SBK-L-11154, Response to Request for Additional Information to License Renewal Application - Set 15 and LRA Changes

Response to Request for Additional Information to License Renewal Application - Set 15 and LRA Changes

ML11227A023
Person / Time
Site:	Seabrook   (CPPR-0136, NPF-086)
Issue date:	08/11/2011
From:	Freeman P NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-11154
Download: ML11227A023 (31)

Text

NExTera ENERG

• S SEA BROO K August 11, 2011 SBK-L-11154 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 Request for Additional Information - Set 15

References:

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

2. NRC Letter "Request for Additional Information For the Review of the Seabrook Station License Renewal Application" (TAC NO. ME4028) - Request for Additional Information Set 15," June 29, 2011. (Accession Number ML11178A338)

3. NextEra Energy Seabrook, LLC letter SBK-L-10204, "Seabrook Station Response to Request for Additional Information, NextEra Energy Seabrook License Renewal Application Aging Management Programs - Set 1", December 17, 2010. (Accession Number ML103540534)

4. NextEra Energy Seabrook, LLC letter SBK-L-1 1063, "Seabrook Station Response to Request for Additional Information, NextEra Energy Seabrook License Renewal Application - Set 13", April 14, 2011. (Accession Number ML1 f108A131)

5. NextEra Energy Seabrook, LLC letter SBK-L-11015, "Seabrook Station Response to Request for Additional Information, NextEra Energy Seabrook License Renewal Application-Sets 6, 7 and 8", February 3, 2011. (Accession Number ML110380081)

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 C6de of Federal Regulations, Title 10, Parts 50, 51, and 54.

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

United States Nuclear Regulatory Commission SBK-L-11154 / Page 2 In Reference 2, the NRC requested additional information in order to complete its review of the License Renewal Application (LRA). The requests are a follow-up to responses provided in References 3 and 4. 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.

In Reference 5, the response to RAI 3.3.2.15-1 was inadvertently duplicated as the same response to RAI 3.3.2.15-01. The correct response to RAI 3.3.2.15-1 is provided in Enclosure 2 of this letter.

Commitment numbers 50 and 51 are revised and commitments 67 and 68 added. There are no other new or revised regulatory commitments contained in this letter. Enclosure 3 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.

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, NextEy Seabrook, LLC.

Paul 0. Freeman Site Vice President

Enclosures:

- Response to Request for Additional Information Seabrook Station License Renewal Application, Set # 15 and Associated LRA Changes -

Revised NextEra Energy Seabrook response to RAI 3.3.2.15-1 provided in letter SBK-L-11015 dated February 3,2011 -

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-I 1154 / Page 3 cc:

W.M. Dean, G. E. Miller, W. J. Raymond, R. A. Plasse Jr.,

M. Wentzel, NRC Region I Administrator NRC Project Manager, Project Directorate 1-2 NRC Resident Inspector NRC Project Manager, License Renewal 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-11154 / Page 4 NEx'7era-SEABROOK I, Kenneth J. Browne, Plant General Manager 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 Ili-day of August, 2011 KMefiieth J. Browne Plant General Manager

-1ý4 Notary Pub Ic

Enclosure I to SBK-L-11154 Response to Request for Additional Information Seabrook Station License Renewal Application Set 15 and Associated LRA Changes

United States Nuclear Regulatory Commission Page 2 of 13 SBK-L-11154/ Enclosure 1 Request for Additional Information (RAI) Follow-up B.2.1.27-1:

Background:

By letter dated April 14, 2011, NextEra Energy Seabrook (the applicant) responded to a staff request for additional information (RAI) regarding testing of the containment liner for possible loss of material from the concrete side of the liner. In the response, the applicant committed to ultrasonic testing (UT) of the containment liner at 100 intervals around the accessible circumference of the containment near the moisture barrier at the -

26' elevation. The applicant committed to finishing the UT no later than December 31, 2015. The applicant further stated that in accordance with ASME Section XI, Subsection IWE 1241 (a), Seabrook will designate the area of the containment liner that is within 10 inches of the moisture barrier at the containment basement floor for examination.

Issue:

IWE 1241(a) requires augmented examination of the containment liner surface area in accordance with Table IWE-2500-1, examination category E-C. Item E.4.12 of Table IWE-2500-1 requires 100% UT measurement of the area designated for augmented examination during each inspection period until the areas examined remain essentially unchanged for three consecutive inspection periods. In the RAI response, the applicant did not explain why a one-time UT examination at 100 increments (-36 measurements) to be completed by December 31, 2015, was appropriate in lieu of IWE-1241(a) and Table IWE-2500-1 requirements.

Request:

Provide technical justification for not following the requirements of IWE-1241(a) and Table IWE-2500-1 for performing UT examination of 100% of the area designated for augmented examination during each inspection period until the area remains essentially unchanged for three consecutive inspection periods. The staff is concerned that the December 31, 2015, deadline for one-time UT examination and the spacing of the UT measurements at 10' increments around the containment circumference may not be able to detect and establish a trend of the potential degradation of the liner plate over the long term.

NextEra Enermy Seabrook Response The containment liner plate in the vicinity of the moisture barrier at Seabrook Station has not exhibited any evidence indicative of loss of material or conditions likely to cause accelerated degradation that would require an ASME Section XI, Subsection IWE repair or augmented examination per IWE 1241 (a).

Seabrook Station, refueling outage 14 (OR14) Containment Liner Examination Summary, dated May 1, 2011 reported 83 indications found in the vicinity of the moisture barrier. These 83 indications in the vicinity of the moisture barrier were on the concrete floor, the moisture barrier, and the containment liner. They consisted of: 1)

United States Nuclear Regulatory Commission Page 3 of 13 SBK-L-11154/ Enclosure 1 apparent lack of bonding, and degradation of the moisture barrier, 2) concrete spalling and chipping, and coating chipping and blistering on the floor; and 3) coating chipping, blistering/cracking, and in 5 indications of minor surface corrosion on the liner plate. All of the indications were minor in nature less than 1 sq. inch area and does not meet the requirements of IWE -1240 for augment examination.

All OR14 indications have been evaluated and accepted by engineering, corrective measures, such as recoating performed, or approved by engineering for remediation during refueling outage OR15 in 2012. The evaluations and corrective actions are documented in the 'corrective action process.

As stated above Seabrook Station performed an inspection of the containment liner around the moister barrier in 2011 during OR14, and found no areas that required IWE repairs or augmented examination. Seabrook Station will additionally confirm no loss of material of the liner plate around the moister barrier. This confirmatory process will include ultrasonic thickness (UT) examinations of 3600 of the accessible liner plate in a band extending from the moisture barrier at el. -26', to ten inches above the moisture barrier. The examination process will perform 50 UT's at approximately equal spacing around the accessible circumference of the liner plate.

There are no current indications that require ASME Section XI, Subsection IWE repair or augmented examination. To allow for process development, planning, and scheduling, the initial confirmatory examination will be conducted within the next two refueling outages, OR15 in 2012 and OR16 in 2014. In the absence of any positive indication of material loss being identified during the initial examination, confirmatory examinations will be repeated on an interval of every five refueling outages. Based on the above discussion, the following changes are made to the LRA:

1) License Renewal Application Appendix B, Section B.2.1.27, page B-151, as changed by RAI B2.1.27-1 in SBK-L-11063, is further changed to read as follows:

Enhancements Searook will pcrfrm ultraseonice thiekaess (UT) testing of the liner plate insi cotainment fory ross Of materia n the ronicete side of the liner) The testing will be subject to ASME Section X!, Subasetaion eniE acceptance criteria. The UT testing ta2gets the area near-the mcisture barrier at el. 26 and a nominal 10' inwrefrens a5ound the accessible circumfereene of tontaiiment.

This will be completed by December 31, 2015 and at intoveals of no mre than, five refueling outages there NextEra Energy Seabrook commits to initiating a confirmatory examination process to initially and periodically veri[y the soundness of the liner plate. The confirmatory process will include ultrasonic thickness (UT) examinations of 3600 of the accessible liner plate in a band extending from the moisture barrier at el. -

26'Y to ten inches above the moisture barrier. The examination process will perform 50 UT's at approximately equal spacing around the accessible circumference of the liner plate as disc ussed above.

United States Nuclear Regulatory Commission SBK-L-11154/ Enclosure 1 Page 4 of 13 NextEra Energy Seabrook will conduct confirmatory UT examinations of the containment liner plate in the vicinity of the moisture barrier for loss of material within the next two refueling outages, CR15 or CR16. In the absence of any positive indication of material loss being identified during the initial examination, confirmatory examinations will be repeated at five refueling outage intervals.

2) License Renewal Application Appendix A, Section A.3, page A-43, as changed by RAI B2.1.27-1 in SBK-L-1 1063, is further changed and added to, as follows:

No.

PROGRAM or TOPIC COMMITMENT UFSAR SCHEDULE LOCATION leater-thant Dee... br. 31, 2015 Within the next two Perform UT testing of the containment liner refueling outages, 500ASME Section XI, OR5oOR16, and

50.

Subsection IWE plate in the vicinity of the moisture barrier for A.2.1.27 repeated at intervals loss of material.

of no more than five refueling outages Request for Additional Information (RAI) Follow-up B.2.1.27-2:

Background:

By letter dated April 14, 2011, the applicant responded to a staff RAI regarding UT examinations of the containment liner below the fuel transfer tube which had been exposed to borated water leakage. In the response, the applicant stated that the area was subject to UT examinations and had been examined and accepted.

Issue:

The applicant provided no information about when the UT examinations had been conducted or the results of the examinations. It is not clear if the containment liner plate below the fuel transfer tube that has been exposed to the borated water leakage was designated for augmented examination in accordance with IWE-1241(a). In addition, the RAI response did not provide the timing for the initial, and three subsequent consecutive, examinations to comply with IWE-1240 and Table IWE-2500-1 requirements.-

Request:

Provide the dates and results of the UT examinations of the containment liner plate area below the fuel transfer tube. If any of the values were below the minimum wall thickness, explain how the areas were repaired or evaluated.

United States Nuclear Regulatory Commission Page 5 of 13 SBK-L-I 1154/ Enclosure 1 NextEra Energy Seabrook Response The containment liner plate at the fuel transfer tube penetration (PEN-X62) was subject to a VT-3 examination under Subsection IWE on October 15, 2009. Five indications were subjected to UT examination, power tool cleaned, and recoated. None of these indications had measured values less than nominal wall thickness, however, the area was incorrectly designated for ASME Section XI, Subsection IWE-1241 augmented inspection.

The first of three planned consecutive augmented inspections was performed in April 2011. The entire surface area at fuel transfer tube penetration (PEN-X62) was subjected to visual examination, and areas around the five indications identified in October 2009 were subjected to UT scans. The area of the prior minimum thickness reading of 0.411 inches was re-measured as ranging from 0.400 to 0.409 inches. The thickness measured is greater than tnom (0.375 inches); no repairs are required.

Request for Additional Information (RAI) Follow-up B.2.1.31-1:

Background:

By letter dated April 14, 2011, the applicant responded to a staff RAI regarding concrete degradation due to groundwater in-leakage and the occurrence of Alkali-Silica Reaction (ASR) in the concrete. The applicant stated that an extent of condition investigation regarding the ASR degradation was on-going, along with the development of a long range aging management plan. The applicant explained that the plan would not be fully developed and implemented until December 2013. The applicant's response also listed several American Society for Testing and Materials (ASTM) standards that would be.

used to estimate the ASR reaction rate.

Issue:

The applicant provided no specific information about the applicability of the original operability determination conducted when ASR was initially identified. The response also lacked specific information about what tests (laboratory and in-situ) would be conducted and when. The response also made no mention of how possible reductions in concrete shear strength were being estimated and addressed. In addition, the RAI response stated that cores were being taken in accordance with American Concrete Institute (ACI) 228.1R-03; however, it did not address the statistical validity and size of core samples taken or planned at each location.

Request:

1. Explain if the current operability determination remains valid until the long term aging management plan is developed and implemented.

United States Nuclear Regulatory Commission Page 6 of 13 SBK-L-11154/ Enclosure 1

2. Explain how the concrete tests and evaluations performed so far can be used to establish a trend in degradation of the affected structures until the long term aging management plan is implemented.

3. Provide detailed and comprehensive information regarding the planned approach to addressing ASR degradation throughout the site. The description of the actions planned to test, evaluate, and mitigate ASR in the RAI response do not provide sufficient details for the staff to determine if the aging of the structures will be adequately managed during the period of extended operation.

At a minimum include a discussion of the following:

a.

The locations where monitoring or sampling will be conducted, and how these results will be used to address other susceptible locations.

b.

The frequency of the monitoring and sampling to establish a trend in degradation of the structures and rate of ASR, and why the provided frequency is adequate.

c.

Detailed information about the planned in-situ monitoring or testing and laboratory testing. This should include the test method, frequency, and schedule.

d.

How the number of concrete samples taken or planned from each structure will ensure statistical validity.

e.

How the length of core samples taken or planned will account for variation of ASR across the wall thickness.

f.

How the extent of degradation/corrosion of rebars will be established in the ASR affected areas during the period of extended operation.

g.

How the reduction in load carrying capacity in the steel embedments and anchors be established in the ASR affected areas during the period of extended operation.

h.

How the results of the petrographic examination will be used to determine quantitative damage in concrete and rate of degradation for the period of extended operation.

i.

Plans, if any, for relative humidity and temperature measurements of affected concrete areas over the long term.

j.

Plans to perform stiffness damage tests to estimate the expansion attained to date in ASR affected concrete.

k.

How the current and future rate of expansion of concrete will be determined to ensure that bond between the rebar and concrete is effective over the long term.

1.

How the results of concrete compressive strength and modulus of elasticity conducted so far will be adjusted to account for future degradation during the

United States Nuclear Regulatory Commission Page 7 of 13 SBK-L-11154/ Enclosure 1 period of extended operation.

4. Explain how the possibility of a reduction in shear strength capacity due to ASR degradation is being evaluated and addressed since core samples are not being used to establish the tensile strength of concrete. The response should include a discussion of how the possible reduction is being quantified and how the reduction is shown to be acceptable for the period of extended operation.

NextEra Energy Seabrook Response

1. The current operability determination is expected to remain valid but may require modification, as discussed below. A comprehensive plan to evaluate and address ASR concrete degradation, and develop and implement a long term monitoring plan is ongoing, (See Item 2 response below).

As required by 10 CFR § 54.30(a), if information / results are identified, that impact the current operability determination, they will be evaluated and addressed accordingly. If the reviews show that there is not reasonable assurance that during the current license term, concrete affected by Alkali Silica Reaction is in compliance with applicable design codes, then NextEra Energy Seabrook will take measures under its current license, as appropriate, to ensure that the intended function of those systems, structures or components will be maintained in accordance with the current licensing basis ("CLB") throughout the term of its current license. Thus, by regulation, compliance with the CLB must be maintained until the long term aging management plan is developed and implemented.

As noted in the current operability determination, the areas of concrete affected by Alkali Silica Reaction are in compliance with the applicable design codes stated in the CLB. Structural integrity of the affected structures is fully qualified and all system, structures, and components housed within the structures are capable of performing their design function. The long term effects of the ASR condition are being monitored by the Structures Monitoring Program and the status of the condition is included in the Structures Health Report which, reports the results of subsequent investigations and testing to the Plant Health Committee. Should the condition degrade further, a higher level of qualification analysis will be employed to demonstrate that significant margin exists for operability.

2. Detailed and comprehensive information regarding the planned approach to addressing ASR degradation throughout the site will be included in an engineering, evaluation scheduled to complete in March 2012. The content of the evaluation will include: discussion of degradation mechanisms in concrete, identification of areas susceptible to ASR, progress of in-situ testing of concrete and impact on current licensing basis calculations and analyses, progress of lab testing to establish ASR degradation rates in concrete, and mitigation techniques. Specific questions presented in Follow-up RAI B.2.1.31-1, items 2 through 4, will be addressed in this evaluation.

NextEra Energy Seabrook plans to update the structures monitoring program, as

United States Nuclear Regulatory Commission Page 8 of 13 SBK-L-11154/Enclosure 1 appropriate, to manage aging related to ASR in concrete structures based on the engineering evaluation results.

3. Discussion Items 3a through 31 will be addressed in the evaluation described in Item 2 above.

4. See Item 2 response above.

Request for Additional Information (RAI) Follow-up B.2.1.31-4:

Background:

By letter dated April 14, 2011, the applicant responded to a staff RAI regarding past spent fuel pool (SFP) leakage and explained that a concrete core would be taken by December 31, 2015, in an area that had been wetted by the leakage. The applicant further stated that the SFP leak-off system is routinely hydro-lazed to ensure that it is free-flowing. During a conference call on May 31, 2011, the applicant also noted that SFP leakage had been detected during the spring 2011 refueling outage.

Issue:

1. The applicant did not explain why December 31, 2015, was an acceptable deadline for the concrete core, nor did they commit to taking the core.

2. The applicant did not identify, or justify, a frequency for hydro-lazing the leak-off system. The applicant also did not commit to continuing the hydro-lazing during the period of extended operation.

3. The applicant has not provided the staff with information on the new operating experience regarding the recent SFP leakage.

Request:

1. Provide technical justification for the adequacy of the December 31, 2015, deadline for the SFP concrete core, or provide a new deadline and appropriate justification.

Commit to complete the core by the proposed deadline.

2. Identify the frequency that the leak-off system is ensured to be free-flowing. Provide technical justification for the frequency and commit to maintain the leak-off system free-flowing for the remainder of the operating term, including the period of extended operation.

3. Provide information on the recent leakage from the SFP. Include when the leakage was identified, the amount of leakage, the probable leakage path and source, and how the leakage is being addressed. Explain whether or not the leakage is contained within the leak-off system and provide technical justification for this conclusion. Also

United States Nuclear Regulatory Commission Page 9 of 13 SBK-L-11154/ Enclosure I provide results of any chemical analysis (e.g., pH, iron content, etc.) that has been done on the leakage in the past and whether or not periodic chemical analysis will be performed on the leakage in the future.

NextEra Energy Seabrook Response

1. Seabrook Station does not have continuous borated water leakage from the spent fuel pool. Currently, any leakage from the spent fuel pool collects in a steel catch basin installed in the sump and does not come in contact with concrete.

NextEra Energy Seabrook commits to confirm the absence of embedded steel corrosion by performing a shallow core sample in an area subjected to wetting of borated water during the time frame of the spent fuel pool leakage. The core samples will be examined for degradation of concrete from borated water and also expose rebar to detect any degradation such as loss of material.

As demonstrated by examination of concrete cores from the Connecticut Yankee spent fuel pool and Salem Nuclear Generating Station referenced in the "Safety Evaluation Report Related to the License Renewal of Salem Nuclear Generating Station Units 1 and 2" (ADAMS Accession Number ML11164A051), the structural capacity will not be significantly affected by exposure to borated water. In addition, borated water is not in continuous contact with concrete at Seabrook Station. Hence performing a confirmatory core bore and exposing rebar by December 31, 2015 is adequate.

2. Seabrook Station currently performs hydro-lazing of the spent fuel pool leakoff lines at a 4 l/2 year frequency and will maintain this throughout the period of extended operation. Leak-off is recorded once a month on a spent fuel pool leakage spread sheet. The System Engineer monitors the leak-off telltale drains via the spread sheet for unusual leakage or lack thereof, which could be an indicator of blockage.

Monitoring will continue throughout the period of extended operation.

3. Currently the spent fuel pool leakoff collection is analyzed for gamma and tritium activity monthly. On April 6, 2011, tritium activity concentration measured in Spent Fuel Pool (SFP) zone 6 tell-tale leakage collection pipe indicated a step increase from 2.58E-5 ýtCi/ml to 7.87E-3 iCi/ml.

The increased leak rate occurred coincident with refilling of the cask loading pool which had previously been drained to support maintenance and testing of the spent fuel transfer system equipment. The tritium activity concentration increased by about a factor of 300 and the calculated pool leak rate was 1.2 gpd. Subsequent measurements identified the leak rate peaked at approximately 2.57 gpd on 4/10/2011, after which leakage decreased to the current level of 0.016 gpd (approximately 2oz. per day) by 5/9/2011.

United States Nuclear Regulatory Commission Page 10 of 13 SBK-L-11154/ Enclosure I On average, about 10 gallons per day of groundwater leaks out of the zone 6 tell-tale collection pipe. This groundwater has background contamination from tritium that is diffusing out of the concrete that was originally contaminated from the pool leakage identified in 1999. That leakage was terminated in 2004 with the application of the first non-metallic liner. Groundwater leakage is monitored by the Structures Monitoring Program.

Fuel pool volumetric leakage is estimated by taking the ratio of the leak-off line tritium concentration to the pool tritium concentration and multiplying that value by the amount of zone 6 leakage pumped out from the collection tank. In this particular instance, the only leak-off line that indicated any leakage was zone 6.

There are several potential causes for the increased leakage, and each is discussed below. Those include: -

A new SS liner plate leak in an area not lined with the non-metallic liner.

A failure in the new non-metallic liner at the same location as a SS liner failure.

A skimmer pit leak Corrective actions are being implemented and are on going, such as:

Determining if the skimmer pit(s) are the source of the current leakage.

Verifying integrity of cask loading area non-metallic liner through drain down and inspection.

Revising procedures for cask filling to limit pool level.

Revising the maintenance work order for removal sequence of the weir gate, or reduce the height of the weir gate.

Determining whether the current design of skimmer pits is appropriate and what changes need to be made to prevent leakage out of the pit.

Perform a qualified visual inspection of the weld at the skimmer plate to discharge line interface to determine whether there is actually a seal weld.

The above corrective actions are scheduled to be completed by 12/31/2011.

As explained in response #2, the leakoff lines are hydro-lazed every 4 12 years and the System Engineer monitor's the leak-off telltale drains via a collection spread sheet for unusual leakage conditions.

United States Nuclear Regulatory Commission Page 11 of 13 SBK-L-11154/Enclosure 1 Currently the spent fuel pool leakoff collection is'analyzed for gamma and tritium monthly. The program will be enhanced to perform sampling for chlorides, sulfates, pH and iron for four quarters (for seasonal variations) of one year once every 5 'years.

The information from these samples will be incorporated into the Structures Monitoring Program assessments.

Based on the above discussion, the following changes are made to the LRA:

1) License Renewal Application Appendix B, Section B.2.1.31, page B-169, is revised to add Enhancement Id and 3 as follows:

1.d Perform a confirmatory core bore and expose rebar in an area under the catch basin in spent fuel pool leakage sump.

3. Enhance procedure to perform chemistry sample of the spent fuel pool leakoff collection points.

a. Procedure CP 3.1, "Primary Chemistry Control Program" will be enhanced to include chemistry sampling of the spent fuel pool leakoff collection point for chlorides, sulfates, pH and iron for four quarters of one year once every 5 years.

2) License Renewal Application Appendix A, Section commitments as follows:

A.3,, is revised to add No.

PROGRAM or TOPIC COMMITMENT UFSAR SCHEDULE LOCATION Perform one shallow core bore in an area 67 Structures Monitoring that was continuously wetted from borated No later than 67 Ptrogturam Monito water to be examined for concrete A.2.1.31 December 3,21 Program degradation and also expose rebar to detect December 31, 2015 any degradation such as loss of material.

Perform sampling at the spent fuel pool leakoff collection point for chlorides, 68 Structures Monitoring sulfates, pH and iron for four quarters of A.2.1.31 Starting January Program one year once every 5 years.

2014

United States Nuclear Regulatory Commission Page 12 of 13 SBK-L-11154/ Enclosure 1 Request for Additional Information (RAI) Follow-up B.2.1.28-3:

Background:

By letter dated December 17, 2010, the applicant responded to RAI B.2.1.28-3 regarding possible testing of the containment concrete. In the response, the applicant enhanced the ASME Section XI, Subsection IWL AMP to include confirmatory testing of the containment concrete to determine the compressive strength, the presence or absence of ASR, the concrete modulus of elasticity, and the presence or absence of rebar degradation. The applicant committed to complete the testing prior to the period of extended operation.

Issue:

During several conversations with the staff during the license renewal inspection the week of April 4, 2011, as well as conference calls on April 27, and May 31, 2011, the applicant indicated that they did not want to remove core bores from the containment.

However, the staff is unaware of any method other than core bores that can be used to determine all the concrete properties discussed in Commitment 51 in the letter dated December 17, 2010. In addition, it is not clear how the possible degradation/corrosion of the rebar will be established. Furthermore, one time tests prior to the period of extended operation in 2030 can be used to establish a trend during the period of extended operation.

Request:

1. Verify whether or not the enhancement, and Commitment 51, regarding testing to confirm containment concrete properties, made in the letter dated December 17, 2010, is still valid.

2. If Commitment 51 is still valid as stated in the letter dated December 17, 2010, explain how these properties (compressive strength, presence of ASR, modulus of elasticity, presence of rebar degradation) can be verified without taking core samples.

3.

Provide details of the plans to monitor the extent of cracking and expansion in concrete.

Justify why it is appropriate to wait until the period of extended operation, in 2030, to verify whether or not ASR is occurring in the containment and to begin trending possible degradation.

NextEra Energy Seabrook Response:

1. Program enhancements and commitment to confirmatory testing cannot be made until the aging effects of ASR are fully understood. Information regarding the planned approach to addressing ASR degradation throughout the site will be included in an engineering evaluation scheduled to complete in March 2012. The content of the evaluation will include: discussion of degradation mechanisms in concrete,

United States Nuclear Regulatory Commission SBK-L-11154/ Enclosure 1 Page 13 of 13 identification of areas susceptible to ASR, results of in-situ testing of concrete and impact on current licensing basis calculations and analyses, progress of lab testing to establish ASR degradation rates in concrete, and mitigation techniques. Specific questions presented in Follow-up. RAI B.2.1.31-1, items 2 through 4, and Follow-up RAI B.2.1.28-3 will be addressed in this evaluation.

Based on this discussion, the following changes are made-to the LRA, as amended by previous correspondence:

1) Program enhancement made in NextEra Seabrook letter SBK-L-10204 (Reference

3) to License Renewal Application Appendix B,.Section B.2.1.28, page B-156 is revised as follows:

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2. Seabrook will pertormf contiiffatery testing and evaluation ef th-e Containmnent Stft~etu~vere oncr-et~e. The testing and evaluation will determfine the eoncrete comp

.Reiv strength, the prcscnee or absence of Alkali Silic-a Reaction (ASR), the concrete modulus of elastieitýy, and the prcscnee or-absence of rcbar degradation. The testing and evaluateion will be completed prior to the period of extended operation.

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2) Commitment No. 51 made in NextEra Seabrook letter SBK-L-10204 (Reference

3) to License Renewal Application Appendix A, Section A.3, page A-43 is revised as follows:

UFSAR SCHEDULE No.

PROGRAM or TOPIC COMMITMENT LOCATION LOCATION Number Not Used PFior4o he 51 Pe*rfor confirmatory, testing and cvaluation of the A212 ARSME S-ectfion-RXI, Contai-nment Structufi-re concrete etne Subsection IWL eper.e

2.

See response to item 1.

3.

Plans to monitor the extent of cracking and expansion in concrete will be included in the engineering evaluation described in item 1.

to SBK-L-11154 Changes to the Seabrook Station License Renewal Application Revised NextEra Energy Seabrook response to RAI 3.3.2.15-1 provided in letter SBK-L-11015 dated February 3, 2011

United States Nuclear Regulatory Commission Page 2 of 3 SBK-L-1 1154/Enclosure 2 The following NextEra Energy Seabrook response replaces the response to RAI 3.3.2.15-1 provided in Reference 5.

In NRC Letter "Request for Additional Information Related to the Review of the Seabrook Station License Renewal Application (TAC NO. ME4028)

Aging Management Review - Set 6" dated January 5, 2011 (ML103420585), there were two RAIs with similar numbering. One was RAI 3.3.2.15-1 and the second one was RAI 3.3.2.15-01. The response to RAI 3.3.2.15-1 was inadvertently duplicated as the same response to RAI 3.3.2.15-01; The correct response to RAI 3.3.2.15-1 is as follows:

Request for Additional Information (RAI) 3.3.2.15-1

Background:

The GALL Report does not contain an aging management review for heat exchangers exposed to steam affected by reduction in heat transfer.

For other materials and environments, the GALL Report typically suggests using a water chemistry program in conjunction with an inspection program for managing the reduction of heat transfer aging effect. In LRA Table 3.3.2-15, the applicant has indicated that reduction of heat transfer is an aging effect relevant to heat exchanger tubes exposed to steam. The applicant has indicated it plans to use the Water Chemistry Program to manage this aging effect.

Issue:

The applicant has indicated in the LRA that it plans to only use a water chemistry program to manage the reduction of heat transfer for heat exchanger tubes exposed to reactor coolant. This appears to be in contrast to the GALL Report typical management of this aging effect. It is not clear to the staff how management of water chemistry alone will ensure that reduction of heat transfer is appropriately managed.

Request:

Justify how the Water Chemistry Program alone is sufficient to determine that heat exchanger tubes are not affected by reduction of heat transfer when exposed to steam.

NextEra Seabrook Station Response The Fire Protection system heat exchanger tubes for FP-E-46 & FP-E-47 have an external environment of steam, which is converted from potable water. PWR secondary plant water chemistry program is not applicable to potable water. Therefore, instead of the Water Chemistry Program, the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program is a more appropriate program for age managing these heat exchangers. Seabrook Station has a preventive maintenance activity already in place to clean and inspect the external surfaces of these heat exchanger tubes. The frequency of this maintenance activity is approximately every 4 years. Cleaning and inspection of the

United States Nuclear Regulatory Commission SBK-L-11154/Enclosure 2 Page 3 of 3 external surfaces of the heat exchanger tubes ensure that reduction of heat transfer is managed. Based on this discussion, the following changes were made to the License Renewal Application.

In Table 3.3.2-15, on page 3.3-304, the 7 th row is revised as follows:

Inspection of Heat Heat Internal Exchanger Transfer Surfaces in Components Stainless Steam Reduction Miscellaneous of Heat Piping and None None G, 8 Steel (External)

Transfer Ducting 47 Tubes)

Pressure Components Boundary Water Chemisti*'y Program Please reference response to RAI 3.3.2.15-01 for additional changes that were made to heat exchanger FP-E-46 and 47 components. Additionally, Note 8 in the above table was added in response to RAI 3.3.2.15-01.

to SBK-L-11154 LRA Appendix A - Final Safety Report Supplement Table A.3 License Renewal'Commitment List

United States Nuclear Regulatory Commission SBK-L-115ý4 / Enclosure 3 A.3 LICENSE RENEWAL COMMITMENT LIST Page 2 of 11 UFSAR No.

PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE Program to be implemented prior to the period of extended operation. Inspection plan to be An inspection plan for Reactor Vessel Internals will be submitted to NRC not later than

1.

PWR Vessel Internals submitted for NRC review and approval.

2 years after receipt of the A.2.1.7 renewed license or not less than 24 months prior to the period of extended operation, whichever comes first.

Closed-Cycle Cooling Enhance the program to include visual inspection for cracking, Prior to the period of extended

2.

Water loss of material and fouling when the in-scope systems are A.2.1.12 operation Sopened for maintenance.

Inspection of Overhead Heavy Enhance the program to monitor general corrosion on the crane Prior to the period of extended

3.

Load and Light Load and trolley structural components and the effects of wear on the A.2.1.13 (Related to Refueling) rails in the rail system.

operation Handling Systems Inspection of Overhead Heavy Prior to the period of extended

4.

Load and Light Load Enhance the program to list additional cranes for monitoring.

A.2.1.13 operation (Related to Refueling)

Handling Systems Enhance the program to include an annual air quality test Prior to the period of extended Compressed Air requirement for the Diesel Generator compressed air sub A.2.1.14 operaton

5.

Monitoring system.

operation Enhance the program to perform visual inspection of Prior to the period of extended penetration seals by a fire protection qualified inspector.

operation.

United States Nuclear Regulatory Commission SBK-L-I 1154 / Enclosure 3 Page 3 of 11 UFSAR No.

PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE Enhance the program to add inspection requirements such as Prior to the period of extended

7.

Fire Protection spalling, and loss of material caused by freeze-thaw, chemical A.2.1.15 operation.

attack, and reaction with aggregates by qualified inspector.

Enhance the program to include the performance of visual Prior to the period of extended

8.

Fire Protection inspection of fire-rated doors by a fire protection qualified A.2.1.15 operation.

inspector.

Enhance the program to include NEPA 25 guidance for "where sprinklers have been in place for 50 years, they shall be Prior to the period of extended

9.

Fire Water System replaced or representative samples from one or more sample A.2.1.16 peration.

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

Enhance the program to include the performance of periodic

10.

Fire Water System flow testing of the fire water system in accordance with the A.2.1.16 Prior to the period of extended guidance of NFPA 25.

operation.

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 if a representative number of Within ten years prior to the

11.

Fire Water System inspections have been performed prior to the period of A.2.1.16 Witin tenyea pritoth extended operation. If a representative number of inspections period of extended operation.

have not been 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.

Enhance the program to include components and aging effects

12.

TanksbSteel required by the Aboveground Steel Tanks.

A.2.1.17 Prior to the period of extended Tanksoperation.

13.

Aboveground Steel Enhance the program to include an ultrasonic inspection and Within ten years prior to the

13.

TAnksegroun Steel evaluation of the internal bottom surface of the two Fire A.2.1.17 perio tended prion.

TanksProtection Water Storage Tanks.extended operation.

United States Nuclear Regulatory Commission SBK-L-1 1154 / Enclosure 3 Page 4 of I 1 UFSAR No.

PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE Enhance program to add requirements to 1) sample and

14.

Fuel Oil Chemistry analyze new fuel deliveries for biodiesel prior to offloading to A.2.1.18 Prior to the period of extended the Auxiliary Boiler fuel oil storage tank and 2) periodically operation.

sample stored fuel in the Auxiliary Boiler fuel oil storage tank.

Enhance the program to add requirements to check for the

15.

Fuel Oil Chemistry presence of water in the Auxiliary Boiler fuel oil storage tank at A.2.1.18 operaton.

least once per quarter and to remove water as necessary.

Enhance the program to require draining, cleaning and Prior to the period of extended

16.

Fuel Oil Chemistry inspection of the diesel fire pump fuel oil day tanks on a A.2.1.18 operation.

frequency of at least once every ten years.

Enhance the program to require ultrasonic thickness measurement of the tank bottom during the 10-year draining, Prior to the period of extended

17.

Fuel Oil Chemistry cleaning and inspection of the Diesel Generator fuel oil storage A.2.1.18 operation.

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

Reactor Vessel Enhance the program to specify that all pulled and tested Prior to the period of extended

18.

Sureillance capsules, unless discarded before August 31, 2000, are placed A.2.1.19 operation.

Surveillance in storage.

oeain Enhance the program to specify that if plant operations exceed the limitations or bounds defined by the Reactor Vessel

19.

Reactor Vessel Surveillance Program, such as operating at a lower cold leg A.2.1.19 Prior to the period of extended Surveillance temperature or higher fluence, the impact of plant operation operation.

changes on the extent of Reactor Vessel embrittlement will be evaluated and the NRC will be notified.

United States Nuclear Regulatory Commission SBK-L-11154 / Enclosure 3 Page 5 of 1 I UFSAR No.

PROGRAM or TOPIC COMMITMENT 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 an outage in which the

20.

Reactor Vessel capsule receives a neutron fluence that meets the schedule A.2.1.19 Prior to the period of extended Surveillance requirements of 10 CFR 50 Appendix H and ASTM E185-82 operation.

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 removed,

21.

Reactor Vessel without the intent to test it, is stored in a manner which A.2.1.19 Prior to the period of extended Surveillance maintains it in a condition which would permit its future use, operation.

including during the period of extended operation.

Within ten years prior to the

22.

One-Time Inspection Implement the One Time Inspection Program.

A.2.1.20 perio tended prio n.

period of extended operation.

Implement the Selective Leaching of Materials Program. The Selective Leaching of program will include a one-time inspection of selected Within five years prior to the

23.

Materials components where selective leaching has not been identified A.2.1.21 period of extended operation.

and periodic inspections of selected components where selective leaching has been identified.

Within ten years prior to

24.

Buried Piping And Implement the Buried Piping And Tanks Inspection Program.

A.2.1.22 entering the period of extended Tanks Inspection operation One-Time Inspection Implement the One-Time Inspection of ASME Code Class 1 Within ten years prior to the

25.

of ASME Code Class Imlmn h n-ieIseto fAM oeCas1A.2.1

.23

25. 1 Sma Bore-Pipin Small Bore-Piping Program.

period of extended operation.

1 Small Bore-Piping Enhance-the program to specifically address the scope of the program, relevant degradation mechanisms and effects of

26.

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

insulation, the training requirements for inspectors and the required periodic reviews to determine program effectiveness.

United States Nuclear Regulatory Commission SBK-L-1 1154 / Enclosure 3 Page 6 of I1 UFSAR No.

PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE Inspection of Internal Implement the Inspection of Internal Surfaces in Miscellaneous Prior to the period of extended

27.

Miscellaneous Piping Piping and Ducting Components Program.

A.2.1.25 and Ducting operation.

Components Lubricating Oil Enhance the program to add required equipment, lube oil Prior to the period of extended

28.

Ana tis analysis required, sampling frequency, and periodic oil A.2.1.26 operation.

Analysis changes.

29.

Lubricating Oil Enhance the program to sample the oil for the Switchyard SF6 A.2.1.26 Prior to the period of extended Analysis compressors and the Reactor Coolant pump oil collection tanks.

operation.

Enhance the program to require the performance of a one-time

30.

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

extended operation.

ASME Section Xl, Enhance procedure to include the definition of "Responsible Prior to the period of extended Subsection IWL Engineer".

operation.

Structures Monitoring Enhance procedure to add the aging effects, additional Prior to the period of extended

32.

locations, inspection frequency and ultrasonic test A.2.1.31 operation.

Program requirements.

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

Program_____________

concrete.

United States Nuclear Regulatory Commission SBK-L-11154 / Enclosure 3 Page 7 of 11 UFSAR No.

PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE Electrical Cables and Connections Not Subjections tot 1Implement the Electrical Cables and Connections Not Subject 50.49 Environmental to 10 CFR 50.49 Environmental Qualification Requirements A.2.1.32 operation.

Qualification program.

Requirements Electrical Cables and Connections Not Subject to 10 CFR Implement the Electrical Cables and Connections Not Subject 5.50.49 Environmental peetteEetiaCalsadCneiosNtSbctPrior to the period of extended Qualification to 10 CFR 50.49 Environmental Qualification Requirements A.2.1.33 operaton.

Requirements Used Used in Instrumentation Circuits program.

operation.

in Instrumentation Circuits Inaccessible Power Cables Not Subject to 10 CFR 50.49 Implement the Inaccessible Power Cables Not Subject to 10 Prior to the period of extended 36.A2134 oeain Environmental CFR 50.49 Environmental Qualification Requirements program.

operation.

Qualification Requirements

37.

Metal Enclosed Bus Implement the Metal Enclosed Bus program.

A.2.1.35 Prior to the period of extended operation.

Prior to the period of extended

38.

Fuse Holders Implement the Fuse Holders program.

A.2.1.36 operaton.

operation.

Electrical Cable Connections Not

39.

Subject to 10 CFR Implement the Electrical Cable Connections Not Subject to 10 A.2.1.37 Prior to the period of extended 50.49 Environmental CFR 50.49 Environmental Qualification Requirements program.

operation.

Qualification Requirements Prior to the period of extended

40.

345 KV SF6 Bus Implement the 345 KV SF6 Bus program.

A.2.2.1 operaton.

operation.

United States Nuclear Regulatory SBK-L-1 1154 / Enclosure 3 Commission Page 8 of I I UFSAR No.

PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE Metal Fatigue of Enhance the program to include additional transients beyond Prior to the period of extended

41.

Reactor Coolant A.2.3.1 PRessure Bound those defined in the Technical Specifications and UFSAR.

operation.

Pressure Boundary Metal Fatigue of Enhance the program to implement a software program, to Prior to the period of extended

42.

Reactor Coolant count transients to monitor cumulative usage on selected A.2.3.1 operation.

Pressure Boundary components.

Pressure -

The updated analyses will be Temperature Limits, Seabrook Station will submit updates to the P-T curves and submitted at the appropriate

43.

including Low LTOP limits to the NRC at the appropriate time to comply with A.2.4.1.4 time to comply with 10 CFR 50 Overpressure 10 CFR 50 Appendix G.

Appendix G, Fracture Protection Limits Toughness Requirements.

NextEra Seabrook will perform a review of design basis ASME Class 1 component fatigue evaluations to determine whether the NUREG/CR-6260-based components that have been evaluated for the effects of the reactor coolant environment on fatigue usage are the limiting components for the Seabrook plant configuration. If more limiting components are identified, the most limiting component will be evaluated for the effects of the reactor coolant environment on fatigue usage. If the limiting location identified consists of nickel alloy, the environmentally-assisted fatigue calculation for nickel alloy will be performed Environmentally-using the rules of NUREG/CR-6909.

At least two years prior to

44.

Assisted Fatigue (1) Consistent with the Metal Fatigue of Reactor Coolant A.2.4.2.3 entering the period of extended Pressure Boundary Program Seabrook Station will update the Operation.

Analyses (TLAA) 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 valid for the period of extended operation or from an analysis using an NRC-approved version of the ASME code or NRC-approved alternative (e.g., NRC-approved code case).

(2) If acceptable CUFs cannot be demonstrated for all the selected locations, then additional plant-specific locations will be evaluated. For the additional plant-specific locations, if CUF,

United States Nuclear Regulatory Commission SBK-L-1 1154 / Enclosure 3 Page 9 of 11 UFSAR No.

PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE 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 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).

45.

Number Not Used Protective Coating Enhance the program by designating and qualifying an Prior to the period of extended

46.

Monitoring and Inspector Coordinator and an Inspection Results Evaluator.

A.2.1.38 operation Maintenance Enhance the program by including, "Instruments and Equipment Protective Coating needed for inspection may include, but not be limited to, Prior to the period of etended

47.

Monitoring and flashlight, spotlights, marker pen, mirror, measuring tape, A.2.1.38 operation Maintenance magnifier, binoculars, camera with or without wide angle lens, and self sealing polyethylene sample bags."

Protective Coating Prior to the period of extended

48.

Monitoring and Enhance the program to include a review of the previous t A.2.1.38 operation Maintenance monitoring reports.

Protective Coating Enhance the program to require that the inspection report is to Prior to the period of extended

49.

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

hNo l;ter than December 31, 2015 Within the next two refueling outages, 0R15 or ASME Section XI, Perform UT testing of the containment liner plate in the vicinity A.2.1.27 0R16, and repeated at intervals Subsection IWE of the moisture barrier for loss of material.

of nd re t

ivervals of no more than five refueling outages

United States Nuclear Regulatory Commission SBK-L-I 1154 / Enclosure 3 Page 10 of I I UFSAR No.

PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE Number Not Used PerfoFrm confirmatory teSting and evaluation of the Containment Prior to the period of extene

51.

ASME Section I Structure concret-e X.28 opeFation

52.

ASME Section XI Containment Structure, from elevation -30 feet to +20 feet, in a A.2.1.28 By 2013 Subsection IWL dewatered state.

Reactor Head Replace the spare reactor head closure stud(s) manufactured Prior to the period of extended 53.Cosure Studs from the bar that has a yield strength > 150 ksi with ones that A.2.1.3 Operation.

Closure Studsdo not exceed 150 ksi.

oeain Unless an alternate repair criteria changing the ASME code boundary is permanently approved by the NRC, or the Program to be submitted to Steam Generator Seabrook Station steam generators are changed to eliminate NRC at least 24 months prior to 54m Tube Integrity PWSCC-susceptible tube-to-tubesheet welds, submit a plant-A.2. 1.10 the period of extended specific aging management program to manage the potential aging effect of cracking due to PWSCC at least twenty-four months prior to entering the Period of Extended Operation.

55.

Steam Generator Seabrook will perform an inspection of each steam generator to A.2. 1.10 Prior to entering the period of Tube Integrity assess the condition of the divider plate assembly.

extended operation Closed-Cycle Cooling Revise the station program documents to reflectthe EPRI Prior to entering the period of

56.

Water System Guideline operating ranges and Action Level values for A.2.1.12 extended operation.

hydrazine and sulfates.

Closed-Cycle Cooling Revise the station program documents to reflect the EPRI Prior to entering the period of

57.

Guideline operating ranges and Action Level values for Diesel A.2.1.12 extended operation.

Generator Cooling Water Jacket pH.

Update Technical Requirement Program 5.1, (Diesel Fuel Oil Prior to the period of extended

58.

Fuel Oil Chemistry Testing Program) ASTM standards to ASTM D2709-96 and A.2.1.18 operation.

ASTM D4057-95 required by the GALL XI.M30 Rev 1 The Nickel Alloy Aging Nozzles and Penetrations program will Prior to the period of extended 59.Nickel Alloy Nozzles implement applicable Bulletins, Generic Letters, and staff A.2.2.3 operaton.

.and Penetrations accepted industry guidelines.

Buried Implement the design change replacing the buried Auxiliary Prior to entering the period of

60.

Tanks Piping and Boiler supply piping with a pipe-within-pipe configuration with A.2.1.22 extended operation.

sInspection leak indication capability.

Within ten years prior to

61.

Compressed Air Replace the flexible hoses associated with the Diesel Generator A.2.1.14 entering the period of extended Monitoring Program air compressors on a frequency of every 10 years.

operation.

United States Nuclear Regulatory Commission SBK-L-1 1154 / Enclosure 3 Page I Iof 11 UFSAR SHDL No.

PROGRAM or TOPIC COMMITMENT LOCATION SCHEDULE Enhance the program to include a statement that sampling Prior to entering the period of

62.

Water Chemistry frequencies are increased when chemistry action levels are A.2.1.2 extended operation.

exceeded.

Ensure that the quarterly CVCS Charging Pump testing is continued during the PEO. Additionally, add a precaution to the

63.

Flow Induced Erosion test procedure to state that an increase in the CVCS Charging N/A Prior to the period of extended Pump mini flow above the acceptance criteria may be indicative operation of erosion of the mini flow orifice as described in LER 50-275/94-023.

Soil analysis shall be performed prior to entering the period of extended operation to determine the corrosivity of the soil in the

.64.

Buried Piping and vicinity of non-cathodically protected steel pipe within the scope A.2.1.22 Prior to entering the period of of this program. If the initial analysis shows the soil to be non-extended operation.

Tanks Inspection corrosive, this analysis will be re-performed every ten years thereafter.

Implement measures to ensure that the movable incore Prior to entering the period of

65.

Flux Thimble Tube detectors are not returned to service during the period of N/A extended operation extended operation.

Enhance the current station operating experience review process implemented in response to NUREG 0737 Task I.C.5-Procedures for Feedback of Operating Experience to Plant Staff (UFSAR L11.9.1) to include future reviews of plant-specific and Within ten years prior to

66.

Operating Experience industry operating experience in order to confirm the N/A entering the period of extended effectiveness of the license renewal aging management operation.

programs and to determine the need for programs to be enhanced or the need to develop new aging management programs.

Perform one shallow core bore in an area that was

67.

Structures continuously wetted from borated water to be examined for A.2.1.31 No later than December 31, Monitoring Program concrete degradation and also expose rebar to detect any 2015 degradation such as loss of material.

Perform sampling at the spent fuel pool leakoff collection

68.

Structures point for chlorides, sulfates, pH and iron for four quarters A.2.1.31 Starting January 2014 Monitoring Program of one year once every 5 years.