ML12174A232
| ML12174A232 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 01/20/2012 |
| From: | Chris Miller Division of Reactor Safety I |
| To: | Freeman P NextEra Energy Seabrook |
| References | |
| FOIA/PA-2012-0119 IR-11-010 | |
| Download: ML12174A232 (23) | |
See also: IR 05000443/2011010
Text
Inspector and Tech Reviewer Areas to focus on:
1. Cover letter messages and request
2. Executive Summary (ES) for appropriateness and need
3. Length of Open URI section 40A5.2
4. Summary of POD assumptions issues from TIA in ES and 40A5
5. What to address in plans is only in cover letter (management integration of inspector observations
to TIA)
6. No immediate safety concerns in ES only.
7. List of ACRONYMS and Reference list ???????
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
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 on January 2 0 th 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. The focus of this inspection was activities surrounding your development actions
related to the Alkali-Silica Reaction (ASR) problem occurring in safety related structures and
other structures of regulatory importance (covered by the maintenance rule). In particular, we
reviewed your Prompt Operability Determinations for certain structures based on best available
information. At the beginning of the inspection report period, we noted some areas that still
needed to be addressed based on available information and NextEra satisfactorily addressed
them with revisions to the documents.
On January 20, 2011 a final exit meeting was conducted and lead by Mr. Richard J. Conte,
Chief Engineering Branch No. 1 of my staff. During the meeting, my staff summarized the
change in status of the new findings and our plans to issue a Task Interface Agreement
between Region I and the Office of Nuclear Reactor Regulation simultaneously with this report.
The TIA was placed in the public document room (ADAMS Accession No. MLXXXXXXX). The
purpose of the Task Interface Agreement was for the NRR staff to identify the review criteria in
evaluating the operability determination for the "B" Electrical Tunnel affected by ASR (part of the
Control Building) in assistance to the Region I staff by addressing questions we had on the
matter.
Also on January 2 0 1h, we focused on and summarized observations on your plans with respect
to the unwritten assumptions in your operability determinations. The NRC staff noted that these
I
P. Freeman 2
determinations listed no assumptions in the applicable sections and that the design basis code
ACI 318-1971 was based on empirical data for determining certain parameters that were a part
of the design bases. Also Poison's ratio on concrete cores were not being tested or evaluated
and this ratio was in the UFSAR. The assumption of this empirical data was that the
relationships were for ASR free concrete. Specific areas for which your plans do not address
unwritten assumptions being made in the prompt operability determinations were list in section
40A5.2
In consultation with our technical reviewers in headquarters and to address the current
shortcomings on unwritten assumptions for your operability determinations, we have determine
that your plans do not sufficiently provide information related to: 1) condition assessment
(extent and characterization); 2) cause of the ASR as it impacts current degradation and
operability; 3) estimate of expansion to date and current expansion rate; 4) interim structural
assessment as it impacts current operability vs. longer term structural assessment; and, longer
term monitoring ensure operability in the near future vs. longer term of the duration of the
license (1-2 years vs. longer); and, 5) short term mitigation or needed remedial actions. This is
in distinction to your overall comprehensive plan for the problem.
Accordingly, we request that you provide your plans to address the above issues within 30 days
of the date of this inspection report. We noted that, from the exit meeting of January 2 0 th you
have agreed to this request and to review the report in 15 days and let us know of your plans to
honor our request or identify the need for a management meeting. We further request that,
should a management meeting be needed on these issues, it should be conducted within 30
days of the date of this report and a final response time will be negotiated at the management
meeting. If your root cause evaluation scheduled for Feb. 2012 and the associate corrective
action plan for this significant condition adverse to quality addresses the above, please use
them to respond to our request.
Also, the report documents two NRC-identified findings of very low significance (Green) that
were determined to involve 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. If 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 Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington, DC 20555-
0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement,
United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC
Resident Inspector 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 I, and the NRC Resident Inspector at Seabrook.
P. Freeman 2
In accordance with 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 (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.qov/readinq-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
Christopher G. Miller, Director
Division of Reactor Safety
Docket No.: 50-443
License No.: NPF-86
Enclosure:
Inspection Report No. 05000443/201110
w/Attachment: Supplemental Information
cc w/encl: Distribution via ListServ
P. Freeman 2
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and 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 (PARS) component of
NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.qov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
Christopher Miller, Director
Division of Reactor Safety
Docket No.: 50-443
License No.: NPF-86
Enclosure:
Inspection Report No. 05000443/201110
w/Attachment: Supplemental Information
cc w/encl: Distribution via ListServ
Distribution w/encl: (via e-mail)
W. Dean, RA W. Raymond, DRP, SRI
D. Lew, DRA J. Johnson, DRP, RI
D. Roberts, DRP A. Cass, DRP, Resident OA
D. Ayres, DRP M. Franke, RI, OEDO
C. Miller, DRS RidsNrrPMSeabrook Resource
P. Wilson, DRS RidsNrrDorlLpll-2 Resource
A. Burritt, DRP ROPreports Resource
R. Montgomery, DRP M. Modes, DRS
SUNSI Review Complete: (Reviewer's Initials)
DOCUMENT NAME: G:\DRS\Engineering Branch 1\-- MModes\20111213 050442_201110
Seabrook IP711158A ASR Follow UP.docx
After declaring this document "An Official Agency Record" it will be released to the Public.
To receive a copy of this document, indicate in the box: "C" = Copy without
attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy
OFFICE RI/DRSI RI/DRS RI/DRP RI/DRS RI/DRS
NAME SChaudhary MModes/ ABurritt/ RConte/ CMiller/
DATE 1/ /12 1/ /12 12/ /12 12/ /12 12/ /12
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 25 - September 30,
November 15-17, 2011 (Illinois)
November 28-29, 2011
January 20, 2012 (Conference Call)
Inspectors: M. Modes, Senior Reactor Inspector, Region I
S. Chaudhary, Reactor Inspector, Region I
W. Raymond, Senior Resident Inspector, Seabrook
Atif Shaikh, Reactor Inspector, Region III
Accompanied by: A. Sheikh, Senior Structural Engineer, NRR
G. Thomas, Structural Engineer, NRR
Approved by: Richard J. Conte, Chief
Engineering Branch 1
Division of Reactor Safety
Enclosure
SUMMARY OF FINDINGS
IR 05000443/2011010; 9/25/2011 - 12/2/2011; Seabrook Station (IP 7111115 and IP7111117).
This report covers an inspection by regional inspectors, and resident staff; with assistance from
NRR structural specialists. Two Green findings were identified. The significance of most
findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process". The cross-cutting aspects for the findings
were determined using IMC 0310, "Components Within Cross-Cutting Areas." Findings for
which the Significance Determination Process does not apply may be Green, or be assigned a
severity level after 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 regarding NextEra's operability determinations for
Category I structures with a reduced concrete modulus of elasticity caused by alkali-silica
reaction in the concrete. NextEra did not adequately evaluate with available information the
effects of the reduced concrete modulus with respect to key aspects of structural design as
described in the Updated Final Safety Analysis Report (UFSAR). Specifically, NextEra did not
initially fully evaluate the effects of the reduced modulus on the dynamic response of Category I
structures to seismic events relative to global response; the changes in the structural natural
frequency; and, the effects on attached systems, components and anchors. Further, NextEra
did not adequately evaluate the adequacy of shear capacity of electric tunnel concrete
walls without shear reinforcement to resist lateral forces during seismic events.
The failure to fully evaluate the degraded and nonconforming concrete modulus condition as
required by procedure EN-AA-203-1 001 was a performance deficiency. The performance
deficiency was associated with the Mitigating Systems cornerstone and was determined to be
more than minor based on a comparison with Appendix E.3.i of IMC 0612 because it adversely
affected the cornerstone objective to ensure the availability, reliability and capability of systems
that respond to initiating events in order to prevent core damage. The issue was evaluated
using IMC 0609, "Significance Determination Process" (SDP), 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. Specifically, 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. (Section 1 R1 5)
Severity Level IV. The inspectors identified a non-cited violation of 10 CFR 50.59(d)(1) because
NextEra did not provide an evaluation that adequately documented why implementing a design
change to address an identified reduction in the concrete modulus of elasticity for several
Category I concrete structures, 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 UFSAR. Specifically, NextEra issued EC272057 on April 25, 2011,
to address reduced concrete modulus of elasticity in the control building electric tunnel and the
Enclosure
t
containment enclosure building, but did not complete a 10 CFR 50.59 evaluation prior to
implementing changes to the facility as described in the modification.
The failure to evaluate changes to the facility as described in EC272057 was contrary to
10 CPR 50.59(d)(1) and was a performance deficiency warranting a significance evaluation in
accordance with the NRC Enforcement Manual for Traditional Enforcement and Inspection
Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix B, "Issue
Disposition Screening." The violation was determined to be more than minor in accordance with
IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," because the
inspector could not reasonably determine that the changes would not have ultimately required
prior NRC approval. The finding was evaluated using the SDP in accordance with IMC 0609,
"Significance Determination Process," and determined to be potentially risk significant due to a
design deficiency confirmed not to result in a loss of operability. In accordance with Section
6.1.d.2 of the NRC Enforcement Policy, this violation is categorized as Severity Level IV
because the resulting changes were evaluated by the SDP as having very low safety
significance (Green). 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 5059 screen for EC272057. (Section 1R17)
Executive Summary
The focus of this inspection was activities surrounding development actions related to the Alkali-
Silica Reaction (ASR) problem occurring in safety related structures and other structures of
regulatory importance (covered by the maintenance rule). In particular, NRC staff reviewed the
Prompt Operability Determinations for certain structures based on best available information. At
the beginning of the inspection report period, we noted some areas that still needed to be
addressed based on available information and these issues were satisfactorily addressed with
revisions to the documents.
Prior NRC review of this area was documented in the following Inspection Reports
05000443/2010004, 2010005, 2011002, 2011003, and 2011007 and the results were
summarized in the Background section of this report.
One unresolved items was closed and sufficient information was obtained in order to determine
the performance deficiency - 50.59 screening to accept as-is conditions for reduced modulus of
elasticity for certain safety related structures. Another unresolved item was left open - Prompt
Operability Determinations for certain safety related structure with the ASR problem. Another
finding of very low safety significance was determined in the course of the operability
determination review. Both findings were summarized above.
The inspectors observed that NextEra plans for the ASR problem were not addressing unwritten
assumptions in the operability determinations. These operability determinations listed no
assumptions in the applicable sections. The design basis code ACI 318-1971 was based on
empirical data for determining certain parameters that were a part of the design bases.
Poisson's ratio on concrete cores were not being tested or evaluated and this ratio was in the
UFSAR. The assumption of this empirical data was that the relationships were for ASR free
concrete. Specific areas for which the plans do not address unwritten assumptions being made
in the prompt operability determinations were list in section 40A5.2
Enclosure
To address the above and in conjunction with the NRR technical reviewers, the inspectors noted
that the current plans were not finalized and what was existing did not specifically address the
unwritten assumptions in the following areas: 1) condition assessment (extent and
characterization); 2) cause of the ASR as it impacts current degradation and operability; 3)
estimate of expansion to date and current expansion rate; 4) interim structural assessment as it
impacts current operability vs. longer term structural assessment; and, longer term monitoring
ensure operability in the near future vs. longer term of the duration of the license (1-2 years vs.
longer); and, 5) short term mitigation or needed remedial actions.
The NRC staff summarized why there is no immediate safety concerns for the existing
conditions: walkdowns confirm no significant degradation, no visual evidence of distortion nor
visible evidence of rebar corrosion; overall evidence of sufficient stiffness remaining; No
appreciable evidence of cracking where found, in isolated sections of the wall; degradation
appears to be localized; We know from best available research that the ASR rate slowly
progresses and there is some evidence that it has progressed to a plateau but it needs to be
confirmed by testing; parameters obtained for compressive strength and modulus of elasticity
indicate robust design to strength of the concrete poured (4K psi concrete used in buildings only
needing 3K psi concrete and no significant negative shift on seismic analysis.
On January 20, 2011 a final exit meeting was conducted. During the meeting, NRC staff
summarized the change in status of the new findings and plans to issue a Task Interface
Agreement between Region I and the Office of Nuclear Reactor Regulation simultaneously with
this report. The TIA was placed in the public document room (ADAMS Accession No.
MLXXXXXXX). The purpose of the Task Interface Agreement was for the NRR staff to identify
the review criteria in evaluating the operability determination for the "B" Electrical Tunnel
affected by ASTR (part of the Control Building) in assistance to the Region I staff by addressing
questions related to the problem.
Enclosure
2
REPORT DETAILS
Background
In June 2009, NextEra conducted walkdowns of structures within the scope of license renewal
as part of license renewal application preparations. In June, 2010 the License Renewal
Application (LRA) was received by the agency. In October 2010, the License Renewal Audit
results noted the alkali-silica reaction (ASR) problem and pointed to need for a good number of
requests for additional information in this area since the issue was newly discovered for the site
(noted as a area to address in the GALL, Generic Aging Lessons Learned, Revision 1). In the
Fall of 2010, NextEra performed an Immediate and Prompt Operability Review (POD) based on
core samples taken in Control Building in August 2010. In November 2010, Inspection Report 05000443/2010004 and, in February 2011, Inspection Report 2010005, followed developments
in this area from an operability viewpoint. In these reports, no findings were noted since
laboratory results determined compressive strength and modulus of elasticity met UFSAR
values (degradation into reserve strength not design margin). In May 2011, Inspection Report 05000443/2011002 identified two noncited violations of very low safety significance in the
structures monitoring area with respect to the maintenance rule (10 CFR 50.65 a(1 ) and b(2)).
Also in May 2011, License Renewal Inspection (IP71002) Report 05000443/2011007 had an
overall result: "Except for Structures Monitoring Program, results support a reasonable
assurance determination for license renewal."
As the year progressed, NextEra continued to identify and characterize the below-grade
structures at Seabrook having experienced groundwater infiltration and a resultant reduction in
concrete material properties. NextEra determined the degraded concrete condition was most
likely due to distress from ASR in the concrete. ASR is a chemical reaction in concrete over
time between the alkaline cement paste and reactive non-crystalline silica which is found in
common coarse aggregates. The reaction only occurs in the presence of water and forms a gel
that expands, forming micro-cracks that change the strength of concrete system. NextEra is in
the discovery phase of condition assessment along with extent of condition reviews in order to
support other building operability determinations and a plan to conduct an engineering
evaluation which would appear to constitute a long term operability review (40 year license
term) along with mitigation and monitoring measures.
The appearance of ASR degradation in safety related concrete structures was the first noted in
the nuclear industry in the United States and could be significant related to preservation of
reserve capacity with building design loads as reflected in the current licensing basis during
normal operations and aging management over the period of extended plant operation. While
the problem was noted as an aging effect in a license renewal topical report, the appearance of
ASR reflected a newly discovered aging effect at Seabrook that needs to be managed for
In August 2011, Inspection Report 05000443/2011003 addressed a noncited violation of very
low safety significance related to the untimely Initial and Prompt ODs for results on extent of
condition review for other buildings affected by ASR. Two unresolved items was also opened,
one dealing with a potential inadequate screen in accordance with 10 CFR 50.59 for accepting
the reduce parameter found on compressive strength and modulus of elasticity for the "B"
Electrical Tunnel and the Containment Enclosure Building. The other unresolved dealt with the
open prompt operability determinations associated with the "B" Electrical Tunnel and the
Enclosure
3
building subjected to an extent of conditions review. In August of 2011 a Task Interface
Agreement (TIA) was issued between Region I and the Office of Nuclear Reactor Regulation on
the ASR issue. The response to the TIA is publicly available (ADAMS Accession No.
MLxxxxxxxxx). The purpose of this inspection was to followup on the unresolved items and to
work with the NRR technical reviewers in developing the answers to the questions posed in the
TIA.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1 R15 Operability Determinations and Functionality Assessments (71111.15 - 0 samples)
a. Inspection Scope
As a part of the review of an unresolved item (see section 4OA5.2), the inspectors, in
conjunction with technical reviewers from the Office of Nuclear Reactor Regulation,
reviewed the adequacy of operability determinations for the below grade concrete walls
of seismic Category I buildings affected by alkali-silica reaction. The review focused on
the specific design for the buildings affected by an alkali silica reaction and operability
was addressed by two prompt operability determinations, one, to address the control
building and the other to address the following extent of conditions with five other
building/structures: Containment Enclosure Building, Emergency Diesel Generator
Buildings (fuel oil rooms), Residual Heat Removal Equipment Vaults, Emergency
Feedwater Pump House, and the Radiological Control Access (RCA) tunnel.
b. Findings
Inadequate Operability Determinations
Introduction. In April 2011, NextEra identified a degraded and nonconforming condition
related to reduced modulus of elasticity for buildings housing safety related equipment,
but did not thoroughly evaluate potential impacts in accordance with the requirements in
NextEra procedure EN-AA-203. Specifically, the evaluation did not consider the affect of
the reduced modulus on: overall stiffness and resulting seismic response, concrete
material property changes affecting natural frequency and therefore the amplitude of the
seismic response, and the performance of systems and components attached to the
affected structures during seismic events.
Description. In April 2011 NextEra determined that certain below grade concrete walls
were affected by alkali-silica reaction (ASR). The analysis of concrete cores showed a
reduced concrete modulus of elasticity in the control building / electric tunnel
(AR581434), containment enclosure building (AR1644074) and three other seismic
Category I buildings (AR1664399). The lowest measured modulus was about 40% less
than the design value of 3.62E+03 ksi.
NextEra completed prompt operability determinations for the affected Category I
concrete structures (reference ARs 581434, 1644074 and 1664399) as required by
NextEra Procedure EN-AA-203-1001, "Operability Determinations/Functional
Enclosure
4
Assessments." In accordance with the procedure EN-AA-203-1001, a POD 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 SSC design basis safety functions; and evaluation of the effects of the
degraded condition on the ability of the SSC to meet its specified function and
performance requirements.
During the week of September 28, 2011, the inspectors in conjunction with technical
reviewers from the Office of Nuclear Reactor Regulation reviewed NextEra's competed
PODs for the ASR-affected Category I concrete structures. It was determined that the
evaluations were not complete with respect to available information since NextEra did
not evaluate the degraded condition with respect to key aspects of the structure design
as described in UFSAR. Specifically, the initial PODs did not adequately address the
effects of the reduced modulus of elasticity in the following areas:
1. As a result of further review of the design bases, the NRC staff determined that the
walls below grade in the Control Building 'B' Electrical Tunnel do not contain shear
reinforcement to resist dynamic lateral forces acting on the wall during a design basis
earthquake. The design of the wall intentionally depends on the strength of the
concrete alone to resist these dynamic forces and this information was not evaluated
for the degraded conditions. The same was true for the Diesel Generator Building.
The modulus of elasticity of concrete was a function of concrete compressive
strength which is generally higher in the as-cast condition than assumed in the
design. The concrete used in construction of Seabrook structures was formulated to
have a design strength greater than 3000 psi. However, as stated in the Seabrook
Updated Final Safety Analysis Report, Revision 12, Section 3.8, "While variability in
concrete modulus has no significant effect on structural design, it influences
structural stiffness and natural frequency, and, subsequently, the amplified response
spectra of the seismic analysis."
2. Changes in the modulus of elasticity affect the material concrete properties and,
therefore, the natural frequency of the structure, which affects how the buildings are
analyzed in the seismic analysis. NRC reviews determined that the prompt
operability determinations addressed the dynamic response of the structures in a
qualitative manner noting that the ASR impacted walls are below grade and the
structural loadings would be governed by the ground response spectra assumed in
the original design. The initial evaluations did not sufficiently address the impact of
the reduced modulus on the structure natural frequencies in a quantitative manner to
validate that the structure response would remain rigid or that there would be a
amplification of the response. Specifically, the initial evaluation did not verify there
.would be no amplification of the motions beyond those in a ground response spectra
as assumed in the seismic analysis per UFSAR 3.7(B).2.
3. The initial POD addressed the effects of the reduced modulus on components
housed within the structures, such as pipe supports, cable trays and component
support anchors. NRC review determined that the initial evaluations addressed the
impacts on the internal components in a qualitative manner, but did not verify the
equipment performance would remain bounded by the analysis in the original design
as described in UFSAR 3.7 and 3.8. Specifically, the initial evaluation did not verify
Enclosure
IL
5
there would be no amplification of the motions beyond those in a ground response
spectra as assumed in the seismic analysis per UFSAR 3.7.(B).2. Further, the initial
evaluation did not evaluate the potential impacts on anchor or wall shear capacities
caused by ASR induced changes in material properties beyond that allowed for as
described in UFSAR 3.8.
In response to the NRC-identified issues, NextEra completed' additional evaluations that
determined the structures and other affected systems and components remained
functional for design basis conditions. On October 14, 2011, NextEra completed
Calculation C-S-1-10163, and revisions to the PODs for AR581434 (CB/ ET) and
AR1664399 (CEB and other Category I Structures). The NRC determined that
NextEra's additional analysis and revisions to the PODs adequately addressed the
concerns discussed above. Specifically, the analysis confirmed a minor impact on the
overall response of the structure during a seismic event, a small effect on the structure's
natural frequencies that results in no appreciable amplification of the ground response
during a seismic event, and no impact on the ability of the equipment anchors to perform
their function due to the quality of the concrete and construction methods used.
Analysis. The inspectors determined that not following a self imposed standard, not
completely analyzing the effects of the reduced modulus of elasticity on Category I
structures based on available information, per procedure EN-AA-203-1001 was a
performance deficiency. Specifically, because the reduced modulus affected the
dynamic response of Category I structures to seismic events relative to global response,
changes in natural frequency and the effects on attached systems and components,
procedure EN-AA-203-1 001 required that these impacts be evaluated as part of the
prompt operability determination. This performance deficiency was associated with the
Mitigating Systems cornerstone and was determined to be more than minor because,
based on a comparison with Examples 3.i of Appendix E of IMC 0612, it adversely
affected the cornerstone objective to ensure the availability, reliability and capability of
systems that respond to initiating events in order to prevent core damage. Specifically,
the effects of the reduced modulus of elasticity on the dynamic response of Category I
structures to seismic loading required further evaluation to demonstrate the structures
and enclosed systems remained functional as described in the licensing and design
bases. The issue was evaluated using IMC 0609, "Significance Determination Process"
(SDP), and was determined to be of very low safety significance (Green). Specifically,
when evaluated under IMC 0609, Attachment 4, the performance deficiency was not a
design or qualification deficiency resulting in an actual loss of safety function, was not a
loss of a barrier function, and was not potentially risk significant for external events. 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 thoroughly evaluate conditions adverse to
quality, including evaluating the effects of the reduced concrete modulus for impact on
operability of the affected structures.
Enforcement. Because this finding does not involve a violation and has very low safety
significance, it is identified as FIN 05000443/2011-10-01, Incomplete Operability
Determination for Degraded Concrete Structures Housing Safety-Related Equipment
based on available information.
Enclosure
6
1R17 Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications
(71111.17- 0 sample)
a. Inspection Scope
As a part of the review of an unresolved item (see section 40A5.1), the inspector
reviewed EC272057, dated April 25, 2011, for adequacy in which the EC was a design
change to address reduced concrete modulus of elasticity in the control building electric
tunnel and the containment enclosure building. The review was to determine if only a
50.59 screening was acceptable to accept "as-is" conditions for this concrete material
property which was degraded from the design bases as reflected in the UFSAR
apparently due to the ASR problem.
b. Finding
Inadequate 50.59 Screen Evaluation for EC272057
Introduction: A Severity Level IV NCV of 10 CFR 50.59(d)(1), "Changes, Tests, and
Experiments," was identified because NextEra did not provide an evaluation that
adequately documented why implementing a design change to address an identified
reduction in the concrete modulus of elasticity for several Category I concrete structures,
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). Specifically, NextEra issued
EC272057 on April 25, 2011, to address reduced concrete modulus of elasticity in the
control building electric tunnel and the containment enclosure building, but did not
complete a 10 CFR 50.59 evaluation prior to implementing changes to the facility as
described in the modification.
Description: NextEra determined that certain below grade concrete walls were affected
by alkali-silica reaction (ASR). The analysis of concrete cores taken from ASR affected
areas, showed a reduced concrete modulus of elasticity in the control building / electric
tunnel (AR581434), containment enclosure building (AR1644074) and four other seismic
Category I buildings (AR1664399). The lowest measured modulus was about 40% less
than the design value of 3.62E+03 ksi.
Enclosure
7
On April 25, 2011, NextEra issued EC272057, "Concrete Modulus of Elasticity
Evaluation," to address the reduced modulus. EC272057 dispositioned the degraded
condition as "use-as-is," and incorporated the degraded condition into the design basis.
In a safety evaluation screen for EC272057, NextEra concluded the change did not
require a complete evaluation per 50.59(c)(2) because adequate design margin existed
and there was no adverse affect on an UFSAR described design function.
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. It 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 50.59(c)(2) 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 of
the 5059 Resource Manual. The concrete modulus of elasticity is a design value
specified in both the Seabrook UFSAR and the ACI 318 Building Code for the applicable
plant structures. The reduced modulus of elasticity caused by the ASR occurring in
impacted concrete walls has the potential to affect the flexural capacity and dynamic
response of the impacted structures. Therefore, the inspectors determined that there
was sufficient evidence that the reduction in the modulus of elasticity was caused by the
ASR and it was an "adverse affect" as described in Section 5.2.2 of the 5059 Resource
Manual and thus required further evaluation per 50.59(c)(2). The additional evaluation
required by 10 CFR 50.59 d(1) was needed to at least ensure that 10 CFR 50.59 c(2) (ii)
and (iv) criteria were not met and, therefore, there would be a need for a license
amendment per 10 CFR 50.90. The criterion c(2)(ii) and iv) deal with 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 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.
On October 14, 2011, NextEra issued additional information to support of its engineering
evaluation of the ASR impacted structures, including Calculation C-S-1 -10163, the
Prompt Operability Determination for AR581434 (CB/ ET) Revision 1, and the Prompt
Operability Determination for AR1664399 (CEB and other Category I Structures)
Revision 1. The reduced modulus caused the concrete to have increased flexure, but
the results of NextEra's additional evaluations confirmed that the reduction in capacity
was minimal and the resultant stresses on the steel and concrete caused by the ASR
degradation remained below the design stress limits with margin. Similarly, the affect of
the reduced modulus also reduced the natural frequency of the structures, but the
additional evaluation again determined that the shift in natural frequency was minimal
and remained well above the ground response peak frequency range such that the
response of the structures remained rigid. Therefore, although the effect of the ASR on
the impacted walls was to reduce the design modulus parameter, the structural integrity
Enclosure
remained fully intact under all design loads, and the buildings remained operable.
NextEra actions continue to review the degraded concrete issue within the corrective
action program, including the effects on the long term reliability of the structures. See
Section 40A5 of this report for further NRC reviews of the revised operability
determinations for ASR impacted structures.
Analysis The inspectors determined that the failure to evaluate changes to the facility
as described in EC272057 was contrary to 10 CFR 50.59(d)(1) and was a performance
deficiency warranting a significance evaluation in accordance with the NRC Enforcement
Manual for Traditional Enforcement and Inspection Manual Chapter (IMC) 0612, "Power
Reactor Inspection Reports," Appendix B, "Issue Disposition Screening." The violation
was determined to be more than minor in accordance with IMC 0612, "Power Reactor
Inspection Reports," Appendix B, "Issue 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,
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," Tables 3b and 4a, for the Mitigating
Systems Cornerstone. The inspector answered "Yes" to Question 5 under the Mitigating
Systems Cornerstone column of the Phase 1 worksheet because the inspector
concluded that the finding screened as potentially risk significant due to a design or
qualification deficiency confirmed not to result in a loss of operability or functionality. In
accordance with Section 6.1 .d.2 of the NRC Enforcement Policy, this violation is
categorized as Severity Level IV because the resulting changes were evaluated by the
SDP as having very low safety significance (Green). Further evaluation determined that
the structures remained operable despite the degraded modulus condition. Upon
removal of EC272057 from the design basis on September 22, 2011, the issue no longer
required an evaluation per 10 CFR 50.59(a)(2).
NextEra personnel did not complete the 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). 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 50.59 Resource Manual when preparing the 50.59 screen for EC272057.
Enforcement Title 10 CFR 50.59, "Changes, Tests, and Experiments," Section (d)(1)
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 I structures did not present a
Enclosure
I
9
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 05000443/2011010-02, Failure to Properly Complete a 5059 Screen for EC272057).
4. OTHER ACTIVITIES
40A5 Other Activities
.1 (Closed) Unresolved Item 05000443/2011003-02, 50.59 Evaluation for Accepting
Reduced Modulus of Elasticity in Certain Safety-Related Structures Affected by ASR
a. Inspection Scope
The inspector reviewed EC272057, dated April 25, 2011, for adequacy in which the EC
addressed reduced concrete modulus of elasticity in the control building electric tunnel
and the containment enclosure building. The review was to determine if only a 50.59
screening was acceptable to accept "as-is" conditions for this concrete material property
which was degraded from the design bases as reflected in the UFSAR.
b. Observations/Findings
This issue was closed since the inspectors identified one Severity Level IV NCV, as
described in Section 1 R1 7 of this report.
.2 (Open) Unresolved Item 05000443/2011003-03, Open Operability Determinations for
Safety-Related Structures Affected by ASR
a. Inspection Scope
The scope of this review was to update NextEra actions. to date. The inspectors
reviewed the prompt operability determination for the control building, and for the extent
of conditions review for other buildings affected by the alkali silica reaction. The
inspectors utilized site records and interviews to develop the design basis for the safety
related structures in greater detail than summarized in section 3.8 of the Updated Final
Safety Analysis Report. Additionally, this review was assessed progress in the
development of a plan and schedule to address inspection activities, in-situ and
laboratory testing to address the alkali-silica-reaction degradation with specific focus on
the Control Building as a test case for review.
With respect to laboratory conditions 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 0 F) observed during preparation and testing; and 4) core samples
were adequately stored and labeled in individual bags. Particular care was 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
Enclosure
10
verified: 1) caliper (model 500-505-10, serial #0014816) calibration document and
calibration sticker on the caliper; and, 2) compression machine (model CM5000-D, serial
- 11005) calibration document and calibration sticker on the machine. With respect to
test technician qualifications, the inspector also verified qualification records (Level 2
qualification for concrete testing up to date). The inspector also reviewed the Altran
Commercial Grade Dedication Plan 10-0076 Part 05 - In Field Check List - the check list
was in hand during the preparation and testing.
During the week of November 28, 2011, the inspector continued to review historical
documentation from the construction phase of the plant, correlation 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 was discussed and
reviewed with cognizant engineering and management personnel. 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 IWL inspection of the containment, and the results of the
licensee's efforts in inspection and mapping of 'crazed' cracking in containment and
enclosure building wall and the adequacy and validity of the documented results.
b. Observations/Findinqs
The inspectors identified one finding which was addressed in section 1 R1 5 of this report.
In summary of the finding, the operability determinations did not fully evaluate design
information that was available on the potential effects of the reduced modulus of
elasticity in the following areas: overall stiffness and resulting seismic response,
concrete material property changes affecting natural frequency and therefore the
amplitude of the seismic response, and the performance of systems and components
attached to the affected structures during seismic events. Procedure EN-AA-203-1001
required that these impacts be evaluated as part of the prompt operability determination
based on available information.
The unresolved item was kept open because the operabilitiy determinations have not
been finalized. Several other observations were made in order to update this item. The
reviewers noted that NextEra had engaged knowledgeable vendors, appropriate
consultants, and recognized experts for testing, analysis, and evaluation of the effects of
alkali-silica-reaction, on the serviceability and safety of the affected structures. Although
the NextEra plan has not been refined before the NRC staff may be able to determine
that if it meets the rigor expected of an Appendix B Quality Assurance program, a
preliminary schedule of actions generally consistent with the proposed plans from the
contractors was underdevelopment. It was noted that some of the elements of an aging
management program had been developed and a final one was under development
putting for putting in place, including monitoring and trending. Additional core samples
were planned, the operability determination was to be updated as new information,
analysis, and assessments became available on an as needed basis. The extent of
condition was being comprehensively determined along with building initial assessments
through the work of its consultants and contractors. NextEra was also developing and
scheduling actions to determine where the alkali-silica-reaction was on the alkali
Enclosure
11
depletion curve (relates to extent of potential future degradation), although the tests may
take up to two years to complete, and provide reliable data.
On October 14, 2011, NextEra issued additional information in support of its engineering
evaluation of the alkali-silica-reaction impacted structures, including Calculation C-S-1-
10163, the Prompt Operability Determination for AR581434 (for the Control Building and
Electrical Tunnel), Revision 1, and the Prompt Operability Determination for AR1664399
(other Category I Structures), Revision 1. In C-S-1-10163, the fundamental frequency
was evaluated using the measured modulus of elasticity determined in concrete core
samples taken from the building walls. The calculation evaluated the impact of the
reduced modulus (compared to the design value) on the wall stiffness with respect to the
ground response spectra for the Seabrook site. The building response frequency was
calculated using the principles and equations of engineering mechanics for a uniformly
loaded fixed-fixed beam model (a simple span fixed at both ends during a seismic
event). The effect of the reduced modulus was similarly evaluated to assess the impact
on the natural frequency of the structures. The seismic analysis, for Seabrook described
in Updated Final Safety Analysis Report Section 3.7(B).2, was used in the design of
Category I Structures, systems and components at Seabrook.
The inspectors completed a detailed review of C-S-1-10163 and verified that the
calculation inputs were supported by plant data and the design references cited in the
calculation. No inadequacies were identified. The results of C-S-1-10163 supported
NextEra's conclusions in the revised prompt operability determinations, AR 581434 and
AR1 664399.
Also, during the week of November 14, 2011, a Region III inspector reviewed laboratory
testing for compressive strength on fifteen concrete core samples taken from the control
building in the October 2011 time frame. The testing was conducted at a laboratory in
Northbrook, Illinois. The scope of this review was as noted above. For the testing the
week of November 14, 2011, 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 during the cutting
phase to be used by Seabrook for further petrography in the near future. Sample
preparation (capping) was done in accordance with ASTM C617. Compression testing
was done in accordance with ASTM C39. No concerns were noted with respect to
quality control during all aspects of compression testing.
Other observations were made during the week of November 14, 2011. Multiple
laboratory engineers, licensee engineer, and Altran engineer were involved in making
call on fracture patterns. All but one of the obtained compressive strengths were fairly
consistent with previous lab's results (2010, 2011 data). Core sample L5-C exhibited
highest compressive strength of 6610 PSI whereas the previous lab's data identifies
strength at 3950 PSI. This core compressive strength value was the only apparent
outlier amongst the data set. All 15 destroyed cores are to be shipped back to Seabrook
later today including the cut samples to be used for petrography.
During the week of November 28, 2011, the IWL Examination Report for the Primary
Containment in October 2011, recorded information related to concrete conditions to
ensure no unacceptable surface conditions for cracking (greater than 40 mils) and report
Enclosure
12
on other conditions such spalling and discoloration conditions. Related to this review
the inspector noted in AR 01641413 an evaluation of crazed cracking on the exterior
surface of the primary containment at azimuth 3150 and elevation (-)30 feet, 00 inches.
While several factors were identified by NextEra in support of structural integrity of this
structure, it was noted the continued evaluation would be done in accordance with the
extent of condition review per AR 574120 which identified the loss of concrete strength
due to alkali-silica reaction (ASR) in other buildings noted herein. The inspector noted
that no specific cause for the crazed cracking was identified which could be due ASR or
other mechanisms. Further the inspector questioned the reliance on the April 2008 10
CFR 50 Appendix J, Type A Test at 49.6 psig that showed no evidence of cracking
greater than 40 mils without the use of a before and after crack mapping effort. No
unacceptable conditions were found and the extent of condition review noted above is a
part of this open unresolved item.
During this inspection, it was determined that NextEra and its contractor were
conducting a remodeling effort on the Containment Enclosure Building (an extent of
condition building reviewed in AR 01664399) using current data from core sample and
in-situ reviews such as crack mapping etc. The purpose of this remodeling was to
conduct a seismic reanalysis to demonstrate the effects of the reduced modulus on
structural response because of the unique design of the building and because of the
need to address a global response vs. a localized response. The completion of this
review was not expected until early 2012. NRC review determined that the initial
evaluation for the CEB did not address the response of the entire structure to seismic
loading comparable to the methods described in UFSAR 3.8 and how the induced
seismic stresses would shift between the concrete and the steel in adjoining sections of
the structure. In response, NextEra pointed out that they began development of an
analytical model to reanalyze the CEB using the as-measured elastic modulus (40%
reduced) applied to that ASR-impacted sections. The results of this analysis will be
further reviewed as a part of an Engineering Evaluation scheduled to be completed by
March 2012.
Notwithstanding the acceptable revised operability determination based on available
information, the inspectors noted that these determinations listed no assumptions in the
applicable sections. The inspectors also noted that design basis code ACI 318-1971
was based on empirical data for determining certain parameters that are a part of the
design bases. Also Poison's ratio on concrete cores are not being tested or evaluated
and this ratio was in the UFSAR. The assumption of this empirical data was that the
relationships were for ASR free concrete. Further, based on a review of the
implementation schedule from contractor-submitted plans that do not have NextEra
approval as of Nov. 2 9 th, the inspectors noted that the plans do not address the following
which are directly related to addressing the unwritten assumptions being made in the
prompt operability determinations:
1. The plans do not appear to test concrete cores for the following key design
parameters from the design basis code ACI 318-1971, as for tensile and shear
strength, rebar bond strength and Poison's ratio.
2. The plans do not appear to address nondestructive testing to assess the current
progression of the ASR expansion rate before the destructive tests of the concrete
cores.
Enclosure
13
3. There was a apparent lack a clear framework for concrete core sampling in the
buildings to ensure how representative the core sampling addressing the need for
random core sampling in distinction to smart sampling on worst case
conditions/doing a bounding calculation along addressing the impact of too much
core boring and re-grouting on the building structural integrity.
4. The plans do not appear to address potential effects of other degradations from an
aggressive groundwater environment along with the presence of ASR.
In summary at the close of the inspection, NextEra continued to work on: their plans and
implementing schedule, building initial assessments along with evaluation results for
additional core sampling; identifying the in-situ and out-of-situ testing (concrete core
samples and nondestructive testing of concrete core samples) for the structure areas
affected by alkali-silica-reaction; need to address key design parameters for the
buildings, such as compressive strength, tensile strength, bond strength (between rebar
and the concrete), modulus of elasticity and Poisson's Ratio in terms of how alkali-silica-
reaction has affected the non-alkali-silica-reaction functional relationship between these
parameters per the design code ACI 318-1971; and, remodeling efforts on the
Containment Enclosure Building.
Overall, this area remains open pending further project and test plan development by
NextEra and further NRC staff review of the final operability determinations on or about
March 2012.
4OA6 Meetings, Including 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, 2011 a final exit meeting was conducted and lead 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.
Enclosure
14
ATTACHMENT: SUPPLEMENTARY INFORMATION
Enclosure
A-1
ATTACHMENT
SUPPLEMENTARY INFORMATION
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
Key Manager Participants for Teleconference of January 12, 2012
NRC Staff
C. Miller, Director Division of Reactor Safety (DRS), Region I
R. Conte, Chief Engineering Branch No. 1, DRS, Region I
A. Burritt, Chief Reactor Projects Branch No. 3, Region I
LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED
Opened:
05000443/2011-010-01 FIN Inadequate Operability Determination for Degraded
Concrete Structures Housing Safety-Related Equipment
05000443/2011-010-02 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
05000443/2011-003-03 NCV Prompt Operability Determination for Safety Related
Structures affected by ASR.
Documents Reviewed:
Attachment
A
A-2
LIST OF ACRONYMS
AR ????
IMC Inspection Manual Chapter
NCV Non-Cited Violation
NRC U.S. Nuclear Regulatory Commission
NRR Nuclear ????
TS Technical Specification
URI Unresolved Item
Attachment