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{{Adams
#REDIRECT [[IR 05000443/2011010]]
| number = ML120480066
| issue date = 03/26/2012
| title = IR 05000443-11-010; NextEra Energy Seabrook, LLC; 9/25/2011 - 12/2/2011 Seabrook Station (Problem Identification and Resolution; Follow-up to Operability and Plant Modifications)
| author name = Miller C G
| author affiliation = NRC/RGN-I/DRS
| addressee name = Freeman P
| addressee affiliation = NextEra Energy Seabrook, LLC
| docket = 05000443
| license number = NPF-086
| contact person =
| case reference number = G20120266, SECY-2012-0196
| document report number = IR-11-010
| document type = Inspection Report, Letter
| page count = 22
}}
See also: [[followed by::IR 05000443/2011010]]
 
=Text=
{{#Wiki_filter: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-lm 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 l 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 l 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
}}

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