Enclosure 4 to this Letter Contains Proprietary Information

Withhold EncloSUre 4 from Public Disclosure in Accordance with 10 CFR 2.390

N Exera®

ENERGQ*

SEABROOK

June 30, 2015

10 CFR 54

SBK-L-15107

Docket No. 50-443

U. S. Nuclear Regulatory Commission

Attn: Document Control Desk

Washington, DC 20555-0001

Seabrook Station

Response to Requests for Additional Information for the Review of the

Seabrook Station, License Renewal Application- SET 23 (TAC NO. ME4028)

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

References:

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

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

2. NRC Letter, Requests For Additional Information for the Review of the Seabrook Station,

License Renewal Application- Set 23 (TAC NO. ME4028), November 25,2014 (Accession

Number ML14301A316)

3. NextEra Energy Seabrook, LLC letter SBK-L-15024, "Response to Requests for

Additional Information for the Review of the Seabrook Station, License Renewal

Application- SET 23 (TAC NO. ME4028) Relating to the Alkali-Silica Reaction (ASR)

Monitoring Program," February 23, 2015. (Accession Number ML15062A042)

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

renewed facility operating license for Seabrook Station Unit 1 in accordance with the Code of

Federal Regulations, Title 10, Parts 50, 51, and 54.

In Reference 3 NextEra provided responses to the information requested in Reference 2 for RAIs

B.2.1.31A-2(a) and B.2.1.31A-7.

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

SBK-L-15107 / Page 2

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

B.2.1.3 1A-5(a) and B.2.1.3 1A-6(a). Within these responses, changes to LRA Appendix A -

Updated UFSAR Supplement, and Appendix B - Aging Management Programs associated with

the Alkali-Silica Reaction Monitoring Program are provided. The changes are explained, and

where appropriate to facilitate understanding, portions of the LRA are repeated with the change

highlighted by strikethroughs for deleted text and bolded italics for inserted text. In some

instances the entire text of a section has been replaced or added. In these cases a note is included

in the introduction indicating the replacement of the entire text of the section.

Enclosure 2 provides the revised LRA Appendix A - Updated Final Safety Analysis Report

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

commitment (83) and one new commitment (91). Commitment 52 status has been changed to

"Complete" as the ongoing dewatering activities have been incorporated into the station's

preventative maintenance program to maintain the exterior surface of the Containment Structure,

from elevation -30 feet to +20 feet, in a dewatered state. In addition, based on staff review,

twelve commitments (24, 44, 55, 56, 57, 60, 61, 64, 65, 71, 72, and 73) were changed to

incorporate consistent terminology of"...prior to the period of extended operation..."

Enclosures 3/4 are provided to support the review of this RAI response. Enclosure 3 is provided

as non-proprietary version of the referenced report. Enclosure 4 provides the proprietary version

as described below.

Enclosure 4 to this letter, MPR-4153, Revision 1: Seabrook Station- Approach for Determining

Through-Thickness Expansion from Alkali-Silica Reaction, MPR Associates, Inc., Alexandria,

VA, June 2015 (Proprietary), contains NextEra Energy Seabrook proprietary information. This

letter is supported by an affidavit signed by NextEra Energy Seabrook, setting forth the basis on

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

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

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

accordance 10 CFR 2.390.
U.S. Nuclear Regulatory Commission

SBK-L- 15107 / Page 3

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

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

If you have any questions regarding this correspondence, please contact Mr. Michael Ossing

Licensing Manager, at (603) 773-7512.

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

Executed on June 0, 2015

Sincerely,

Dean Curtland

Site Vice President

NextEra Energy Seabrook, LLC

Enclosure 1 - Seabrook Station License Renewal Application, Response to Requests for

Additional Information - Set 23, RAIs Relating to the Alkali-Silica Reaction

(ASR) Monitoring Program for the Seabrook Station License Renewal

Application

Enclosure 2 - LRA Appendix A - Final Safety Analysis Report Supplement Table A.3, License

Renewal Commitment List Updated to Reflect Changes to Date

Enclosure 3 - MPR-4153, Seabrook Station- Approach for Estimating Through-Thickness

Expansion from Alkali-Silica Reaction, MPR Associates, Inc., Alexandria, VA,

2015. (Non-Proprietary)

Enclosure 4 - MPR-4153, Revision 1: Seabrook Station- Approach for Determining ThroughThickness Expansion from Alkali-Silica Reaction, MPR Associates, Inc.,

Alexandria, VA, June 2015. (Proprietary)

Enclosure 5 Affidavit in Support of Application for Withholding Proprietary Information from

Public Disclosure
U.S. Nuclear Regulatory Commission

SBK-L-15107 / Page 4

cc: D. H. Dorman

J. G. Lamb

P. C. Cataldo

R. A. Plasse

L. M. James

NRC Region I Administrator

NRC Project Manager

NRC Senior Resident Inspector

NRC Project Manager, License Renewal

NRC Project Manager, License Renewal

Mr. Perry Plummer

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- 15107

Seabrook Station License Renewal Application

Requests for Additional Information - Set 23

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

for the

Seabrook Station License Renewal Application
U.S. Nuclear Regulatory Commission

SBK-L-15107 /Enclosure 1/Page 2

RAI B.2.1.31A-5(a)

Background

In its response to RAI B.2.1.31A-4, dated May 15, 2014, the applicant clarified that the largescale testing program does not have a role in establishing the technical basis for the ASR

Monitoring Program, and with respect to the monitoring parameters, the large-scale testing

program is "confirmatory in nature." However, the testing is referenced in and appears to be

relied upon to address the staff's concerns regarding the "parameters monitored or inspected"

"detection of aging effects," "monitoring and trending," and "acceptance criteria" program

elements.

In response to RAI B.2.1.31A-5, dated May 15, 2014, part (1) regarding the adequacy of the

crack indexing methodology to monitor volumetric expansion, the applicant indicated that the

magnitude of expansion in the reinforced direction (the in-plane, or "x-y", direction which is the

direction being monitored on the concrete surface onsite) will be smaller than in a non-reinforced

direction (the "through-wall" or "transverse" direction not currently being monitored). The

applicant also stated that:-2- [b]ased upon a preliminary review of large-scale testing results to

date, the test specimens appear to confirm CCI as an appropriate parameter for monitoring ASR

progression.... However, expansion in the test specimens in x and y plane appears to level off

over time and is smaller in magnitude than the transverse direction. The applicant further stated

that "the test specimens continue to generate information for other parameters including

transverse direction expansion, material property changes (e.g., modulus) and the results will be

used to confirm and adjust monitoring parameters...." In response to RAI B.2.1.31A-5, part (2)

regarding a proposed method or technique to monitor expansion in the transverse direction, the

applicant committed (Commitment No. 83) to install instrumentation in "representative sample

areas" to monitor expansion in the transverse direction. The applicant stated that "expansion data

from deep pins installed in test specimens will be used to correlate with transverse expansion

measurements from instruments to be installed in plant structures," and that if determined

necessary, "the instrument and pin expansion data will be used to establish acceptance criteria

and monitoring frequencies for monitoring transverse expansion.

Nuclear Regulatory Commission (NRC) guidance as stated in Standard Review Plan for License

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

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

monitoring these parameters will ensure adequate aging management.

Issue

The staff has identified the following concerns:

1. The applicant stated, in its response dated May 15, 2014, to part (1) of RAI B.2.1.31A-5, that a

larger number of smaller cracks would be anticipated in the restrained direction relative to a

smaller number of larger width cracks in an unrestrained direction for the same amount of

ASR progression. However, the applicant did not provide any information regarding the

structural implications of this phenomenon.

2. It is not clear to the staff how measuring Combined Cracking Index (CCI) and crack width as

parameters will correlate the aging effect of cracking due to expansion from reaction with
U.S. Nuclear Regulatory Commission

SBK-L-15107 /Enclosure 1/Page 3

aggregates to the degradation of ASR-affected structures in order to evaluate whether there is

an impact on the structure's intended function. Also, confinement due to reinforcement

detailing may affect the CCI measurement between structures. The staff is not clear how it can

be confirmed that the same CCI value measured on different structures represents the same

degree of ASR degradation.

3. In its response dated May 15, 2014, to part 2 of RAI B.2.1.31A-5, the applicant refers to

"representative sample areas" that it plans to monitor expansion due to ASR in the out-ofplane direction. However, the response did not provide any further information as to how the

sample will be determined. SRP-LR Section A. 1.2.3.4 states that when sampling is used to

represent a larger population size, the basis for the sample size should be provided, and should

be biased toward locations most susceptible to the specific aging effect. The staff does not

have sufficient information to determine whether the sample will be adequate to ensure that

the areas instrumented for through-wall expansion data will be bounding of the ASR-related

degradation at the plant.

4. Commitment No. 83 states that instrumentation to monitor expansion due to ASR in the outof-plane direction will be installed "to determine whether there is a need to enhance the

program to monitor expansion in the out-of-plane direction." Commitment 83 also states that

this evaluation will take place "prior to the period of extended operation;" however, the

implementation schedule did not contain sufficient detail for the staff to review and evaluate.

The staff is concerned that depending on when this activity is performed, if it is determined

that it is necessary to monitor out-of-plane expansion; there may not be sufficient time to

acquire enough information to determine whether the expansion could impact structural

function. The RAI response did not provide any information regarding schedule and duration

of activities, and associated technical bases, in order for the staff to determine whether the

time of implementation provides for adequate monitoring and trending during the Period of

Extended Operation (PEO).

5. The applicant's RAI response states "expansion data from deep pins installed in test

specimens will be used to correlate with out-of-plane expansion measurements from

instruments to be installed in plant structures." It also states that "[i]f an additional monitoring

parameter in the out-of-plane direction is determined to be necessary as part of the [ASR]

Monitoring Program, NextEra will enhance the [program] ... to establish acceptance criteria

and monitoring frequencies for monitoring out of plane expansion." It is not clear how the

expansion data from deep pins in test specimens will be correlated to data from

instrumentation installed in plant structures because there is no expansion data measured to

date for onsite structures, and it is not clear how the test specimens simulate the boundary

conditions of the structural components (e.g., walls) of the onsite structures.

6. Since the majority of the walls of the ASR-affected structures at Seabrook are not restrained in

the transverse direction, expansion in the transverse direction could lead to laminar cracking

which could propagate with time. It is not clear whether the program accounts for the

potential for laminar cracking and its implication with regard to structural and anchor bolt

capacity or if there is a technical basis for not considering the potential for delamination.
U.S. Nuclear Regulatory Commission

SBK-L- 15107 /Enclosure 1/Page 4

7. During a February 21, 2013, public meeting with the applicant, the staff expressed its concern

regarding the ability to correlate CCI to loss of strength, a parameter not monitored as part of

the ASR Monitoring Program. As documented in the meeting summary (Agency wide

Documents Access and Management System (ADAMS) Accession No. ML13066A488), the

applicant responded to that concern stating that the selection of specific acceptance criteria

was based on its review of industry experience. The large-scale testing, which it is currently

pursuing, will provide results that will be used for its structural evaluation to determine

operability and for assessing and evaluating plant structures impacted by ASR. The staff is

concerned that there is no aspect of the ASR Monitoring Program, either periodic or one-time,

that would validate the assumed changes in material properties being established through the

large-scale testing program. It is unclear to the staff whether the applicant will have enough

information to perform an adequate structural evaluation if warranted.

At this time, it appears that multiple elements of the program have yet to be fully established

and may need to be amended; therefore, the staff does not have sufficient information to make

a determination that the effects of cracking due to expansion from reaction with aggregates

will be adequately managed during the period of extended operation.

Request

Considering the staff concerns identified in the "Issue" section, provide information with

technical basis to: (1) demonstrate the adequacy of the parameters proposed to be monitored or

inspected by the program to manage the effects of aging due to ASR; and (2) clearly establish the

link between the parameters that will be monitored and how monitoring these parameters will

ensure adequate aging management such that intended function will be maintained during the

period of extended operation.

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

NextEra's monitoring plan for ASR has two objectives. First, the degree and rate of progression

of ASR over time is monitored and trended to prevent the loss of intended function. Second, the

monitored parameters are used to correlate the structural test data obtained from large scale

testing program at the University of Texas at Austin, Ferguson Structural Engineering

Laboratory (FSEL) to Seabrook Station structures. Expansion (strain) is the primary monitoring

parameter, because the potential for impact on structural performance from ASR is a result of the

internal expansion that leads to concrete cracking [References: 1, 2 &3]. Seabrook Station's

approach to monitoring expansion is to use a combined crack index (CCI) for in-plane (x-y

direction) expansion and extensometers for out-of-plane (through-the-thickness or z direction)

expansion.

The large scale testing program at the FSEL validated the use of expansion to monitor the level

and rate of ASR progression. Preferential ASR expansion is expected in the unrestrained or less

restrained direction. Consistent with this expectation, observation of the large scale test

specimens has confirmed expansion to be significantly more pronounced in the out-of-plane

direction. The FSEL tests include the direct measurement and monitoring of strain in both in-
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SBK-L-15107 /Enclosure 1/ Page 5

plane and out-of-plane directions via cast in place metal pins. Expansion in the in-plane direction

has been observed to plateau (i.e., minimal change over time) at some point while ASR

expansion in the out-of-plane direction continues to progress. [Reference: 6].

NextEra has been using a Combined Cracking Index (CCI) method to monitor expansion in the

in-plane directions for approximately two years. CCI is monitored in the same manner for the

plant as it is for the test specimens. CCI correlates well with strain in the in-plane directions and

the ability to visually detect cracking in exposed surfaces makes 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.

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. No structural impacts from ASR have been seen at these plateau

levels in the large scale testing program at FSEL. In fact, no adverse structural impacts have

been identified at any level of tested ASR expansion to date.

In addition to the three tier approach for monitoring the CCI for the in-plane direction, ASR

expansion in the out-of-plane direction will be monitored by borehole extensometers installed in

drilled core bore holes. Expansion is proven to be the most effective parameter for long term

monitoring [Reference: 6]. Out-of-plane expansion is also the best parameter for correlating the

degree of ASR progression in the test specimens to the in-situ plant structures. The ASR aging

management program will be enhanced to include measurement of out-of-plane expansion. A

representative sample of out-of-plane monitoring areas will be chosen to bound different

environmental conditions and levels of ASR. The locations in plant structures that will be

selected, for the installation of the extensometers to monitor out-of-plane expansion are

representative of various plant exposure conditions, levels of ASR progression, and structural

details. Extensometers will also be installed in selected control locations not impacted by ASR

for comparison of instrument data. The selected locations for monitoring out-of-plane expansion

consider the following criteria.

" Exposure; Ambient weather exposure and below grade ground water exposure

" Level of ASR progression based on CCI; Tier 3, Tier 2, Tier 1 (control instruments)

" Thickness of concrete element; 1.5 foot, 2 foot, 2.5 foot, 3 foot, 3.5 foot, 4 foot, 6 foot

* Type of concrete element; walls and slabs

" Design compressive strength of the concrete; 3 ksi and 4 ksi

The large scale testing program at FSEL determines the actual structural performance at various

levels of ASR expansion. The acceptance criteria for structural performance will be determined

from the test results by correlating the degree of monitored ASR expansion at the plant to the

plot of tested structural performance of the test specimens versus ASR expansion [Reference: 6].

The instruments to measure out-of-plane expansion were not installed during original

construction. This means that an estimate of accumulated strain to date is required to be added to

any monitoring of strain going forward. A correlation curve for out-of-plane expansion and
U.S. Nuclear Regulatory Commission

SBK-L-15107 /Enclosure 1/ Page 6

percent reduction in tested modulus was developed for Seabrook Station [Reference: 6]. The test

data used for this correlation were obtained from full scale test specimens that were designed to

be representative of the concrete and reinforcement details at Seabrook Station. Furthermore, the

correlation compares well with trends noted in the literature of other published studies of free

expansion [Reference: 6]. Using this correlation to determine through-thickness expansion to

date, the structural impact of ASR for any level of out-of-plane expansion can be determined.

The uncertainties associated with this correlation will be included in the evaluation of structural

implications for a given level of ASR expansion. When installing the extensometers at the plant,

the removed cores will be tested for modulus. Total out-of-plane expansion is determined as the

sum of the pre-instrument expansion (determined from modulus loss correlation) and measured

expansion from the extensometers. Levels of ASR expansion are within the acceptance criteria

for structural performance when the monitored quantitative value of ASR expansion is less than

the ASR expansion level successfully tested in structural specimens at FSEL.

The observed strains due to ASR are very small in magnitude and adequately monitored by CCI

and extensometers for the purpose of observing rate of progression and for correlating the

structural strength impacts from the large scale testing program. However recent operating

experience at NextEra Seabrook has shown that over large distances, and with the right building

geometry, discemable dimension changes in a structure can result due to ASR. While this is not

a structural capacity concern, it can result in relative displacement between structures. Additional

monitoring of this relative displacement potential and its impact to plant systems and

components will be included in the ASR Monitoring program. Specifically, this will include

additional monitoring focused on identifying signs of relative building deformation (e.g., fire

seal displacement, seismic gap width changes, pipe/conduit misalignments at penetrations or

between adjacent structures, bent or displaced pipe/conduit and supports, doorway

misalignments). Critical building geometry locations where the potential for deformation is

likely will be monitored for displacement via laser targets and gap measurements to validate that

design margin remains acceptable. If a monitoring location is found outside of its original design

the location will be restored to be within the original design or reanalyzed for continued

operability.

NextEra Energy Response to Issue No. 1

There are no direct structural implications resulting from this phenomenon. The ASR structural

implications are best correlated to expansion or strain. Strain in a given direction is approximated

by the summation of crack widths over distance. For a given strain, more reinforcement will

generally result in a relatively larger number of narrower cracks than for the same strain with

lesser reinforcement. The large scale test specimens at FSEL include reinforcement details with

#9 and #11 bars at different spacing. This is representative of the reinforcement detailing for

walls at Seabrook Station. No disceruable difference in in-plane crack width or crack pattern

was seen between these types of reinforcement differences. CCI was found to correlate with

strain in both the x and y in-plane directions. In the out-of-plane direction there is no

reinforcement (also the case for limiting structures at the plant). In this case a significant
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SBK-L-15107 /Enclosure 1/Page 7

difference in individual crack width was noted, with essentially one or a few wide cracks

perpendicular to the direction of expansion. The structural testing is being correlated to the outof-plane expansion and therefore, implicitly includes any structural impact of preferential

expansion in this direction.

CCI is used as a surrogate for accumulated strain from ASR expansion in the in-plane directions.

Cracking in the in-plane direction as a monitoring parameter for the plant was originally based

on industry recommendations [Reference: 1 and 3]. A second factor in the selection of CCI is a

means to provide an estimate of cumulative expansion to date, whereas other methods for

monitoring in-plane expansion such as imbedded pins only provide information of expansion that

occurs after the pins are installed.

NextEra Energy Response to Issue No. 2

The large scale test program at FSEL includes direct measurement of in-plane strain in the

specimens to validate the correlations of CCI to internal expansion. CCI was confirmed to

correlate well with respect to in-plane strain. The CCI values measured on the test specimens are

applicable to CCI values measured at Seabrook Station based on the following considerations:

(1) the specimens are designed to be representative of structures at Seabrook Station; (2)

reinforcement detailing, which is the primary means of confinement, is representative of

structures at Seabrook Station; and (3) the CCI procedure used at the laboratory is the same as

that used at Seabrook Station.

CCI has not been observed to be impacted by nominal differences in reinforcement detailing

between the longitudinal and perpendicular directions of a given specimen (No. 9 bar versus No.

11 bar), or by differences in the longitudinal reinforcement ratio for the shear test specimens

versus the reinforcement anchorage test specimens. The reinforcement bars used in the test

specimens duplicate the plant details and are much larger than is typical outside the nuclear

industry. The measured in-plane strain was much lower in magnitude than the strain

perpendicular to the reinforcement and was shown to level off over time. Additionally, the outof-plane expansion was of significantly greater magnitude than the in-plane expansion.

Therefore, the long-term ASR monitoring at the site will include out-of-plane expansion in

addition to in-plane expansion. Out-of-plane expansion will be the principal parameter for

correlating the beam test structural results to the plant. CCI will be used in the initial assessment

of ASR and will provide correlation for the anchor bolt program, since out-of-plane expansion is

not as influential as in-plane expansion (cracking) for anchor bolt performance.

NextEra Energy Response to Issue No. 3

Not all ASR affected areas will be directly monitored for out-of-plane expansion. The monitored

locations will be selected to bound the most advanced ASR areas and environmental conditions,

as well as include locations with lower levels of ASR and control locations (no ASR). ASR in
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SBK-L- 15107 /Enclosure 1 / Page 8

the out-of-plane direction will be monitored by at least thirty-four (34) Geokon model A-4,

single snap ring anchor, rod-type borehole extensometers installed in bore holes. The instrument

type and installation is based on work done at the FSEL in the instrument beam portion of the

testing. The locations in plant structures that were selected for the installation of the

extensometers to monitor out-of-plane expansion are representative of various plant exposure

conditions, levels of ASR distress, and structural details. Extensometers will also be installed in

selected control locations not impacted by ASR for comparison of instrument data. The selected

locations bound the following criteria.

0 Exposure; Ambient weather exposure and below grade ground water exposure

0 Level of ASR progression based on CCI; Tier 3, Tier 2, Tier 1 (control instruments)

0 Thickness of concrete element; 1.5 foot, 2 foot, 2.5 foot, 3 foot, 3.5 foot, 4 foot, 6 foot

• Type of concrete element; walls and slabs

* Design compressive strength of the concrete; 3 ksi and 4 ksi

Below is a breakdown of number of instruments to be installed in each of the above criteria,

totaling 34 locations.

Expostire QuanfitN of

I Instruments

Below-Grade

Groundwater

Exposure

BTow-Grade No

Water Control

TOTAL

10

6

34

2ft 17

2 V2ft 4

3ft 2

4ft 7

6ft 2

TOTAL 34

3 ksi 24

4 ksi 10

TOTAL 34

Tier 1 (control 15

Tier 2 10

TierO3 9

TOTAL 34

Wall 31

Slab 3

TOTAL 34
U.S. Nuclear Regulatory Commission

SBK-L-15107 /Enclosure 1/Page 9

Using the Operating Experience Element of the ASR Monitoring Program, the sample size will

be increased if necessary.

NextEra Energy Response to Issue No. 4

It has been concluded that monitoring of out-of-plane expansion is required prior to entering the

period of extended operation and continuing through the period of extended operation. The

instrumentation type and installation details have already been developed and these instruments

will be installed in thirty-four (34) sample locations in 2016. . Data collection will commence

upon installation in accordance with the ASR Monitoring Program and will continue through the

period of extended operation. Collection of data starting in 2016 through the start of extended

operations in 2030 provides sufficient time for monitoring and trending of ASR progression.

NextEra Energy Response to Request No. 5

The deep pins that will be installed are extensometers installed in bore holes in the affected

walls. The instrument type and mounting is based on mockups developed at FSEL for this

program. These instruments provide the best correlation to direct measurements of out-of-plane

expansion. Because the instruments to measure out-of-plane expansion were not installed during

original construction, an estimate of accumulated strain to date is required. A correlation curve

for out-of-plane expansion and percent reduction in tested modulus was developed for Seabrook

Station [Reference: 6]. The test data used for this correlation were obtained from full scale test

specimens that were designed to be as close as practical to the concrete and reinforcement details

at Seabrook Station. Furthermore, the correlation compares well with trends noted in the

literature of other published studies of free expansion [Reference: 6]. Based on this correlation,

the structural impact of ASR for any level of out-of-plane expansion can be determined. The

uncertainties associated with this correlation will be included in the evaluation of structural

implications for a given level of ASR expansion. When installing the extensometers at the plant

site, the removed cores will be tested for modulus. Total out-of-plane expansion is determined

as the sum of the pre-instrument expansion (determined from modulus loss correlation) and

measured expansion from the extensometers.

NextEra Energy Response to Issue No. 6

As stated previously, the expansion in the unreinforced out-of-plane direction has been seen to

be much greater than in the reinforced in-plane direction. In the test specimens this has resulted

in the formation of one or a few relatively large surface cracks perpendicular to the expansion.

All investigated cracks have been shown to be mostly a surface phenomenon by direct sectioning

of affected specimens and not leading to delamination [Reference: 7]. Cracks in this orientation

are not observable by inspection of the interior surface and/or exterior surface of in-situ walls at

the plant. Additionally, core bores removed to install the out-of-plane instruments will be

thoroughly inspected to rule out the laminar cracking phenomenon if it's present in the plant.

Thirty-four (34) concrete cores have been removed to date from ASR impacted concrete and
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SBK-L-15107 /Enclosure 1/Page 10

visually examined. No evidence of laminar cracking in any of the concrete cores has been

identified to date.

NextEra Energy Response to Issue No. 7

Monitoring the potential loss of structural strength in ASR affected structures is the primary goal

of the monitoring program. However this parameter cannot be monitored directly. Expansion

due to ASR (in the through-the-thickness direction) will be monitored and correlated to the

structural impacts quantified in the large scale testing program.

Although concrete cores will be removed from affected structures and tested, those test results

will not be indicative of material strength changes. The expansion (strain) due to ASR and

resulting cracking may degrade the tested material properties of the concrete samples removed

[References: 1, 2 &3]. The potential for impact on structural performance from ASR is also a

result of the internal expansion that leads to concrete cracking. Consequently, the tested material

strength properties in removed concrete core samples that typically demonstrate the most

significant changes are elastic modulus and tensile strength. Tested compressive strength may

also be affected, but less rapidly and less severely. While the progression of ASR results in a

reduction in tested material properties of the removed core samples, there is not necessarily a

corresponding decrease in the actual in-situ structural performance. As discussed in previous

reports on ASR at Seabrook Station [References: 4&5], cores removed from a reinforced ASR

affected structure for testing are no longer confined by the reinforcement and therefore, do not

represent the structural context of the in-situ condition. Structural context is of particular

importance with an ongoing internal expansion mechanism like ASR where the confining

reinforcement may function more as pre-stressed or post tensioned reinforcement. Therefore, in

ASR affected structures, the material properties obtained from removed cores have limited

applicability for directly evaluating the actual structural capacity. These core-based material tests

do however, remain indicative of the amount of ASR progression (expansion and cracking) and

are useful additional monitoring parameters. Monitoring will include core samples taken when

the extensometers are installed and these will be tested for compressive strength and elastic

modulus.

Large scale specimens that duplicate the reinforcement detailing of the subject structure retain

the structural context during testing. The large scale testing program conducted for Seabrook at

FSEL provides direct measurement of the structural performance at various levels of ASR

progression with the structural context of the specimen intact. The test specimens were designed

with the same concrete strength and reinforcement details as the structures at Seabrook Station.

ASR progression was accelerated in the test specimens both by adding additional aggregate

reactivity to the original mix design and conditioning at elevated ambient temperature and

humidity. The testing methodology is the same as was used in development of the ACI 318 code

relationships. Structural testing of the specimens was performed at several ASR levels (based on

monitored strains, CCI, material core tests and petrography). The structural performance of the
U.S. Nuclear Regulatory Commission

SBK-L-15107 /Enclosure 1/Page 11

plant at any given level of ASR can then be determined from the test results by correlating the

ASR expansion at the plant to the plot of ASR expansion versus tested structural performance of

the test specimens [Reference: 6]. This will provide reasonable assurance that the aging effects

due to ASR expansion can be adequately managed during the period of extended operation to

prevent the loss of the intended function of structural performance.. Furthermore, FSEL testing

to date has not shown any negative structural implications due to the advanced progression of

ASR in structural components designed to be representative of Seabrook Station.

References:

1. Institution of Structural Engineers, Structural Effects of Alkali-Silica Reaction: Technical

Guidance on the Appraisal of Existing Structures, London, UK, 1992.

2. Bayrak, 0., "Structural Implications of ASR: State of the Art," July 2014.

3. Fournier, B. et al, FHWA-HIF-09-004, Report on the Diagnosis, Prognosis, and Mitigation of

Alkali-Silica Reaction in Transportation Structures, January 2010.

4. MPR-3727, Seabrook Station: Impact of Alkali-Silica Reaction on Concrete Structures and

Attachments, Rev. 1, MPR Associates, Inc., Alexandria, VA, 2012.

5. MPR-3848, Seabrook Station - Approach for Shear and Reinforcement Anchorage Testing of

Concrete Affected by Alkali-Silica Reaction, April 2013.

6. MPR-4153, Revision 1: Seabrook Station- Approach for Determining Through-Thickness

Expansion from Alkali-Silica Reaction, MPR Associates, Inc., Alexandria, VA, June 2015.

(Proprietary)

7. Seabrook/MPR ASR Test Programs at FSEL: Destructive Examination of Anchor Test

Specimen AN-02, MPR DRN 0326-0058-171, Revision 0, January 22, 2014.

Based on the above clarification, LRA Appendix B and Appendix A have been revised as

follows:

a) LRA Appendix B, Section B.2.1.31A, Alkali Silica Reaction (ASR) Monitoring Program, the

following new paragraphs have been added to the end of Element 3 (Parameters

Monitored/Inspected)
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SBK-L-15107 /Enclosure 1/Page 12

ELEMENT 3 - PARAMETERS MONITORED/INSPECTED

Monitoring Expansion in the Out-Of-Plane Direction

CCI correlates well with strain in the in-plane directions and the ability to visually detect

cracking in exposed surfaces makes it an effective initial detection parameter. CCI's

limitation for heavily reinforced structures is that it has been observed to plateau at a

relatively low level of accumulated strain. 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.

Monitoring Building Deformation

The observed strains due to ASR are of very small magnitude and adequately monitored by

CCI and extensometers for the purpose of observing rate of progression and for

correlating the structural strength impacts from the large scale testing program. However,

over large distances, and with the right building geometry, these strains due to ASR can

result in discernable dimensional changes in a structure. While this is not a structural

capacity concern, it can result in relative displacement between structures. Additional

monitoring of this relative displacement potential and its impact to plant systems and

components will be included in the ASR Monitoring Program. Specifically, this will

include additional monitoring focused on identifying signs of relative displacement or

building deformation (e.g., fire seal displacement, seismic gap width changes, pipe/conduit

misalignments at penetrations or between adjacent structures, bent or displaced

pipe/conduit and supports, doorway misalignments). Critical building geometry locations

where the potentialfor deformation is likely will be monitored for displacement via laser

targets and gap measurements.

b) LRA Appendix B, Section B.2.1.31A, Alkali Silica Reaction (ASR) Monitoring Program, the

following new paragraphs have been added to the end of Element 5 (Monitoring and

Trending)
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SBK-L-15107 /Enclosure 1/ Page 13

ELEMENT 5 - MONITORING AND TRENDING

Monitorinz Expansion in the Out-Of-Plane Direction

The deep pins that will be installed are extensometers installed in bore holes in the affected

walls. The instrument type and mounting is based on mockups developed at FSEL for this

program. These instruments provide the best correlation to direct measurements of out-ofplane expansion. Because the instruments to measure out-of-plane expansion were not

installed during original construction, an estimate of accumulated strain to date is required.

A correlation curve for out-of-plane expansion andpercent reduction in tested modulus was

developed for Seabrook Station. The test data used for this correlation were obtained from full

scale test specimens that were designed to be as close as practical to the concrete and

reinforcement details at Seabrook Station. Furthermore, the correlation compares well with

trends noted in the literature of other published studies offree expansion [Reference: 6J.

Using this correlation to determine through-thickness expansion to date,, the structural impact

of ASR for any level of out-of-plane expansion can be determined. The uncertainties

associated with this correlation will be included in the evaluation of structural implications for

a given level of ASR expansion. When installing the extensometers at the plant site, the

removed cores will be tested for modulus. Total out-of-plane expansion is determined as the

sum of thepre-instrument expansion (determined from modulus loss correlation) and

measured expansion from the extensometers.

Monitoring' Building' Deformation

Monitoring of relative displacement potential and its impact to plant systems and components

will include additional monitoring focused on identifying signs of relative building

deformation (e.g., fire seal displacement, seismic gap width changes, pipe/conduit

misalignments at penetrations or between adjacent structures, bent or displaced pipe/conduit

and supports, doorway misalignments). Critical building geometry locations where the

potential for deformation is likely will be monitored for displacement via laser targets and gap

measurements to validate that the established design margin remains acceptable.
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SBK-L-15107 /Enclosure 1/Page 14

c) LRA Appendix B, Section B.2.1.31A, Alkali Silica Reaction (ASR) Monitoring Program, the

following new paragraph has been added to the end of Element 6 (Acceptance Criteria).

ELEMENT 6 - ACCEPTANCE CRITERIA

Monitoring Expansion in the Out-Of-Plane Direction

Total out-of-plane expansion is determined as the sum of the pre-instrument expansion

(determined from modulus loss correlation) and measured expansion from the extensometers.

Levels of ASR expansion are within the acceptance criteria for structural performance when

the monitored quantitative value of ASR expansion is less than the ASR expansion level

successfully tested in structural specimens at FSEL.

Monitorin' Building Deformation

Monitoring of relative displacement potential and its impact to plant systems and components

will include additional monitoring focused on identifying signs of relative building

deformation (e.g., fire seal displacement, seismic gap width changes, pipe/conduit

misalignments at penetrations or between adjacent structures, bent or displaced pipe/conduit

and supports, doorway misalignments). Critical building geometry locations where the

potentialfor deformation is likely will be monitored for displacement via laser targets and gap

measurements to validate that the established design margin remains acceptable. If a

monitoring location is found outside of its original design, the location will be restored to be

within the original design or reanalyzed for continued operability.

d) LRA Appendix A, Section A.2.1.31A, Alkali Silica Reaction (ASR) Monitoring Program,

the following new paragraphs have been added to the end of the program description as

follows:

Program Description

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. 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. Although the observed

strains due to ASR are of very small magnitude and adequately monitored by CCI and

extensometers, over large distances and with the right building geometry, they can result in
U.S. Nuclear Regulatory Commission

SBK-L-15107 /Enclosure 1/Page 15

discernable dimension changes in a structure. While this is not a structural capacity

concern, it can result in relative displacement between structures. Additional monitoring of

this relative displacement potential and its impact to plant systems and components will be

included in the ASR Monitoring Program. Specifically, this will include additional

monitoring focused on identifying signs of relative displacement or building deformation

(e.g., fire seal displacement, seismic gap width changes, pipe/conduit misalignments at

penetrations or between adjacent structures, bent or displaced pipe/conduit and supports,

doorway misalignments). Critical building geometry locations where the potential for

deformation is likely will be monitored for displacement via laser targets and gap

measurements.

e) LRA Appendix A, Section A.3, Commitment #83 has been revised and a new commitment

#91 has been added as follows:

Note: The entire text of Commitment #83 has been revised.

PROGRAM UF'SAR

No. COMMITMENT SCHEDULE

or TOPIC LOCATION

Install extensometers in representative sample

areas of structures to monitor expansion due to

alkali-silica reaction in the out-of-plane

direction. Install at least thirty-four (34)

extensometers by December 31, 2016. Using

Alkali-Silica the Operating Experience Element of the ASR

83 Reaction Monitoring Program, determine if increase in A.2.1.3 1A December 31, 2016

Monitoring sample size is required.

Monitoring expansion in the out-of-plane

direction will commence upon installation of

the extensometers in 2016 and continue

through the period of extended operation.

In building geometry locations where the Within 10years

Alkali-Silica potentialfor deformation is likely, enhance the prior to the period

91 Reaction program to monitor for building displacement A.2.1.31A of extended

Monitoring using laser targets and by taking gap operation.

measurements.
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SBK-L-15107/Enclosure 1/Page 16

B.2.1.31A-6(a)

Background

In its response to RAI B.2.1.31A-6, dated May 15, 2014, the applicant stated that a separate

evaluation for inaccessible areas has not been performed; however, the opportunistic or focused

inspections performed every 5 years will establish the baseline for future monitoring and

trending. The applicant also stated that the concrete materials used to produce concrete at the site

was the same for both accessible and inaccessible areas of structures; therefore, the performance

and aging would be the same.

Regarding the evaluation of inaccessible areas, the applicant stated in its response to RAI

B.2.1.31-1, dated March 30, 2012, that "it is expected that where there is cracking in the interior

exposed walls, the level of ASR-related distress will be similar on the exterior wall surface." The

applicant also stated in its response to RAI B.2.1.31-9 dated November 2, 2012, that "since these

[accessible] surfaces have the least confinement, the expansion will be most pronounced on the

exposed surfaces."

Issue

In the absence of an evaluation for the acceptability of inaccessible areas, where conditions in

accessible areas indicate the presence of ASR, the staff cannot verify that the effects in

accessible areas would be the same for inaccessible areas. The rate of expansion on the exterior

surface of structures, some of which are directly exposed to groundwater, could be greater than

the observed planar ("x-y," or "in-plane") expansion on the interior surfaces of structures.

Although the ASR Monitoring Program was revised to include inspection of inaccessible areas

meeting the criteria described in the program, the staff is concerned that the opportunistic or

focused inspections may not be representative of inaccessible areas of accessible concrete

exhibiting the highest levels of ASR degradation.

Request

For inaccessible areas of concrete structures, state whether actions will be taken to confirm the

validity of the expectation that the magnitude of cracking due to expansion from reaction with

aggregates is not greater than that observed on the accessible surfaces. Include a discussion of

any planned actions. If not, explain how, absent an evaluation of inaccessible areas, the program

will ensure ASR is not more severe in inaccessible areas.

NextEra Energy Seabrook Response to RAI B.2.1.31A-6(a)

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. These

inspections will establish the baseline inspection for future monitoring and trending of

inaccessible areas of concrete. 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.
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SBK-L-15107 /Enclosure 1/Page 17

To date, NextEra has not observed ASR in-accessible areas greater than that observed in

accessible areas and does not expect to observe such expansion in the future. In general these

areas are not accessible because they are buried and have no accessible interior spaces. The

environmental conditions that affect ASR development are those related to alkali transport and

silica solubility. Temperature and humidity (in this case ground water) are the most significant.

The buried concrete will be subject to ground water on all sides. Most accessible areas have

groundwater on one side with an adjacent interior space. This arrangement allows for flow

through the concrete with an alternating wet dry surface on one side. This tends to facilitate

alkali transfer and higher ASR progression has been seen in these conditions as opposed to fully

or constantly wetted conditions.

With respect to ambient temperature, in general the higher the temperature the more soluble the

silica and the faster ASR will progress. The below grade accessible areas generally have a heated

interior space that means the concrete is warmer than the surrounding backfill material. The

inaccessible below grade concrete will essentially be at the constant cool temperature of the

surrounding backfill material. The ambient temperature and humidity conditions are no harsher

for ASR than the observable concrete and so the rates of ASR progression are bounded.

The same three Tier evaluation process and acceptance criteria specified in the ASR Monitoring

Program will be applied during the initial assessment of inaccessible areas of concrete. These

areas will also be monitored and trended at the same frequency as the accessible areas as

specified in the ASR Monitoring Program. In other words, if a Tier 2 or 3 ASR location is

identified during an opportunistic or focused inspection of an inaccessible area, the re-inspection

of this ASR location would be every 2.5 years for a Tier 2 location and every 6 months for a

Tier 3 location.

Several inaccessible areas have been inspected during the last 30 months and results to date have

confirmed instances where ASR is present. However, the levels of ASR observed have been

consistent with that observed in accessible areas of the plant. Typical inaccessible areas

inspected include underground electrical manholes, GSU transformer foundations, GSU

transformer containment structures, underground SW pipe access vault, and below grade backfill

concrete.

Based on the information above Element 10 of the ASR AMP is appended as follows:

An opportunistic Structures Monitoring inspection was conducted on the underground Unit 2

Circulating Water Pipe Access Vault on June 3rd, 2015. On opening the vault it was found to

have been flooded to approximately 15foot elevation. Overall the condition of the concrete

was found to be in good condition and deemed acceptable. Minor cracking was present

(mostly on the top surface concrete) but did not exceed 0.025 inches in width and appeared to

be shallow in depth for all notable instances. No visible map cracking, dark staining or gel

exudation indicative of ASR was noted.
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SBK-L-15107 /Enclosure I/Page 18

An opportunistic Structures Monitoring inspection was conducted on concrete structures

associated with the Generator Step-up Transformer Units (GSU) on March 19th, 2014 and

June 2 6th, 2014for "A" and "C", respectively. The inspections encompassed both the GSU

foundation and its respective containment structure. Both pit areas that were inspected

showed characteristics that are suggestive of ASR (map cracking, and dark staining). The

indications were noted on the inside face and top of the oil containment structure only. No

indications were noted on the foundations. The nominal width of the pattern cracking

appeared to be less than 2 mils (0.002 inches), the minimum measurable crack width that can

be reliably and accurately measured. Therefore, the ASR pattern cracking on top of the oil

containment walls are classified as ASR Tier 2 -Qualitative and being monitored on a 30-

month basis. The top of the south wall at 1-ED-X-1A (GSU) has a modified CCI grid (due to

size restrictions) and will be classified as ASR Tier 2- Quantitative and will be re-inspected on

a 30-month basis as well In addition, areas of spalled concrete were found on the GSU

foundations and were promptly remediated.

Enclosure 2 to SBK-L-15107
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- 15107 /Enclosure2/ Page 2

A.3 LICENSE RENEWAL COMMITMENT LIST

UFSAR

PROGRAM or TOPIC COMMITMENT SCHEDULE

LOCATION

Provide confirmation and acceptability of the implementation of

PWR Vessel Internals MRP-227-A by addressing the plant-specific Applicant/Licensee A.2.1.7 Complete

Action Items outlined in section 4.2 of the NRC SER.

Closed-Cycle Cooling Enhance the program to include visual inspection for cracking, loss

Water of material and fouling when the in-scope systems are opened for A.2.1.12 Prior to the period of extended maintenance. operation.

Inspection of Overhead

Heavy Load and Light Enhance the program to monitor general corrosion on the crane and Prior to the period of extended

Load (Related to trolley structural components and the effects of wear on the rails in A.2.1.13 operation.

Refueling) Handling the rail system.

Systems

Inspection of Overhead

Heavy Load and Light Prior to the period of extended

Load (Related to Enhance the program to list additional cranes for monitoring. A.2.1.13 operation.

Refueling) Handling

Systems

Compressed Air Enhance the program to include an annual air quality test requirement A.2.1.14 Prior to the period of extended

Monitoring for the Diesel Generator compressed air sub system. operation.

Fire Protection Enhance the program to perform visual inspection of penetration A.2.1.15 Prior to the period of extended seals by a fire protection qualified inspector. operation.

Enhance the program to add inspection requirements such as spalling, Prior to the period of extended

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

reaction with aggregates by qualified inspector.

Fire Protection Enhance the program to include the performance of visual inspection A.2.1.15 Prior to the period of extended of fire-rated doors by a fire protection qualified inspector. operation.
U.S. Nuclear Regulatory Commission

SBK-L-15107 /Enclosure2/ Page 3

Enhance the program to include NFPA 25 (2011 Edition) guidance

for "where sprinklers have been in place for 50 years, they shall be

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

shall be submitted to a recognized testing laboratory for field service

testing".

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

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

NFPA 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. Within ten years prior to the

11. Fire Water System These inspections will be documented and trended to determine if a A.2.1.16 period of extended operation.

representative number of inspections have been performed prior to

the period of extended operation. If a representative number of

inspections have not been performed prior to the period of extended

operation, focused inspections will be conducted. These inspections

will commence during the ten year period prior to the period of

extended operation and continue through the period of extended

operation.

Enhance the program to include 1) In-scope outdoor tanks, except fire

water storage tanks, constructed on soil or concrete, 2) Indoor large

volume storage tanks (greater than 100,000 gallons) designed to nearatmospheric internal pressures, sit on concrete or soil, and exposed

internally to water, 3) Visual, surface, and volumetric examinations of

12. Aboveground Steel the outside and inside surfaces for managing the aging effects of loss of A.2.1.17 Within 10 years prior to the

Tanks material and cracking, 4) External visual examinations to monitor period of extended operation.

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.
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SBK-L-15107 /Enclosure2/ Page 4

Enhance the program to perform exterior inspection of the fire water

storage tanks annually for signs of degradation and include an Within ten years prior to the

13. Fire Water System ultrasonic inspection and evaluation of the internal bottom surface of A.2.1.16 period of extended operation.

the two Fire Protection Water Storage Tanks per the guidance

provided in NFPA 25 (2011 Edition).

Enhance program to add requirements to 1) sample and analyze new

14. Fuel Oil Chemistry fuel deliveries for biodiesel prior to offloading to the Auxiliary Boiler A.2.1.18 Prior to the period of extended

fuel oil storage tank and 2) periodically sample stored fuel in the A operation.

Auxiliary Boiler fuel oil storage tank.

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

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

quarter and to remove water as necessary.

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

16. Fuel Oil Chemistry the diesel fire pump fuel oil day tanks on a frequency of at least once A.2.1.18 operation.

every ten years.

Enhance the program to require ultrasonic thickness measurement of

the tank bottom during the 10-year draining, cleaning and inspection Prior to the period of extended

17. Fuel Oil Chemistry of the Diesel Generator fuel oil storage tanks, Diesel Generator fuel A.2.1.18 operation.

oil day tanks, diesel fire pump fuel oil day tanks and auxiliary boiler

fuel oil storage tank.

Reactor Vessel Enhance the program to specify that all pulled and tested capsules, A.2.1.19 Prior to the period of extended

18. Surveillance unless discarded before August 31, 2000, are placed in storage. operation.

Enhance the program to specify that if plant operations exceed the

limitations or bounds defined by the Reactor Vessel Surveillance

Reactor Vessel Program, such as operating at a lower cold leg temperature or higher A.2.1.19 Prior to the period of extended

19. Surveillance fluence, the impact of plant operation changes on the extent of operation.

Reactor Vessel embrittlement will be evaluated and the NRC will be

notified.
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SBK-L- 15107 /Enclosure2/ Page 5

20.

Reactor Vessel

Surveillance

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.

A.2.1.19 Prior to the period of extended

operation.

Enhance the program to ensure that any capsule removed, without the

Reactor Vessel intent to test it, is stored in a manner which maintains it in a condition A.2.1.19 Prior to the period of extended

21. Surveillance which would permit its future use, including during the period of operation.

extended operation.

Within ten years prior to the 22. One-Time Inspection Implement the One Time Inspection Program. A.2. 1.20 peidoexnedprao. period of extended operation.

Implement the Selective Leaching of Materials Program. The

Selective Leaching of program will include a one-time inspection of selected components Within five years prior to the

23. Materials where selective leaching has not been identified and periodic A.2.1.21 period of extended operation.

inspections of selected components where selective leaching has been

identified.

Buried Piping And Implement the Buried Piping And Tanks Inspection Program. Within ten years prior to

24. Tnk I A.2.1.22 entering the period of extended

Tans Inspectionoprtn operation

tion Implement the One-Time Inspection of ASME Code Class I Small Within ten years prior to the

25, of ASME Code Class 1 Bore-Piping Program. A.2.1.23 period of extended operation.

Small Bore-Piping

Enhance the program to specifically address the scope of the

External Surfaces program, relevant degradation mechanisms and effects of interest, the Prior to the period of extended

26. Monitoring refueling outage inspection frequency, the training requirements for A.2.1.24 operation.

inspectors, and the required periodic reviews to determine program

effectiveness.
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SBK-L- 15107 /Enclosure2/ Page 6

Inspection of Internal

Surfaces in 27. Miscellaneous Piping Implement the Inspection of Internal Surfaces in Miscellaneous A.2.1.25 Prior to the period of extended

and Ducting Piping and Ducting Components Program. operation.

Components

Lubricating Oil Enhance the program to add required equipment, lube oil analysis Prior to the period of extended

28. Analysis required, sampling frequency, and periodic oil changes. A2.1.26 operation.

Lubricating Oil Enhance the program to sample the oil for the Reactor Coolant pump A.2.1.26 Prior to the period of extended

29. Analysis oil collection tanks. operation.

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

Lubricating Oil ultrasonic thickness measurement of the lower portion of the Reactor Prior to the period of extended

30. Analysis Coolant pump oil collection tanks prior to the period of extended A.2.1.26 operation.

operation.

ASME Section XI, Enhance procedure to include the definition of "Responsible A.2.1.28 Prior to the period of extended

31. Subsection IWL Engineer". operation.

Structures Monitoring Enhance procedure to add the aging effects, additional locations, A.2.1.31 Prior to the period of extended

32. Program inspection frequency and ultrasonic test requirements. operation.

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

Program planning excavation work that would expose inaccessible concrete. A.2.1.31 operation.
U.S. Nuclear Regulatory Commission

SBK-L- 15107 /Enclosure2/ Page 7

Electrical Cables and

Connections Not

Subject to 10 CFR Implement the Electrical Cables and Connections Not Subject to 10 Prior to the period of extended

50.49 Environmental CFR 50.49 Environmental Qualification Requirements program. A.2.1.32 operation.

Qualification

Requirements

Electrical Cables and

Connections Not

Subject to 10 CFR 50.49 Environmental Implement the Electrical Cables and Connections Not Subject to 10 Prior to the period of extended

35. Qualification CFR 50.49 Environmental Qualification Requirements Used in A.2.1.33 operation.

Requirements Used in Instrumentation Circuits program.

Instrumentation

Circuits

Inaccessible Power

Cables Not Subject to

10 CFR 50.49 Implement the Inaccessible Power Cables Not Subject to 10 CFR Prior to the period of extended

36. Environmental 50.49 Environmental Qualification Requirements program. A2.1.34 operation.

Qualification

Requirements

Prior to the period of extended

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

operation.

Prior to the period of extended

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

operation.
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SBK-L-15107 /Enclosure2/ Page 8

Electrical Cable

Connections Not

Subject to 10 CFR Implement the Electrical Cable Connections Not Subject to 10 CFR Prior to the period of extended

50.49 Enviromnental 50.49 Environmental Qualification Requirements program. A.2.1.37 operation.

Qualification

Requirements

40. 345 KV SF6 Bus Implement the 345 KV SF6 Bus program. A.2.2.1 Prior to the period of extended

operation.

Metal Fatigue ofPrototeprdofxend 4eac Cat Enhance the program to include additional transients beyond those Prior to the period of extended

PRessur Pressure Boundary Bound defined in the Technical Specifications and UFSAR. operation.

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

42. Reactor Coolant transients to monitor cumulative usage on selected components A.2.3. operation.

Pressure Boundary

Pressure -Temperature The updated analyses will be

Limits, including Low Seabrook Station will submit updates to the P-T curves and LTOP submitted at the appropriate time

43. Temperature limits to the NRC at the appropriate time to comply with 10 CFR 50 A.2.4.1.4 to comply with 10 CFR 50

Overpressure Appendix G. Appendix G, Fracture

Protection Limits Toughness Requirements.
U.S. Nuclear Regulatory Commission

SBK-L-15107 /Enclosure2/ Page 9

44.

EnvironmentallyAssisted Fatigue

Analyses (TLAA)

NextEra Seabrook will perform a review of design basis ASME Class

1 component fatigue evaluations to determine whether the

NUREG/CR-6260-based components that have been evaluated for

the effects of the reactor coolant environment on fatigue usage are the

limiting components for the Seabrook plant configuration. If more

limiting components are identified, the most limiting component will

be evaluated for the effects of the reactor coolant environment on

fatigue usage. If the limiting location identified consists of nickel

alloy, the environmentally-assisted fatigue calculation for nickel alloy

will be performed using the rules of NUREG/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 factors to valid CUFs determined from an existing

fatigue analysis valid for the period of extended operation or from an

analysis using an NRC-approved version of the ASME code or NRCapproved alternative (e.g., NRC-approved code case).

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

locations, then additional plant-specific locations will be evaluated.

For the additional plant-specific locations, if CUF, including

environmental effects is greater than 1.0, then Corrective Actions will

be initiated, in accordance with the Metal Fatigue of Reactor Coolant

Pressure Boundary Program, B.2.3. 1. Corrective Actions will

include inspection, repair, or replacement of the affected locations

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

managed by an inspection program that has been reviewed and

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

the affected locations at inspection intervals to be determined by a

method accepted by the NRC).

A.2.4.2.3

At least two years prior to

entering the period of extended

operation.
U.S. Nuclear Regulatory Commission

SBK-L- 15107 /Enclosure2/ Page 10

45. Number Not Used

Protective Coating Enhance the program by designating and qualifying an Inspector Prior to the period of extended

46. Monitoring and Coordinator and an Inspection Results Evaluator. A.2.1.38 operation.

Maintenance

Enhance the program by including, "Instruments and Equipment

Protective Coating needed for inspection may include, but not be limited to, flashlight, Prior to the period of extended

47. Monitoring and spotlights, marker pen, mirror, measuring tape, magnifier, binoculars, A.2.1.38 operation.

Maintenance camera with or without wide angle lens, and self sealing polyethylene

sample bags."

Protective Coating Enhance the program to include a review of the previous two Prior to the period of extended

48. Monitoring and monitoring reports. operation.

Maintenance

Protective Coating Enhance the program to require that the inspection report is to be

49. Monitoring and evaluated by the responsible evaluation personnel, who is to prepare A.2.1.38 Prior to the period of extended

Maintenance a summary of findings and recommendations for future surveillance operation.

or repair.

Baseline inspections were

completed during OR16. Repeat

ASMEPerform UT of the accessible areas of the containment liner plate in containment liner UT thickness

50. Su Section the vicinity of the moisture barrier for loss of material. Perform A.2.1.27 opportunistic UT of inaccessible areas.at intervals of no

more than five (5) refueling

outages.

51. Number Not Used

ASME Section XI, Implement measures to maintain the exterior surface of the

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

Reactor Head Closure Replace the spare reactor head closure stud(s) manufactured from the Prior to the period of extended

53. Studs bar that has a yield strength > 150 ksi with ones that do not exceed A.2.1.3 operation.

150 ksi.
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SBK-L- 15107 /Enclosure2/ Page 11

54.

Steam Generator Tube

Integrity

NextEra will address the potential for cracking of the primary to

secondary pressure boundary due to PWSCC of tube-to-tubesheet

welds using one of the following two options:

1) Perform a one-time inspection of a representative sample of tubeto-tubesheet welds in all steam generators to determine if PWSCC

cracking is present and, if cracking is identified, resolve the condition

through engineering evaluation justifying continued operation or

repair the condition, as appropriate, and establish an ongoing

monitoring program to perform routine tube-to-tubesheet weld

inspections for the remaining life of the steam generators, or

2) Perform an analytical evaluation showing that the structural

integrity of the steam generator tube-to-tubesheet interface is

adequately maintaining the pressure boundary in the presence of

tube-to-tubesheet weld cracking, or redefining the pressure boundary

in which the tube-to-tubesheet weld is no longer included and,

therefore, is not required for reactor coolant pressure boundary

function. The redefinition of the reactor coolant pressure boundary

must be approved by the NRC as part of a license amendment

request.

A.2. 1.10 Complete

Within five years prior to

Steam Generator Tube Seabrook will perform an inspection of each steam generator to A.2.1.10 entering the period of extended

55. Integrity assess the condition of the divider plate assembly. operation.

Closed-Cycle Cooling Revise the station program documents to reflect the EPRI Guideline A.2.1.12 Prior to entering the period of

56. Water System operating ranges and Action Level values for hydrazine and sulfates. extended operation.

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

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

Cooling Water Jacket pH.

Update Technical Requirement Program 5.1, (Diesel Fuel Oil Testing Prior to the period of extended

58. Fuel Oil Chemistry Program) ASTM standards to ASTM D2709-96 and ASTM D4057- A.2.1.18 operation.

95 required by the GALL XI.M30 Rev 1
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Nickel Alloy Nozzles The Nickel Alloy Aging Nozzles and Penetrations program will Prior to the period of extended

59 implement applicable Bulletins, Generic Letters, and staff accepted A.2.2.3 operation. .and Penetrations industry guidelines.

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

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

detection capability.

Compressed Air Replace the flexible hoses associated with the Diesel Generator air Within ten years prior to

6.A.2.1.14 entering the period of extended

61. Monitoring Program compressors on a frequency of every 10 years. operation.

Proprationteprido.etne

62. Water Chemistry Enhance the program to include a statement that sampling A.2.1.2 Prior to the period of extended

frequencies are increased when chemistry action levels are exceeded. operation.

Ensure that the quarterly CVCS Charging Pump testing is continued

during the PEO. Additionally, add a precaution to the test procedure Prior to the period of extended

63. Flow Induced Erosion to state that an increase in the CVCS Charging Pump mini flow N/A prratote

above the acceptance criteria may be indicative of erosion of the mini operation.

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 of non-cathodically protected steel pipe within the scope of A.2.1.22 Prior to entering the period of

Tanks InspedcPipionga this program. If the initial analysis shows the soil to be non- extended operation. Tanks Inspection corrosive, this analysis will be re-performed every ten years

thereafter.

65. Flux Thimble Tube Implement measures to ensure that the movable incore detectors are Prior to entering the period of

not returned to service during the period of extended operation. N/A extended operation.

66. Number Not Used
U.S. Nuclear Regulatory Commission

SBK-L-15107 /Enclosure2/ Page 13

Perform one shallow core bore in an area that was continuously

wetted from borated water to be examined for concrete degradation

Structures Monitoring and also expose rebar to detect any degradation such as loss of No later than December 31,

67. Program material. The removed core will also be subjected to petrographic A2131 2015.

examination for concrete degradation due to ASR per ASTM

Standard Practice C856.

Quarterly Preventive

Structures Monitoring Perform sampling at the leakoff collection points for chlorides, A.2.1.31 Maintenance Activity

68. Program sulfates, pH and iron once every three months. Implemented

Open-Cycle Cooling Replace the Diesel Generator Heat Exchanger Plastisol PVC lined A.2.1.11

69. Water System Service Water piping with piping fabricated from AL6XN material. Complete.

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

Closed-Cycle Cooling determine whether the loss of material due to cavitation induced Within ten years prior to the

70. Water System erosion has been eliminated or whether this remains an issue in the A.2.1.12 period of extended operation.

primary component cooling water system.

Alkali-Silica Reaction Implement the Alkali-Silica Reaction (ASR) Monitoring Program. 71. (ASR) Monitoring Testing will be performed to confirm that parameters being Prior to entering the period of

7ro ntrg monitored and acceptance criteria used are appropriate to manage the extended operation.

Program effects of ASR.

Flow-Accelerated Enhance the program to include management of wall thinning caused A.2.1.8 Prior to entering the period of

72. Corrosion by mechanisms other than FAC. extended operation.

Inspection of Internal Enhance the program to include performance of focused

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

73. Miscellaneous Piping maximum of 25, of each identified material, environment, and aging A.2.1.25 Prior to entering the period of

and Ducting effect combinations during each 10 year period in the period of extended operation.

Components extended operation.
U.S. Nuclear Regulatory Commission

SBK-L- 15107 /Enclosure2/ Page 14

Enhance the program to perform sprinkler inspections annually per

the guidance provided in NFPA 25 (2011 Edition). Inspection will

ensure that sprinklers are free of corrosion, foreign materials, paint,

Fi Water System and physical damage and installed in the proper orientation (e.g., Prior to the period of extended

upright, pendant, or sidewall). Any sprinkler that is painted, corroded, operation.

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 NFPA 25 (2011 Edition) and b) If the presence Prior to the period of extended

75. Fire Water System of sufficient foreign organic or inorganic material to obstruct pipe or A.2.1.16 operation.

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 NFPA 25 (2011 Edition), Section 14.2.2.

Enhance the Program to conduct the following activities annually per

the guidance provided in NFPA 25 (2011 Edition). Prior to the period of extended

76. Fire Water System * main drain tests A.2..16 operation.

* deluge valve trip tests

* fire water storage tank exterior surface inspections
U.S. Nuclear Regulatory Commission

SBK-L- 15107 /Enclosure2/ Page 15

The Fire Water System Program will be enhanced to include the

following requirements related to the main drain testing per the

guidance provided in NFPA 25 (2011 Edition).

77. Fire Water System * The requirement that if there is a 10 percent reduction in full flow A.2.1.16 Prior to the period of extended

pressure when compared to the original acceptance tests or operation.

previously performed tests, the cause of the reduction shall be

identified and corrected if necessary.

0 Recording the time taken for the supply water pressure to return

to the original static (nonflowing) pressure.

Enhance the program to include periodic inspections of in-scope

insulated components for possible corrosion under insulation. A

External Surfaces sample of outdoor component surfaces that are insulated and a Prior to the period of extended 78. Monitoring sample of indoor insulated components exposed to condensation (due A.2.1.24 operation.

to the in-scope component 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 7. Open-Cycle Cooling cotns.nng oso cotn .nert. A...1 Within 10 years prior to the

Water System coatings/linings for loss of coating integrity. A.2.1.11 period of extended operation.

Enhance the program to include visual inspection of internal

80. Fire Water System coatings/linings for loss of coating integrity. A.2.1.16 period Within of 10 extended yearsdprioretotthe operation.

Enhance the program to include visual inspection of internal

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

Inspection of Internal

Surfaces in Enhance the program to include visual inspection of internal

82. Miscellaneous Piping coatings/linings for loss of coating integrity. A.2.1.25 Within 10 years prior to the

and Ducting period of extended operation.

Components
U.S. Nuclear Regulatory Commission

SBK-L- 15107 /Enclosure2/ Page 16

install instrumcntation in repr-esentative sample areas of stcrut-es to

monitor qexpanfsion due to ailkai silica reactioni in the out of plane

direction. Evaluate instr ument and pin expansion data under- the

Operating Exper-ience Elemnent of the Alkali Silica Reaction

Monitoring Program to determiniewhether there is a need t enhance

the progr-am to monitor- expansion in the out of plane difection. if the

evalutationi concludes that out of plane monitor-ing is neeesaý,,

establish acceptance cr-iter-ia and monitoring frequencies for

expansion in the out of plane direction using the instrument and pin

83. Alkali-Silica Reaction

Monitoring

expansion data.

Install extensometers in representative sample areas of structures

to monitor expansion due to alkali-silica reaction in the out-ofplane direction. Install at least thirty-four (34) extensometers by

December 31, 2016. Using the Operating Experience Element of

the ASR Monitoring Program, determine if increase in sample size

is required.

Monitoring expansion in the out-of-plane direction will commence

upon installation of the extensometers in 2016 and continue

through the period of extended operation.

A.2.1.3 1A

Prior to the periodl o! extended

operation.7

December 31, 2016.

84. ASME Section XI, Evaluate the acceptability of inaccessible areas for structures within A.2.1.28 Prior to the period of extended Subsection IWL the scope of ASME Section XI, Subsection IWL Program. operation.

Enhance the program to perform additional tests and inspections on

the Fire Water Storage Tanks as specified in Section 9.2.7 of NFPA

25 (2011 Edition) in the event that it is required by Section 9.2.6.4, A2.16 Prior to the period of extended

85. Fire Water System which states "Steel tanks exhibiting signs of interior pitting, operation.

corrosion, or failure of coating shall be tested in accordance with

9.2.7."

Enhance the program to include disassembly, inspection, and A.2.1.16 Prior to the period of extended 86. Fire Water System cleaning of the mainline strainers every 5 years. operation.
U.S. Nuclear Regulatory Commission

SBK-L- 15107 /Enclosure2/ Page 17

87. Fire Water System

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.

A.2.1.16

Prior to the period of extended

operation.

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 Within five years prior to the

88. Fire Water System drains and associated piping are configured to completely drain the A.2.1.16 Witin fiexyea prior t

piping, and c) normally-dry piping that could have been wetted by

inadvertent system actuations or those that occur after a fire are

restored to a dry state as part of the suppression system restoration.

Inspection of Internal Incorporate Coating Service Level III requirements into the RCP

Surfaces in Motor Refurbishment Specificationfor the internal painting of the

89. Miscellaneous Piping motor upper bearing coolers and motor air coolers. All four RCPs A.2.1.25 prratote

and Ducting will be refurbished and replaced using the Coating Service Level III

Components requirements prior to entering the period of extended operation.

Implement the PWR Vessel Internals Program. The program will be

90. PWR Vessel Internals implemented in accordance with MRP-227-A (Pressurized Water A.2.1.7 Prior to the period of extended

Reactor Internals Inspection and Evaluation Guidelines) and NEI 03- operation

08 (Guideline for the Management of Materials Issues).

Alkali-Silica Reaction In building geometry locations where deformation is likely, Within 10 years prior to the

91. Monitoring enhance the program to monitor for building deformation using A.2.1.31A period of extended operation.

laser targets and by taking gap measurements.