ML120480066
| ML120480066 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 03/26/2012 |
| From: | Chris Miller Division of Reactor Safety I |
| To: | Freeman P NextEra Energy Seabrook |
| References | |
| G20120266, SECY-2012-0196 IR-11-010 | |
| Download: ML120480066 (22) | |
See also: IR 05000443/2011010
Text
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).
,aA
Sincerely,
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 "N"=No
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
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
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
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No.: 50-443
License No.: NPF-86
Report No.: 05000443/2011010
Licensee: NextEra Energy Seabrook, LLC
Facility: Seabrook Station
Location: Seabrook, NH 03874
Dates: September 26-September 30, 2011
November 15-17, 201 1 (Northbrook, lllinois)
November 28-December 1, 2011
January 20,2012 (Conference Call)
Inspectors: 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
Accompanied by: A. Sheikh, Senior Structural Engineer, Office of Nuclear
Reactor Regulation (NRR)
G. Thomas, Structural Engineer, NRR
Approved by: 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
a. 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.
b. 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:
Enclosure
3
(1) 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;
(2) Basis for the representativeness of concrete core sampling in the buildings for
those taken to date and those to be taken, should they occur;
(s) lmpact of core boring and re-grouting on the building structural integrity; and
(4) 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
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.
40A5 Other Activities
.1 (Open) Unresolved ltem 05000443/2011003-03. Open Operabilitv Determinations for
Safetv-Related Stru ctu res Affected bv Al kal i-Sil ica Reaction
a. 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:
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
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.
.2 (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
a. 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,
b. Observations
This issue was closed based on the revocation of the EC, and on the Severity Level lV
NCV, as noted below.
c. 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.
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 URI Adequacy of Corrective Actions Associated with
Calculation Methods for Alkali-Silica Reaction lssue
05000443/201 1-010-02 FIN lnadequate Operability Determination for Degraded
Concrete Structures Housing Safety-Related Equipment
05000443/201 1 -01 0-03 NCV Failure to Properly Complete a 50.59 Screen for
Closed:
05000443/2011-003-02 URI Review of 50.59 screening to accept-as-is reduced values
for concrete properties in safety related structures.
Updated:
05000443/201 1 -003-03 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)
LIST OF ACRONYMS
AR Action Request
ACt American Concrete Institute
ASR Alkali-Silica Reaction
ASME American Society of Mechanical Engineers
CB Control Building
CEB Containment Enclosure Building
CFR Code of Federal Regulations
CR Corrective Action
DRS Division of Reactor Safety
EC Engineering Change
EN Procedural Notice for Engineering Department
FIN Finding
tMc Inspection Manual Chapter
IP Inspection Procedure
KSI Kilo-pounds per square inch
LRA License Renewal Application
NCV Non-Cited Violation
NRC U.S. Nuclear Regulatory Commission
Attachment
A-5
NRR Office of Nuclear Reactor Regulation
OD Operabil ity Determ ination
POD Prompt Operability Determination
psi Pounds per square inch (absolute)
PSIG Pounds per square inch (gage)
RCA Radiological Controlled Area
SDP Significance Determination Process
SR Safety Related
SSC Structure, System, or Component
TIA Task lnterface Agreement
TS Technical Specification
UFSAR Updated Final Safety Analysis Report
URI Unresolved ltem
USAR Updated Safety Analysis Report
Attachment