ML111360432
| ML111360432 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 05/23/2011 |
| From: | Conte R J Engineering Region 1 Branch 1 |
| To: | Freeman P NextEra Energy Seabrook |
| References | |
| IR-11-007 | |
| Download: ML111360432 (31) | |
See also: IR 05000443/2011007
Text
UNITED STATES NUCLEAR REGU LATORY COMMISSION
REGION I 475 ALLENDALE
ROAD KlNG OF PRUSSlA. PA 19406-1415
l{ay 23, 20IL Mr. Paul Freeman Site Vice President NextEra Energy Seabrook LLC P. O. Box 300 Seabrook, NH 03874 SUBJECT: NEXTERA ENERGY SEABROOK - NRC LICENSE RENEWAL INSPECTION
REPORT 05000443/201
1 007 Dear Mr.On April 8, 2011, the NRC completed
the onsite portion of the inspection
of your application
for license renewal of Seabrook Station. The NRC inspection
is one of several inputs into the NRC review process for license renewal applications.
The enclosed report documents
the results of the inspection, which were discussed
on March 28rh and April 8th with members of your staff.The purpose of this inspection
was to examine the plant activities
and documents
that support the application
for a renewed license of Seabrook Station. lnspectors
reviewed the screening and scoping of non-safety
related systems, structures, and components, as required in 10 CFR 54.4(a)(2), to determine
if the proposed aging management
programs are capable of reasonably
managing the effects of aging.The inspection
team concluded
screening
and scoping of non-safety
related systems, structures, and components, was implemented
as required in 10 CFR 54.4(a)(2), and the aging management
portion of the license renewal activities
were conducted
as described
in the License Renewal Application.
We noted that your staff continued
to develop an appropriate
initial response to the aging effect of the alkali-silica
reaction in certain concrete structures
of Seabrook Station. Because your investigation
and testing was ongoing and you were not currently
in a position to propose a new or revised aging management
program, the inspection
team was unable to arrive at a conclusion
about the adequacy of your aging management
review for the alkali-silica
reaction issue. As part of the ongoing review of your application
for a renewed license, you should continue to inform the Division of License Renewal as you develop your response to the alkali-silica reaction issue. With assistance
from our Headquarters
Office, Region I will review those key points in the implementation
of your project plan associated
with this issue to ensure the current licensing
bases is maintained, a key assumption
in the license renewal process.Except for the alkali-silica
reaction issue, the inspection
results support a conclusion
of reasonable
assurance
with respect to managing the effects of aging in the systems, structures, and components
identified
in your application.
The inspection
also concluded
the documentation
supporting
the application
was in an auditable
and retrievable
form.
P. Freeman In accordance
with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure
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 Website at http://www.nrc.qov/readinq-
rm/adams.html (the Public Electronic
Reading Room).Sincerely, 6LA.-/Petu
Richard J. Conte, Chief Engineering
Branch 1 Division of Reactor Safety Docket No. 50-443 License No. NPF-86 Enclosure:
Inspection
Report0500044312011007
cc Mencl: Distribution
via ListServ
P. Freeman In accordance
with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure
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 Website at http://www.nrc.qovireadinq-
rmladams.html (the Public Electronic
Reading Room).Sincerely,/RN Richard J. Conte, Chief Engineering
Branch 1 Division of Reactor Safety Docket No. 50-443 License No. NPF-86 Enclosure:
cc Mencl: Distribution
Mencl: (VlA E-MAIL)W. Dean, RA D. Lew, DRA P. Wilson, DRS A. Burrit, DRP C. LaRegina, DRP I nspection
Report 05000443/201
1 007 Distribution
via ListServ A. Williams, Rl OEDO ROPreports@nrc.gov
D. Bearde, DRS Region I Docket Room (with concurrences)
SUNSI Review Gomplete: MCM/RJC (Reviewer's
lnitials)ADAMS ACC#MLI11360432
DOCUMENT NAME: G:\DRS\Engineering
Branch 1\_Technical
lmportance\Seabrook
Concrete\SbkLRl
Rpts\05000443
201 1 007 lP7 1 OO2 Sbrk I nsp Rpt Final. 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 attachmenVenclosure "E" = Copy Wth attachmenVenclosure "N" = No ost18t11 OFFICIAL RECORD COPY
Docket No: License No: Report No: Licensee: Facility: Location: U. S. NUCLEAR REGULATORY
COMMISSION
REGION I 50-443 NPF-86 05000443/2011007
NextEra Energy Seabrook LLC Seabrook Station Seabrook, NH March 7-11,21-25, and April 4-8,2011 M. Modes, Team Leader, Division of Reactor Safety (DRS)G. Meyer, Sr. Reactor Inspector, DRS S. Chaudhary, Reactor Inspector, DRS J. Lilliendahl, Reactor Inspector, DRS Richard J. Conte, Chief Engineering
Branch 1 Division of Reactor Safety
SUMMARY OF FINDINGS lR 0500044312011007;
March 7-11,21-25, and April 4-8,2011, Seabrook Station; Inspection
of the Scoping of Non-Safety
Systems and the Proposed Aging Management
Procedures
for the NextEra Energy Seabrook LLC Application
for Renewed License for Seabrook Station.This inspection
of license renewal activities
was performed
by four regional office engineering
inspectors.
The inspection
was conducted
in accordance
with NRC Manual Chapter 2516 and NRC lnspection
Procedure
71002. This inspection
did not identify any "findings" as defined in NRC Manual Chapter 0612. The inspection
team concluded
screening
and scoping of non-safety related systems, structures, and components, were implemented
as required in 10 CFR 54.4(a)(2), and the aging management
portions of the license renewal activities
were conducted as described
in the License RenewalApplication.
Except for the alkali-silica
reaction issue, the inspection
results support a conclusion
of reasonable
assurance
with respect to managing the effects of aging in the systems, structures, and components
identified
in your application.
The inspection
concluded
the documentation
supporting
the application
was in an auditable
and retrievable
form.
40.A2 1 REPORT DETAILS Other - License Renewal Inspection
Scope This inspection
was conducted
by NRC Region I based inspectors
in order to evaluate the thoroughness
and accuracy of the screening
and scoping of non-safety
related systems, structures, and components, as required in 10 CFR 54.4(a)(2)
and to evaluate whether aging management
programs will be capable of managing the identified
aging effect in a reasonable
manner.The team selected a number of systems for review, using the NRC accepted guidance;
in order to determine
if the methodology
applied by the applicant
appropriately
captured the non-safety
systems affecting
the safety functions
of a system, component, or structure within the scope of license renewal.The team selected a sample of aging management
programs to verify the adequacy of the applicant's
documentation
and implementation
activities.
The selected aging management
programs were reviewed to determine
whether the proposed aging management
implementing
process would adequately
manage the effects of aging on the system.The team selected risk significant
systems and conducted
a review of the Aging Management
Basis documents
for each selected system to determine
if the applicant
had adequately
applied the Aging Management
Programs to ensure that reasonable
assurance
exists for the monitoring
of aging effects on the selected systems.The team reviewed supporting
documentation
and interviewed
applicant
personnel
to confirm the accuracy of the license renewal application
conclusions.
For a sample of plant systems and structures, the team performed
visual examinations
of accessible
portions of the systems to observe aging effects.Scopinq of Non Safetv-Related
Svstems. Structures.
and Components
under 10 CFR 54.4 (a) (2)For scoping the team reviewed program guidance procedures
and summaries
of scoping results for Seabrook Station to assess the thoroughness
and accuracy of the methods used to bring systems, structures, and components
within the scope of license renewal into the application, including
non-safety-related
systems, structures, and components, as required in 10 CFR 54.4 (a)(2). The team determined
that the procedures
were consistent
with the NRC accepted guidance in Sections 3, 4, and 5 of Appendix F to Nuclear Energy Institute (NEl) 95-10, Rev. 6, "lndustry
Guideline
for lmplementing
the Requirements
of 10 CFR Part 54," (Section 3: non-safety-related
systems, structures, and components
within scope of the current licensing
basis; Section 4: non-safety-related
systems, structures, and components
directly connected
to safety-related
systems, structures, and components;
and Section 5: non-safety-related
systems, structures, and components
not directly connected
to safety-related
systems, structures, and a.Enclosure
2 components).
The team noted that scoping guidance was not clear regarding
structural
descriptions.
By drawing reviews and in-plant walk-downs, the team identified
that the few scoping errors related to the guidance inconsistencies
were conservative, i.e., components
were placed within the scope of license renewalwhich
were not required to be included.
Subsequently, the applicant
revised the scoping guidance, and the team reviewed the revised guidance.The team reviewed the set of license renewal drawings submitted
with the Seabrook Station License Renewal Application, which was color-coded
to indicate non-safety
related systems and components
in scope for license renewal. The drawings included numerous explanatory
notes, which described
the basis for scoping determinations
on the drawings.
The team interviewed
personnel, reviewed license renewal program documents, and independently
inspected
numerous areas within Seabrook Station, to confirm that appropriate
non-safety-related
systems, structures, and components
had been included within the license renewal scope; that systems, structures, and components
excluded from the license renewal scope had an acceptable
basis; and that the boundary for determining
license renewal scope within the systems, including
seismic supports and anchors, was appropriate.
The Seabrook Station in-plant areas reviewed included the following:. Turbine Building o Primary Auxiliary
Building. East Main Steam & Feedwater
Pipe Chase o West Main Steam & Feedwater
Pipe Chase. Control Building. Service Water Pumphouse e Emergency
Pumphouse
and Pre-Action
Valve Building o Steam Generator
Blowdown Building o Emergency
Diesel Generator
Room B. RCATunnel. Tank Farm Area For systems, structures, and components
selected regarding
spatial interaction (failure of non-safety-related
components
adversely
affecting
adjacent safety-related
components), the team determined
the in-plant configuration
was accurately
and acceptably
categorized
within the license renewal program documents.
The team determined
the personnel
involved in the process were knowledgeable
and appropriately
trained.For systems, structures, and components
selected regarding
structural
interaction (seismic design of safety-related
components
dependent
upon non-safety-related
components), the team determined
that structural
boundaries
had been accurately
determined
and categorized
within the license renewal program documents.
The team determined
that the applicant
had thoroughly
reviewed applicable
isometric
drawings to determine
the seismic design boundaries
and had correctly
included the applicable
components
in the license renewal application, based on the inspector's
independent
Enclosure
3 review of a sample of the isometric
drawings and the seismic boundary determinations
combined with in-plant review of the configurations.
ln summary, the team concluded
that the applicant
had implemented
an acceptable
method of scoping of non-safety-related
systems, structures, and components
and that this method resulted in accurate scoping determinations.
Proorams 8.2.1.9 Bolting Inteqritv The Seabrook Station Bolting Integrity
Program is an existing program that manages the aging effects of cracking due to stress corrosion
cracking, loss of material due to general, crevice, pitting, and galvanic corrosion;
Microbiologically
induced corrosion, fouling and wear; and loss of preload due to thermal effects, gasket creep, and self-loosening
associated
with bolted connections.
The program manages these aging effects through the performance
of periodic inspections.
The program also includes repair/replacement
controls for ASME Section Xl related bolting and generic guidance for material selection, thread lubrication
and assembly of bolted joints.The inspector
reviewed the program basis documents, program description, baseline inspection
results, subsequent
inspection
results for trending, and implementing
procedures
to determine
the scope and technical
adequacy of the Program. Also, the team reviewed action requests (ARs) to assess the adequacy of evaluations
of findings, and resolution
of concerns, if any, identified
in these inspections.
The inspector
noted that the program follows the guidelines
and recommendations
provided in NUREG-1339, "Resolution
of Generic Safety lssue 29; Bolting Degradation
or Failure of Bolting in Nuclear Power Plants", EPRI NP-5769, "Degradation
and Failure of Bolting in Nuclear Power Plants" (with the exceptions
noted in NUREG- 1339), and EPRI TR-104213, "Bolted Joint Maintenance
and Application
Guide" for comprehensive
bolting maintenance.
However, indications
of aging identified
in ASME pressure retaining bolting during In-service
Inspection
are evaluated
per ASME Section Xl, Subsections
3600. lndications
of aging identified
in other pressure retaining
bolting, nuclear steam supply system component
supports, or structural
bolting are evaluated
through the Corrective
Action Program, This program covers bolting within the scope of license renewal, including:
1. safety- related bolting, 2. bolting for nuclear steam supply system component
supports, 3. bolting for other pressure retaining
components, including
non-safety
related bolting; and, 4. structural
bolting.The aging management
of reactor head closure studs is addressed
by Seabrook Station Reactor Head Closure Studs Program (8.2.1.3)
and is not part of this program l Enclosure
4 B.2.1.13 lnspection
of Overhead Heavy Load And Liqht Load (Related To Refuelinq)
Handlino Svstems The Seabrook Station Inspection
of Overhead Heavy Load and Light Load (Related to Refueling)
Handling Systems Program is an existing program that will be enhanced to manage the aging effects of loss of material due to general corrosion
and due to wear of structural
components
of lifting systems and the effects of loss of material due to wear on the rails in the rail system, for lifting systems within the scope of license renewal.Included in scope are those cranes encompassed
by the Seabrook Station commitments
to NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," plus two cranes related to fuel handling.The team reviewed the program basis documents, program description, baseline inspection
results, subsequent
inspection
results for trending, and implementing
procedures
to determine
the scope and technical
adequacy of the Program. Also, the team reviewed ARs and work orders to assess the adequacy of evaluations
of findings, and resolution
of concerns, if any, identified
in these inspections.
The team noted that the program manages loss of material due to general corrosion
on structural
steel members and rails of the cranes within the scope of license renewal including
the structural
steel members of the bridges, trolleys and monorails.
The program also manages loss of material due to wear on rails. Only the structural
portions of the in-scope cranes and monorails
are in the scope of this program. The individual
components
of these overhead handling systems that are subject to periodic replacement, or those which perform their intended function through moving parts or a change in configuration, are not in the scope of this program.Structural
inspections
are conducted
under the Seabrook Station lifting systems manual.Periodic inspections
are conducted
at the frequencies, and include the applicable
items, delineated
in ANSI 830.2, "Overhead
and Gantry Cranes," ANSI B30.1 1, "Monorails
and Under hung Cranes," ANSI 830.16, " Overhead Hoists (Under-hung)," and ANSI 830.17,"Overhead
and Gantry Cranes (Top Running Bridge, Single Girder, Under-hung
Hoist)" for a periodic inspection
and in accordance
with the manufacturer's
recommendations.
Inspections
are conducted
yearly. All periodic inspections
are documented
on work orders.The enhancement
to the program includes: 1. The Seabrook Station lnspection
of Overhead Heavy Load and Light Load (Related to Refueling)
Handling Systems Program Lifting System Manualwill
be enhanced to include monitoring
of general corrosion
on the crane and trolley structural
components
and the effects of wear on the rails in the rail system;2. The Seabrook Station Inspection
of Overhead Heavy Load and Light Load (Related to Refueling)
Handling Systems Program Lifting Systems Manualwill
be enhanced to list additional
cranes related to the refueling
handling system.Enclosure
5 8.2.1.16 Fire Water Svstem The Fire Water System Program is an existing program modified to manage the effects of aging on fire water system components
through detailed inspections.
Specifically, the program manages the following
aging effects: loss of material due to general, crevice, pitting, galvanic, and microbiologically
influenced
corrosion;
fouling; and reduction
of heat transfer due to fouling of the Fire Water System components.
The Seabrook Station Fire Water System Program manages aging of the following system components:
sprinklers, nozzles, fittings, filters, valves, hydrants, hose stations, flow gages and flow elements, pumps, standpipes, aboveground
and underground
piping and components, water storage tanks, and heat exchangers.
The Seabrook Station Fire Protection
Manual incorporates
activities, such as inspections, flushing, and testing, that serve to prevent or manage aging of the fire water system.Specific examples include: inspections
of fire hydrants, fire hydrant hose hydrostatic
tests, gasket inspections, and fire hydrant flow tests.The Seabrook Station procedures
are being enhanced to require the following:
inspection
sampling or replacing
of sprinklers
after 50 years of service, flow testing of the fire water system in accordance
with National Fire Protection
Association (NFPA) 25 guidelines, and periodic visual or volumetric
inspection
of the internal surface of the fire protection
system.The team interviewed
the system engineer to understand
historical
and current conditions
of the system. The team reviewed the current program and existing maintenance/surveillance
procedures
to verify the adequacy of the program for detecting and managing aging effects. The team reviewed condition
reports to verify that all known aging effects will be managed by the new program. The team conducted
a walkdown of accessible
portions of system including
the electrical
area, cable spreading room, water storage tanks, and fire pumps to assess the material condition
of the accessible
fire water system piping.Based on questions
from the team, the applicant
modified the application
to specify that the flow testing will be done in accordance
with the 2002 version of NFPA 25. (License Renewal Application
change letter SBK-L-1 1063). Also, based on questions
from the team, the applicant
modified the application
to correct the types of fire water buried piping. The original application
stated that the fire water piping was polyvinylchloride
and carbon steel. The correct materials
were determined
to be fiberglass
and carbon steel (License Renewal Application
change letter SBK-L-1 1054).l Enclosure
6 B.2.1.17 Aboveqround
Steel Tanks The Aboveground
SteelTanks
program is a new program used to manage the aging effects of the external surfaces of five aboveground
steel tanks within the scope of license renewal. The five tanks within scope are: r Auxiliary
Boiler Fuel Oil Storage Tank 1-AB-TK-29
o Fire Protection
Fuel OilTank 1-FP-TK-3S-A
r Fire Protection
Fuel OilTank 1-FP-TK-3S-B
o Fire Protection
Water Storage Tank 1-FP-TK-36-A
o Fire Protection
Water Storage Tank 1-FP-TK-36-B
The Auxiliary
Boiler Fuel Oil Storage Tank 1'AB-TK-?9
has been abandoned.
lt is included in the application
as part of the planning to renovate the tank and return it to service. All the tanks have protective
The Fire Protection
Water Storage Tanks are placed on a concrete pad, leveled using oiled sand, and the edges caulked.The inspector
walked-down
each of the above tanks. The path chosen by NextEra to monitor this area was tank level monitoring.
For example, blistered
paint, with rust and rust stains was noted on Fire Protection
Storage Tanks. The tank bottom to concrete pad intersection
was caulked; however, there was evidence of cracking and peeling of the caulk. Moisture was present at this intersection
and it was not possible to tell if the water was from the tank or local inclement
weather conditions.
The inspector
verified the blistered
paint with rust, and rust staining was noted in the corrective
action program.The inspector
also determined, as evidenced
by the documented
results, that daily operator surveillance
included the water level of the Fire Protection
Storage Tanks. lf the moisture at the bottom of the tank represented
a leak, it would be reflected
in an unanticipated
change in level.The Aboveground
Steel Tanks program is credited with managing loss of material on the tank external surfaces including
the exterior bottom surface of tanks that is not accessible
for direct visual inspection.
The outer surfaces of the tanks, up to the surface in contact with the concrete foundation, are managed by visual inspection.
Ultrasonic
thickness gauging will be used to monitor loss of material on the inaccessible
tank bottom external surfaces.8.2.1.20 One-Time lnspection
The One-Time Inspection
Program is a new, one-time program for Seabrook Station that will be implemented
prior to the period of extended operation.
The program will verify the effectiveness
of other aging management
programs, including
Water Chemistry, Fuel Oil Chemistry, and Lubricating
OilAnalysis
Programs, by reviewing
various aging effects for impact. Where corrosion
resistant
materials
and/or non-corrosive
environments
exist, the One-Time Inspection
Program is intended to verify that an aging management
program is not needed during the period of extended operation
by confirming
that aging effects are not occurring
or are occurring
in a manner that does not affect the safety function of systems, structures, and components.
Non-destructive
examinations
will be performed Enclosure
7 by qualified
personnel
using procedures
and processes
consistent
with the approved plant procedures
and appropriate
industry standards.
The team reviewed application
section 8.2.1.20, results of the NRC aging management
program audit, and applicant
responses
to requests for additional
information (RAls).The team reviewed the aging management
program basis document and draft implementing
guidance, discussed
the planned activities
with the responsible
staff, including
sampling plan, and reviewed a sample of corrective
action program documents for applicable
components.
8.2.1.21 Selective
Leachino of Materials The Selective
Leaching of Materials
Program is a new, onetime program for Seabrook Station that will be implemented
prior to the period of extended operation.
The program is credited with managing the aging of components
made of gray cast iron, copper alloys with greater than 15olo zinc, and aluminum bronze with greater than 8% aluminum, exposed to raw water, treated water, and soil environments, which may lead to the selective
leaching of material constituents, e.9., graphitization
and dezincification.
The program will include a one-time visual inspection
and hardness measurement
test of selected components
that may be susceptible
to selective
leaching to determine
whether loss of material due to selective
leaching is occurring, and whether the leaching process will affect the ability of the components
to perform their intended function during the period of extended operation.
ln 1998 Seabrook operating
experience
identified
selective leaching on aluminum bronze components
in sea water. As such, Seabrook will include periodic inspections
for selective
leaching of aluminum bronze as part of this aging management
program.The team reviewed application
section 8.2.1.21, results of the NRC aging management
program audit, and applicant
responses
to requests for additional
information (RAls).The team reviewed the aging management
program basis document and draft implementing
guidance, discussed
the planned activities
with the responsible
staff, including
sampling plan, and reviewed a sample of corrective
action program documents for applicable
components
and for corrective
actions to the selective
leaching of aluminum bronze.B.2.1.22 Buried Pipinq and Tanks Inspection
The Seabrook Station Buried Piping and Tanks Inspection
Program is a new program that includes coating, cathodic protection, and backfill quality as preventive
measures to mitigate corrosion.
Periodic inspections
manage the aging effects of corrosion
on buried piping in the scope of license renewal. Buried steel and stainless
piping has an external protective
coating consisting
of coal-tar primer, coal-tar enamel, asbestos felt or fibrous glass mat, and a wrapping of kraft paper or coat of whitewash.
Some hot-applied
tape coating was also used. Coatings were fabricated
and applied in accordance
with the requirements
of American Water Works Association
specification
C203 and this required"holiday" (flaws in coating) testing.Enclosure
8 Backfill was applied in accordance
with Seabrook Specification
9763-8-1, "Bedding, Backfilling
and Compaction
for Miscellaneous
Non Safety Related Piping" and 9763-8-5"Bedding, Backfilling
and Compaction
for Safety Related Systems and Structures".
Except for the allowance
of backfill at a size of 1/z" the backfill is equal to or better than the GALL Revision 2 proposal of ASTM D 448-08 Size 67. As a consequence, NextEra is proposing
inspection
in conformance
with an acceptable
backfill limit until a discovery is made of coating damage. For steel with cathodic protection, they propose 1 inspection.
lf backfill damage is discovered, they will increase this by another 3 samples.For steel without cathodic protection, they propose 4 inspections;
and if backfill damage is discovered, they will expand by another 4 inspections.
The team reviewed cathodic protection
system reports and determined
the system was in disrepair
since being identified
as unreliable
in 1993. The system was not restored until 2007 when a survey found that only 620/o of the areas surveyed were being mitigated
During the first quarter of 2009 the cathodic protection
system was finally categorized
as green (or satisfactory
condition).
system was made a Maintenance
Rule (10 CFR 50.65) System during the same quarter.There is an adequate historical
basis to conclude that buried piping was adequately
protected, and the backfill correctly
specified
and filled, during construction.
There is an absence of buried piping problems at the site. Because there was an absence of a consistent
for a period of 1993 to 2009, it is appropriate
for NextEra to inspect buried piping by excavation
to corroborate
the historical
basis.B.2.1.23 One-Time Inspection
of ASME Code Class 1 Small Bore Pipinq The One-Time Inspection
of ASME Code Class 1 Small Bore Piping Program is a new program that manages the aging effect of cracking in stainless
steel small-bore
ASME Code Class 1 piping less that 4 inches nominal pipe size, including
pipe, fittings, and branch connections.
Seabrook has not experienced
a small bore piping failure due to stress corrosion
or thermal and mechanical
loading. The small bore piping selected for insonification
is based on EPRI Report 1011955, "Management
of Thermal Fatigue in Normally Stagnant Non-lsolable
Reactor Coolant System Branch Lines (MRP-146)", issued June 2005 and the supplementalguidance
issued in EPRI Report 1018330,"Management
of Thermal Fatigue in Normally Stagnant Non-isolable
Reactor Coolant System Branch Lines - Supplemental
Guidance (MRP-1465)
issued December of 2008.Using these criteria the applicant
has identified
448 welds, of which 157 are socket welds (including
58 in-core instrument
guide tube welds) and 291 butt welds. In this population
there are 6 small bore stagnant segments susceptible
to thermalfatigue.
These are in the two charging lines and four high head safety injection
lines. These locations
are monitored.
Twenty-Nine
(29) welds (4 socket and 25 butt welds) have been identified
in the 448 candidates
as vulnerable
to cracking.
These will be tested using ultrasonic
inspection
not sooner than 10 years before the extended period of operation.
Enclosure
9 B.2.1.25 Inspection
of lnternal Surfaces in Miscellaneous
Pipinq and Ductino Components
The Inspection
of Internal Surfaces in Miscellaneous
Piping and Ducting Components (lnternal
Surfaces)
Program is a new program that will inspect the internals
of piping, piping components, ducting, and other components
of various materials
to manage the aging effects of cracking, loss of material, reduction
of heat transfer, and hardening
of elastomers.
The inspections
of opportunity
will occur during maintenance
and surveillance
activities
when systems are opened.The team reviewed application
section 8.2.1.25, draft NRC aging management
program audit, and applicant
responses
to requests for additional
information (RAls). The team reviewed the aging management
program basis document, operating
experience
review documents, draft implementing
guidance, and relevant condition
reports. The team interviewed
applicable
plant personnel.
B.2.1.26 Lubricatinq
Oil Analvsis The Lubricating
OilAnalysis
Program is an existing program, which maintains
oil systems free of contaminants (primarily
water and particulates), thereby preserving
an environment
that is not conducive
to loss of material, cracking, or fouling. The applicant performs sampling, analysis, and trending of results on numerous systems to provide an early indication
of adverse equipment
condition
in the lubricating
oil environment.
The applicant
samples the lubricating
oil for most of the affected equipment
on frequencies
recommended
by the vendor.The team reviewed application
section 8.2.1.26, draft NRC aging management
program audit, and applicant
responses
to requests for additional
information (RAls). The team reviewed the aging management
program basis document, operating
experience
review documents, existing procedures, relevant condition
reports, and system health reports.The team interviewed
plant personnel
and sampled oil measurement
results and trending within the applicant's
database.
Further, the team performed
walk downs of the lubricating
oil components
of B emergency
diesel generator.
The team identified
an issue regarding
the existing lubricating
oil practice on testing for water content. Specifically, the applicant
tests for water content on lubricating
oil for pumps and motors when these components
are water-cooled
and have the potential
for water contamination.
Nonetheless, the team identified
that the lubricating
oil and hydraulic
fluid samples of charging pump 1-CS-P-128
were not being tested for water content despite the pump being water-cooled.
The applicant
issued Action Request 01632769 to correct the testing for water content on this pump, to confirm test packages for other components
are correct, and to review the testing for water content of all pumps and motors as part of the enhancement
to the program to provide a program attachment
with the required equipment
and the specified
sample analyses and frequency.
Enclosure
10 B.2.1.27 ASME Section Xl. Subsection
IWE The ASME Section Xl, Subsection
program is an existing program, credited in the LRA, which provides for inspecting
the reactor building liner plate and related components
for loss of material, loss of pressure retaining
bolting preload, cracking due to cyclic loading, loss of sealing, and leakage through seals, gaskets and moisture barriers in accordance
with ASME Section Xl. Areas of the reactor building adjacent to the moisture barrier and the moisture barrier are subject to augmented examination.
The team reviewed applicable
procedures, the latest lnservice
Inspection
program results and interviewed
the Inservice
lnspection
program manager. The team reviewed a sample of recent corrective
action reports from Section IWE examinations.
The team concluded
that the Inservice
Inspection
program was in place, had been implemented, was an on-going program subject to NRC review, and included the elements identified
in the license renewalapplication.
8.2.1.28 ASME Section Xl. Subsection
IWL The Seabrook Station ASME Section Xl, Subsection
IWL Program is an existing program that manages the aging effects of cracking, loss of bond, loss of material (spalling, scaling) due to corrosion
of embedded steel, expansion
and cracking due to reaction with aggregates, increase in porosity and permeability, cracking, loss of material (spalling, scaling) due to aggressive
chemical attack, and increase in porosity and permeability, loss of strength due to leaching of calcium hydroxide.
The team reviewed the program basis documents, program description, baseline inspection
results, subsequent
inspection
results for trending, and implementing
procedures
to determine
the scope and technical
adequacy of the Program. Also, the team reviewed ARs to assess the adequacy of evaluations
of findings, and resolution
of concerns, if any, identified
in these inspections.
The team observed that the program complies with the requirements
of ASME Section Xl, Sub-Section
lWL, "Requirements
for Class CC Concrete Components
of Light-Water
Cooled Power Plants". The components
examination
contained
in 10 CFR 50.55a in accordance
with ASME Boiler and Pressure Vessel Code, Section Xl, Subsection
IWL managed by the program include steel reinforced
concrete for the Seabrook Station containment
building and complies with the requirement
for examination
contained
in 10 CFR 50.55a in accordance
with ASME Boiler and Pressure Vessel Code, Section Xl, Subsection
lWL.The primary inspection
method used at Seabrook Station is W-1C visual examination, W-3C visualexamination, and alternative
examination
methods (in accordance
with IWA-2240).
The Seabrook Station ASME Section Xl, Subsection
IWL Program provides acceptance
criteria and corrective
actions for each exam type. The team noted, for this program and the structures
monitoring
program, a technically
acceptable
trending system was not implemented
to establish
the status of observed cracks (stable or active), and I Enclosure
11 qualification
and certification
of inspectors/examiners
was not explicitly
established
and documented
to assure assignment
of qualified
individuals
for inspection.
The inspection
personnel
selection
is left to the supervisor
of the group. Also, there was a lack of clear quantitative
acceptance/evaluation
criteria established
by the procedure
to assure consistency
in observation, evaluation, and assessment
of inspection
results by different inspectors
and technical
personnellengineers
and at different
times. This program will be further enhanced with revised implementing
procedures
to include definition
of"Responsible
Enginee/'(letter
SBK-L-10204, RAl 8.2.1.28-3, Commitment
No. 31) and trending information
and acceptance
criteria (same letter, RAI 8.2.1 .28-1).Concrete degradation
due to alkai-silica
reaction is an aging effect that was recentlydiscovered
for Seabrook Station. In addition to the control building, it had been noted in other buildings
such as Emergency
Diesel Generator
Building, and the Residual Heat Removal Vault (see additional
details in the section b of this report). The Team reviewed applicant
photographs
of pattern cracking on the primary containment
wall in the annulus region. The annulus region appears to have had approximately
six feet of water for an extended period of time due to groundwater
infiltration.
NextEra plans to keep the area drained (Letter SBK-L-11063
commitnment
No. 52) and to review, analyze, and assess the effect of this condition
in order to determine
the cause on the primary containment (AR 01641413, Crazed Crack Pattern On Containment
In Annulus Area).8.2.1.31 Structures
Monitorinq
Prooram The Structures
Monitoring
Program at Seabrook Station is an existing program that is to be further enhanced to be consistent
with guidance set forth in 10 CFR 50.65,"Requirements
for Monitoring
the Effectiveness
of Maintenance
at Nuclear Power Plants", NUMARC 93-01, "lndustry
Guidelines
for Monitoring
the Effectiveness
of Maintenance
at Nuclear Power Plants", and Regulatory
Guide 1.160, Rev. 2, "Monitoring
the Effectiveness
of Maintenance
at Nuclear Power Plants". This program is described
in Appendix B, Section 2.39 tor the license renewal application.
The applicant
uses the structural
monitoring
program to monitor the condition
of structures
and structural
components
within scope of the Maintenance
Rule, thereby providing
reasonable
assurance
that there is no loss of intended function of structure
or structural
component.
As noted in the application, the program will be enhanced to include: additional
structures
and structural
components
identified
in the license renewalaging
management
review, add aging effects, additional
locations, inspection
frequency, and ultrasonic
test requirements
and enhancements
for procedures
to include inspection
opportunities
when planning excavation
work that would expose inaccessible
concrete.
Enhancements
to the Structural
Monitoring
Program will be implemented
prior to the period of extended operation.
Aging effects or material degradation
in concrete identified
within the scope of the Structures
Monitoring
Program such as loss of material, cracking, change in material properties, and loss of form are detected by visual inspection
of external surfaces prior to the loss of the structure's
or component's
intended function.The team reviewed the Aging Management
Program description
for the Structural
Monitoring
Program, the Program Evaluation
Document for the Structural
Monitoring
Program, engineering
documents, inspection
reports, condition
reports, corrective
action Enclosure
12 documents, work request documents, site procedures, and related references
used to manage the aging effects on the structures.
During the inspection
the team conducted
a general walkthrough
inspection
of the site, including
the turbine building, reactor containment
building, diesel generator
building, control room, the intake structure, and other applicable
structures, systems, and components
related to the Structural
Monitoring
Program. The team held discussions
with applicant's
supervisory
and technical personnel
to verify that areas where signs of degradation, such as spalling, cracking, leakage through concrete walls, corrosion
of steel members, deterioration
of structural
materials
and other aging effects, had been identified
and documented.
Also, the team verified that the applicant
maintains
appropriate (photographic
and/or written)documentation
of these inspections
to facilitate
effective
monitoring
and trending of structural
deficiencies
and degradations.
Through the review of documents, walkthrough
inspections, and discussions
with engineering
and plant personnel, the inspector
identified
some weaknesses
in the structural
program. Similar to the IWL program, the inspector observed the need for clarification
on acceptance
criteria and the responsible
engineer performing
inspections.
The applicant
agreed to the needed changes as noted in the IWL program 8.2.1.27 (previous
section).As noted in the IWL program, concrete degradation
due to alkai-silica
reaction is an aging effect that was recently discovered
for Seabrook Station (see additional
details in the section b of this report).8.2.1 .32 Electrical
Cables and Connections
Not Subiect to 10 CFR 50.49 EQ Requirements
The Electrical
Cables and Connections
Not Subject To 10 CFR 50.49 Environmental
Qualification
Requirements
Program is a new program that will manage the aging effects of embrittlement, cracking, discoloration
or surface contamination
leading to reduced insulation
resistance
or electrical
failure of accessible
cables and connections
due to exposure to an adverse localized
environment
caused by heat, radiation
or moisture in the presence of oxygen. This program applies to accessible
cables and connections
installed
in in-scope structures.
This program will visually inspect accessible
electrical
cables and connections
installed
in adverse localized
environments
at least once every 10 years. The first inspection
for license renewal is to be completed
before the period of extended operation.
An adverse localized
environment
is defined as a condition
in a limited plant area that is significantly
more severe than the specified
service environment (i.e. temperature, radiation, or moisture)
for the cable or connections.
The team conducted
walkdowns
to observe cable and connector
conditions
in potential adverse localized
environments.
The team reviewed condition
reports and interviewed
plant personnelto
assess historical
and current conditions.
The team reviewed the draft program documents
to verify the program will be able to manage aging effects.Enclosure
13 8.2.1.34 Inaccessible
Power Cables Not Subiect To 10 CFR 50.49 EQ Requirements
The Inaccessible
Power Cables Not Subject to 10 CFR 50.49 Environmental
Qualification
Requirements
Program is a new program that will manage the aging effects of localized damage and breakdown
of insulation
leading to electricalfailure
of inaccessible
power cables (400V and higher) due to adverse localized
environments
caused by exposure to significant
moisture.
Seabrook Station defines an adverse localized
environment
for power cables as exposure to moisture for more than a few days.The Seabrook Station program includes periodic inspections
of manholes containing
in-scope cables. The inspection
focuses on water collection
in cable manholes, and draining water, as needed. The frequency
of manhole inspections
for accumulated
water and subsequent
pumping will be based on plant specific operating
experience, The maximum time between inspections
will be no more than one year.ln addition to periodic manhole inspections, in-scope cables are tested to provide an indication
of the condition
of the conductor
insulation.
The specific type of test performed will be determined
prior to the initial test, and is a proven test for detecting
deterioration
of the insulation
system due to wetting, such as power factor, partial discharge, or polarization
index or other testing that is state-of-the-art
at the time the test is performed.
Cable testing will be performed
prior to entering the period of extended operation
and at least every six years thereafter.
Overall actions are to test cables and keep them dry. Seabrook has had, and continues to get, some water in their manholes.
NextEra is taking corrective
actions by increasing
the inspection
frequency
and pumping frequently
to prevent submergence
of safety-related cables. They are committing
to having initial inspections
done and adjust inspection/pumping
frequencies
based on experience.
The team interviewed
the responsible
system engineer to understand
the proposed program and power cable operating
experience
at Seabrook.
The team reviewed data from previous manhole inspections
to verify the established
inspection
frequencies
are commensurate
with operating
experience.
The team observed the inspection
of a below-ground manhole at Seabrook to assess the process for inspections
and the material condition
of the manhole. The team reviewed system health reports and condition reports for historical
operating
experience
and program guidance for cable condition monitoring
to assess the adequacy of the proposed program to manage aging effects.B.2.1.35 Metal Enclosed Bus The Metal Enclosed Bus Program is a new program that will manage the following
aging effects of in-scope metal enclosed buses: loosening
of bolted connections
due to thermal cycling and ohmic heating; hardening
and loss of strength due to elastomer
degradation;
loss of material due to general corrosion;
and embrittlement, cracking, melting, swelling, or discoloration
due to overheating
or aging degradation
This new program will be implemented
prior to entering the period of extended operation and at least once every 10 years thereafter.
Enclosure
14 The internal portions of the in-scope metal enclosed bus enclosures
will be visually inspected
for aging degradation
of insulating
material and for cracks, corrosion, foreign debris, excessive
dust buildup, and evidence of moisture intrusion.
The bus insulation
will be visually inspected
for signs of embrittlement, cracking, melting, swelling, or discoloration, which may indicate overheating
or aging degradation.
The internal bus supports will be visually inspected
for structural
integrity
and signs of cracks. The accessible
bus sections will be inspected
for loose connections
using thermography
from outside the metal enclosed bus through the viewport, while the bus is energized.
The team reviewed previous work orders for inspection
and cleaning activities
for metal enclosed buses. The team interviewed
the associated
system engineer and reviewed condition
reports to assess the historical
and current condition
of the metal enclosed buses. The team conducted
a walkdown of accessible
portions of the metal enclosed buses to evaluate the exterior condition
of the buses and the operating
environment.
8.2.2.1 34 5 kV SFG Bus The Seabrook Station 345kV SF6 Bus Program is a new plant-specific
program that will manage the following
aging effects on the 345kV SF6 Bus: loss of pressure boundary due to elastomer
degradation;
loss of material due to pitting; crevice and galvanic corrosion;
and loss of function due to unacceptable
air, moisture or sulfur dioxide (SOz)levels.Sulfur Hexafluoride (SF6) is an inert gas used to insulate bus conductors.
The program will inspect for corrosion
on the exterior of the bus duct housing, test for leaks at elastomers, and periodically
test gas samples to determine
air, moisture and SOz levels.Inspections, leak testing, and gas sampling will be done prior to entering the period of extended operation
and at least once every six months thereafter.
The team reviewed previous work orders for maintenance
activities
associated
with inspections
of the SF6 buses and SFo gas monitoring.
The team interviewed
the associated
system engineer and reviewed condition
reports to assess the historical
and current condition
of the SFo buses. The team reviewed system health reports to verify that any aging effects are being adequately
managed. The team conducted
a walkdown of the SF6 buses to evaluate the exterior condition
of the buses and the operating environment.
B.2.2.2 Boral Monitorinq
The Boral Monitoring
Program is an existing program used to monitor the condition
of the material used in spent fuel pools for reactivity
control. Boral is the brand name for a sheet of uniformly
distributed
boron carbide in an alloy 1 100 aluminum matrix with a thin aluminum clad on both sides. The predecessor
to Boral is Boraflex, a similar material susceptible
to radiolytic
degradation.
Boraflex is used in the first six sets of racks at Seabrook.
The Boraflex utilized in the initial six racks is not credited in the criticality
analyses and is not credited for license renewal.Enclosure
15 The aging affect requiring
management
is a reduction
in neutron absorbing
capacity, a change in dimensions, and a loss of material due to the affects of the spent fuel pool environment.
Boral exposed to treated borated water is the subject of Draft LR-ISG-2009-01, "Staff Guidance Regarding
Plant Specific Aging Management
Revieft and Aging Management
Program for Neutron-Absorbing
Material in Spent Fuel Pools" The team reviewed the program documents, reviewed various corrective
actions, and interviewed
the responsible
engineers.
B.2.2.3 Nickel-Allov
Nozzles and Penetrations
The Nickel-Alloy
Nozzles and Penetrations
Program is an existing program that manages cracking, due to primary water stress corrosion, of the nickel based alloy pressure boundary and structural
components
exposed to the reactor coolant. This includes Pressurizer
Nozzles, Steam Generator
Channel Head Drain Tube and Welds, Reactor Vessel Core Support Pan/Lug, and Clevis Inserts, Reactor Vessel Hot and Cold Leg Nozzles, and the Reactor Vessel Bottom Mounted lnstrumentation
The program has been in existence, in various forms, since 2004 when Seabrook responded to NRC Bulletin 2004-01 "lnspection
of Alloy 8211821600
Materials
Used in the Fabrication
of Pressurizer
and Steam Space Piping Connections
at Pressurized
Water Reactors".
The management
of this aging affect has been refined since the phenomena
was first described
and has culminated
in the Electric Power Research lnstitute
sponsored
program MRP-139 "Material
Reliability
Program: Primary System Piping Butt Weld lnspection
and Evaluation
Guideline".
Seabrook's
draft "Reactor Coolant System Materials
Degradation
Management
Program" is structured
around the primary goal of mitigating
material degradation
of the reactor coolant system pressure boundary and reactor vessel internals.
The program is intended to manage the "Steam Generator
Program", Thermal Fatigue Management
Program","Alloy 600 Program", "Boric Acid Program", "Reactor Vessel lnternals
Program", and the"ASME Section Xl Program (NDE, lSl, Repair/Replacement)".
The management
program includes an appendix titled "Westinghouse
Proprietary
Information", which identifies
potential
Alloy 600/821182locations
in the primary pressure boundary components
of the Westinghouse
designed Nuclear Steam Supply System.Svstem Review In distinction
to the above noted program review, a system review was chosen by the team as a different
approach to ensure comprehensive
coverage of aging effects. The Residual Heat Removal System was chosen since the most likely initiating
event, at Seabrook, is a station black out and a dominate system for station black out response is the Residual Heat Removal System. The approach is to walk down the system in the plant and question how aging effects are covered and verify that coverage based on a review of the application, program descriptions, and if available
implementing
procedures.
Materials
identified
for this system are Cast Austenitic
Stainless
Steel, Glass, Stainless Steel, and Steel in the external environments
of indoor air that may included borated and Enclosure
16 non-borated
water leakage and Closed Cycle Cooling Water. The internalenvironments
are various treated and untreated
water, lubricating
oil, and reactor coolant.This results in the possible or experienced
aging affects of cracking, (cyclic, stress corrosion, thermal, loaded, and fatigue) and corrosion (boric acid, crevice, galvanic, general, and pitting), loss of preload, and fouling.The applicant, in turn, proposes the following
programs: ASME Section Xl Subsections
lWB, lWC, and IWD Program Bolting Integrity
Program Boric Acid Program Closed-Cycle
Cooling Water System Program External Surfaces Monitoring
Program Lubricating
Oil Analysis Program One'Time Inspection
of ASME Code Class Small Bore Piping One-Time Inspection
Program Water Chemistry
Program The ASME Section Xl Subsections
lWB, lWC, and IWD program, the Boric Acid Program are reviewed at every outage under the NRC's Reactor Oversight
Program using inspection
procedure
1P71111.08P "lSl Inspection".
The Water Chemistry
Program is part of the same procedure
by way of the Steam Generator
inspection
portion. The Bolting Integrity
Program, One-Time Inspection
of Code Class Small Bore Piping, and One-Time lnspection
are covered elsewhere
in this report.Of interest was a note in the System Walk-down
Report, in 2008, recording
the presence of water intrusion
associated
with "several supports in the vault stairuvell" and the observation
the "conditions
are slowly becoming worse as calcium accumulates." WO 0844358 was initiated
to verify the bolting integrity.
The work order incorrectly
compared the testing of anchors submerged
in raw water in a manhole with the anchors supporting
the RHR piping inserted into a calcium carbonate
degraded wall and concluded, based on the submerged
bolting, that the bolting in the RHR anchors were acceptable (AR 01633206).
This comparison
did not take into account the additional
concern of a recently discovered
alkaline silica degradation
associated
with the calcium carbonate degraded wall and the issue of anchor bolting integrity
was not revisited
subsequent
to the discovery
of alkali silica degradation.
WO 0844358 was translated, during a database change, into Condition
Report 08-15902 and closed on the basis of the comparison (two different
material environmental
conditions)
even though the condition
report contained
a proposal to randomly sample the bolts and perform a calibrated
torque test. The implications
of the NRC BulletinT9-02
anchor bolt integrity
program were never considered
during the evolution.
lnitially, these erroneous
comparisons, and incomplete
analysis, indicate a weakness in the NextEra's
program for identifying
and tracking the recently discovered
aging effects at the site. The revised analysis resulted in satisfactory
conditions
and the learning needed in dealing with aging effects to support license renewal (AR 01633206).
Enclosure
b.17 The inspector
walked-down
the RHR system from the outlet of RHR Pump P-8A, at elevation
54"-4", to the inlet of RHR Heat Exchanger
E-gA, at elevation
-31"-0", pausing at each support to carefully
inspect the visual appearance
of the bare piping revealed by the gaps in insulation.
The inspector
did not identify any evidence of aging that was not already considered
by the applicant
and adequately
covered by an existing of proposed program.Observations
and Findinqs Alkali-Silica
Reaction Aqinq Effect at Seabrook Station To assess the material condition
of concrete structures
in the plant; and to acquire, verify, and validate the design basis of structural
design, the applicant
personnel
performed civil/structuralwalk-down
inspections.
The Residual Heat Removal Equipment
Vaults, A and B Electrical
Tunnels, Radiological
Controlled
Area Walkway, and Service Water pump house was included in the walk-down
inspection
and assessment.
The observations
and findings were documented
in the License Renewal Project issue tracking report number 15. The walk-down
inspections
discovered
the following
plant material conditions; (a) large amount of groundwater
infiltration, (b) large amount of calcium carbonate
deposits, (c) corroded steel supports, base plates and piping, (d) corroded anchor bolts, (e) pooling of water and (f) cracking and spalling of concrete.The inspection
further noted that the below grade, exterior walls in the Control Building B Electrical
Tunnel at elevation
(-) 20'- 00" have random cracking and for several years have been saturated
by groundwater
infiltration.
The severity of the cracking and groundwater
infiltration
varies from location to location.
The groundwater
infiltration
has produced large, tightly adherent deposits of calcium oxide/carbonate
at certain locations
on the walls and pooling of groundwater
on the floor slab sometimes
to a depth of 2-inches.
The groundwater
has also produced smaller, loose deposits of calcium salts at most other crack locations.
The observations
and findings from the walk-down
inspections
were reviewed by applicant's
Design Engineering
Organization
and it was determined
that the concrete walls in the B-Electrical
Tunnel exhibited
the most extensive
distressed
condition
as determined
by the applicant
and required further investigation.
Specifically, the below grade exterior walls in the Control Building B Electrical
Tunnel at elevation
(-) 20' - 00" were selected due to the presence of fine, random cracking and, because, for over 10 to 15 years had remained in saturated
condition
by groundwater
infiltration.
The severity of the cracking and groundwater
infiltration
varied from location to location.
The groundwater
infiltration
had produced large, tightly adherent deposits of calcium oxide at certain locations
on the walls and pooling of groundwater
on the floor slab sometimes
to a depth of 2-inches.
The groundwater
has also produced smaller, loose deposits of calcium oxide at most other crack locations.
To assess the integrity
of cracked concrete and prolonged
groundwater
saturation, the applicant
contracted
with vendors to perform Penetration
Resistance
Testing (also referred to as Windsor Probe Test), and also to obtain concrete core specimens
at designated
locations
in four below grade, exterior walls of the B Electrical
Tunnel. The concrete core Enclosure
18 specimens
were subjected
to compressive
testing by the vendor and selected sections of the core specimens
were provided to another vendor for Petrographic
examinations.
The results Penetration
Resistance
Tests (PRT) for the control building indicated
an average concrete compressive
strength of 5340 psi and the concrete core testing indicated
an average compressive
strength of 4790 psi. PRT performed
in 1979 indicated
an average concrete compressive
strength of 6750 psi and the concrete test cylinders
that were cast during the placement
of the walls in February 1979 indicated
an average 28-day compressive
strength of 6120 psi. At each of the six (6) locations, three (3) individual
replicate
Resistance
Tests as recommended
per ACI 228.1R, Tables 5.2 and 5.5 has been performed
for a total of eighteen (18) Penetration
Resistance
Tests. Each of the eighteen (18) PRTs required three (3) firmly embedded probes as recommended
in ASTM C 803-03, paragraph
8.1.2for a total of fifty-four
(54) probes. The PRTs shall be performed
per ASTM C 803-03 standard, utilizing
Windsor Probe Test System per foreign print no. 100561.At each of six (6) locations, core drilled and removed two (2), 4-inch nominaldiameter
concrete core specimens
as recommended
in ACI 228.1R, paragraph
4.3.2. A totalof twelve (12) concrete core specimens
will be obtained as recommended
in ACI 228.1R paragraph
4.3.2to develop an adequate strength relationship
between the PRTs and the in-situ compressive
strength of the concrete.
The concrete core specimens
has been obtained per the method specified
in ASTM C 42-04 and compression
tested in the ME&T laboratory
per ASTM C 39-09. The length of the concrete core specimens "as removed" were12 to 16-inches
maximum. This provided adequate specimen lengths for compression
testing and Petrographic
examinations.
All of the walls in the B Electrical
Tunnel included in this study were 2-foot in thicKness
per drawing 101345, thus the concrete core drilling did not penetrate
through the walls or contacted
the two layers of reinforcement
on the outer-face of the walls.A comparison
of the 2010 concrete compression
test results to the 1979 concrete compression
test results indicated
a 21.7 percent reduction
in the compressive
strength of the concrete.
The reduction
in compressive
strength is most likely due to alkali-silica
reaction in the concrete which was detected in Petrographic
examinations
of four of the concrete core samples removed from the CB walls. lt was reported that the four concrete core samples had moderate to severe Alkali-Silica
Reaction in the concrete.
Alkali-Silica
Reaction is a reaction that occurs over time in concrete between the alkaline cement paste and reactive non-crystalline
silica which is found in many common coarse aggregates.
The reaction produces a gel substance
which expands and causes micro-cracking or fissures in and surrounding
the coarse aggregates.
The micro-cracking
typically
progresses
and extends into the cement paste thus compromising
the quality and integrity
of the concrete.
The presence of water, irrespective
of water chemistry (i.e., aggressive
or non-aggressive), is required for Alkali-Silica
Reaction to develop and to continue to propagate
in the hardened concrete.
Without the presence of water, Alkali-Silica Reaction will not develop or continue to propagate
in hardened concrete.
Alkali-Silica Reaction often results in a reduction
in both strength and elasticity
of the concrete;both of which were noted in the sample concrete cores analyzed for Seabrook.Enclosure
19 The reduction
in compressive
strength raises questions
regarding
the effect on modulus of elasticity, and flexural and shear capacity of concrete structural
members. ln addition the modulus of elasticity
affects the dynamic response of Structures.
The applicant
is considering
the structure
dynamic response in their analyses.In accordance
with Inspection
Procedure
71002 and Inspection
Manual Chapter 2516, a key assumption
of license renewal is that the current licensing
bases is to be maintained.
The above discussion
indicated
that this may not be true if operability
of the safety related structures
cannot be maintained.
The NRC inspection
report 0500044312011002, issued May 12,2011, addresses
current licensing
bases issues along with an extent of condition review planned by the applicant.
With respect to the aging management
review for this aging effect at the station, the applicant
provided a summary of their plans in a response for additional
information
associated
with the Division of License Renewal review in a letter dated April 14, 2011 (letter SBK-L-11063).
Overall Findinos The team concluded
screening
and scoping of non-safety
related systems, structures, and components, was implemented
as required in 10 CFR 54.4(a)(2), and the aging management
portion of the license renewal activities
were conducted
as described
in the License Renewal Application.
The inspection
concluded
the documentation
supporting
the application
was in an auditable
and retrievable
form. Except for the alkali-silica
reaction issue, the inspection
results support a conclusion
of reasonable
assurance
with respect to managing the effects of aging in the systems, structures, and components
identified
in the application.
Enclosure
A-1 ATTACHMENT
SUPPLEMENTAL
INFORMATION
KEY POINTS OF CONTACT Applicant
Personnel E. Metcalf Plant Manager M. Collins Design Engineering
Manager M. O'Keefe Seabrook Station Licensing
Manager R. Cliche License Renewal Project Manager P. Tutinas License Renewal Project Electrical
Lead A. Kodal License Renewal Project Mechanical
Lead K. Chew License Renewal Project CivilStructural
Lead LIST OF DOCUMENTS
REVIEWED General License Renewal Documents NRC lnspection
Procedure
71002; License Renewal Inspection
NRC AMP Audit Report (results)SBK-L-10192, Seabrook Station, Response to RAls, Set ?, X,2Q10 SBK-L-10204, Seabrook Station, Response to RAls, Set ?, December 17 ,2Q10 SBK-L-11002, Seabrook Station, Response to RAls, Set 4, January 13,2011 SBK-L-11003, Seabrook Station, Response to RAls, Set 5, January 13,2011 SBK-L-11015, Seabrook Station, Response to RAls, Set ?, X,2011 SBK-L-1 1027, Seabrook Station, Response to RAls, Set 9, X,2011 Updated Final Safety Analysis Report, Section 3.7(8).3.13
License Renewal Basis Documents LRAM-ELEC, Aging Management
Review Report: Electrical
Components
and Commodities, Rev 1 LRAP-EI, Aging Management
Program Basis Document:
Electrical
Cables and Connections
Not Subject to 10 CFR 50.49 Environmental
Qualification
Requirements, Rev 2 and Rev 3 LRAP-E3, Aging Management
Program Basis Document:
Inaccessible
Power Cables Not Subject to 10 CFR 50.49 Environmental
Qualification
Requirements
Program, Rev 2 LRAP-E3, Aging Management
Program Basis Document:
Metal Enclosed Bus, Rev 1 LRAP-M027, Aging Management
Program Basis Document:
Fire Water System, Rev 1 LRAP-M032, Aging Management
Program Basis Document:
One-Time lnspection, Revision 1 LRAP-M033, Aging Management
Program Basis Document:
Selective
Leaching of Materials, Revision 1 LRAP-M033, Aging Management
Program Basis Document:
Selective
Leaching of Materials, Revision 2 (Draft)I Attachment
A-2 LRAP-M038, Aging Management
Program Basis Document:
lnspection
of lnternalSurfaces
in Miscellaneous
Piping and Ducting Components, Revision 1 LRAP-M039, Aging Management
Program Basis Document:
Lubricating
OilAnalysis, Revision 1 LRAP-SF6, Aging Management
Program Basis Document:
345kV SF6 Bus, Rev 1 LRSP-ELEC, Scoping and Screening
Report: Electrical
Systems, Components, and Commodities, Rev 2 LRTR-NSAS, Technical
Report - Non-Safety
Affecting
Safety, Revision 3 LRTR-NSAS, Technical
Report - Non-Safety
Affecting
Safety, Revision 4 lmplementino
Procedures
CP 3.3, Closed Cooling Water Systems, Chemistry
Control Program, Rev 20 ER-AA-106, Cable Condition
Monitoring
Program, Rev 1 ES1807.020, Machinery
OilAnalysis, Revision 0 FP 3.1, Fire Protection
Maintenance
and Surveillance
Testing, Rev 3 LN0560.10, SFO Dewpoint Check, Rev 2 1N0560.11, SFO SO2 and Purity Sample, Rev 7 ON0443.54, Non-safety
Related Deluge and Sprinkler
Systems 18 Month lnspection, Rev 4, Change 8 AN1242.01, Loss of lnstrumentAir, Revision 12 030443.66, Safety Related Spray and Sprinkler
Systems 18 Month Flow and System Alarms Test, Rev 4, Change 9 OX0443.04, Fire Protection
System Annual Flush, Rev 6 Change I OX0443.12, Fire Protection
Dry Pipe Spray and Sprinkler
Systems 18 Month Inspection, Rev 6, Change 4 OX0443.19, Yard Hydrant Hose House Monthly Inspection, Rev 6 Change 4 OX0443.20, Yard Hydrant Semi-Annual
lnspection
and Functional
Test, Rev 6, Change 6 OX0443.21, Yard Fire Hydrant Hose Houses Annual Hose Replacement
and Gasket lnspection, Rev 6, Change 2 PEG'10, System Walkdowns, Rev 18 PEG-265, Cable Condition
Monitoring, Rev 0 SSCP, Chemistry
Manual, Rev 64 Draft lmplementinq
Procedures
LRTR-INT, Technical
Report - lnspection
of Internal Surfaces, Revision 0 (Draft)LRTR-OTI, Technical
Report - One-Time lnspection, Revision 0 (Draft)LRTR-SEL, Technical
Report - Selective
Leaching of Materials, Revision 0 (Draft)Technical
Reports EE-07-018, Response to GL 2001-01, Rev 0 Engineering
Evaluationg4-41, Submerged
Electrical
Cables and Supports, Dated 1l39l95 Technical
Report "Buried Piping and Tanks lnspection
Program" LRTR-BP Revision 0 Attachment
A-3 Work Orders 0080886 01 81964 0187223 0234295 0242456 0301 31 1 031 0880 0317696 0401697 0401699 0401728 0406534 0414066 0417588 0431657 0443640 0444321 0519953 0526073 0603042 4702705 0716257 0716258 0718994 0719543 0720390 0727117 0727135 0727136 0727137 0727138 081 3420 0827061 0827184 0827185 0831312 0831 31 3 0831583 0835656 98C3889 99A5575 I Attachment
A-4 Work Order Package 00611225 01, "Reference
Maintenance - Auxilliary
Boiler Tank Manway Leakage" Work Order Package 00616970 01, "The Outside of FP-TK-36A
Has Peeling Paint and Rust TK" Work Order Package 00616971 01, "The Outside of FP-TK-368
Has Peeling Paint and Rust TK" Work Order Package 00791046 01, "Diesel Fire Pump Fuel Oil Tank Water Removal" Work Order Package 00791057 01, "Diesel Fire Pump Fuel Oil Tank Water Removal" Action Request 00207755 "Seabrook
Station License Renewal lmplementation
Actions" Completed
Surveillance
Tests 12 oil sample analysis results from Herguth Labs Reference
Documents Materials
Reliability
Program: Primary System Piping Butt Weld Inspection
and Evaluation
Guidelines (MRP-139)
1010087, August 2005 NEI 96-03, Guideline
for Monitoring
the Condition
of Structures
at Nuclear Power Plants, 1996 ACI 201.1R-92, Guide for Making a Condition
Survey of Concrete in Service, American Concrete Institute ACI 349.3R-96, Evaluation
of Existing Nuclear Safety- Related Concrete Structures, American Concrete lnstitute
ACI 531-79, Concrete Masonry Structures, Design and Construction, American Concrete lnstitute Hope Creek Update Final Safety Analysis Report, Section 7.2.1.36 Materials
Reliability
Program: Primary System Piping Butt Weld Inspection
and Evaluation
Guidelines (MRP-139)
1010087, August 2005 NEI 09-14, Revision 0; Guidelines
For The Management
Of Buried Piping lntegrity, 01110 EPRI Final Report 1016456, 121Q8; Recommendations
for an Effective
Program to Controlthe
Degradation
of Buried Piping Drawinos Complete set of submitted
license renewal drawings 1-AS-2301-2, Auxiliary
Steam Piping, Revision 4 1-AS-5198-02, Auxiliary
Steam Piping, Revision 3 1-DM-D20355, Demineralized
Water Distribution
Detail, Revision 17 9763-F-310248, Underground
Duct Plan, Rev 13 9763-F-802807-641.20C, Piping - Combustible
Gas lsometric, Revision 0 9763-F-802807S, Sheets 15, 155, 16; Pipe Support Details, Revision 68 9763-F-202753-610.60, Service Air lsometric, Revision 0 9763-M-202751S, Sheets 43, 43S, 74,745,74A;
Support Details, Revision 25A Attachment
A-5 9763-M-212368S, Sheets 15, 155, 16; Support Details, Revision 11B 9763-M-212368S, Sheets 17, 175,18, 18A; Support Details, Revision 23A 9763-M-2123685, Sheets 19, 195; Support Details, Revision 208 9763-M-2123685, Sheets 36, 365, 37; Support Details, Revision 128 9763-M-2123685, Sheets 53, 53S, 54 - 57; Support Details, Revision 24A 9763-M-8029133, Sheets 49, 49S, 50, 51, 52; Support Details, Revision 11B 1-NHY-310002, Unit Electrical
Distribution
One Line Diagram, Rev 40 1-NHY-505084, Instrument
Air Installation - DualAir Supply, Revision 6 PID-1-WLD-820224, Waste Processing
Liquid Drains - RCA Walkway Details, Revision 7 License Renewal PID Drawing PID-1-RH-1R20663
License Renewal PID Drawing PID-1-SI-LR20446
License Renewal PID Drawing PID-1-Sl-LR20447
License Renewal PID Drawing PID-1-Sl-LR20448
License Renewal PID Drawing PID-1-Sl-LR20449
License Renewal PID Drawing PID-1-Sl-1R20450
License Renewal Pl D Drawing PID-1 -WLD-LR20221
License Renewal Pl D Drawing Pl D- 1 -VSL-LR2O77
6 License Renewal PID Drawing PID-1-CBS-1R20233
License Renewal PID Drawing PID-1-CS-LR20722
License Renewal PID Drawing PID-1-CS-LR20725
License Renewal PID Drawing PID-1-RC-LR20841
License Renewal PID Drawing PID-1-RC-LR20844
License Renewal PID Drawing PID-1-RH-1R20662
Corrective
Action Documents 198495 95-33705 98-00804 98-01661 99-12562 00-05286 01-04204 01-04373 01-07417 01-08751 01-08770 01-02389 01-13429 02-01 989 02-02211 02-03132 02-05112 02-05698 02-08670 02-08671 02-13425 02-15177 02-17027 03-03536 03-07418 04-1 1389 04-12631 05-04768 05-05078 05-07548 05-07730 05-09832 05-1 3056 05-15093 05-041 1 5 06-08855 06-11121 07-03741 07-05144 07-09377 07-12356 07-14158 07-1 5599 07-14047 Attachment
A-6 08-05795 08-06033 08-06080 08-06088 08-1 31 73 08-01461 08-01468 08-13706 08-15277 09-01489 09-01 520 09-207352 00-216968 00-590824 01-63276 Apparent Cause Evaluation
for B EDG rocker arm lube oil tank fuel dilution Apparent Cause Evaluation
for supply jug oil contamination
with water Apparent Cause Evaluation
for aluminum bronze fittings in sea water piping systems Miscellaneous
09CAR029, Change Authorization
Request: De-Watering
System for Safety Related Cable Vaults, Dated 6/25109 Keyword searches of CRs for Karl Fischer, water contamination, cast iron, graphitization, dezincification, de-alloy, and leaching Fire Protection
System Walk Down Report Plant Engineering
Guidelines
System Walkdowns
PEG-10 Revision 19 Roving NSO Log Operations
Routine Tours, 0210912011
Buried Piping Program ER-AA-102 Buried Piping Program ER-AA-1 02-1000 Mechanical
Maintenance
Procedure "Application
of Repair and Protective
Coating(s)" MS0517.12
Rev. 04, Chg. 03 Svstem Health Reports System Health Reports, Switchyard
System, Dated 111109 through 12131110 Cable Program Health Report, Dated 1011log through 12131110 Predictive
Maintenance
Program Health Report, Quarter 4,2007 to Quarter 3, 2008 Predictive
Maintenance
Program Health Report, Quarter 4,2OOg to Quarter 2,2010 Buried Piping Program Health Report - 4n Quarter 2008 through 3'o Quarter 2010 Cathodic Protection
System Health Report 1't Quarter 2004 through 3'o Quarter 2010 Above Ground Steel Tanks Program Health Report 1010112008 - 12/3112008
Above Ground Steel Tanks Program Health Report 0110112009 - 03/3112009
Above Ground SteelTanks
Program Health Report 0410112009 - 06/30/2009
Above Ground Steel Tanks Program Health Report 0710112009 - 09/30/2009
Above Ground Steel Tanks Program Health Report 10/01/2009 - 1213112009
Above Ground Steel Tanks Program Health Report 0110112010 - 0313112010
Above Ground SteelTanks
Program Health Report 0410112010 - 06/30/2010
Attachment
A-7 Above Ground SteelTanks
Program Health Report 0710112010 - 09/30/2010
Above Ground Steel Tanks Program Health Report 10lO1l201A - 1213112010
RHR System Health Report 1UA112010 - 1213112010
RHR System Health Report 2010-04 RHR System Walk-Down
Report 0210812011
RHR System Walk-Down
Report 0410112010
RHR System Walk-Down
Report 06/30/2010
Attachment