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: 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

D. Lew, DRA D. Bearde, DRS

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

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

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

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0406534

0414066

0417588

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0443640

0444321

0519953

0526073

0603042

4702705

0716257

0716258

0718994

0719543

0720390

0727117

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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