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 ML 12312A0 17).

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).

Enclosure 1 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

Enclosure 1 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 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

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 established in

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 accelerated ASR confirms

parameters being monitored are appropriate to 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

Implement the Alkali-Silica Reaction

(ASR) Monitoring Program. Testing

Alkali-Silica Reaction will be performed to confirm that

71 (ASR) Monitoring parameters being monitored and

Program acceptance criteria used are

appropriate to manage the effects of

ASR.

SCHEDULE

Prior to entering the

period of extended

operation.
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. inE mirocacks 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 reactive forms 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 resulting from 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 resulting from 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 SafetyRelated 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

evaluation)

Structural Evaluation 1.0 mm/m or greater 1.0 mm or greater

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 nonnuclear 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-HIF09-004, "Report on the Diagnosis, Prognosis, and Mitigation of Alkali-Silica

Reaction (ASR) in Transportation Structures. " 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 Structures and 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."

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

Basis for use of the crack index methodology is as follows:

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 approximated as 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 account for strain in the un-cracked concrete between

cracks (Ebd, the Cracking Index provides a reasonable approximation of the

total strain applied to the concrete after crack initiation.

Rebar in Tension -

AL = L2 - L1

Cracks L2 LI

Figure A-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 consolidate the expansion assessment to a single parameter.

Large Scale Testing

While the monitoring program action levels are currently based on the generic

information available in the literature, a full scale testing program has been

undertaken to refine the impacts for structures similar 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

reinforcement anchorage. Specific objectives of the testing are as follows:

" Determine the extent to which the shear performance, development length

of reinforcement, and flexural stiffness of reinforced concrete beams are

affected as a function of ASR-related expansion.

" Determine the effectiveness of retrofit techniques in the enhancement of

shear performance and on the development length in reinforced concrete

beams.

Large scale destructive testing of concrete beams with accelerated ASR will be

conducted to determine actual structural impact of ASR. Structural performance

will be established based on correlation between the structural testing results and

observed expansion levels/crack mapping. Large scale tests will confirm that

parameters being monitored are appropriate to manage the effects of ASR and

that acceptance criteria used 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 correlate the accumulated expansion 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 rating and the damage rating index, 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 monitoring program,

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 generating facilities, " 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 resulting from the expansive gel. These strains produce

the associated cracking.

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

Transportation Structures. " 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 and Attachments.

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 Structures and 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

direct physical manifestation of the expansion induced by ASR within the core of

the structural member.

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 several factors. The surface or cover concrete extends beyond the steel

reinforcing bars. Because this surface is not within the steel reinforced part 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 appropriate and reliable diagnostic tool for

monitoring the 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-HIF09-004, "Report on the Diagnosis, Prognosis, and Mitigation of Alkali-Silica

Reaction (ASR) in Transportation Structures. " 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 Structures and 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." "Evaluation of 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 generating facilities," 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 nonaggressive 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 BElectrical 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 BElectrical 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

degradation from ASR, including estimation of in situ expansion. The

results are presented in MPR-3 727, 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 "Further Evaluation" 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 inspection frequency of 2 2 years.

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.

7. In October 2012, a structural evaluation 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 is fully capable of meeting all its design functions.
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

SBK-L-13162 / Enclosure 2

Page 2 of 14

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 nanc they 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. Heated to 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

SBK-L- 13162 / Enclosure 2

Page 3 of 14

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

SBK-L- 13162 / Enclosure 2

Page 4 of 14

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.

Enhance the program to specify that I all pulled and tested 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

SBK-L-13162 / Enclosure 2

Page 5 of 14

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

SBK-L-13 162 / Enclosure 2

Page 6 of 14

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

SBK-L-13162 / Enclosure 2

Page 7 of 14

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

10 CFR 50.49 Environmental to 10 CFR 50.49 Environmental Qualification Requirements A.2.1.32 Prior to the period of extended 10 CR 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 3 notcesuibje tow10 CFR 5049 Implement the Inaccessible Power Cables Not Subject to 10 3.Not Subject to 10 CFR 50.49 CR5.9EvrnetlQaictoneqrmnsA2134 Prior to the period of extended 36.Envronentl Qaliicaion CFR 50.49 Environmental Qualification Requirements A.2.1.34 oeain

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

SBK-L- 13162 / Enclosure 2

Page 8 of 14

UFSAR

No. PROGRAM or TOPIC COMMITMENT SCHEDULE

LOCATION

Electrical Notra Cable Connections S bjet to 10 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.

P The 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

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UFSAR

No. PROGRAM or TOPIC COMMITMENT SCHEDULE

LOCATION

44. Environmentally-Assisted

Fatigue Analyses (TLAA)

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 environmentallyassisted 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

environment. This includes applying the appropriate Fen

factors to valid CUFs determined from an existing fatigue

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).

A.2.4.2.3 At least two years prior to entering

the period of extended operation.
United States Nuclear Regulatory Commission

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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

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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-totubesheet 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

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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

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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

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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 parameters being monitored and acceptance criteria used are extended operation.

appropriate to manage the effects of ASR