC).

United States Nuclear Regulatory Commission Page 10 of 14 SBK-L-13162 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION

45. Number Not Used

46. Protective Coating Enhance the program by designating and qualifying an A.2.1.38 Prior to the period of extended Monitoring and Maintenance Inspector Coordinator and an Inspection Results Evaluator. operation Enhance the program by including, "Instruments and Protective Coating Equipment needed for inspection may include, but not be Prior to the period of extended

47. Moniting ating limited to, flashlight, spotlights, marker pen, mirror, measuring A.2.1.38 operation Monitoring and Maintenance tape, magnifier, binoculars, camera with or without wide angle lens, and self sealing polyethylene sample bags."

48. Protective Coating Enhance the program to include a review of the previous two A.2.1.38 Prior to the period of extended Monitoring and Maintenance monitoring reports. operation Enhance the program to require that the inspection report is to

49. Protective Coating be evaluated by the responsible evaluation personnel, who is to A.2.1.38 Prior to the period of extended Monitoring and Maintenance prepare a summary of findings and recommendations for future operation surveillance or repair.

Within the next two refueling outages, ORI 5 or OR16, and

50. ASME Section XI, Perform UT testing of the containment liner plate in the vicinity A.2.1.27 repeated at intervals of no more than Subsection IWE of the moisture barrier for loss of material, five refueling outages

51. Number Not Used ASME Section XI, Implement measures to maintain the exterior surface of the

52. SubsectiontiwL Containment Structure, from elevation -30 feet to +20 feet, in a A.2.1.28 Ongoing dewatered state.

United States Nuclear Regulatory Commission Page 11 of 14 SBK-L-13162 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Replace the spare reactor head closure stud(s) manufactured Prior to the period of extended

53. Reactor Head Closure Studs from the bar that has a yield strength > 150 ksi with ones that A.2.1.3 operation.

do not exceed 150 ksi.

NextEra will address the potential for cracking of the primary to secondary pressure boundary due to PWSCC of tube-to-tubesheet welds using one of the following two options:

1) Perform a one-time inspection of a representative sample of tube-to-tubesheet welds in all steam generators to determine if PWSCC cracking is present and, if cracking is identified, resolve the condition through engineering evaluation justifying continued operation or repair the condition, as appropriate, and establish an ongoing monitoring program to perform routine Steam Generator Tube tube-to-tubesheet.weld inspections for the remaining life of the A.2.1.10 Complete Integrity steam generators, or

2) Perform an analytical evaluation showing that the structural integrity of the steam generator tube-to-tubesheet interface is adequately maintaining the pressure boundary in the presence of tube-to-tubesheet weld cracking, or redefining the pressure boundary in which the tube-to-tubesheet weld is no longer included and, therefore, is not required for reactor coolant pressure boundary function. The redefinition of the rector coolant pressure boundary must be approved by the NRC as part of a license amendment request.

55. Steam Generator Tube Seabrook will perform an inspection of each steam generator to A.2.1.10 Within five years prior to entering Integrity assess the condition of the divider plate assembly. the period of extended operation.

Closed-Cycle Cooling Water Revise the station program documents to reflect the EPRI Prior to entering the period of

56. Syst e Guideline operating ranges and Action Level values for A.2.1.12 extended operation.

System hydrazine and sulfates.

United States Nuclear Regulatory Commission Page 12 of 14 SBK-L-13162 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Closed-Cycle Cooling Water Revise the station program documents to reflect the EPRI Prior to entering the period of

57. System Guideline operating ranges and Action Level values for Diesel A.2.1.12 extended operation.

Generator Cooling Water Jacket pH.

Update Technical Requirement Program 5,1, (Diesel Fuel Oil

58. Fuel Oil Chemistry Testing Program) ASTM standards to ASTM D2709-96 and A.2.1.18 prratote ASTM D4057-95 required by the GALL XI.M30 Rev 1 operation.

Nickel Alloy Nozzles and The Nickel Alloy Aging Nozzles and Penetrations program will Prior to the period of extended

59. implement applicable Bulletins, Generic Letters, and staff A.2.2.3 operation.

.Penetrations accepted industry guidelines.

Buried Piping and Tanks Implement the design change replacing the buried Auxiliary Prior to entering the period of

60. Inspection Boiler supply piping with a pipe-within-pipe configuration with A.2.1.22 extended operation.

Inspection leak detection capability.

61. Compressed Air Monitoring Replace the flexible hoses associated with the Diesel Generator A.2.1.14 Within ten years prior to entering Program air compressors on a frequency of every 10 years. the period of extended operation.

Enhance the program to include a statement that sampling Prior to the period of extended

62. Water Chemistry frequencies are increased when chemistry action levels are A.2.1.2 operation.

exceeded.

Ensure that the quarterly CVCS Charging Pump testing is continued during the PEO. Additionally, add a precaution to the

63. Flow Induced Erosion test procedure to state that an increase in the CVCS Charging N/A Prior to the period of extended Pump mini flow above the acceptance criteria may be operation indicative of erosion of the mini flow orifice as described in LER 50-275/94-023.

United States Nuclear Regulatory Commission Page 13 of 14 SBK-L-13162 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Soil analysis shall be performed prior to entering the period of extended operation to determine the corrosivity of the soil in the vicinity of non-cathodically protected steel pipe within the Prior to entering the period of scope of this program. If the initial analysis shows the soil to extended operation.

Inspection be non-corrosive, this analysis will be re-performed every ten years thereafter.

Implement measures to ensure that the movable incore

65. Flux Thimble Tube detectors are not returned to service during the period of extended operation. N/A extended operation

66. Number Not Used Perform one shallow core bore in an area that was continuously

67. Structures Monitoring wetted from borated water to be examined for concrete Program degradation and also expose rebar to detect any degradation A.2.1.31 No later than December 31, 2015 such as loss of material.

68. Structures Monitoring Perform sampling at the leakoff collection points for chlorides, A.2.1.31 Starting January 2014 Program sulfates, pH and iron once every three months.

Open-Cycle Cooling Water Replace the Diesel Generator Heat Exchanger Plastisol PVC Prior to the period of extended

69. System lined Service Water piping with piping fabricated from AL6XN A.2.1.11 operation.

material.

Inspect the piping downstream of CC-V-444 and CC-V-446 to

70. Closed-Cycle Cooling Water determine whether the loss of material due to cavitation A.2.1.12 Within ten years prior to the period System induced erosion has been eliminated or whether this remains an of extended operation.

issue in the primary component cooling water system.

United States Nuclear Regulatory Commission Page 14 of 14 SBK-L-13162 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT SCHEDULE LOCATION Implement the Alkali-Silica Reaction (ASR) Monitoring

71. Alkali-Silica Reaction (ASR) Program. Testing will be performed to confirm that A.2I.31A Prior to entering the period of Monitoring Program parametersbeing monitored and acceptance criteria used are extended operation.

appropriateto manage the effects of ASR