SBK-L-11003, Response to Request for Additional Information NextEra Energy Seabrook License Renewal Application Aging Management Programs - Set 5

Response to Request for Additional Information NextEra Energy Seabrook License Renewal Application Aging Management Programs - Set 5

ML110140587
Person / Time
Site:	Seabrook
Issue date:	01/13/2011
From:	Freeman P NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-11003
Download: ML110140587 (28)

Text

NEXTera ENERG.Y.C

.,,,SEBROK January 13, 2011 SBK-L- 11003 Docket No. 50-443 U.S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852 Seabrook Station Response to Request for Additional Information NextEra Energy Seabrook License Renewal Application Aging Management Programs - Set 5

References:

1. NextEra Energy Seabrook, LLC letter SBK-L-10077, "Seabrook Station Application for Renewed Operating License," May 25, 2010. (Accession Number ML101590099)

2. NextEra Energy Seabrook, LLC letter SBK-L- 10179, "Supplement to the NextEra Energy Seabrook, LLC, Seabrook Station License Renewal Application", October 29, 2010.

(Accession Number ML10306002)

3. NRC Letter "Request for Additional Information Related to the Review of the Seabrook Station License Renewal Application (TAC NO. ME4028) - Aging Management Programs" December 14, 2010 (Accession Number ML103420273),

In Reference 1, NextEra Energy Seabrook, LLC (NextEra) submitted an application for a renewed facility operating license for Seabrook Station Unit 1 in accordance with the Code of Federal Regulations, Title 10, Parts 50, 51, and 54.

In Reference 2, NextEra submitted changes to the Seabrook Station Unit No. 1 License Renewal Application (LRA) in regards to the Buried Piping and Tanks Inspection Program.

In Reference 3, the NRC requested additional information in order to complete its review of the License Renewal Application (LRA). Enclosure 1 contains NextEra's response to the request for additional information and associated changes made to the LRA. For clarity, deleted LRA text is highlighted by strikethroughs and inserted texts highlighted by bold italics.

NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874

United States Nuclear Regulatory Commission SBK-L-1 1003 / Page2 Commitment number 59 is added to the License Renewal Commitment List. There are no other new or revised regulatory commitments contained in this letter. Enclosure 2 provides a revised LRA Appendix A - Final Safety Report Supplement-Table A.3, License Renewal Commitment List, updated to reflect the license renewal commitment changes made in NextEra Energy Seabrook correspondence to date. In addition to new commitment 59 contained in this letter, commitment numbers 53, 54, 55, 56, 57 and 58 that were added in NextEra Energy Seabrook, LLC letter SBK-L- 11002, "Response to Request for Additional Information, Aging Management Programs - Set 4" are included.

If there are any questions or additional information is needed, please contact Mr. Richard R.Cliche, License Renewal Project Manager, at (603) 773-7003.

If you have any questions regarding this correspondence, please contact Mr. Michael O'Keefe, Licensing Manager, at (603) 773-7745.

Sincerely, NextEra Energy Seabrook, LLC.

Paul 0. Freeman Site Vice President

Enclosures:

- Response to Request for Additional Information Seabrook Station License Renewal Application Aging Management Programs and Associated LRA Changes - LRA Appendix A - Final Safety Report Supplement Table A.3, License Renewal Commitment List, updated to reflect the license renewal commitment changes made in NextEra Seabrook correspondence to date.

United States Nuclear Regulatory Commission SBK-L- 11003 Page 3 cc:

W.M. Dean, NRC Region I Administrator G. E. Miller, NRC Project Manager, Project Directorate 1-2 W. J. Raymond, NRC Resident Inspector R. A. Plasse Jr., NRC Project Manager, License Renewal M. Wentzel, NRC Project Manager, License Renewal Mr. Christopher M. Pope Director Homeland Security and Emergency Management New Hampshire Department of Safety Division of Homeland Security and Emergency Management Bureau of Emergency Management 33 Hazen Drive Concord, NH 03305 John Giarrusso, Jr., Nuclear Preparedness Manager The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399

United States Nuclear Regulatory Commission SBK-L-11003 / Page 4 NE)xera I, Paul 0. Freeman, Site Vice President of NextEra Energy Seabrook, LLC hereby affirm that the information and statements contained within are based on facts and circumstances which are true and accurate to the best of my knowledge and belief.

Sworn and Subscribed Before me this i3 dayof " .. ,2011 Paul 0. Freeman Site Vice President Notary Pubic

United States Nuclear Regulatory Commission Page 1 of 14 SBK-L- 11003 /Enclosure 1 Enclosure 1 to SBK-L-11003 Response to Request for Additional Information Seabrook Station License Renewal Application Aging Management Programs

,and Associated LRA Changes

United States Nuclear Regulatory Commission Page 2 of 14 SBK-L-11003 / Enclosure 1 Request for Additional Information (RAI) B.2.1.20-1

Background

Generic Aging Lessons Learned (GALL) aging management program (AMP)XI.M32, "One-Time Inspection," states in element 4, "detection of aging effects," that the inspection includes a representative sample of the system population, and, where practical, focuses on the bounding or lead components most susceptible to aging due to time in service, severity of operating conditions, and lowest design margin.

License renewal application (LRA) Section B.2.1.20, One-Time Inspection, states that the inspection sample -includes locations where the most severe aging effect(s) would be expected to occur. The inspection population will be based on such aspects of the systems and components as similarity of materials of construction, operating environment, and aging effects. The sample size will be based on such aspects of the systems and components as the specific aging effect, location, system, and structure design, materials of construction, service environment, or previous failure history. The selection criteria will include stagnant or low-flow areas.

Issue Due to the uncertainty in determining the most susceptible locations and the potential for aging to occur in other locations, the staff noted that large sample sizes (at least 20%) may be required in order to adequately confirm an aging effect is not occurring. The applicant's One-Time Inspection Program did not include specific information regarding how the selected set of components to be sampled or the sample size will be determined.

Request Provide specific information regarding how the selected set of components to be sampled will be determined and the size of the sample of components that will be. inspected.

NextEra Enermy Seabrook Response:

The sample population for the One-Time Inspection will be approximately twenty percent of the components having the same material, environment and aging effect combinations, not to exceed twenty five components.

Based on the above discussion, the following change is being made to the Seabrook Station License Renewal Application:

United States Nuclear Regulatory Commission Page 3 of 14 SBK-L- 11003 / Enclosure 1

1. In Section B.2.1.20, on page B-1 18, change the second paragraph from the bottom as follows:

This program will perform a one-time inspection of selected components determined to be most susceptible* to the potential degradation mechanisms. The components to be inspected will be chosen from the systems within the scope of the Seabrook Station Water Chemistry Program, Seabrook Station Fuel Oil Chemistry Program, and the Seabrook Station Lubricating Oil Analysis Program. From these groups of components, a sample of the population will be selected for inspection as part of the Seabrook Station One-Time Inspection Program. The inspection sample population will be approximately twenty percent of the components based on such aspects of the systems and components as similarity of materials of construction, operating environment, and aging effects, not to exceed twentyfive components. The sample size will be based on such aspects of the systems and components as the specific aging effect, location, system, and structure design, materials of construction, service environment, or previous failure history. The selection criteria will include stagnant or low-flow areas.

Request for Additional Information (RAI) B.2.1.22-1

Background

LRA Section B.2.1.22 states that the cathodic protection system protects the service water, diesel generator cooling water and instrument air piping systems as well as portions of the fire protection and control building air handling system. The staff noted that the auxiliary boiler, auxiliary steam condensate, auxiliary steam heating, condensate, feed water, and plant floor drains as well as portions of the control building air handling and fire protection systems are not provided with cathodic protection. The applicant stated that opportunistic and/or directed visual inspections will be performed in areas with the highest likelihood of corrosion problems or areas with a history of corrosion problems.

Issue For the cathodically protected portions of in-scope buried piping, given that coatings can be missing, degraded or nonconforming (e.g., holidays),.the staff believes that a cathodic protection system is most effective when it is available at least 90% of the time or not frequently removed from service. The LRA does not contain details on the availability of the cathodic protection system.

Given that several in-scope buried piping systems do not have cathodic protection, selection of inspection quantities and locations is particularly important to ensure that the most susceptible locations are being inspected. The applicant did not provide details on how it will determine localized data (e.g.,

soil pH, composition of the soil, water table, chemical runoff probability, soil resistivity, and potential for stray currents) or localized corrosion rates in order to inform its inspection quantities and locations.

As a result of not having this information, the staff cannot make a determination'that the number and locations of planned buried piping inspections is sufficient to maintain the pipe wall thickness at or above design minimum values throughout the period of extended operation.

United States Nuclear Regulatory Commission Page 4 of 14 SBK-L- 11003 / Enclosure 1 Request Note: Although gray cast iron is included within the scope of the GALL Report Section IX definition of steel, the below request does not apply to piping segments constructed of this material in the fire protection system.

For the in-scope buried piping systems that are protected by a cathodic protection system, state the availability of the cathodic protection system, and if portions of the system are not available 90% of the time or will be allowed to be out of service for greater than 90 days in any given year, justify how the piping will meet or exceed the minimum design wall thickness throughout the period of extended operation.

For the in-scope buried systems and portions of systems constructed of steel that are not cathodically protected:

a. State the lengths of the in-scope buried portions of piping for each system that is not provided with or only portions provided with cathodic protection.

b. Provide details on plant-specific data of localized soil conditions (e.g., pH, composition of the soil, water, table, chemical runoff probability, soil resistivity, potential for stray currents), plant-specific operating experience, and localized corrosion rates that will be utilized to optimize inspection quantities and if this data does not exist, state what samples will be taken and how they will be utilized in selecting inspection locations.

c. Justify the basis of the inspection population size (i.e., linear feet of buried piping) in relation to standard industrial sampling methods so as to maintain pipe wall thickness at or above design minimum values throughout the period of extended operation.

NextEra Energv Seabrook Response:

1. The 3rd quarter 2010 Buried Piping Integrity Program Health Report indicates that 98 % of the Cathodic Protection system was within the NACE recommendations for pipe to soil potentials. The availability of the cathodic protection system segments is monitored routinely. Should the availability of the respective cathodic protection system be found to be less than 90%, or to deviate from the criteria specified by NACE for a period of more than 90 days, the steel piping in that portion of the system will be treated as having no cathodic protection and the number of inspections per interval increased accordingly. The number of inspections per interval is specified in the Seabrook Station Buried Piping and Tanks Inspection Program (Reference LRA Supplement dated October 29, 2010, Enclosure 1, Page 12 of 18).

2a. The table below provides the approximate length of in-scope buried piping, by system, that is not provided with cathodic protection or only portions of which are provided with cathodic protection.

This table does not include the Auxiliary Boiler fuel supply piping as that piping currently exists as a temporary modification, with no buried piping, pending final design change.

United States Nuclear Regulatory Commission Page 5 of 14 SBK-L- 11003 / Enclosure 1 Safety Cathodically Total Lerngh Reatedý (!:i,:

feet)': i Protected Yes 90 CBA Steel Yes No 970 Yes 3986 FP Steel No No 1996 CO Steel No No 152 DG Steel No No 50 FW Steel Yes No 147 DF Steel No No 250 2b. Plant-specific data of localized soil conditions, plant-specific operating experience, and localized corrosion rates that will be utilized to optimize inspection quantities and locations is not available.

Prior to the initial set of buried pipe inspections performed under this program, plant-specific data of soil conditions will be collected and used as one input to select the most susceptible locations for inspection. The FPL Nuclear Fleet Buried Piping Examination Procedure requires the following samples to be taken in the area of planned direct inspections of buried pipe.

1. Pipe-to-soil potentials

2. Soil resistivity

3. Soil Samples

4. Water samples

5. Measurement of under-film liquid pH

6. MIC samples if applicable 2c. The sample size for in-scope steel buried pipe inspected under this program is increased from the number of inspections, specified in the Seabrook Station Buried Piping and Tanks Inspection Program by a factor of 4 for systems or portions of systems that are not cathodically protected.

Without cathodic protection, the role of adequate coating becomes more important. Therefore, should evidence of coating damage attributable to backfill materials be identified, the sample size for in-scope steel buried pipe is increased by a factor of 8 times that for cathodically protected piping (Reference LRA Supplement dated October 29, 2010, Enclosure 1, Page 12 of 18). Guidance from NUREG-1801, Revision 2, was utilized in' establishing the scope and frequency of the buried piping inspections. The number of inspections to be performed, based on pipe material, cathodic protection and backfill criteria of NUREG-1801 are incorporated into the tables in the Seabrook Station Buried Piping and Tanks Inspection Program.

United States Nuclear Regulatory Commission Page 6 of 14 SBK-L- i 1003 /Enclosure 1 Request for Additional Information (RAI) B.2.1.22-2

Background

In footnote 5 to the Buried Pipe Inspection Locations chart in the "detection of aging effects" program element, the applicant stated that if during inspections of a particular material type, damage to coatings or base materials is determined to have been caused by backfill, the backfill will be considered to be "inadequate" for that material type only. The staff noted that the number of proscribed inspections increases if backfill is determined to be inadequate. In the "preventive action" program element, the applicant did not state that backfill, requirements were dependent on the material type of the buried pipe.

Issue Given that backfill specifications are not dependent on the material type of the buried pipe, the staff believes that the results of inspections of backfill quality for a given piping material (e.g., steel) should be applied to the inspection sample size of the other material types (e.g., stainless steel, polymeric).

Request Justify why an inadequate backfill determination for a single material type inspection should not be applied to the other material types.

NextEra Enerev Seabrook ResDonse:

After further evaluation, backfill materials and backfill procedures would. not be expected to be different when installing pipe of different materials. The License Renewal Application, Appendix B is revised to delete, in NOTE 4 of the Buried Piping Inspection Locations table, the words "for that material type only."

Based on the above discussion, the following-change is being made to the Seabrook Station License Renewal Application:

1. In Section B.2.1.22, as submitted in License Renewal Application Supplement 1, dated October 29, 2010 (SBK-L- 10179), on Enclosure 1 page 12, Note 4 of the Buried Piping Inspection Locations table is revised as follows:

United States Nuclear Regulatory Commission Page 7 of 14 SBK-L- 11003 / Enclosure 1

4. The effectiveness of backfill materials and processes will be determined by the condition of coatings and base materials noted during inspections. If damage to the coatings or base materials are determined to have been caused by the backfill, the backfill will be considered to be "inadequate" (for the purpose of this program) for that material only.

Reguest for Additional Information (RAI) B.2.1.22-3

Background

In LRA Section B.2.1.22 the applicant stated that the service water system contains inaccessible submerged steel piping exposed to raw water in two vaults, one with four fifteen foot lines and one with a piping segment less than ten feet long. The applicant also stated that the piping is cathodically protected and coated. During the AMP audit the applicant stated that these vaults are normally filled with raw water but are periodically drained to conduct inspections. The applicant further stated that the vault containing this piping gradually re-fills with raw water as a result of in-leakage of groundwater.

Issue The staff does not have sufficient information to determine how visual inspections of the external surfaces of the piping (i.e., inspection of the coatings) will detect corrosion of the piping that can occur either due to the permeability of the coating or undetected holidays in the coating. Coatings, in this case, could mask an on-going corrosion issue.

Request State how the corrosion of the piping that could occur either due to permeability of the coating or undetected holidays in the coating will be detected.

NextEra Energv Seabrook Response:

The piping in the Service Water vault, is protected by coal-tar enamel and felt wrap coating in accordance with station procedures. The specific requirements of the station piping specification require an initial, manufacturer applied coat of epoxy primer on steel piping. This specification also specifies that the pipe be commercially coated and wrapped with 1) a coat of cold-applied coal-tar primer, 2) a coat of hot-applied coal-tar enamel into which is bonded asbestos felt or fibrous glass mat, and 3) a final wrap of draft paper or coat of whitewash. The coal-tar primer, coal-tar enamel, asbestos felt or fibrous glass mat, and kraft paper/whitewash are required to meet the requirements of AWWA Specification C203. Following coating, all coated surfaces are tested with an electric holiday tester and

United States Nuclear Regulatory Commission Page 8 of 14 SBK-L- 11003 / Enclosure 1 all holidays, faults, or missing places are repaired.

To facilitate access to the piping, removable spools-in the vaults were fabricated from the existing pipe.

The new welds and areas adjacent to the pipe cuts and the new flanges were coated using a Keeler and Long 1000 Kolormastic system and a Tapecoat 20 primer and wrap io complete the transition from the existing coal-tar coating system. As'stated in the respective design change document, the "service water vault is not designed to be a water tight structure and will be flooded with ground water during normal plant operation. Entry into the vault will require initial pump down and may require continuous portable sump pump operation. The painting system chosen for service water piping within the vault is designed to protect the pipe from long term external corrosion based on continuous immersion in brackish, stagnant water."

The following are excerpts from a technical paper from Plastics Division, Allied Chemical Corporation, Morristown, N. J., titled "Hot Applied Coal Tar Coating."

"High temperature coal-tar pitch is practically inert to the action of water and neither absorbs or transmits it. High temperature coal-tar pitch is highly resistant to attack by bacteria and fungi. This property, together with its moisture resistance, make it eminently suitable for roofing; waterproofing; coating of buried steel pipe lines to protect them from corrosion action of wet soil; lining of water pipes, tanks, etc."

"Water absorption of coal-tar enamels is extremely low. NACE Committee T-6A on Thermoplastic Coal Tar Base Linings reports that after 6 years immersion, coal-tar enamels, at approximately 100 mils thickness, show an absorption of only 1.7 to 2.3 gms per square foot or 0.5 to 0.6% by weight."

The paper provides anecdotal evidence of the effectiveness of coal-tar as a coating material.

"Coal tar pitch shows extremely low moisture absorption, is highly resistant to bacterial deterioration, and highly resistant to soil chemicals. As a result, the coating remains practically unchanged through years of service. The pipe and the bond remains firm and strong throughout long years of burial. Coal tar coated pipelines have been dug up after being in service for 20-30 years and more and we find the coating unchanged and the bond strong."

Based on the information provided in the design change document, excerpts from industry technical papers, and over 15 years of operational experience with this specific piping exposed to a raw-water environment with no failures, the suitability of the applied coating system is considered to be adequate to prevent corrosion of the underlying pipe. The periodic visual inspections of this pipe described in the Seabrook Station Buried Piping and Tanks Inspection Program are capable of detecting damage or degradation of the coating, thereby effectively age managing this piping. Any degradation of the coating is reported in the Seabrook Corrective Action Program and repairs will be made as appropriate.

United States Nuclear Regulatory Commission Page 9 of 14 SBK-L- 11003 / Enclosure 1 Reguest for Additional Information (RAI) B.2.1.22-4

Background

In LRA Section B.2.1.22, the applicant stated that in November 2000, buried in-scope fuel supply piping leaked as a result of damaged wrapping. The applicant also stated that during subsequent extent of condition inspections of the same piping system, further damage was discovered, ultimately leading to a decision to not return this buried fuel supply line to service.

In footnote 5 to the Buried Pipe Inspection Locations chart in the "detection of aging effects" program element, the applicant stated, "This, line is not in use and has been drained and flushed and is awaiting replacement per a design change. The inspection criteria for the replacement piping will be determined based [on] material selection, coating, cathodic protection and quality of backfill.". During the audit the applicant stated that temporary fuel oil piping and tanks have been installed to replace the effected buried fuel oil supply piping and the associated aboveground fuel oil storage tank.

Issue Given that the leakage from the buried fuel oil supply piping was the result of corrosion due to degraded wrapping, what extent of condition reviews were conducted to evaluate the condition of the wrapping for other in-scope systems containing non-cathodically protected, buried piping.

The staff noted that temporary piping and tanks are currently installed to support the in-scope function once served by the buried fuel oil supply piping and the associated above-ground fuel oil storage tank.

The staff also noted that the LRA does not discuss if/when this temporary arrangement will be replaced with a permanent arrangement or how either the temporary piping and tanks or permanent arrangement will be age managed through the period of extended operation.

Request

1. Provide details, (if any) regarding what extent of condition reviews, beyond the inspections performed of the fuel oil piping system, were conducted to determine the extent. of coating damage in other in-scope, non-cathodically protected, buried piping systems.

2. Given that temporary piping and tanks are currently installed to support the in-scope function once served by the buried fuel oil supply piping and the associated abo>veground fuel oil storage tank, describe how either the temporary piping and tanks or a possible future permanent arrangement will be age managed through the period of extended operation.

United States Nuclear Regulatory Commission Page 10 of 14 SBK-L- 11003 / Enclosure 1 NextEra Enermy Seabrook Response:

1. The extent of condition review performed as part of the root cause evaluation did not result in inspection of other buried piping as a result of this incident.

The root cause evaluation of the failed fuel supply piping stated that "[t]he cause of this event was the fact that the bituminous wrapping in a single location on the Auxiliary Boiler fuel oil supply line, AB-5145-Al-4, had been damaged. The damage resulted in the subsequent degradation of the line local to the damaged wrap area in the form of pipe wall thinning due to corrosion. The thinning resulted in a through wall leak the size of a pinhole and the noted fuel oil leakage." The evaluation further stated that "[it] is not possible to determine when this damaged occurred except to say that this pipe has been buried for approximately twenty years and it would appear that this damage occurred during original installation."

2.' Implementation of the final design change replacing the piping associated with the above-ground fuel oil storage tank will be completed prior to the period of extended operation. The design for buried portions of the system will include a pipe-within-pipe configuration with leak detection capability. Portions of that buried piping that are in-scope for license renewal will be included in the Seabrook Station Buried Piping and Tanks Inspection Program. Portions of that final design that are above-ground, including tanks, will be evaluated in accordance with the License Renewal Rule, 10 CFR 54, and age managed under the appropriate programs (e.g., B.2.1.24 External Surfaces Monitoring, B.2.1.18 Fuel Oil Chemistry, B.2.1.17 Above Ground Steel Tanks) through the period of extended operation.

I Based on the above discussion, the following changes to the License Renewal Application have been made.

1. The following commitment is added to Appendix A, Section A.3:

No.PROGRAM COMMITMENT or TOPIC LOCATION Buried Implement the design change Priorto enteringthe Piping and replacing the buriedAuxiliaryBoiler period of extended 59 Tanks supply piping with a pipe-within-pipe A.2.1.22 operation.

Inspection configuration with leak detection capability.

2. In Section B.2.1.22 (as submitted in the Seabrook Station License, Renewal Application Supplement dated October 29, 2010, SBK-L-10179), on page 7 of Enclosure 1, the following paragraph is added following the last paragraph of Element 1 - Scope of Program:

United States Nuclear Regulatory Commission Page 11 of 14 SBK-L- 11003 / Enclosure 1 Implementation of the final design change replacing the piping associated with the above-groundfuel oil storage tank will be completed prior to the period of extended operation. The design for buried portions of the system will include a pipe-within-pipe configuration with leak detection capability. Portions of that buried piping that are in-scopefor license renewal will be included in the Seabrook Station Buried Piping and Tanks Inspection Program.

Portions of that final design that are above-ground, including tanks, will be evaluated in accordance with the License Renewal Rule, 10 CFR 54, and age managed under the appropriate programs (e.g., B.2.1.24 External Surfaces Monitoring, B.2.1.18 Fuel Oil Chemistry, B.2.1.17 Above GroundSteel Tanks) through the period of extended operation.

Request for Additional Information (RAI) B.2.1.30-1

Background

GALL Report (NUREG-1801), AMP XI.S4, "10 CFR Part 50, Appendix J," Element 4 states that a containment LRT program is effective in detecting degradation of containment shells, liners, and components that compromise the containment pressure boundary, including seals and gaskets. While the calculation of leakage rates demonstrates the leak-tightness and structural integrity of the containment, it does not by itself provide information that would indicate that aging degradation has initiated or that the capacity of the containment may have been reduced for other types of loads, such as seismic loading. This would be achieved with the additional implementation of an acceptable containment inservice inspection program as described in XI.S1 and XI.S2. In addition, 10 CFR Part 50, Appendix J requires a general inspection of internal and external surfaces of the containment prior to a Type A test Issue According to the applicant, the containment surfaces were inspected prior to the most recent Type A test which was performed in 2008 using the "Complex Procedure" for reactor containment integrated leakage rate testing. This "Complex Procedure" states that visual inspections of the exposed interior and exterior surfaces of the containment vessel and the containment enclosure building will be performed.

Based on a -review of the procedure, the staff noted that the containment inspection section of the procedure does not specify examination methods for conducting internal and external inspections that are consistent with ASME Section XI, Subsections IWE and IWL requirements.

Request The applicant is requested to provide the following information:

1. Describe the methods and procedures used to conduct a general inspection of internal and external surfaces of the containment prior to the most recent Type A test.

United States Nuclear Regulatory Commission Page 12of 14 SBK-L- 11003 / Enclosure 1

2. Indicate whether these methods and procedures are consistent with the containment inservice inspection programs described in GALL AMP XI.S1 and XI.S2.

3. Describe the method being used to ensure that internal and external containment inspections are being implemented as described in GALL AMP XI.S1 and XI.S2 and consistent with element 4 of GALL AMP XI.S4, "10 CFR Part 50, Appendix J."

The staff needs the above information to confirm that the effects of aging of the concrete containment will be adequately managed so that it's intended function will be maintained consistent with the current licensing basis for the period of extended operation, as required by 10 CFR 54.21 (a)(3).

NextEra Energy Seabrook Response:

1. The "Reactor Containment Integrated Leakage Rate Test - Type A" complex procedure was used to perform the Type 'A' test in 2008. The procedure contains a prerequisite which states: "A general inspection of the accessible interior and exterior surfaces of the containment structure and components is complete".

The inspection requires that: "The structural integrity shall be determined by a visual inspection of the exposed accessible interior and exterior surfaces of the containment vessel. The inspection shall be performed to verify no apparent changes in appearance of the surfaces or other abnormal degradation. Any abnormal degradation of the containment vessel detected during the above required inspections shall be reported to the Commission in a Special Report pursuant to T.S. 6.8.2 within 15 days."

The general inspection of the accessible interior and exterior surfaces of the containment structure is a requirement from 10 CFR Part 50, Appendix J, subsection V.A, and is in addition to, and totally separate from, the inspection requirements of ASME Section XI, Subsections IWE and IWL.

2. The Seabrook Appendix J Program is consistent with the requirements of NUREG- 1801, XI-S4 10 CFR Part 50, Appendix J. The containment Inspection required by 10 CFR Part 50, Appendix J is in addition to and separate from the requirements of NUREG-1801, Aging Management -Programs XI.S1 and XI.S2, ASME Section XI, Subsections IWE & IWL.

3. As discussed in the Seabrook Station License Renewal Application both the XI.S I and the XI.S2 are consistent with NUREG 1801. The inspections conducted as part of the 10 CFR Part 50, Appendix J Program are consistent with NUREG 1801 XI.S4.

Request for Additional Information (RAI) B.2.1.30-2

Background

GALL Report (NUREG-1801), AMP XI.S4, "10 CFR Part 50, Appendix J," states that Appendix J provides two options, A and B, either of which can be chosen to meet the requirements of a

United States Nuclear Regulatory Commission Page 13 of 14 SBK-L- 11003 / Enclosure 1 containment LRT program. Under Option A, all of the testing must be performed on a periodic interval.

.Option B is a performance-based approach. More detailed information for Option B. is provided in Regulatory Guide (RG) 1.163 and NEI 94-01, Rev. 0.

Nuclear Energy Institute (NEI) 94-01 states that a general visual inspection of the accessible interior and exterior surfaces of the containment system for structural deterioration which may affect the containment leak-tight integrity must be conducted prior to each test, and at a periodic interval between tests based on the performance of the containment system. In addition, NEI 94-01 recommended that these inspections be performed in conjunction or coordinated with the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWE/IWL required examinations.

Regulations for codes and standards in 10 CFR 50.55a. require personnel that examine containment concrete surfaces and tendon hardware, wires, or strands must meet the qualification provisions in IWA-2300 and that the "owner-defined" personnel qualification provisions in IWL-2310(d) are not approved for use.

In LRA B.2.1.30, the applicant states that the Seabrook Station Containment Leakage Rate Testing Program, required by Seabrook Station Technical Specification, implements Option B. In addition, the Seabrook Station Leakage Test Reference is based on the guidance provided in NEI 94-01 and ANSI/

ANS-56.8-1994 with the restrictions identified in RG 1.163.

Issue During the audit of element 4, the staff reviewed the "Complex Procedure" for reactor containment integrated leakage* rate testing and qualification guidance for personnel who conducted visual examinations of concrete containment surfaces. The staff concluded that the qualification of personnel who conduct visual examinations of concrete containment surfaces should be consistent with qualification provisions in IWA-2300 as required by 10 CFR 50.55a.

Request The applicant is requested to provide plans and schedule that will ensure that (1) personnel who perform visual examinations of concrete containment surfaces to comply with the'applicant's commitment to implement Option B for integrated leakage rate tests are qualified in accordance with IWA-2300 requirements and (2) the applicant's 10 CFR Part 50, Appendix J AMP is consistent with GALL AMP XI.S4, "10 CFR Part 50, Appendix J."

NextEra Enermy Seabrook Response:

Personnel performing the visual examinations of concrete surfaces prior to Integrated Leak Rate Test are not qualified in accordance with IWA-2300. There is no requirement in 10 CFR 50.55a that Appendix J general visual inspection personnel be qualified per IWA-2300. This is a general inspection

United States Nuclear Regulatory Commission Page 14 of 14 SBK-L- 11003 / Enclosure 1 of the containment surface for any apparent degradation that would cause failure of the Integrated leak Rate Test. Since there are no requirements for inspection qualifications, the Appendix J aging management program is consistent with GALL AMP XI.S4, "10 CFR Part 50, Appendix J.

Personnel performing the visual examinations required by the ASME Section XI, Subsections IWE and IWL are qualified to the requirements of IWA-2300.

United States Nuclear Regulatory Commission Page 1 of 10 SBK-L- I1003 / Enclosure 2 Enclosure 2 to SBK-L-11003 LRA Appendix A - Final Safety Report Supplement Table A.3 License Renewal Commitment List

United States Nuclear Regulatory Commission Page 2 of 10 SBK-L-1 1003 / Enclosure 2 A.3 'LICENSE RENEWAL COMMITMENT LIST No. PROGRAM or TOPIC COMMITMENT UFSAR SCHEDULE Program to be An inspection plan for Reactor Vessel Internals will be implemented prior to the submitted for NRC review and approval at least twenty- period of extended

1. PWR Vessel Internals four months prior to entering the period of extended A.2.1.7 operation. Inspection plan operation. to be submitted to NRC not less than 24 months prior to the period of extended operation.

2 Enhance the program to include visual inspection for Prior. to the period of

2. Water cracking, loss of material and fouling when the in-scope A.2.1.12 extended operation systems are opened for maintenance.

Inspection of Overhead Heavy Load and Light Enhance the program to monitor general corrosion on the Prior to the period of

3. Load (Related to crane and trolley structural components and the effects of A.2.1.13 exte e perion Refueling) Handling wear on the rails in the rail system. extended operation Systems Inspection of Overhead Heavy Load and Light Enhance the program to list additional cranes for Prior to the periAod of

4. Load (Related to Enhan tothe g r A.2.1.13 period Refueling) Handling monitoring. extended operation Systems Enhance the program to include an annual air quality test Compressed rin g Air requirement s ys te m . , for the Diesel Generator compressed air sub Prior

5. Mo n ito 5.A2..4 A .2 .1 .14e x tond the extended period of doperation o ra i .

.6. Fire Protection Enhance the program to perform visual inspection of A.2.1.15 Prior to the period of penetration seals by a fire protection qualified inspector, extended operation.

United States Nuclear Regulatory Commission Page 3 of 10 SBK-L-1 1003 / Enclosure 2 No. PROGRAM or TOPIC COMMITMENT UFSAR LOCATION SCHEDULE Enhance the program to add inspection requirements

7. Fire Protection such as spalling, and loss of material caused by freeze- A.2.1.15 Prior to the period of thaw, chemical attack, and reaction with aggregates by extended operation.

qualified inspector.

8. Enhance the program to include the performance of Prior to the period of Fire Protection visual inspection of fire-rated doors by a fire protection A.2.1.15 qualified inspector.

Enhance the program to include NFPA 25 guidance for Fr .where sprinklers have been in place for 50 years, they Prior to the period of Fire Water System shall be replaced or representative samples from one or A.2.1.16 extended operation.

more sample areas shall be submitted to a recognized testing laboratory for field service testing".

10. Enhance the program to include the performance of Prior to the period of Fire Water System periodic flow testing of the fire water system in A.2.1.16 extended operation.

accordance with the guidance of NFPA 25.

Enhance the program to include the performance of periodic visual or volumetric inspection of the internal surface of the fire protection system upon each entry to the system for routine or corrective maintenance. These inspections will be documented and trended to determine Within ten years prior to Fire Water System 11'. if a representative number of inspections have been A.2.1.16 the period of extended performed prior to the period of extended operation. If a operiond representative number of inspections have not been operation.

performed prior to the period of extended operation, focused inspections will be conducted. These inspections will be performed within ten years prior to the period of extended operation.

12. Aboveground Steel Enhance the program to include components and aging Prior to the period of Tanks effects required by the Aboveground Steel Tanks. A.2.1.17 extended operation.

United States Nuclear Regulatory Commission Page 4 of 10 SBK-L- 11003 / Enclosure 2 No. PROGRAM or TOPIC COMMITMENT

• UFSAR LOCATION SCHEDULE Enhance the program to include an ultrasonic inspection Within ten years prior to 13.

Aand evaluation of the internal bottom surface of the two A.2.1.17 the period of extended Fire Protection Water Storage Tanks. operation.

Enhance program to add requirements to 1) sample and

14. analyze new fuel deliveries for biodiesel prior to Priorto the period of Fuel Oil Chemistry offloading to the Auxiliary Boiler fuel oil storage tank and A.2.1.18 extended operation.

2) periodically sample stored fuel in the Auxiliary Boiler fuel oil storage tank.

Enhance the program to add requirements to check for

15. Fuel Oil Chemistry the presence of water in the Auxiliary Boiler fuel oil A.2.1.18 Prior to the period of storage tank at least once per quarter and to remove 1extended operation.

water as necessary.

16. Enhance Fuel Oil Chemistry inspectionthe program of the dieseltofire require pumpdraining, cleaning fuel oil day tanks and on a A.2.1.18 Prior to theoperation.

extended period of frequency of at least once every ten years.

Enhance the program to require ultrasonic thickness measurement of the tank bottom during the 10-year 17 draining, cleaning and inspection of the Diesel Generator Prior to the period of Fuel OilChemistry

17. fuel oil storage tanks, Diesel Generator fuel oil day tanks, A.2.1.18 extended operation.

diesel fire pump fuel oil day tanks and auxiliary boiler fuel oil storage tank.

18. Reactor Vessel Enhance the program to specify that all pulled and tested Prior to the period of Su.ReilanctrVessel capsules, unless discarded before August 31, 2000, are A.2.1.19 exte e perion.

Surveillance placed in storage. extended operation.

Enhance the program to specify that.if plant operations exceed the limitations or bounds defined by the Reactor Vessel Surveillance Program, such as operating at a

19. Reactor Vessel lower cold leg temperature or higher fluence, the impact A.2.1.19 Prior to the period of Surveillance of plant operation changes on the extent of Reactor extended operation.

Vessel embrittlement will be evaluated and the NRC will be notified.

United States Nuclear Regulatory Commission Page 5 of 10 SBK-L-1 1003 / Enclosure 2 No. PROGRAM or TOPIC COMMITMENT UFSAR LOCATION SCHEDULE Enhance the program as necessary to ensure the appropriate withdrawal schedule for capsules remaining in the vessel such that one capsule will be withdrawn at

20. Reactor Vessel an outage in which the capsule receives a neutron Prior to the period of Surveillance fluence that meets the schedule requirements of 10 CFR A.2.1.19 extended operation.

50 Appendix H and ASTM E185-82 and that bounds the 60-year fluence, and the remaining capsule(s) will be removed from the vessel unless determined to provide meaningful metallurgical data.

Enhance the program to ensure that any capsule

21. 'Reactor Vessel removed, without the intent to test it, is stored in a Prior to the period of Surveillance manner permit itswhich futuremaintains it in a condition which would A.2.1.19 Prientotheperion.

use, including during the period of extended operation.

extended operation.

• 22. Within ten years prior to One-Time Inspection Implement the One Time Inspection Program. A.2.1.20 the period of extended operation.

Implement the Selective Leaching of Materials Program.

The program will include a one-time inspection of Within five years Iprior to

23. Selective Leaching of selected components where selective leaching has not A.2.1.21 the period of extended Materials been identified and periodic inspections of selected components where selective leaching has been identified.

24. Buried Piping And Implement the Buried Piping And Tanks Inspection Within ten years prior to Tanks Pinpeiong AProgram. A.2.1.22 entering the period of extended operation

United States Nuclear Regulatory Commission Page 6 of 10 SBK-L-1 1003 / Enclosure 2 No. PROGRAM or TOPIC COMMITMENT. UFSAR LOCATION SCHEDULE One-Time Inspection of Implement the One-Time Inspection of ASME Code Within ten years prior to

25. ASME Code-Class 1 I t A.2.1.23 the period of extended Small Bore-Piping Class 1 Small Bore-Piping Smal Program.

Boe-Piingoperation.

Enhance the program to specifically address the scope of the program, relevant degradation mechanisms and

26. External Surfaces effects of interest, the refueling outage inspection Moniterng frequency, the inspections of opportunity for possible A,2.1.24 Prior to the period of Monitoring corrosion under insulation, the training requirements for extended operation.

inspectors and the required periodic reviews to determine program effectiveness.

Inspection of Internal

27. Surfaces in - Implement the Inspection of Internal Surfaces in Prior to theperiod of Miscellaneous Piping Miscellaneous Piping and Ducting Components Program. A.2.1.25 extended operation.

and Ducting Components

28. Enhance the program to add required equipment, lube oil Prior to the period of Lubricating Oil Analysis analysis

" required, sampling frequency, and periodic oil chagesextended A.2.1.26 operation.

changes.

Enhance the program to sample the oil for the Prior to the period of

29. Lubricating Oil Analysis Switchyard SF 6 compressors and the Reactor Coolant A.2.1.26 extended operation.

pump oil collection tanks.

Enhance the program to require the performance of a

30. one-time ultrasonic thickness measurement of the lower Prior to the period of Lubricating Oil Analysis portion of the Reactor Coolant pump oil collection tanks A.2.1.26 extended operation.

prior to the period of extended operation.

31. ASME Section XI, Enhance procedure to include the definition of Prior to the period of Subsection IWL "Responsible Engineer". A.2.1.28 extended operation.

United States Nuclear Regulatory Commission Page 7 of 10 SBK-L-1 1003 / Enclosure 2 No. PROGRAM or TOPIC COMMITMENT UFSAR LOCATION SCHEDULE Enhance procedure to add the aging effects, additional Prior to the period of

32. Structures Monitoring locations, inspection frequency and ultrasonic test A.2.1.31 Program. requirements extended operation.

33. Structures Monitoring Enhance procedure to include inspection of opportunity Prior to the period of Program' when planning excavation work that would expose A.2.1.31 extended operation.

inaccessible concrete.

Electrical Cables and Connections 3.Sbect to Not Implement the Electrical Cables and Connections Not Prior to the period of

34. Subject to 10 CFR Subject to 10 CFR 50.49 Environmental Qualification A.2.1.32 exte e perion.

50.49 Environmental Rqientprga.extended operation.

Qualification Requirements program.

Requirements Electrical Cables and Connections Not Subject to 10 CFR Implement the Electrical Cables and Connections Not Prior to the period of

35. 50.49 Environmental- Subject to 10 CFR 50.49 Environmental Qualification A.2.1.33 extended operation.

Qualification Requirements Used in Instrumentation Circuits program.

Requirements Used in Instrumentation Circuits Inaccessible Power Cables Not Subject to Implement the Inaccessible Power Cables Not Subject to Prior to the period of

36. 10 CFR 50.49 10CFR 50.49 Environmental Qualification Requirements A.2.1.34 Environmental 1ro 50.4 extended operation.

Qualification program.

Requirements

37. Prior to the period of Metal Enclosed Bus Implement the Metal Enclosed Bus program. A.2.1.35 extended operation.

38. FPrior to the period of Fuse e.Fuse Holders program. A.2.1.36 extended operation.

Electrical Cable Connections Not Implement the Electrical Cable Connections Not Subject Prior to the period of

39. Subject to 10 CFR to 10 CFR 50.49 Environmental Qualification A.2.1.37 extne perion.

50.49 Environmental, extended operation.

Qualification SRequirements

United States Nuclear Regulatory Commission Page 8 of 10 SBK-L-1 1003 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT LOCAT LOCATION SCHEDULE,

40. 345 KV SF 6 Bus Implement the 345 KV SF 6 Bus program. A.2.2..1 Prior to the period of extended operation.

41. Metal Fatigue of Enhance the program to include additional transients Prior to the period of Reactor Coolant beyond those defined in the Technical Specifications and A.2.3.1 extended operation. -

Pressure Boundary UFSAR.

Metal Fatigue of Enhance the program to implement a software program,

42. Reactor Coolant to count transients to monitor cumulative usage on A.2.3.1 Prior to the period of Pressure Boundary selected components. extended operation.

Pressure -Temperature The updated analyses be submitted at the will Limits, including Low Seabrook Station will submit updates to the P-T curves appropriate time to comply

43. Temperature and LTOP limits to the NRC at the appropriate time to A.2.4.1.4 w ith 10 CF 50 mppd comply with 10 CFR 50 Appendix G.

Overpressure Protection LimitsG, with 10 CFR 50 Appendix Fracture Toughness Limits Requirements.

(1) Consistent with the Metal Fatigue of Reactor Coolant Pressure Boundary Program Seabrook Station will update the fatigue usage calculations using refined fatigue analyses, if necessary, to determine acceptable CUFs (i.e., less than 1.0) when accounting for the effects of the reactor water environment. This includes applying the appropriate Fen factors to valid CUFs determined from an existing fatigue analysis validfor the period of extended operation or from an analysis using an NRC-Environmentally- approved version of the ASME code or NRC-approved At least two years prior to alternative (e.g., NRC-approved code case). A.2.4.2.3 entering the period of

44. Assisted Analyses Fatigue A2423 etrn (TLAA) (2) If acceptable CUFs cannot be demonstrated for all the extended hoperation.

eido selected locations, then additional plant-specific locations will be evaluated. For the additional plant-specific locations, if CUF, including environmental effects is greater than 1.0, then Corrective Actions will be initiated, in accordance with the Metal Fatigue of Reactor Coolant Pressure Boundary Program, B.2.3.1. Corrective Actions will include inspection, repair, or replacement of the affected locations before exceeding a CUF of 1.0 or the effects of fatigue will be managed by an inspection

United States Nuclear Regulatory Commission Page 9 of 10 SBK-L-1 1003 / Enclosure 2 UFSAR No. PROGRAM or TOPIC COMMITMENT LOCAT LOCATION SCHEDULE program that has been reviewed and approved by the NRC (e.g., periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method accepted by the NRC).

Mechanical Equipment Revise Mechanical Equipment Qualification Files. A245.9 Prior to the period of Qualification extended operation.

Protective Coating Enhance the program by designating and qualifying an Prior to the period of

46. Monitoring and Inspector Coordinator and an Inspection Results A.2.1.38 extended Operation Maintenance Evaluator.

Enhance the program by including, "Instruments and Protective Coating Equipment limited needed for to, flashlight, inspection spotlights, may include, marker but not be pen, mirror, Prior to the period of

47. Monitoring and measuring tape, magnifier, binoculars, camera with or extended operation Maintenance without wide angle lens, and self sealing polyethylene sample bags."

Protective Coating Enhance the. program to include a review of the previous

48. Monitoring and A.2.1.38 Prior to the period of Maintenance two monitoring reports. extended operation Protective Coating Enhance the program to require that the inspection report Prior to the period of

49. Monitoring and is to be evaluated by the responsible evaluation A.2.1.38 extended operation Maintenance personnel, who is to prepare a summary of findings and recommendations for future surveillance or repair.

ASME Section XI, Perform testing of the containment liner plate for loss of Prior to the period of

50. Subsection IWE material. A.2.1.17 extendedoperation.

ASME Section Xl, Perform confirmatory testing and evaluation of the Prior to the period of

51. Subsection IWL Containment Structure concrete A.2.1.28 extended operation

United States Nuclear Regulatory Commission Page 10 of 10 SBK-L-1 1003 / Enclosure 2 No. PROGRAM or TOPIC COMMITMENT LOFSAR LOCATION SCHEDULE Implement measures to maintain the exterior surface of

52. ASME Section.XI, the Containment Structure, from elevation -30 feet to +20 A.2.1.28 Prior to the period of Subsection IWL feet, in a dewatered state. extended operation Reactor Head Closure The Unit 2 reactorhead closure stud(s) manufactured Prior to the period of

53. from the barthat has a yield strength > 150 ksi will be A.2.1.3 extendedhoperion.

Studs replaced. extended operation.

Unless a permanentalternaterepaircriteriachanging the ASME code boundary is approved by the NRC, or Program to be the Seabrook Station steam generatorsare changed submitted to NRC at Steam GeneratorTube to eliminate PWSCC-susceptible tube-to-tubesheet A

54. Integrity welds, submit a plant-specificaging management A.2.1.10 least 24 months priorto program to manage the potentialaging effect of theperiodof extended cracking due to PWSCC at least twenty-four months operation.

prior to entering the Periodof Extended Operation.

Seabrook will perform an inspectionprior to entering period of extended operationsof each steam Steam GeneratorTube generatorto assess the condition of the divider plate A2110 Priorto entering the Steam assembly unless operating experience and/or period of extended Integrity analyticalresults show that crack propagationinto operation RCS pressure boundary is not possible, then the inspections need not be performed.

5.Closed-Cycle Cooli+ng Revise the station program documents to reflect the Priorto entering the

56. Cloed-Cylem CEPRI Guideline operatingranges and Action Level A.2.1.12 period of extended Water System values for hydrazine and sulfates. operation.

Update Technical Requirement Program 5.1, (Diesel

57. Fuel Oil Chemistry Fuel Oil Testing Program)ASTM standardsto ASTM A.2.1.18 Priorto the periodof D2709-96 and ASTM D4057-95 requiredby the GALL extended operation.

XI.M30 Rev I

58. kel And PeNetrat s The Nickel Alloy Aging Nozzles and Penetrations

58. and Penetrations program will implement applicableBulletins, Generic A.2.2.3 extended operation.

58._andPenetrations Letters, and staff acceptedindustry guidelines.

Buried Piping and Implement the design change replacingthe buried Priorto entering the Tanks Inspection Auxiliary Boiler supply piping with a pipe-within-pipe A.2.1.22 period of extended configuration with leak indication capability, operation.