ML110700018
| ML110700018 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 03/30/2011 |
| From: | Plasse R License Renewal Projects Branch 2 |
| To: | Freeman P, O'Keefe M NextEra Energy Seabrook |
| Plasse, Richard A. DLR/RPB2, 415-1427 | |
| References | |
| TAC ME4028 | |
| Download: ML110700018 (14) | |
Text
REG UUNITED STATES NUCLEAR REGULATORY COMMISSION "'01'..<.
WASHINGTON, D.C. 20555-0001
<< 0 I-0 <II l:
§ March 30, I') 0' ****1< 'Ii' Mr. Paul Site Vice c/o Mr. Michael NextEra Energy Seabrook, P.O. Box Seabrook, NH REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE REVIEW OF THE SEABROOK STATION LICENSE RENEWAL APPLICATION (TAC NUMBER ME4028)
Dear Mr. Freeman:
By letter dated May 25,2010, NextEra Energy Seabrook, LLC submitted an application pursuant to Title 10 of the Code of Federal Regulations Part 54, to renew Operating License NPF-86 for Seabrook Station, Unit 1, for review by the U.S. Nuclear Regulatory Commission (NRC or the staff). The staff is reviewing the information contained in the license renewal application and has identified, in the enclosure, areas where additional information is needed to complete the review. The request for additional information was discussed with Mr. Rick Cliche, and a mutually agreeable date for the response is within 30 days from the date of this letter. If you have any questions.
please contact me at 301-415-1427 or bye-mail at Richard.Plasse@nrc.gov.
Sincerely, AJ91Vi r U . -Richard A Plasse, Project Manager Projects Branch 2 Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-443 As cc w/encl: Listserv -I SEABROOK LICENSE RENEWAL REQUEST FOR ADDITIONAL INFORMATION SET RAI2.3.3.1S-1
Background:
The LRA drawing PID-1-FP-LR20270 shows that sprinkler systems at locations C-4 to H-4 are out of scope (Le., not colored in red). In a letter dated December 3,2010, NextEra Energy responded to RAJ 2.3.3.15-1 by stating the sprinkler systems located on drawing PID-1-FP-LR20270, locations C-4 to H-4, are not in scope of license renewal because they do not provide a function credited in the Appendix R safe shutdown analysis and do not provide a pressure boundary function needed to support the Appendix R suppression systems. Issue: NextEra's response to RAI 2.3.3.15-1 may be inconsistent with the updated final safety analysis report (UFSAR) Revision 13, Section 9.5.1.2(c)(7), "Manually Operated Pre-Action Sprinkler Systems," which states that manually operated sprinkler systems are provided for areas containing turbine bearings and lube oil piping from turbine bearings to guard. Request: The fire suppression systems discussed above appear to have been credited in the approved fire protection program (UFSAR Section 9.5.1) for the fire suppression activities.
Based on its review, the staff does not find the applicant's response to RAI 2.3.3.15-1 acceptable.
The applicant explains that the fire protection systems in question are not credited to meet the requirements of Appendix R for achieving safe-shutdown in the event of a fire. However, the staff finds that the applicant's analysis of fire protection regulation does not completely capture the fire protection SSCs required for compliance with Title 10 of the Code of Federal Regulations (10 CFR) 50.48. The scope of structure systems and components (SSCs) required for compliance with 10 CFR 50.48 and general design criteria 3 (GDC 3) goes beyond preserving the ability to maintain safe-shutdown in the event of a fire. GDC 3 states in part, "Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety." Furthermore, the general requirements provided in GDC 3 to minimize the adverse effects of fires on SSCs important to safety establish a general level of protection which is afforded to aI/ systems, not only where required to prevent a loss of safe shutdown capability.
10 CFR 50.48(a) states, "Each operating nuclear power plant must have a fire protection plan that satisfies Criterion 3 of Appendix A of this part." The term "important to safety" encompasses a broader scope of equipment than safety-related and safe-shutdown equipment.
Though there is a focus on the protection of safety-related equipment or safe-shutdown equipment, this does not imply that there is any exclusion of equipment which protects safety related equipment.
ENCLOSURE
-2 For example, in accordance with 10 CFR 50.48, some portions of suppression systems may be required in plant areas where a fire could result in the release of radioactive materials to the environment, even if no safety-related or safe-shutdown equipment is located in that particular fire area. The staff finds this contrary to the UFSAR which includes the original Seabrook Station fire protection SE as the CLB. The staff requests that the applicant verify whether fire suppression systems discussed are in the scope of license renewal in accordance with 10 CFR 54.4(a) and subject to an AMR in accordance with 10 CFR 54.21(a)(1).
If they are excluded from the scope of license renewal and not subject to an AMR, the staff requests that the applicant provide justification for the exclusion.
Follow-up RAI3.1.1.60-01/02
Background:
By letter dated January 5, 2011, the staff issued two RAls to the applicant.
RAI 3.1.1-60-01 requested that the applicant justify not including an applicable aging management review (AMR) line item to manage loss of material due to wear in the nickel-alloy flux thimble tubes and to justify why a Flux Thimble Tube Inspection Program is not credited to manage loss of material due to wear for these nickel-alloy flux thimble tubes. RAI 3.1.1-60-02 requested that the applicant justify the use of the PWR Vessel Internals Program to manage cracking in the flux thimble tubes, considering that MRP-227 Rev. 0 does not contain recommendations for managing cracking in Westinghouse-design flux thimble tubes. In its response dated February 3, 2011, the applicant stated that its design is unique and can accommodate both fixed and movable incore detectors.
The applicant also stated that since its Operating Cycle 5, the moveable incore detectors have not been used and were placed in a lay-up condition during Refueling Outage 7 (fall of 2000). The applicant also stated that since Refueling Outage 7, as part of a design change, the seal table tubing between the inner calibration tubing and the isolation valves has been removed and the inner calibration tube has been capped. The applicant further stated that, based on the unique design features of the incore detectors, the aging effects managed by GALL AMP XI.M37, do not apply to Seabrook Station. The applicant also stated that the movable flux thimbles do not have a license renewal intended function and the line items referencing the flux thimble tubes will be deleted from LRA Tables 2.3.1-3 and 3.1.2-3. Issue: During its review of the applicant's responses to RAls 3.1.1-60-01 and 3.1.1-60-02, the staff noted that the flux thimbles for the moveable incore detectors, if left in a permanent lay-up condition, would not be subject to flow induced vibrations, and therefore would not be subject to wear. However the staff also noted that, under the plants CLB and design basis, the applicant has the option to place the movable incore detectors back in service and the flux thimbles will once again provide a pressure boundary function.
-3 Request: Since the applicant has the option to place the movable incore detectors back in service, justify the deletion of the AMR line items associated with cracking of the flux thimble tubes from LRA Table 3.1.2-3. Also, justify why an aging management program is not required to manage loss of material due to wear of the flux thimbles if the movable incore detectors were placed back into service. FoIIOW-UI2 RAI 3.3.2.3.4-1
Background:
In its response to RAt 3.3.2.3.4-1, the applicant stated that the fiberglass piping components in the chlorination system exposed to raw water (LRA Table 3.3.2-4) are not subject to aging because they are constructed of either an epoxy resin or vinyl ester type of material for which the system's design temperature is below the aging limit of 170F. Issue: The staff noted that based on an independent review of corrosion data, chlorine can cause aging in either epoxy resin or vinyl ester based fiberglass components.
The staff noted that the applicant's 2009 Environmental Monitoring Report (ADAMS Accession No. ML1033602980) states that the sodium hypochlorite level used for the circulating water system is 15%; however, the staff recognizes that the chlorine level in the in-scope fiberglass piping may not be as high as that in the storage tank, or the in-scope piping may not be associated with the sodium hypochlorite injection source. The staff cannot make a determination of no aging effects without knowing the chlorine level in the in-scope fiberglass piping. Request: State the chlorine concentration, in ppm, for in-scope piping components in the chlorination system and state why no aging effect will occur or propose an aging management program for the components.
FoIIOW-UI2 RAJ 4.3-1 b
Background:
In its response to RAI 4.3-1 the applicant stated that the fatigue conformance of ASME Class 1 valves was demonstrated by performing an "umbrella" fatigue analysis of the piping system containing the valves (Le., in accordance with ASME Section III, Subsection NB-3650).
However, UFSAR Section 3.9(N).1.4(e) states that the pressure boundary portions of Class 1 valves in the reactor coolant system were designed and analyzed according to the valve design requirements of ASME Section III, NB-3500 (edition 1971 including 1972 Addenda).
Issue: It is not clear why the fatigue analyses of Class 1 valves was performed in accordance with ASME Section III Subsection NB-3650, "Analysis of Piping Products," instead of the requirements of Subsection NB-3500 as indicated in UFSAR Section 3.9(N).1.4(e).
Request: Clarify and explain why Class 1 valve fatigue conformance was demonstrated by performing an "umbrella" fatigue analysiS using ASME Section III Subsection NB-3650 instead of Subsection NB-3500. Justify that the "umbrella" fatigue analysis using ASME Section III Subsection NB-3650 is equivalent to or more conservative than the NB-3500 analysis.
Follow-up RAI 4.3.1-1 b
Background:
In its response to RAI4.3.1-1, the applicant stated that in LRA Table 4.3.1-3, the rows for Unit Loading Between 0% and 15% Power and Unit Unloading Between 15% and 0% Power were revised, and the 60-year projected cycles for Unit Loading and Unloading are 70 and 65, respectively.
Issue: The revised values for the 60-year projected cycles for Unit Loading and Unloading between 0% and 15% Power are inconsistent with the projected values for the other transients.
In LRA Table 4.3.1-3, all the projected values for other transients have been linearly extrapolated from 18.6-year to 60-year operation with a ratio of 3.22 or higher (the ratio between 60-year projected cycle and current number of cycle is 3.22). For Unit Loading and Unit Unloading transients, the ratios between 60-year projected cycle and current number of cycle are 2.6 and 2.5, respectively.
Request: Provide the basis and justify the 60-year projected values for the Unit Loading and Unit Unloading transients.
RAI4.3.1-2 Background and Issue: For the "Feedwater Heaters Out of Service" transient in LRA Table 4.3.1-3,the number of nuclear steam supply system (NSSS) Design Cycles is "2000" with footnote (5) indicating that the original design analysis number is assumed to be the anticipated number of cycles at the end of the period of extended operation.
However, the number of 60-year projected cycles for this transient is listed as "39" in the Table 4.3.1-3.
-5 LRA Table 4.3.1-2 identified three Emergency Transients that are reactor coolant system design transients.
The applicant did not provide the number of current cycles and the number of 50-year projected cycles for these three Emergency Transients in LRA Table 4.3.1-3. Request: (1) For consistency, revise LRA Table 4.3.1-3 to reflect the proper 50-year projected cycles for the "Feedwater Heaters Out of Service" transient or justify why footnote (5) is not applicable to this transient.
(2) Provide the number of current cycles and the number of 50-year projected cycles for the three Emergency Transients in LRA Table 4.3.1-3 or justify why cycle projections are not needed. Clarify whether these transients will be monitored under the Metal Fatigue of Reactor Coolant Pressure Boundary Program or justify why the transients need not be monitored.
FOllow-u,e RAt 4.3.3-1b
Background:
In its response to RAI 4.3.3-1 the applicant stated that the intent is to disposition fatigue of the vessel internals using 10 CFR 54.21 (c}(1}(i).
The applicant added that, based on generic and plant-specific analyses, the following were identified as fatigue limiting locations:
lower support columns, core barrel nozzle, lower core plate, and upper core plate. The fatigue cumulative usuage factor (CUFs) for these locations were all shown to be less than 1.0. The applicant further stated that the effects of fatigue at these locations will also be monitored by cycle counting under its Metal Fatigue of Reactor Coolant Pressure Boundary Program to verify that the number of design cycles assumed in the analyses will not be exceeded during the period of extended operation.
Issue: The applicant did not update LRA Section 4.3.3 and applicable LRA Appendix A section reflecting that these TLAA have been dis positioned in accordance with 10 CFR 54.21(c)(1)(i) and 10 CFR 54.21(c)(1}(iii) using the Metal Fatigue of Reactor Coolant Pressure Boundary Program. Request: For consistency in the LRA, revise LRA Section 4.3.3 indicating that, for the reactor vessel internal components, the TLAA disposition is in accordance with 10 CFR 54.21 (c}(1)(i) and 10 CFR 54.21 (c)(1)(iii) using the Metal Fatigue of Reactor Coolant Pressure Boundary Program.
-6 Follow-up RAI 4.1-1 b
Background:
In its response to RAI 4.1-1, the applicant stated that a flow-induced vibration (FlY) analysis is not part of the CLB for its reactor vessel internal (RVI) components.
Also, in its response to RAI 4.1-2, the applicant indicated that fluence-dependent reduction offracture toughness of vessel internals is not analyzed as part of the current licensing design basis. Issue: Part 1 -LRA Table 4.1-3 still indicates that an FIV analysis and loss of fracture toughness (ductility reduction) analysis are part of the CLB, and being TLAAs, for the RVI components and these analyses are addressed in LRA Section 4.3.3. The applicant did not revise LRA Table 4.1-3 to indicate that FIV and ductility reductionlloss of fracture toughness analyses are not TLAA. Furthermore, the staff noted that LRA Section 4.3.3 does not include any discussion regarding how the FIV and ductility reductionlloss of fracture toughness analyses factor into the CUF calculations for the RVI core support structure components.
Part 2 -The response to RAI 4.1-1 indicated that there are "further analyses performed for the Seabrook Station rea*:tor internals" for FIV of the RVI components.
However, the applicant's response did not provide any comparison of these FIV analyses to the NRC's six criteria for time-limited aging analysis (TLAAs) in 10 CFR 54.3. Therefore, the staff cannot determine whether these further analyses need to be identified as TLAA for the LRA. Reguest: Part 1 -For consistency in the LRA, the staff requests that the applicant either revise LRA Table 4.1-3 to identify that FIV and ductility reductionlloss of fracture toughness analyses are not TLAAs and provide the justification for making these changes to LRA Table 4.1-3; or amend LRA Section 4.3.3 to clarify how the CUF calculations account for and bound any considerations of FIV in the RVI core support structures and/or reduction in ductility or fracture toughness properties for the materials that the core support structures are fabricated from. Part 2 -Provide the basis and justify why these further analyses for the RVI components not conform to the definition of a TLAA in 10 CFR 54.3. Follow-up RAI 4.3.4-1 b
Background:
In its response to RAI4.3.4-1 request (1), the applicant clarified that the hot leg surge nozzle-to-pipe weld was evaluated to be the limiting location in the surge line. The staff notes that the surge nozzle-to-pipe weld consists of nozzle, nozzle-to-safe end weld, safe end, safe end-to-pipe weld, and the pipe.
Issue: It is not clear to the staff whether the fatigue limiting CUF evaluations were performed for the nozzle-to-safe end weld or the safe end-to-pipe weld. In footnote (1) of Table 1 of the applicant's response to RAI 4.3-1, the highest fatigue usage location is identified as nozzle transition and safe end, whereas in Table 2 of the applicant's response to RAt 4.3.2-1 the highest fatigue usage location is identified as nozzle safe end-to pipe weld. Furthermore, the staff notes that there are inconsistencies in the values of CUF listed in various tables: -In LRA Table 4.3.4-1, for the hot leg surge nozzle-to-pipe weld, the 60-year CUFs is 0.2844 in air and 3.428 in reactor coolant environment.
Fen is 12.05. -In RAI4.3.4-1 response, for the hot leg surge nozzle safe-end, the 60-year CUFs is 0.2844 in air and 3.2848 in reactor coolant environment.
Fen is 11.55. Reguest: (1) Resolve or justify the inconsistencies in the reported values of CUF. Revise the LRA sections and Tables accordingly.
(2) Clarify the fatigue limiting location of the hot leg surge nozzle. Revise the LRA sections and Tables accordingly.
Follow-ue RAt 4.3.5-1 b
Background:
In LRA Section 4.3.5, the applicant described the TLAA for steam generator tube fatigue in the U-bend region resulting from flow-induced vibrations (FIV). In its response to RAI 4.3.5-1, the applicant amended its LRA Section 4.3.5. The amendment separated the original TLAA description into two LRA Sections:
4.3.5 to deal with fatigue and a new Section 4.7.15 to deal with wear -both caused by FIV. The applicant also changed the disposition of Section 4.3.5 to 10 CFR 54.21(c)(1)(iii).
Issue: The staff notes that the applicant did not remove wear-related (loss of material) discussion in the amended Section 4.3.5. The staff also notes that the title of Section 4.3.5 remains unchanged, which still indicated the loss of material as part of the TLAA of Section 4.3.5. With regard to the fatigue issue, the staff does not find the applicant's change of disposition of TLAA for steam generator tube fatigue due to the FIV acceptable.
The staff notes that, with the existing fatigue evaluation in Section 4.3.5, the applicant demonstrated that the CUF is well below the acceptance value of 1.0 for the period of extended operation.
The staff finds the 10 CFR 54.21 (c)(1 )(i) disposition appropriate for steam generator tube fatigue. The staff noted that the Steam Generator Tube Integrity Program cannot be substituted for an ASME Code
-Section III fatigue evaluation unless justifications are provided to demonstrate that the Steam Generator Tube Integrity Program addresses the fatigue-related CUF analysis.
The staff also could not confirm the adequacy of UFSAR supplement summary of the TLAA in Section 4.3.5 because the applicant did not amend the applicable UFSAR supplement description.
Request: (a) Revise Section 4.3.5 and move the wear-related (loss of material) discussion to Section 4.7.15. Limit the title and discussion in Section 4.3.5 to the fatigue issue. Revise Table 4.1-1 to make it consistent with the revised text of 4.3.5. (b) Justify the TLAA disposition of the steam generator tube fatigue due to FIV in accordance with 10 CFR 54.21 (c)(1)(iii), or revise the TLAA disposition of the steam generator tube fatigue TLAA due to FIV in accordance with 10 CFR 54.21(c)(1}(i). (c) Provide an updated UFSAR supplement section LRA Appendix A consistent with the amended TLAA Section 4.3.5. Follow-up RAI4.7.1S-1
Background:
In its response to RAI 4.7.15-1, the applicant dispositioned the steam generator tube wear TLAA in accordance with 10 CFR 54.21(c)(1)(i).
The applicant stated that the basis for its disposition of the wear TLAA was the stretch power uprate (SPU), from 3411 MWt to 3587 MWt, previously approved in staff's SER (ADAMS Accession No. ML050140453).
The staff noted that the applicant received a 1.7% Measurement Uncertainty Recapture approval, from 3587 MWt to 3648 MWt, on May 22, 2006 (ADAMS Accession No. ML061360034).
Issue: The staff noted that the SPU approved in the staff's SER (ADAMS Accession No. ML050140453) is for 5.2% power increase and it was discussed in the revised LRA Section 4.3.5. However, Section 4.7.15 indicated that the power uprate is 7.4%. It is not clear if the actual power uprate is for 5.2% or for 7.4%. The staff noted that, in the staff's SER for the SPU, tube wear increased from approximately 0.003 inches to approximately 0.005 inches at the 5.2% up rated condition.
The staff could not confirm the adequacy of UFSAR supplement summary for the TLAA for the Steam Generator Tubes wear due to FIV for Section 4.7.15 because the applicant did not add a new section in LRA Appendix A for LRA Section 4.7.15. Request: (a) Clarify or reconcile the actual power uprate applicable for the period of extended operation and amend LRA Sections 4.3.5 and 4.7.15 accordingly.
-9 (b) Provide an updated UFSAR supplement section in LRA Appendix A consistent with the added TLAA Section 4.7.15. Follow-up RAI4.3.7-1b 8ackground:
In its respond to RAI4.3.7-1, the applicant stated that there are 831.1 piping, piping components, and piping elements that are within the scope of the license renewal. Issue: The applicant did not amend LRA Section 4.3.7 to include piping and piping components that were designed in accordance with 831.1 rules as part of the non-class 1 components.
The staff noted that that only ASME Section III Class 2 and 3 piping and piping components are considered as non-class 1 in LRA Section 4.3.7. Request: For consistency, revise LRA Section 4.3.7 and applicable LRA Appendix A Section indicating that piping and piping components that were designed in accordance with 831.1 rules are included as part of the Non-class 1 piping and piping components.
Follow-up RAI4.7.9-1b 8ackground:
In its response to RAI4.7.9-1, the applicant stated that there is no specific aging effect identified for canopy seal clamp assemblies.
The applicant stated that there is an aging effect identified for the Head Adapters since a fatigue analysis was developed using design transients over the current operating term. In LRA Section 4.7.9, the applicant stated that the canopy seal clamp assemblies were designed for a 40-year design life on the basis of meeting stress limits. Issue: From the review of LRA Table 3.1.2-2, the staff did not find an AMR line item that addresses head adapters and the associated aging effect. Furthermore, in the revised LRA Section 4.7.9, the applicant stated that the canopy seal clamp assemblies were designed for a 40-year design life and fatigue analysis is performed for the head adapters.
The applicant has not identified, in the LRA, the relationship between the canopy seal clamp assemblies and the head adapters.
The applicant also has not demonstrated how the head adapter's 60-year evaluation support the canopy seal clamp assemblies design life basis of meeting stress limits. Furthermore, the UFSAR Supplement Appendix A.2.4.5.7 was not revised to reflect the change made to LRA Section 4.7.9 summary description.
Request: (1) Identify the AMR line item, in the LRA Section 3 Tables, that is applicable to the head adapters or justify that an AMR line item is not needed for the head adapters.
(2) Clarify and explain how the head adapters' 50-year TLAA evaluation support the canopy seal clamp assemblies design life basis of meeting stress limits. (3) Provide an updated UFSAR supplement section in LRA Appendix A consistent with the change in LRA Section 4.7.9. Follow-up RAI4.7.11-1b
Background:
In its response to RAI 4.7.11 dated February 3, 2011, the applicant stated that only normal service radiation exposure was subjected to a TLAA. The applicant clarified that it has a calculation of total integrated radiation dose design values for a 50-year plant life for various environmental zones and the calculation has been used to evaluate the 50-year dose impact on equipment in their respective zones. The applicant stated that the 50-year design dose values were compared to the current design dose limits of the equipment and it was determined that the 50-year dose limits are bounded by the existing equipment design dose limits. The applicant's disposition of this TLAA in accordance with 10 CFR 54.21 (c)(1)(ii) indicates that the effect of aging on the intended functions of equipment have been projected to be bounded by existing equipment design limits. The staff finds the disposition not acceptable because the existing analyses (equipment design dose limits) has not been revised and extended.
Issue: While the applicant stated that the 50-year dose limits are bounded by the existing equipment design dose limit, it has not provided the projected 50-year doses for all the zones and the dose limits of the equipment within the scope of mechanical equipment qualification (MEa). Without such information, the staff cannot evaluate the adequacy of the TLAA of normal service radiation exposure in MEa. LRA Section 4.7.11, as amended by letter dated February 3, 2011, does not discuss the detail regarding the calculated dose limits and equipment design dose limits for normal service radiation exposure.
Furthermore, the applicant did not amend Commitment No. 45. Commitment No. 45, as its currently stated, did not identify what portion of the MEa files will be revised and what is the acceptance criteria of such revision.
The applicant demonstrated that the existing analyses (equipment design dose limits) are bounding for the projected 50-year doses for all zones. The staff noted that this demonstration of the normal service radiation exposure is consistent with a disposition in accordance with 10 CFR 54.21 (c)(1)(i).
SRP-LR Section 4.7.3.1.1 states that for the disposition of 10 CFR 54.21 (c)(1)(i), the existing analyses should be shown to be bounding during the period of extended operation.
-11 Request: (1) Provide the design dose limits of the equipment within the scope of MEa and the calculated total integrated radiation dose SO-year doses for all the zones to justify that TLAA of normal service radiation exposure in MEa has been properly dispositioned.
(2) Amend LRA Section 4.7.11 and provide sufficient detail to support the TLAA disposition of normal service radiation exposure in MEa. Revise Commitment No. 45 to identify the information to be revised and the acceptance criteria of the revision or justify why the existing Commitment No. 45 is acceptable.
(3) Amend the disposition of TLAA of normal service radiation exposure to MEa to 10 CFR 54.21 (c)(1)(i) or justify why the existing TLAA disposition is acceptable.
If LRA Section 4.7.11 is amended as a result of RAI 4.7.11-1b, provide an updated UFSAR supplement section in LRA Appendix A consistent with the revisions.
March 30, 2011 Mr. Paul Freeman Site Vice President clo Mr. Michael O'Keefe NextEra Energy Seabrook, LLC P.O. Box 300 Seabrook, NH 03874 REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE REVIEW OF THE SEABROOK STATION LICENSE RENEWAL APPLICATION (TAC NUMBER ME4028)
Dear Mr. Freeman:
By letter dated May 25,2010, NextEra Energy Seabrook, LLC submitted an application pursuant to Title 10 of the Code of Federal Regulations Part 54, to renew Operating License NPF-86 for Seabrook Station, Unit 1, for review by the U.S. Nuclear Regulatory Commission (NRC or the staff). The staff is reviewing the information contained in the license renewal application and has identified, in the enclosure, areas where additional information is needed to complete the review. The request for additional information was discussed with Mr. Rick Cliche, and a mutually agreeable date for the response is within 30 days from the date of this letter. If you have any questions, please contact me at 301-415-1427 or bye-mail at Richard.Plasse@nrc.gov.
Sincerely, IRA! Richard A. Plasse, Project Manager Projects Branch 2 Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-443
Enclosure:
As stated cc w/encl: Listserv DISTRIBUTION:
See next page ADAMS Accession Number: ML
RPlasse MSpencer DWrona RPlasse DATE 03/17/2011 03/1712011 03/21/2011 03/30/2011 03/30/2011 OFFICIAL RECORD COpy Letter to Paul Freeman from Richard A. Plasse dated March 3D, 2011
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE REVIEW OF THE SEABROOK STATION LICENSE RENEWAL APPLICATION (TAC NUMBER ME4028) DISTRIBUTION:
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