ML093511025

Request for Additional Information Extended Power Uprate License Amendment Request

ML093511025
Person / Time
Site:	Nine Mile Point
Issue date:	12/23/2009
From:	Richard Guzman Division of Operating Reactor Licensing
To:	Belcher S Nine Mile Point
Guzman R, NRR/DORL, 415-1030
References
TAC ME1476
Download: ML093511025 (11)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 23, 2009 Mr. Samuel L. Belcher Vice President Nine Mile Point Nine Mile Point Nuclear Station, LLC P.O. Box 63 Lycoming, NY 13093 SUB~IECT: REQUEST FOR ADDITIONAL INFORMATION REGARDING NINE MILE POINT NUCLEAR STATION, UNIT NO.2 - RE: THE LICENSE AMENDMENT REQUEST FOR EXTENDED POWER UPRATE OPERATION (TAC NO. ME1476)

Dear Mr. Belcher:

By letter dated May 27, 2009, as supplemented on August 28,2009, Nine Mile Point Nuclear Station, LLC, submitted for Nuclear Regulatory Commission (NRC) staff review and approval, a proposed license amendment requesting an increase in the maximum steady-state power level from 3467 megawatts thermal (MWt) to 3988 MWt for Nine Mile Point Unit NO.2 extended power uprate operation.

The NRC staff is reviewing the information provided in that letter and has determined that additional information is needed to support its review. Enclosed is the NRC staff's request for additional information (RAI). The RAI was discussed with your staff on December 14,17, and 21,2009, and it was agreed that your response would be provided within 60 days from the date of this letter.

Sincerely, If2LeV.~

Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-410

Enclosure:

As stated cc w/encl: Distribution via Listserv

REQUEST FOR ADDITIONAL INFORMATION (RAI)

NINE MILE POINT NUCLEAR STATION, LLC NINE MILE POINT, UNIT NO.2 (NMP2)

LICENSE AMENDMENT REQUEST RE: EXTENDED POWER UPRATE (EPU)

DOCKET NO. 50-410 The Nuclear Regulatory Commission (NRC) staff is reviewing the Nine Mile Point Nuclear Station, LLC (NMPNS or the licensee) license amendment request (LAR) application dated May 27,2009, as supplemented on August 28,2009. The NRC staff has determined that additional information requested below will be needed to support its review.

A. Balance of Plant

1. In the turbine generator technical evaluation section, NMPNS discusses that the overspeed calculation for the turbine generator compares both the entrapped steam energy contained within the turbine and its piping and the sensitivity of the rotor train.

The entrapped energy is expected by NMPNS to increase in EPU conditions. NMPNS also indicates that a hardware modification design and implementation process establishes the overspeed trip settings for the turbine generator. However, NMPNS neither explains how the expected increase in entrapped energy will affect the ability to maintain turbine speed within an acceptable range nor discusses how the specific hardware modification design and implementation process will affect the overspeed trip settings under EPU conditions.

Explain how the increase in entrapped energy in the turbine will affect the ability to maintain turbine speed within an acceptable range. In addition, describe how the changes to both the hardware design and implementation process will affect the turbine generator overspeed trip settings during EPU conditions.

2. NMPNS discusses in the main condenser evacuation system (MCES) technical evaluation section that no changes were made to the following areas of the MCES: (1) the condenser air removal system; (2) the parameters of the physical size of the primary condenser and evacuation times; and (3) the holdup time to the pump discharge line.

However, in the conclusion, NMPNS states that required changes to the MCES were assessed and evaluated for EPU conditions. This statement contradicts the assessments made in the technical evaluation.

Clarify what required changes were made to the MCES, and how they were assessed for EPU conditions to continue to satisfy General Design Criteria (GDC)-60 requirements.

3. In the spent fuel pool cooling and cleanup system (SFPCCS) technical evaluation section, NMPNS discusses that during the normal refueling outages at NMP2, post-EPU implementation, existing administrative controls and procedural limitations will be used to maintain the increase of decay heat for cycle specific full-core offload within design limits.

These administrative controls and procedural limitations for SFPCCS are neither Enclosure

-2 described in the LAR nor are they found in Section 9.1.3 of Revision 20 to the NMP2 Updated Final Safety Analysis Report (UFSAR). It is not clear how the existing administrative controls and procedural limitations will be used to maintain the SFPCCS within design limits with the increased decay heat load due to EPU conditions.

Describe the administrative controls and procedural limitations needed for SFPCCS and how they will be used for EPU conditions to continue maintaining the design limits for the SFPCCS during cycle specific full-core offload, as required by GDC-44.

B, Probabilistic Risk Assessment (PRA)

1. Review Standard (RS)-001, Attachment 1 to Matrix 13, Section 3.1, requires the licensee to address the impacts of EPU on components and system reliability and response times. Attachment 11, Section 2.13.1.2.2 of the Safety Analysis Report (SAR) addresses the reliability impacts, but does not identify impacts on response times. Please provide a discussion of the impacts of the proposed EPU on component and system response times.

2. RS-001, Attachment 1 to Matrix 13, Section 3.1, requires the licensee to describe how it ensures that the PRA adequately models the as-built, as-operated plant, and that the analyses supporting the EPU adequately reflects how the plant will be operated and configured for EPU conditions. Section 2.13 of the SAR does not show how the PRA adequately models the as-built, as-operated plant nor does it show how the PRA will be maintained post-EPU. Please provide a discussion of the missing information.

3. RS-001, Attachment 1 to Matrix 13, Section 3.1 and 3.2, requires the licensee to specifically address vulnerabilities, weaknesses, or review findings identified in the Individual Plant Examination (lPE), Individual Plant External Events Examination (IPEEE), and/or independent/industry peer review findings, that could impact the PRA results and conclusions. It further requires the licensee to present the overall findings of the peer review (by element) and to discuss low-rated elements, and any findings and observations that could potentially impact the licensee's proposed EPU. Attachment 11, Table 2.13-1 and Table 2.13-2 of the SAR shows the peer review 'findings and dispositions to the IPE, IPEEE, and peer reviews, but does not address how these findings may affect the proposed EPU. Please provide a discussion of the missing information.

4. As part of the IPE process and in response to staff's questions for clarification, NMP2 committed to six hardware and procedural enhancements. Address how these commitments were met, and if they have not been met, state how it affects the proposed EPU.

5. EPU SAR Attachment 11, Section 2.13.1.2.1 states no significant change to electrical reliability was identified with the EPU. Provide an explanation that summarizes which analyses or methodology was used for this conclusion and whether there are any changes in a loss of offsite power (LOOP) initiating event frequency or transformer reliabilityllife expectancy due to EPU.

6. EPU SAR Attachment 11, Section 2.13.1.2.1 proposes no significant impact on internal flooding initiator frequencies due to the EPU. Since higher flow rates can contribute to

-3 changes in initiating event for floods, please provide a more thorough justification for your conclusion on how EPU flow rates will not affect internal flooding initiator frequencies or provide a sensitivity analysis addressing this potential increase.

7. EPU SAR Attachment 11, Section 2.13.1.5 provides a sensitivity calculation for a stuck open relief valve, and states that there is a 1a-percent increase in expected demands, post-EPU, due to the power increase. While the sensitivity calculation demonstrates minor changes in core damage frequency (CDF) and large early release frequency (LERF), the probability of a stuck-open relief valve given a transient initiator would increase due to an increase in the number of safety relief valve (SRV) cycles. Discuss how the common cause failure (CCF) probability has changed due to the EPU and what methodology was utilized to determine the new probability for stuck-open relief valves.

8. EPU SAR Attachment 11, Section 2.13.1.2.3 states that the pressure following a plant trip with an anticipated transient without scram (ATWS) post-EPU will increase, but the number of open SRVs is not expected to change. Discuss the basis for why SRV success criteria remains the same for ATWS transient events between pre- and post EPU. Describe any calculations and the methodology used for this assessment.

9. Describe any new operator actions developed due to the EPU, and describe the methodology utilized to determine the error probability associated with the new actions.

10. EPU SAR Attachment 11, Section 2.13.1.2.4 proposes a screening criteria for operator actions based on Fussel-Vesely (FV) > 0.01 and Risk Achievement Worth (RAW) > 2.

The staff finds an FV> 0.005 as a more appropriate criteria to identify important operator actions for EPU. Provide a change in human error probability (HEP) assessment pre and post-EPU for all operator actions impacted by the EPU that are either under 30 minutes; have an FV > 0.005; or have an RAW> 2.

11. EPU SAR Attachment 11, Table 2.13-7 presents an operator action evaluation pre- and post-EPU. This table does not provide the basis or methodology for this analysis, nor parameters or models used. Submit an updated table including additional results from RAI Question B.1a that includes the methodology utilized to determine the updated HEP results.

12. EPU SAR Attachment 11, Table 2.13-2, Element IE, Sub-element 7 highlights exclusion of Break Outside Containment (BOC) as an initiator. Parts of the secondary systems could be considered wear areas in the flow accelerated corrosion (FAC) program.

Provide a more thorough assessment of why BOC is not included as an initiator and the potential risk impacts due to EPU.

13. EPU SAR Attachment 11, Table 2.13-2, Element DA, Sub-element 15 states that the probability of a stuck open relief valve (SORV) conditional on its need to open for various transient initiators is not modeled. The staff does not understand how transients with an SORV are accounted for in the NMP2 PRA. The EPU increases the probability of SORV openings, therefore; provide a more thorough explanation explaining how SORV conditional probability is modeled in the NMP2 PRA and its EPU impact.

14. EPU SAR Attachment 11, Section 2.13.1.3.1 provides information on seismic events.

Explain and provide a seismic risk impact of any new vulnerabilities due to EPU

-4 implementation, and whether modifications will affect structures or component anchoring mechanisms.

15. EPU SAR Attachment 11, Section 2.13.1.4 discusses EPU impact on shutdown. The attachment states there is negligible impact on suppression pool cooling capacity, blow down loads, and reactor pressure vessel (RPV) over pressure success criteria. Discuss the risk impact of SRV actuation success criteria during shutdown post-EPU. In addition:

a. Explain how the EPU affects the scheduling of outage activities.

b. Provide additional information regarding the reliability and availability of equipment used for shutdown operations.

c. Explain how the EPU affects the availability of equipment or instrumentation used for contingency plans.

d. Explain how the EPU affects the ability of the operator to close containment.

e. Describe the plant's shutdown risk management philosophies, processes, and controls relied upon to ensure that the risk impact of the EPU for shutdown operations is not significant.

C. Electrical Engineering

1. In Attachment 8, the licensee provided an evaluation of grid stability based on the EPU of NMP2. To account for seasonal variations in station performance, the maximum output of the station used for the grid stability study was 1380 MWe at 233 MVAR lagging (vs. 1368.9 MWe at 278 MVAR lagging). Provide a detailed discussion as to how the reactive output values are developed, including whether or not turbine-generator equipment conditions are factored into these values (de-rating). In addition, if turbine generator equipment conditions (other than seasonal variations) during operation result in restrictions of generator output (real and/or reactive capability less than reactive capability curve), provide a detailed discussion as to how this information impacts the operational grid contingency analyses. Also, provide a detailed discussion regarding the capability of the main generator to supply reactive output in accordance with the reactive capability curve.

2. In Attachment 8 of the May 27,2009, LAR, the licensee states that because the EPU will be accomplished using the existing main generator and step-up transformers, there will be no change in the short circuit current contribution of NMP2 to system faults, and the overall system fault duty is not impacted. Provide a detailed discussion as to the basis for that statement, including whether actual fault studies for overall system fault duty were conducted using the increased inertia constant for the new high pressure rotor.

3. Provide a list and description of components being added to your 10 CFR 50.49 program due to the proposed EPU (if applicable). Confirm that these components are qualified for the environmental conditions they are expected to be exposed to.

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4. In Attachment 3 of the May 27,2009, LAR, the licensee states that shielding will be installed to reduce post-accident dose to components in Group III, enough to extend the qualified life sufficient to meet environmental qualification (EO) program requirements.

Provide a detailed discussion of the specific shielding designs addressing how a 32%

reduction (minimum) in exposure is achieved, and the resulting doses to Group III equipment (demonstrating qualification of said equipment).

5. In Attachment 3, the licensee provided a worst case Plant EO Enveloping Temperature profile for all zones and elevations. Provide a more detailed discussion and profiles, addressing accident pressure and humidity for all zones and elevations.

D. Reactor Systems

1. Figure 2.8-22 provides a plot of the main steam isolation valve closure with flux scram transient. The Vessel Pressure Rise is plotted in pounds per square-inch (psi), but scaled in percent rated (of something). Please quantify the bases, e.g., rated power (MWth) or nominal pressure (psi), for the plots of Vessel Pressure Rise and other parameters in this Figure.

2. Explain why the vessel steam flow is oscillating while all other parameters appear to hold comparatively steady throughout the main steam isolation valve closure with flux scram transient.

3. Summarize the results of the plant-specific loss of feedwater (LOFW) analysis.

(a) Address the modeling tool used, describe the sequence of events, discuss any additional failures assumed - beyond the high-pressure coolant injection (HPCI)

- in the analysis and provide plots of the significant parameters to demonstrate performance relative to the applicable acceptance criteria.

(b) Compare analytic assumptions to those used in the current licensing basis (CLB).

If applicable, provide equipment out-of-service assumptions used in this analysis and discuss whether these assumptions change in the EPU analysis as compared to the CLB analysis.

(c) Verify that any proposed changes in the Technical Specifications (TSs) will reflect the analysis assumptions used in the EPU analyses.

4. Provide plots for the limiting ATWS event (for both EPU and current licensed thermal power (CLTP) conditions) consistent with Tables 2.8-5 and 2.8-6 of NEDC-33351P, which show the bottom vessel pressure and indicate when the Standby Liquid Control (SLC) System starts injecting.

5. With respect to the ATWS analysis results, indicate when the emergency core cooling system (ECCS) begins to deliver flow, and evaluate its effect upon core boron concentration and core reactivity.

6. Table 2.8-12 of NEDC-33351P shows the Boral rack results based on zero degradation penalty. Are there any Boraflex racks in use at NMP2 to store new or spent fuel?

7. Does the maximum k-effective (k-eff) shown in Table 2.8-12 of NEDC-33351P account for the limiting accident scenario? What is the limiting accident considered?

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8. Show that the analysis of the limiting accident scenario considered the bounding depletion parameter values relative to projected EPU operating conditions. The response should consider the effects of applicable depletion parameters, such as the moderator density, fuel temperature, specific power and operating history, and burnable poisons.

9. What methods and analytic codes will be used to evaluate the LOFW heating (LOFWH) for EPU reload licensing analyses at NMP2?

10. Evaluate the LOFWH transient at EPU conditions at NMP2 to demonstrate conformance to fuel thermal-mechanical acceptance criteria. Provide transient results including those pertaining to fuel thermal mechanical performance, and specifically address the potential for pellet-cladding interaction and pellet-cladding mechanical interaction.

11. What methods and analytical codes will be used to analyze the Rod Withdrawal Error event both at low-power and at power for the NMP2 EPU reload licensing analysis?

12. What methods and analytical codes will be used in the reload licensing analysis to predict the fuel and system response to a Control Rod Drop Accident at EPU conditions?

13. Do the calculated cladding oxidation levels account for pre-existing oxidation? If not, what amount of pre-existing oxidation will exist on the limiting bundle? How much pre existing oxidation will exist on the more highly exposed bundles? Does the transient oxidation result presented consider oxidation on both surfaces of the fuel cladding?

14. In Section 2.8.4.4, it is stated that, "containment pressures for EPU events increased slightly above the CLTP analyzed pressures, but remained below the existing TS peak containment internal pressure, Pa." What is the value that precedes Pa?

15. In Section 2.8.4.6, it is stated that, "the NMP2 recirculation loop jet pump flow mismatch TS limits do not change because these limits are based on rated core flow, which is not affected by EPU, and the flow mismatch limits are not affected because a detailed ECCS evaluation was not required." Describe how it was determined that a detailed ECCS evaluation was not required.

16. If screening criteria are employed in the reload analyses, provide these screening criteria. If screening criteria are exceeded on a cycle-specific basis, what recourse is taken to demonstrate compliance?

17. Indicate whether credit was taken for containment backpressure in the loss-of-coolant accident (LOCA) analyses of record, and whether the said credit will be used in the EPU LOCA analyses.

E. Thermal and Hydraulic Design

1. With respect to the proposed EPU, provide a description of:

(a) the implementation status of the Long-Term (LIT) Stability Solution in NMP2, and any effects or impacts of the EPU on the LIT stability implementation (b) the power range neutron monitoring system, oscillation power range monitor

-7 (OPRM) operability and surveillance requirements, and the backup stability protection (SSP) implementation in NMP2.

(c) the NMP2 OPRM operating experience and the lessons learned from the 2003 instability event.

2. For the SSP calculations, describe how the stability curves for scram region and controlled entry region shown in Figure 2.8-21 of NEDO-33351 P are calculated for EPU conditions. Specifically, provide the associated feedwater temperature assumptions that allow the use of the same decay ratio criteria shown in Table 2.8-2 for the scram and exit boundary.

3. Will the Option III hardware implemented in NMP2 have the DSS-CD software installed for testing purpose? What are the testing plans?

4. Will the Delta CPR/Initial CPR Versus Oscillation Magnitude (DIVOM) curve be implemented as cycle-specific in NMP2? If the generic DIVOM slope will not be used, provide a reference to the cycle-specific DIVOM analysis methodology that will be used.

5. In September 2006, the Hope Creek plant experienced a half-scram indication from the Option III hardware while withdrawing peripheral control rods in low-power bundles.

Hope Creek implemented recommendations for speed of rod withdrawal inside the armed region. Have these recommendations been incorporated in the NMP2 operator training?

6. Assuming a conservative OPRM setpoint of 1.15, provide the hot-spot fuel temperature as function of time before the scram. Evaluate this fuel temperature oscillation against pellet-clad interaction (PCI) limits. Assume the steady-state fuel conditions before the oscillations are those of point ICA-A1 of Figure 2.8-21 of NEDC-33351 P (the highest power point in the SSP scram region).

7. Provide plant-specific information relevant to ATWS, specifically:

(a) location of the boron injection (b) a description of the SLC system actuation logic and its operability requirements, (c) boron enrichment level (d) turbine bypass capacity (e) location of the steam extraction points for FW heaters.

8. Provide a summary of the ATWS emergency operating procedure (EOP) actions.

Provide justification that the EOPs are sufficient to suppress potential thermal-hydraulic instabilities during ATWS events.

9. The NMP2 EOPs quote a 25 degree Fahrenheit (OF) value for the increase in suppression pool temperature during a reactor pressure vessel ATWS blowdown.

Provide the bases for this 25 of value.

10. Describe any effects or impacts, if any, of the proposed EPU on suppression pool cooling, EOPs, or ECCS (NPSH) requirements

11. Please provide a short description of the simulator neutronic core model.

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12. During an October 28,2009, staff audit, the NRC staff observed a series of ATWS scenarios in the NMP2 training simulator. The EPU power ATWS scenario was limited because the simulator had not been fully programmed for these conditions. When the NMP2 simulator is upgraded for EPU conditions, provide results of an MSIV isolation ATWS scenario at both CLTP and EPU conditions to compare the impact of EPU conditions.

13. Provide a list of approved methodologies used to support the calculation for Section 2.8.3 of the proposed EPU LAR.

F. Health Physics and Human Performance

1. In Section 2.11.1.1 (page 2-373) of NEDO-33351, "Safety Analysis Report for Nine Mile Point Nuclear Station Unit 2 Constant Pressure Power Uprate (PUSAR) (non-proprietary version)," it is stated that the abnormal operating procedure for feedwater failure, N2 SOP-6, be changed such that the reactor recirculation runback logic will be modified to initiate the runback immediately upon a feedwater pump trip. What does this mean to the operator and the associated times available and time required for the operator to complete the actions required in this procedure?

2. In Section 2.11.1.1 (page 2-373) of NEDO-33351 PUSAR, it is stated that the Loss of Condenser Vacuum abnormal operating procedure, N2-S0P-9, will be revised such that the Turbine Back Pressure Alarm Limit for EPU full power operation will be slightly less restrictive and will be incorporated in the Loss of Vacuum SOP. The alarm set point will be changed to allow operation closer to the trip set point. What is the margin between the new alarm setpoint proposed under EPU conditions and the trip setpoint under EPU conditions? What, if any, is the impact on the operator's ability to complete the necessary actions before the trip setpoint?

3. In Section 2.11.1.1 (page 2-373) of NEDO-33351 PUSAR, it is stated that the Main Condenser Tube Rupture / Condensate High Conductivity abnormal operating procedure, N2-S0P-10, has been revised. Steps were added to verify closure of the condensate demineralizer bypass valve installed as part of EPU modifications. Does this change mean that the current procedures have a step that requires the operator to verify the closure of the original bypass valves, and the change to this procedure for proposed EPU conditions will add a step for the operator to verify the state of the heater drain pump? Will the procedure be updated with a provision that states that if the heater drain pump is out of service, that instead of verifying closure of the original bypass valves, the operator should verify the EPU bypass valves are closed? In the aforementioned situation, would the operators be verifying that the original bypass valves are opened, so that the condensate pre-filters are not bypassed? Please clarify the changes to operator actions.

4. In Section 2.11.1.2 (pages 2-373 and 2-374) of NEDO-33351 PUSAR, it is stated that the Combustible Gas Control in Containment scenario assumes that the operators actuate containment sprays within 30 minutes after the LOCA (UFSAR 6.2.5.1). It further states in this section that operators initiate the hydrogen recombiners in 32.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The hydrogen recombiner initiation time is reduced by 10.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. CLTP initiation time is 43.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. This time reduction is due to a change in the analyzed value of percent of hydrogen and oxygen concentrations in the containment at which the recombiner system

- 9 is started following the accident. How many other actions are the operators expected to complete during this time span which has been reduced by 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. Was this time change analyzed with regards to the ability of operators to complete the necessary actions? Was there a validation process associated with this time change? Is it assumed that the operators will initiate core spray and initiate the hydrogen combiners in the same sequence of actions?

5. In Section 2.11.1.2 (2-374) of NEDO-33351 PUSAR, it is stated that the EOPs are symptom-based procedures. These procedures, as written, do not have specific time constraints, or time limits associated with their execution. EPU conditions will result in greater decay heat loads. The actual EOP actions performed by operators are not changed. Those actions that remove decay heat will be influenced. Have these actions been identified, and how will they be influenced by the proposed EPU conditions? It is also stated in this section that the NMP2 simulator will be modified for EPU changes.

Will these modifications occur prior to implementation? Have the changes been identified? This section of NEDO-33351 PUSAR also states decay heat changes and their effects on EOP execution have been reviewed, and appropriate training will be provided to the operators. UFSAR action times that would be executed under the guidance of EOPs are included in the evaluation of this section. Does this evaluation of UFSAR action times refer to the entire list of changes to operator actions sensitive to power uprate or is this a reference to an evaluation located in the UFSAR. Please clarify.

G. Vessels & Internals Integrity

1. The submittal refers to BWRVIP-74 on pages 2-3 and in Table 2.1-3. Is BWRVIP 74 the proper reference or should it be referred to the SER for BWRVIP-05? Does NMP2 plan to submit an updated request for relief reflecting the data shown in Table 2.1-3, which will reflect the higher fluence at 54 EFPY for the EPU conditions? If so, please provide as a regulatory commitment.

2. Since the applicant stated that it is incorporating hydrogen water chemistry (HWC) and noble metal chemical addition (NMCA) programs, the staff requests the licensee to identify the method of controlling HWC/NMCA in the reactor vessel. Provide details on the methods for determining the effectiveness of HWC/NMCA by using the following parameters:

(a) electrochemical potential (ECP),

(b) feedwater hydrogen flow, (c) main steam oxygen content, and (d) hydrogen loxygen molar ratio

December 23, 2009 Mr. Samuel L. Belcher Vice President Nine Mile Point Nine Mile Point Nuclear Station, LLC P.O. Box 63 Lycoming, NY 13093 SUB..IECT: REQUEST FOR ADDITIONAL INFORMATION REGARDING NINE MILE POINT NUCLEAR STATION, UNIT NO.2 - RE: THE LICENSE AMENDMENT REQUEST FOR EXTENDED POWER UPRATE OPERATION (TAC NO. ME1476)

Dear Mr. Belcher:

By letter dated May 27,2009, as supplemented on August 28,2009, Nine Mile Point Nuclear Station, LLC, submitted for Nuclear Regulatory Commission (NRC) staff review and approval, a proposed license amendment requesting an increase in the maximum steady-state power level from 3467 megawatts thermal (MWt) to 3988 MWt for Nine Mile Point Unit NO.2 extended power uprate operation.

The NRC staff is reviewing the information provided in that letter and has determined that additional information is needed to support its review. Enclosed is the NRC staff's request for additional information (RAI). The RAI was discussed with your staff on December 14,17, and 21, 2009, and it was agreed that your response would be provided within 60 days from the date of this letter.

Sincerely, IRA!

Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-410

Enclosure:

As stated cc w/encl: Distribution via Listserv DISTRIBUTION:

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