ML090840073

Issuance of Amendment Emergency Diesel Generator Surveillance Test

ML090840073
Person / Time
Site:	Indian Point
Issue date:	04/22/2009
From:	Boska J Plant Licensing Branch 1
To:	Entergy Nuclear Operations
Boska J, NRR, 301-415-2901
References
TAC MD9214, FOIA/PA-2016-0148
Download: ML090840073 (15)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 22, 2009 Vice President, Operations Entergy Nuclear Operations, Inc.

Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249

SUBJECT:

INDIAN POINT NUCLEAR GENERATING UNIT NO.2 - ISSUANCE OF AMENDMENT RE: EMERGENCY DIESEL GENERATOR SURVEILLANCE TEST (TAC NO. MD9214)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 259 to Facility Operating License No.

DPR-26 for the Indian Point Nuclear Generating Unit NO.2. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated July 9, 2008, as supplemented by letters dated September 29, October 3, and October 8, 2008, and February 6, 2009.

The amendment revises the TSs by revising the test acceptance criteria specified in the TS surveillance requirement for the emergency diesel generator endurance test.

A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Sincerely, J hn P. Boska, Senior Project Manager lant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-247

Enclosures:

1. Amendment No. 259 to DPR-26

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY NUCLEAR INDIAN POINT 2, LLC ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-247 INDIAN POINT NUCLEAR GENERATING UNIT NO.2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 259 License No. DPR-26

1. The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Entergy Nuclear Operations, Inc. (the licensee) dated July 9, 2008, as supplemented by letters dated September 29, October 3, and October 8, 2008, and February 6, 2009, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-26 is hereby amended to read as follows:

- 2 (2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 259, are hereby incorporated in the license. ENO shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION Douglas V. Pickett, Acting Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: April 22, 2009

ATTACHMENT TO LICENSE AMENDMENT NO. 259 FACILITY OPERATING LICENSE NO. DPR-26 DOCKET NO. 50-247 Replace the following page of the License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page Insert Page 3

3 Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page Insert Page 3.8.1-8 3.8.1-8

- 3 instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

ENG pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive Amdt. 42 10-17-78 apparatus or components; (5)

ENG pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

Amdt. 220 09-06-01 C. This amended license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level ENG is authorized to operate the facility at steady state Amdt. 241 reactor core power levels not in excess of 3216 10-27-04 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendices A and S, as revised through Amendment No. 259, are hereby incorporated in the license. ENG shall operate the facility in accordance with the Technical Specifications.

(3) The following conditions relate to the amendment approving the conversion to Improved Standard Technical Specifications:

1. This amendment authorizes the relocation of certain Technical Specification requirements and detailed information to licensee-controlled documents as described in Table R, "Relocated Technical Specifications from the CTS," and Table LA, "Removed Details and Less Restrictive Administrative Changes to the CTS" attached to the NRC staffs Safety Evaluation enclosed with this amendment. The relocation of requirements and detailed information shall be completed on or before the implementation of this amendment.

Amendment No. 259

3.8.1 AC Sources - Operating SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.10

- NOTES

1.

Momentary transients outside the load and power factor ranges do not invalidate this test.

2.

This SR shall not normally be performed in MODE 1 or 2. However, this Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.

3.

If performed with DG synchronized with offsite power, it shall be performed at a power factor of ~ 0.88 for DG 21,

~ 0.87 for DG 22, and ~ 0.88 for DG 23. However, if grid conditions do not permit, the power factor limit is not required to be met. Under this condition the power factor shall be maintained as close to the limit as practicable.

Verify each DG operating at a power factor as stated in Note 3 operates for ~ 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s:

a.

For ~ 105 minutes and ~ 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> loaded

~ 2050 kW and ~ 2100 kW, followed by

b. For ~ 10 minutes and ~ 15 minutes loaded

~ 2270 kW and ~ 2300 kW, followed by

c.

For the remaining hours of the test loaded

~ 1700 kW and ~ 1750 kW.

24 rnonths'"

SR 3.8.1.11

. NOTE*

Load sequence timers associated with equipment that has automatic initiation capability disabled are not required to be OPERABLE.

Verify each load sequence timer relay functions within the required design interval.

24 months (1) The surveillance interval is extended, on a one time basis, to 48 months, with a 6 month grace period, following the testing in refueling outage 17 (spring 2006) based on testing performed under administrative controls in accordance with Administrative Letter 98-10 during refueling outage 18 (spring 2008) that satisfy the intent of the surveillance.

INDIAN POINT 2 3.8.1-8 Amendment No. 259

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 259 TO FACILITY OPERATING LICENSE NO. DPR-26 ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO.2 DOCKET NO. 50-247

1.0 INTRODUCTION

By letter dated July 9, 2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML081980160), as supplemented by letters dated September 29, 2008 (ADAMS Accession No. ML082760288), October 3, 2008 (ADAMS Accession No. ML082820162), October 8,2008 (ADAMS Accession No. ML082890535), and February 6,2009 (ADAMS Accession No. ML090620325), Entergy Nuclear Operations, Inc, (the licensee) requested an amendment to Appendix A, Technical Specifications (TSs), of the Facility Operating License for Indian Point Nuclear Generating Unit NO.2 (IP2). The proposed changes would revise the test acceptance criteria specified in TS Surveillance Requirement (SR) 3.8.1.10 for the emergency diesel generator (EDG) endurance test surveillance. Specifically, the licensee has proposed revising the load ranges and power factors specified for the endurance test for consistency with the associated plant safety analyses. The supplements dated September 29, October 3, and October 8, 2008, and February 6, 2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration.

2.0 REGULATORY EVALUATION

The following explains the applicability of General Design Criteria (GDC) for IP2. The construction permit for IP2 was issued by the Atomic Energy Commission (AEC) on October 14, 1966, and the operating license was issued on September 28, 1973. The plant GDC are listed in the Updated Final Safety Analysis Report (UFSAR) Chapter 1.3, "General Design Criteria,"

with more details given in the applicable UFSAR sections. The AEC published the final rule that added Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," in the Federal Register (36 FR 3255) on February 20, 1971, with the rule effective on May 21, 1971. In accordance with an NRC staff requirements memorandum from S. J. Chilk to J. M. Taylor, "SECY-92-223 - Resolution of Deviations Identified During the Systematic Evaluation Program," dated September 18, 1992 (ADAMS Accession No. ML003763736), the Commission decided not to apply the Appendix A GDC to plants with construction permits issued prior to May 21, 1971. Therefore, the GDC which constitute the licensing bases for IP2 are those in the UFSAR.

- 2 As discussed in the UFSAR, the licensees for IP2 have made some changes to the facilities over the life of the units that have committed to some of the GDCs from 10 CFR Part 50, Appendix A. The extent to which the Appendix A GDC have been invoked can be found in specific sections of the UFSAR and in other IP2 licensing basis documentation, such as license amendments.

Based on a review of UFSAR Section 8.1, the NRC staff identified the following UFSAR GDCs as being applicable to the proposed amendment:

An emergency power source shall be provided and designed with adequate independency, redundancy, capacity, and testability to permit the functioning of the engineered safety features and protection systems required to avoid undue risk to the health and safety of the public. This power source shall provide this capacity assuming a failure of a single component (GDC 39 and GDC 24).

The following NRC requirements and guidance documents are also applicable to the NRC staff's review of the licensee's amendment request:

Paragraph 50.36(c)(2)(ii) of 10 CFR, "Technical specifications," requires that "[a] technical specification limiting condition for operation [LCO] of a nuclear reactor must be established for each item meeting one or more of the [criteria set forth in 10 CFR 50.36(c)(2)(ii)(A)-(D)]."

Paragraph 50.36(c)(3) of 10 CFR, "Technical specifications," requires that TSs include SRs, which "are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met."

While IP2 is not currently committed to either of the following NRC guidance documents, the NRC staff used these documents as a technical reference during its review of this license amendment request (LAR):

Safety Guide 9, "Selection of Diesel Generator Set Capacity for Standby Power Supplies,"

describes an acceptable basis for the selection of EDG sets of sufficient capacity and margin to implement GDC 17.

Regulatory Guide (RG) 1.9, "Selection, Design, Qualification, and Testing of Emergency Diesel Generators Units Used As Class 1E Onsite Electric Power Systems at Nuclear Power Plants,"

provides guidance with respect to design and testing of safety-related EDGs.

3.0 TECHNICAL EVALUATION

3.1 Description of the IP2 Alternating Current (AC) Electrical Power System The IP2 engineered safety feature (ESF) buses include 480 volt (V) switchqear buses 2A, 3A, 5A, and 6A and numerous motor control center buses. The 480 V ESF switchgear buses are normally supplied from the non-safety 6.9 kilovolt (kV) buses.

The required ESF equipment circuits are supplied electrical power from the 480 V ESF switchgear buses. The normal source of power for buses 5A and 6A is the 138 kV system (via

- 3 the station auxiliary transformer, 6.9 kV buses 5 and 6, and station service transformers); since the normal source of power to these buses is not the main generator, no transfer is required in the event of a generator trip. Buses 2A and 3A are normally supplied from 6.9 kV buses 2 and 3. During power operation, 6.9 kv buses 2 and 3 are normally supplied from the unit auxiliary transformer, which is supplied from the main generator. In the event of a generator trip, there is a "dead-fast" transfer of the 6.9 kV buses to the station auxiliary transformer. In the event of a loss of offsite power, one EDG provides emergency power to bus 5A, one to bus 6A, and the third EDG to buses 2A and 3A. Each EDG automatically starts on a safety injection signal or upon an undervoltage condition on the associated 480 V switchgear bus.

Loads required for safe shutdown and accident mitigation are supplied from the 480 V ESF switchgear buses and from certain 480 V motor control centers. In the event of loss-of-offsite power (LOOP), all loads are stripped from the 480 V buses, the EDGs are started, and required loads are sequenced on to the ESF 480 V buses.

The EDG sets (EDGs 21,22, and 23) provide three sources of onsite emergency electrical power. Each set is an Alco Model 16-251-E engine coupled to a Westinghouse 900 revolution per minute, 3-phase, 60-cycle, 480 V generator. According to the IP2 UFSAR, the EDG units are capable of supplying 1750 kilowatts (kW) (continuous), 2300 kW for a half hour in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, and 2100 kW for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in any 24-hour period. There is a sequential limitation whereby it is unacceptable to operate an EDG for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at 2100 kW followed by operation at 2300 kW for a half hour. The IP2 UFSAR further notes that operation at any other combination of the above ratings is acceptable.

Any two IP2 EDGs are capable of sequentially starting and supplying the power requirement of at least one complete set of safeguards equipment. The EDGs are installed in a seismic class 1 structure. The EDGs are capable of starting and load sequencing within 10 seconds after the initial start signal. The EDGs have the capability of being fully loaded within 30 seconds after the start-of-Ioad sequencing.

3.2 Evaluation of Proposed Changes During an inspection at IP2, NRC inspectors questioned the adequacy of the load ranges specified in SR 3.8.1.10 to demonstrate the capability of the EDGs to operate at the peak loading conditions identified in the plant safety analyses for the limiting design-basis accident (DBA). As a result of the inspection findings, the licensee acknowledged the need to submit a license amendment request to establish new load ranges that would envelop the peak accident loads. While reviewing the above changes, the licensee also determined that a change to the power factor test value would also be appropriate.

In its letter dated July 9, 2008, the licensee proposed a license amendment to the TSs for IP2.

In its September 29, 2008, response to an NRC staff request for additional information (RAI), the licensee modified its initial license amendment request. The proposed changes would revise the test acceptance criteria specified in TS SR 3.8.1.10 for the EDG endurance test surveillance.

Specifically, the licensee has proposed revising the load ranges and power factor values specified for the endurance test for consistency with the associated plant safety analyses.

The NRC staff noted that the timing of operator manual actions during an accident would affect the EDG loading. In an RAI dated January 9, 2009, the NRC staff requested that the licensee

- 4 describe how the timing assumed for manual actions was confirmed to be appropriate for IP2.

In a response dated February 6, 2009, the licensee replied that when the initial EDG loading study was performed (WCAP-12655, "Emergency Diesel Generator Loading Study for Indian Point Unit 2), operators verified the times that had been used in the WCAP. Since there have been some changes to the EDG loading study and to the emergency operating procedures since that time, in December 2008, the licensee evaluated the timing of operator manual actions using the IP2 plant-reference simulator. A large loss-of-coolant accident (LOCA) scenario was run, with LOOP, safety injection, containment spray actuation, and the loss of EDG-23. The licensee reported that the timing of the operator manual actions was monitored, and was consistent with those assumed in the EDG loading study. The NRC staff notes that the scenario selected provided an acceptable exercise of the operator manual actions to confirm that the manual actions are feasible and consistent with the assumptions of the loading study.

The NRC staff reviewed and evaluated each of the proposed changes to the IP2 TSs as follows:

(The proposed TS changes are identified by underlined text for addition and strikeout text for deletion.)

3.2.1 TS 3.8.1 Change (1)

Note 3 of SR 3.8.1.10 will be modified as follows:

If performed with DG [diesel generator] synchronized with offsite power, it shall be performed at a power factor of ~ 0.88 for DG 21, ~0.87 for DG 22, and ~ 0.88 for DG 23~. However, if grid conditions do not permit, the power factor limit is not required to be met. Under this condition the power factor shall be maintained as close to the limit as practicable.

Evaluation of TS 3.8.1 Change (1)

SR 3.8.1.10 requires that each EDG be started and loaded for a specified period of time at specified loading conditions, which include kW output and power factor.

As mentioned previously, the EDG and associated electrical distribution is a 480 V system.

Since the 480 V loads that would be powered under an accident scenario are not able to be run during surveillance testing, the EDG loading is accomplished by connecting to the offsite electrical power grid. According to the licensee, this involves reversing the power flow in the transformers from the 480 V buses to the 6.9 kV buses and then additional step-up to either 13.8 kV or 138 kV, depending on the feeder circuits that are available between IP2 and the grid.

During its review of the electrical loading study to address the kW limit issue, the licensee determined that there is margin between the existing TS power factor test requirement and the analysis power factor for the limiting load scenarios. The licensee contends that the proposed change will eliminate unnecessary conservatism from the test and provide greater ability to perform the test without crediting the TS note regarding the limitations on power factor caused by grid conditions.

As part of its evaluation, the NRC staff reviewed the licensee's EDG loading studies (WCAP 12655, Revision 2, "Emergency Diesel Generator Loading Study for Indian Point Unit 2," and FEX-00083-00, "Dynamic Loading of Emergency Diesel Generators"), switchgear analysis, and power factor analysis for IP2. The staff's review of the IP2 EDG loading studies was limited to assessing the IP2 EDG and the associated switchgear margins that were available considering

- 5 design factors such as load size, loading sequence, and maximum operating temperature.

During its review, the NRC staff requested the licensee to explain (1) the discrepancy between the TS allowable frequency variation and the assumed frequency impact in WCAP-12655 and (2) the ability of the EDG to adequately perform its design function while loaded under the worst case frequency scenario (i.e., frequency variation of +2%). The NRC staff also requested the licensee to provide assurance that the EDG can perform its design function while loaded during the worst-case voltage scenario (i.e., 428 volts minimum or 500 volts maximum), as identified in TS SRs 3.8.1.2, 3.8.1.12, and 3.8.1.13. In response to this request, the licensee noted that the discrepancy with TS allowable frequency variation was noted in responses to a Condition Report IP2-2006-06850, and that a separate LAR is being processed to address required TS changes.

In the meantime, the licensee has implemented administrative controls to specify an acceptance criterion of +/- 0.3 hertz (Hz) (+/- 0.5%). The licensee stated that the voltage range is based on the EDG running in parallel mode (in parallel with the normal transformer power supply) and that during the EDG-only mode of operation the maximum voltage regulation would be 0.5% (2.4 V).

This variation should have very little effect on the performance of the EDG or on running motor loads. During accident conditions, the EDG only supplies bus loads using the EDG-only mode of operation.

3.2.2 TS 3.8.1 Change (2)

SR 3.8.1.10 will be modified as follows:

Verify each DG operating at a power factor as stated in Note 3 ~ operates for ~8 hours:

a.

For> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> loaded> 1837 kVV and < 1925 kW and For> 105 minutes and :s; 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> loaded> 2050 kW and < 2100 kW, followed by

b.

For the remaining hours of the test loaded> 1575 kVV and < 1750 kW. For> 10 minutes and < 15 minutes loaded> 2270 kW and < 2300 kW, followed by

c.

For the remaining hours of the test loaded ~ 1700 kW and :s; 1750 kW.

Evaluation of TS 3.8.1 Change (2)

As mentioned previously, the licensee acknowledged the need to submit an LAR to establish new load ranges that would envelop the peak accident loads as a result of an NRC inspection.

In its letter dated July 9,2008, the licensee proposed first loading the EDG at the 1/2 hour rating then at the 2-hour rating followed by operation at the continuous rating. Furthermore, the licensee proposed adding new Note A in TS Bases Section 3.8.1 that would state that operation at the overload ratings is allowed only for s 2300 kW (1/2 hour) followed by s 2100 kW (2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />),

not vice versa.

Based on its review of the EDG loading study for IP2 (WCAP-12655, Revision 2), the NRC staff noted that the peak load (approximately 2300 kW) can occur at different times during the LOCA scenario depending upon the size of break (approximately 40 minutes after the LOCA for a large-break LOCA and aporoximately 120 minutes after the LOCA for a small-break LOCA).

The NRC staff informed the licensee that the EDG and the associated switchgear must be capable of supplying the peak load at any time (i.e., loading order should not be a factor) during either LOCA scenario anclthat the EDG endurance run test load profile must envelop the design

- 6 basis assumptions (Le., the worst-case load profile) for the nuclear power plant. Subsequently, in letters dated September 29, 2008, and February 6, 2009, the licensee revised the proposed SR to envelop the peak loading conditions identified in the current version of the IP2 EDG loading study. The NRC staff understands that this profile is not representative of all accident loading scenarios. However, based on its review of the IP2 EDG loading study, the NRC staff finds that the revised SR is an acceptable representation of worst-case conditions. The NRC staff finds that the proposed change will ensure that the EDG and the switchgear associated with the output breaker of the EDG will perform their intended design function.

3.3 Conclusion The NRC staff's review of the IP2 EDG loading studies was limited to assessing the IP2 EDG and the associated switchgear margins that are available, considering design factors such as load size, loading sequence, and maximum operating temperature. The NRC staff finds that the proposed change will ensure that the EDG and the switchgear associated with the EDG output breaker will perform their intended design function.

Based on the above evaluation, the NRC staff concludes the proposed revisions to the IP2 TSs provide reasonable assurance of the continued availability of the required EDG capacity to maintain the reactor in a safe condition after an anticipated operational occurrence or a postulated DBA. The NRC staff also concludes that the proposed TS changes are in accordance with 10 CFR 50.36 and the requirements of UFSAR GDCs 24 and 39. Therefore, the NRC staff finds the proposed changes acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (73 FR 52416). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

- 7

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: M. McConnell R. Wolfgang Date: April 22, 2009

April 22, 2009 Vice President, Operations Entergy Nuclear Operations, Inc.

Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249

SUBJECT:

INDIAN POINT NUCLEAR GENERATING UNIT NO.2 - ISSUANCE OF AMENDMENT RE: EMERGENCY DIESEL GENERATOR SURVEILLANCE TEST (TAC NO. MD9214)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 259 to Facility Operating License No.

DPR-26 for the Indian Point Nuclear Generating Unit No.2. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated July 9, 2008, as supplemented by letters dated September 29, October 3, and October 8, 2008, and February 6, 2009.

The amendment revises the TSs by revising the test acceptance criteria specified in the TS surveillance requirement for the emergency diesel generator endurance test.

A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Sincerely, IRAI John P. Boska, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-247

Enclosures:

1. Amendment No. 259 to DPR-26

2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

See next page ADAMS ACCESSION NO'.. ML090840073

• See memo dated 3/20/09 OFFICE LPL1-1/PM LPL 1-1/LA EEEB/BC IOLB/BC OGC NAME

~IBoska SLittie GWilson*

LVick for NSalgado STurk DATE 4/20/09 4/6/09 3/20/09 4/9/09 4/20/09 OFFICE LPL1-1/BC(A)

NAME DPickett DATE 4/21/09 OFFICIAL RECORD COPY

DATED: April 22, 2009 AMENDMENT NO. 259 TO FACILITY OPERATING LICENSE NO. DPR-26 INDIAN POINT UNIT 2 DISTRIBUTION:

PUBLIC LPL1-1R/F RidsNrrDorl RidsNrrDorlLpl1-1 RidsNrrDorlDpr RidsNrrPMJBoska RidsNrrLASLittle (paper copy)

RidsNrrDeEeeb RidsNrrDciCptb RidsNrrDirslolb RidsNrrDirsltsb RidsRgn1 MailCenter RidsOGCRp RidsAcrsAcnw MailCTR MMcConnell RWolfgang GLapinsky cc: Distribution via Listserv