UNITED STATES

NUCLEAR REGU LATORY COMMISSION

REGION I

475 ALLENDALE ROAD

KING OF PRUSSIA. PA 19406.1415

l4arch 26, 2012

Mr. Paul Freeman

Site Vice President, North Region

Seabrook Nuclear Power Plant

NextEra Energy Seabrook, LLC

c/o Mr. Michael O'Keefe

P.O. Box 300

Seabrook, NH 03874

SUBJECT: SEABROOK STATION - NRC INSPECTION REPORT 05000443/2011010

RELATED TO ALKALI-SILICA REACTION ISSUE IN SAFETY RELATED

STRUCTURES

Dear Mr. Freeman:

On January 20,2012, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection

at Seabrook Station. The enclosed inspection report documents the inspection results, which

were discussed at the exit meeting with you and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and

compliance with the Commission's rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed

personnel. In conjunction with the follow-up of two unresolved items, the focus of this inspection

was a review of activities involving NextEra's analysis and evaluation related to addressing the

Alkali-Silica Reaction (ASR) issue occurring in safety related and other important to safety

concrete structures. As a part of this inspection, we reviewed your original and revised Prompt

Operability Determinations (POD) for certain affected structures.

During the exit meeting, Mr. Richard J. Conte, Chief Engineering Branch 1, summarized the

findings and observations. In addition, he discussed NRC observations regarding your planned

correCtive actions and assumptions being made in the NextEra operability determinations. The

inspectors conctuded that these structures can currently perform their safety related functions

despite the observed degradation due to ASR. However the NRC still has concerns associated

with long term operability, therefore additional information is needed to determine: 1) how

various characteristics of the concrete may be affected by ASR; 2) the related effects on other

elements of the structures, such as rebar, due to groundwater in-leakage; and 3) the rate of

progression of the ASR in structures at the site. lt is our understanding that these specific areas

are being addressed in a comprehensive corrective action plan that was still being finalized by

your organization at the end of the inspection.

Therefore, we request that you summarize your plans to address the above issues at a

management meeting to be conducted April 23, 2012, at NRC Headquarters in Rockville, MD.

At the meeting you should be prepared to focus on the following technical issues: 1) describe

which applicable American Concrete Institute (ACl) 318 code relationships are affected by ASR
P. Freeman

and your plans to ensure the applicable licensing and design bases remain valid; 2) describe

your comprehensive plans to understand the related effects and overall progression of ASR, its

cause, and actions to correct and/or mitigate the issue; and, 3) provide a timeline for key

actions, including those to address long term operability, how the degradation affects the design

basis, and longer term management of the ASR issue. During the meeting we will discuss your

overall corrective action plans, including the documents to be submitted to the NRC on the

docket.

Also, the report documents two NRC-identified findings of very low significance (Green) one of

which involved a violation of NRC requirements, Because of the very low safety significance,

and because they are entered into your corrective action program, the NRC is treating these

findings as non-cited violations, consistent with Section 2.3.2 of the NRC Enforcement Policy. lf

you contest any non-cited violations in this report, you should provide a response within 30 days

of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear

Regulatory Commission, ATTN.: Document Control Desk, Washington, DC 20555-0001; with

copies to the RegionalAdministrator, Region l; the Director, Office of Enforcement, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident lnspector at

Seabrook Station. In addition, if you disagree with the cross-cutting aspect assigned to any

finding in this report, you should provide a response within 30 days of the date of this inspection

report, with the basis for your disagreement, to the Regional Administrator, Region l, and the

NRC Resident Inspector at Seabrook.

ln accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390 of the NRC's

"Rules of Practice," a copy of this letter, its enclosure, and your response (if any), will be

available electronically for public inspection in the NRC Public Document Room or from the

Publicly Available Records component of the NRC's document system, Agencyrvide Documents

Access and Management System (ADAMS). The ADAMS is accessible from the NRC Web site

at http:/lwww.nrc.govlreading-rmladams.html (the Public Electronic Reading Room).

Sincerely, ,aA Christopher G. Miller, Director

Division of Reactor Safety

Docket No.: 50-443

License No.: NPF-86

Enclosure:

lnspection Report No. 050004431201 1010

w/Attachment: Supplemental Information

cc w/encl: Distribution via ListServ
P. Freeman 2

and your plans to ensure the applicable licensing and design bases remain valid; 2) describe

your comprehensive plans to understand the related effects and overall progression of ASR, its

cause, and actions to correct and/or mitigate the issue; and, 3) provide a timeline for key

actions, including those to address long term operability, how the degradation affects the design

basis, and longer term management of the ASR issue. During the meeting we will discuss your

overall corrective action plans, including the documents to be submitted to the NRC on the

docket.

Also, the report documents two NRC-identified findings of very low significance (Green) one of

which involved a violation of NRC requirements. Because of the very low safety significance,

and because they are entered into your corrective action program, the NRC is treating these

findings as non-cited violations, consistent with Section 2.3.2 of the NRC Enforcement Policy. lf

you contest any non-cited violations in this report, you should provide a response within 30 days

of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear

Regulatory Commission, ATTN.: Document Control Desk, Washington, DC 20555-0001; with

copies to the Regional Administrator, Region l; the Director, Office of Enforcement, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident lnspector at

Seabrook Station. ln addition, if you disagree with the cross-cutting aspect assigned to any

finding in this report, you should provide a response within 30 days of the date of this inspection

report, with the basis for your disagreement, to the Regional Administrator, Region l, and the

NRC Resident Inspector at Seabrook.

ln accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390 of the NRC's

"Rules of Practice," a copy of this letter, its enclosure, and your response (if any), will be

available electronically for public inspection in the NRC Public Document Room or from the

Publicly Available Records component of the NRC's document system, Agency,vide Documents

Access and Management System (ADAMS). The ADAMS is accessible from the NRC Web site

at http://www.nrc.gov/reading-rm/adams.html(the Public Electronic Reading Room).

Sincerely,

/RN

Christopher G. Miller, Director

Division of Reactor Safety

Docket No.: 50-443

License No.: NPF-86

Enclosure:

lnspection Report No. 050004 431201 1O1O

w/Attachment: Supplemental Information

cc w/encl: Distribution via ListServ

DISTRIBUTION: See Next Page ADAMS ACCESSION NO. M1120480066

SUNSI Review Complete: RJC (Reviewer's lnitials)

DOCUMENT NAME: ClMyFiles\Checkout\o5000443 2011010 Seabrook StandaloneforASR lssueFlNAL.dOCX

After declaring this document "An Official Agency Record" it will be released to the Public.

To receive a copy of without

OFFICE RI/DRS I RI/DRS I RI/DRP RI/DRS RI/DRS

NAME MModes/HG WSchmidVCGC ABurritt RConte CMiller

DATE 2t23t12 2t23t12 3t23t12 3t26t12 3t26t12

OFFICIAL RECORD COPY

"N"=No
P. Freeman

DISTRIBUTION w/encl: (via e-mail)

W. Dean, RA

D. Lew, DRA

J. Tappert, DRP

J. Clifford, DRP

C. Miller, DRS

P. Wilson, DRS

A. Burritt, DRP

L. Cline, DRP

A. Turilin, DRP

R. Montgomery, DRP

W. Raymond, DRP, SRI

J. Johnson, DRP, Rl

A. Cass, DRP, Resident OA

L. Chang, Rl, OEDO

RidsNrrPMSeabrook Resource

Rids N rrDorlLpll -2 Resou rce

ROPreports Resource

R. Conte, DRS

M. Modes, DRS
Docket No.:

License No.:

Report No.:

Licensee:

Facility:

Location:

Dates:

Inspectors:

Accompanied by:

Approved by:

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

50-443

NPF-86

05000443/2011010

NextEra Energy Seabrook, LLC

Seabrook Station

Seabrook, NH 03874

September 26-September 30, 2011

November 15-17, 201 1 (Northbrook, lllinois)

November 28-December 1, 2011

January 20,2012 (Conference Call)

M. Modes, Senior Reactor lnspector, Region I

S. Chaudhary, Reactor Inspector, Region I

W. Raymond, Senior Resident Inspector, Seabrook

Atif Shaikh, Reactor Inspector, Region lll

A. Sheikh, Senior Structural Engineer, Office of Nuclear

Reactor Regulation (NRR)

G. Thomas, Structural Engineer, NRR

Richard J. Conte, Chief

Engineering Branch 1

Division of Reactor Safety

Enclosure
SUMMARY OF FINDINGS

lR 0500044312011010;91251201 1 - 121212011; Seabrook Station (Problem ldentification and

Resolution; Follow-up to Operability and Plant Modifications).

This report covers an inspection by regional inspectors and resident staff, with assistance from

the Office of Nuclear Reactor Regulation (NRR) structural specialists. Two Green findings were

identified. The significance of most findings is indicated by their color (Green, White, Yellow,

Red) using lnspection Manual Chapter (lMC) 0609, "Significance Determination Process"

(SDP). The cross-cutting aspects for the findings were determined using IMC 0310,

"Components Within Cross-Cutting Areas." Findings for which the SDP does not apply may be

Green, or be assigned a severity levelafter NRC management review. The NRC's program for

overseeing the safe operation of commercial nuclear power reactors is described in

NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006.

Cornerstone: Mitigating Systems

Green. The inspectors identified a finding in that NextEra failed to fully evaluate potential

structural and seismic response impacts in accordance with the requirements in NextEra

procedure EN-AA-1001 after identifying a degraded and nonconforming condition related to

degraded conditions for some safety related structures due to Alkali-Silica Reaction (ASR).

Specifically, the evaluation did not consider the following effects due to changed properties of

concrete, is reflected in reduced values of the modulus of elasticity as measured directly from

concrete core samples: 1) building natural frequency in the dynamic response; 2) performance

of anchorages and embedment of systems and components attached to the structures; and,

3) shear strength or capacity of affected structures and the dynamic/flexural response especially

those buildings without corresponding shear reinforcement'

The failure to conduct adequate prompt operability determinations per procedure

EN-AA-203-1001 for degraded and nonconforming conditions associated with ASR was a

performance deficiency relative to a self imposed standard. Specifically, the prompt operability

determinations conducted following the identification of ASR in safety-related structures did not

completely analyze the effects of the reduced modulus of elasticity on the dynamic and flexural

response of the structures to seismic events for certain conditions. This performance deficiency

was associated with the design control aspect of the Mitigating Systems cornerstone; and,

based on a comparison to Example 3.i of Appendix E of IMC 0612, it was determined to be

more than minor. Specificatty, the failure to conduct adequate operability determinations

adversely affected the Mitigating Systems cornerstone objective to ensure the availability,

reliability, and capability of systems that respond to initiating events to prevent undesirable

consequences because it required an additional evaluation to confirm that the design bases was

met. The issue was evaluated using IMC 0609, "significance Determination Process," and was

determined to be of very low safety significance (Green). Specifically, when evaluated under

IMC 0609, Attachment 4, the performance deficiency was a design or qualification deficiency

confirmed not to result in an actual loss of safety function. The finding had a cross cutting

aspect in the area of problem identification and resolution, P.1(c), related to ensuring that issues

potentially impacting nuclear safety are thoroughly evaluated. Specifically, NextEra did not fully

evaluate conditions adverse to quality, including evaluating the effects of the reduced concrete

modulus of elasticity for impact on operability of the affected structures. (Section 4OA5.1.c)

Enclosure
Severitv Level lV. The inspectors identified a Severity Level lV non-cited violation (NCV) of

Title 10 of the Code of Federal Regulations (10 CFR) 50.59(dX1), "Changes, Tests, and

Experiments," because NextEra did not adequately evaluate a "use-as-is" determination,

resulting in a defacto design change, for certain ASR impacted safety related structures.

Specifically, NextEra did not complete a 10 CFR 50.59 evaluation, to ensure that the identified

reduction in concrete modulus of elasticity did not present a more than minimal increase in the

likelihood of the occurrence of a malfunction of a structure, system, or component (SSC)

important to safety previously evaluated in the updated safety analysis report (USAR) prior to

implementing changes to the facility as described in the engineering change EC272057 issued

on April 25,2011.

The failure to evaluate changes to the facility as described in EC272057 was contrary to

10 CFR 50.59(dX1) and was a performance deficiency warranting a significance evaluation in

accordance with the NRC Enforcement Manualfor Traditional Enforcement and IMC 0612,

"Power Reactor Inspection Reports," Appendix B, "lssue Disposition Screening." The violation

was determined to be more than minor in accordance with IMC 0612, "Power Reactor

Inspection Reports," Appendix B, "lssue Screening," because it could not reasonably be

determined that the changes would not have ultimately required prior NRC approval. In

accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this violation is categorized as

Severity Level lV because the resulting changes were evaluated by the SDP as having very low

safety significance (Green), because it was a design or qualification deficiency confirmed not to

result in an actual loss of safety function and because further evaluation determined that the

structures remained operable despite the degraded modulus condition. The finding had a cross

cutting aspect in the area of human performance - work practices, H.4(b), because NextEra

personnel did not follow procedures. Specifically, NextEra personnel did not follow the

requirements of Section 5.2.2 of the 5059 Resource Manual when preparing the 50.59 screen

tor EC272057. (Section 4OA5.2.c)

iii Enclosure
REPORT DETAILS

Backqround

In June 2009, NextEra conducted walk downs of structures within the scope of license renewal

as part of license renewal application preparations. ln June 2010, the License Renewal

Application (LRA) was received by the agency. ln October 2010, the NRC staff noted that the

licensee was beginning to formulate actions associated with both finalizing the operability

determination for the control building (CB) and starting an extent of conditions review of other

areas that may be subject to the alkali-silica reaction (ASR) degradation.

The ASR is a chemical reaction in concrete, which occurs over time in the presence of water,

between the alkaline cement paste and reactive non-crystalline silica that is found in some

common coarse aggregates. In the presence of water, the ASR for.ms a gel that expands,

causing micro-craclis that change the physical structural propertiesl of the concrete, including

compressive and tensile strength, modulus of elasticity, and Poisson Ratio. At Seabrook the

below-grade concrete structures have experienced groundwater infiltration.

ln the summer of 2010, NextEra performed an lmmediate and Prompt Operability Determination

(POD) for the CB "B" electrical tunnel structure based on core samples taken from the building,

lnspection Report 05000443/2010004, issued November 1, 2010, documented the NRC review

of the POD with no findings.

On May 12,2011, Inspection Report 05000443/2011002 identified two non-cited violations

(NCV) of very low safety significance related to maintenance rule (Title 10 of the Code of

Federal Regulations (10 CFR) 50.65 a(1) and b(2)) monitoring of structures. One of the NCVs

related NextEra's failure to properly monitor the structural performance of the CB resulting in

degraded conditions - 10 CFR 50.65 (aX1) (NCV 2011-002-01). Also in May 2011, License

Renewal Inspection Report 05000443/2011007 (1P71002) reflected an overall inspection result

as follows: "Except for Structures Monitoring Program, results support a reasonable assurance

determination for license renewal." The structure monitoring program had not addressed the

ASR condition.

On August 12, 2011, lnspection Report 05000443/2011003, identified a NCV of very low safety

significance related to the untimely operability determinations regarding the extent of condition

review for other buildings affected by ASR. The report also identified two unresolved items

(URl) related to the operability determinations. Specifically the report identified: 1) the need for

additional information related to open operability determinations, one for the CB "8" electrical

tunnel, and the other operability determinations for the extent of conditions review for five other

areas/structures with evidence of ASR (URl 201 1-003-03); and,2) potential inadequate

screening in accordance with 10 CFR 50.59 for accepting the reduced values found on

compressive strength and modulus of elasticity for the "B" Electrical Tunnel and the

Containment Enclosure Building (URl 201 1-003-02).

1 Material properties defined in the supplemental section of this report

Enclosure
2

ln September 2011, Region I obtained assistance from the Office of Nuclear Reactor Regulation

(NRR) through a Task Interface Agreement (TlA) in order to assist in the review of the open

PODs.

4. OTHER ACTIVITIES

4OI2 Problem ldentification and Resolution (71152- 1 sample)

Annual Sample: Corrective Actions Associated with Alkali-Silica Reaction in Safetv

Related Structures

Inspection Scope

This review was to assess progress in the development of a corrective plan and

implementing schedule to address the ASR degradation issue including: initial

assessments of all buildings potentially affected by the problem; root or apparent cause

of the problem; control of in-situ testing such as crack mappingiindexing; control of

contractor testing and laboratory test facilities in accordance with quality assurance

requirements; and any mitigation or long term monitoring actions. The inspectors

reviewed laboratory testing to address the ASR degradation with specific focus on the

CB ("8" Electrical Tunnel). Laboratory testing was observed during the week of

November 14,2011, to ensure proper sample controls, test preparation, and conduct of

the test.

During the week of November 28,2011, the inspectors reviewed historical

documentation from the construction phase of the plant, correlations between the

concrete strength value determined by the recent core samples, and the original strength

values determined at the time of concrete placement. The licensee's projected plan and

schedule for further studies and assessment of the ASR problem were discussed and

reviewed with cognizant engineering and management personnel. lnspectors also

reviewed the licensee's control of contractors and laboratory facilities used to analyze

concrete core samples. The inspectors reviewed the licensee's procedures for

administration and control of engineering and testing service vendors and contactors.

Additionally, the inspector reviewed the results and documentation of American Society

of Mechanical Engineers (ASME) Code Section IWL inspection of the containment.

Findinqs and Observations

No findings were identified. The inspector noted that a comprehensive corrective action

was still under development. NextEra classified this issue as a significant condition

adverse to quality and was in the progress of completing a root cause analysis, which

was scheduled to be completed in February 2012 in order to support an Engineering

Evaluation in March 2012. The inspectors noted that NextEra's plans to date did not

address some key issues related to ASR that include but are not limited to:

a.

b.

Enclosure
(1)

3

Need for other concrete core testing (i.e., low stress range stiffness damage

tests) to assess expansion-to-date or severity of degradation in the critical

direction of the thickness with no rebar ties and lesser resistance to expansion;

Basis for the representativeness of concrete core sampling in the buildings for

those taken to date and those to be taken, should they occur;

lmpact of core boring and re-grouting on the building structural integrity; and

Potential effects of other degradation mechanisms from an "aggressive"

groundwater environment along with the presence of ASR.

Methods Used in Evaluatinq Structural Inteqritv

The NRC staff noted that the methods used in evaluating structural integrity for the

selected buildings were based on the correct design basis code ACI 318-1971.

However, the mathematical relationships in this code were based on empiricaldata, from

testing of non-degraded concrete, for determining key ratios that are a part of the design

bases and used for determining tensile and shear strength or capacity in addition to

compressive strength. These strength values were important in the building loading

analysis during normal or upset conditions such as for seismic events. More importantly,

while some testing for the modulus of elasticity was done, it was not clear if the plans

would result in additional testing of concrete cores for this parameter or any independent

testing associated with other key design parameters such as Poisson's Ratio, shear

modulus, or bulk modulus. With these parameters known, various strengths or

capacities can be determined such as for tensile and shear strength. In addition, the

plans that the inspectors reviewed did not address variation in mechanical properties of

the concrete in different directions due to ASR cracking nor the effect of the ASR

expansion on stresses in the rebar. These parameters were important in order to ensure

that the current licensing and design basis was maintained. The licensee

representatives agreed to address the assumptions or establish relationships for the

current conditions at Seabrook. Accordingly this area is unresolved pending completion

of license actions as noted above and further NRC staff review

(URf 05000 4/.312011 01 0-01, Corrective Actions Associated with calculation

methods used to address the ASR lssue)

Control of ContractorsA/endors and Laboratorv Testinq

The reviewers noted that NextEra had engaged knowledgeable vendors, appropriate

consultants, and experts for testing, analysis, and evaluation of the effects of ASR on the

serviceability and safety of the affected structures. Also, during the week of

November 14,2011, a Region lll inspectorreviewed laboratorytestingforcompressive

strength on 15 concrete core samples taken from the CB "8" electricaltunnel in the

October 2011 time frame. This testing was being completed to resolve discrepant

information for compressive strength testing between two different contractors.

The testing was conducted at a laboratory in Northbrook, lllinois. All 15 core samples

were compression tested. Photographs were taken for all core samples prior to loading

for compression test and after fracture. Three cores had small length samples cut from

them to be used by Seabrook for further petrography. Sample preparation (capping)

was done in accordance with American Society for Testing and Materials (ASTM) C617.

Enclosure

(2)

(s)

(4)
40A5

4

Compression testing was done in accordance with ASTM C39. With respect to

laboratory conditions for testing of concrete cores, the inspector verified: 1) organized

and clean working area during both sample preparation (measurements and cutting)and

compression testing; 2) adequate lighting available at all times; 3) ambient room

temperature (- 68"F) observed during preparation and testing; and 4) core samples

were adequately stored and labeled in individual bags.

The inspector observed the care taken to ensure only one core was handled at any

given time so as not to confuse cores during measurements, cutting, and testing. With

respect to equipment calibration, the inspector verified proper equipment documentation

and calibration. With respect to test technician qualifications, the inspector also verified

qualification records. The inspector also reviewed the Altran Commercial Grade

Dedication Plan.

No concerns were noted with respect to quality control during all aspects of compression

testing. All 15 destroyed cores were shipped back to Seabrook including the cut

samples to be used for petrography. These results were to be evaluated by NextEra.

Other Activities

(Open) Unresolved ltem 05000443/2011003-03. Open Operabilitv Determinations for

Safetv-Related Stru ctu res Affected bv Al kal i-Sil ica Reaction

Insoection Scope

The NRC staff reviewed NextEra actions to develop finalized operability determinations

along with the review for extent of conditions. The review included the open aspects as

documented in the originating inspection report for which NextEra was to provide

additional information related to: 1) effect of the reduced modulus of elasticity on natural

frequency of the structures (applied to CB - "8" Electrical tunnel and other structures

being evaluated in the extent of conditions review such as for the Containment

Enclosure Building (CEB); 2) the effect of the modulus of elasticity on structure flexural

response as related to components attached to the structures, such as pipe and cable

trays supports and their anchor bolts; 3) related effects from increased flexure of building

on the loading and seismic effects on safety related pipes and cable tray supports; and,

4) effect of reduced parameters on the whole building (global) response of the CEB

structure to seismic loads including further information of the effect on stress and strain

in the concrete and rebar system. With respect to numbers 1 and 2 above, the

inspectors reviewed the operability determinations for the below listed safety related

structures degraded by ASR. The inspectors verified the basis for why the Radiological

Control Area tunnel was confirmed to not be affected by ASR. The inspectors reviewed

operability determinations for the following buildings:

.1

a.

Enclosure
5

. Control Building - "8" Electrical Tunnel,

o Containment Enclosure Building,

. Diesel Generator Fuel OilTank Rooms,

. Residual Heat Removal Equipment Vaults, and

r Emergency Feedwater Pump House.

The inspectors utilized site records and interviews to determine the design basis for the

safety related structures in addition to those summarized in Sections 3.7 and 3.8 of the

Updated Final Safety Analysis Report (UFSAR).

b. Observations

For the open aspects of numbers 1 and 2 above, a finding was identified and addressed

in Section 4OA5.1.c. This section also noted a new issue identified by NRC staff related

to shear reinforcement for the walls of the CB and the diesel generator building.

The open aspects of numbers 3 and 4 were updated but not completely resolved due to

the need to obtain additional information. At the beginning of the inspection, the NRC

staff review determined that the initial evaluation for the CEB did not address the open

aspects of numbers 1 and 2 above; and, in particular, the response of the entire

structure (whole building) to seismic loading comparable to the methods described in

UFSAR 3.8. This included how the induced seismic stresses would shift between the

concrete and the steel in adjoining sections of the structure. In response, NextEra noted

that these issues would be factored into the analytical model (finite element analysis) to

reanalyze the CEB using the as-measured worst case elastic modulus applied to

ASR-|m pacted sections.

Revision 1 of the applicable operability determination for the CEB provided additional

quantitative and qualitative analysis, for the available information, which addressed

groundwater intrusion limited to less than 25 percent of the perimeter of the below grade

portion of the building; the effect of the reduced modulus on the natural frequency; and

the effect on shear capacity that indicated that the dynamic and flexural response had a

minimaleffect.

In conclusion, this area remained open pending further developments and completion of

licensee actions as noted above and further NRC staff review. While this unresolved

item remains open, the NRC staff determined that the affected safety-related structures

can currently perform their safety functions. This conclusion was based on the following:

. Conservative safety load factors in controlling load conditions and engineering

conservatisms in design provide reasonable expectation that affected structures

can perform their safety function, despite the current licensing basis design

margin being reduced by the change of mechanical properties;

. Field walk-downs confirm no visible indication of significant deformation,

distortion, or displacement of structures, or rebar corrosion;

. Evidence of ASR limited to localized areas in the concrete walls; and

Enclosure
6

. Progression of ASR degradation occurs slowly based on existing operating

experience and published literature, and the licensee continues to monitor.

This unresolved item related to operability of ASR affected safety related buildings

remained open for NextEra to evaluate ASR effect on cable and pipe loadings

(number 3) and evaluate ASR effect on the CEB whole building response (number 4).

c. Findinq Related to Operabilitv Determinations and Functionalitv Assessments -

Inadequate Operabilitv Determinations

lntroduction. The inspector identified a finding in that NextEra failed to fully evaluate

potential structural and seismic response impacts in accordance with the requirements in

NextEra Procedure EN-AA-1001 after identifying degraded and nonconforming condition

related to reduced concrete modulus of elasticity due to ASR degradation for safety

related structures. The evaluation did not consider the following effects due to changed

properties of concrete as measured directly from building concrete core samples:

building natural frequency in the dynamic response; performance of anchorages and

embedment of systems and components attached to the structures; and shear strength

or capacity of affected structures and the dynamic/flexural response especially for those

building walls without corresponding shear reinforcement.

Description. NextEra analysis of concrete cores samples taken following the April 2011

determination that certain below grade concrete walls in safety related structures were

affected by ASR, indicated a reduced modulus of elasticity and compressive strength.

Although the compressive strength reduction was viewed by NextEra as slight and

acceptable, the lowest measured modulus was about 40 percent less than the design

value of 3,620 kpsi.

NextEra completed operability determinations for certain affected safety-related concrete

structures as required by NextEra Procedure EN-AA-203-1001 , 

"Operability

DeterminationsiFunctional Assessments." ln accordance with the Procedure

EN-AA-203-1001, an operability determination must include: identification of current

licensing basis functions and performance requirements as listed in the UFSAR;

identification of the minimum design basis values necessary to satisfy the structure,

system, or component (SSC) design basis safety functions; and evaluation of the effects

of the degraded condition on the ability of the SSCs to meet its specified function and

performance requirements.

During the week of September 26,2011, NRC staff determined that the completed

operability determinations were not sufficient in that they did not address the impact of

the degraded condition on key aspects of the structure design as described in UFSAR.

Specifically, NextEra failed to address the ASR induced effects of the reduced modulus

of elasticity on seismic dynamic and flexural response in the following areas:

r Building naturalfrequency in the dynamic response;

. Performance of anchorages and embedment of systems and components

attached to the structures affected by ASR; and

Enclosure
7

. Shear capacity of affected walls especially for those buildings without

corresponding shear reinforcement such as for the CB and the emergency diesel

generator building.

NextEra performed additional reviews and updated the operability determinations for the

affected areas in response to these concerns, on October 14,2011. The licensee

determined that the structures and other affected systems and components remained

functional for design basis conditions but were degraded.

The NRC reviewed the updated operability determinations and associated calculations

determining that the additional areas needing evaluation were addressed and that the

structures remained "operable but degraded." The previous determination indicated that

the evaluated structures were "operable." Specifically, NextEra used quantitative and

qualitative information with respect to the degraded concrete conditions as noted below.

With respect to dynamic response and the change in the natural frequency of the

structures, licensee's additional evaluation determined that the shift in naturalfrequency

was minimal and remained well above the ground response peak frequency range such

that the response of the structures remained rigid. With respect to the ability of the

equipment anchors and embedment to perform their function, the licensee's additional

evaluation noted that there was no appreciable impact. The licensee also determined

that the impact on the flexural capacity of seismic buildings with respect to shear stress

was minimal, and the resultant stresses on the steel and concrete remained below the

design stress limits with margin.

Following review, the inspector determined there was a reasonable expectation that the

structural integrity remained intact under design loads, and the buildings remained

operable but degraded. NextEra continued to review the degraded concrete issue within

the corrective action program, including the effects on the long term reliability of the

structures.

Analvsis. The inspectors determined that NextEra's failure to conduct adequate prompt

operability determinations per Procedure EN-M-203-1001 for degraded and

nonconforming conditions associated with ASR was a performance deficiency relative to

a self imposed standard. Specifically, the operability determinations conducted following

identification of ASR in safety-related structures did not completely analyze the effects of

the reduced modulus on the dynamic and flexural response of safety related structures

to seismic events along with the effect on attached systems and components. This

performance deficiency was associated with the design control aspect of the Mitigating

Systems cornerstone; and, based on a comparison to Example 3.i of Appendix E of

IMC 0612, it was determined to be more than minor. The issue was evaluated using

IMC 0609, "significance Determination Process," and was determined to be of very low

safety significance (Green). The finding had a cross cutting aspect in the area of

problem identification and resolution, P.1(c), related to ensuring that issues potentially

impacting nuclear safety are thoroughly evaluated. NextEra did not thoroughly evaluate

conditions adverse to quality, including evaluating the effects of the reduced concrete

modulus for impact on operability of the affected structures.

Enclosure
.2

8

Enforcement. Because this finding does not involve a violation and has very low safety

significance, it is identified as FIN 05000443/2011-10-02, lncomplete Operability

Determ i nation for Degraded Concrete Stru ctu res Hous i n g Safety-Re lated

Equipment.

(Closed) Unresolved ltem 05000443/201 1003-02. 50.59 Evaluation for Acceptinq

Reduced Modulus of Elasticitv in Certain Safetv-Related Structures Affected bv

Alkali-Silica Reaction

Inspection Scope

As part of the review of this unresolved item, the inspectors continued to review

EC272057, dated April 25, 2011, for adequacy in which the engineering change (EC)

was a design change to address reduced concrete modulus of elasticity in the CB

electric tunnel and the containment enclosure building. The review was to determine if

only a 10 CFR 50.59 screening was adequate to accept "as-is" conditions for this

concrete material property. The inspector reviewed NextEra's revocation of this EC,

Observations

This issue was closed based on the revocation of the EC, and on the Severity Level lV

NCV, as noted below.

Findino Related to Evaluations of Chanqes. Tests, or Exoeriments and Permanent Plant

Modifications - Inadequate 50.59 Screen Evaluation for EC272057

Introduction. The inspectors identified a Severity Level lV NCV of 10 CFR 50,59(dX1),

"Changes, Tests, and Experiments," because NextEra did not adequately evaluate a

"uSe-aS-iS" determination for the ASR impacted Category | concrete structures.

Specifically, NextEra did not complete a 10 CFR 50.59 evaluation, to ensure that the

identified reduction in concrete modulus of elasticity did not present a more than minimal

increase in the likelihood of the occurrence of a malfunction of a SSC important to safety

previously evaluated in the updated safety analysis report (USAR) prior to implementing

changes to the facility as described in the engineering change EC272057 issued on

April25, 2011.

Description. On April 25,2011, NextEra issued EC272057, "Concrete Modulus of

Elasticity Evaluation," to address the reduced concrete modulus in the CB, the "B"

electric tunnel, the containment enclosure building, the diesel generator fuel oil tank

rooms, the residual heat removal equipment vaults, and emergency feedwater pump

house. EC272057 dispositioned the degraded condition as "use-as-is" and incorporated

the degraded condition into the design basis. In a safety evaluation screen for

EC272Q57, NextEra concluded the change did not require a complete evaluation per

10 CFR 50.59(c)(2) because adequate design margin existed and there was no adverse

affect on an UFSAR described design function.

a.

b.

c.

Enclosure
9

10 CFR 50.59 requires licensees to evaluate whether NRC approval is required for

proposed changes to the facility. The Seabrook 5059 Resource Manual defines the

process for completing 10 CFR 50.59 evaluations for changes, tests, and experiments

completed at Seabrook. lt includes a screening process that defines criteria used to

determine whether a full 10 CFR 50.59 evaluation must be performed for each

applicable change, test, or experiment. NextEra screened EC272057 in accordance

with the guidance in the 5059 Resource Manual and concluded that the change did not

require a full evaluation per 10 CFR 50.59(cX2) because adequate design margin

existed and there were no adverse affects on the UFSAR described design functions.

The inspectors reviewed EC272057 and determined that NextEra's 50.59 Screen for

EC272057 did not correctly address "adverse affects" as described in Section 5.2.2 ot

the 5059 Resource Manual. The concrete modulus of elasticity is a design value

specified in both the Seabrook UFSAR and the ACI 318 - 1971 Building Code for the

applicable plant structures. The inspectors determined that the reduced modulus of

elasticity caused by the ASR could have had an "adverse affect" on the flexural and

dynamic response of the impacted structures and, as such, required further evaluation

per 10 CFR 50.59(cX2 (ii) and (iv). The criterion c(2)(ii) and (iv) dealwith the change

resulting in more than minimal increase in the likelihood of occurrence or in the

consequences of a malfunction of an SSC important to safety previously evaluated in the

UFSAR. In response to the inspectors' concerns regarding the adequacy of the

10 CFR 50.59 evaluation, NextEra rescinded the design change EC272057 from the

design basis on September 22,2011, and initiated additional evaluations of the ASR

affected structures.

NextEra personnel did not complete the 10 CFR 50.59 screen properly because they

misunderstood the guidance in the 50.59 Resource Manual regarding the need to screen

in changes in design parameters which impact the design function acceptance criteria

(Resource Manual Section 5.2.2).

Analvsis. The inspectors determined that the failure to evaluate changes to the facility

as described in EC272057 was contrary to 10 CFR 50.59(dX1) and was a performance

deficiency warranting a significance evaluation in accordance with the NRC Enforcement

Manual for Traditional Enforcement and IMC 0612, "Power Reactor lnspection Reports,"

Appendix B, "lssue Disposition Screening." The violation was determined to be more

than minor in accordance with IMC 0612, "Power Reactor Inspection Reports,"

Appendix B, "lssue Screening," because the inspector could not reasonably determine

that the changes would not have ultimately required prior NRC approval.

Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process

instead of the SDP because they are considered to be violations that could potentially

impede or impact the regulatory process. However, if possible, the underlying technical

issue is evaluated under the SDP to determine the severity of the violation. In this case,

for Mitigating Systems, the inspector determined the finding could be evaluated using

the SDP in accordance with IMC 0609, "Significance Determination Process,"

Attachment 0609.04, "Phase 1 - Initial Screening and Characterization of Findings." The

issue was determined to be of very low safety significance (Green) because it was a

design or qualification deficiency confirmed not to result in an actual loss of safety

Enclosure
10

function, because further evaluation determined that the structures remained operable

despite the degraded modulus condition. In accordance with Section 6.1.d.2 of the NRC

Enforcement Policy, this violation is categorized as Severity Level lV because the

resulting changes were evaluated by the SDP as having very low safety significance

(Green). Upon removal of EC272057 from the design basis on September 22, 2011, the

issue no longer required an evaluation per 10 CFR 50.59(aX2).

The finding had a cross cutting aspect in the area of human performance - work

practices, H.4(b), because NextEra personnel did not follow procedures. Specifically,

NextEra personnel did not address "adverse effects" as required by Section 5.2.2 of the

50.59 Resource Manual when preparing the 10 CFR 50.59 screen for EC272057.

Enforcement. Title 10 CFR 50.59, "Changes, Tests, and Experiments," Section (dX1)

states, in part, that the licensee shall maintain records of changes in the facility or

procedures, and that the records must include a written evaluation that provides the

bases for the determination that the change does not require a license amendment

pursuant to paragraph 10 CFR 50.59(c)(2). Contrary to the above, from April 25 to

September 22, 2011, NextEra did not provide an evaluation that adequately documented

why the reduced concrete modulus of elasticity in Category | structures did not present a

more than minimal increase in the likelihood of occurrence of a malfunction of a SSC

important to safety previously evaluated in the USAR. Because this failure to properly

evaluate a proposed change is of very low safety significance and has been entered into

the licensee's Corrective Action Program (CR1647722), this violation is being treated as

an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy.

(NCV 050004/32011010-03, Failure to Properly Complete a 50.59 Screen).

4046 Meetinqs. Includins Exit

On September 30 and December 2, 2011, the inspectors presented the interim results of

this inspection to Mr. P. Freeman, Site Vice President, and Seabrook Station staff. The

inspectors also confirmed with NextEra that no proprietary information was retained by

inspectors during the course of the inspection.

On January 20,2012, a flnal exit meeting was conducted and led by

Mr. Richard J. Conte, Chief Engineering Branch No, 1. Others involved in this

conference are noted on the list of contacts. During the meeting, the NRC staff's final

disposition of the unresolved items and new findings were summarized. Other

comments and questions were communicated to NextEra management with respect to

the ASR problem in safety related structures.

ATTACHMENT: SUPPLEMENTARY I NFORMATION

Enclosure
A-1

SUPPLEMENTARY I N FORMATION

KEY POINTS OF CONTACT

Licensee Personnel

B. Brown, Supervisor, Civil Engineering

V. Brown, Senior Licensing Analyst

K. Browne, Plant General Manager

J. Esteves, Plant Engineering

P. Freeman, Site Vice President

P. Gurney, Reactor Engineering Supervisor

M. Collins, Manager, Design Engineering

M. O'Keefe, Licensing Manager

Kev Participants for Teleconference of Januarv 20. 2012

NextEra Attendees:

Paul Freeman, Site Vice President

Mike O'Keefe, Licensing Manager

Mike Collins, Design Engineering Manager

Rick Clich6, License Renewal Project Manager

Ted Vassallo, Design Engineering

Paul Willoughby, Licensing

Ken Chew, License Renewal

Al Griffith, Public Communications

NRC Staff:

Christopher Miller, Division of Reactor Safety, Region I

Richard Conte, Division of Reactor Safety, Region I

Suresh Chaudhary, Division of Reactor Safety, Region I

Art Burritt, Division of Reactor Projects, Region I

Bill Raymond, Division of Reactor Projects, Region I

John Lamb, Division of Operating Reactor Licensing, NRR

Abdul Sheikh, Division of License Renewal, NRR

George Thomas, Division of Engineering, NRR

Raj Auluck, Division of License Renewal, NRR

Attachment
A-2

LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED

Opened/Closed:

05000443/201 1-010-01

05000443/201 1-010-02

05000443/201 1 -01 0-03

Closed:

05000443/2011-003-02

Updated:

05000443/201 1 -003-03

URI Adequacy of Corrective Actions Associated with

Calculation Methods for Alkali-Silica Reaction lssue

FIN lnadequate Operability Determination for Degraded

Concrete Structures Housing Safety-Related Equipment

NCV Failure to Properly Complete a 50.59 Screen for

EC272057

URI Review of 50.59 screening to accept-as-is reduced values

for concrete properties in safety related structures.

URI Prompt Operability Determination for Safety Related

Structures affected by ASR.

Definitions

(American Concrete Institute (ACl) Terminoloqv)

Poisson's ratio, y: The ratio of transverse strain (perpendicular to the applied load) to the axial

strain (in the direction of the applied load).

Modulus of Elasticitv. E: The ratio of the normal stress to the corresponding strain for tensile or

compressive stress below the proportional limit of the material.

Shear Modulus or Modulus of Riqiditv. G: The ratio of unit shearing stress to the corresponding

unit shearing strain.

Bulk Modulus. K: The ratio of the change in average stress to the change in unit volume.

Note: The above parameters are four simplified elastic constants defining a material exhibiting

"elastic" behavior.

Other Definitions:

Stress: The force per unit area (compressiveltensile, transverse or shear).

Strain: In a given direction (transverse or axial) is the change in dimension under load to the

original dimension in the direction under consideration,

Compressive Strenoth: Capacity of a material or structure to withstand axial pushing forces.

When the limit of compressive strength is reached, materials fail.

Tensile Strenqth: Capacity of a material or structure to withstand axial pulling forces. When the

limit of tensile strength is reached, materials fail.

Attachment
A-3

Shear Strenqth: Capacity of a material or structure to withstand forces parallel to a surface area

that could cause sliding failure of the material. When the limit of shear strength is

reached, materials fail.

Bond Strenqth: The resistance to separation of mortar and concrete from reinforcing and other

materials with which it is in contact.

LIST OF DOCUMENTS REVIEWED

Prompt Operability Determination (POD) AR 581434, Reduced Concrete Modulus of Elasticity

below Grade in 'B' Electrical Tunnel Exterior Walls, Revision 0, June 27,2011, and

Revision 1, October 14,2011

POD AR 1664399, Reduced Concrete Modulus of Elasticity Below Grade in Containment

Enclosure Building, RHR Equipment Vaults, EFW Pump House, and Diesel Generator

Fuel OilTank Rooms, Revision 0, June 27,2011, and Revision 1, October 14,2011

Catculation C-S-1-10163, Rev. 0, Fundamental Frequency of ASR Effected Walls,

October 14,2011

Calculation C-S-1-10159, Rev. 0,'B'ElectricalTunnel Transverse Shear Evaluation Supplement

to Calculation CD-20

Calculation C-S-1-10150, Rev. 0, Effects of Reduced Modulus of Elasticity -'B' Electrical

Tunnel Exterior Walls

Calculation CD-20-CALC, UE Control and Diesel Generator Building Design of Material

and Walls below Grade for Electrical Tunnel and the Control Building (Original Design

Calculation)

Drawings for Control Building Concrete (ElectricalTunnel) 9763-F-111342,9763-F'111343

and 9763-F-111345

EC 145305, Condition Assessment of Control Building Concrete

AR1641413, Evaluation of Containment with Craze Cracking in Concrete, April 2Q,2011

AR1644074, Concrete Test Results for Containment Enclosure Building, April 21, 2011

AR 574120, Preliminary Test Results of Control Building Concrete

AR 581434 Test Results from Control Building Concrete Modulus Testing (Results of

petrographic analysis of four of the 12 CB cores identified the presence of moderate

to severe ASR in the concrete)

EC250348, Revision 002, Condition Assessment of Building Concrete

Attachment
A-4

AR 01625775, Revision 000, Petrographic Analysis of Concrete Cores from Seabrook Station

System Description No. SD-66, Revision 2, System Description for Structural Design Criteria

for Public Service Company of New Hampshire, Seabrook Station, Unit Nos. 1 and 2,

3102184.

Seabrook UFSAR, Revision 12, Section 3.8.4, Other Seismic Category 1 Structures

Letter dated 6-29-2011 from Richard Plasse, USNRC, to Mr. Paul Freeman, NextEra Energy

Seabrook, LLC - Request for Additional Information for the Review of Seabrook Station

License Renewal Application (Specifically Follow-up to RAl B.2.1.31-1 on pages 2-3)

(M11117843380)

NextEra Energy Letter SBK-L-1 1154 to USNRC dated 8-11-2011, Docket No. 50-443, Seabrook

Station Response to Request for Additional lnformation - NextEra Energy Seabrook

License

NextEra Energy Letter SBK-L-1 1063 to USNRC dated 4-14-2011, Docket No. 50-443, Seabrook

Station Response to Request for Additional Information - NextEra Energy Seabrook

License Renewal Application Request for Additional Information - Set 13 (Specifically

Responses to Follow-up to RAl 8.2.1 .31-1 and -2 on pages 4-7) (ML1 1 108A1310)

NextEra Energy Letter SBK-L-10204to USNRC dated 12-17-2010, Docket No. 50-443,

Seabrook Station Response to Request for Additional Information - NextEra Energy

Seabrook License RenewalApplication Aging Management Programs (Specifically

Responses to RAI 8.2.1 .31-1 , -2 and -3 on pages 36-39) (M11035405340)

AR

ACt

ASR

ASME

CB

CEB

CFR

CR

DRS

EC

EN

FIN

tMc

IP

KSI

LRA

NCV

NRC

LIST OF ACRONYMS

Action Request

American Concrete Institute

Alkali-Silica Reaction

American Society of Mechanical Engineers

Control Building

Containment Enclosure Building

Code of Federal Regulations

Corrective Action

Division of Reactor Safety

Engineering Change

Procedural Notice for Engineering Department

Finding

Inspection Manual Chapter

Inspection Procedure

Kilo-pounds per square inch

License Renewal Application

Non-Cited Violation

U.S. Nuclear Regulatory Commission

Attachment
NRR

OD

POD

psi

PSIG

RCA

SDP

SR

SSC

TIA

TS

UFSAR

URI

USAR

A-5

Office of Nuclear Reactor Regulation

Operabil ity Determ ination

Prompt Operability Determination

Pounds per square inch (absolute)

Pounds per square inch (gage)

Radiological Controlled Area

Significance Determination Process

Safety Related

Structure, System, or Component

Task lnterface Agreement

Technical Specification

Updated Final Safety Analysis Report

Unresolved ltem

Updated Safety Analysis Report

Attachment