ML111360432
| ML111360432 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 05/23/2011 |
| From: | Conte R 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
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: I nspection Report 05000443/201 1 007
cc Mencl: Distribution via ListServ
Distribution Mencl: (VlA E-MAIL) A. Williams, Rl OEDO
W. Dean, RA ROPreports@nrc.gov
P. Wilson, DRS Region I Docket Room (with concurrences)
A. Burrit, DRP
C. LaRegina, DRP
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 nsp Rpt Final. docxI
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
U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No: 50-443
License No: NPF-86
Report No: 05000443/2011007
Licensee: NextEra Energy Seabrook LLC
Facility: Seabrook Station
Location: 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.
1
REPORT DETAILS
40.A2 Other - License Renewal
a. 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
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
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 Feedwater 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
Enclosure
l
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 penetration 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).
Enclosure
6
B.2.1.17 Aboveqround Steel Tanks
The Aboveground SteelTanks aging management 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 coatings. 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 by
cathodic protection. During the first quarter of 2009 the cathodic protection system was
finally categorized as green (or satisfactory condition). The cathodic protection 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 cathodic protection 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 IWE aging management 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
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 aging management 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 Penetrations. 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 Penetrations 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 aging management 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
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.
b. 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 Penetration 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)
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
Attachment
I
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 02-17027
95-33705 03-03536
98-00804 03-07418
98-01661 04-1 1389
99-12562 04-12631
00-05286 05-04768
01-04204 05-05078
01-04373 05-07548
01-07417 05-07730
01-08751 05-09832
01-08770 05-1 3056
01-02389 05-15093
01-13429 05-041 1 5
02-01 989 06-08855
02-02211 06-11121
02-03132 07-03741
02-05112 07-05144
02-05698 07-09377
02-08670 07-12356
02-08671 07-14158
02-13425 07-1 5599
02-15177 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