Enclosure 2 to this Letter Contains Proprietary Information 

Withhold Enclosure 2 from Public Disclosure in Accordance with 10 CFR 2.390 

NEXTera® 

ENERGY~ 

August 9, 2016 

10 CFR54 

SEABROOK 

SBK-L-16013 

Docket No. 50-443 

U. S. Nuclear Regulatory Commission 

Attn: Document Control Desk 

Washington, DC 20555-0001 

References: 

Seabrook Station 

Seabrook Station, License Renewal Application Relating to the 

Alkali-Silica Reaction CASR) Monitoring Program 

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

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

2. NextEra Energy Seabrook, LLC letter SBK-L-15202, "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," December 3, 2015. (Accession Number ML15343A470) 

3. U.S. Nuclear Regulatory Commission Public Meeting Summary, Public Meeting with 

NextEra Energy Seabrook, LLC Regarding Aging Management of Alkali-Silica Reaction 

Pertaining to the License Renewal of Seabrook Station, Unit 1.;April 28, 2016. (Accession 

Number ML16146Al 76) 

In Reference 1, NextEra Energy Seabrook, LLC (NextEra Energy Seabrook) submitted an 

application for a renewed facility operating license for Seabrook Station Unit 1 in accordance 

with the Code of Federal Regulations, Title 10, Parts 50, 51, and 54. 

In Reference 2, NextEra Energy Seabrook responded to the staff Request for Additional 

Information refated to Alkali Silica Reaction. As documented in Reference 3 a meeting for 

NextEra Energy Seabrook, LLC (NextEra) to discuss their plan to address the U.S. Nuclear 

Regulatory Commission (NRC) staffs concerns related to the aging management of alkali-silica 

reaction (ASR) for Seabrook was conducted. f:)!> 

The enclosures provide additional clarifying information as discussed with the staff. Enclosure ~ O 

1 provides a general description of the Containment and Containment Enclosure Building and 

1 

I 1 ( 

discussion of clarifying changes to proposed commitments. Jr-{ ~ \ 

NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874 "1 ~ R 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 I Page 2 

Enclosure 3 provides a revised LRA Appendix A - Updated Final Safety Analysis Report Section 

A.2.l.31A for Alkali-Silica Reaction and A.2.l.31B for Building Deformation. Enclosure 4 

provides a revised LRA Appendix B Section B.2.l.31A for Alkali-Silica Reaction (ASR) and 

B.2.l.31B for Building Deformation Aging Management Programs. These two revisions 

supersede the respective previously submitted sections to the LRA. The enclosed ASR and 

Building Deformation aging management programs are intended to assure that the effects of 

aging will be adequately managed so that the intended function(s) will be maintained consistent 

with the current licensing basis (CLB) for the period of extended operation. NextEra recently 

requested an amendment of the Seabrook operating license (LAR) to modify its CLB to include 

methodologies to evaluate concrete affected by ASR in letter SBK-L-i 6071. The enclosed AMPs 

are consistent with the methodologies presented in the LAR. Any changes to those 

methodologies resulting from the NRC's review of the LAR will be reflected in the AMPs 

through updating the CLB or as operating experience, as appropriate. 

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

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

Commitments 71, 83 and 91. 

This letter is supported by an affidavit signed by NextEra Energy Seabrook (Enclosure 6), 

setting forth the basis on which the information in Enclosure 2 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. The information provided in Enclosure 2 

does not affect the non-proprietary version of information previously submitted in Reference 2. 

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. Ken Browne 

Licensing Manager, at (603) 773-7932. 

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

Executed on August q-k, 2016 

Sincerely, 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 I Page 3 

Enclosures: 

Enclosure 1 - Seabrook Station License Renewal Application Relating to the Alkali-Silica 

Reaction (ASR) Monitoring Program for the Seabrook Station License Renewal 

Application Clarifying Information 

Enclosure 2 - Seabrook Station License Renewal Application Seabrook Station License 

Renewal Application Resubmittal of SBK-L-15202 Enclosure 1 -Figures 17-19 

(Proprietary) 

Enclosure 3 - Seabrook Station Updated Final Safety Analysis Report Section A.2.1.3 lA for 

Alkali-Silica Reaction and Section A.2.1.3 lB for Building Deformation 

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

Reaction and Section B.2.1.3 lB for Building Deformation Aging Monitoring 

Programs 

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 

cc: D. H. Dorman 

J.C. Poole 

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 1 to SBK-L-16013 

Seabrook Station License Renewal Application 

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

for the Seabrook Station License Renewal Application 

Clarifying Information 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 1/ Page 2 

Containment Structures 

During discussions with the staff it was noted that some external stakeholders are not familiar with 

the unique design of Seabrook Containment Structures, particularly the relationship of the 

Containment Enclosure Building (CEB) and Containment. To assist the staff in addressing this 

topic, the following general description is provided. Additional information related to the CEB 

and Containment can be found in the License Renewal Application (Reference 1) Section 2.4.2 

and Seabrook Station Updated Final Safety Analysis Report Sections 1.2.2, 3.8 and 6.5.3. 

Containment 

The containment building is a reinforced concrete structure that is designed in accordance with the 

requirements of Section III of the American Society of Mechanical Engineers (ASME) Boiler & 

Pressure Vessel Code (1975 Edition). The containment structure completely encloses a Reactor 

Coolant System and is a seismic Category I reinforced concrete structure in the form of a right 

vertical cylinder with a hemispherical dome and flat foundation mat founded on bedrock. The 

inside face is lined with a welded carbon steel plate, providing a high degree of leak tightness. A 

protective 4-ft thick concrete mat that forms the floor of the Containment protects the liner over 

the foundation mat. The containment structure provides biological shielding for normal and 

accident conditions. 

Containment penetrations are provided in the lower portion of the structure, and consist of a 

personnel lock and an equipment hatch/personnel lock, a fuel transfer tube, and piping, electrical, 

instrumentation, and ventilation penetrations. All penetrations are anchored to sleeves (or to 

barrels) which are embedded in the concrete containment wall. This embedment is accomplished 

by means of an engineered anchorage system that is welded to the sleeve (or barrel) which is, in 

turn, welded to the locally thickened liner. The Containment is heavily reinforced with tri-axial 

reinforcement. 

The containment structure provides the primary containment. 

Containment Enclosure Building 

The Containment Enclosure Building_ was designed and constructed to comply with the 

1971 edition of ACI 318, Building Code Requirements for Reinforced Concrete (Reference 

Seabrook Station UFSAR Section 3.8.4.) The containment enclosure surrounds the containment 

structure and is designed in a similar configuration as a vertical right cylindrical seismic Category 

I, reinforced concrete structure with dome and ring base. 

The containment enclosure is designed to entrap, filter and then discharge any leakage from the 

containment structure. To accomplish this, the space between the containment enclosure and the 

containment structure, as well as the penetration and safety-related pump areas, are maintained at a 

negative pressure following a loss-of-coolant accident, by fans which take suction from the 

containment enclosure and exhaust to atmosphere through charcoal filters. To ensure air tightness 

for the negative pressure, leakage through all joints and penetrations has been minimized. 

The containment enclosure provides the secondary containment. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 1/ fage 3 

Commitment 71 

Commitment 71 was originally submitted when testing at the University of Texas Ferguson 

Structural Engineering Laboratory was still in progress. As structural testing has been completed 

the commitment has been updated to confirm the results from testing will be applied to the ASR 

Monitoring Program described in B.2.1.3 lA of the License Renewal Application. 

Commitment 71 has been updated to reflect the addition of a new plant specific aging management 

program and ongoing actions based on the completed testing. 

NextEra has completed testing at the 

University of Texas Ferguson Structural 

Engineering Laboratory which 

demonstrates the parameters being 

monitored and acceptance criteria used 

Alkali-Silica are appropriate to manage the effects of 

Reaction (ASR) ASR. 

Monitoring NextEra wi:ll-Iimplement the Alkali-Silica A.2.l.31A Prior to entering the 

71 Program/ Reaction (ASR) Monitoring Program and A.2.1.31B period of extended 

Building Building Deformation Monitoring operation. Deformation Program described in B.2.1.31A and Monitoring B.2.1.31B of the License Renewal 

Program Application. 

+esting will be fl6Ffermea te een:f:irm that 

flaFaHl:eteFS being ffiSHitSFea ana ae66fltaHG6 

GFitefia usea aFe aflflFSflFiate te manage the 

effects efAgR: 

Commitment 83 

Commitment #83 was restructured to ensure a more precise description that extensometers will be 

installed in all Tier 3 locations that are subject to possible thru wall expansion due to ASR 

progression in two dimensional reinforced structures. Monitoring of out-of-plane expansion will 

commence with the installation of the extensometer and continue on a six month basis. 

Enhance the ASR AMP to install 

extensometers in all Tier 3 areas of two 

dimensional reinforced structures to 

monitor expansion due to alkali-silica 

Alkali-Silica reaction in the out-of-plane direction. 

83. Reaction A.2.l.31A December 31,2016. 

Monitoring Monitoring expansion in the out-of-plane 

direction will commence upon installation 

of the extensometers ~ and continue 

011 a six month frequency through the 

period of extended operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 1/ Page 4 

Commitment 91 

Commitment 91 has revised to reflect a new plant specific Aging Management Program 

for Building Deformation. 

Implement the Building Deformation 

Monitoring Program 

Enhance the AKR ,<\gmg Management 

Structures Monitoring 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 largeAl*ali ~miea 

scale test programs, as appropriate. 

Evaluations for structural deformation will 

Reaetion also consider the impact to functionality of 1\.2.1.311\ 91 Building affected systems and components (e.g., March 15, 2020 

A.2.1.31B Deformation conduit expansion joints). NextEra will 

Monitoring evaluate the specific circumstances against 

the design basis of the affected system or 

component. 

Enhance the AS& Building Deformation 

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. 
Enclosure 3 to SBK-L- 16013 

Seabrook Station Updated Final Safety Analysis Report 

Section A.2.1.3 lA for Alkali-Silica Reaction and 

Section A.2.1.31 B for Building Deformation 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /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. 

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. 

Monitoring of crack growth is used to assess the in-plane expansion associated with ASR and to 

specify monitoring intervals. 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 of20-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/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 Yi year 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 3/ Page 3 

(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, NextEra will take such action under 

the Operating Experience element of the Alkali-Silica Reaction Monitoring Program. 

Tier Structural Monitoring Recommendation for lndiyidual Criteria Program Category Concrete Components 

Structural Evaluation 

Unacceptable (requires Implement enhanced ASR 1.0 nun/m or greater Combined 3 monitoring, such as through- further evaluation) wall expansion monitoring using Cracking Index (CCI) 

Extensometers. 

0.5 nun/m or greater CCI 

Quantitative Monitoring and CI of greater than 0.5 nun/min 

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 nun/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 an acceptance criterion of 1 mm/m provides sufficient margin. 

For heavily reinforced structures, CCI is limited in 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-of-plane direction will be 

monitored by borehole extensometers installed in drilled core bore holes. In the selected 

locations, cores will be removed for modulus testing to establish the level of through-thickness 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 3/ Page 4 

expansion to date. Instruments ( extensometers) will be placed in the resulting bore holes to 

monitor expansion in this direction going forward so the limits specified below are maintained. 

Structural Design Issue Criteria1 

Flexure & reinforcement anchorage See FP#101020 - Section 2.1 

Shear See FP#101020 - Section 2.1 

Anchor bolts and structural attachments See FP#101020 - Section 2.1 

A.2.1.31B BUILDING DEFORMATION MONITORING 

The Building Deformation Monitoring Program is a plant specific program implemented under 

the existing Maintenance Rule Structures Monitoring Program. Building Deformation is an aging 

mechanism that may occur as a result of other aging effects of concrete. Building Deformation at 

Seabrook is primarily a result of the alkali silica reaction (ASR) but can also result from 

swelling, creep, and shrinkage. Building deformation can cause components within the structures 

to move such that their intended :functions may be impacted. 

The Building Deformation Monitoring Program uses visual inspections associated with the 

Structures Monitoring Program and cracking measurements associated with the Alkali-Silica 

Reaction program to identify buildings that are experiencing deformation. The first inspection is 

a baseline to identify areas that are exhibiting surface cracking. The surface cracking will be 

characterized and analytically documented. This inspection will also identify any local areas that 

are exhibiting deformation. The extent of surface cracking will be input into an analytical model. 

This model will determine the extent of building deformation and the frequency of required 

visual inspections. 

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, 

additional inspections and/or analysis) to the concrete or components are required. Specific 

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

1 Expansion Limit Criteria is considered proprietary to NextEra Seabrook. FP #101020 MPR-4288, Revision 0, "Seabrook 

Station: Impact of Alkali-Silica Reaction on the Structural Design Evaluations," July 2016, was previously submitted to the NRC 

in SBL-L-16071; License Amendment Request 16-03; Revise Current Licensing Basis to Adopt a Methodology for the Analysis 

of Seismic Category I; Structures with Concrete Affected by Alkali-Silica Reaction; Dated August 1, 2016 
Enclosure 4 to SBK-L-16013 

Seabrook Station License Renewal Application 

Section B.2.1.3 lA for Alkali-Silica Reaction (ASR) and 

Section B.2.1.3 lB for Building Deformation 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 4/ Page 2 

B.2.1.31A ALKALI-SILICA REACTION 

PROGRAM DESCRIPTION 

The Alkali-Silica Reaction Aging Management Program is a new plant specific program 

being implemented under the existing Maintenance Rule Structures Monitoring Program 

that will manage the aging effects related to Alkali-Silica Reaction of each structure and 

component subject to an Aging Management Review, so that the intended function(s) will 

be maintained consistent with the current licensing basis for the period of extended 

operation. 

Alkali-Silica Reaction 

Alkali-Silica Reaction (ASR) is an aging mechanism that may occur in concrete under 

certain circumstances. It is a reaction between the alkaline cement and reactive forms 

of silicate material (if present) in the aggregate. The reaction, which requires moisture 

to proceed, produces an expansive gel material. This expansion results in strains in the 

material that can produce micro-cracking in the aggregate and in the cement paste. The 

potential impact of ASR on the structural strength and anchorage capacity of concrete 

is a consequence of strains resulting from the expansive gel. These strains produce the 

associated cracking. Because the ASR mechanism requires the presence of moisture in 

the concrete, ASR has been predominantly detected in groundwater impacted portions 

of below grade structures, with limited impact to exterior surfaces of above grade 

structures. 

alkali cement + 

reactive aggregate 

Impact of Confinement 

expansive gel 

ASR Expansion Mechanism 

cracking of the 

aggregate and paste 

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 pre-loading a bolted joint. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 4/ Page 3 

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 m original construction at Seabrook utilized an 

aggregate that was susceptible to ASR, which was not known at the time. Although 

the testing was conducted in accordance with the ASTM C289 standards, the standard 

was subsequently identified as limited in its ability to predict long term ASR for 

moderate to low reactive aggregates. ASTM C289 has since been withdrawn. 

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 inscope 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 

compressive strength and modulus of elasticity, and subjected to petrographic 

examinations. While the results showed that both compressive strength and modulus of 

elasticity had declined, the structures were determined to be within design basis and 

remain operable. 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 portions of below-grade structures, with limited impact to exterior surfaces of above 

grade structures. 

Large-Scale Testing Program 

The structural assessment 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 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 4/ Page 4 

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. 

The need for Seabrook-specific testing was driven by limitations in the publicly 

available test data related to ASR effects on structures. Most research on ASR has 

focused on the science and kinetics of ASR, rather than engineering research on 

structural implications. Although structural testing of ASR-affected test specimens has 

been performed, the application of the conclusions to a specific structure can be 

challenged by lack of representativeness in the data (e.g., small-scale specimens; poor 

test methods; different reinforcement ~onfiguration). The large-scale test programs 

undertaken by NextEra provided data on the limit states that were essential for 

evaluating seismic Category I structures at Seabrook Station. The data produced from 

these programs were a significant improvement from the data in published literature 

sources, because test data across the range of ASR levels were obtained using a 

common methodology and identical test specimens. The results were used to assess 

the structural limit states and to inform the assessment of design considerations. 

The large-scale test programs included testing of specimens that reflected the 

characteristics of ASR-affected structures at Seabrook Station. Tests were completed 

at various levels of ASR cracking to assess the impact on selected limit states. The 

extent of ASR cracking in the test specimens was quantified by measuring the 

expansion in the in-plane and through-thickness specimen dimensions. The in-plane 

dimension refers to measurements taken in a plane parallel to the underlying 

reinforcement bars. There was no reinforcement in the test specimen 

through-thickness direction (perpendicular to the in-plane direction). ASR expansion 

measurements were monitored throughout testing. The test programs assessed all 

relevant limit states except compression (i.e., flexure and reinforcement anchorage, 

shear, and anchor bolts and structural attachments to concrete). The results of the test 

program demonstrated that none of the assessed limit states are reduced by ASR when 

ASR expansion levels in plant structures are below those evaluated in the large-scale 

test programs. 

The effect of ASR on compressive strength was not assessed in the large-scale test 

program. An evaluation of compression using existing data from published literature 

sources was performed. The evaluation concluded that ASR expansion in reinforced 

concrete results in compressive load that should be combined with other loads in 

design calculations. However, ASR does not reduce the structural capacity of 

compression elements. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 4/ Page 5 

The specimens used in the large-scale test programs experienced levels of ASR that 

bound ASR levels currently found in Seabrook structures (i.e., are more severe than at 

Seabrook), but the number of available test specimens and nature of the testing 

prohibited testing out to ASR levels where there was a clear change in limit state 

capacity. Because there is not testing data for these more advanced levels of ASR, 

periodic monitoring of ASR at Seabrook is necessary to ensure that the conclusions of 

the large-scale test program remain valid - that the level of ASR does not exceed that· 

considered under the test programs. 

The overall conclusion from analyses of structural limit states is that limit state 

capacity is not degraded when small amounts of ASR expansion are present in 

structures. Presently, the ASR expansion levels in Seabrook structures are below the 

levels at which limit state capacities are reduced. 

One of the objectives of the test program was to identify effective methods for 

monitoring ASR. The program concluded that monitoring the in-plane and throughthickness expansion is effective for characterizing the significance of ASR in 

structures. A Combined Cracking Index (CCI) methodology based on crack width 

summation was shown to be effective for in-plane expansion monitoring. Snap ring 

borehole extensometers (SRBEs) provided accurate and reliable measurements for 

monitoring through-thickness expansion. 

Results from the large-scale testing program are also used to support evaluations of 

structures subjected to deformation. These evaluations are discussed in the Building 

Deformation Monitoring Program in LRA Section B.2.1.31 B. 

PROGRAM ELEMENTS 

The following provides the results of the evaluation of each program element against the 

10 elements described in Appendix A ofNUREG-1800 Rev. 1, "Standard Review Plan for 

Review of License Renewal Applications for Nuclear Power Plants". 

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 the License Renewal Structures Monitoring Program. License 

Renewal concrete structures within the scope of this program include: 

Category I Structures 

• Containment Building (including equipment hatch missile shield) 

• Containment Enclosure Building 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 4/ Page 6 

• 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 

• TankFarm 

• 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 

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 1-FP-TK-35-A, 1-FP-TK-35-B, 1-FP-TK-36-A, 1-FPTK-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 

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. 
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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 performance 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 

W alkdowns of the station are performed on a periodic basis (SMP walkdowns, 

Systems Walkdowns, 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. 

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 Alkali-Silica 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." 

Inspection of inaccessible areas of concrete will be performed during opportunistic 

or focused inspections for buried concrete performed under the Maintenance Rule 

every 5 years. The concrete materials used to produce the concrete placed in 

inaccessible areas were the same as the concrete materials used to produce the 

concrete placed in accessible areas. Thus, the performance and aging of inaccessible 

concrete would be the same as the performance and aging of accessible concrete. 

Since the concrete mix and aggregates used at Seabrook Station is consistent 

between structures it is assumed unless demonstrated otherwise that ASR can be 

present. 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. 
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Expansion 

For ASR-affected surfaces at Seabrook Station, NextEra will monitor the effects of 

ASR expansion by obtaining measurements in both the in-plane (X&Y directions) 

and through-thickness directions CZ-direction). Specifically, Seabrook will be 

monitoring the Combined Cracking Index (CCI) for in-plane expansion and 

extensometer measurements for through-thickness 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 or vertical sides of a 

section of the ASR-affected concrete surface of predefined dimensions. Seabrook 

uses a grid size of 20 inches by 30 inches. The CI in a given direction is converted 

and reported in units of mm/m. 

The Cis are used to establish the Combined Cracking Index (CCI). The CCI 

estimates expansion on a concrete surface using measurements of crack widths 

along a pre-determined length or grid. The CCI is calculated by summing the 

crack widths crossing all reference grid lines and dividing the result by the sum of 

all gridline lengths. Criteria used in assessment of expansion is expressed in terms 

of CCI and based on recommendations provided in MPR-3727, "Seabrook Station: 

Impact of Alkali-Silica Reaction on Concrete Structures and Attachments" and 

supported by the test programs. 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 twodimensional reinforcement mats provide confinement in the in-plane directions, 

and through-thickness expansion dominates. 

Data analysis from the large-scale test program has been· completed and thresholds 

established based on the test reports. The thresholds are based on the structure as a 

whole so if localized extensive ASR or macro cracking is experienced in particular 

areas of the structure, then the entire structure is assumed to be susceptible to 

similar degradation. The overall methodology for using in-plane and throughthickness expansion values for various aspects of the monitoring program is 

summarized as follows: 
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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-destructive 

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). 

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 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 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, 

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

Crack Width Summation 

Crack width summation is a simple methodology for initial assessment of ASRaffected 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. 

The engineering strain in a structural member at the time of crack initiation ( Ecr) is 

equivalent to the tensile strength of the concrete divided by the elastic modulus ( Ecr 

= O"t IE). 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 (a ::::::: Ecr + Cl). Figure 1 depicts a concrete 

specimen with rebar being put in tension resulting in cracking. 
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Concrete has little strain capacity; therefore, in ASR-affected concrete, the crack 

widths comprise most of the expansion (~L). 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 is 

minimal. 

Figure 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. The CCI is also used to measure 

the effects of associated rebar strain. 

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 inplane direction plateaus at low expansion levels, while expansion in the throughthickness direction continues to increase. Seabrook will install the extensometers 

in Tier 3 and other selected locations to measure expansion in the throughthickness 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 
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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. 

A correlation relating expansion to reduction in elastic modulus was developed 

from the large scale testing program data. The correlation 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 throughthickness expansion going forwards. The measurements are the parameter to be 

monitored. The elastic modulus will not be monitored going forward; Preinstrument expansion is calculated initially to establish expansion to date and is not 

repeated. 

Structural Limit States 

The applicable design codes provide methodologies to calculate structural 

capacities for the various limit states and loading conditions applicable to 

Seabrook Station. Each relevant limit state was evaluated using published 

literature and the results of the MPR/FSEL large-scale test programs that used 

specimens designed and fabricated to represent reinforced concrete at Seabrook 

Station. The following guidance applies for structural evaluations of ASRaffected concrete structures at Seabrook Station: 

• Flexure/Reinforcement Anchorage - Based on the MPR/FSEL large-scale 

test program results, structural evaluations should consider that there has been 

no adverse impact on flexural capacity and reinforcement anchorage 

(development length) performance, provided that through-thickness expansion is 

at or below bounding conditions of the large scale testing and expansion 

behavior is comparable to the test specimens. 

• Shear - Based on the MPR/FSEL large-scale test program results, structural 

evaluations should consider that there has been no adverse impact on shear 

capacity, provided that through-thickness expansion is at or below bounding 

conditions of the large scale testing and expansion behavior is comparable to the 

test specimens. 
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• Anchors and Embedments - Based on the MPRIFSEL large-scale test 

program results, structural evaluations should consider that there is no adverse 

effect to post-installed or cast in place anchor/embedment capacity, provided 

that in-plane expansions remain at or below limits established by large scale 

testing. Through-thickness expansion is not relevant for anchor/embedment 

capacity. 

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, frequently 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). 

In-Plane Expansion 

As previously discussed in Element 3, Seabrook uses the CCI methodology to 

monitor the expansion of ASR affected areas in the in-plane 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 
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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 most-severe 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 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 in-plane expansion and the through-thickness expansion is due to the 

reinforcement detailing and the resulting difference in confinement between the inplane 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. Similarly, for 

unreinforced concrete backfill, expansion occurs in all directions. 

Out-of- Plane Expansion 

The need for out-of-plane expansion monitoring is triggered by a CCI exceeding 1 

mmlm. 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). 
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Elastic Modulus 

To determine expansion to date at a location selected for 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. 

Concrete cores will be removed from all Tier 3 locations for material property 

testing. Cores removed for property testing will have the approximate dimensions 

of 4" diameter x 8" length and will be tested in accordance with ASTM C39 for 

Compressive Strength and C469 for Elastic Modulus. The cores will be taken 

perpendicular to the reinforcement mat. 

A visual examination of the cores will confirm there is no mid-plane crack or edgeeffect cracking. 

Snap-Ring Borehole Extensometer 

Seabrook will install Snap-Ring Borehole Extensometers (SRBEs) at the station to 

monitor through-thickness expansion. The Large Scale Testing 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 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 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 a core bore at each Tier 3 location. The elastic 

modulus will only be determined at the time of core removal to determine preinstrument expansion to date. Additionally, mid-plane or edge-effect cracking will 

be visually observed at the time of core removal. 
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ELEMENT 5 - Monitoring and Trending 

The progression of ASR degradation of the concrete is an important consideration for 

assessing the long term implications of ASR and specifying monitoring intervals. The 

most reliable means for establishing the progression of ASR degradation is to monitor 

expansion of the in situ concrete. 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. 

ELEMENT 6 - Acceptance Criteria 

Identification of the typical symptoms indicative of ASR generates the need to initially 

start monitoring the area using CCI. For the structures subject to ASR monitoring, 

rebar strain as a result of ASR induced stresses and ASR induced stresses in 

combination with design bases loads will _be verified to be within code allowable 

limits. 

In-Plane 

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). Deficienyies 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 Yi year ( 6-

month) inspection frequency and added to the through-thickness expansion 
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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. Tier 1 structures do not display signs of ASR and are monitored 

consistent with the Structures Monitoring Program. 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. 

Recommendation for Structural Monitoring Tier Individual Concrete Criteria Program Category Components 

Structural Evaluation 

Implement enhanced ASR 

3 Unacceptable (requires monitoring, such as 1.0 mm/m or greater Combined 

further evaluation) through-wall expansion Cracking Index (CCI) 

monitoring using 

Extensometers. 

0.5 mmlm or greater CCI 

Quantitative Monitoring and CI of greater less than 0.5 

Trending mmlm in the vertical and 

· horizontal directions. 

2 Acceptable with 

Any area with visual presence of Deficiencies 

ASR (as defined in FHWA-HIFQualitative Monitoring 12-022) accompanied by a CI of 

less than 0.5 mmlm in the vertical 

and horizontal directions. 

Routine inspection as Area has no indications of pattern 

1 Acceptable prescribed by the Structural cracking or water ingress. No 

Monitoring Program visual symptoms of ASR. 
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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). 

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 

have been established by the large scale test program test reports, and are 

summarized in the Table below. Maintaining these limits is assured by periodically 

measuring through-thickness expansion in areas affected by ASR. 

Effect of ASR on Structural Limit States 

Structural Design Issue Criteria2 

Flexure & reinforcement anchorage See FP#101020 - Section 2.1 

Shear See FP#101020 - Section 2.1 

Anchor bolts and structural See FP#101020 - Section 2.1 

attachments 

ELEMENT 7 - Corrective Actions 

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. The NextEra Energy Quality Assurance Program and Nuclear Fleet procedures 

will be utilized to meet Element 7 Corrective Actions. 

ELEMENT 8 - Confirmation Process 

The FPL/NextEra Energy Quality Assurance Program and Nuclear Fleet procedures will 

be utilized to meet Element 8 Confirmation Process. 

ELEMENT 9 - Administrative Controls 

The FPL/NextEra Energy Quality Assurance Program and Nuclear Fleet procedures will 

be utilized to meet Element 9 Administrative Controls. 

2 Expansion Limit Criteria is considered proprietary to NextEra Seabrook. FP #101020 MPR-4288, Revision 0, "Seabrook 

Station: Impact of Alkali-Silica Reaction on the Structural Design Evaluations," July 2016, was previously submitted to the NRC 

in SBL-L-16071; License Amendment Request 16-03; Revise Current Licensing Basis to Adopt a Methodology for the Analysis 

of Seismic Category I; Structures with Concrete Affected by Alkali-Silica Reaction; Dated August 1, 2016 
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SBK-L-16013 /Enclosure 4/ Page 18 

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. 

Newly Identified Operating Experience (OE) 

Seabrook will update the Aging Management Program for any new plant-specific or 

industry OE. This includes ongoing industry studies performed both nationally and 

internationally. Research data taken from these studies will be used to enhance the 

ASR program, if applicable. In addition NextEra Seabrook has submitted a License 

Amendment Request to the Commission in accordance with 10CFR50.90 to 

incorporate a revised methodology related to ASR material properties and building 

deformation analysis for review and approval. NextEra will incorporate changes 

related to this LAR submittal as necessary to maintain alignment of the aging 

management program to the current license basis. 

Groundwater Operating Experience 

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. Some of these areas have 

shown signs of leaching, cracking, and efflorescence on the concrete due to the 

infiltration. During the early 1990's an evaluation was conducted to assess the effect 

of the groundwater infiltration on the serviceability of the concrete walls. That 

evaluation concluded that there would be no deleterious effect, based on the design 

and placement of the concrete and on the non-aggressive nature of the groundwater. 

In 2009, NextEra tested seasonal groundwater samples to support the development of 

a License Renewal Application. The results showed some of the groundwater to be 

aggressive. Ground water testing performed in November 2008 and September 2009 

found pH values between 6.01 and 7.51, chloride values between 19 ppm and 3900 

ppm, and sulfate values between 10 ppm and 100 ppm. Aggressive chemical attack 

becomes a concern when environmental conditions exceed threshold values 

(Chlorides> 500 ppm, Sulfates >1500 ppm, or pH< 5.5). Based on determination of 

aggressive ground water and observed efflorescence on the concrete surface, NextEra 

initiated a comprehensive review of possible effects to concrete of in-scope 

structures. 
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ASR Identification OE 

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. 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 the expected value 

except for the Diesel Generator, Containment Enclosure Buildings, Emergency 

Feedwater Pumphouse, and the Equipment Vaults, which were less than the expected 

value in localized areas. 

Evaluation of the 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. 

EXCEPTIONS TO NUREG-1800 

None 
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ENHANCEMENTS 

• Implement the Alkali-Silica Reaction (ASR) Monitoring 

• Revise the Seabrook Structural Monitoring Procedure EDS 36180 to include Alkali-silica 

reaction description, aging effects, inspection criteria, acceptance criteria. 

• Revise the Seabrook ASME Section XI, Subsection IWL Program ES 1807 .031 to include 

Alkali-silica reaction aging effects. 

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 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 CCL 

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. 

The Seabrook Station ASR Monitoring Program provides reasonable assurance that the 

effects of aging of in-scope concrete structures due to the presence of Alkali-Silica reaction 

will be managed to ensure the structures continue to perform their intended function 

consistent with the current licensing basis for the period of extended operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 4/ Page 21 

B.2.1.31B BUILDING DEFORMATION 

PROGRAM DESCRIPTION 

The Building Deformation Monitoring Program is a new plant specific program being 

implemented under the existing Maintenance Rule Structures Monitoring Program. Building 

Deformation is an aging mechanism that may occur as a result of other aging effects of 

concrete. Building Deformation at Seabrook is primarily a result of the alkali silica reaction 

(ASR) described in LRA section B.2.1.3 lA but can also result from swelling, creep, and 

. shrinkage. Building deformation can cause components within the structures to move such 

that their intended functions may be impacted. 

The Building Deformation Monitoring Program uses visual inspections associated with the 

Structures Monitoring Program and cracking measurements associated with the Alkali-Silica 

Reaction program to identify buildings that are experiencing ·deformation. The first 

inspection is a baseline to identify areas that are exhibiting surface cracking. The surface 

cracking will be characterized and analytically documented. This inspection will also identify 

any local areas that are exhibiting deformation. The extent of surface cracking will be input 

into an analytical model. This model will determine the extent of building deformation and 

the :frequency of required visual inspections. 

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 or components are required. Specific criteria for selecting effective 

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

PROGRAM ELEMENTS, 

The following provides the results of the evaluation of each program element against the 10 

elements described in Appendix A of NUREG-1800 Rev. 1, "Standard Review Plan for 

Review of License Renewal Applications for Nuclear Power Plants". 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 4/ Page 22 

ELEMENT 1 - Scope of Program 

The Seabrook Building Deformation Monitoring Program provides for management of 

the effect of building deformation on Seismic Category 1 structures and ·associated 

components within the scope of license renewal. Program scope includes components 

within the scope of license renewal contained in concrete structures within the scope of 

the Structures Monitoring Program. Concrete 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 

• TankFarm 

• 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 

e Circulating Water Pump House Building (below elevation 21 '-0) 

• Safety Related Electrical Manholes and Duct Banks 

• Pre-Action Valve Building 

ELEMENT 2 - Preventive Actions 

There are no preventive actions specified in the Seabrook Station Structures Monitoring 

Program, which includes implementation of NUREG-1801 XI.S5, XI.S6, and XI.S7. 

These are monitoring programs only. Similarly, the Building Deformation Monitoring 

Program does not rely on preventive actions. 

ELEMENT 3 - Parameters Monitored/Inspected 

A three step process will be used to initially screen for deformation and to analyze the 

effects on structures for the self-straining loads from ASR expansion, creep, shrinkage, 

and swelling. Each stage of the process would have increasing levels of rigor. 
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SBK-L-16013 /Enclosure 4/ Page 23 

Stage One- Susceptibility Screening Evaluation: Each of the seismic Category I 

structures are screened for susceptibility to structural deformation caused by ASR 

using existing field data and conservative hand calculations. 

Stage Two - Analytical Evaluation: An analytical evaluation is performed for 

structures that the Stage One Susceptibility Screening Evaluation identifies as 

susceptible to deformation, but do not satisfy ACI 318-71 acceptance criteria. A 

finite element model of the structure is used to estimate structural demands due to 

self-straining loads, while all other demands are taken from existing design 

calculations. Additional field data is obtained to provide input to the analysis. The 

evaluation verifies compliance with ACI 318-71 using the same criteria as the 

original design. 

Stage Three - Detailed Evaluation: A detailed design confirmation calculation is 

performed when the Stage Two Analytical Evaluation concludes that some area of a 

structure does not satisfy ACI 318-71 acceptance criteria or when the structure has 

sufficient deformation that may impact demands computed in the original design. 

The detailed evaluation uses the Stage Two finite element model to compute demands 

due to self-straining loads as well as all other design loads. In the Stage Three 

evaluation, consideration is given to cracked section properties, self-limiting 

secondary stresses, and the redistribution of structural demands when sufficient 

ductility is available. 

All three stages of the evaluation process use the original design acceptance criteria 

given in the UFSAR Chapter 3 including separation of structures by seismic gaps. 

Each analysis stage will determine threshold monitoring limits to define the 

monitoring frequency and criteria for re-evaluating structures with deformation. The 

threshold monitoring limits are described below. 

Stage One: Susceptibility Screening Evaluation 

N extEra has conducted walkdowns of selected in scope structures and plant 

equipment to identify items of interest and evaluate the items through the Corrective 

Action Program for their impact on plant operations. NextEra will perform future 

walkdowns for all in scope structures. Inspection data from these walkdowns and 

other measurements obtained for ASR-affected structures will be reviewed to 

determine if deformation is occurring and to identify potential locations and 

directions of movement or deformation. The data that will be collected includes 

measurements of relative building movements, equipment misalignments, and 

concrete cracking indexes. ASR monitoring grids, which are used to measure the 

strain in reinforced concrete, were installed on structures throughout the facility. The 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 4/ Page 24 

monitoring grids were installed at the most severe locations for ASR cracking, and 

therefore, provide a conservative estimate of the strain in the structure. After 

re"."iewing existing field data, a walkthrough inspection will be performed to verify 

field conditions and determine if ASR expansion only affected localized regions of 

the structure or whether the structure has experienced global deformation of structural 

members. 

In the susceptibility screening process, conservative estimates of deformations and 

strains based on the field data are used to estimate demands caused by self-straining 

loads for critical locations in the structure. Self-straining loads include four 

components: ASR, creep, shrinkage, and swelling. Based on guidance in ACI 318-

71, creep, shrinkage and swelling are included with the dead load. The ASR demands 

(identified as "Sa" herein) are factored and then combined with demands due to 

design loads for critical load combinations in the current licensing basis. An 

evaluation is performed using strength acceptance criteria as described in the current 

licensing basis. 

For screening evaluations that conclude a structure fully complies with the strength 

acceptance criteria, the critical locations of the structure are re-evaluated for a higher 

level of ASR demand to determine the maximum allowable, factored self-straining 

loads at which the structure meets the design acceptance criteria. A set of monitoring 

elements (consisting of strain measurements, deformation measurements, seismic gap 

measurements, and/or other quantifiable behaviors) is established along with 

threshold limits for each monitoring element. The threshold limits are defined as the 

maximum measurement for each monitoring element that results in a factored selfstraining load equal to the factored self-straining load at the structural design limit 

(with factored design basis loads included). The threshold limit for the monitoring 

elements defined in Stage One is equal to the set of monitoring element 

measurements that produce a factored ASR demand that is 90% of the factored selfstraining load at the acceptance limit. If a structure monitoring element measurement 

obtained from walkdowns and other monitoring activities exceeds the monitoring 

threshold limit, then a Stage Two Analytical Evaluation is required. 

A structure is classified ·as Stage 1 if the Susceptibility Screening Evaluation 

concludes that the structure satisfies the strength acceptance criteria and the structure 

monitoring element measurements are less than the Stage One threshold limits. The 

Susceptibility Screening Evaluation for Stage 1 structures is summarized in a 

calculation package that supplements the original design calculation. The calculation 

package also documents the set of monitoring measurements and the threshold limits 

for the monitoring process. The monitoring measurements and the threshold limits are 

incorporated into the Seabrook Structures Monitoring Program to periodically assess 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 4/ Page 25 

the condition of structures and verify that the structure meets the design acceptance 

criteria. 

Stage 1: 

Screening 

Evaluation 

Original Design 

Demands & 

Capacities 

Field Observations 

Yes 

Define Threshold 

for Monitoring 

Stage Two: Analytical Evaluation 

Evaluate 

Responses due to 

ASR loads (S ) 

a 

Stage 2: Analytical 

Evaluation 

For structures that cannot be shown to meet the ACI 318-71 acceptance criteria using 

the conservative methods of the Susceptibility Screening Evaluation or monitoring 

measurements indicate high Sa demands, an Analytical Evaluation is required. The 

Analytical Evaluation uses more accurate methods to quantify demands due to selfstraining loads. Also, additional inspections are performed to measure structural 

strains and deformations at a broader range of critical locations of the structure. 

These measurements would be used to compute the self-straining loads with more 

accuracy than possible using the inputs from the Susceptibility Screening Evaluation 

process. 

An ANSYS finite element model (FEM) of the structure is created based on design 

drawings and uncracked design section properties. The model is initially 

benchmarked to the original design analysis of the structure with only the current 

licensing basis loads. The FEM is then calibrated such that the deformations and 

strains due to unfactored sustained loads and self-straining loads are consistent with 

field measurements. The FEM is used to compute the structural demands due to ASR 

loads (Sa). The self-restraining demands from finite element analysis are factored and 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 4/ Page 26 

then combined with demands due to factored design loads from the original design 

calculations for the load combinations described in the current licensing basis. The 

structural demand in critical regions of the structure are evaluated using strength 

acceptance criteria described in the current licensing basis. The methods used for the 

Stage Two analysis are unchanged from the original design analyses with the 

exception of accounting for the self-straining loads in the analysis and the use of the 

ANSYS software program for computing the sustained and self-straining loads. 

Structures that satisfy the Analytical Evaluation acceptance criteria are re-evaluated 

for a higher level of Sa to compute the maximum allowable self-straining loads on the 

structure. The maximum allowable loads correspond to the maximum, factored selfstraining loads at which the structure meets the design acceptance criteria. A set of 

monitoring measurements are identified and threshold limits are set for each 

measurement based on the maximum allowable self-straining load. The threshold 

limits for each monitoring element defined in Stage Two are determined by scaling 

all measurements proportionally such that a factored self-restraining demand equal to 

95% of the value at the design acceptance limit is achieved. 

A structure is classified as Stage 2 if the Analytical Evaluation concludes that the 

structure satisfies the strength acceptance criteria and the structure monitoring 

element measurements are less than the Stage Two threshold limits. The Analytical 

Evaluation calculation for Stage 2 structures supplements the original design 

calculation. The monitoring measurements, measuring locations, and threshold limits 

for monitoring are also included in the supplement to the calculation. 

Monitoring elements may include strain measurements, measurements of the relative 

displacement between structures, component specific measurements (e.g. gap 

measurements) and other quantifiable parameters. The threshold limits are defined as 

the maximum allowable measurement for each monitoring parameters that limits the 

self-straining loads to some fraction of the maximum allowable self-straining load. 

The monitoring measurements and the threshold limits are used to periodically assess 

the condition of structures and verify that the structure meets the design acceptance 

criteria. 
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SBK-L-16013 /Enclosure 4/ Page 27 

"S " Inputs Based on a 

Field Measurements and 

other self straining loads 

Original Design Demands 

1 & Capacities 

Stage 2: Analytical Evaluation 

Finite Correlate with Field 

Element I - -· -,, Observations 

Modeling 

Calculate S a 

/''~ 

~'~ Adequate ) Margin with "S II 

a 

/ 

Yes 

Define Threshold for 

Monitoring 

No Stage 3: 

Detailed 

Evaluation 

Stage Three: Detailed Evaluation 

Structures that do not meet the acceptance criteria of the Stage Two Analytical 

Evaluation are analyzed by a Detailed Evaluation. In the Detailed Evaluation, Sa 

demands and the loads from creep, shrinkage and swelling are recomputed using the 

Stage Two FEM. Structural demands due to design loads are recomputed by applying 

design demands (i.e. wind, seismic, hydrostatic pressure, etc.) to the FEM. A detailed 

structural evaluation is performed for all load combinations described in the licensing 

basis. The structure is evaluated using strength acceptance criteria described in the 

current licensing basis. Consideration is given to force and moment redistribution in 

regions with localized overstresses and sufficient ductility. In the Stage Three 

evaluation, consideration is given to cracked section properties, self-limiting 

secondary stresses, and the redistribution of structural demands when sufficient 

ductility is available. The 100-40-40 percent rule in NRC Regulatory Guide 1.92, 

Revision 3, is used as an alternative to the SRSS method for combining three 

directional seismic loading in the analysis of seismic, Category I structures that are 

deformed by the effects of ASR. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 4/ Page 28 

Structures that meet the acceptance criteria of the Detailed Evaluation are reevaluated for a higher level of self-straining load to establish the threshold limits for 

each monitoring element measurement. A similar process is used as described in the 

Stage Two Analytical Evaluation above. The threshold limit for each monitoring 

element defined in Stage Three is equal to the limit for the monitoring element 

measurement that produces a factored Sa load at the design acceptance limit. 

The Detailed Evaluation is summarized in a design calculation package that will 

supersede the original design calculation. The calculation package documents the set 

of monitoring measurements and the threshold limit of the monitoring measurements 

for the structural monitoring program to verify that the structure to meets the design 

acceptance criteria. 

Original Design 

Loads Inputs 

Example 

Stage 3: Detailed Evaluation 

Finite Element 

Modeling 

.I. 

. Evaluate Using Total Factored Design 

Demands Including "S " 

a 

• Cracked section 

• Redistribution of Self-limiting 

load 

• 100-40-40 Combination of 

seismic components 

Define Th reshold for Monitoring 

Refined "S " Inputs Based 

on Additional Field 

Measurements 

An example of the process of determining threshold limits for subsequent monitoring 

(if required for a particular structure) is described below. For the containment 

enclosure building, seismic gap measurements and annulus width measurements can 

be used to monitor deformation of the structure. Specific locations are chosen and 

threshold limits are set for these locations to ensure license renewal intended 
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SBK-L-16013 /Enclosure 4/ Page 29 

functions are met. The calculation of these threshold limits is defined and evaluated 

using the following equations: 

Where: 

TM::; TL n 

TM = LI dn,field - dn,design I X (~) i=O 

n 

TL = L [I dn,baseline - dn,design I X kn,thf ] X (~) i=O 

d 

k _ n,FEA,1.2 

n,thf - d 

n,FEA,baseline 

TM = Average deformation for locations in threshold measurement set 

TL= Threshold Limit 

n =Number of measurement locations in threshold measurement set 

dn,field = Field measurement of threshold measurement n at time of monitoring 

dn,design =Design dimension of threshold measurement n 

dn,baseline = Field measurement of threshold measurement n at time when TL is 

established and CEB evaluation is performed 

dFEA,i.2 = Radial deformation of the CEB at location of threshold measurement n 

due to unfactored sustained loads plus unfactored self-straining loads with a 1.2 

threshold factor 

dn,FEA,baseline = Radial deformation of the CEB at location of threshold 

measurement n due to unfactored sustained loads plus unfactored self-straining 

loads without threshold factor amplification 

For each threshold measurement, a method will be established to perform the 

measurement in a repeatable way. It is particularly important to perform the 

measurement in a well-defined location; otherwise, seemingly small deviations in the 

concrete surfaces can have a significant impact on the repeatability of the threshold 

measurements. For some of the locations in the containment enclosure building, a 

repeatable measurement method has already been established and a baseline 

measurement has been obtained. Other locations have been measured in the past, but 

have not been measured in a suitably repeatable way for continued monitoring. Once 

a suitable baseline measurement is established for all locations in the each structure, 

then the average threshold limit can be computed. An example projected value of the 

threshold limit is provided in Table below. It should be noted that the values in the 

table are presented as an example and are not intended to be applicable to actual 

locations. 
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Example Threshold Limits 

Measurement ID 1 2 3 4 

Measurement Type Seismic Seismic Annulus Annulus 

Gap Gap Width Width 

Measurement Azimuth 180 305 220 240 

Measurement Elevation +5.5 ft +21 ft +9 ft +9 ft 

Relative-to Structure CB Personnel CB CB Hatch 

Direction of deformation Inward Inward Inward Inward 

Measurement taken from 

Inside or Outside of Inside Outside Inside Inside 

Annulus 

Baseline Measurement April April 2016 April 2016 April 2016 Date and Report Reference 2016 

r:xx1 [XX] [XX] [XX] 

dn,baseline , in. 1.5 1.97 51.00 52.00 

dn,design , in. 3.0 3.0 54.00 54.00 

Baseline Measurement, in. 

1.50 1.03 3.00 2.00 I dn,b~eline - dn,desi.an I 

dn,FEA,baseline , in. -0.34 -0.65 -I.OJ -0.41 

dn,FEA 1.2 , in. -0.34 -0.69 -1.18 -0.48 

kn,thf 1.01 1.05 1.17 1.17 

Local Threshold Limit, in. 

I dn,baseline - dn,design I 1.52 1.08 3.51 2.34 

X kn,thf 

Average of Baseline Measurements 1.88 in. 

Threshold Limit (based on projected baseline 2.11 in. values) 

Summary 

In summary, the process will classify affected structures into one of three categories: 

(1) structures with minimal amounts of deformation that do not affect the structural 

capacity as determined in the original design analysis; (2) structures with elevated 

levels of deformation that are shown to be acceptable using Finite Element Analysis 

(FEA) but still meeting the original design basis requirements when ASR effects are 

included; and (3) structures with significant deformation that are analyzed and shown 

to meet the requirements of the code of record using the methods described herein. 

This approach is consistent with guidance in ACI 349.3R-1996 used to establish the 

inspection criteria for the Structures Monitoring Program. The ASR deformation 

categories do not necessarily correspond to the criteria used to characterize ASR 

cracking in structures that is discussed in LRA section B.2.1.3 IA. That is, a Stage 2 

structure does not necessarily have ASR cracking that is classified as Tier 2. 

Structures will be monitored based on the most limiting parameter for monitoring 

from either the ASR Monitoring Program or the Building Deformation Monitoring 

Program. The building deformation monitoring frequency for structures for each 

Stage are summarized in Table 1. 
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Table 1 Structure D~formation Monitoring Requirements 

Stage Deformation Evaluation Stage Monitoring Interval 

1 Screening assessment 3 years 

2 Analytical Evaluation 18 months 

3 Detailed Evaluation 6 months 

The monitoring frequencies in Table are based on guidelines developed for 

inspecting transportation structures with ASR degradation. The guidance 

recommends inspections from six months to 5 years depending on the age of the 

damage to the structure and the rate of change in degradation. The interval for 

recording monitoring elements for deformation for each structure can be increased to 

the interval in the next lower Stage (i.e., Stage 3 to Stage 2 and Stage 2 to Stage 1) if 

no change in measurements are observed for 3 years. Stage 1 structures that have 

shown no change in deformation for 10 years may increase the inspection interval to 

once every 5 years. Structures that show no evidence of building deformation will 

continue to be inspected with a frequency as established by the Structures Monitoring 

Program. 

Components Impacted by Structural Deformation 

With deformation, an aging effect of concern is component functionality and 

structural interferences. Condition walkdowns are performed with a focus on safetyrelated 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 

are not limited to, the following: 

• Distorted flexible couplings 

• Non-parallel pipe/conduit/HY 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 

• Misaligned pipes in penetrations 

• Roof membranes and weather seals degraded 

• Electrical box, panel, or fitting distorted 
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SBK-L-16013 /Enclosure 4/ Page 32 

Component specific features may indicate irreversible deformation of the affected 

component or irreversible plastic deformation of the structure such as rebar yielding 

or rebar slip. If these features are observed, then they will be documented in the 

corrective action process so that future monitoring walkdowns will observe the same 

features. Inspections of these features are in addition to the installed monitoring 

elements such as strain measurements and measurements of the relative deformation 

between structures. All of these measurements will be performed at the frequency 

described in Table 1. 

The walkdowns will be performed in accordance with the Structures Monitoring 

Program and ASME Section XI, Subsection IWL Program documents. Seabrook will 

update the walkdown guidance documents as necessary to accommodate new 

Operating Experience (OE) identified during the walkdowns. 

ELEMENT 4 - Detection of Aging Effects 

As discussed in Element 3 baseline walkdowns to identify the potential effects caused by 

building deformation will be performed. The results of the baseline walkdowns will be 

used to determine the key assumptions in the structural analysis in addition to 

determining the monitoring frequencies for equipment impacted by building deformation. 

Subsequent monitoring will be performed as part of future Structural Monitoring 

Program (SMP) walkdowns. The inspection frequencies identified by Table 1 will be 

applied in locations where symptoms of deformation are identified; otherwise, the 

inspection frequency will follow the requirements of the SMP. 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. 

ELEMENT 5 - Monitoring and Trending 

Once the inspection frequencies are determined as described by Element 3, visual 

inspections will be used to monitor and trend future building deformation. Any new 

indications of building deformation will be placed in the corrective action program, and 

evaluations will be performed to determine if inspection frequencies should be changed 

to ensure that future effects of degradation would be identified before loss of 

components' intended function. 

ELEMENT 6 - Acceptance Criteria . 

A systematic approach to evaluation of structures and components impacted by ASR expansion 

and building deformation is utilized. A structural model is developed where ASR induced 

expansion is applied to the structure developing force, moments, and displacements that are 

attributed to the effects of ASR. The added load due to ASR is then combined with other CLB 
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SBK-L-16013 /Enclosure 4/ Page 33 

loads (deadweight, wind, hydrostatic, seismic, etc.). Resultant load combinations are then 

evaluated to validate compliance with structural design code requirements 

Specific quantitative criteria to ensure component-intended functions will be maintained under all 

design conditions and is condition and location specific. Field observations of 

distorted/misaligned components and local structural deformation indicated by strain 

measurements or relative building movements are evaluated utilizing the existing acceptance 

criteria or design code specified for the design function of the component. 

NextEra will determine appropriate criteria based on the walkdown results and the particular 

geometry and configuration in the area of interest. The criteria will include margin to trigger 

action prior to loss of intended function whether that action is an additional inspection or 

repair/replacement of the component. 

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 

The FPL/NextEra Energy Quality Assurance Program and Nuclear Fleet procedures will be 

utilized to meet Element 8 Confirmation Process. 

ELEMENT 9 - Administrative Controls 

The FPL/NextEra Energy Quality Assurance Program and Nuclear Fleet procedures will be 

utilized to meet Element 9 Administrative Controls. 

ELEMENT 10 - Operating Experience 

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 an 

engineering firm to perform visual assessments of accessible areas surrounding the 
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SBK-L-16013 /Enclosure 4/ Page 34 

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 the CEB. Parametric studies using a linear finite element model of the 

CEB with boundary conditions modeling parameters appropriate for estimating 

structural deflections and deformed shapes were performed. 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 (FEA) were generally similar to field measurements. The 

results of the FEA 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 were performed to identify the above symptoms that may have 

been missed during the Structural Monitoring Program Walkdowns that were 

conducted prior to this discovery. The items identified were entered Seabrook's 

Corrective Action Program. 

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. 

Plant Specific Operating Experience 

AR 02044627 notes that the as-measured width of seismic isolation gaps is less than 

the nominal value of 3 inches specified on concrete drawings for isolation between 
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SBK-L-16013 /Enclosure 4/ Page 3 5 

structures. There are a total of 93 as-measured gaps less than 3 inches between the 

following abutting structures: Containment Building, Containment Enclosure 

Building, Mechanical Penetration Area, West Main Steam and Feed Water Pipe 

Chase, Electrical Penetration Area and Emergency Feed Water Pump House. Initial 

finite element analysis completed determined that the deformation is attributed to 

ASR expansion and creep. The compensatory measure implemented requires 

measuring seismic isolation gaps every six months. 

AR 2114299 documents that a seismic isolation joint located on an expansion boot 

near ductwork in the Containment Enclosure Building is vertically misaligned by 

approximately 2". The boot appeared to be in good shape and not dry or cracking. 

The AR determined that the cause of the misalignment is building deformation of the 

Containment Enclosure Building. The engineering evaluation concluded that the 

displaced ducts resulted in some slipping of the expansion joint material relative to 

the clamp at the areas of highest relative movement and that there is reasonable 

assurance that the joint material would most likely slip rather than tear or elongate 

during a seismic event. The condition was found acceptable as is and no loss of 

intended function was identified. 

AR 02107225 documents a deformed and misaligned flexible coupling on a conduit 

located in the West Pipe Chase area. Based on a field walkdown, the coupling was 

misaligned by 1. 7 5" which is greater than the established 1.25" acceptable limit. The 

cause of the misalignment was building deformation. Therefore, engineering analysis 

was performed to ensure that the enclosed cable can continue to perform its safety 

function. Even though the cable could continue to perform its safety function, the 

flexible conduit was repaired to restore design margin. 

AR 02129621 documents the seismic gap between Containment and the CEB 

horizontal cantilevered concrete shield block at Azimuth 23 0 elevation 22' is less the 

the minimum required seismic gap of .277 inches. The cause of the reduced gap was 

building deformation. An engineering analysis was performed to ensure that the 

structural remains operable while steps are taken to restore to design requirements. 

Newly Identified Operating Experience (OE) 

Seabrook will update the Aging Management Program for any new plant-specific or 

industry OE. This includes ongoing industry studies performed both nationally and 

internationally. Research data taken from these studies will be used to enhance the 

Building Deformation Monitoring Program, if applicable. In addition NextEra 

Seabrook has submitted a License Amendment Request to the Commission in 

accordance with 1 OCFR50.90 to incorporate a revised methodology related to ASR 

material properties and building deformation analysis for review and approval. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 4/ Page 3 6 I 

N extEra will incorporate changes related to this LAR submittal as necessary to 

maintain alignment of the aging management program to the current license basis. 

EXCEPTIONS TO NUREG-1800 

None 

ENHANCEMENTS 

Implement the Building Deformation (BD) Monitoring Program 

The Seabrook Structural Monitoring Procedure EDS 36180 will be revised to include 

building deformation aging effects, inspection criteria, and acceptance criteria. 

CONCLUSION 

To manage the aging effects of building deformation due to ASR, swell, creep, and 

expansion, the existing Structures Monitoring Program and ASME Section XI, Subsection 

IWL Program, have been augmented by this plant specific Building Deformation Monitoring 

Program. This program will perform baseline inspections to determine the extent of 

deformation, input the inspection results into an analytical model, and use this model to 

determine the projected rate of future deformation, and set inspection frequencies both to 

ensure that the calculated deformation rate is valid, and the established monitoring 

frequencies ensure that intended functions for structures and components will be maintained. 
Enclosure 5 to SBK-L-16013 

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-16013 /Enclosure 5/ Page 2 

A.3 LICENSE RENEW AL COMMITMENT LIST 

No. PROGRAM or TOPIC COMMITMENT 

Provide confirmation and acceptability of the 

1. PWR Vessel Internals implementation ofMRP-227-A by addressing the plantspecific Applicant/Licensee Action Items outlined in section 

4.2 of the NRC SER. 

Enhance the program to include visual inspection for 

2. Closed-Cycle Cooling Water cracking, loss of material and fouling when the in-scope 

systems are opened for maintenance. 

Inspection of Overhead Heavy . Enhance the program to monitor general corrosion on the Load and Light Load (Related 

3. to Refueling) Handling crane and trolley structural components and the effects of 

Systems wear on the rails in the rail system. 

Inspection of Overhead Heavy 

4. 

Load and Light Load (Related Enhance the program to list additional cranes for monitoring. to Refueling) Handling 

Systems 

Enhance the program to include an annual air quality test 

5. Compressed Air Monitoring requirement 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 

7. Fire Protection spalling, and loss of material caused by freeze-thaw, 

chemical attack, and reaction with aggregates by 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. 

A.2.1.13 Prior to the period of extended 

operation. 

Prior to the period of extended A.2.1.14 operation. 

A.2.1.15 Prior to the period of extended 

operation. 

A.2.1.15 Prior to the period of extended 

operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 51 Page 3 

8. Fire Protection 

9. Fire Water System 

10. Fire Water System 

11. Fire Water System 

Enhance the program to include the performance of visual 

inspection of fire-rated doors by a fire protection qualified 

inspector. 

Enhance the program to include NFPA 25 (2011 Edition) 

guidance for "where sprinklers have been in place for 50 

years, they shall be 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 testing of the fire water system in accordance with the 

guidance ofNFP 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. 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 

Prior to the period of extended A.2.1.15 operation. 

Prior to the period of extended A.2.1.16 operation. 

Prior to the period of extended A.2.1.16 operation. 

A.2.1.16 Within ten years prior to the 

period of extended operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 5/ Page 4 

12. Aboveground Steel Tanks 

13. Fire Water System 

14. Fuel Oil Chemistry 

15. Fuel Oil Chemistry 

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 near-atmospheric internal 

pressures, sit on concrete or soil, and exposed internally to 

water, 3) Visual, surface, and volumetric examinations of the 

outside and inside surfaces for managing the aging effects of 

loss of 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. 

Enhance the program to perform exterior inspection of the 

fire water storage tanks annually for signs of degradation and 

include an ultrasonic inspection and evaluation of the 

internal bottom surface of the two Fire Protection Water 

Storage Tanks per the guidance provided in NFP A 25 (2011 

Edition). 

Enhance program to add requirements to 1) sample and 

analyze new 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 of water in the Auxiliary Boiler fuel oil storage tank 

at least once per quarter and to remove water as necessary. 

Within 10 years prior to the A.2.1.17 period of extended operation. 

A.2.1.16 Within ten years 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. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 5/ Page 5 

16. Fuel Oil Chemistry 

17. Fuel Oil Chemistry 

18. Reactor Vessel Surveillance 

19. Reactor Vessel Surveillance 

20. Reactor Vessel Surveillance 

21. Reactor Vessel Surveillance 

Enhance the program to require draining, cleaning and 

inspection of 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 IO-year draining, 

cleaning and inspection of the Diesel Generator fuel oil 

storage tanks, Diesel Generator fuel oil day tanks, diesel fire 

pump fuel oil day tanks and auxiliary boiler fuel oil storage 

tank. 

Enhance the program to specify that all pulled and tested 

capsules, 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 Program, such as operating at a lower 

cold leg temperature or higher fluence, the impact of plant 

operation changes on the extent of Reactor Vessel 

embrittlement will be evaluated and the NRC will be 

notified. 

Enhance the program as necessary to ensure the appropriate 

withdrawal schedule for capsules remaining in the vessel 

such that one capsule will be withdrawn at an outage in 

which the capsule receives a neutron fluence that meets the 

schedule requirements of 10 CFR 50 Appendix H and 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 intent to test it, is stored in a manner which 

maintains it in a condition which would permit its future use, 

including during the period of extended operation. 

Prior to the period of extended 

A.2.1.18 operation. 

A.2.1.18 

Prior to the period of extended 

operation. 

Prior to the period of extended A.2.1.19 operation. 

Prior to the period of extended A.2.1.19 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-16013 /Enclosure 5/ Page 6 

22. One-Time Inspection 

Selective Leaching of 

23. Materials 

Buried Piping And Tanks 

24. Inspection 

One-Time Inspection of 

25. ASME Code Class 1 Small 

Bore-Piping 

26. External Surfaces Monitoring 

Inspection of Internal Surfaces 

27. in Miscellaneous Piping and 

Ducting Components 

28. Lubricating Oil Analysis 

29. Lubricating Oil Analysis 

30. Lubricating Oil Analysis 

Implement the One Time Inspection Program. 

Implement the Selective Leaching of Materials Program. The 

program will include a one-time inspection of selected 

components where selective leaching has not been identified 

and periodic inspections of selected components where 

selective leaching has been identified. 

Implement the Buried Piping And Tanks Inspection 

Program. 

Implement the One-Time Inspection of ASME Code Class 1 

Small Bore-Piping Program. 

Enhance the program to specifically address the scope of the 

program, relevant degradation mechanisms and effects of 

interest, the refueling outage inspection frequency, the 

training requirements for inspectors and the required periodic 

reviews to determine program effectiveness. 

Implement the Inspection oflnternal Surfaces in 

Miscellaneous Piping and Ducting Components Program. 

Enhance the program to add required equipment, lube oil 

analysis required, sampling frequency, and periodic oil 

changes. 

Enhance the program to sample the oil for the Reactor 

Coolant pump oil collection tanks. 

Enhance the program to require the performance of a onetime ultrasonic thickness measurement of the lower portion 

of the Reactor Coolant pump oil collection tanks 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. 

Within ten years prior to the A.2.1.22 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. 

Prior to the period of extended A.2.1.25 operation. 

Prior to the period of extended A.2.1.26 operation. 

A.2.1.26 Prior to the period of extended 

operation. 

A.2.1.26 Prior to the period of extended 

operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 5/ Page 7 

31. 

ASME Section XI, Subsection 

IWL 

32. Structures Monitoring Program 

33. Structures Monitoring Program 

Electrical Cables and 

Connections Not Subject to 10 

34. CFR 50.49 Environmental 

Qualification Requirements 

Electrical Cables and 

Connections Not Subject to 10 

CFR 50.49 Environmental 

35. Qualification Requirements 

Used in Instrumentation 

Circuits 

Inaccessible Power Cables Not 

Subject to 10 CFR 50.49 

36. Environmental Qualification 

Requirements 

37. Metal Enclosed Bus 

38. Fuse Holders 

Enhance procedure to include the definition of "Responsible A.2.1.28 Prior to the period of extended 

Engineer". operation. 

Enhance procedure to add the aging effects, additional Prior to the period of extended locations, inspection frequency and ultrasonic test A.2.1.31 operation. requirements. 

Enhance procedure to include inspection of opportunity Prior to the period of extended when planning excavation work that would expose A.2.1.31 

inaccessible concrete. operation. 

Implement the Electrical Cables and Connections Not Prior to the period of extended Subject to 10 CFR 50.49 Environmental Qualification A.2.1.32 operation. Requirements program. 

Implement the Electrical Cables and Connections Not Prior to the period of extended 

Subject to 10 CFR 50.49 Environmental Qualification A.2.1.33 operation. Requirements Used in Instrumentation Circuits program. 

Implement the Inaccessible Power Cables Not Subject to 10 Prior to the period of extended CFR 50.49 Environmental Qualification Requirements A.2.1.34 operation. program. 

Implement the Metal Enclosed Bus program. A.2.1.35 Prior to the period of extended 

operation. 

Implement the Fuse Holders program. A.2.1.36 Prior to the period of extended 

operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 5/ Page 8 

Electrical Cable Connections 

Not Subject to 10 CFR 50.49 

39. Environmental Qualification 

Requirements 

40. 345 KV SF6 Bus 

41. 

Metal Fatigue of Reactor 

Coolant Pressure Boundary 

Metal Fatigue of Reactor 

42. Coolant Pressure Boundary 

Pressure -Temperature Limits, 

43. including Low Temperature 

Overpressure Protection Limits 

Implement the Electrical Cable Connections Not Subject to Prior to the period of extended 10 CFR 50.49 Environmental Qualification Requirements A.2.1.37 operation. program. 

Implement the 345 KV SF6 Bus program. A.2.2.1 Prior to the period of extended 

operation. 

Enhance the program to include additional transients beyond A.2.3.l Prior to the period of extended 

those defined in the Technical Specifications and UFSAR. operation. 

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

count transients to monitor cumulative usage on selected A.2.3.1 operation. components. 

The updated analyses will be 

Seabrook Station will submit updates to the P-T curves and submitted at the appropriate 

L TOP limits to the NRC at the appropriate time to comply A.2.4.1.4 time to comply with 10 CFR 

with 10 CFR 50 Appendix G. 50 Appendix G, Fracture 

Toughness Requirements. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 5/ Page 9 

Environmentally-Assisted 44. Fatigue Analyses (TLAA) 

N extEra Seabrook will perform a review of design basis 

ASME Class 1 component fatigue evaluations to determine 

whether the NUREG/CR-6260-based components that have 

been evaluated for the effects of the reactor coolant 

environment on fatigue usage are the limiting components 

for the Seabrook plant configuration. If more limiting 

components are identified, the most limiting component will 

be evaluated for the effects of the reactor coolant 

environment on fatigue usage. If the limiting location 

identified consists of nickel alloy, the environmentallyassisted fatigue calculation for nickel alloy will be performed 

using the rules 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 of the reactor water 

environment. This includes applying the appropriate Fen 

A.2.4.2.3 At least two years prior to the 

factors to valid CUFs determined from an existing fatigue period of extended operation. 

analysis valid for the period of extended operation or from an 

analysis using an NRC-approved version of the ASME code 

or NRC-approved alternative (e.g., NRC-approved code 

case). 

(2) If acceptable CUFs cannot be demonstrated for all the 

selected locations, then additional plant-specific locations 

will be evaluated. For the additional plant-specific locations, 

if CUF, including environmental effects is greater than 1.0, 

then Corrective Actions will be initiated, in accordance with 

the Metal Fatigue of Reactor Coolant Pressure Boundary 

Program, B.2.3.1. Corrective Actions will include 

inspection, repair, or replacement of the affected locations 

before exceeding a CUF of 1.0 or the effects of fatigue will 

be managed by an inspection program that has been reviewed 

and approved by the NRC (e.g., periodic non-destructive 

examination of the affected locations at inspection intervals 

to be determined by a method accepted by the NRC). 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 5/ Page 10 

45. Number Not Used 

46. 

Protective Coating Monitoring 

and Maintenance 

47. 

Protective Coating Monitoring 

and Maintenance 

48. 

Protective Coating Monitoring 

and Maintenance 

49. 

Protective Coating Monitoring 

and Maintenance 

50. 

ASME Section XI, Subsection 

IWE 

51. Number Not Used 

52. 

ASME Section XI, Subsection 

IWL 

53. Reactor Head Closure Studs 

Enhance the program by designating and qualifying an 

Inspector Coordinator and an Inspection Results Evaluator. 

Enhance the program by including, "Instruments and 

Equipment needed for inspection may include, but not be 

limited to, flashlight, spotlights, marker pen, mirror, 

measuring tape, magnifier, binoculars, camera with or 

without wide angle lens, and self sealing polyethylene 

sample bags." 

Enhance the program to include a review of the previous two 

monitoring reports. 

Enhance the program to require that the inspection report is 

to be evaluated by the responsible evaluation personnel, who 

is to prepare a summary of findings and recommendations 

for future surveillance or repair. 

Perform UT of the accessible areas of the containment liner 

plate in the vicinity of the moisture barrier for loss of 

material. Perform opportunistic UT of inaccessible areas. 

Implement measures to maintain the exterior surface of the 

Containment Structure, from elevation -30 feet to +20 feet, in 

a dewatered state. 

Replace the spare reactor head closure stud(s) manufactured 

from the bar that has a yield strength> 150 ksi with ones that 

do not exceed 150 ksi. 

A.2.1.38 Prior to the period of extended 

operation. 

A.2.1.38 Prior to the period of extended 

operation. 

A.2.1.38 Prior to the period of extended 

operation. 

A.2.1.38 Prior to the period of extended 

operation. 

Baseline inspections were 

completed during ORI 6. 

A.2.1.27 Repeat containment liner UT 

thickness examinations at 

intervals of no more than five 

(5) refueling outages. 

A.2.1.28 Complete 

Prior to the period of extended A.2.1.3 operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 5/ Page 11 

Steam Generator Tube 

54. Integrity 

Steam Generator Tube 

55. Integrity 

Closed-Cycle Cooling Water 

56. System 

Closed-Cycle Cooling Water 

57. System 

NextEra will address the potential for cracking of the 

primary to secondary pressure boundary due to PWSCC of 

tube-to-tubesheet welds using one of the following two 

options: 

1) Perform a one-time inspection of a representative sample 

oftube~to-tubesheet welds in all steam generators to 

determine if PWSCC cracking is present and, if cracking is 

identified, resolve the condition through engineering 

evaluation justifying continued operation or repair the 

condition, as appropriate, and establish an ongoing 

monitoring program to perform routine tube-to-tubesheet 

weld inspections for the remaining life of the steam A.2.1.10 Complete 

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 oftube-to-tubesheet weld cracking, or 

redefining the pressure boundary in which the tube-totubesheet 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. 

Seabrook will perform an inspection of each steam generator A.2.1.10 

Within five years prior to the 

to assess the condition of the divider plate assembly. period of extended operation. 

Revise the station program documents to reflect the EPRl Prior to the period of extended 

Guideline operating ranges and Action Level values for A.2.1.12 operation. hydrazine and sulfates. 

Revise the station program documents to reflect the EPRl Prior to the period of extended 

Guideline operating ranges and Action Level values for A.2.1.12 operation. Diesel Generator Cooling Water Jacket pH. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 5/ Page 12 

58. Fuel Oil Chemistry 

Nickel Alloy Nozzles and 

59. Penetrations 

Buried Piping and Tanks 

60. Inspection 

61. 

Compressed Air Monitoring 

Program 

62. Water Chemistry 

63. Flow Induced Erosion 

64. Buried Piping and Tanks 

Inspection 

65. Flux Thimble Tube 

66. Number Not Used 

Update Technical Requirement Program 5.1, (Diesel Fuel Oil Prior to the period of extended Testing Program) ASTM standards to ASTM D2709-96 and A.2.1.18 

ASTM D4057-95 required by the GALL Xl.M30 Rev 1 operation. 

The Nickel Alloy Aging Nozzles and Penetrations program Prior to the period of extended will implement applicable Bulletins, Generic Letters, and A.2.2.3 operation. staff accepted industry guidelines. 

Implement the design change replacing the buried Auxiliary Prior to the period of extended Boiler supply piping with a pipe-within-pipe configuration A.2.1.22 operation. with leak detection capability. 

Replace the flexible hoses associated with the Diesel A.2.1.14 Within ten years prior to the 

Generator air compressors on a frequency of every 10 years. period of extended operation. 

Enhance the program to include a statement that sampling Prior to the period of extended frequencies are increased when chemistry action levels are A.2.1.2 operation. exceeded. 

Ensure that the quarterly CVCS Charging Pump testing is 

continued during the PEO. Additionally, add a precaution to WA 

the test procedure to state that an increase in the eves Prior to the period of extended 

Charging Pump mini flow above the acceptance criteria may A.2.1.2 operation. 

be indicative of erosion of the mini flow orifice as described 

in LER 50-275/94-023. 

Soil analysis shall be performed prior to entering the period 

of extended operation to determine the corrosivity of the soil 

in the vicinity ofnon-cathodically protected steel pipe within A.2.1.22 Prior to the period of extended 

the scope of this program. If the initial analysis shows the operation. 

soil to be non-corrosive, this analysis will be re-performed 

every ten years thereafter. 

Implement measures to ensure that the movable incore Prior to the period of extended detectors are not returned to service during the period of 

extended operation. NIA operation. - In Progress 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 5/ Page 13 

67. Structures Monitoring Program 

68. Structures Monitoring Program 

Open-Cycle Cooling Water 

69. System 

Closed-Cycle Cooling Water 

70. System 

Alkali-Silica Reaction (ASR) 

Monitoring Program I 

71. Building Deformation 

Monitoring Program 

72. Flow-Accelerated Corrosion 

Perform one shallow core bore in an area that was 

continuously wetted from borated water to- be examined for 

concrete degradation and also expose rebar to detect any 

degradation such as loss of material. The removed core will 

also be subjected to petrographic examination for concrete 

degradation due to ASR per ASTM Standard Practice C856. 

Perform sampling at the leak off collection points for 

chlorides, sulfates, pH and iron once every three months. 

Replace the Diesel Generator Heat Exchanger Plastisol PVC 

lined Service Water piping with piping fabricated from 

AL6XN material. 

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

to determine whether the loss of material due to cavitation 

induced erosion has been eliminated or whether this remains 

an issue in the primary component cooling wat~r system. 

NextEra has completed testing at the University of Texas 

Ferguson Structural Engineering Laboratory which 

demonstrates the parameters being monitored and 

acceptance criteria used are appropriate to manage the 

effects of ASR. 

NextEra will-Iimplement the Alkali-Silica Reaction (ASR) 

Monitoring Program and Building Deformation Monitoring 

Program described in B.2.J.31A and B.2.l.31B of the 

License Renewal Application. 

+esting 0wiU be peFfefHled te eentifm that parnmeteFs being 

meniteFed and aeeeptanee eFiteFia used am ap13rn13Fiate te 

manage the effeets efAgR: 

Enhance the program to include management of wall 

thinning caused by mechanisms other than F AC. 

A.2.1.31 Complete 

A.2.1.31 Complete 

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 

A.2.l.31B operation. 

A.2.1.8 

Prior to the period of extended 

operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 5/ Page 14 

Inspection oflntemal Surfaces 

73. in Miscellaneous Piping and 

Ducting Components 

74. Fire Water System 

75. Fire Water System 

76. Fire Water System 

Enhance the program to include performance of focused 

examinations to provide a representative sample of 20%, or a 

maximum of25, of each identified material, environment, 

and aging effect combinations during each 10 year period in 

the period of extended operation. 

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, 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 guidance provided in NFP A 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 NFP A 25 (2011 

Edition). 

• main drain tests 

• deluge valve trip tests 

• fire water storage tank exterior surface inspections 

A.2.1.25 Prior to the period of extended 

operation. 

Prior to the period of extended A.2.1.16 operation. 

A.2.1.16 Prior to the period of extended 

operation. 

-

A.2.1.16 

Prior to the period of extended 

operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 5/ Page 15 

77. Fire Water System 

78. External Surfaces Monitoring 

Open-Cycle Cooling Water 79. System 

80. Fire Water System 

81. Fuel Oil Chemistry 

Inspection oflnternal Surfaces 

82. in Miscellaneous Piping and 

Ducting Components 

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). 

• The requirement that if there is a 10 percent reduction in Prior to the period of extended full flow pressure when compared to the original A.2.1.16 operation. 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 inscope insulated components for possible corrosion under 

insulation. A sample of outdoor component surfaces that are 

insulated and a sample of indoor insulated components A.2.1.24 Prior to the period of extended 

exposed to condensation (due to the in-scope component operation. 

being operated below the dew point), will be periodically 

inspected every 10 years during the period of extended 

operation. 

Enhance the program to include visual inspection of internal 

Within 10 years prior to the coatings/linings for loss of coating integrity. A.2.1.11 period of extended operation. 

Enhance the program to include visual inspection of internal 

Within 10 years prior to the coatings/linings for loss of coating integrity. A.2.1.16 period of extended operation. 

Enhance the program to include visual inspection of internal 

Within 10 years prior to the coatings/linings for loss of coating integrity. A.2.1.18 period of extended operation. 

Enhance the program to include visual inspection of internal 

Within 10 years prior to the coatings/linings for loss of coating integrity. A.2.1.25 period of extended operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 5/ Page 16 

83. Alkali-Silica Reaction 

Monitoring 

ASME Section XI, Subsection 84. 

IWL 

85. Fire Water System 

86. Fire Water System 

87. Fire Water System 

88. Fire Water System 

Enhance the ASR AMP to install extensometers in all Tier 3 

areas of two dimensional reinforced structures to monitor 

expansion due to alkali-silica reaction in the out-of-plane 

direction. 

Monitoring expansion in the out-of-plane direction will 

commence upon installation of the extensometers in 2016 

and continue on a six month frequency through the period 

of extended operation. 

Evaluate the acceptability of inaccessible areas for structures 

within 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 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." 

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 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 drains and associated piping are 

configured to completely drain the piping, and c) normallydry 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. 

Prior to the period of extended A.2.1.28 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. 

A.2.1.16 Within five years prior to the 

period of extended operation. 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 5/ Page 17 

Inspection of Internal Surfaces 

89. in Miscellaneous Piping and 

Ducting Components 

90. PWR Vessel Internals 

Alkali Silica Reaction 

91 Building Deformation 

Monitoring 

Incorporate Coating Service Level III requirements into the 

RCP Motor Refurbishment Specification for the internal 

painting of the motor upper bearing coolers and motor air Prior to the period of extended 

coolers. All four RCP motors will be refurbished and A.2.1.25 operation. 

replaced using the Coating Service Level III requirements 

prior to entering the period of extended operation. 

Implement the PWR Vessel Internals Program. The program 

will be implemented in accordance with MRP-227-A Prior to the period of extended 

(Pressurized Water Reactor Internals Inspection and A.2.1.7 operation Evaluation Guidelines) and NEI 03-08 (Guideline for the 

Management of Materials Issues). 

Implement the Building Deformation Monitoring Program 

Enhance the lA3R Aging Management Structures 

Monitoring 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 

A.'.2.LHA consider the impact to functionality of affected systems and March 15, 2020 

components (e.g., conduit expansion joints). NextEra will A.2.1.31B 

evaluate the specific circumstances against the design basis 

of the affected system or component. 

Enhance the A8R Building Deformation 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. 
Enclosure 6 to SBK-L-16013 

Next Era Seabrook: Application for Withholding 

Proprietary Information from Public Disclosure 

Affidavit in Support of Application for Withholding 

Proprietary Information from Public Disclosure 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 6/ Page 2 

NEXTera .. 

EN~~~~ 

NextEra Energy Seabrook, LLC 

AFFIDAVIT IN SUPPORT OF APPLICATION FOR WITHHOLDING 

PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE 

County of Rockingham ) 

) 

State of New Hampshire ) 

I, Eric McCartney, 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 Enclosure 2 ofNextEra 

Energy Seabrook's letter SBK-L-16013, Eric McCartney (NextEra Energy Seabrook) to 

U.S. Nuclear Regulatory Commission, entitled "Seabrook Station, License Renewal 

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

(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 NextEra 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 
U.S. Nuclear Regulatory Commission 

SBK-L-16013 /Enclosure 5/ Page 3 

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 NextEra 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. 

Subscribed and sworn to before me 

this 9#.. day of August, 2016. 

~~~'~ Notary Public ' ( 

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