ML111360432

IR 05000443-11-007; 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

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 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

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 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

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).l 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 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

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

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

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)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