SBK-L-11063, Response to Request for Additional Information NextEra Energy Seabrook License Renewal Application Request for Additional Information - Set 13
| ML11108A131 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 04/14/2011 |
| From: | Freeman P NextEra Energy Seabrook |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| SBK-L-11063 | |
| Download: ML11108A131 (34) | |
Text
NExTera EN ERGYdM April 14, 2011 SBK-L-1 1063 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 13
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 Related to the Review of the Seabrook Station License Renewal Application (TAC NO. ME4028) - Request for Additional Information Set 13," March 17, 2011. (Accession Number MLI 10350630)
- 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)
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, 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.
Based on discussion with the Staff, NextEra Energy Seabrook has made changes to License Renewal Application, which are contained in Enclosure 2 of this letter.
NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874
United States Nuclear Regulatory Commission SBK-L-1 1063 / Page 2 Commitment numbers 50 and 52 are revised.
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, Licensing Manager, at (603) 773-7745.
please contact Mr. Michael O'Keefe, Sincerely, NextEra Energy Seabrook, LLC.
'6 0o,
ý Paul 0. Freeman Site Vice President
Enclosures:
Response to Request for Additional Information Seabrook Station License Renewal Application, Set # 13 and Associated LRA Changes Changes to License Renewal Application based on NRC Staff discussions 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-1 1063 / 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-11063 / Page 4 NEXTera ENERC-G7YI 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 9"" day of 2011
- Y&f Paul 0. Freeman Site Vice President "I&
111OayvPublic 7
SBK-L-1 1063 Seabrook Station Response to Request for Additional Information NextEra Energy Seabrook License Renewal Application Request for Additional Information - Set 13 Hamrick, S.
e-mail Ross, M.
e-mail Mashhadi, M.
e-mail Dryden, M. S.
e-mail Brown, A.
e-mail Cliche, R.
e-mail Dunn, B.
e-mail Carley, E.
e-mail Collins, M.
e-mail Metcalf, E.
e-mail Noble, R e-mail Letter Distribution e-mail File 0018 GLC RMD ORM to SBK-L-11063 Response to Request for Additional Information Seabrook Station License Renewal Application Set 13 and Associated LRA Changes
United States Nuclear Regulatory Commission Page 2 of 9 SBK-L-1 1063 / Enclosure 1 Request for Additional Information (RAI) Follow-up B.2.1.27 -1
Background:
By letter dated December 17, 2010, the applicant responded to RAI B.2.1.27-1 and stated that Seabrook will perform testing of the containment liner plate for loss of material on the concrete side of the liner. The testing will be conducted in accordance with approved ASME Section XI, Subsection IWE methodology, and will be completed prior to the period of extended operation.
Issue:
The applicant has committed to performing testing of the containment liner plate for the loss of material on the side of the concrete; however, it is not clear how this testing will be performed.
Request:
Provide details regarding the testing to be performed to determine the loss of material on the concrete side of the liner plate. Include a description of the nondestructive testing methods and locations where thickness measurements will be obtained, and explain why the measurement locations will provide an adequate representation of liner plate locations that may be degraded.
NextEra Energy Seabrook Response:
Seabrook will perform ultrasonic thickness (UT) testing of the liner plate inside containment for loss of material on the concrete side of the liner. The testing will be subject to ASME Section XI, Subsection IWE acceptance criteria (the code currently in use at Seabrook Station is the 2004 ASME code), and will be completed by December 31, 2015.
Under IWE 1241(a), Seabrook will designate the area of the containment liner that is within ten inches of the moisture barrier at the containment basement floor for examination. This is the lowest accessible point on the liner. For potential degradation due to moisture, the lowest point is the most susceptible. Locations spaced at 100 increments (approximately every 12 feet) of accessible circumference, or locations showing visible signs of degradation, will be tested with one or more readings taken at each location. The examination will be repeated at intervals of no more than five refueling outages.
If any indications are found that show a loss of material exceeding 10% of nominal thickness (ASME Code acceptance criteria), an Engineering Evaluation of deficient indications will be performed and actions planned accordingly. The following changes are made to the LRA:
United States Nuclear Regulatory Commission Page 3 of 9 SBK-L-11063 / Enclosure 1
- 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-10204, is further changed to read as follows:
Enhancements Seabrook will perform ultrasonic thickness (UT) testing of the co..ntai..en liner plate inside containment for loss of material on the concrete side of the liner. The testing will be subject to conducted in accordance with approved ASME Section XI, Subsection IWE acceptance criteria. methodology, and will be ompicte prier to the period of extended operatin,. The UT testing targets the area near the moisture barrier at el. -26 and at nominal 100 increments around the accessible circumference of containment. This will be completed by December 31, 2015 and at intervals of no more than five refueling outages thereafter.
- 2) License Renewal Application Appendix A, Section A.3, page A-43, as changed by RAI B2.1.27-1 in SBK-L-10204, is further changed and added to, as follows:
ASME Perform UTtesting of the containment A.2.1.-1-27 Prior to the peri*d et Section XI, liner plate in the vicinity of the moisture extended eperatin..
Subsection barrier for loss of material.
No later titan IW
- 50.
1 IWE December 31, 2015 and repeated at intervals of no more than five refueling outages Request for Additional Information (RAI) Follow-up B.2.1.27 -2
Background:
By letter dated December 17, 2010, the applicant responded to RAI B.2.1.27-2 and stated that the liner plate around the fuel transfer tube has been identified in the ISI program for augmented inspection in accordance with the 1995 Edition with 1996 Addenda of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)
Code,Section XI, Subsection IWE-2420(b) and (c).
Issue:
The ASME 1995 Edition with 1996 Addenda,Section XI, Subsection IWE-2420(b) and (c) states that reexamination of degraded areas is no longer required if these. areas remains essentially unchanged for three consecutive inspection periods. However, it is not clear from the applicant's response if the containment liner plate around the fuel transfer tube is still exposed to the borated water leakage. Exposure to borated water can promote corrosion of the liner plate and adversely affect the ability of the liner to perform its intended function.
United States Nuclear Regulatory Commission Page 4 of 9 SBK-L-11063 / Enclosure 1 Request:
Describe steps that are being taken to monitor the liner plate thickness around the transfer tube and/or efforts to address the leakage of borated water.
NextEra Enermy Seabrook Response:
The leak path into the fuel transfer tube vault has been repaired and the borated water leakage stopped. The areas of the containment liner plate that had showed signs of deficiency (loss of material) have been examined and accepted. The areas are subject to IWE required augmented UT examinations for the next three exam cycles. If no further degradation (loss of material) is observed during those three cycles, the subject area will return to normal visual IWE inspections. These visual inspections would be able to identify any further leakage of borated water.
Request for Additional Information (RAI) Follow-up B.2.1.31-1
Background:
By letter dated December 17, 2010, the applicant responded to RAI B.2.1.31-1 regarding concrete degradation due to groundwater in-leakage and explained that recent cores had shown significant reductions in concrete compressive strength and modulus of elasticity.
The applicant stated that a prompt operability determination concluded the affected areas were in compliance with the design code and that an extent of condition investigation was ongoing. The applicant further stated that any necessary future remediation will be identified and conducted through the corrective action program.
Issue:
The response lacked information regarding the extent of condition assessment including approximate completion dates and probable path forward.
Request:
Provide additional information regarding the extent of the condition investigation, including the following:
- 1. Any additional tests planned or results of investigations conducted since the initial RAI response was submitted.
- 2. An estimated timeframe for the extent of condition investigation.
- 3. A proposed path forward, including the location and timing of future tests as well as proposed remedial actions based on available information.
United States Nuclear Regulatory Commission Page 5 of 9 SBK-L-1 1063 / Enclosure 1
- 4. How the investigation / path forward will ensure the adequacy of the concrete during the period of extended operation.
NextEra Energy Seabrook Response:
- 1. With respect to additional testing or examinations completed since the initial RAI response was submitted, Seabrook is completing the extent of condition assessment in support of the prompt operability determination. This consisted of removing core bores for testing from five additional suspect building location areas. Selection of locations for cores was based on below grade areas that exhibited groundwater inleakage and surface cracking indicative of ASR being present. Sampling of cores was based on American Concrete Institute (ACI) standard ACI 228.1R-03 "In-Place Methods to Estimate Concrete Strength" The samples are being prepared for shipment to a lab for analysis. Testing will determine compressive strength (ASTM C 42-04 "Standard Test Method for Obtaining and Testing Drilled Cores and Sawed Beams of Concrete"), modulus of elasticity (ASTM C 469-02 "Standard Test Method for Static Modulus of Elasticity and Poisson's Ratio of Concrete in Compression")
and examination for the presence of Alkali Silica Reaction (ASTM C 856-04 "Petrographic Examination of Hardened Concrete"). Testing of concrete cores for tensile properties are not performed as tensile properties of concrete are not used in the design of concrete structures at Seabrook Station.
Test results are used as input to reconcile existing calculations and analyses to ensure that concrete structures satisfy all design basis conditions (i.e. deadweight, wind, seismic, etc). Results of the testing are expected in May 2011.
- 2. The extent of condition assessment to support the prompt operability determination is scheduled to complete in June 2011.
The comprehensive long term extent of condition assessment is detailed in the action plan discussed in response number 3 below.
- 3. Utilizing the knowledge gained from testing completed to date, an action plan has been developed to
- Identify other areas at Seabrook that are potentially susceptible to ASR
- Complete testing of concrete in other susceptible areas. Testing will determine compressive strength (ASTM C 42-04 "Standard Test Method for Obtaining and Testing Drilled Cores and Sawed Beams of Concrete"), modulus of elasticity (ASTM C 469-02 "Standard Test Method for Static Modulus of Elasticity and Poisson's Ratio of Concrete in Compression"), and examination for the presence of Alkaline Silica Reaction (ASTM C 856-04 "Petrographic Examination of Hardened Concrete").
Based on test results, reconcile existing calculations and analyses to ensure that concrete structures satisfy all design basis conditions (i.e. deadweight, wind, seismic, etc).
United States Nuclear Regulatory Commission Page 6 of 9 SBK-L-1 1063 / Enclosure 1 Perform Lab tests with in-situ concrete material to determine how ASR degradation mechanism propagates. Tests to be completed per ASTM and other appropriate standards include:
Alkali reactivity tests of coarse aggregates per:
ASTM C 1260 "Mortar Bar Expansion Test" - Short duration testing (16 Days)
ASTM C 1293 "Concrete Prism" - Long duration testing (1-2 years)
Other tests subjecting cores to an accelerated aging process to establish a rate of ASR degradation Issue Engineering Evaluation "Alkali-Silica Reaction of Concrete at Seabrook Station", tentatively scheduled for March 2012. Content of evaluation includes:
Technical discussion on ASR degradation mechanism in concrete Identification of areas susceptible to ASR Results to date of in-situ testing of concrete and impact on Current Licensing Basis (CLB) calculations and analyses Results of Lab testing to establish ASR degradation rate in concrete.
Mitigation Techniques o Based on test results of concrete samples and inspection of plant areas, update the Structures Monitoring Program (SMP) to include:
Type of monitoring required to detect ASR degradation such as ASR indicative crack pattern, surface acid etching, and presence of water.
The frequency of monitoring areas impacted by ASR, utilizing the multi-tiered process as described in the SMP for evaluating deficiencies.
- Develop a long range plan to implement mitigation measures to arrest degradation attributed to ASR. Utilizing rate for progression of ASR concrete degradation, prioritize areas to be remediated. Develop mitigation techniques to divert groundwater from the below grade structures utilizing industry input on waterproofing technology and insights gained from the new groundwater fate and transport study (the study of groundwater distribution and movement) completed for the Seabrook site.
Implementation of the action plan is scheduled to be completed in December 2013
- 4. The Seabrook Structural Monitoring Procedure will be revised to include actions for inspection and monitoring of concrete due to ASR. The Extent of Condition review will provide the plant staff with the scope of the effects of ASR on concrete structural elements at Seabrook. Actions will be taken on the basis of the extent of condition to keep the structures within the limits of the current design bases. The Seabrook Structural Monitoring Procedure will be revised to include direction on monitoring the presence of ASR.
United States Nuclear Regulatory Commission Page 7 of 9 SBK-L-1 1063 / Enclosure I Request for Additional Information (RAI) Follow-up B.2.1.31-2
Background:
By letter dated December 17, 2010, the applicant responded to RAI B.2.2.31-2 and explained that components affected by groundwater in-leakage are managed under the Structures Monitoring Program which implements the Structural Engineering Standard Technical Procedure issued in March 2010. The program covers "building structural steel" and instructs the inspectors to look for degradation such as corrosion, peeling paint, excessive deflection of members, etc.
Issue:
Although the procedure was updated in March 2010, the staff noted several areas of degradation due to in-leakage during walkdowns in October 2010. The staff needs more information on how this will be addressed during the period of extended operation.
Request:
Explain what actions will be taken when degradation is noted in areas prone to in-leakage and whether or not additional actions are taken to monitor these areas (e.g., more frequent inspections).
NextEra Energy Seabrook Response:
The Engineering Department Standard (EDS), 'Structural Monitoring Program',
delineates the activities for Periodic Inspections and Deficiency Follow-up Inspections, specifies three tiers of acceptance criteria, and directs actions and enhanced inspections to be taken in response to deficiency identification.
Periodic Inspections are conducted on Harsh environment structures once every five years. New deficiencies identified in structural steel during a Periodic Inspection are documented on the inspection report and evaluated per the three tier acceptance criteria as described in the Engineering Department Standard (EDS), 'Structural Monitoring Program'. Deficiencies that do not meet the acceptance criteria are entered into the Corrective Action Program. Corrective Actions are entered in the Work Control Program for implementation. Deficiencies determined to be acceptable by engineering review are trended for evidence of further degradation.
Deficiencies being repaired or trended are subject to Deficiency Follow-up Inspections (DFI) at a minimum every 21/2 years. These may be Increased Frequency DFIs, Routine Frequency DFIs, or Deficiency Repair DFIs.
The Seabrook Structural Monitoring Procedure has been revised to include specific direction on monitoring for the presence of water in-leakage.
United States Nuclear Regulatory Commission Page 8 of 9 SBK-L-11063 / Enclosure 1 Request for Additional Information (RAI) Follow-up B.2.1.31-4
Background:
By letter dated December 17, 2010, the applicant responded to RAI B.2.1.31-4 and explained that spent fuel pool leakage has migrated through the surrounding concrete in the past. The applicant further stated that the leakage was stopped in 2004 after the application of a nonmetallic liner to the spent fuel pool.
Issue:
The applicant did not provide adequate justification for its conclusion that the leakage has stopped and that no through-wall leakage is occurring. In addition, based on industry operating experience with failures of spent fuel pool nonmetallic coatings, the staff is not confident that the nonmetallic liner is an appropriate long-term fix.
Request:
- 1. Discuss what measures will be taken to ensure the adequacy of the concrete and rebars exposed to SFP leakage, including the possibility of core bores from known leakage locations prior to or during the period of extended operation.
- 2. Explain how the conclusion was reached that through-wall leakage is not occurring, especially in inaccessible areas. Include a discussion of any additional inspections that will be conducted during the period of extended operation to verify that leakage is not occurring.
- 3. If the nonmetallic liner is relied upon to stop leakage, explain what measures will be taken to ensure the adequacy of the liner during the period of extended operation.
NextEra Energy Seabrook Response:
- 1. Seabrook will perform a core-bore test for compression strength of the concrete and also expose rebar to detect any degradation such as loss of material. This will take place in an area subjected to wetting during the time frame of the leakage.
Examination, testing, and evaluation will be complete no later than December 31, 2015. An action request (AR) has been created to track and control the core bore and rebar condition assessment.
- 2. The Root Cause Investigation of the Spent Fuel Pool leak concluded that the leakage was from the transfer canal and cask handling areas. These two areas share water with the Spent Fuel Pool but can be isolated from it. The transfer canal and the cask handling area were isolated and pumped down to the bottom of the transfer canal and the leakage stopped, based on the leak off system discharge. The liner in these two areas was visually inspected and no evidence of a through wall opening was observed.
United States Nuclear Regulatory Commission Page9of9 SBK-L-11063 / Enclosure I areas was visually inspected and no evidence of a through wall opening was observed.
The leak off system is routinely hydro-lazed to ensure that it is free-flowing. It is the path of least resistance for any water that may be between the liner plate and the concrete wall. Any leakage from the transfer canal and cask handling area (or the Spent Fuel Pool) will drain to the leak collection sump. By procedure, the leak collection sump is periodically sampled and tested for chemical composition -
evidence of boron or tritium content.
- 3. The non-metallic lining that seals the stainless steel pool liner is replaced on the basis of condition monitoring that will identify when the end of life for the material has been reached or degradation is observed. This is consistent with NUREG-1801 for the treatment of consumables.
The following activities will be maintained during the Period of Extended Operation:
- a. Continued monitoring of the liner coating coupon system under the PM "Visual Inspection of Coupons Coated with SEFR".
- b. Continued sampling and analysis of leak system effluent under procedure "Spent Fuel Pool (SFP) Leakage Collection Program".
to SBK-L-11063 Changes to the Seabrook Station License Renewal Application Associated with NRC Staff Discussions
United States Nuclear Regulatory Commission SBK-L-1 1063/ Enclosure 2 Changes to Section 3.3 and 3.4 of the LRA Page 2 of 7 In the LRA, the material for the flame arrestor installed in the Auxiliary Boiler system should have been listed as aluminum instead of steel. The material discrepancy was identified during the Seabrook Environment and Material Sample Audit conducted during the week of September 20, 2010. As a result, the following change was made to the LRA.
- 1) In Table 3.3.2-1, on page 3.3-131, the 6th and 7th rows are revised as follows:
- 2) In Table 3.4.1, on page 3.4-31, item number 3.4.1-30 is revised as follows:
3.4. 1-30 Steel piping, piping components, and piping elements exposed to air-outdoor (internal) or condensation (internal)
Loss of material due to general, pitting, and crevice corrosion Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components No Components in the Auxi4iaDy Be4ei-, Control Building Air Handling, and Service Water systems have aligned with this line item based on material, environment, and aging effect.
Consistent with NUREG-1801 with exceptions. The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Program (with exceptions), B.2.1.25, will be used to manage loss of material due to general, pitting, and crevice corrosion in steel piping components exposed to air outdoor (internal)in the Auxiliar=- Beiler, Control Building Air Handling, Main Steam, and Service Water systems, and steel fan housing exposed to air-outdoor (internal)in the Control Building Air Handling System, and steel piping components exposed to condensation (internal) in the Main Steam system.
United States Nuclear Regulatory Commission Page 3 of 7 SBK-L-1 1063/ Enclosure 2 Changes to Table 3.5.2.4 of the LRA The Structures Monitoring Program utilizes visual inspection to detect deterioration of elastomers, non-metallic fire proofing, aluminum, etc. Specific inspection methods and frequency of these inspections performed for these materials used in roofing of primary structures were clarified in discussion with the Staff on April 7, 2011 during the IP71002 Inspection. These clarifications are documented as follows:
- 1. The elastomeric roofs are walked down and inspected by a contract roofing company (Mayo Roofing Company) that specializes in roofing and has specific industry expertise. The physical walkdown (walking on the roof) and visual inspection will help in determining the condition (separation, environmental degradation and water in.
leakage due to weathering) of the roof.
Aluminum will be visually inspected like all metallic materials (using the material aging effects).
Non-metallic fire proofing, which is a sprayed on cementitious material, is examined the same as concrete.
- 2. The frequency of inspection is based on environment (Harsh - 5 year inspection and Mild -10 year inspection) and operating experience that may increase inspection frequency.
There are no inspector qualifications for the roof inspectors, because the inspectors are contract roofers who have specific industry expertise in elastomeric roofs.
The inspector qualifications for the metallic and concrete inspectors are specified in the Structures Monitoring Program:
"Individuals conducting the inspection and reviewing the results are to possess expertise in the design and inspection of steel, concrete and masonry structures.
This individual is to be either a licensed Professional Engineer experienced in this area, or shall be working under the direction of a licensed Professional Engineer experienced in this area."
- 3. The acceptance criteria will be based on the Engineering Department Standard for Structural Monitoring Program which describes the aging effects and evaluation criteria based on ACI 349 three-tiered hierarchy and quantitative limits.
Upon review, Seabrook Station found that Table 3.5.2.4 in LRA Chapter 3 inadvertently included an Aging Effect Requiring Management of Crack Initiation and Growth for Component Type MYS-Aluminum STATION BLACKOUT STRUCTURES Exposed to Weather. The Aging Effect should be Cracking. The following change has been made to the LRA:
United States Nuclear Regulatory Commission Page 4 of 7 SBK-L-1 1063/ Enclosure 2
- 1) LRA Chapter 3, Table 3.5.2.4, page 3.5-138 is changed to read as follows:
MYS - Aluminum G*aek Structures Struutcturesat~n lI.B2-7 35l5 STATION BLACKOUT Structural Aluminum (Exteora iltate.13-67 STRUCTURES Exposed to Support Air (External) aMd GreA;h Monitoring (Tp-6)
Weather Cracking Program Changes to Appendix B, Section B.2.1.16 of the LRA During the IP71002 Inspection of March 8-25, 2011, clarification was provided to the Staff on the applicable NFPA 25 edition and fire hose house inspection intervals for the Fire Water System Program. These clarifications are documented as follows:
The LRA Appendix B, Section B.2.1.16, references to NFPA 25 is changed to read, "in accordance with the guidance referenced in NFPA 25, 2002 Edition".
- 1) License Renewal Application, Appendix B, Section B.2.1.16, last paragraph on page B-99 is changed to read as follows:
Procedures are established to test and inspect fire protection piping and components for indications of degradation. The Fire Water System Program will be enhanced to perform periodic flow testing of the fire water system in accordance with the guidance referenced in National Fire Protection Association (NFPA) 25, 2002 Edition guidelines.
- 2) License Renewal Application, Appendix B, Section B.2.1.16, section titled "Enhancements" on pages B-100 and B-101 is changed to read as follows:
Enhancements The following enhancements will be made prior to entering the period of extended operation.
- 1. The Seabrook Station Fire Water System Program will be enhanced to include the guidance referenced in National Fire Protection Association (NFPA) 25, 2002 Edition, for "'where sprinklers have been in place for 50 years, they will be replaced or representative samples fr'om one or more sample areas will be submitted to a recognized testing laboratory for field service testing". This sampling will be performed every 10 years after the initial field service testing to ensure that signs of degradation, such as corrosion, are detected in a timely manner.
Program Elements Affected. Element 4 (Detection ofAging Effects)
United States Nuclear Regulatory Commission Page 5 of 7 SBK-L-1 1063/ Enclosure 2
- 2. The Seabrook Station Fire Water System Program will be enhanced to include the performance of periodic flow testing of the fire water system in accordance with the guidance referenced in NFPA 25, 2002 Edition The LRA Appendix B, Section B,2.1.16, clarification is made to fire hose house inspection intervals and hydrant inspection acceptance criteria.
- 1) License Renewal Application, Appendix B, Section B.2.1.16, paragraph four through six on page B-100 are changed to read as follows:
Seabrook Station procedures require the performance of a visual inspection of fire hose houses, an inspection to ensure required equipment is present at each hose house, a hydrant inspection and operability test, a fire hydrant hose hydrostatic tests, and a hose replacement and gasket inspection and r.pla.ement monthly, semi annually and annually.
Seabrook Station procedures require the performance of cleaning and a tube inspection of the Fire Pump House Heat Exchangers.
Acceptance criteria are defined in the Seabrook Station procedures used for performing tests and inspections of the Fire Water System Program.
Sprinkler inspections are acceptable if there is no indication of biofouling in the sprinkler system. Piping inspections and tests are acceptable if there are no indications of unacceptable signs of degradation such as corrosion, microbiologically influenced corrosion or biofouling and that the fire protection system is able to maintain required pressure. Hydrant inspections are acceptable if there is no indication of degradation; sueh as eerrosien.
Clarification to RAI B.2.1.28-3 Response The response to RAI B.2.1.28-3 provided in SBK-L-10204, dated December 17, 2010 (Reference 3) information was provided to confirm that the effects of aging of the concrete containment will be adequately managed so that its intended function would be maintained consistent with the current licensing basis for the period of extended operation. As discussed with the Staff on April 7, 2011 during the IP71002 Inspection, additional changes to the LRA have been made to support this response. The LRA has been revised to include material/environment of concrete in raw water for the containment building.
- 1) LRA Chapter 2, Page 2.4-10 is changed by adding the following row to Table 2.4-2:
CONCRETE IN RAW WA TER Missile Barrier Shelter, Protection Structural Pressure Barrier Structural Support
United States Nuclear Regulatory Commission Page 6 of 7 SBK-L-1 1063/ Enclosure 2
- 2) LRA Chapter 3, Page 3.5-97 is changed by adding the following four rows to Table 3.5.2-2, as shown on the following page:
CNT-CS-Reinforced Cracking due ASME Section XI, Concrete Exposed to Missile Conc Raw Water to epansion Subsection IWL 3.5.1-15 A
Ret E t
teBarrier (External) reaction with Program(C-04)
Raw Water Barregere(Pr-04am CNT-CS-Reinforced Cracking due ASME Section XI, IIA 1-3 Concrete Exposed to
- Shelter, Concrete Raw Water to expansion!
Subsection IWL (C-0 3.5.1-15 A
Raw Water Protection (External) reaction with Program (C-04) aggregates CNT-CS-Reinforced Structural Rracking due ASME Section XI, Concrete Exposed to Pressure Concrete Raw Water to epansion/
Subsection IWL 3.5.1-15 A
Raw Water Barrier (External) reaction with Program (C-04) aggregates CNT-CS-Reinforced Cracking due ASME Section XI, IIA1-3 Concrete Exposed to Structural onc Raw Water to pansion Subsection IWL (C-0 3.5.1-15 A
Raw Water Support (External) reaction with Program (C-04) aggregates
- 3) License Renewal Application Appendix A, Section A.3, page A-43, as changed by RAI B2.1.28-3 in as follows:
SBK-L-10204, is further changed so that Commitment No. 52 read ASME Implement measures to Section XI, maintain the exterior surface Prior to the period e
- 52.
Subsection of the Containment Structure, A.2.1.28 extended operatien.
i on from elevation -30 feet to +20 By December 31, 2012 IWL feet, dewatered.
I Clarification to RAI B.2.1.17-1 Resnonse In response to RAI B.2.1.17-1 provided in SBK-L-10204, dated December 17, 2010 (Reference 3) it was stated that the Structures Monitoring Program will include an external surface inspection of the aboveground steel tanks 1-FP-TK-35-A, 1-FP-TK B, 1-FP-TK-36-A, l-FP-TK-36-B, and 1-AB-TK-29.
This inspection will inspect the paint or coating for cracking, flaking, or peeling. Based on discussions with the Staff on April 7, 2011, the interval at which tank coatings are being inspected for cracking, flaking, or peeling has been clarified. There are no changes being made to the LRA.
Clarification is as follows:
The Engineer Department Standard for Structural Monitoring Program calls for the performance of an external visual (and tactile, where required) inspection for any signs of degradation including coatings, sealants, and caulking of the Fire fuel oil tanks, Fire water tanks and Auxiliary boiler fuel oil tank l-FP-TK-35-A, 1-FP-TK-35-B, 1-FP-TK-36-A, 1-FPTK-36-B and l-AB-TK-29 on 5 year frequency.
United States Nuclear Regulatory Commission Page 7 of 7 SBK-L-11063/ Enclosure 2 In addition to the Structural Monitoring inspection the Fire Protection program calls for the performance of inspections of the Fire oil fuel tanks and Fire water tanks 1-FP-TK-35-A, 1-FP-TK-35-B, 1-FP-TK-36-A, 1-FP-TK-36-B and Auxiliary boiler fuel oil tank 1-AB-TK-29. The inspections include visual inspection of coatings and tactile examination of caulking for tanks that have a caulking seal between the tank and foundation. The inspections are performed on a quarterly basis per the System Walkdown Engineering Guidelines.
to SBK-L-11063 LRA Appendix A - Final Safety Report Supplement Table A.3 License Renewal Commitment List
United States Nuclear Regulatory Commission SBK-L-1 1063 / Enclosure 3 Page 2 of 13 A.3 LICENSE RENEWAL COMMITMENT LIST No.
PROGRAM or TOPIC COMMITMENT UFSAR LOCATION SCHEDULE Program to be implemented prior to the An inspection plan for Reactor Vessel Internals will be period of extended submitted for NRC review and approval at least twenty-four operation. Inspection
- 1.
PWR Vessel Internals months prior to entering the period of extended operation.
A.2.1.7 plan to be submitted to NRC not less than 24 months prior to the period of extended operation.
Enhance the program to include visual inspection for
- 2.
Closed-Cycle Cooling cracking, loss of material and fouling when the in-scope A.2.1.12 Prior to the period of Waterextended operation systems are opened for maintenance.
extendedoperation 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 Refueling) Handling wear on the rails in the rail system.
Systems Inspection of Overhead Heavy Load and Light Enhance the program to list additional cranes for Prior to the period of
- 4.
Load (Related to monitoring.
A.2.1.13 extended operation Refueling) Handling Systems Enhance the program to include an annual air quality test 5.Compressed Air requirement for the Diesel Generator compressed air sub Prior to the period of Monitoring system.
A.2.1.14 extended operation
United States Nuclear Regulatory Commission SBK-L-1 1063 / Enclosure 3 Page 3 of 13 No.
PROGRAM or TOPIC COMMITMENT UFSAR LOCATION SCHEDULE
- 6.
Fire Protection Enhance the program to perform visual inspection of A. 2.l.15 Prior to the period of penetration seals by a fire protection qualified inspector.
extended operation.
Enhance the program to add inspection requirements such
- 7.
Fire Protection as spalling, and loss of material caused by freeze-thaw, A.2.1.15 Prioreto the period of chemical attack, and reaction with aggregates by qualified extended operation.
inspector.
- 8.
Enhance the program to include the performance of visual Prior to the period of Fire Protection inspection of fire-rated doors by a fire protection qualified A.2.1.15 extended operation.
inspector.
Enhance the program to include NFPA 25 guidance for
- 9.
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 accordance A.2.1.16 extended operation.
with the guidance of NFPA 25.
United States Nuclear Regulatory Commission SBK-L-1 1063 / Enclosure 3 Page 4 of 13 No.
PROGRAM or TOPIC COMMITMENT UFSAR LOCATION SCHEDULE 11.
Fire Water System 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 inspections have been performed prior to the period of extended operation. If a representative number of inspections 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.
A.2.1.16 Within ten years prior to the period of extended operation.
Enhance the program to include components and aging Prior to the period of
- 12. Aboveground Steel effects required by the Aboveground Steel Tanks.
A.2.1.17 extended operation.
Tanksexeddoeain Enhance the program to include an ultrasonic inspection Within ten years prior to Tak and evaluation of the internal bottom surface of the two Fire A.2.1.17 the period of extended Protection Water Storage Tanks.
operation.
Enhance program to add requirements to 1) sample and
- 14.
analyze new fuel deliveries for biodiesel prior to offloading Prior to the period of Fuel Oil Chemistry to the Auxiliary Boiler fuel oil storage tank and 2)
A.2.1.18 extended operation.
periodically 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 A.2.1.18 Prior to the period of tank at least once per quarter and to remove water as extended operation.
necessary.
United States Nuclear Regulatory Commission SBK-L-1 1063 / Enclosure 3 Page 5 of 13 No.
PROGRAM or TOPIC COMMITMENT UFSAR LOCATION SCHEDULE
- 16.
Enhance the program to require draining, cleaning and Prior to the period of Fuel Oil Chemistry inspection of the diesel fire pump fuel oil day tanks on a A.2.1.18 extended 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
- 17.
Fl 0i Cht draining, cleaning and inspection of the Diesel Generator A.2.1.18 Prior to the period of 17Fuel Oil Chemistry fuel oil storage tanks, Diesel Generator fuel oil day tanks, 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 Surveillance capsules, unless discarded before August 31, 2000, are A.2.1.19 extended operation.
placed in storage.
Enhance the program to specify that if plant operations exceed the limitations or bounds defined by the Reactor
- 19. Reactor Vessel Vessel Surveillance Program, such as operating at a lower Prior to the period of Surveillance cold leg temperature or higher fluence, the impact of plant A.2.1.19 extended operation.
operation changes on the extent of Reactor Vessel embrittlement will be evaluated and the NRC will be notified.
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
- 20. Reactor Vessel outage in which the capsule receives a neutron fluence that Prior to the period of Surveillance meets the schedule requirements of 10 CFR 50 Appendix A.2.1.19 extended operation.
H and ASTM El 85-82 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.
United States Nuclear Regulatory Commission SBK-L-1 1063 / Enclosure 3 Page 6 of 13 No.
PROGRAM or TOPIC COMMITMENT UFSAR LOCATION SCHEDULE 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 Surveillance maintains it in a condition which would permit its future use, extended operation.
including during the period of 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.
- 23. Selective Leaching of The program will include a one-time inspection of selected Within five years prior to Materials components where selective leaching has not been A.2.1.21 the period of extended identified and periodic inspections of selected components operation.
where selective leaching has been identified.
- 24.
Buried Piping And Tanks Implement the Buried Piping And Tanks Inspection Within ten years prior to Inspection Program.
A.2.1.22 entering the period of extended operation
- 25. One-Time Inspection of Implement the One-Time Inspection of ASME Code Class Within ten years prior to ASME Code Class 1 1A.2.1.23 the period of extended Small Bore-Piping operation.
Enhance the program to specifically address the scope of the program, relevant degradation mechanisms and effects
- 26. External Surfaces of interest, the refueling outage inspection frequency, the Prior to the period of Monitoring inspections of opportunity for possible corrosion under A.2.1.24 extended operation.
insulation, the training requirements for inspectors and the required periodic reviews to determine program effectiveness.
United States Nuclear Regulatory Commission SBK-L-11063 / Enclosure 3 Page 7 of 13 No.
PROGRAM or TOPIC COMMITMENT UFSAR LOCATION SCHEDULE Inspection of Internal
- 27. Surfaces in Implement the Inspection of Internal Surfaces in Prior to the period of andouc Miscellaneous Piping and Ducting Components Program.
A.2.1.25 extended operation.
and Ducting Components
- 28. L Enhance the program to add required equipment, lube oil Prior to the period of Lubricating Oil Analysis analysis required, sampling frequency, and periodic oil A.2.1.26 extended operation.
changes.
- 29.
Enhance the program to sample the oil for the Switchyard Prior to the period of Lubricating Oil Analysis SF6 compressors and the Reactor Coolant pump oil A.2.1.26 extended operation.
collection tanks.
Enhance the program to require the performance of a one-
- 30. Lubricating Oil Analysis time ultrasonic thickness measurement of the lower portion A.2.1.26 Prior to the period of of the Reactor Coolant pump oil collection tanks prior to the extended operation.
period of extended operation.
- 31. ASME Section XI, Enhance procedure to include the definition of A.2.1.28 Prior to the period of Subsection IWL "Responsible Engineer"'.
extended operation.
- 32. Structures Monitoring Enhance procedure to add the aging effects, additional Prior to the period of Program locations, inspection frequency and ultrasonic test A.2.1.31 extended operation.
requirements.
Enhance procedure to include inspection of opportunity Prior to the period of
- 33. Structures Monitoring when planning excavation work that would expose A.2.1.31 exte e perion.
Program inaccessible concrete.
extended operation.
Electrical Cables and Connections Not Subject Implement the Electrical Cables and Connections Not
- 34. to 10 CFR 50.49 Subject to 10 CFR 50.49 Environmental Qualification A.2.1.32 extended operation.
Environmental Requirements program.
Qualification
____Requirements________________________________________________
United States Nuclear Regulatory Commission SBK-L-1 1063 / Enclosure 3 Page 8 of 13 No.
PROGRAM or TOPIC COMMITMENT UFSAR LOCATION SCHEDULE Electrical Cables and Connections Not Subject to 10 CFR 50.49 Implement the Electrical Cables and Connections Not Prior to the period of 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 10 CFR 50.49 Environmental Qualification Requirements A.2.1.34 Environmental extended operation.
Qualification program.
Requirements
- 37. Metal Enclosed Bus Implement the Metal Enclosed Bus program.
A.2.1.35 Prior to the period of extended operation.
- 38. Fuse Holders Implement the Fuse Holders program.
A.2.1.36 Prior to the period of I
extended operation.
Electrical Cable Connections Not Subject Implement the Electrical Cable Connections Not Subject to
- 39. to 10 CFR 50.49 10 CFR 50.49 Environmental Qualification Requirements A.2.1.37 Prior to the period of Environmental extended operation.
Qualification program.
Requirements
A.2.2.1 Prior to the period of extended operation.
- 41.
Metal Fatigue of Reactor Enhance the program to include additional transients Prior to the period of Coolant Pressure beyond those defined in the Technical Specifications and A.2.3.1 extended operation.
Boundary UFSAR.
United States Nuclear Regulatory Commission SBK-L-I 1063 / Enclosure 3 Page 9 of 13 No.
PROGRAM or TOPIC COMMITMENT UFSAR LOCATION SCHEDULE Metal Fatigue of Reactor Enhance the program to implement a software program, to Prior to the period of
- 42. Coolant Pressure count transients to monitor cumulative usage on selected A.2.3.1 extended operation.
Boundary components.
The updated analyses will Pressure -Temperature be submitted at the
.43. Limits, including Low Seabrook Station will submit updates to the P-T curves and appropriate time to Temperature LTOP limits to the NRC at the appropriate time to comply A.2.4.1.4 comply with 10 CFR 50 Overpressure Protection with 10 CFR 50 Appendix G.
Appendix G, Fracture 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 Environmentally-identified consists of nickel alloy, the environmentally-At least two years prior to
- 44. Assisted Fatigue assisted fatigue calculation for nickel alloy will be A.2.4.2.3 entering the period of Analyses (TLAA) performed using the rules of NUREG/CR-6909.
extended operation.
(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 valid for the period of extended operation or from an analysis using an NRC-approved version of the
United States Nuclear Regulatory Commission SBK-L-1 1063 / Enclosure 3 Page 10 of 13 No.
PROGRAM or TOPIC COMMITMENT UFSAR LOCATION SCHEDULE 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, 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 bv a method acceDted bv the NRC).
Mechanical Equipment Prior to the period of Qualification Revise MechanicQualification ualification Files.
A.2.4.5.9 extended operation.
Protective Coating Enhance the program by designating and qualifying an Prior to the period of Pr. otetoivean A.2.1.38 etne prto
- 46. Monitoring and Inspector Coordinator and an Inspection Results Evaluator.
extended operation Maintenance Enhance the program by including, "Instruments and Protective Coating Equipment needed for inspection may include, but not be Co an g limited to, flashlight, spotlights, marker pen, mirror, A.2.1.38 Prior to the period of
- 47. Monitoring and measuring tape, magnifier, binoculars, camera with or extended operation without wide angle lens, and self sealing polyethylene sample bags."
Protctiv CoaingPrior to the period of Protective Coating Enhance the program to include a review of the previous A.2. 1.38 extended operation
- 48. Monitoring and two monitoring reports.
Maintenance
United States Nuclear Regulatory Commission SBK-L-1 1063 / Enclosure 3 Page 11 of 13 No.
PROGRAM or TOPIC COMMITMENT UFSAR LOCATION SCHEDULE 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 personnel, A.2.1.38 extended operation Maintenance who is to prepare a summary of findings and recommendations for future surveillance or repair.
Perform UTtesting of the containment liner plate in the A.2.1.4-27 Prir to the period oe ASME Section XI, vicinity of the moisture barrier for loss of material.
extended operation. No ASMEbsection X,
later than December 31,
- 50. Subsection IWE 2015 and repeated at intervals of no more than five refueling outages ASME Section XI, Perform confirmatory testing and evaluation of the Prior to the period of 51m Subsection IWL Containment Structure concrete A.2.1.28 extended operation ASME Section XI, Implement measures to maintain the exterior surface of the Prior to the period o
- 52.
Containment Structure, from elevation -30 feet to +20 feet, A.2.1.28 extended operation. By Subsection IWL in a dewatered state.
December 31, 2012 Reactor Head Closure Replace the spare reactor head closure stud(s)
Prior to the period of
- 53. ReactorHeadClosure manufactured from the bar that has a yield strength > 150 A.2.1.3 exte e perion.
Studs ksi with ones that do not exceed 150 ksi.
extended operation.
Unless an alternate repair criteria changing the ASME code boundary is permanently approved by the NRC, or the Seabrook Station steam generators are changed to Program to be submitted
- 54. Steam Generator Tube eliminate PWSCC-susceptible tube-to-tubesheet welds, A.2.1.10 to NRC at least 24 Integrity submit a plant-specific aging management program to months prior to the period manage the potential aging effect of cracking due to of extended operation.
PWSCC at least twenty-four months prior to entering the Period of Extended Operation.
United States Nuclear Regulatory Commission SBK-L-1 1063 / Enclosure 3 Page 12 of 13 No.
PROGRAM or TOPIC COMMITMENT UFSAR LOCATION SCHEDULE Steam Generator Tube Seabrook will perform an inspection of each steam Prior to entering the
- 55.
generator to assess the condition of the divider plate A.2.1.10 period of extended Integrity assembly.
operation Closed-Cycle Cooling Revise the station program documents to reflect the EPRI Prior to entering the
- 56. Water System Guideline operating ranges and Action Level values for A.2.1.12 period of extended hydrazine and sulfates.
operation.
Closed-Cycle Cooling Revise the station program documents to reflect the EPRI Prior to entering the
- 57. Water-Cyste Guideline operating ranges and Action Level values for A.2.1.12 period of extended Water System Diesel Generator Cooling Water Jacket pH.
operation.
Update Technical Requirement Program 5.1, (Diesel Fuel
and ASTM D4057-95 required by the GALL XI.M30 Rev 1 Nickel Alloy Nozzles and The Nickel Alloy Aging Nozzles and Penetrations program Prior to the period of
- 59. Penetrations will implement applicable Bulletins, Generic Letters, and A.2.2.3 extended operation.
staff accepted industry guidelines.
Buried Piping and Tanks Implement the design change replacing the buried Auxiliary Prior to entering the
- 60. Inspection Boiler supply piping with a pipe-within-pipe configuration A.2.1.22 period of extended with leak indication capability, operation.
Compressed Air Replace the flexible hoses associated with the Diesel Within ten years prior to 61.Monitoring Program Generator air compressors on a frequency of every 10 A.2.1.14 entering the period of years.
extended operation.
Enhance the program to include a statement that sampling Prior to entering the
- 62. Water Chemistry frequencies are increased when chemistry action levels are A.2.1.2 period of extended exceeded.
operation.
Ensure that the quarterly CVCS Charging Pump testing is Prior to the period of continued during the PEO. Additionally, add a precaution to N/A extended operation Flow Induced Erosion the test procedure to state that an increase in the CVCS
- 63.
Charging Pump mini flow above the acceptance criteria may be indicative of erosion of the mini flow orifice as described in LER 50-275/94-023.
United States Nuclear Regulatory Commission SBK-L-1 1063 / Enclosure 3 Page 13 of 13 No.
PROGRAM or TOPIC COMMITMENT UFSAR LOCATION SCHEDULE Soil analysis shall be performed prior to entering the period A.2.1.22 Prior to entering the Buried Piping and Tanks of extended operation to determine the corrosivity of the period of extended
- 64. Inspection soil in the vicinity of non-cathodically protected steel pipe operation.
within the scope of this program. If the initial analysis shows the soil to be non-corrosive, this analysis will be re-performed every ten years thereafter.