Enclosure 1 to this Letter Contains Proprietary Information 

Withhold Enclosure 1 from Public Disclosure in Accordance with 10 CFR 2.390 

U.S. Nuclear Regulatory Commission 

Attn: Document Control Desk 

Washington, DC 20555-0001 

Seabrook Station 

NEx-iera® 

ENERGYe 

SEABROOK 

December 3, 2015 

10 CFR 54 

SBK-L-15202 

Docket No. 50-443 

Response to Requests for Additional Information for the Review of the 

Seabrook Station, License Renewal Application- SET 25 (TAC NO. ME4028) 

Relating to the Alkali-Silica Reaction (ASR) Monitoring Program 

References: 

1. N extEra Energy Seabrook, LLC letter SBK-L-10077, "Seabrook Station Application for 

Renewed Operating License," May 25, 2010. (Accession Number ML101590099) 

2. NRC Letter, Request For Additional Information Related to the Review of the Seabrook 

Station License Renewal Application- Set 25 (TAC NO. ME4028), October 2, 2015 

(Accession Number ML15251A333) 

In Reference 1, NextEra Energy Seabrook, LLC (NextEra) submitted ai;i 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. 

In Reference 2, the staff of the U.S. Nuclear Regulatory Commission identified areas where 

additional information is needed to complete the review of the license renewal application. 

Enclosure 1 contains NextEra responses to the information requested in Reference 2. Enclosure 

1 to this letter contains information proprietary to NextEra Energy Seabrook. This letter is 

supported by an affidavit signed by NextEra Energy Seabrook (Enclosure 6), setting forth the 

basis on which the information may be withheld from public disclosure by the Commission and 

addressing the considerations listed in 10 CFR 2.390(b)(4). Accordingly, it is respectfully 

requested that the information which is proprietary be withheld from public disclosure in 

accordance 10 CFR 2.390. 

NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 I Page 2 

Enclosure 2 provides a non-proprietary version of Enclosure 1. 

Enclosure 3 and Enclosure 4 provide the revised LRA Appendix A - Updated Final Safety 

Analysis Report Supplement Section A.2.1.3 lA for Alkali-Silica Reaction and LRA Appendix B 

Section B.2.1.3 lA for Alkali-Silica Reaction (ASR) Aging Management Program, respectively. 

These two revisions supersede the respective previously submitted sections to the LRA. 

Enclosure 5 provides the revised LRA Appendix A - Updated Final Safety Analysis Report 

Supplement Table A.3, License Renewal Commitment List. This letter contains two revised 

Commitments 83 and 91. 

If there are any questions or additional information is needed, please contact Mr. Edward J. 

Carley, Engineering Supervisor - License Renewal, at (603) 773-7957. 

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 December 2-_, 2015 

s~:t£/J D;); Curtland 

Site Vice President 

N extEra Energy Seabrook, LLC 

Enclosure 1 - Seabrook Station License Renewal Application Requests for Additional 

Information - Set 25, Response to RAis Relating to the Alkali-Silica Reaction 

(ASR) Monitoring Program for the Seabrook Station License Renewal 

Application (Proprietary) 

Enclosure 2 - Seabrook Station License Renewal Application Requests for Additional 

Information - Set 25, Response to RAis Relating to the Alkali-Silica Reaction 

(ASR) Monitoring Program for the Seabrook Station License Renewal 

Application (Non-Proprietary) 

Enclosure 3 - Seabrook Station Updated Final Safety Analysis Report Supplement Section 

A.2.1.31 A for Alkali-Silica Reaction 

Enclosure 4 - Seabrook Station Updated License Renewal Application Section B.2.1.3 lA for 

Alkali-Silica Reaction (ASR) Aging Management Program 

Enclosure 5 - LRA Appendix A- Final Safety Report Supplement Table A.3, License 

Renewal Commitment List Updated to Reflect Changes to Date. 

Enclosure 6 - NextEra Seabrook, Application for Withholding Proprietary Information from 

Public Disclosure and Affidavit 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 I Page 3 

cc: D. H. Dorman 

J. G. Lamb 

P. C. Cataldo 

T. M. Tran 

L. M. James 

Mr. Perry Plummer 

NRC Region I Administrator 

NRC Project Manager 

NRC Senior Resident Inspector 

NRC Project Manager, License Renewal 

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 

Mr. John Giarrusso, Jr., Nuclear Preparedness Manager 

The Commonwealth of Massachusetts 

Emergency Management Agency 

400 Worcester Road 

Framingham, MA 01702-5399 
Enclosure 2 to SBK-L- 15202 

Seabrook Station License Renewal Application 

Requests for Additional Information - Set 25 

Response to RAis Relating to the Alkali-Silica Reaction (ASR) Monitoring Program 

for the 

Seabrook Station License Renewal Application 

(Non-Proprietary) 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 2/ Page 2 

RAI B.2.1.31A-8: Addressing Recent Operating Experience 

Background: 

The regulation in 10 CFR 54.21(a)(3) requires applicants to demonstrate that the effects of 

aging will be adequately managed so that intended functions will be maintained consistent with 

the current licensing basis during the period of extended operation. 

LR-ISG~201 l-05 "Ongoing Review of Operating Experience" recommends that Aging 

Management Programs (AMPs) be informed and enhanced when necessary through the 

systematic and ongoing review of both plant-specific and industry operating experience (OE), as 

discussed therein. The mark-up in LR-ISG-2011-05 for Section A.1.2.3 .10 "Operating 

Experience;' of SRP-LR Appendix A.l "Aging Management Review - Generic (Branch 

Technical Position RLSB-1)" states, in part: · 

Consideration of future plant-specific and industry operating experience relating 

to AMPS should be discussed. The ongoing review of operating experience may 

identify areas where AMPs should be enhanced or new AMPs developed. As 

such, an applicant should ensure that it has adequate processes to monitor and 

evaluate plant-specific and industry operating experience related to aging 

management to ensure that the AMPs are effective in managing aging effects for 

which they are credited. The AMPs are informed by this review of operating 

experience on an ongoing basis, regardless of the AMP's implementation 

schedule. The ongoing review of operating experience information should 

provide objective evidence to support the conclusion that the effects of aging are 

managed adequately so that the structure- and component-intended function(s) 

will be maintained during the period of extended operation. 

LR-ISG~2011-05 also states that "Currently available operating experience applicable to new 

programs should also be discussed. . . . Thus, when developing the elements for new 

programs, an applicant should consider the impact of relevant operating experience from 

implementation of its existing AMPs and from generic industry operating experience." 

As documented in Seabrook Station, Unit No. 1 - Integrated Inspection Report 

05000443/2015002 (Agency wide Documents Access and Management System (ADAMS) 

Accession No. ML15217 A256), Seabrook Station has recently discovered operating experience 

in structures affected by alkali-silica reaction (ASR), as described below, that may be potentially 

attributable to aging effects causing global bulk expansion of concrete. The report indicates that 

the applicant's evaluation of the degraded conditions confirmed that the identified deformation is 

due to bulk expansion from long-term cumulative effects of ASR and strain associated with 

creep. 

- -- --- - ------- -- -
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 2/ Page 3 

During walkdown of plant structures, and from NRC Inspection Reports 05000443/2015002, 

05000443/2014005 (ADAMS Accession No. ML15037A172), 05000443/2014003 

(ADAMS Accession No. ML14212A458), and 05000443/2014009 (ADAMS Accession 

No. ML14349A751), the NRC staff noted the following: 

Relative deformation (differential movement) of the Containment Enclosure Building 

(CEB) indicated by changes in the 3-inch seismic gaps or annulus gap between adjacent 

structures, damaged fire seals, misalignment of conduits/piping at penetrations or between 

adjacent structures, deformed flexible conduit couplings, bent small pipes/conduits and 

supports, etc. 

Discrete wide horizontal cracking, spalling, doorway misalignments etc., in the Residual 

Heat Removal (RHR) and Containment Spray (CS) Vault. 

Cracking, displacements or other indications of structural conditions adverse to quality 

associated with the Fuel Storage Building (FSB). 

The applicant's response to RAI B.2.l.31A-5(a), by letter dated June 30, 2015, includes 

changes to the "parameters monitored or inspected," "monitoring and trending," and 

"acceptance criteria" program elements of the ASR Monitoring Program in License Renewal 

Application (LRA) Section B.2.1.3 lA, with qualitative descriptions for "monitoring building 

deformation" and a related new Commitment No. 91 which states: "In building geometry 

locations where the potential for deformation is likely, enhance the program to monitor for 

building displacement using laser targets and by taking gap measurements." The 

implementation schedule for Commitment No. 91 is described as "within 10 years prior to the 

period of extended operation." 

Issue: 

Regarding aging management of structure and component as they relate to the ASR Monitoring 

Program or the Structure Monitoring Program or both, the wording in the applicant's response to 

RAI B.2.l .31A-5(a) appear to be unclear, relative to (a) monitoring building deformation, (b) the 

corresponding Commitment No. 91, and (c) implementation schedule (the descriptions in the 

response and new Commitment No. 91 appear to address the recently discovered operating 

experience related to the CEB). The staff identified the following concerns: 

[Issue 1] The applicant's response does not include an update to the "Operating Experience" 

program element of the ASR Monitoring Program describing the review and evaluation of the 

cause and impact of any of the relevant recent operating experience, described in the 

"Background" section and other operating experience (if any), to determine how the AMP 

will be affected or new AMPs developed to ensure adequate aging management. The 

"Operating Experience" program element of the revised AMP appears to be incomplete in 

addressing the recent potentially ASR-related operating experience. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 2/ Page 4 

[Issue 2] The AMP does not appear to provide information that "commits" ongoing and 

future review of all relevant plant-specific and industry operating experience related to ASR 

as is recommended in LR-ISG-2011-05 for Section A.l.2.3.10 "Operating Experience" of 

SRP-LR Appendix A.I "Aging Management Review- Generic (Branch Technical Position 

RLSB-1). 

[Issue 3] It is not clear whether global aging effects of ASR such as potentially irreversible 

deformation, relative movement, displacements, or wide discrete cracking that manifests in 

the global direction of least restraint at the structural system level (as opposed to the 

structural component level) are addressed in the large-scale testing program that is being 

used to form the basis for structural functionality through the period of extended operation. 

It is also not clear if the applicant plans to address the global behavior of the structure by 

nonlinear analyses that simulate the kinetics of ASR. 

[Issue 4] Though the applicant stated that it will monitor building deformation aging effect 

by monitoring critical building geometry locations "for displacement via laser targets and gap 

measurements," it is not clear (i) whether the observed indications are irreversible 

deformations (ii) what acceptance criteria the applicant has determined to be appropriate in 

terms of gap measurements to detect damage prior to a loss of intended function, (iii) how a 

structural evaluation would be performed, and (iv) how the results of the evaluation will 

affect the program. 

[Issue 5] It is not clear that the applicant has fully characterized the phenomena observed in 

the recent operating experience and its implications to the current design basis in terms of 

impact to original calculations, etc. Although the applicant performed a temporary "prompt 

operability determination" (POD) that structures have been determined operable under the 

current operating license, the staff has not received any information to support a basis for 

long term functionality of affected structures and structural systems. Information is needed 

with regard to the current and long-term effects of this recent OE, for the staffs review. 

[Issue 6] The revised program appears only to address differential lateral movement between 

structures observed in areas such as the containment enclosure ventilation area slab-to-wall 

interface. It does not appear to address other recently identified operating experience of large 

horizontal cracks in the RHR vault and indications in the FSB. It is not clear whether these 

issues have been evaluated to be related, if there has been an evaluation to 

determine the mechanism and structural implications of large macro cracking in these areas, 

as well as any impact on aging management of those structures. It is also not clear whether 

there has been an evaluation to determine if rebar has yielded in these areas, how during or 

prior to the period of extended operation rebar yielding would be considered and/or evaluated, 

and the implications for service through the period of extended operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 2/Page 5 

[Issue 7] It is not clear if all potential aging effects of the ASR mechanism have been 

identified for monitoring in the AMP, including, but not limited to, those that involve 

irreversible deformation, relative movement, displacement, and discrete cracking. 

[Issue 8] The applicant's response to RAI B.2.l.31A-5(a) asserts that the observed building 

deformation is not a structural capacity concern but does not provide any supporting basis. 

Commitment No. 91 does not appear to consider the basis for service through the period of 

extended operation for the CEB structures, and does not provide information for staff to 

evaluate the adequacy of this statement. Further, the intent of the wording of the 

implementation schedule for the commitment is unclear and, as written, conveys that it will 

be implemented during the 10-year period prior to the period of extended operation. It is not 

clear how the timeliness of this commitment would ensure adequate management of aging 

effects of ASR at the global structure (structural system) level for the period of extended 

operation. 

The staff does not have sufficient information to determine whether aging effects of ASR from 

the recent operating experience have been adequately evaluated and incorporated into the AMP 

to provide adequate aging management during the period of extended operation. 

Request: 

The staff requests information for resolution of the issues in the issue section. 

1) Describe, including technical basis, how the appropriate program elements of the ASR 

Monitoring Program, the Structures Monitoring Program, or other applicable AMP will 

adequately account for all relevant recent plant-specific operating experience described in 

the "Background" section, to address the concerns described in the "Issues" section. The 

information should provide objective evidence to support the conclusion that the aging 

effects of ASR will be managed adequately such that structure- and component-intended 

functions will be maintained during the period of extended operation. The information 

should also include an update to the "operating experience" program element to describe 

the recent operating experience in sufficient technical detail. 

2) If the applicant determines that no modifications or enhancements to the ASR Monitoring 

Program, the Structures Monitoring Program, or other applicable AMP are necessary based 

on the operating experience described as Issue items 1 and 2, explain, with sufficient 

technical detail, the basis for that determination. 

3) Update the LRA program elements and Updated Final Safety Analysis Report (UFSAR) 

supplement, as applicable, based on the response. To facilitate efficient staff review, 

provide the updated ASR Monitoring Program and associated UFSAR supplement, in its 

entirety. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 2/ Page 6 

NextEra Energy Seabrook Response to RAI B.2.1.31A-8 

1. NextEra has prepared a revised version of the ASR AMP (Enclosure 4) to incorporate new 

operating experience identified at Seabrook Station related to building deformation. Specific 

features of the revised AMP that address the concerns described in the "Issues" section ofRAI 

B.2.1.31A-8 include the following: 

91. 

• (Issue 1) Updated Element 10 "Operating Experience" of the ASR AMP to include recent 

observations of expansion leading to building deformation in the Containment Enclosure 

Building (CEB) that have been attributed to ASR. Specifically, the root cause of the event 

was determined to be in-plane expansion of the CEB concrete produced by ASR and ASR. 

expansion in the backfill concrete coincident with a unique building configuration. 

NextEra is in the process ofrevising the structural analysis of the CEB to include the effect 

of ASR expansion with validated deformations from completed walkdowns. 

LRA Appendix A, Section A.3, Commitment #91 has been revised as follows: 

Enhance the ASR Aging Management Program 

to require structural evaluations be performed 

on buildings and components affected by 

deformation as necessary to ensure that the 

structural function is maintained. Evaluations 

of structures will validate structural 

performance against the design basis, and may 

use results from the large-scale test programs, 

as appropriate. Evaluations for structural 

deformation will also consider the impact to 

Alkali-Silica functionality of affected systems and 

Reaction components (e.g., conduit expansion joints). A.2.l.31A Marcil 15, 2020 

Monitoring NextEra will evaluate the specific circumstances 

against the design basis of the affected system 

or component. 

Enhance tile ASR AMP to include additional 

parameters to be monitored based on tile results 

of tile CEB Root Cause, Structural Evaluation 

and walk downs. Additional parameters 

monitored will include: alignment of ducting, 

conduit, and piping; seal integrity; laser target 

measurements; key seismic gap measurements; 

and additional instrumentation. 

• NextEra is currently evaluating observations of expansion resulting in building deformation 

in the Residual Heat Removal and Containment Spray Equipment Vault (i.e., the RHR 

Equipment Vault) and the Fuel Storage Building (FSB). Although ASR is a potential 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 2/ Page 7 

reason for these observations, the engineering evaluation has not been completed. NextEra 

will update the Operating Experience program element of the ASR AMP as necessary, 

pending the results of the evaluations of the RHR Equipment Vault and the FSB. 

• (Issue 2) Added a requirement to program for reviewing plant-specific and industry 

operating experience related to ASR to confirm the effectiveness of the AMP. This 

statement is consistent with the guidance from the Standard Review Plan for License 

Renewal, Appendix A, Section A.1.2.3.10 (Reference 4). 

This operating experience review will include any new information identified from the 

follow-up actions for the CEB resulting from the root cause evaluation and the conclusions 

from the ongoing evaluations of observed expansion relating to building deformation 

affecting the RHR Equipment Vault and the FSB. 

• (Issue 3) Clarified the role of the large-scale testing program for the evaluation of 

expansion relating to building deformation. Specifically, the large scale testing program 

provides insights on expansion behavior that can be used for engineering evaluations. For 

structural evaluations of components affected by deformation, the large scale testing 

program also provides information to supplement the original design code (ACl-318) in 

support of structural evaluations for shear capacity, reinforcement anchorage, and anchor 

capacity. Building deformation could affect the function of structures and equipment in 

many ways (e.g., creating interferences between components, closing or widening seismic 

gaps, degrading fire seals or seismic gap seals, displacing piping, ducting or conduit). If 

such impacts occur, NextEra will address them based on the specific equipment 

configuration in the location of interest. 

RAJ B.2. l .3 lA-8 stated that it is not clear if NextEra plans to address the global behavior 

of the structure by nonlinear analyses that simulate the kinetics of ASR. NextEra will 

perform analyses of ASR-affected structures using analytical tools that are appropriate for 

the specific circumstances, potentially including nonlinear structural analyses. NextEra 

does not plan to perform any analyses regarding the development of ASR in concrete and 

the kinetics of the associated processes (e.g., chemistry modeling). 

• (Issue 4) Clarified that the indications of building deformation may reflect irreversible 

deformation. 

• (Issue 4) Explained the approach for determining acceptance criteria for building 

deformation. Specifically, NextEra will perform walkdowns to identify indications of 

deformation and obtain measurements from areas affected by ASR. Guidance for 

performing these walkdowns states that spaces should be checked for the following 

symptoms: 

o Seismic gap distance different than expected. (All walkdowns include specific 

measurement of seismic gaps for monitoring.) 

o Conduit, duct, or piping exhibiting deformation, displacement or misalignment of 

expansion joints, displacement at support locations, or displaced or misaligned 

supports 

o Gate or door misalignment 

---------
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 2/ Page 8 

o Seismic gap seal or fire seal degradation 

o Any other observations of building deformation. (The list of specific symptoms is 

not exhaustive. The configuration of a particular location may enable monitoring of 

a different parameter than listed above.) 

o NextEra will identify specific limits for measurable symptoms on a locationspecific basis to reflect the particular geometry and configuration in the area of 

interest. The limit will include margin to trigger action prior to loss of intended 

function. 

• (Issue 4) Stated that structural evaluations will be performed on buildings and components 

affected by deformation as necessary to ensure that the structural function is maintained. 

Evaluations of structures will validate structural performance against the design basis, and 

may use results from the large-scale test programs, as appropriate. 

Evaluations for structural deformation will also consider the impact to functionality of 

affected systems and components (e.g., conduit expansion joints). NextEra will evaluate 

the specific circumstances against the design basis of the affected system or component. 

• (Issue 4) Stated that results from structural evaluations will be used to determine whether 

additional corrective actions (e.g., repairs) to the concrete are required. Specific criteria for 

selecting effective corrective actions will be evaluated on a location-specific basis. 

Acceptance criteria may be adjusted based on results of structural evaluations. 

• (Issue 5) Included detection of conditions similar to the symptoms observed in the 

Containment Enclosure Building (CEB) that are potentially attributable to ASR (e.g., 

displacement, relative movement, etc.). NextEra is currently resolving Prompt Operability 

Determinations (PODs) in several locations at Seabrook Station with the ultimate goal of 

reconciling the current structural condition to the original design basis. Analyses of the 

CEB are in progress. Evaluations of the RHR Equipment Vault and the FSB are currently 

characterizing the observed symptoms and how they are influenced by ASR. NextEra will 

update the AMP with any subsequent conclusions from these evaluations or any other 

plant-specific or industry operating experience related to ASR-affected concrete. 

• (Issue 6) The guidance for performing walkdowns includes direction to look for symptoms 

of ASR that have been identified or postulated thus far at Seabrook Station. Such 

symptoms include: 

o Seismic gap distance different than expected (e.g., relative to design configuration). 

o Conduit, duct, or piping exhibiting deformation, displacement or misalignment of 

expansion joints, displacement at support locations, or displaced or misaligned 

supports 

o Gate or door misalignment 

o Seismic gap seal or fire seal degradation 

o Recognizing that the list of symptoms is not exhaustive, the guidance also directs 

that walkdowns look for any other symptoms of building deformation particular to 

the location being evaluated. Guidance for performing walkdowns will be updated 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 2/ Page 9 

to reflect newly identified or postulated symptoms of ASR, as appropriate. Because 

the evaluation of the RHR Equipment Vault and the FSB are ongoing and the 

observed deformation has not yet been conclusively attributed to ASR, the 

walkdown guidance has not been updated to reflect observations in these locations. 

With regard to effects of rebar yielding in ASR-affected structures, refer to the 

response to RAI B.2.l.31A-5(a4). 

• (Issue 7) Stated all known potential aging effects of the ASR mechanism, including effects 

of deformation discussed in the previous bullet. Any additional effects (e.g., identified by 

operating experience) will be added to the AMP and the walkdown guidance as necessary. 

• (Issue 8) Clarified that structural implications of building deformation will be evaluated on 

a case-by-case basis. For expansion within a structural component, NextEra will evaluate 

the structural implications by evaluating the condition against the design basis, which may 

involve use of the large scale test program results. This process is identical to any 

structural component with observed ASR. Additional implications to component or system 

functionality resulting from the observed building deformation (e.g., seal degradation) will 

be evaluated considering the specific geometry of the location of interest. 

• (Issue 8) Updated Commitment 91 to reflect implementation of the ASR AMP no later than 

March 15, 2020. 

2. NextEra has performed a complete revision (Enclosure 4) of the ASR AMP to incorporate 

the most up-to-date information on plans for managing ASR at Seabrook Station. This 

latest revision includes the recent operating experience regarding building deformation 

attributed to ASR. Further revisions to the ASR AMP will be made, as necessary in 

accordance with Element 10, and will include pending resolution of the ongoing technical 

evaluations of the CEB, the RHR Equipment Vault and the FSB. 

3. NextEra has updated the LRA program elements and UFSAR supplement (Enclosure 3) as 

necessary to reflect the discussion above. 

RAI B.2.1.31A-5(al): Justify Representativeness o(Large-scale Test Data to Actual Structure 

Background: 

SRP-LR, Section A.1.2.3.3, states that the "parameters monitored or inspected" program 

element should provide a link between the parameters that will be monitored and how 

monitoring these parameters will ensure adequate aging management. 

In its response to RAI B.2.l.31A-5(a), dated June 30, 2015, the applicant included as Enclosure 

4 the report MPR-4153, Revision 1, "Seabrook Station-Approach for Determining ThroughThickness Expansion from Alkali-Silica Reaction (Proprietary)" as the technical basis for the 

proposed methodology to quantitatively relate the observed ASR effects in existing plant 

structures at Seabrook Station to the results of the large-scale test program at the Ferguson 

Structural Engineering Laboratory (FSEL). Chapter 2 of this document discusses expansion 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 2/ Page 10 

behavior in the test specimens, including formation of a large discrete crack on each specimen 

face between the reinforcement mats. The propagation of this crack dominates the 

through-thickness expansion measurements on the surface, whether measured between the deep 

pins or across the width, and causes bending of the deep pins. Section 2.2.1 of the report 

states that "Once the large crack forms, expansion measured using the embedded rods is 

governed by the increase in crack width. Expansion in the regions outside the embedded rods 

remains relatively unchanged. Therefore, expansion must be calculated based on the total 

width of the beam, rather than the distance between the rods, to appropriately characterize the 

expansion." Section 5.2 of calculation 0326-0062-CLC-03 appended to the report includes an 

expansion measurement correction equation (from Reference 21 of the report). This equation is 

based on average measured through-thickness expansions across the total width using 9 sets 

of measurements across the breadth, intended to correct and appropriately characterize the 

expansion measured between the pins. 

With regard to correlation between strain measurements in large-scale test beams versus 

Seabrook structures, the boundary conditions of the large scale test beams are such that they are 

free to expand at the top, bottom and all edges, causing a large crack to form in the center of the 

beam thickness around the perimeter. Currently it appears that the applicant is using deep pin 

measurements taken at the free edge of the concrete (with a correction) to determine total 

expansion (and using this total expansion as a surrogate for strain). 

The applicant indicated in its revised ASR Monitoring Program "Monitoring and Trending" 

program element that ASR expansion-to-date on Seabrook structures will be determined by 

directly comparing the reduction in elastic modulus (normalized modulus) on the structure to a 

correlating relationship between through-thickness expansion and reduction in elastic modulus 

and through-thickness expansion developed based on measured data on the large-scale 

specimens. 

Issue: 

It is not clear ifthe expansion measured on the test specimens is representative of ASR 

expansion and its potential effects on Seabrook structures to which the test results will be 

applied for reasons below: 

1) Expansion measured on test beams may be overestimated: ASR expansion at Seabrook is 

occurring in walls that are restrained at the four boundary edges, whereas the large-scale 

specimens with free edges have large cracking at the top and bottom that do not extend 

through the depth of the concrete. As indicated in Reference 21 to MPR-4153, this large 

cracking along the middle of the width is an edge effect unique to the design of the test 

specimen that may not be representative of Z-direction expansion of the entire specimen, and 

this effect is not likely to occur in actual structures with different restraints at the boundary. 

Even though the applicant stated that there is correction accounting for the crack by 

normalizing the expansion along the entire thickness of the beam as opposed to just between 

the deep pins, it appears that the measured top and bottom "expansions" across the width are 

predominantly a measure of the surface deformation from propagation of the 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 2/ Page 11 

wide discrete surface crack developed in the top and bottom faces of the test specimen, and 

the measurements at these two surfaces are possibly outliers, and thus not representative 

of ASR expansion on the specimen. It appears that this methodology would likely result in 

significantly larger measurements in the top and bottom (which the applicant has indicated 

are the areas currently being measured) that could cause the expansion to be inflated versus 

the true through-thickness measurements. Therefore, it is not clear that the through- wall 

expansion measured from the freely expanding deep pins in the large-scale specimens is a 

true representation of expansion in the through-wall direction for these specimens. The 

expansion measurements used for correlation and comparison to Seabrook appear not to have 

an appropriate correction for this phenomenon, nor is there information to indicate that 

measurements are being taken at the center of the specimen. Thus, it is not clear whether the 

ASR strain measurements in the large-scale specimens are accurate for direct comparison. 

Further, it appears that cores taken from the test specimens for measurement of elastic 

modulus at the time of expansion measurement do not traverse the large surface crack; 

therefore, its effect is not captured in the measured elastic modulus used for developing the 

correlation with measured expansion. Therefore, it is not clear whether outlier "expansion" 

measurements (i.e., on the surface with pins) in the large scale specimens can be directly 

associated with elastic modulus reduction of cores taken from the center where the large 

crack and free expansion is not occurring. 

The methodology proposed by the applicant is such that this measured expansion value is 

applied directly to elastic modulus reduction factors (i.e., normalized modulus) which will 

be direct surrogates for expansion to-date in Seabrook structures; it is not clear whether this 

method of measuring expansion will accurately represent Seabrook expansion to date. The 

through-thickness expansions measured on the width oftest specimen (even with the 

correction) appears to potentially overestimate the ASR expansion. This measured expansion 

on the specimen may not be conservative in developing the relationship between expansion 

and normalized modulus used to correlate the large scale test data to the actual structure and 

may need to be further corrected. 

2) Boundary conditions of Seabrook structures and components: For the large-scale beam test 

specimens, there is no external restraint and the progression of ASR is likely to be relatively 

more uniform in the specimen because of the significantly aggressive reactive material in the 

concrete mix uniformly distributed through the volume to facilitate accelerated expansion. 

The boundary conditions (constraints at the boundary edges) of the test specimens are 

generally free on all the edges. The test specimens do not appear to represent the boundary 

conditions of the structural components (e.g., walls) and other external restraints on Seabrook 

structures. In addition to internal confinement provided by the reinforcement, the boundary 

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conditions of the monolithically constructed structural system and components and other 

external features, such as concrete structural fill in 

below-grade areas, may also provide significant external restraint that influences the overall 

expansion and distribution of aging effects (e.g., cracking) from ASR in the Seabrook 

structures. Further, in the actual structures, ASR results in differential volumetric expansion 

that is initiated and propagated randomly in a non-uniform maniler. As a result of these 

boundary conditions, the volumetric expansion (i.e., swelling of the gel) will re-orient in the 

direction(s) ofleast restraint, which at the component level may be the out-of-plane 

direction, but likely more focused in the more unrestrained surface layers near the 

reinforcement mat; or at the structural system level in the global direction of least restraint 

(e.g., could be vertical or some other dimension). It is not clear whether through-wall 

measurements are fully representative of volumetric expansion. Since the boundary 

conditions are different between the large-scale specimens and the Seabrook walls, resulting 

in potentially different through-wall expansion behavior, the staff needs additional 

information to determine whether through wall measurements as the single monitoring 

parameter are sufficient and a basis for direct comparison of large-scale test specimens to 

Seabrook structures and structural components that are also subject to external restraint. 

3) The descriptions in LRA Sections A.2.l.31A and B.2.l.31A, as revised by applicant's 

response to RAI B.2.l.31A-5(a), do not appear to include reference to the technical basis 

document(s) (e.g., Report MPR-4153) used by the revised AMP to correlate the large-scale 

test program results to Seabrook structures. 

Request: 

1) Considering the issues above, provide information, with supporting objective evidence, to 

demonstrate that ASR expansions observed and measured (even after proposed correction) on 

the large-scale test specimens at FSEL are representative of ASR effects on actual Seabrook 

structures, and appropriate for correlation of results between test specimens and actual 

Seabrook structures. 

2) Since ASR results in volumetric expansion that can reorient in direction(s) ofleast restraint, 

explain why out-of-plane expansion (and not volumetric strain) is considered the appropriate 

parameter to correlate test results to Seabrook structures. 

3) Update the ASR Monitoring Program elements and UFSAR supplement, as applicable, 

based on the response, and also to incorporate by reference the document(s) (e.g., Report 

MPR-4153) used as the technical basis for methodologies used by the AMP to correlate the 

large-scale test program results to Seabrook structures. 

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NextEra Energy Seabrook Response to RAI B.2.1.31A-5(al) 

1. The large-scale test programs were designed to provide results that represent performance of 

ASR-affected reinforced concrete at Seabrook Station to the maximum extent practical. 

During the course of testing, expansion phenomena were investigated in several different ways 

to confirm that the approach for applying the test results to Seabrook Station is appropriate. 

RAI B.2.l.31A-(al) posed questions about specific aspects of the test program to establish that 

results are appropriately representative of Seabrook Station. The RAI focuses on the 

representativeness of the test specimens with regard to expansion behavior, and determination 

of expansion in the specimens and at the plant. The discussion below addresses these 

questions in several parts: 

• Test Specimen Design Philosophy and Boundary Conditions 

• Sectioning of Test Specimens (to assess observed expansion behavior) 

• Expansion Monitoring Methodology 

• Methodology for Obtaining Cores from Test Specimens 

• Observations from Material Property Testing and In-Plane Expansion Monitoring 

The last subsection in the response to Request # 1 includes conclusions that integrate the 

content of these parts into a summary of the reasons that the large scale test program results are 

representative of the expected performance of ASR-affected reinforced concrete at Seabrook 

Station. 

Test Specimen Design Philosophy and Boundary Conditions 

The interim structural assessment (Reference 5) of ASR-affected structures at Seabrook Station 

considered the various limit states for reinforced concrete and applied available literature data to 

evaluate structural capacity. This evaluation identified that the limited available data for shear 

capacity and reinforcement anchorage for ASR-affected reinforced concrete with twodimensional reinforcement mats were not representative of Seabrook Station. This conclusion 

was driven largely by the facts that 

(1) the literature data for reinforcement anchorage were from a test method that ACI indicates 

is inappropriate and 

(2) the literature data for shear capacity were from test specimens only inches in size. 

Additionally, no data were available on anchor bolt capacity on reinforced concrete with 

two dimensional reinforcement mats like Seabrook Station. 
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Based on the lack of representative literature data, NextEra initiated large-scale test programs to 

evaluate shear capacity, reinforcement anchorage, and anchor bolt capacity of ASR-affected 

reinforced concrete. 

The large-scale test programs were designed to maximize representativeness of reinforced concrete 

structures at Seabrook Station using a reasonable number oftest specimens. The specimens were 

designed to be representative of a reference location (i.e., the B electrical tunnel). To enable application 

of the results to other structures, the test results are compared to provisions in ACI 318-71 to determine 

an appropriate reduction factor to reflect ASR development for the structural capacity calculated using 

code equations. This approach is appropriate, because the test program methodology is consistent with 

the experimental methods used to generate the data on which the code equations from ACI 318 are 

based. 

Seabrook Station includes many reinforced concrete structures and the specific configuration of 

structural components within those structures varies widely. Because the approach for the test 

programs supplements (rather than replaces) the design code, results from appropriately 

representative test specimens may be applied for reinforced concrete structures throughout 

Seabrook Station. Additionally, the test specimen design represents actual structures at 

Seabrook Station to the maximum extent practical. Specific features of the test specimens that 

support this objective include the following: 

• Typical test specimens are long, - in height, and- in width. These 

dimensions are significantly larger than the existing literature data and are therefore a much 

better representation of structural performance in actual structures at Seabrook Station. 

• The specimens included two-dimensional reinforcement mats, one along each longitudinal 

face, with no shear reinforcement. 

• Stirrups, consisting of-rebar, tie together the reinforcement mats on the longitudinal 

edges of each shear and reinforcement anchorage test specimen, which provides 

through-thickness restraint at the ends of the specimen. The primary purpose of the stirrups 

is to ensure that the failure mode of each beam supports the limit state of interest. However, 

these stirrups also mimic the context of an actual structure where the interface with adjacent 

components provides through-thickness restraint at the component edges. The stirrups are 

outside of the structural test region and do not provide through-thickness restraint that 

influences the structural test. 

No additional features are required to represent below-grade exterior walls where distributed load 

on one side of the wall would exist due to the presence of ground water or backfill material. 

Applied load on only one side of the structure does not provide restraint that would inhibit ASR 

development. Therefore, the test specimens adequately represent below-grade exterior 

components. 
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Figure 1 illustrates the dimensions and reinforcement configuration of a typical reinforcement 

anchorage test specimen. Tue design of typical shear specimens and anchor test specimens is 

comparable. 

Figure 1. Diagram of Example Reinforcement Anchorage Test Specimen 

The concrete mix design and storage of the test specimens accelerated development of ASR and 

promoted relatively uniform ASR development in the test specimens. The concrete mix design for 

the test specimens was based on specifications used at Seabrook Station (compressive strength, 

coarse aggregate grading and type, water-to-cement ratio, cement type, aggregate proportions, 

etc.), but also includes features to promote accelerated ASR development including highly 

siliceous aggregate material, high alkali cement, and admixture. In addition, 

the test specimens were stored in an enviroillllental conditioning facility that resulted in exposure 

to elevated temperahrre and humidity. Tue relatively uniform ASR in the test specimens bounds 

the effects of ASR at Seabrook Station, where ASR development is not uniform within a single 

structural component, but is monitored at the location of greatest apparent expansion. 

Sectio11i11g of Test Specimens 

As ASR progressed in the test specimens, a large crack was noted in the center of the surfaces of 

the beam that were between the reinforcement mats. This large crack is not representative of 

expansion behavior of actual structures at Seabrook Station; which have a network of members 

that are either cast together or integrally cast with special joint reinforcing details. In an actual 

strnctme, a vertical wall with two-dimensional reinforcement will be confined in the 
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through-thickness direction at its intersection with neighboring members (i.e., atthe top and 

bottom with floor and ceiling slabs, at the sides with perpendicular walls, and uniformly along the 

wall face by the subgrade for below grade external walls). The confinement provided by the 

network of members in a structure are likely sufficient to preclude a large, through-thickness 

crack. 

To confirm that this large crack was an edge effect that did not compromise the representativeness 

of the test area, the laboratory sectioned the beam cross section (i.e., cut with a saw) to assess the 

depth of the crack. Test specimens sectioned as part of this effort included one shear test specimen 

(post-load testing) and one anchor program test specimen (which has a similar design as the shear 

and reinforcement anchorage specimens). In both cases, the laboratory observed that the large 

crack penetrated only a few inches into the specimen height. 

Figures 2 and 3 are photographs of the large crack from the shear test specimen located between 

the embedded pins used as points of reference for through-thickness expansion measurements. 

Figure 2. Photograph of Large Crack from Surface Between Reinforcement Mats 
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Figure 3. Photograph of Large Crack from Specimen Cross-Section (as Cut with Saw) 

Sectioning of the selected test specimens provided assurance that the large crack was an edge 

effect that did not compromise the representativeness of the test specimens. However, it was not 

clear from these observations whether the large crack at the specimen edges had affected the 

ability to measure expansion in the through-thickness direction using the embedded pins. The 

large crack concentrated expansion between the embedded pins, rather than distributing the 

expansion across the entire specimen width, as would be expected in actual structures at Seabrook 

Station. Damage incurred to the specimens by the sectioning process and the immediate expansion 

after sawing resulting from relaxation of confinement prevented quantitative evaluation of the 

sectioned specimen. Therefore, additional monitoring was implemented to better characterize 

through-thickness expansion. 

Expansion Monitoring Methodology 

As discussed in MPR-4153 (Reference 6) and in RAI B.2.1.31A-5(al), expansion of the reinforced 

concrete beams in the shear and reinforcement anchorage test programs was initially determined 

by measuring the distance between - pins that were embedded in the concrete during specimen 

fabr'ication. A correction equation was used to adjust this measured value for the full width of the 

test specimen. The cumulative expansion at a given point in time is the difference between the 

initial value and the measurement at a given time. (Reference 6: MPR-4153 Section 2.1.2.) 
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To support development of the correction equation, additional through-thickness expansion 

measurements were. performed along the entire height of test specimens using a 

laboratory-fabricated frame. Specifically, the frame fit around the test specimen and enabled 

repeatable measurements of through-thickness (i.e., z-direction) expansion at- points along the 

height of the beam. These measurements provide an independent check of the through-thickness 

expansion observed using the embedded pins. Figure 4 provides a diagram of this frame (i.e., the 

z-frame) and the embedded pins. 

Figure 4. Diagram of Z-Frame Measurement Equipment and Embedded Pins 

The z-frame methodology was implemented in May 2015, at which point• of the test specimens 

had not yet been tested. Table 1 summarizes expansion measurements from these• specimens 

using both the embedded pin and z-frame methodologies. The corrected measurements from the 

embedded pins agree closely with the average z-frame expansion measurements. 
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Figures 5 through 14 provide the expansion profile for each test specimen using the_ 

measurement locations along the z-frame. The red line in each plot corresponds to the measured 

value (uncorrected) from the embedded pins. The- blue dots represent the- z-frame 

measurements and the dashed blue line is average of the z-frame measurements. 

These figures show that the expansion measured near the edge of the beam (i.e., where the large 

crack exists) is consistent with the expansion measured over the entire beam height. Therefore, the 

corrected expansion values obtained from the embedded pins are representative of expansion 

throughout the test specimen. 
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Figure 5. Expansion Profile of Specimen• (as Measured with Z-Frame) 

Figure 6. Expansion Profile of Specimen• (as Measured with Z-Frame) 
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Figure 7. Expansion Profile of Specimen• (as Measured with Z-Frame) 

Figure 8. Expansion Profile of Specimen• (as Measured with Z-Frame) 
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Figure 9. Expansion Profile of Specimen• (as Measured with Z-Frame) 

Figure 10. Expansion Profile of Specimen.(as Measured with Z-Frame) 
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Figure 11. Expansion Profile of Specimen• (as Measured with Z-Frame) 

Figure 12. Expansion Profile of Specimen• (as Measured with Z-Frame) 

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Figure 13. Expansion Profile of Specimen• (as Measured with Z-Frame) 

Figure 14. Expansion Profile of Specimen• (as Measured with Z-Frame) 
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The z-frame data and the observations from sectioning indicate that while total expansion in the 

through-thickness direction is consistent across the profile of the test specimen, the cracking 

behavior is different. These observations suggest that along the specimen edges, expansion is 

concentrated in the large crack; whereas away from the edges, expansion is distributed into finer 

cracks across the specimen cross-section. Figure 15 illustrates this expansion behavior. 

Figure 15. Expansion Behavior of Test Specimens 

Methodology for Obtaining Cores from Test Specimens 

The methodology for determining expansion to date at Seabrook Station (i.e., normalized elastic 

modulus) requires obtaining cores to determine elastic modulus of ASR-affected concrete. The 

correlation in MPR-4153 (Reference 6) is based on elastic modulus data from cores removed from 

the test specimens. The laboratory removed cores from the center portion of the test specimens, 

where expansion is distributed across the beam width in many fine cracks. Therefore, the cores 

would include expanded concrete that would be representative of the bulk specimen and of a 

structural component at Seabrook Station, which would not have free edges. Figure 16 illustrates 

the location of coring from the test specimens. Coring at Seabrook Station will also be in the 

through-thickness direction. 
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Figure 16. Coring Location in Test Specimens 

Observations from Material Property Testing and In-Plane Expansion Monitoring 

In addition to expansion monitoring, the laboratory has been monitoring material properties of the 

test specimens as an independent means for monitoring progression of ASR. Figures 17, 18, and 

19 show the material properties as a function of through-thickness expansion for test specimens 

from the shear test program, reinforcement anchorage test program, and the instrumentation 

program. 
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Figure 17. Normalized Compressive Strength of Test Specimens 

Figure 18. Normalized Elastic Modulus of Test Specimens 
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Figure 19. Normalized Tensile Strength of Test Specimens 

Figures 17 through 19 identify no discernible difference between the test specimens over the 

course of aging, despite the differences in the design of the test specimens: 

• The shear test specimens included- stirrups~ rebar) at each end of the beam, the 

reinforcement anchorage test specimens included stirrups~ rebar) at each end 

of the beam, and the instrumentation beam did not have any stirrups. 

• The shear and reinforcement anchorage test specimens were •••I in height and the 

instrumentation test specimen wa in height. 

• The reinforcement anchorage test specimens and the instrumentation test specimen included 

longitudinal reinforcement at a interval. The shear test specimens included 

longitudinal reinforcement at a-interval. Therefore, the shear test specimens had a 

higher reinforcement ratio in the longitudinal direction. 

The consistent relationship between material properties and expansion for the various beam 

designs suggests that the specific boundary conditions of a particular specimen design do not affect 

the aging mechanism or the measured material properties. Therefore, results can be broadly 

applied to Seabrook Station where similar differences in configuration exist in reinforced concrete 

structures around the site. 

Expansion monitoring information in the in-plane directions also supports the conclusion that the 

test program results are applicable to the range of reinforced concrete structures at Seabrook 
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Station. FSEL monitored expansion in the in-plane directions using both embedded pins and 

combined cracking index. Despite the difference in reinforcement ratio between the shear test 

specimens and the reinforcement anchorage test specimens, the in-plane expansion of all 

specimens has stabilized at approximatel.)11······· 

Conclusions on Representativeness of Test Results 

Based on the discussions above, the through-thickness expansion observed and measured (with 

correction) on the large-scale test specimens is representative of ASR effects on actual structures at 

Seabrook Station and is appropriate for correlation of structural test results. Key conclusions from 

these discussions include the following: 

• The approach for the test programs is to supplement the design code for Seabrook Station, 

ACI 318-71, by determining an appropriate reduction factor to reflect ASR development for 

the structural capacity calculated using code equations. This approach is appropriate, 

because the test program methodology is consistent with the methods used to generate the 

data on which the code equations are based. Because the approach for the test programs 

supplements (rather than replaces) the design code, results from appropriately representative 

test specimens may be applied for reinforced concrete· structures throughout Seabrook 

Station that conform to the design code. 

• The design of the large-scale test specimens is significantly more representative of Seabrook 

Station than the small specimens used for testing in publicly-available literature. The 

specimens were designed to represent a range of structural configurations that enables 

application of test results to all reinforced concrete structures at Seabrook Station. 

• Sectioning of selected beam specimens confirmed that the large cracking on the surface of 

beam specimens between the reinforcement mats is an edge effect that is not expected to 

occur in structures at Seabrook Station. 

• Z-frame measurements demonstrate that expansion measured with embedded pins, as 

corrected for gauge length, is a strong representation of expansion throughout the shear and 

reinforcement anchorage test specimens. 

• Expansion away from the edges of the specimen is characterized by fine cracking across the 

beam width, so cores obtained from this region are sufficiently representative of the bulk 

condition of the test specimen. 

• Test data show that differences in specimen geometry and reinforcement do not affect aging 

of the concrete by ASR. This conclusion supports general applicability of the test results to 

structures at Seabrook Station. 

2. For ASR-affected surfaces at Seabrook Station, NextEra will monitor volumetric effects of 

ASR expansion by obtaining measurements in both the in-plane and through-thiC?kness 

directions. Specifically, the ASR AMP specifies monitoring combined cracking index (CCI) 

for in-plane expansion and extensometer measurements for through-thickness expansion. Data 

analysis from the large-scale test programs is ongoing and monitoring thresholds for the ASR 
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AMP will be developed based on the test reports. However, the overall methodology for using 

in-plane and through-thickness expansion values for various aspects of the ASR AMP is 

summarized as follows. 

• Initial screening for ASR will be performed using CCI only. CCI values exceeding a defined 

low-level threshold will trigger additional actions. CCI is a relatively simple, non-destructive 

method for monitoring cracking that appropriately characterizes expansion until expansion 

reorients in the direction ofleast restraint (i.e., the through-thickness direction at Seabrook 

Station). 

The test programs indicated that direction of expansion is not significantly affected by the 

reinforcement when expansion is at or belo~ Beyond this expansion level, the twodimensional reinforcement mats provide confinement in the in-plane directions, and throughthickness expansion dominates. Figure 20 describes the observed expansion of a 

representative specimen. 

Figure 20. Expansion of Specimen A51 

1 Figure 20 is for illustrative purposes only. All measurements described in Figure 20 were performed using an 

approved test procedure. However, periodic monitoring of expansion is considered for information only, whereas the 

measurements at the time of testing are formal test measurements. 
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• For anchor performance, the large scale test programs show that ASR does not have an effect 

until in-plane expansion reaches a sufficiently high level. Therefore, the ASR AMP specifies 

that ifthe CCI exceeds a specified threshold, additional evaluation must be performed to 

justify continued acceptability of the structure. 

This approach is based on the fact that anchor performance is sensitive to in-plane expansion, . 

but not through-thickness expansion. In-plane expansion creates microcracks parallel to the 

axis of an anchor, mainly in the concrete cover. These microcracks perpendicular to the 

concrete surface have the potential to provide a preferential failure path within a potential 

breakout cone, leading to degraded anchor performance. 

Through-thickness expansion has the potential to create microcracks perpendicular to the 

axis of an anchor. These potential microcracks that open parallel to the concrete surface do 

not provide a preferential failure path to result in degraded anchor performance. An anchor 

loaded in tension would compress the through-thickness expansion and close any potential 

microcracks within the area of influence of that anchor. Without a 'short-circuit' of the 

breakout cone, through-thickness expansion is a non-factor in anchor performance. 

• For shear and reinforcement anchorage performance, the large scale test programs show that 

ASR does not have an effect until at least the maximum through-thickness expansion 

observed in the test programs. Therefore the ASR AMP specifies that if through-thickness 

expansion exceeds a specified threshold, additional evaluation must be performed to justify 

continued acceptability of the structure. NextEra will determine through-thickness expansion 

from the sum of the calculated pre-instrument expansion (using the methodology ofMPR4153) and subsequent measurements with an extensometer. 

This approach is based on the observed expansion behavior of the test specimens, in which 

expansion in the through-thickness direction was dominant. (See Figure 20.) It is not 

necessary to also incorporate in-plane expansion, which represents a very small proportion of 

the volumetric strain. 

3. NextEra has performed a complete revision of the ASR AMP (Enclosure 4) to incorporate the 

most up-to-date information on plans for managing ASR at Seabrook Station. This latest 

revision includes the currently available information from the large scale test programs and 

plans for using the final data from the test programs, as discussed above. The ASR AMP and 

the UFSAR supplement have also been updated to reflect the discussion above and the 

content ofMPR-4153. After the test program reports have been issued, the ASR AMP will be 

updated to reference the test program reports for the specific thresholds for expansion that 

trigger further evaluation in accordance with Element 10 "Operating Experience". 
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RAI B.2.1.31A-5(a2): Uncertainties and Variations Associated with Estimate o(Normalized 

Modulus 

Background: 

In its response to RAI B.2.l.31A-5(a), dated June 30, 2015, the applicant included as Enclosure 

4 the report MPR-4153, Revision 1, "Seabrook Station-Approach for Determining ThroughThickness Expansion from Alkali-Silica Reaction (Proprietary)" as the technical basis for the 

proposed methodology to quantitatively relate the extent alkali-silica reaction (ASR) in existing 

plant structures at Seabrook Station to the results of the large-scale test program at the FSEL. 

Figure 3-3 and Equation 1 of the report show the development of the correlation 

between normalized concrete modulus and expansion data from the test specimens. Based on 

Sections 3.1.1and3.2.l ofreport MPR-4153, the normalized modulus value used in Figure 3-3 

and Equation I of the report for developing the relationship between normalized modulus and 

through-thickness expansion, is the ratio (Et/ EO) of the elastic modulus by testing cores 

extracted from the specimen at the time of expansion measurement (Et) to the 28-day elastic 

modulus (EO) obtained by testing cylinders (8 inches in height and 4 inches in diameter), that 

were molded at the time of specimen fabrication, at age 28 days. 

Section 3.3 of report MPR-4153 describes two approaches for obtaining elastic modulus at 28 

days in Seabrook structures for the purpose of normalization and determining the expansion to 

date. "Approach l" uses 28-day elastic modulus calculated using the American Concrete 

Institute (ACI) 318 equation based on original concrete cylinder strength (f' c) measured at 28 

days. For "Approach 2," MPR-4153 states that it plans to test the elastic modulus of the cores 

obtained during installation of extensometers at "control locations where ASR has not affected 

the structure." MPR-4153 states that the applicant "should evaluate selection of a 

representative reference core on a case-by-case basis. 

In both approaches, there are variations introduced in the estimate 28-day elastic modulus from 

the use of the ACI 318 equation based on 28-day cylinder strength ("Approach l") or measured 

elastic modulus from cores because of change (i.e., increase) in concrete strength with age 

("Approach 2"). In both cases, the normalized modulus at the time of installation of 

extensometers is determined as the ratio of the measured modulus on cores extracted at the 

installation location to the 28-day modulus estimate from Approach 1, or Approach 2, or both. 

Section 3.3.1 ofMPR-4153, states: 

NextEra has retrieved records for concrete fabrication from original construction for 

selected buildings ... For structural assessment of particular concrete members; 

application of values from [MPR Calculation 0326-0062-CLC-02] will need to be 

evaluated on a case-by-case basis to determine whether the available data are 

sufficiently representative of the concrete being evaluated. NextEra may need to 

retrieve additional original construction records to implement this approach. 

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Section 4.2 of MPR-4153 discusses an uncertainty value related to the approach for 

determining original elastic modulus but does not give any information regarding the 

methodology for determining this uncertainty value. 

Issue: 

Report MPR-4153 does not provide the size, orientation, and locations of cores extracted 

from test specimens to be used in conducting tests for concrete material properties at the 

time of expansion measurement and load test. 

It is not clear if using the concrete material properties (i.e., modulus); from cylinder test, data 

at 28-days and core test data from the large scale beam specimens at the time of expansion 

measurement is appropriate for normalization of elastic modulus to determine ASR 

degradation. It is not clear whether the error associated with variations such as type of 

specimen (cylinder vs core) and size (diameter, d; height, h; and hid ratio) oftest specimens 

used may influence the values obtained for the normalized modulus were considered in the 

calculation of uncertainty or in the development of the correlating factors between expansion 

and normalized modulus. 

It is not clear what value off c (mean, median, cir probability of exceedance consistent with 

the statistical basis in the ACI 318 code) will be used in the ACI 318 equation for arriving at 

the 28-day elastic modulus based on 28-day test data of cylinder strength. 

With regard to the statement from Section 3.3.1 of MPR-4153 referenced in the 

"Background" section above, it is not clear how the applicant will assess the data and the 

criteria to be used to determine whether the available data are "sufficiently representative." 

It is not clear whether original data is available and if not, how the inputs to the modulus 

calculation will be determined. 

Uncertainties and variations associated with the recommended correlating approach based 

on estimate of normalized elastic modulus on the test specimens and on Seabrook structures 

are not addressed and it is not clear whether they need to be addressed statistically in a 

bounding manner. 

Request: 

The staff requests information for resolution of the issues in the issue section. 

1) Provide additional information regarding the methodology used to determine the uncertainty 

value associated with developing the empirical relationship between concrete material 

property and measured expansion. Include the basis for determination of the uncertainty 
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value used for the normalized correlation between expansion and reduction in elastic 

modulus. 

2) Discuss whether and how the methodology considered factors such as size variation, 

orientation and locations of cores from the test specimens used in measuring concrete 

material properties, including elastic modulus. In addition, discuss whether and how the 

methodology considered uncertainty or errors associated with comparison between concrete 

cylinders 28-day test data and extracted cores. If these factors were not considered, 

provide technical justification that such consideration is not needed. 

3) Clarify, with the basis, what value off c (mean, median, or probability of exceedance 

consistent with the statistical basis in the ACI 318 code) will be used in the ACI 318 equation 

in "Approach l" for arriving at the 28-day elastic modulus based on 28-day test data of 

cylinder strength of Seabrook structures. 

4) With regard to the statement from Section 3.3.1 of report MPR-4153 referenced in the 

"Background" section above, clarify how the applicant will assess the data and the criteria to 

be used to determine whether the available data are "sufficiently representative." Also 

clarify whether "original data" referenced in that statement are available and if not, how the 

inputs to the modulus calculation will be determined. 

NextEra Energy Seabrook Response to RAI B.2.1.31A-5(a2) 

Summary of MPR-4153 Methodology 

The correlation in MPR-4153 relates normalized elastic modulus to through-thickness expansion. 

Normalized elastic modulus is the ratio of the elastic modulus at the time of interest (i.e., from 

material property testing of the removed cores) to the original elastic modulus. Elastic modulus 

was the material property selected for the correlation (rather than compressive strength or splitting 

tensile strength), because elastic modulus was the most sensitive to expansion and the most 

repeatable. 

MPR-4153 recommends two different methods for determining the original elastic modulus, as 

summarized below: 

• Approach 1 - Use original construction records to determine the 28-day compressive strength 

of the concrete. Apply the ACI 318 correlation relating compressive strength to elastic 

modulus to calculate the 28-day elastic modulus, which is used as the original elastic 

modulus. 

• Approach 2 - Identify a location from the same concrete placement as the concrete surface of 

interest that does not appear to be affected by ASR. Obtain cores from this location and 
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perform material property testing to determine elastic modulus of non-ASR-affected 

concrete, This non-ASR core is used as the original elastic modulus. 

NextEra will select an appropriate approach to determine the original elastic modulus based on the 

specific circumstances for each location of interest. NextEra would prefer to use Approach 1, 

because it requires fewer cores. However, this approach may not be practical in all cases, 

depending on the effort required to retrieve the compressive strength test records for a specific 

concrete placement. Approach 2 provides an alternate methodology. 

Pending Update to MPR-4153 

At the time MPR-4153 was published, only a portion of the test data were available, as testing had 

not yet been performed on• of the• test specimens. Data from all of the test specimens and the 

final report from the laboratory will be transmitted by test reports and commercially dedicated by 

the end of2015. After all data are available (i.e., in early 2016), MPR-4153 will be updated and a 

treatment for uncertainty will be included. (The data not included in the previously-transmitted 

version ofMPR-4153 were for expansion levels markedly higher than our estimate of the 

expansion levels currently exhibited at Seabrook Station. Therefore, the correlation is not 

expected to significantly change in the region of the curve that reflects the current conditions at the 

plant.) 

NextEra plans to install extensometers at Seabrook Station to monitor expansion in selected 

locations affected by ASR. Installation of extensometers requires core boring into the concrete 

surface in question. The methodology specified in MPR-4153 relies on elastic modulus data from 

concrete cores of an ASR-affected surface to determine the expansion that had occurred. At 

Seabrook Station, the cores removed to enable extensometer installation will also be tested to 

determine the elastic modulus of the concrete. 

1 & 2. The methodology described in MPR-4153 considered the factors identified by RAI 

B.2.l.31A-5(a2) for contributions to uncertainty. As discussed above, Revision 1 ofMPR4153 did not include a quantitative evaluation of uncertainty. A qualitative discussion of 

each contributor to uncertainty identified by RAI B.2.l.31A-5(a2) is provided as follows: 

• Size Variation - Cores obtained from large scale test specimens were approximately 

4.0 inches in diameter. At Seabrook Station, all new cores obtained for the ASR program 

will be obtained with a bore that will yield a core with a diameter of approximately 

4.0 inches. 

Cores obtained from the large scale test specimens were trimmed to a length of 

approximately 8.0 inches to satisfy the target length-to-diameter ratio of 2.0 (which is the 

median of the 1.9 to 2.1 range specified by ASTM C42). Cores obtained from Seabrook 

Station will also be trimmed to approximately 8.0 inches. 

Because the approximate dimensions of the cores obtained from the large scale test 

specimens and the cores from Seabrook Station will be identical, uncertainty associated with 
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size variation will be negligible. Further, material tests of specimens with these dimensions 

do not require corrections for reduced size specimens. 

• Orientation - For the large scale test program, the laboratory obtained most cores by drilling 

into the face of the test specimen representative of the interior wall at Seabrook Station (i.e., 

perpendicular to the face with two inches of cover above the reinforcement mat). Most cores 

obtained from Seabrook Station will also be from interior wall surfaces, including all 

surfaces monitored below grade. NextEra will install selected extensometers above grade on 

exterior concrete surfaces, which have three inches of cover above the reinforcement mat. 

This small difference in cover will not have a significant effect on confinement of the 

concrete in the core. 

The orientation of cores obtained from the large scale test specimens and cores obtained from 

Seabrook Station will be the same (i.e., perpendicular to the embedded reinforcement mats). 

Therefore, there is no uncertainty related to orientation of the core relative to the dominant 

direction of expansion. 

• Locations of Cores - For the large-scale test program, the laboratory obtained cores from 

locations that were in the central portion of the specimens, through the openings in the 

reinforcement mats. As discussed in the response to RAI B.2.l.31A-5(al), these locations 

are not subject to edge effects. Additionally, laboratory procedures required strict controls 

on specimen treatment (e.g., exposure in the environmental conditioning facility), so 

exposure conditions were consistent across the entire specimen. Therefore, variability in 

ASR development within a test specimen is expected to be low. 

At Seabrook Station, ASR-related expansion is not typically consistent across a single 

concrete member. The locations for extensometer placement (and therefore coring) will be 

the areas that have the greatest symptoms of ASR-related expansion. This approach will 

conservatively characterize the elastic modulus of the concrete member in question. Because 

this approach is inherently conservative, quantitative treatment of uncertainty for core 

location may not be necessary. 

• Cylinders vs. Cores - The data used for preparing the correlation in MPR-4153, Approach 1, 

and Approach 2 all determine elastic modulus at the time of interest (i.e., the numerator) by 

testing removed cores. Differences in determination of original elastic modulus (i.e., the 

denominator) may contribute to uncertainty of the methodology described in MPR-4153. 

For the data used to prepare the correlation in MPR-4153, the original elastic modulus 

is the average elastic modulus test result from cylinders tested 28 days after specimen 

fabrication. 

For Approach 1 at Seabrook Station, the original elastic modulus would be calculated 

from the average compressive strength test result of cylinders tested 28 days after the 

concrete was placed. 

For Approach 2 at Seabrook Station, the original elastic modulus would be the 

average elastic modulus test result of "reference" cores obtained at the time of 

extensometer installation. The reference cores will be obtained from a nearby area 

from the same concrete placement that does not exhibit signs of ASR. 
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For Approach 2, the coring process can cause internal damage to the concrete in the core and 

impact measured material properties. The compressive strength of a core is expected to be 

lower than the compressive strength of a cylinder from the concrete it represents because of 

damage incurred during the coring process. Elastic modulus of a core may also be biased 

low for the same reason. However, for Approach 2, the concrete will have had decades of 

additional time to cure, which affects material properties, albeit at a slow rate. This factor 

increases the compressive strength relative to a 28-day cylinder and may also bias elastic 

modulus high for the same reason. These factors have opposite effects on the normalized 

material property and are potentially offsetting. The upcoming revision ofMPR-4153 will 

include an evaluation of the combined effect of these factors. 

While Approach I does not require coring to determine original elastic modulus, NextEra 

notes that the use of the ACI correlation in Approach 1 for calculating elastic modulus from 

compressive strength introduces uncertainty. ACI 318 states that this correlation provides a 

calculated elastic modulus to within ±20%. Calculation 0326-0062-CLC-01 (Appendix B of 

MPR-4153) demonstrated that this range was consistent with the data from the large scale 

test programs, which supports application of the ACI correlation to concrete at Seabrook 

Station. 

Revision 2 ofMPR-4153 will include expanded discussions of the above issues. 

MPR-4153 Section 4.2 acknowledges the potential influence of uncertainty with the correlation 

between normalized modulus and expansion, focusing on the sensitivity of expansion to the 

measured modulus. This section notes that the sensitivity of the correlation is higher when 

normalized elastic modulus is less than., so uncertainty would have a greater impact in that 

region of the relationship. 

3. As noted in RAI B.2.l.31A-5(a2), the uncertainty associated with Approach I from MPR4153 may be different than the uncertainty associated with Approach 2. The factors 

associated with each approach will be considered in the application of the methodology at 

Seabrook Station to ensure that the estimate of expansion to date is conservative. 

Treatment of the compressive strength data as an input for the Approach 1 methodology for 

determining original elastic modulus will depend on the available information for the 

location in question. 

• IfNextEra traces the 28-day cylinder compressive strength results from an original 

construction record to a particular concrete member, the average 28-day compressive 

strength value of those cylinders will be used. It is appropriate to use the mean as the 

objective is to estimate expansion to date; it is not used directly in a structural evaluation. 

• If a broader data set is applied (e.g., the statistical results for compressive strength for all 

concrete placements of a given mix design), then an appropriately conservative approach 

will be selected and justified (e.g., applying a compressive strength value higher than the 

average by a certain factor of the standard deviation). 
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4. Approach 1 requires that the input data for compressive strength of28-day cylinders be 

representative of the concrete being evaluated. 

• IfNextEra establishes traceability from the original construction record to the specific 

concrete member being evaluated, then the data are sufficiently representative to enable use 

of the average compressive strength. 

• IfNextEra uses a broader data sample that is not directly linked to the specific concrete 

member being evaluated (e.g., compressive strength test results from other locations in the 

same building, or the statistical results for compressive strength for all concrete pours of a · 

given mix design), then an evaluation for representativeness will be performed to ensure that 

the calculation is appropriately conservative. The specific criteria associated with this 

evaluation will vary depending on the data set and its intended application. 

Ultimately, if NextEra concludes that the available data are not sufficiently representative, 

Approach 2 provides an alternative method to obtain the information on original elastic modulus. 

NextEra has researched original construction records throughout its efforts to evaluate 

ASR-affected concrete at Seabrook Station, and has demonstrated that these records are available 

and retrievable. However, the experience to date indicates that the effort to identify the concrete 

batch for a specific location (i.e., building, wall, and elevation) and retrieve the specific test record 

may be impractical for all locations to be evaluated. 

RAI B.2.1.31A-5(a3): Representative Sample (or Monitoring Through-wall Expansion 

Background: 

The applicant's response to RAI B.2.l.31A-5(a), dated June 30, 2015, states that the ASR 

Monitoring Program will be enhanced (revised Commitment No. 83) to install borehole 

extensometers in at least 34 representative locations that the applicant has stated are 

representative of the ASR on site such that the program will monitor, trend, and assess ASR 

expansion in the out-of-plane (through-wall) direction. 

Section A. l .2.3 .4 "Detection of Aging Effects" of Appendix A. l "Aging Management Review 

-Generic (Branch Technical Position RLSB-1) of SRP-LR states, in part: 

For a condition monitoring program, when sampling is used to represent a larger 

population of SCs, applicants should provide the basis for the inspection 

population and sample size. The inspection population should be based on such 

aspects of the SCs as a similarity of materials of construction, fabrication, 

procurement, design, installation, operating environment, or aging effects. The 

sample size should be based on such aspects of the SCs as the specific aging 

effect, location, existing technical information, system and structure design, 

materials of construction, service environment, or previous failure history. The 
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samples should be biased t9ward locations most susceptible to the specific aging 

effect of concern in the period of extended operation. Provisions on expanding the 

sample size when degradation is detected in the initial sample should also be 

included. 

Section A.l.2.3.4 "Detection of Aging Effects" of the SRP-LR also states that this program 

element describes "when," "where," and "how" program data are collected (i.e., all aspects of 

activities to collect data as part of the program), including how frequently evaluated. The section 

further states that the discussion should provide information that links the parameters to be 

monitored or inspected to the aging effects being detected and managed prior to loss of function. 

Issue: 

1) The response to RAJ B.2.l.31A-5(a), dated June 30, 2015, does not include an update to the 

"detection of aging effects" program element of the ASR Monitoring Program with the 

information recommended in Section A.1.2.3.4 of SRP-LR (as described in the Background 

section) with regard to the representative sample of locations and aspects of activities to 

collect and evaluate data in the program enhancement to monitor out-of-plane ASR 

expansion. 

2) In the breakdown of number of instruments to be installed provided in the applicant's 

response to RAJ B.2.l.31A-5(a), Issue No. 3, almost 50 percent of the extensometers are to 

be installed in "Tier l" ASR locations and over 50 percent will be installed in "Ambient 

Weather Conditions." The staff needs additional information to determine whether the 

chosen locations and sample will be sufficient to manage through-wall cracking. 

3) ASR damage is likely to progress more quickly in areas of high humidity and temperature. It 

is not clear, from the tables given in the RAJ response, whether the most severe areas are 

well-represented. It is also not clear whether the areas that exhibit the highest combined 

crack index (CCI) are the most severely affected areas, and the criteria that will be used to 

determine if an increase in sample size is required. 

4) The response to RAJ B.2.l.31A-5(a) does not appear to address the durability and long-term 

reliability of the borehole extensometer proposed to be used to monitor out-of-plane 

expansion, and how the measurements would continue if an extensometer became nonfunctional in service. The response also does not discuss the data acquisition system that will 

be used to gather and process data from the proposed extensometers to ensure adequate aging 

management. 
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Request: 

The staff requests information for resolution of the issues in the Issue section with regard to the 

representative sample for monitoring through-wall expansion. Also, the staff requests the 

applicant to update the "detection of aging effects" program element to the ASR Monitoring 

Program and UFSAR supplement, as applicable and appropriate (based on the response). 

1) Provide an update to the "detection of aging effects" program element of the ASR 

Monitoring Program with information recommended in Section A.1.2.3.4 of SRP-LR (as 

stated in the background section) to describe the basis for (a) the inspection population and · 

sample size and (b) all aspects of activities to collect data, including how frequently the data 

are sampled and evaluated, regarding representative sample of locations in the program 

enhancement to monitor out-of-plane ASR expansion. 

2) Provide additional information to demonstrate the adequacy of sampling locations for 

through-wall cracking management (i.e., determination of whether the chosen locations and 

sample will be sufficient to manage through-wall cracking). 

3) Provide additional information to demonstrate the bounding conditions of the aging 

management measures relative to (a) sampling most severe locations (correlation with CCI) 

and (b) triggering point for increasing sample size. 

4) Provide additional information to demonstrate the reliability of the aging management 

measures relative to durability of supporting instrumentation (e.g., borehole extensometer 

and data acquisition system). 

NextEra Energy Response to RAI B.2.l.31A-5(a3) 

1. NextEra has revised the approach for monitoring through-thickness expansion to install 

extensometers in all Tier 3 locations. These values have CCI of greater than 1 mm/m, -

The revised ASR AMP incorporates this change. 
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SBK-L-15202 /Enclosure 2/ Page 41 

83. 

LRA Appendix A, Section A.3, Commitment #83 has been revised as follows: 

Enllance tile ASR AMP to install extensometers 

in Tier 3 areas of structures to monitor 

expansion due to alkali-silica reaction in· tile 

Alkali-Silica out-of-plane direction. 

Reaction A.2.l.31A December 31, 2016. 

Monitoring Monitoring expansion in the out-of-plane 

direction will commence upon installation of the 

extensometers in 2016 and continue through the 

period of extended operation. 

As previously established for Tier 3 monitoring; NextEra plans to perform semiannual 

monitoring of extensometers to provide a baseline for variability in extensometer readings. 

Specific thresholds for triggering additional evaluation will be identified following evaluation 

of the large scale test program results. A description of the monitoring approach and its basis 

are included in the response to RAJ B.2.l.31A-5(al). 

2. As discussed above, the ASR AMP (Enclosure 4) has been updated to specify installation of 

extensometers at all Tier 3 locations. NextEra currently plans to install extensometers in all 

Tier 3 locations, which will ensure that the most limiting locations for ASR development are 

being monitored. NextEra also plans to install extensometers in selected locations where CCI 

is less than 1.0 mm/m. 

3. As discussed above, the ASR AMP (Enclosure 4) has been updated to specify installation of 

extensometers at all Tier 3 locations. This change ensures that the locations with greatest ASR 

development are being monitored. A triggering point for increasing the sample size for 

extensometers in Tier 3 locations is not necessary, because all Tier 3 locations will be 

monitored. 

4. NextEra will install Snap-Ring Borehole Extensometers (SRBEs) at Seabrook Station to 

monitor through-thickness expansion. The Instrument Evaluation Program evaluated 

performance of the SRBEs in a test specimen representative of the concrete at Seabrook 

Station over a one-year period. The SRBE provided accurate measurements ofthroughthickness expansion throughout the test program and did not exhibit any problems related to 

reliability. The test program involved specimen conditioning through temperature cycles and 

extended exposure to high humidity. Successful operation in these conditions demonstrates 

that the SRBEs will perform satisfactorily in the conditions expected at Seabrook Station. 

The SRBE consists of a graphite rod that is held in place by an anchor placed in the borehole. 

Measurements are performed by using a depth micrometer to measure the distance from a 

reference anchor at the surface of the concrete to the end of the rod. There is no data 

acquisition system used to obtain measurements. The SRBE design contains no electronics 
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and does not require calibration. Therefore, failure of the SRBE is unlikely. In the event that 

an SRBE did fail (e.g., an anchor broke loose), NextEra could install another SRBE nearby to 

the failed location and continue expansion monitoring. 

RAI B.2.1.31A-5(a4): CCI as a surrogate for ASR expansion in the in-plane direction 

Background: 

The applicant's response to RAI B.2.l.31A-5(a), dated June 30, 2015, for Issue 1 states that 

"CCI [Combined Crack Index] is used as a surrogate for accumulated strain from ASR 

expansion in the in-plane directions." 

In the changes to LRA Section B.2.1.3 lA by letter dated September 13, 2013, under "Program 

Description," the applicant provided the basis for use of the crack index methodology on page 8 

of 18 of Enclosure I. The applicant states "The total strain in the concrete can be approximated 

as the sum of the strain at crack initiation plus the crack index (c = Ecr +CI) ... " The applicant 

provided an illustration in Figure A-1 of the expansion in ASR-affected concrete and goes on to 

conclude that "the Cracking Index (CI) provides a reasonable approximation of the total strain 

applied to the concrete after crack initiation." SRP-LR, Section A.1.2.3.3, states that the 

"parameters monitored or inspected" program element should provide a link between the 

parameters that will be monitored and how monitoring these parameters will ensure adequate 

aging management. 

Issue: 

While continued surface crack indexing may be a necessary component for a complete 

measure of ASR expansion, which is volumetric in nature, there is a lack of clarity in the ASR 

Monitoring Program with regard to the following. 

It is not clear whether the applicant's statement referenced in the "Background" section is 

indicating that the large-scale test program directly measures rebar strain and correlates to 

CCL For its review, the staff needs additional information in support of the statement that 

CCI is a "surrogate for accumulated strain." 

In order to maintain strain compatibility between concrete and rebar in ASR-affected 

concrete, the strain in the rebar will be equal to the strain applied to the concrete. 

Therefore, a CCI of 1 mm/m (which is the acceptance criteria used in the program to 

distinguish between Tier 2 and Tier 3 criteria for initiation of a structural evaluation 

corresponds to 1 millistrain of expansion resulting in rebar stress of 29 ksi. The staff is 

concerned that this may represent a significant magnitude of rebar stress due to ASR 

expansion that has not been or will not be accounted for during the period of extended 
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operation. The program does not address rebar strain limits as compared to ASR 

progression through the period of extended operation to ensure the concrete maintains its 

design functionality. 

The current AMP uses the Tier 1-2-3 approach to determine monitoring :frequency and it 

relies upon CCI in two directions. It is not clear if and how CCI will be used to 

characterize ASR progression during the period of extended operation. Also, there is lack 

of clarity with regard to the correlation of CCI with anchor bolt capacity. 

It is not clear whether in-plane cracking measurements will be used, qualitatively or 

quantitatively, in conjunction with through-wall expansion measurement data, to characterize 

ASR severity or progression. 

Request: 

The staff requests information for resolution of the issues in the Issue section. 

1) Clarify the physical significance of CCI as a surrogate for strain in rebar due to ASR as 

explained in the applicant's revision to LRA Section B.2.l.31A by letter dated 

September 13, 2013, and whether and how the corresponding rebar stress is accounted for. 

2) Clarify whether the large-scale test program directly measures rebar strain and correlates to 

CCI. Provide the technical basis as to how CCI correlates to stress in the rebar and how 

rebar stress is quantified. 

3) Describe the role of CCI in the large scale test program with regard to correlating the test 

results to Seabrook structures and components, including evaluating impact on anchor 

capacity. 

4) Clarify, with supporting basis, if and how the program will account for information 

gained from the elastic modulus testing in terms of comparison of ASR severity for 

monitoring purposes. 

5) Clarify whether in-plane cracking measurements will be used, qualitatively or quantitatively, 

in conjunction with through-wall expansion measurement data, to characterize ASR severity 

or progression. 

6) Update program elements and UFSAR supplement, as applicable, based on the responses. 
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NextEra Energy Seabrook Response to RAI B.2.1.31A-5(a4) 

For ASR-affected surfaces at Seabrook Station, NextEra will monitor volumetric effects of ASR 

expansion by obtaining measurements in both the in-plane and through-thickness directions. 

Specifically, the ASR AMP specifies monitoring combined cracking index (CCI) for in-plane 

expansion and extensometer measurements for through-thickness expansion. 

1. Concrete has minimal capacity to expand before cracking. Therefore, measurement of 

concrete expansion can be well approximated by crack width summation (e.g., CCI). While 

true engineering strain is represented by the sum of material elongation and crack widths, the 

crack width term rapidly dominates the overall expansion. 

As shown in response to RAJ B.2.l.31A-5(al); Figure 20, CCI values agree closely with the 

observed expansion from embedded pins. The expansion values measured using embedded 

pins are a better measure of true engineering strain because these measurements reflect both 

material elongation and crack width. However, because of the close agreement with CCI, 

results from the large-scale test program for expansion monitoring support use of CCI as an 

approximation for in-plane expansion. 

The bond between the concrete and the reinforcement prevents relative displacement (i.e., 

slip), so expansion of a reinforced concrete member will have an equal effect on the concrete 

and the reinforcement. Concrete expansion and rebar expansion are therefore identical, and 

CCI is also a reasonable approximation for rebar strain. 

In reinforced concrete, expansion of concrete due to ASR is restrained by the steel 

reinforcement. As ASR develops, the result is a tensile stress in the reinforcement and a 

compressive stress in concrete, which is necessary for structural equilibrium. External tensile 

loading will not result in increased tensile stress in rebar until the concrete compressive stress 

is overcome. This effect is known as pre-stressing. 

The structural tests inherently account for rebar strain, which has occurred in the ASR-affected 

test specimens. However, structural test results from the large-scale test programs indicate no 

loss of shear capacity or reinforcement anchorage due to ASR at expansion levels observed in 

the test programs. This observation demonstrates the effect of ASR-induced pre-stressing, 

which results in no loss of structural capacity even though rebar strain has occurred. A more 

specific treatment of the pre-stressing effect will be included in the test report of the shear and 

reinforcement anchorage test programs. This report will be published after completion of the 

test programs in December 2015. 

2. The large-scale test programs include measurements of CCI on the concrete surfaces parallel to 

the reinforcement mats in addition to direct measurement of concrete expansion using 

embedded pins. As discussed in the response to Request #1, these CCI measurements provide 

a reasonable approximation of the total strain in the concrete and the embedded reinforcement 
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SBK-L-15202 /Enclosure 2/ Page 45 

in the in-plane directions after crack initiation. The large scale test programs did not include a 

direct measurement of rebar strain. 

Rebar stress in the test specimens can be determined based on the CCI and the steel elastic 

modulus. However, the key quantitative measure from the large scale test programs was 

structural capacity of the reinforced concrete test specimens. As discussed in the response to 

Request #1, ASR-induced expansion did not result in a reduction in structural capacity, even 

though ASR-induced expansion does result in rebar stress. The reason for this observation is 

the pre-stressing effect. 

3. CCI was one of the methods for monitoring expansion in test specimens in the large-scale test 

program. At Seabrook Station, CCI will be used for initial screening of locations potentially 

affected by ASR and for evaluating the capacity of anchors in ASR-affected concrete. CCI 

will not be used for evaluation of ASR-affected concrete for shear capacity or reinforcement 

anchorage. See response to Request #2 ofRAI B.2.l.31A-5(al) for additional discussion. 

4. Elastic modulus testing of cores obtained during installation of extensometers at Seabrook 

Station will be used to calculate through-thickness expansion to date of the concrete 

component being evaluated. The calculated pre-instrument expansion will be added to the 

measurement obtained from the extensometer to determine total expansion. NextEra will 

compare this total expansion to the specific threshold defined in the ASR AMP to determine if 

additional structural evaluation is necessary. 

Because direct measurement of the extensometer will be used for monitoring in the future, 

monitoring for degradation of elastic modulus is not required beyond the initial coring at the 

time of extensometer installation. Specific limits on elastic modulus are also not required, 

since the total expansion criterion will address the monitored parameter (i.e., expansion) that 

elastic modulus test results are used to determine. Additionally, the elastic modulus values 

obtained at the time of extensometer installation will not be used as part of structural 

evaluations, because they are not representative of structural performance of ASR-affected 

reinforced concrete components. 

5. As previously discussed, NextEra will monitor volumetric effects of ASR expansion by 

obtaining measurements in both the in-plane and through-thickness directions. Specifically, 

the ASR AMP specifies monitoring combined cracking index (CCI) for in-plane expansion and 

extensometer measurements for through-thickness expansion. 

The ASR AMP specifies that NextEra will continue monitoring ASR in all locations where an 

extensometer is installed. However, the criterion associated with shear capacity and 

reinforcement anchorage is based exclusively on through-thickness expansion. As discussed in 

the response to Request #2 in RAI B.2.l.31A-5(al), combination of the in-plane and through-
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SBK-L-15202 /Enclosure 2/ Page 46 

thickness values is not necessary because the in-plarie expansion represents such a small 

proportion of the overall expansion. 

6. NextEra has performed a complete revision of the ASR AMP (Enclosure 4) to incorporate the 

most up-to-date information on plans for managing ASR at Seabrook Station. This latest 

revision incorporates portions of the discussion above. NextEra has also updated the 

associated UFSAR Supplement (Enclosure 3) to reflect the responses above. 

References 

1. U.S. Nuclear Regulatory Commission Letter, "Request for Additional Information Related 

to the Review of the Seabrook Station License Renewal Application - Set 25 (TAC No. 

ME4028)," dated October 2, 2015. 

2. NextEra Energy letter SBK-L-15024, "Response to Requests for Additional Information for 

the Review of the Seabrook Station, License Renewal Application - SET 23 (TAC NO. ME 

4028) Relating to the Alkali-Silica Reaction (ASR) Monitoring Program," dated February 

23, 2015. 

3. NextEra Energy letter SBK-L-15107, "Response to Requests for Additional Information for 

the Review of the Seabrook Station, License Renewal Application - SET 23 (TAC NO. ME 

4028) Relating to the Alkali-Silica Reaction (ASR) Monitoring Program," dated June 30, 

2015. 

4. NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for 

Nuclear Power Plants," Revision 2. 

5. MPR-3727, "Seabrook Station: Impact of Alkali-Silica Reaction on Concrete Structures and 

Attachments," Revision 1. (Seabrook FP# 100716) 

6. MPR-4153, "Seabrook Station - Approach for Determining Through-Thickness Expansion 

from Alkali-Silica Reaction," Revision 1. (Seabrook FP# 100918) 
Enclosure 3 to SBK-L-15202 

Seabrook Station Updated Final Safety Analysis Report 

Supplement Section A.2.1.3 lA for Alkali-Silica Reaction 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 3/ Page 2 

A.2.1.31A ALKALI-SILICA REACTION (ASR) MONITORING 

The plant specific 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. 

The Structural Monitoring Program performs visual inspections of the concrete structures at 

Seabrook for indications of the presence of alkali-silica reaction (ASR). ASR involves the 

formation of an alkali-silica gel which expands when it absorbs water. This expansion is 

volumetric in nature but is most readily detected by visual observation of cracking on the surface 

of the concrete. This cracking is the result of expansion that is occurring in the in-plane 

directions. Expansion is also occurring perpendicular (through the thickness of the wall) to the 

surface of the wall, but cracking will not be visible in this direction from the accessible surface. 

Cracking on the surface of the concrete 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. Micro-cracking 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 in-plane expansion associated with ASR and to 

specify monitoring intervals. In selected locations, cores will be removed for modulus testing to 

establish the level of through-thickness expansion to date. Instruments ( extensometers) will be 

placed in the resulting bore holes to monitor expansion in this direction going forward. 

ASR is primarily detected by non-intrusive visual observation of cracking on the surface of the 

concrete. The cracking is typically accompanied by the presence of moisture and efflorescence. 

ASR may also be detected or confirmed by removal of concrete cores and subsequent 

petrographic analysis. 

A Combined Cracking Index (CCI) is established at thresholds at which structural evaluation is 

necessary (see table below). The Cracking Index (Cl) is the summation of the crack widths on the 

horizontal or vertical sides of 20-inch by 30-inch grid 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. A CCI of less than the 1.0 mm/m can be deemed acceptable with 

deficiencies (Tier 2). Deficiencies determined to be acceptable with further review are trended 

for evidence of further degradation. The change from qualitative monitoring to quantitative 

monitoring occurs when the Cracking Index (Cl) of the pattern cracking equals or is greater than 

0.5 mm/min the vertical and horizontal directions .. Concrete crack widths less than 0.05 mm 

cannot be accurately measured and reliably repeated with standard, visual inspection equipment. 

A CCI of 1.0 mm/m or greater requires structural evaluation (Tier 3). All locations meeting Tier 

3 criteria will be monitored via CCI on a Yz year (6-month) inspection frequency and added to the 

through-thickness expansion monitoring via extensometers. All locations meeting the Tier 2 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 3/ Page 3 

structures monitoring criteria will be monitored on a 2.5 year (30-month) frequency. CCI 

correlates well with strain in the in-plane directions and the ability to visually detect cracking in 

exposed surfaces making it an effective initial detection parameter. In the event ASR monitoring 

results indicate a need to amend either the monitoring program acceptance criteria or the 

frequency of monitoring; NextEra will take such action under the Operating Experience element 

of the Alkali-Silica Reaction Monitoring Program. 

Tier Structural Monitoring Recommendation for Individual Criteria Program Category Concrete Components 

Structural Evaluation 

Unacceptable (requires Implement enhanced ASR 1.0 mm/m or greater Combined 3 further evaluation) monitoring, such as through- Cracking Index (CCI) wall expansion monitoring using 

Extensometers. 

0.5 mm/m or greater CCI 

Quantitative Monitoring and CI of greater than 0.5 mmlm in 

Trending the vertical and horizontal 

Acceptable with directions. 

2 Any area with visual presence of Deficiencies 

ASR (as defined in FHWA-HIF-12-

Qualitative Monitoring 022) accompanied by a CI ofless 

than 0.5 mm/m in the vertical and 

horizontal directions. 

Routine inspection as prescribed by Area has no indications of pattern 

1 Acceptable the Structural Monitoring Program cracking or water ingress- No visual 

symptoms of ASR 

The Alkali-Silica Reaction Monitoring Program was initially based on published studies 

describing screening methods to determine when structural evaluations of ASR affected concrete 

are appropriate. Large scale destructive testing of concrete beams with accelerated ASR has 

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

that acceptance criterion of 1 mm/m a used provides sufficient margin. 

CCI's limitation for heavily reinforced structures is that in-plane expansion, and therefore CCI, 

has been observed in the large scale test programs to plateau at a relatively low level of 

accumulated strain (approximately 1 mm/m). No structural impacts from ASR have been seen at 

these plateau levels in the large scale testing program at the University of Texas at Austin, 

Ferguson Structural Engineering Laboratory. While CCI remains useful for the detection and 

monitoring of ASR at the initial stages, an additional monitoring parameter in the out-of-plane 

direction is required to monitor more advanced ASR progression. ASR expansion in the out-ofplane direction will be monitored by borehole extensometers installed in drilled core bore holes. 

Although the observed strains due to ASR are of very small magnitude and adequately monitored 

by CCI and extensometers, over large distances and with the right building geometry, they can 

result in discemable dimension changes in a structure. Additional monitoring of this relative 

displacement potential and its impact to plant systems and components is included in the ASR 

Monitoring Program. Specifically, monitoring includes identifying signs of relative displacement 

or building deformation (e.g., fire seal displacement, seismic gap width changes, pipe/conduit 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 3/ Page 4 

misalignments at penetrations or between adjacent structures, bent or displaced pipe/conduit and 

supports, doorway misalignments). Critical building geometry locations where the potential for 

deformation is likely will be monitored for displacement via location-specific techniques. 

--------------- -~- - ~-
Enclosure 4 to SBK-L-15202 

Seabrook Station Updated License Renewal Application Section 

B.2.1.3 IA for Alkali-Silica Reaction (ASR) 

Aging Management Program 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 2 

B.2.1.31A ALKALI-SILICA REACTION MONITORING PROGRAM 

PROGRAM DESCRIPTION 

NextEra Energy Seabrook Operating Experience (OE) indicates that Alkali-Silica Reaction 

(ASR) is present in concrete structures and will require monitoring though the Period of 

Extended Operation (PEO). 

Alkali-Silica Reaction 

ASR is an aging mechanism that may occur in concrete under certain circumstances. It was first 

identified and reported by Thomas Stanton in the late 1930s in the State of California. It is a 

reaction between the alkali ions in the concrete pore solution and reactive forms of silicate 

material (if present) in the aggregate. The reaction, which requires the presence of 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 cement paste. ASR is typically 

characterized by the presence of "map" or "pattern" cracking accompanied by dark staining. 

alkali cement + expansive gel cracking of the 

reactive aggregate aggregate and paste 

ASR Expansion Mechanism 

Impact of Confinement 

Reinforcing steel, loads on the concrete structure (i.e., deadweight of the structure itself), and the 

configuration of the structure provide confinement that restrains in-situ expansion of the gel and 

limits the resulting cracking in concrete. 

Since the impact of ASR on mechanical properties relates to the extent of cracking, restraint of 

the expansion limits the reduction of in-situ mechanical properties and overall degradation of 

structural performance. There is a prestressing effect that occurs when reinforcement restrains 

the expansion caused by ASR. This effect is similar to concrete prestressing or analogous to preloading a bolted joint. 

The concrete prestressing effect is only present when the concrete is confined. If the concrete is 

removed from the stress field, the concrete prestressing effect is lost. For example, a core taken 

from a reinforced concrete structure that has been affected by ASR will lose the confinement 

provided by the reinforcement and concrete surrounding the sample, and therefore is no longer 

representative of the concrete within its structural context. 

Seabrook Station Concrete 

The concrete mix designs used in original construction at Seabrook utilized an aggregate that 

was susceptible to ASR, which was not known at the time. Although the testing was conducted 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 3 

in accordance with the ASTM C289 and C227 standards, these standards were subsequently 

identified as limited in their ability to predict long term ASR for moderate to low reactive 

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 and potential local dimensional changes. 

In 2009, Seabrook tested seasonal groundwater samples to support the development of a License 

Renewal Application. The results showed that the groundwater had become aggressive and 

Seabrook 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). This discovery 

initiated an Extent of Condition evaluation. Because the ASR mechanism requires the presence 

of moisture or very high humidity 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. 

Large-Scale Testing Program 

The interim structural assessment (Reference document MPR-3727) of ASR-affected structures 

at Seabrook Station considered the various limit states for reinforced concrete and applied 

available literature data to evaluate structural capacity. This evaluation identified gaps in the 

publicly available test data and the applicability to the reinforcement concrete at Seabrook. The 

limited available data for shear capacity and reinforcement anchorage for ASR-affected 

reinforced concrete with two-dimensional reinforcement mats were not representative of 

Seabrook Station. This conclusion was driven largely by the facts that the literature data for 

reinforcement anchorage were from a test method that ACI indicates is unrealistic and the 

literature data for shear capacity were from test specimens only inches in size. Additionally, no 

data were available on anchor bolt capacity on reinforced concrete with two dimensional 

reinforcement mats like Seabrook Station. Based on the lack of representative literature data, 

Seabrook initiated large-scale test programs to evaluate shear capacity, reinforcement anchorage, 

and anchor bolt capacity of ASR-affected reinforced concrete. The large scale test programs are 

being executed by MPR Associates and the Ferguson Structural Engineering Laboratory (FSEL) 

at The University of Texas at Austin. 

The large-scale test programs were designed to max1m1ze representativeness of reinforced 

concrete structures at Seabrook Station using a reasonable number of test specimens. The 

specimens were designed to be representative of a reference location (i.e., the B electrical 

tunnel). To enable application of the results to other structures, the test results are compared to 

provisions in ACI 318-71 to determine an appropriate reduction factor to reflect ASR 

development for the structural capacity calculated using code equations. This approach is 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 4 

appropriate, because the test program methodology is consistent with the experimental methods 

used to generate the data on which the code equations from ACI 318 are based. 

Seabrook Station includes many reinforced concrete structures and the specific configuration of 

structural components within those structures varies widely. Because the approach for the test 

programs supplemented (rather than replaced) the design code, results from appropriately 

representative test specimens may be applied for reinforced concrete structures throughout 

Seabrook Station. Additionally, the test specimen design represents actual structures at Seabrook 

Station to the maximum extent practical. The design of the test specimens considered many 

factors to maximize representativeness including, but not limited to, the following: 

• Large specimen sizes to represent actual structures 

• Reinforcement configuration representative to Seabrook 

• Restraint to mimic context of an actual structure 

• Concrete mix design based on Seabrook specifications 

Results from the large-scale testing program can also be used to support evaluations of structures 

subjected to deformation. Specifically, the large-scale testing program provides insights on 

expansion behavior that can be used for engineering evaluations. For structural evaluations of 

components affected by deformation, the large scale testing program can be used in support of 

assessments for shear capacity, reinforcement anchorage, and anchor capacity. Building 

deformation could affect the function of structures and equipment in many ways (e.g., creating 

interferences between components, closing or widening seismic gaps, degrading fire seals or 

seismic gap seals, displacing piping, ducting or conduit). If such impacts occur, NextEra will 

address them based on the specific equipment configuration in the location of interest. 

Monitoring Program 

The plant specific 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. 

The Structural Monitoring Program performs visual inspections of the concrete structures at 

Seabrook for indications of the presence of alkali-silica reaction (ASR). ASR involves the 

formation of an alkali-silica gel which expands when it absorbs water. This expansion is 

volumetric in nature but is most readily detected by visual observation of cracking on the surface 

of the concrete. This cracking is the result of expansion that is occurring in the in-plane 

directions. Expansion is also occurring perpendicular (through the thickness of the wall) to the 

surface of the wall, but cracking will not be visible in this direction from the accessible surface. 

Cracking on the surface of the concrete 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. Micro-cracking 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. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 5 

Monitoring of crack growth is used to assess the in-plane expansion associated with ASR and to 

specify monitoring intervals. In selected locations, cores will be removed for modulus testing to 

establish the level of through-thickness expansion to date. Instruments ( extensometers) will be 

placed in the resulting bore holes to monitor expansion in this direction going forward. 

ASR is primarily detected by non-intrusive visual observation of cracking on the surface of the 

concrete. The cracking is typically accompanied by the presence of moisture and efflorescence. 

ASR may also be detected or confirmed by removal of concrete cores and subsequent 

petrographic analysis. 

A Combined Cracking Index (CCI) is established at thresholds at which structural evaluation is 

necessary (see table below). The Cracking Index (CI) is the summation of the crack widths on the 

horizontal or vertical sides of 20-inch by 30-inch grid 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. A CCI of less than the 1.0 mm/m can be deemed acceptable with 

deficiencies (Tier 2). Deficiencies determined to be acceptable with further review are trended 

for evidence of further degradation. The change from qualitative monitoring to· quantitative 

monitoring occurs when the Cracking Index (CI) of the pattern cracking equals or is greater than 

0.5 mm/m in the vertical and horizontal directions.. Concrete crack widths less than 0.05 mm 

cannot be accurately measured and reliably repeated with standard, visual inspection equipment. 

A CCI of 1.0 mm/m or greater requires structural evaluation (Tier 3). All locations meeting Tier 

3 criteria will be monitored via CCI on a Yz year (6-month) inspection :frequency and added to the 

through~thickness expansion monitoring via extensometers. All locations meeting the Tier 2 

structures monitoring criteria will be monitored on a 2.5 year (30-month) frequency. CCI 

correlates well with strain in the in-plane directions and the ability to visually detect cracking in 

exposed surfaces making it an effective initial detection parameter. In the event ASR monitoring 

results indicate a need to amend either the monitoring program acceptance criteria or the 

frequency of monitoring; N extEra will take such action under the Operating Experience element 

of the Alkali-Silica Reaction Monitoring Program. 
I 1-· 

U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 6 

Tier Structural Monitoring Recommendation for Individual 

Program Category Concrete Components 

Structural Evaluation 

Unacceptable (requires Implement enhanced ASR 

3 monitoring, such as through- further evaluation) wall expansion monitoring 

using Extensometers. 

Quantitative Monitoring and 

Trending 

2 Acceptable with 

Deficiencies 

Qualitative Monitoring 

Routine inspection as prescribed 

1 Acceptable by the Structural Monitoring 

Program 

Criteria 

1.0 mm/m or greater Combined 

Cracking Index (CCI) 

0.5 mmlm or greater CCI 

CI of greater than 0.5 mmlm in the 

vertical and horizontal directions. 

Any area with visual presence of ASR 

(as defined in FHWA-HIF-12-022) 

accompanied by a CI of less than 0.5 

mm/m in the vertical and horizontal 

directions. 

Area has no indications of pattern 

cracking or water ingress- No visual 

symptoms of ASR 

The Alkali-Silica Reaction Monitoring Program was initially based on published studies 

describing screening methods to determine when structural evaluations of ASR affected concrete 

are appropriate. Large scale destructive testing of concrete beams with accelerated ASR has 

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

that acceptance criterion of 1 mm/m a used provides sufficient margin. 

CCI's limitation for heavily reinforced structures is that in-plane expansion, and therefore CCI, 

has been observed in the large scale test programs to plateau at a relatively low level of 

accumulated strain (approximately 1 mm/m). No structural impacts from ASR have been seen at 

these plateau levels in the large scale testing program at the University of Texas at Austin, 

Ferguson Structural Engineering Laboratory. While CCI remains useful for the detection and 

monitoring of ASR at the initial stages, an additional monitoring parameter in the out-of-plane 

direction is required to monitor more advanced ASR progression. ASR expansion in the out-ofplane direction will be monitored by borehole extensometers installed in drilled core bore holes. 

Although the observed strains due to ASR are of very small magnitude and adequately monitored 

by CCI and extensometers, over large distances and with the right building geometry, they can 

result in discemable dimension changes in a structure. Additional monitoring of this relative 

displacement potential and its impact to plant systems and components is included in the ASR 

Monitoring Program. Specifically, monitoring includes identifying signs of relative displacement 

or building deformation (e.g., fire seal displacement, seismic gap width changes, pipe/conduit 

misalignments at penetrations or between adjacent structures, bent or displaced pipe/conduit and 

supports, doorway misalignments). Critical building geometry locations where the potential for 

deformation is likely will be monitored for displacement via location-specific techniques. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 7 

Acceptance Criteria 

The ASR Monitoring Program can essentially be broken down into three (3) subsections, as 

follows: 

• Expansion monitoring - Anchors 

• Expansion monitoring - Structural Capacity 

• Expansion monitoring - Building Deformation 

Expansion Monitoring- Anchor Performance 

For anchor performance, the large scale test programs show that ASR does not have an effect 

until in-plane expansion reaches a sufficiently high level. Therefore, the ASR AMP specifies 

that if the CCI exceeds a specified threshold, additional evaluation must be performed to justify 

continued acceptability of the anchors. The specific threshold will be defined by the results from 

the large scale test program, which will be available at a later date. 

This approach is based on the fact that anchor performance is sensitive to in-plane expansion, but 

not through-thickness expansion. In-plane expansion creates micro-cracks parallel to the axis of 

an anchor, mainly in the concrete cover. These micro-cracks perpendicular to the concrete 

surface have the potential to provide a preferential failure path within a potential breakout cone, 

leading to degraded anchor performance. 

Through-thickness expansion has the potential to create micro-cracks perpendicular to the axis of 

an anchor. These potential micro-cracks that open parallel to the concrete surface do not provide 

a preferential failure path to result in degraded anchor performance. An anchor loaded in tension 

would compress the through-thickness expansion and close any potential micro-cracks within the 

area of influence of that anchor. Without a 'short-circuit' of the breakout cone, throughthickness expansion is a non-factor in anchor performance. 

The flowchart below describes the approach for monitoring expansion to confirm satisfactory 

anchor performance. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 8 

Expansion Monitoring- Structural Capacity 

All Tier 3 locations require a structural evaluation to assess the condition of the structure against 

the design basis. The structural evaluation will include an assessment of shear capacity and 

reinforcement anchorage, and other applicable limit states. 

For shear and reinforcement anchorage performance, the large scale test programs show that, for 

concrete elements without transverse reinforcement, ASR does not have an adverse effect until at 

least the maximum through-thickness expansion observed in the test programs. Therefore, if 

through-thickness expansion in elements without transverse reinforcement exceeds a specified 

threshold, additional evaluation must be performed to justify continued acceptability of the 

structure. The specific threshold will be defined by the results of the large scale test program, 

which will be available at a later date. Seabrook will determine through-thickness expansion 

from the sum of the calculated pre-instrument expansion (using the methodology of MPR-4153) 

and subsequent measurements with an extensometer. 

This approach is based on the observed expansion behavior of the test specimens, in which 

expansion in the through-thickness direction was dominant. It is not necessary to have expansion 

criteria that also incorporate in-plane expansion, which represents a very small proportion of the 

volumetric strain. 

The flowchart below describes the approach for monitoring expansion to confirm shear and 

reinforcement anchorage performance. 

Expansion Monitoring- Building Deformation 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 9 

For building deformation location-specific measurements (e.g. via laser target and gap 

measurements) will be compared against location-specific criteria to evaluate acceptability of the 

condition. 

Structural evaluations will be performed on buildings and components affected by deformation 

as necessary to ensure that the structural function is maintained. Evaluations of structures will 

validate structural performance against the design basis, and may use results from the large-scale 

test programs, as appropriate. 

Evaluations for structural deformation will also consider the impact to functionality of affected 

systems and components (e.g., conduit expansion joints). NextEra will evaluate the specific 

circumstances against the design basis of the affected system or component. 

Structural evaluations will be used to determine whether additional corrective actions (e.g., 

repairs) to the concrete are required. Specific criteria for selecting effective corrective actions 

will be evaluated on a location-specific basis. 

Walk down for Signs of 

Relative Displacement 

~ ¥11 '~1 ;~~ !! 

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, and ASME Section 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 10 

XI, Subsection IWL Program, B.2.1.28 have been augmented by this plant specific Alkali-Silica 

Reaction (ASR) Monitoring Program, B.2.1.3 lA. 

Routine inspections are performed by the Structures Monitoring and the ASME Section XI, 

Subsection IWL Program. Areas that have no visual presence of ASR are considered 

"acceptable" (Tier 1 ). An area with a Combined Cracking Index (CCI) of less than 1.0 mm/m is 

deemed "acceptable with deficiencies" (Tier 2). An area with a CCI of 1.0 mm/m or greater is 

deemed "unacceptable" and requires further evaluation (Tier 3). In addition, an area that meets 

Tier 3 requirements will be monitored for through-thickness expansion in addition to CCI. 

Deformation caused by ASR will be monitored by taking location-specific measurements. These 

measurements will be compared against location-specific criteria to evaluate acceptability of the 

condition. 

Evaluations will be performed under the Seabrook Corrective Action Program (CAP) and an 

appropriate analysis will be performed to evaluate against the design basis of that structure. 

PROGRAM ELEMENTS 

ELEMENT 1- Scope of Program 

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 Structures Monitoring Program. License 

Renewal structures within the scope of this program include: 

• Containment Building (including equipment hatch missile shield) 

• Containment Enclosure Building 

• Containment Enclosure Ventilation Area 

• Service Water Cooling Tower including Switchgear Rooms 

• Control Building 

• Control Building Make-up Air Intake Structures 

• Diesel Generator Building 

• Piping (RCA) Tunnels 

• Main Steam and Feed Water East and West Pipe Chase 

• Waste Processing Building 

• Tank Farm 

• Condensate Storage Tank Enclosure 

• Emergency Feed Water Pump House Building, including Electrical Cable Tunnels 

and Penetration Areas (Control Building to Containment) 

• Fuel Storage Building 

• Primary Auxiliary Building including RHR Vaults 

• Service Water Pump House 

• Service Water Access (Inspection) Vault 

• Circulating Water Pump house Building (below elevation 21 '-0) 

• Safety Related Electrical Manholes and Duct Banks 

• Pre-Action Valve Building 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 11 

• Miscellaneous Non-Category I Yard Structures 

SBO Structure - Transformers and Switch Yard foundations 

Non-Safety-Related Electrical Cable Manhole, Duct Bank Yard Structures 

foundations 

Switchyard and 345 KV Power Transmission foundations 

• Non-Category I Structures 

Turbine Generator Building 

Fire Pump House 

Aboveground exterior tanks l-FP-TK-35-A, l-FP-TK-35-B, 1-FP-TK-36-

A, l-FP-TK-36-B and 1-FP-TK-29 foundations. 

Fire Pump House Boiler Building 

Non-Essential Switchgear Building 

Steam Generator Blowdown Recovery Building 

Intake & discharge Transition Structures 

The Seabrook Station Alkali-Silica Reaction (ASR) Monitoring Program manages the effects of 

cracking due to expansion and reaction with aggregates. 

ELEMENT 2- Preventive Actions 

There are no preventive actions specified in the Seabrook Station Structures Monitoring 

Program, which includes implementation of NUREG-1801 XI.SS, XI.S6, and XI.S7. These are 

monitoring programs only. Similarly, the ASR Monitoring Program does not rely on preventive 

actions. 

ELEMENT 3- Parameters Monitored/ Inspected 

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. 

The strains consequently produce the associated cracking. 

The program focuses on identifying evidence of ASR, which could lead to expansion due to the 

reaction with aggregates. The program reflects published guidance for condition assessment of 

structures and incorporates practices consistent with those used as part of the large-scale testing 

programs. 

Initial screening of ASR 

Walkdowns of the station are performed on a periodic basis (SMP walkdowns, Systems 

Walkdowns, Security Rounds, etc.). Visual symptoms of deterioration are noted and compared to 

those commonly observed on structures affected by ASR. Common visual symptoms of ASR 

include, but are not limited to, "map" or "pattern" cracking and surface discoloration of the 

cement paste surrounding the cracks. 

The cracking is typically accompanied by the presence of moisture and efflorescence. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 12 

Petrographic examination can be performed on a concrete specimen to aid in confirming the 

proposed diagnosis arrived upon from visual inspection of the concrete surface. Typical 

petrographic features of ASR generally consist of the following: 

• Micro-cracking in the aggregates and/or cement paste 

• Reaction rims around the aggregates. 

• Silica gel filling cracks or voids in the sample. 

• Loss of cement paste-aggregate bond. 

The lists of symptoms associated with the initial screening of ASR is consistent with many 

published documents, including but not limited to the Federal Highway Administration (FHWA) 

document FHWA-HIF-09-004, "Report on the Diagnosis, Prognosis, and Mitigation of AlkaliSilica Reaction (ASR) in Transportation Structures", and the Institution of Structural 

Engineering document "Structural Effects of Alkali-Silica Reaction: Technical Guidance on the 

Appraisal of Existing Structures." 

Expansion 

For ASR-affected surfaces at Seabrook Station, NextEra will monitor volumetric effects of ASR 

expansion by obtaining measurements in both the in-plane (X&Y directions) and throughthickness directions CZ-direction). Specifically, Seabrook will be monitoring the combined 

cracking index (CCI) for in-plane expansion and extensometer measurements for throughthickness expansion. Expansion from ASR results in cracking and a change to the material 

properties of the concrete, and eventually requires an evaluation to ensure adequate structural 

performance. 

Expansion is a readily quantifiable parameter and an effective method for determining ASR 

progression. Expansion measurements at Seabrook can be easily obtained in the in-plane 

directions. The Cracking Index (CI) is a quantitative assessment of cracking present in the cover 

concrete of affected structures. A CI measurement is taken on accessible surfaces exhibiting the 

typical ASR symptoms. The CI is the summation of the crack widths on the horizontal and 

vertical sides of a section of the ASR-affected concrete surface of predefined dimensions. 

Seabrook uses a grid size of20 inches by 30 inches. The CI in a given direction is converted and 

reported in units of mm/m. 

The horizontal and vertical Cls are then averaged to obtain a Combined Cracking Index (CCI) 

for each area of interest. Criteria used in assessment of expansion is expressed in terms of CCI 

and based on recommendations provided in MPR-3727, "Seabrook Station: Impact of AlkaliSilica Reaction on Concrete Structures and Attachments." 

Data analysis from the large-scale test programs is ongoing and monitoring thresholds will be 

developed based on the test reports. However, the overall methodology for using in-plane and 

through-thickness expansion values for various aspects of the monitoring program is summarized 

as follows: 

Initial screening for ASR will be performed using CCI only. CCI values exceeding 1 mm/m will 

trigger additional actions. CCI is a relatively simple, non-destrudive method for monitoring 

cracking that appropriately characterizes expansion until expansion reorients in the direction of 

least restraint (i.e., the through-thickness direction at Seabrook Station). 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 13 

The test programs indicated that direction of expansion is not significantly affected by the 

reinforcement when expansion is at or below approximately 1 mm/m. Beyond this expansion 

level, the two-dimensional reinforcement mats provide confinement in the in-plane directions, 

and through-thickness expansion dominates. 

Anchor Performance Monitoring Parameter 

For anchor performance, the large scale test programs show that ASR does not have an effect 

until in-plane expansion reaches a sufficiently high level. Therefore, if the CCI exceeds a 

specified threshold, additional evaluation must be performed to justify continued acceptability of 

the anchors. 

This approach is based on the fact that anchor performance is sensitive to in-plane expansion, but 

not through-thickness expansion. In-plane expansion creates micro-cracks parallel to the axis of 

an anchor, mainly in the concrete cover. These micro-cracks perpendicular to the concrete 

surface have the potential to provide a preferential failure path within a potential breakout cone, 

leading to degraded anchor performance. 

Through-thickness expansion has the potential to create micro-cracks perpendicular to the axis of 

an anchor. These potential micro-cracks that open parallel to the concrete surface do not provide 

a preferential failure path to result in degraded anchor performance. An anchor loaded in tension 

would compress the through-thickness expansion and close any potential micro-cracks within the 

area of influence of that anchor. Without a 'short-circuit' of the breakout cone, throughthickness expansion is a non-factor in anchor performance. 

Shear and Reinforcement Anchorage Performance Monitoring Parameter 

For shear and reinforcement anchorage performance, the large scale test programs show that 

ASR does not have an effect until at least the maximum through-thickness expansion observed in 

the test programs. Therefore, if through-thickness expansion exceeds a specified threshold, 

additional evaluation must be performed to justify continued acceptability of the structure. 

NextEra will determine through-thickness expansion from the sum of the calculated 

pre-instrument expansion (using the methodology of MPR-4153) and subsequent measurements 

with an extensometer. 

This approach is based on the observed expansion behavior of the test specimens, in which 

expansion in the through-thickness direction was dominant. It is not necessary to also 

incorporate in-plane expansion, which represents a very small proportion of the volumetric 

strain. 

Crack Width Summation 

Crack width summation is a simple methodology for initial assessment of ASR-affected 

components and is recommended by publicly available resources. 

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. 

----------------------
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 14 

The engineering strain in a structural member at the time of crack initiation (ccr) is equivalent to 

the tensile strength of the concrete divided by the elastic modulus ( ccr 0 1 I 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 (c :::::: ccr + CI). 

Figure A-1 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 (flL). As a result, the Cracking Index provides a reasonable 

approximation of the total strain applied to the concrete after crack initiation, because strain in 

the un-cracked concrete between cracks (€be) is minimal. 

Figure A-1 . Concrete Specimen put in Tension 

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. 

Change in Elastic Modulus and Extensometer Measurements 

The large scale test program shows that out-of-plane expansion dominates for structures with 

two-dimensional reinforcement mats (like seen at Seabrook Station). 

Data from the structural testing programs have shown that expansion in the in-plane direction 

plateaus at low expansion levels, while expansion in the though-thickness direction continues to 

mcrease. Seabrook will install the extensometers in Tier 3 and other selected locations to 

measure expansion in the through-thickness direction. This approach will enable measuring 

expansion for a given concrete structural member from the time the extensometer is installed and 

going forward. To calculate the total expansion, Seabrook will determine expansion from 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 15 

original construction until the time the extensometers are installed (pre-instrument + 

extensometer measurements). 

The method to determine the total ASR induced through-thickness expansion at each instrument 

location at Seabrook is to use a determined pre-instrument expansion based on the reduction in 

modulus of elasticity. 

MPR developed a correlation relating expansion to reduction in elastic modulus. The correlation 

developed from the large scale testing program data relating expansion to reduction in elastic 

modulus is applicable to reinforced concrete structures at Seabrook. The elastic modulus was 

chosen because the large scale test program showed it to be the most sensitive and most 

repeatable material property. The test data used to generate the correlation were obtained from 

the test specimens that were designed to be as representative as practical of the concrete at 

Seabrook, including the reinforcement detailing. Additionally, comparison against literature data 

shows that the correlation follows a trend that is consistent with other published studies. 

The extensometer measurements will provide direct measurements of through-thickness 

expansion going forwards. The measurements are the parameter to be monitored. The elastic 

modulus will not be monitored going forward; it will be used as a constant in the total expansion 

calculation. 

Deformation 

With ASR induced deformation the aging effects of concern are mostly component functionality 

and structural interferences. Extent of condition walkdowns are in progress with a focus on 

safety-relates components such as pumps, valves, conduits, piping etc. The identification of 

items of interest is entered into the Seabrook Corrective Action Program (CAP) to be 

dispositioned for impact on plant structures. Specific features to look for include, but not limited 

to, the following: 

• Distorted flexible couplings 

• Non-parallel pipe/conduit/RV AC joints 

• Gaps, distortions, or tears in seals 

• Crimped tubing 

• Distorted support members/structural steel 

• Distorted/bent anchor bolts 

• Offset rod hangers 

• Support members exceeding minimum clearance 

• Cracked welds 

• Support embedment plates - not flush with walls 

• Misaligned pipe flanges 

These features may indicate irreversible deformation of the affected component. 

Seabrook will update the walkdown guidance document as necessary to accommodate new 

Operating Experience (OE) identified during the walkdowns. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 16 

ELEMENT 4- Detection of Aging Effects 

Monitoring walkdowns are performed on a periodic basis. The Structural Monitoring Program 

(SMP) walkdowns identify areas that show symptoms of ASR being present. The SMP includes 

periodic visual inspection of structures and components for the detection of aging effects specific 

for that structure. The inspections are completed by qualified individuals at a frequency 

determined by the characteristics of the environment in which the structure is found. A structure 

found in a harsh environment is defined as one that is in an area that is subject to outside ambient 

conditions, very high temperature, high moisture or humidity, frequency large cycling of 

temperatures, frequent exposure to caustic materials, or extremely high radiation levels. For 

structures in these harsh environments, the inspection is conducted on a five (5) year basis (plus 

or minus one year due to outage schedule and two inspections within ten years. Structures not 

located in an area qualifying as a harsh environment are classified as being in a mild 

environment, and are inspected on a ten (10) year basis (plus or minus one year due to outage 

schedule and two inspections within twenty years). Discovery of ASR in an area lends itself to a 

minimum of a 2.5 year (30-month) inspection frequency. 

In-Plane Expansion 

Seabrook uses the CCI methodology to monitor the expansion of ASR affected areas in the inplane direction. A CCI is established at thresholds at which structural evaluation is necessary. 

The CCI of less than 1.0 mm/m can be deemed acceptable with deficiencies (Tier 2). 

Deficiencies determined to be acceptable with further review are trended for evidence of further 

degradation. A CCI of 1.0 mm/m or greater requires structural evaluation (Tier 3). All locations 

meeting Tier 3 will be monitored via CCI on a Yz year (6-month) inspection frequency. All 

locations meeting Tier 2 will be monitored on a 2.5 year (30-month) frequency. In the event 

ASR monitoring results indicate a need to amend either the monitoring program acceptance 

criteria or the frequency of monitoring, N extEra will take such action under the Operating 

Experience element of the Alkali-Silica Reaction Monitoring Program. 

Seabrook has established reference grids that track the CCI of ASR affected areas. These grids 

are 20" x 30" and consist of three parallel vertical lines and two parallel horizontal lines. 

Measurement referenced points (gage points) are installed at the intersections of horizontal and 

vertical lines of the reference grid to allow for long-term monitoring of potential ongoing 

expansion. The CI is obtained from measurements of crack widths along a set of lines drawn on 

the surface of a concrete member. Expansion is documented by measuring the increase in the 

length of the lines used to determine the CI (distance between gage points). A pocket-size crack 

comparator card and an optical comparator are used to take the measurements. 

The location of the CCI reference grid is established in the area that appears to exhibit the mostsevere deterioration due to ASR (accessibility and structure geometry also factor into the 

decision making progress on where to establish a grid). At Seabrook the axes of the reference 

grid/ grids are parallel and perpendicular to the main reinforcement of the associated reinforced 

concrete member. 

CCI correlates well with strain in the in-plane directions and the ability to visually detect 

cracking in exposed surfaces making it an effective initial detection parameter. CCI's limitation 

for heavily reinforced structures is that in-plane expansion (and therefore CCI) has been 

observed to plateau at a relatively low level of accumulated strain in test specimens from the 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 17 

large scale testing program. No structural impacts from ASR have been seen at these plateau 

levels in the large scale testing program. While CCI remains useful for the detection and 

monitoring of ASR at the initial stages, an additional monitoring parameter in the out-of-plane 

direction is required to monitor more advanced ASR progression. The difference between the inplane expansion and the through-thickness expansion is due to the reinforcement detailing and 

the resulting difference in confinement between the in-plane and through-thickness direction. 

Through thickness expansion is less confined due to the fact that there is no reinforcement in that 

direction, therefore, expansion occurs preferentially in the through-thickness direction. 

Out-of- Plane Expansion 

The need for out-of-plane expansion monitoring is triggered by a CCI exceeding 1 mm/m. The 

expansions of the test specimens fabricated and tested at Ferguson Structural Engineering 

Laboratory (FSEL), at The University of Texas at Austin, were significantly more pronounced in 

the through-thickness direction (i.e. perpendicular to the reinforcement mats) than the in-plane 

directions (i.e. on the faces of the specimens parallel to the reinforcement mats). Expansion in 

the in-plane direction plateaus at low levels, while expansion in the through-thickness direction 

continues to increase. 

Elastic Modulus 

To determine expansion prior to instrument installation, Seabrook will be removing concrete 

cores at the location in which the instruments will be installed and testing them for compressive 

strength and elastic modulus. Using the methodology from MPR-4153, the elastic modulus 

values will be used to determine pre-instrument expansion in the through-thickness direction. 

All Tier 3 locations will have a sufficient amount of concrete cores removed for the material 

testing. A minimum of two cores for both compression testing and modulus testing will be 

provided (total of a minimum of four cores). The approximate dimensions of the cores needed 

for testing are 4" diameter x 8" length. The cores will be taken perpendicular to the 

reinforcement mat. 

A visual examination of the cores will confirm there is no mid-plane crack and a petrographic 

exam will be performed to confirm the presence of ASR. 

Snap-Ring Borehole Extensometer 

Seabrook will install Snap-Ring Borehole Extensometers (SRBEs) at the station to monitor 

through-thickness expansion. The Instrument Evaluation Program evaluated performance of the 

SRBEs, along with two other instrument types, in a test specimen representative of the concrete 

at Seabrook Station over a one-year period. The SRBE provided accurate measurements of 

through-thickness expansion throughout the test program and did not exhibit any problems 

related to reliability. The test program involved cycles of extended exposure to high temperature 

and humidity, which bounds the conditions expected at Seabrook Station. 

The SRBE consists of a metal graphite rod that is held in place by an anchor placed in the 

borehole. Measurements are performed by using a depth micrometer to measure the distance 

from a reference anchor at the surface of the concrete to the end of the graphite rod. The SRBE 

design contains no electronics and does not require calibration. Therefore, failure of the SRBE is 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 18 

unlikely. In the event that an SRBE did fail (e.g., an anchor broke loose), Seabrook could install 

another SRBE nearby to the failed location and continue expansion monitoring. This will not 

result in significant loss of data. 

A SRBE will be installed in one of the core bore holes at each location that are produced by the 

removal of the concrete cores being used for material testing. The elastic modulus will only be 

measured at the time of core removal to determine pre-instrument expansion. 

Deformation 

The baseline walkdowns to identify the potential effects caused by building deformation due to 

ASR are in progress. Subsequent monitoring will be performed as part of future Structural 

Monitoring Program (SMP) walkdowns. A 2.5 year (30-month) inspection frequency will be 

applied in location where symptoms of deformation are identified; otherwise, the inspection 

frequency will follow the requirements of the SMP. 

ELEMENT 5- Monitoring and Trending 

Results of walkdowns are initially reviewed by a licensed Professional Engineer (PE) or 

qualified person to determine whether the symptoms shown have potential to be ASR and if CCI 

measurements are needed. 

In-Plane and Out-of-Plane Expansion 

For anchor capacity, shear capacity, and reinforcement anchorage, use in-plane expansion (CCI) 

and out-of-plane expansion (modulus + SRBE measurements) to compare with the test results 

from the Large Scale Testing program. 

ASR is a slow progressing phenomenon. Seabrook will consider the rate at which a location is 

approaching the CCI and expansion limits and take appropriate action to ensure continued 

structural adequacy. 

Deformation 

Location-specific measurements will be compared against location-specific criteria to evaluate 

acceptability of the condition. NextEra will determine appropriate criteria based on the 

walkdown results and the particular geometry and configuration in the area of interest. The limit 

will include margin to trigger action prior to loss of intended function. 

Evaluations will be performed under the Seabrook Corrective Action Program (CAP) and an 

appropriate analysis will be performed to evaluate against the design basis of that structure. 

ASR is a slow progressing phenomenon. Seabrook will consider the rate at which a location is 

approaching the deformation criteria and take appropriate action to ensure continued component 

functionality. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 19 

ELEMENT 6- Acceptance Criteria 

Identification of the typical symptoms indicative of ASR generates the need to initially start 

monitoring the area using CCI. 

In-Plane 

Criterion of lmm/m distinguishes between Tier 2 and Tier 3 locations in relation to CCI. The 

large scale test program shows agreement between embedded pins and CCI, therefore ensuring 

CCI is acceptable. A structural evaluation is needed when the CCI reaches what is classified as 

Tier 3 (CCI> 1 mm/m). CCI is also used to inform the deformation evaluation. 

Out-of-Plane 

In areas in which the CCI is classified as Tier 3 the expansion due to ASR will be monitored in 

the through-thickness direction as well. Specific acceptance criteria will be identified in the 

large scale test program test reports, which will be available at a later date. 

Deformation 

Location-specific criteria based on the current design basis, particular configuration and 

applicable indications. Future walkdowns will be compared to the baseline walkdowns that are 

currently in-progress. 

ELEMENT 7- Corrective Actions 

Structural evaluations are performed to ensure impacted structures are in compliance with the 

Current Licensing Basis are documented in the Corrective Action Program. The NextEra Energy 

Quality Assurance Program and Nuclear Fleet procedures will be utilized to meet Element 7 

Corrective Actions. (Ref: LRA A.1.5 and B.1.3.) 

ELEMENT 8- Confirmation Process 

NextEra Energy Quality Assurance Program and Nuclear Fleet procedures will be utilized to 

meet Element 8 Confirmation Process. (Ref: LRA A.1.5 and B.1.3.) 

ELEMENT 9- Administrative Controls 

NextEra Energy Quality Assurance Program and Nuclear Fleet procedures will be utilized to 

meet Element 9 Administrative Controls. (Ref: LRA A.1.5 and B.1.3.) 

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. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 20 

Historically, NextEra Energy Seabrook has experienced groundwater infiltration through cracks, 

capillaries, pore spaces, seismic isolation joints, and construction joints in the below grade walls 

of concrete structures. 

ASR Identification 

In 2009, NextEra performed a qualitative walkdown of plant structures 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 compressive strength and 

modulus of elasticity, and subjected to petrographic examinations. While the results showed that 

both compressive strength and modulus of elasticity were less than the expected values, which is 

symptomatic of ASR. The results of the petrographic examinations also showed that the samples 

had experienced Alkali-Silica Reaction (ASR). 

NextEra initiated an extent of condition evaluation and concrete core samples were taken from 

five additional areas of the plant 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 th~ 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 affected structures concluded that they are fully capable of performing their 

safety function but margin had been reduced. Material property results from cores removed from 

a reinforced concrete structure do not properly represent the actual structural performance 

because the structural context is lost. However, the areas are potentially subject to further 

degradation of material properties due to the effects of ASR. 

Building Deformation - Containment Enclosure Building (CEB) 

In late 2014, a walkdown was performed to investigate a concern from the NRC that water, 

leaking from SB-V-9, was leaking into the Mechanical Penetration (Mech Pen) area though 

building seals. The walkdown documented that a Mechanical Penetration area seal was found 

tom. The damaged seal was a vertical seismic gap seal between the Containment Enclosure 

Building (CEB) and the Containment Building (CB). It was then stated that the condition of the 

seal and other local evidence indicated that the damage to the seal appeared to be caused by 

relative building movement and not seal degradation (i.e. shrinkage or material deterioration). 

Following the discovery mentioned above, Engineering identified that the damage to the seal was 

caused by CEB outward radial deformation. Seabrook engaged the engineering firm of 

Simpson, Gumpertz, and Heger Inc. (SG&H), to perform visual assessments of accessible areas 

surrounding the CEB to determine the behavior of the CEB, whether the CEB movement is 

localized or widespread, and if other plant structures or components had been impacted. A 

Cause and Effect Diagram was prepared to understand the physical phenomena occurring with 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 4/ Page 21 

the CEB. SG&H was contracted to perform parametric studies using a linear finite element 

model of the CEB with boundary conditions modeling parameters appropriate for estimating 

structural deflections and deformed shapes. The results were compared to in-situ field 

measurements taken between structures and at seismic isolation joints between various 

structures. The deformation patterns simulated by finite element analysis (PEA) were generally 

similar to field measurements. The results of the PEA showed that the deformation of the CEB 

was most likely due to Alkali-Silica Reaction (ASR) expansion in the concrete when combined 

with the expected creep and swelling of the concrete. 

The root cause to the event was determined to be the internal expansion (strain) in the CEB 

concrete produced by ASR in the in-plane direction of the CEB shell and ASR expansion in the 

backfill concrete coincident with a unique building configuration. The Root Cause Evaluation 

identified that there are many different symptoms of building deformation. These include: 

• Conduit, duct, or piping seismic connection deformation 

• Gate or door misalignment 

• Seismic gap seal degradation 

• Seismic gap width variations 

• Fire seal degradation 

(Note: above list is not intended to be all inclusive) 

As a result walkdowns are ongoing in identifying the above symptoms that may have been 

missed during the Structural Monitoring Program Walkdowns that were conducted prior to this 

discovery. The items identified are being tracked via Seabrook's Corrective Action Program. 

Seabrook is in the process of revising the structural analysis of the CEB to include the effect of 

ASR expansion, using the deformations observed during the walkdowns as validation. 

Building Deformation - RHR & FSB 

Seabrook is currently evaluating observations of expansion resulting in building deformation in 

the Residual Heat Removal (RHR) Equipment Vault and the Fuel Storage Building (FSB). 

Because the evaluation of the RHR Equipment Vault and the FSB are ongoing and the observed 

deformation has not yet been conclusively attributed to ASR, the walkdown guidance has not 

been updated to reflect observations in these locations. 

Newly Identified Operating Experience (OE) 

Seabrook will update the Aging Management Program for any new plant-specific or industry 

OE. 
Enclosure 5 to SBK-L-15202 

LRA Appendix A - Final Safety Report Supplement Table A.3, 

License Renewal Commitment List Updated to Reflect Changes to Date 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 51Page2 

A.3 LICENSE RENEW AL COMMITMENT LIST 

No. PROGRAM or TOPIC COMMITMENT 

Provide confirmation and acceptability of the implementation of 

1. PWR Vessel Internals MRP-227-A by addressing the plant-specific Applicant/Licensee 

Action Items outlined in section 4.2 of the NRC SER. 

Closed-Cycle Cooling Enhance the program to include visual inspection for cracking, loss of 

2. material and fouling when the in-scope systems are opened for Water maintenance. 

Inspection of Overhead 

Heavy Load and Light Enhance the program to monitor general corrosion on the crane and 

3. Load (Related to trolley structural components and the effects of wear on the rails in 

Refueling) Handling the rail system. 

Systems 

Inspection of Overhead 

Heavy Load and Light 

4. Load (Related to Enhance the program to list additional cranes for monitoring. 

Refueling) Handling 

Systems 

Compressed Air Enhance the program to include an annual air quality test requirement 

5. Monitoring for the Diesel Generator compressed air sub system. 

6. Fire Protection Enhance the program to perform visual inspection of penetration 

seals by a fire protection qualified inspector. 

Enhance the program to add inspection requirements such as spalling, 

7. Fire Protection and loss of material caused by freeze-thaw, chemical attack, and 

reaction with aggregates by qualified inspector. 

8. Fire Protection Enhance the program to include the performance of visual inspection 

of fire-rated doors by a fire protection qualified inspector. 

UFSAR 

SCHEDULE 

LOCATION 

A.2.1.7 Complete 

Prior to the period of extended A.2.1.12 operation. 

Prior to the period of extended A.2.1.13 operation. 

Prior to the period of extended A.2.1.13 operation. 

A.2.1.14 

Prior to the period of extended 

operation. 

A.2.1.15 

Prior to the period of extended 

operation. 

Prior to the period of extended 

A.2.1.15 operation. 

A.2.1.15 

Prior to the period of extended 

operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 5/ Page 3 

Enhance the program to include NFPA 25 (2011 Edition) guidance 

for "where sprinklers have been in place for 50 years, they shall be 

9. Fire Water System replaced or representative samples from one or more sample areas 

shall be submitted to a recognized testing laboratory for field service 

testing". 

Enhance the program to include the performance of periodic flow 

10. Fire Water System testing of the fire water system in accordance with the guidance of 

NFP A 25 (2011 Edition). 

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 to evaluate wall thickness and inner diameter of the fire 

protection piping ensuring that corrosion product buildup will not 

result in flow blockage due to fouling. Where surface irregularities 

are detected, follow-up volumetric examinations are performed. 

11. Fire Water System These inspections will be documented and trended to determine if a 

representative number of inspections have been performed prior to 

the period of extended 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 commence during the ten year period prior to the period of 

extended operation and continue through the period of extended 

operation. 

Enhance the program to include 1) In-scope outdoor tanks, except fire 

water storage tanks, constructed on soil or concrete, 2) Indoor large 

volume storage tanks (greater than 100,000 gallons) designed to nearatmospheric internal pressures, sit on concrete or soil, and exposed 

Aboveground Steel 

internally to water, 3) Visual, surface, and volumetric examinations of 

12. the outside and inside surfaces for managing the aging effects ofloss of 

Tanks material and cracking, 4) External visual examinations to monitor 

degradation of the protective paint or coating, and 5) Inspection of 

sealant and caulking for degradation by performing visual and tactile 

examination (manual manipulation) consisting of pressing on the 

sealant or caulking to detect a reduction in the resiliency and 

pliability. 

Prior to the period of extended A.2.1.16 operation. 

Prior to the period of extended A.2.1.16 operation. 

Within ten years prior to the A.2.1.16 period of extended operation. 

Within 10 years prior to the A.2.1.17 period of extended operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 5/ Page 4 

Enhance the program to perform exterior inspection of the fire water 

storage tanks annually for signs of degradation and include an 

13. Fire Water System ultrasonic inspection and evaluation of the internal bottom surface of 

the two Fire Protection Water Storage Tanks per the guidance 

provided in NFPA 25 (2011 Edition). 

Enhance program to add requirements to 1) sample and analyze new 

14. Fuel Oil Chemistry fuel deliveries for biodiesel prior to offloading to the Auxiliary Boiler 

fuel oil storage tank and 2) periodically sample stored fuel in the 

Auxiliary Boiler fuel oil storage tank. 

Enhance the program to add requirements to check for the presence 

15. Fuel Oil Chemistry of water in the Auxiliary Boiler fuel oil storage tank at least once per 

quarter and to remove water as necessary. 

Enhance the program to require draining, cleaning and inspection of 

16. Fuel Oil Chemistry the diesel fire pump fuel oil day tanks on a 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, cleaning and inspection 

17. Fuel Oil Chemistry of the Diesel Generator fuel oil storage tanks, Diesel Generator fuel 

oil day tanks, diesel fire pump fuel oil day tanks and auxiliaiy boiler 

fuel oil storage tank. 

Reactor Vessel Enhance the program to specify that all pulled and tested capsules, 

18. Surveillance unless discarded before August 31, 2000, are placed in storage. 

Enhance the program to specify that if plant operations exceed the 

limitations or bounds defined by the Reactor Vessel Surveillance 

Reactor Vessel Program, such as operating at a lower cold leg temperature or higher 19. Surveillance fluence, the impact of plant operation changes on the extent of 

Reactor Vessel embrittlement will be evaluated and the NRC will be 

notified. 

Within ten years prior to the A.2.1.16 period of extended operation. 

A.2.1.18 Prior to the period of extended 

operation. 

Prior to the period of extended 

A.2.1.18 operation. 

Prior to the period of extended A.2.1.18 operation. 

Prior to the period of extended A.2.1.18 operation. 

A.2.1.19 

Prior to the period of extended 

operation. 

A.2.1.19 Prior to the period of extended 

operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 5/ Page 5 

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 capsule 

Reactor Vessel receives a neutron fluence that meets the schedule requirements of 10 

20. Surveillance CFR 50 Appendix Hand ASTM E185-82 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, without the 

Reactor Vessel intent to test it, is stored in a manner which maintains it in a condition 

21. Surveillance which would permit its future use, including during the period of 

extended operation. 

22. One-Time Inspection Implement the One Time Inspection Program. 

Implement the Selective Leaching of Materials Program. The 

Selective Leaching of program will include a one-time inspection of selected components 

23. Materials where selective leaching has not been identified and periodic 

inspections of selected components where selective leaching has been 

identified. 

Buried Piping And Implement the Buried Piping And Tanks Inspection Program. 

24. Tanks Inspection 

One-Time Inspection Implement the One-Time Inspection of ASME Code Class 1 Small 

25. of ASME Code Class 1 Bore-Piping Program. Small Bore-Piping 

Enhance the program to specifically address the scope of the 

External Surfaces program, relevant degradation mechanisms and effects of interest, the 

26. Monitoring refueling outage inspection frequency, the training requirements for 

inspectors, and the required periodic reviews to determine program 

effectiveness. 

A.2.1.19 

Prior to the period of extended 

operation. 

A.2.1.19 

Prior to the period of extended 

operation. 

A.2.1.20 Within ten years prior to the 

period of extended operation. 

Within five years prior to the 

A.2.1.21 period of extended operation. 

A.2.1.22 Within ten years prior to the 

period of extended operation 

Within ten years prior to the 

A.2.1.23 period of extended operation. 

Prior to the period of extended 

A.2.1.24 operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 5/ Page 6 

Inspection of Internal 

Surfaces in Implement the Inspection oflntemal Surfaces in Miscellaneous 

27. Miscellaneous Piping 

and Ducting Piping and Ducting Components Program. 

Components 

28. Lubricating Oil Enhance the program to add required equipment, lube oil analysis 

Analysis required, sampling frequency, and periodic oil changes. 

29. 

Lubricating Oil Enhance the program to sample the oil for the Reactor Coolant pump 

Analysis oil collection tanks. 

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

30. 

Lubricating Oil ultrasonic thickness measurement of the lower portion of the Reactor 

Analysis Coolant pump oil collection tanks prior to the period of extended 

operation. 

31. 

ASME Section XI, Enhance procedure to include the definition of "Responsible 

Subsection IWL Engineer". 

32. 

Structures Monitoring Enhance procedure to add the aging effects, additional locations, 

Program inspection frequency and ultrasonic test requirements. 

33. 

Structures Monitoring Enhance procedure to include inspection of opportunity when 

Program planning excavation work that would expose inaccessible concrete. 

Prior to the period of extended 

A.2.1.25 operation. 

A.2.1.26 

Prior to the period of extended 

operation. 

A.2.1.26 Prior to the period of extended 

operation. 

A.2.1.26 Prior to the period of extended 

operation. 

A.2.1.28 Prior to the period of extended 

operation. 

A.2.1.31 Prior to the period of extended 

operation. 

A.2.1.31 

Prior to the period of extended 

operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 5/ Page 7 

Electrical Cables and 

Connections Not 

Subject to 10 CFR Implement the Electrical Cables and Connections Not Subject to 10 

34. 50.49 Environmental CFR 50.49 Environmental Qualification Requirements program. 

Qualification 

Requirements 

Electrical Cables and 

Connections Not 

Subject to 10 CFR 

Implement the Electrical Cables and Connections Not Subject to 10 50.49 Environmental 

35. Qualification CFR 50.49 Environmental Qualification Requirements Used in 

Requirements Used in Instrumentation Circuits program. 

Instrumentation 

Circuits 

Inaccessible Power 

Cables Not Subject to 

10 CFR 50.49 Implement the Inaccessible Power Cables Not Subject to 10 CFR 

36. Environmental 50.49 Environmental Qualification Requirements program. 

Qualification 

Requirements 

37. Metal Enclosed Bus Implement the Metal Enclosed Bus program. 

38. Fuse Holders Implement the Fuse Holders program. 

A.2.1.32 

Prior to the period of extended 

operation. 

Prior to the period of extended A.2.1.33 operation. 

A.2.1.34 

Prior to the period of extended 

operation. 

A.2.1.35 

Prior to the period of extended 

operation. 

A.2.1.36 Prior to the period of extended 

operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 5/ Page 8 

Electrical Cable 

Connections Not 

Subject to 10 CFR Implement the Electrical Cable Connections Not Subject to 10 CFR 

39. 50.49 Environmental 50.49 Environmental Qualification Requirements program. 

Qualification 

Requirements 

40. 345 KV SF6 Bus Implement the 345 KV SF6 Bus program. 

Metal Fatigue of 

Enhance the program to include additional transients beyond those 

41. Reactor Coolant defined in the Technical Specifications and UFSAR. Pressure Boundary 

Metal Fatigue of Enhance the program to implement a software program, to count 

42. Reactor Coolant 

Pressure Boundary transients to monitor cumulative usage on selected components. 

Pressure -Temperature 

Limits, including Low Seabrook Station will submit updates to the P-T cw-ves and L TOP 

43. Temperature limits to the NRC at the appropriate time to comply with 10 CFR 50 

Overpressure Appendix G. 

Protection Limits 

A.2.1.37 

Prior to the period of extended 

operation. 

A.2.2.1 Prior to the period of extended 

operation. 

Prior to the period of extended A.2.3.l operation. 

Prior to the period of extended A.2.3.1 operation. 

The updated analyses will be 

submitted at the appropriate time 

A.2.4.1.4 to comply with 10 CFR 50 

Appendix G, Fracture 

Toughness Requirements. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 5/ Page 9 

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 ofNUREG/CR-6909. 

(1) 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 

Environmentally- of the reactor water environment. This includes applying the 

44. Assisted Fatigue appropriate Fen factors to valid CUFs determined from an existing 

Analyses (TLAA) fatigue analysis valid for the period of extended operation or from an 

analysis using an NRC-approved version of the ASME code or NRCapproved 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). 

At least two years prior to the A.2.4.2.3 period of extended operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 5/ Page 10 

45. Number Not Used 

Protective Coating Enhance the program by designating and qualifying an Inspector 46. Monitoring and 

Maintenance Coordinator and an Inspection Results Evaluator. 

Enhance the program by including, "Instruments and Equipment 

Protective Coating needed for inspection may include, but not be limited to, flashlight, 

47. Monitoring and spotlights, marker pen, mirror, measuring tape, magnifier, binoculars, 

Maintenance camera with or without wide angle lens, and self sealing polyethylene 

sample bags." 

Protective Coating Enhance the program to include a review of the previous two 

48. Monitoring and 

Maintenance monitoring reports. 

Protective Coating Enhance the program to require that the inspection report is to be 

49. Monitoring and evaluated by the responsible evaluation personnel, who is to prepare a 

Maintenance summary of findings and recommendations for future surveillance or 

repair. 

Perform UT of the accessible areas of the containment liner plate in ASME Section XI, 

50. Subsection IWE the vicinity of the moisture barrier for loss of material. Perform 

opportunistic UT of inaccessible areas. 

51. Number Not Used 

ASME Section XI, Implement measures to maintain the exterior surface of the 

52. Subsection IWL Containment Structure, from elevation -30 feet to +20 feet, in a 

dewatered state. 

Reactor Head Closure Replace the spare reactor head closure stud(s) manufactured from the 

53. Studs bar that has a yield strength> 150 ksi with ones that do not exceed 

150 ksi. 

Prior to the period of extended A.2.1.38 operation. 

Prior to the period of extended 

A.2.1.38 operation. 

Prior to the period of extended A.2.1.38 operation. 

A.2.1.38 Prior to the period of extended 

operation. 

Baseline inspections were 

completed during OR16. Repeat 

containment liner UT thickness A.2.1.27 examinations at intervals of no 

more than five (5) refueling 

outages. 

A.2.l.28 Complete 

Prior to the period of extended 

A.2.1.3 operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 5/ Page 11 

NextEra will address the potential for cracking of the primary to 

secondary pressure boundary due to PWSCC oftube-to-tubesheet 

welds using one of the following two options: 

1) Perform a one-time inspection ofa representative sample of tubeto-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 

Steam Generator Tube monitoring program to perform routine tube-to-tubesheet weld 

54. Integrity inspections for the remaining life of the 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 reactor coolant pressure boundary 

must be approved by the NRC as part of a license amendment 

request. 

Steam Generator Tube Seabrook will perform an inspection of each steam generator to 

55. Integrity assess the condition of the divider plate assembly. 

56. Closed-Cycle Cooling Revise the station program documents to reflect the EPRI Guideline 

Water System operating ranges and Action Level values for hydrazine and sulfates. 

Closed-Cycle Cooling Revise the station program documents to reflect the EPRI Guideline 

57. Water System operating ranges and Action Level values for Diesel Generator 

Cooling Water Jacket pH. 

Update Technical Requirement Program 5.1, (Diesel Fuel Oil Testing 

58. Fuel Oil Chemistry Program) ASTM standards to ASTM D2709-96 and ASTM D4057-

95 required by the GALL Xl.M30 Rev 1 

A.2.1.10 Complete 

A.2.1.10 

Within five years prior to the 

period of extended operation. 

A.2.1.12 

Prior to the period of extended 

operation. 

Prior to the period of extended A.2.1.12 operation. 

Prior to the period of extended 

A.2.1.18 operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 5/ Page 12 

Nickel Alloy Nozzles The Nickel Alloy Aging Nozzles and Penetrations program will 

59. and Penetrations implement applicable Bulletins, Generic Letters, and staff accepted 

industry guidelines. 

Buried Piping and Implement the design change replacing the buried Auxiliary Boiler 

60. Tanks Inspection supply piping with a pipe-within-pipe configuration with leak 

detection capability. 

61. 

Compressed Air Replace the flexible hoses associated with the Diesel Generator air 

Monitoring Program compressors on a frequency of every 10 years. 

62. Water Chemistry Enhance the program to include a statement that sampling 

frequencies are increased when chemistry action levels are exceeded. 

Ensure that the quarterly CVCS Charging Pump testing is continued 

during the PEO. Additionally, add a precaution to the test procedure 

63. Flow Induced Erosion to state that an increase in the CVCS Charging Pump mini flow 

above the acceptance criteria may be indicative of erosion of the mini 

flow orifice as described in LER 50-275194-023. 

Soil analysis shall be performed prior to entering the period of 

extended operation to determine the corrosivity of the soil in the 

64. Buried Piping and 

vicinity ofnon-cathodically protected steel pipe within the scope of 

this program. If the initial analysis shows the soil to be nonTanks Inspection corrosive, this analysis will be re-performed every ten years 

thereafter. 

65. Flux Thimble Tube Implement measures to ensure that the movable incore detectors are 

not returned to service during the period of extended operation. 

66. Number Not Used 

Prior to the period of extended 

A.2.2.3 operation. 

Prior to the period of extended 

A.2.1.22 operation. 

A.2.1.14 

Within ten years prior to the 

period of extended operation. · 

A.2.1.2 

Prior to the period of extended 

operation. 

Prior to the period of extended NIA operation. 

A.2.1.22 

Prior to the period of extended 

operation. 

Prior to the period of extended 

NIA operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 5/ Page 13 

Perform one shallow core bore in an area that was continuously 

wetted from borated water to be examined for concrete degradation 

Structures Monitoring and also expose rebar to detect any degradation such as loss of 

67. Program material. The removed core will also be subjected to petrographic 

examination for concrete degradation due to ASR per ASTM 

Standard Practice C856. 

Structures Monitoring Perform sampling at the leakoff collection points for chlorides, 

68. Program sulfates, pH and iron once every three months. 

69. 

Open-Cycle Cooling Replace the Diesel Generator Heat Exchanger Plastisol PVC lined 

Water System Service Water piping with piping fabricated from AL6XN material. 

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

Closed-Cycle Cooling determine whether the loss of material due to cavitation induced 

70. Water System erosion has been eliminated or whether this remains an issue in the 

primary component cooling water system. 

Alkali-Silica Reaction Implement the Alkali-Silica Reaction (ASR) Monitoring Program. 

(ASR) Monitoring Testing will be performed to confirm that parameters being 71. monitored and acceptance criteria used are appropriate to manage the Program effects of ASR. 

Flow-Accelerated Enhance the program to include management of wall thinning caused 

72. Corrosion by mechanisms other than F AC. 

Inspection of Internal Enhance the program to include performance of focused 

Surfaces in examinations to provide a representative sample of20%, or a 

73. Miscellaneous Piping maximum of 25, of each identified material, environment, and aging 

and Ducting effect combinations during each 10 year period in the period of 

Components extended operation. 

A.2.1.31 

No later than December 31, 

2015. 

Quarterly Preventive 

A.2.1.31 Maintenance Activity 

Implemented 

A.2.1.11 Complete. 

A.2.1.12 Within ten years prior to the 

period of extended operation. 

A.2.l.31A Prior to the period of extended 

operation. 

A.2.1.8 Prior to the period of extended 

operation. 

A.2.1.25 Prior to the period of extended 

operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 5/ Page 14 

Enhance the program to perform sprinkler inspections annually per 

the guidance provided in NFP A 25 (2011 Edition). Inspection will 

ensure that sprinklers are free of corrosion, foreign materials, paint, 

74. Fire Water System and physical damage and installed in the proper orientation (e.g., 

upright, pendant, or sidewall). Any sprinkler that is painted, corroded, 

damaged, loaded, or in the improper orientation, and any glass bulb 

sprinkler where the bulb has emptied, will be evaluated for 

replacement. 

Enhance the program to a) conduct an inspection of piping and 

branch line conditions every 5 years by opening a flushing 

connection at the end of one main and by removing a sprinkler 

toward the end of one branch line for the purpose of inspecting for 

the presence of foreign organic and inorganic material per the 

75. Fire Water System guidance provided in NFPA 25 (2011 Edition) and b) If the presence 

of sufficient foreign organic or inorganic material to obstruct pipe or 

sprinklers is detected during pipe inspections, the material will be 

removed and its source is determined and corrected. 

In buildings having multiple wet pipe systems, every other system 

shall have an internal inspection of piping every 5 years as described 

in NFP A 25 (2011 Edition), Section 14.2.2. 

Enhance the Program to conduct the following activities annually per 

the guidance provided in NFPA 25 (2011 Edition). 

76. Fire Water System • main drain tests 

• deluge valve trip tests 

• fire water storage tank exterior surface inspections 

A.2.1.16 Prior to the period of extended 

operation. 

A.2.1.16 

Prior to the period of extended 

operation. 

Prior to the period of extended 

A.2.1.16 operation . 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 5/ Page 15 

The Fire Water System Program will be enhanced to include the 

following requirements related to the main drain testing per the 

guidance provided in NFPA 25 (2011 Edition). 

77. Fire Water System • The requirement that if there is a 10 percent reduction in full flow 

pressure when compared to the original acceptance tests or 

previously performed tests, the cause of the reduction shall be 

identified and corrected if necessary. 

• Recording the time taken for the supply water pressure to return 

to the original static (nonflowing) pressure. 

Enhance the program to include periodic inspections of in-scope 

insulated components for possible corrosion under insulation. A 

External Surfaces sample of outdoor component surfaces that are insulated and a 

78. Monitoring sample of indoor insulated components exposed to condensation (due 

to the in-scope component being operated below the dew point), will 

be periodically inspected every 10 years during the period of 

extended operation. 

Open-Cycle Cooling 

Enhance the program to include visual inspection of internal 

79. coatings/linings for loss of coating integrity. Water System 

Enhance the program to include visual inspection of internal 

80. Fire Water System coatings/linings for loss of coating integrity. 

Enhance the program to include visual inspection of internal 

81. Fuel Oil Chemistry coatings/linings for loss of coating integrity. 

Inspection of Internal 

Surfaces in Enhance the program to include visual inspection of internal 

82. Miscellaneous Piping coatings/linings for loss of coating integrity. 

and Ducting 

Components ·. 

A.2.1.16 Prior to the period of extended 

operation. 

Prior to the period of extended A.2.1.24 operation. 

Within 10 years prior to the A.2.1.11 period of extended operation. 

Within 10 years prior to the A.2.1.16 period of extended operation. 

Within 10 years prior to the A.2.1.18 period of extended operation. 

Within 10 years prior to the A.2.1.25 period of extended operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 5/ Page 16 

Enhance tile ASR AMP to install extensometers in Tier 3 areas of 

structures to monitor expansion due to alkali-silica reaction in the 

Alkali-Silica Reaction out-of-plane direction. 

83. Monitoring Monitoring expansion in the out-of-plane direction will commence 

upon installation of the extensometers in 2016 and continue through 

the period of extended operation. 

84. 

ASME Section XI, Evaluate the acceptability of inaccessible areas for structures within 

Subsection IWL the scope of ASME Section XI, Subsection IWL Program. 

Enhance the program to perform additional tests and inspections on 

the Fire Water Storage Tanks as specified in Section 9.2.7 ofNFPA 

85. Fire Water System 25 (2011 Edition) in the event that it is required by Section 9.2.6.4, 

which states "Steel tanks exhibiting signs of interior pitting, 

corrosion, or failure of coating shall be tested in accordance with 

9.2.7." 

86. Fire Water System Enhance the program to include disassembly, inspection, and 

cleaning of the mainline strainers every 5 years. 

Increase the frequency of the Open Head Spray Nozzle Air Flow Test 

87. Fire Water System from every 3 years to every refueling outage to be consistent with 

LR-ISG-2012-02, AMP XI.M27, Table 4a. 

Enhance the program to include verification that a) the drain holes 

associated with the transformer deluge system are draining to ensure 

complete drainage of the system after each test, b) the deluge system 

88. Fire Water System drains and associated piping are configured to completely drain the 

piping, and c) normally-dry piping that could have been wetted by 

inadvertent system actuations or those that occur after a fire are 

restored to a dry state as part of the suppression system restoration. 

A.2.l.31A December 31, 2016. 

A.2.1.28 Prior to the period of extended 

operation. 

A.2.1.16 Prior to the period of extended 

operation. 

A.2.1.16 

Prior to the period of extended 

operation. 

Prior to the period of extended 

A.2.1.16 operation. 

Within five years prior to the 

A.2.1.16 period of extended operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-15202 /Enclosure 5/ Page 17 

Inspection of Internal Incorporate Coating Service Level III requirements into the RCP 

Surfaces in Motor Refurbishment Specification for the internal painting of the 

89. Miscellaneous Piping motor upper bearing coolers and motor air coolers. All four RCPs 

and Ducting will be refurbished and replaced using the Coating Service Level III 

Components requirements prior to entering the period of extended operation. 

Implement the PWR Vessel Internals Program. The program will be 

90. PWR Vessel Internals implemented in accordance with MRP-227-A (Pressurized Water 

Reactor Internals Inspection and Evaluation Guidelines) and NEI 03-

08 (Guideline for the Management of Materials Issues). 

Enhance the ASR Aging Management Program to require 

structural evaluations be performed on buildings and components 

affected by deformation as necessary to ensure that the 

structural function is maintained. Evaluations of structures will 

validate structural performance against the design basis, and 

may use results from the large-scale test programs, as 

appropriate. Evaluations for structural deformation will also 

Alkali-Silica Reaction 

consider the impact to functionality of affected systems and 

91. components (e.g., conduit expansion joints). NextEra will 

Monitoring evaluate the specific circumstances against the design basis of the 

affected system or component. 

Enhance the ASR AMP to include additional parameters to be 

monitored based on the results of the CEB Root Cause, Structural 

Evaluation and walk downs. Additional parameters monitored 

will include: alignment of ducting, conduit, and piping; seal 

integrity; laser target measurements; key seismic gap 

measurements; and additional instrumentation. 

Prior to the period of extended A.2.1.25 operation. 

A.2.1.7 

Prior to the period of extended 

operation 

A.2.l.31A March 15, 2020 
Enclosure 6 to SBK-L-15202 

Next Era Seabrook: Application for Withholding 

Proprietary Information from Public Disclosure 

Affidavit in Support of Application for Withholding 

Proprietary Information from Public Disclosure 
NextEra Energy Seabrook, LLC 

AFFIDAVIT IN SUPPORT OF APPLICATION FOR WITHHOLDING 

PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE 

County of Rockingham ) 

) 

State of New Hampshire ) 

I, Dean Curtland, being duly sworn according to law, depose and state the following: 

(1) I am the Site Vice President ofNextEra Energy Seabrook, LLC (NextEra Energy 

Seabrook), and have been delegated the function of reviewing the information described 

in paragraph (3) which is sought to be withheld, and have been authorized to apply for its 

withholding. 

(2) I am making this Affidavit in conjunction with NextEra Energy Seabrook's 

"Application for Withholding Proprietary Information from Public Disclosure" 

accompanying this Affidavit and in conformance with the provisions of 10 CFR Section 

2.390. 

(3) The information sought to be withheld is contained in Enclosures 1 ofNextEra 

Energy Seabrook's letter SBK-L-15202, Dean Curtland (NextEra Energy Seabrook) to 

U.S. Nuclear Regulatory Commission, entitled "Seabrook Station Response to Requests 

for Additional Information for the Review of the Seabrook Station, License Renewal 

Application- SET 25 (TAC NO. ME4028) Relating to the Alkali-Silica Reaction (ASR) 

Monitoring Program," dated November 30, 2015. The NextEra Energy Seabrook 

proprietary information in Enclosure 1 of SBK-L-15202 is identified by enclosing boxes 

(LJ). 

( 4) The information sought to be withheld is considered to be proprietary and 

confidential commercial information because alkali-silica reaction (ASR) is a newlyidentified phenomenon at domestic nuclear plants. The information requested to be 

withheld is the result of several years of intensive N extEra Energy Seabrook effort and 
the expenditure of a considerable sum of money. This information may be marketable in 

the event nuclear facilities or other regulated facilities identify the presence of ASR. In 

order for potential customers to duplicate this information, similar technical programs 

would have to be performed and a significant manpower effort, having the requisite talent 

and experience, would have to be expended. The extent to which this information is 

available to potential customers diminishes NextEra Energy Seabrook's ability to sell 

products and services involving the use of the information. Thus, public disclosure of the 

information sought to be withheld is likely to cause substantial harm to N extEra Energy 

Seabrook's competitive position and NextEra Energy Seabrook has a rational basis for 

considering this information to be confidential commercial information. 

(5) The information sought to be withheld is being submitted to the NRC in 

confidence. 

(6) The information sought to be withheld has, to the best of my knowledge and 

belief, consistently been held in confidence by NextEra Energy Seabrook, has not been 

disclosed publicly, and not been made available in public sources. 

(7) The information is of a sort customarily held in confidence by NextEra Energy 

Seabrook, and is in fact so held. 

(8) All disclosures to third parties, including any required transmittals to the NRC, 

have been or will be pursuant to regulatory provisions and/or confidentiality agreements 

that provide for maintaining the information in confidence. 

I declare that the foregoing affidavit and the matters stated therein are true and correct to 

the best of my knowledge, information, and belief. Further, the affiant sayeth not. 

Subscrib&d and sworn to before me 

this~ day of December, 2015. 

lr(cteM~~ .~)~ Notary Public 

My commission expires 1 oJ l l /a.o 11 

• lli/~~ Dean Curtland 

Site Vice President 

NextEra Energy Seabrook, LLC 

626 Lafayette Road 

Seabrook, New Hampshire 03874 

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