ML14126A809

Issuance of Amendment Regarding Revisions to the Containment Analysis Licensing Basis

ML14126A809
Person / Time
Site:	Indian Point
Issue date:	07/16/2014
From:	Pickett D Plant Licensing Branch 1
To:	Entergy Nuclear Operations
Pickett D, NRR/DORL/LPLI-1, 415-1364
References
TAC MF0590, FOIA/PA-2016-0148
Download: ML14126A809 (10)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 16, 2014

SUBJECT:

INDIAN POINT NUCLEAR GENERATING UNIT NO.2- ISSUANCE OF AMENDMENT REGARDING REVISIONS TO THE CONTAINMENT ANALYSIS LICENSING BASIS (TAC NO. MF0590)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 276 to Facility Operating License No. DPR-26 for the Indian Point Nuclear Generating Unit No.2 (IP2). The amendment consists of changes to the Updated Final Safety Analysis Report (UFSAR) in response to your application dated January 28, 2013, as supplemented by letters dated August 21, 2013, and April 22, 2014.

The amendment authorizes revisions to the IP2 UFSAR to credit four rather than three containment fan cooler units in the containment integrity analysis. A re-analysis of the large break loss-of-coolant accident was performed to correct methodology errors in the long-term mass and energy releases for the containment integrity analysis and crediting four containment fan cooler units for the limiting single failure is necessary to maintain the peak containment pressure within the current analysis of record.

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, Douglas V. Pickett, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-24 7

Enclosures:

1. Amendment No. 276 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 AND ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-247 INDIAN POINT NUCLEAR GENERATING UNIT NO.2 AMENDMENT TO FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATIONS Amendment No. 276 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 January 28, 2013, as supplemented on August 21, 2013, and April 22, 2014, 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 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) Technical Specifications The Technical Specifications contained in Appendices A, Band C, as revised through Amendment No. 276, are hereby incorporated in the license. ENO shall operate the facility in accordance with the Technical Specifications.

3. Further, Facility Operating License No. DPR-26 is hereby amended to authorize a change to the Updated Final Safety Analysis Report (UFSAR) crediting four rather than three containment fan cooler units for the containment integrity analysis as set forth in the license amendment application dated January 28, 2013, as supplemented by letters dated August 21, 2013, and April 22, 2014, and evaluated in the safety evaluation dated July 16, 2014. The licensee shall update the UFSAR by adding a description of this change, as authorized by this amendment, and in accordance with 10 CFR 50.71(e).

4. This license amendment is effective as of the date of its issuance and shall be implemented within 30 days. The UFSAR changes shall be implemented in the next periodic update to the UFSAR in accordance with 10 CFR 50.71 (e).

FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License Date of Issuance: July 16, 2014

ATTACHMENT TO LICENSE AMENDMENT NO. 276 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

instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, Amdt. 42 to receive, possess, and use in amounts as required any 10-17-78 byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5) ENO pursuant to the Act and 10 CFR Parts 30 and 70, to Arndt. 220 possess, but not separate, such byproduct and special 09-06-01 nuclear materials as may be produced by the operation of the facility and Indian Point Nuclear Generating Unit No. 3 (IP3).

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 1: 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 ENO is authorized to operate the facility at steady state Arndt. 241 reactor core power levels not in excess of 3216 10-27-2004 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendices A, B, and C, as revised through Amendment No. 276, are hereby incorporated in the license. ENO 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 TableR, "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. 276

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 276 TO FACILITY OPERATING LICENSE NO. DPR-26 ENTERGY NUCLEAR INDIAN POINT 2, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

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

1.0 INTRODUCTION

By letter dated January 28, 2013 (Reference 1, Agencywide Documents Access and Management System (ADAMS) Accession No. ML13042A243), as supplemented by letters dated August 21, 2013 (Reference 2, ADAMS Accession No. ML13239A477), and April 22, 2014 (Reference 6, ADAMS Accession No. ML14121A113), Entergy Nuclear Operations, Inc.,

the licensee, submitted a request for changes to the Indian Point Nuclear Generating Unit No.2 (IP2) Updated Final Safety Analysis Report (UFSAR) and Technical Specification (TS) Bases.

The changes modify the licensing basis to take credit for four, rather than three, containment fan cooler units (FCUs) for the containment integrity analysis. A re-analysis of the large-break loss-of-coolant accident (LBLOCA) was performed to correct methodology errors in the long-term mass and energy (M&E) releases for the containment integrity analysis and crediting four containment fan cooler units for the limiting single failure is necessary to maintain the peak containment pressure within the current analysis of record.

The supplemental letters dated August 21, 2013, and Apri122, 2014, 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) staff's 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 discussed in the 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 (1 0 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.

As discussed in the UFSAR, the licensee for IP2 has made some changes to the facility over the life of the unit that 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.

The NRC staff's acceptance criteria for the primary containment functional design are based on the following GDCs in 10 CFR 50, Appendix A:

GDC-4, "Environmental and dynamic effects design bases," insofar as it requires that Structures, Systems, and Components (SSCs) important to safety be designed to accommodate the effects of, and to be compatible with, the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents and that such SSCs be protected against dynamic effects.

GDC-16, "Containment design," insofar as it requires that the containment and associated systems be designed to establish an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment, and to assure that the containment design conditions important to safety are not exceeded as long as postulated accident conditions require.

GDC-38, "Containment Heat Removal," insofar as it requires that a Containment Heat Removal System (CHRS) be provided and that its function shall be to rapidly reduce the containment pressure and temperature following a LOCA and maintain them at acceptably low levels.

GDC-50, "Containment Design Basis," insofar as it requires that the containment and its associated heat removal systems be designed so that the containment structure can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated temperature and pressure conditions resulting from any LOCA.

NUREG-0800, "Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Section 6.2.1.3, "Mass and Energy Release Analysis for Postulated Loss-of-Coolant Accidents (LOCAs)," provides criteria for M&E release analysis.

SRP Section 6.2.1.1.A, "PWR Dry Containments, Including Subatmospheric Containments,"

provides criteria for containment integrity analysis for large dry containments In addition, the licensee cited the containment integrity analysis of record for Indian Point Unit No.3 (IP3) as a precedent. Both IP2 and IP3 have similar equipment design and the NRC staff has previously reviewed and approved the IP3 containment integrity analysis that credits four containment FCUs.

3.0 TECHNICAL EVALUATION

IP2 is a pressurized-water reactor (PWR) having a large dry containment. The containment building is a reinforced concrete axisymmetric right vertical cylindrical structure with a flat base

and hemispherical dome lined with a steel plate on the inside. The design objective of the containment structure is to contain all radioactive material, which might be released from the core following a loss-of-coolant accident (LOCA). The containment serves as both a biological shield and a pressure container.

3.1 Mass and Energy Release Analysis Nuclear Safety Advisory Letter 11-05 (Reference 3) identified Westinghouse methodology errors in the long-term M&E releases during a LBLOCA. The licensee identified the following issues in the current long-term LOCA M&E release analysis that used the Westinghouse containment analysis methodology (Reference 4):

(a) The reactor vessel stored energy initially available in the M&E model was based on an incorrect mass of the vessel and barrel/baffle downcomer region.

(b) The long-term LOCA M&E release analysis was initialized at a low (non-conservative) steam generator secondary pressure condition.

(c) As described in the licensee's response to NRC's Containment and Ventilation Branch (SCVB) request for additional information (RAI) SCVB-RAI-8 in Reference 2, errors in the EPITOME computer code underestimated the long term M&E during the post-reflood phase of a LOCA. As a result, at the time of peak pressure, the energy release was underestimated by 3.1-percent and at 3600 seconds the energy release was underestimated by about 2.1-percent.

As described in Nuclear Safety Advisory Letter 11-05, the licensee confirmed that the main steam line break (MSLB) M&E accident analysis was not impacted and, therefore, there are no changes in the MSLB containment response.

The NRC staff requested the licensee to list the computer codes used for the corrected M&E release analysis during the various phases of LOCA (i.e., blowdown, refill, reflood and post-reflood) and justify their use if they differ from those used in the current licensing basis analysis.

In response to question SCVB-RAI-3 in Reference 2, the licensee stated that all computer codes used in the proposed analysis are the same as used in the current licensing basis analysis based on Reference 4 methodology except for the version number for the EPITOME code which is revised for error correction to address the issues in Nuclear Safety Advisory Letter 11-05.

3.2 Containment Integrity Analysis IP2 has three emergency diesel generators (EDGs) designated as DG 21, DG 22, and DG 23, three high head safety injection (HHSI) pumps, two residual heat removal (RHR) pumps, two recirculation pumps, two containment spray (CS) pumps, and five FCUs. DG 21 provides emergency power to one HHSI pump, one CS pump, one recirculation pump, and two FCUs.

DG 22 provides emergency power to one HHSI pump, one RHR pump, and two FCUs. DG 23 provides emergency power to one HHSI pump, one RHR pump, one CS pump, one recirculation pump, and one FCU.

The current containment integrity analysis assumes the single failure of one EDG concurrent with loss of offsite power (LOOP). The analysis is based on the most limiting failure of DG 23, thus making all equipment connected to DG 21 and DG 22 available including four FCUs. For additional conservatism, the current analysis only credits three FCUs instead of four. The licensee stated that a single failure of DG 21 or DG 22 will result in only three FCUs available.

However that would not present a more limiting case, because of the availability of two RHR pumps (single failure of DG 21) and two CS pumps (single failure of DG 22). The single failure of DG 23 is limiting because of the availability of one RHR pump and one containment spray pump only, and therefore four FCUs are available for this case.

The proposed containment integrity analysis also assumes the most limiting failure of DG 23 concurrent with a LOOP. The analysis credits the same equipment as the current analysis except it takes advantage of the fourth available FCU, and therefore credits four FCUs instead of three. The analysis uses the corrected M&E release data. In accordance with the response to SCVB-RAI-4 in Reference 2, the licensee performed the analysis using the currently used COCO computer code (Reference 5). The revised calculated peak containment pressure is 40.72 psig for the Double Ended Hot Leg (DEHL) break LOCA and 45.44 psig for the Double Ended Pump Suction (DEPS) break. The current analysis peak containment pressure is 40.62 psig for the DEHL break LOCA and 45.71 psig for DEPS break LOCA. All the calculated values are below the containment design pressure of 47 psig. The licensee stated that the analysis is consistent with the plant configuration for equipment availability and there are no proposed changes to design and operating procedures.

TS Limiting Condition for Operation 3.6.5 for IP2 specifies the range of containment average air temperature during operating Modes 1 through 4 to be greater than 50 °F and less than or equal to 130 °F. In response to a NRC staff RAI, the licensee performed a sensitivity study (Reference 6) to determine the impact on the peak containment pressure for the two extreme containment average air and heat sinks initial temperatures. For this study, the licensee analyzed the most limiting DEPS break LOCA as described above, and conservatively assumed the accumulator water temperature remained at 130 °F, and the Refueling Water Storage Tank (RWST) and service water temperatures remained at the TS maximum allowable value. The results show that the peak containment pressure, which occurs in the post-reflood period of the LOCA, is 45.44 psig at 1365 seconds assuming an initial containment temperature of 130 °F, and 41.58 psig at 1369.9 seconds assuming an initial containment temperature of 50 °F. The licensee stated that while the colder initial air temperature contributes to a higher air partial pressure, it is more than offset by the colder heat sinks which are able to absorb a greater amount of heat over the long-term transient. The licensee stated that using the maximum higher accumulator temperature and maximum RWST and service water temperatures is conservative because the lower temperatures would enhance the performance of the containment spray system and the FCUs and would result in a lower peak containment pressure.

The NRC staff requested that the licensee list all initial conditions used in the current most limiting LBLOCA analysis and the proposed analysis for the containment peak pressure and temperature response. The licensee was also requested to justify if any of the initial conditions in the proposed analysis is less conservative than the current analysis. In response to SCVB-RAI-1 in Reference 2, the licensee compared the initial conditions between the current most limiting licensing basis LBLOCA analysis against the proposed analysis. The licensee listed all the initial conditions and provided the following justification where conservatism is

reduced in the proposed analysis: (a) the emergency core cooling system (ECCS) flow during the recirculation mode increased from 1864 gpm to 2191 gpm as a result of refining the RHR system hydraulic analysis consistent with plant operations at the start of cold leg recirculation, (b) the total hot leg recirculation flow changed from 822 gpm to 1096 gpm, as a result of an error correction, (c) the RHR heat exchanger overall heat transfer coefficient changed based on the revised ECCS flow, (d) the RHR heat exchanger shell side flow changed from 1326 gpm to 1243 gpm due to error correction and Generic Safety Issue (GSI)-191 calculations, (e) the closed cooling water (CCW) overall heat transfer coefficient increased from 2.40 Btu/hr-°F to 2.41 Btu/hr-°F as a result of round-off, (f) the CCW heat exchanger shell side flow decreased from 4936 gpm to 4592 gpm for the future enhancement of the CCW pump performance margin, and (g) the miscellaneous heat load decreased from 19.675x1 06 Btu/hr to 18.9x1 06 Btu/hr as a result of reduction in the spent fuel pool heat load. The NRC staff finds the changes acceptable, as they represent operating experience, error corrections, future margin enhancements, and incorporation of results from generic safety issue (GSI)-191, "Experimental Studies of Loss-of-Coolant-Accident-Generated Debris Accumulation and Head Loss with Emphasis on the Effects of Calcium Silicate Insulation (NUREG/CR-6874, LA-UR-04-1227)."

The NRC staff requested that the licensee list all assumptions used in the current most limiting LBLOCA analysis and the proposed analysis for the containment peak pressure and temperature response. The licensee was also requested to justify if any of the assumptions in the proposed analysis is less conservative than in the current analysis. In response to SCVB-RAI-2 in Reference 2, the licensee stated that all assumptions used in the proposed analysis are typical in performing the LOCA M&E calculation per the NRC approved Reference 4 methodology. The containment code COCO is currently used to calculate the LOCA containment pressure and temperature responses as well as for containment minimum backpressure analysis for the LOCA analysis found in 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," and 10 CFR 50, Appendix K, "ECCS Evaluation Models." The licensee provided the following justification where conservatism is reduced in the proposed analysis: (a) the major assumption reducing conservatism is crediting four FCUs in the proposed containment analysis instead of three credited in the current analysis, and (b) ECCS recirculation and the timing provided by the RHR pumps have been refined due to GSI-191 considerations.

The NRC staff requested the licensee to describe the impact of changes in the M&E input to the containment gas temperature response for equipment environmental qualification (EEQ). In response to SCVB-RAI-5a in Reference 2, the licensee stated that in the proposed analysis, the containment peak gas temperature for EEQ is determined to be 266.46 °F, compared to the current containment peak gas temperature of 266.81 oF. The licensee stated that the temperature determined in the proposed analysis is less than the current value because of crediting an additional fan cooler which increases the heat removal rate and offsets the effect of the increased heat addition from the M&E release issues in Reference 3. Therefore, there is no impact on LOCA EEQ.

The NRC staff requested the licensee to describe the impact of changes in the M&E input on the containment wall temperature response under a LOCA. In response to SCVB-RAI-5b in Reference 2, the licensee stated that the LOCA containment liner temperature response was calculated with different initial conditions than those in the design basis analysis to address the temperature limits on its insulated and uninsulated portions. The maximum temperature for the uninsulated portions of the steel liner is 247 oF and the maximum temperature increase for the

insulated portion is 80 °F. The licensee's analysis of the liner temperature correcting the Reference 3 errors determined that the temperature of the uninsulated steel shell remains below 247 °F, and the increase of the insulated liner temperature is about 20 °F which is significantly less than 80 oF. Therefore there is no impact on the LOCA containment wall temperature response.

The NRC staff requested the licensee to describe the impact of the changes in M&E input to the following containment analyses: (a) sump water temperature response following a LOCA, and (b) net positive suction head (NPSH) analysis for CS and safety injection pumps during the recirculation phase. In response to SCVB-RAI-6a in Reference 2, the licensee provided a comparison of sump water temperature response of the current and the proposed analysis for a LBLOCA with minimum ECCS flows. The comparison showed that the sump water temperature is slightly lower in the proposed analysis, and therefore does not impact the GSI-191 analyses.

During the recirculation phase of a LOCA, the recirculation and RHR pumps provide CS and cold leg recirculation flow along with the HHSI pumps. In the proposed LOCA containment analysis, the maximum sump temperature in the recirculation mode is determined to be 256.68 °F which is less than the sump temperature of 264.4 °F assumed for the recirculation and RHR pumps available NPSH analysis. The NPSH analysis did not credit containment accident pressure above the saturation pressure at the sump water temperature. The licensee determined that the proposed analysis does not impact the GSI-191 NPSH analyses because the change in sump water level in the proposed analysis was small and bounded by the water level assumed in the GSI-191 NPSH analysis.

The NRC staff requested the licensee to describe the impact of the changes in the M&E release on the minimum containment pressure analyses for ECCS performance using the guidance in SRP 6.2.1.5, "Minimum Containment Pressure Analysis for Emergency Core Cooling System Performance Capability Studies," and Regulatory Guide 1.157, "Best-Estimate Calculations of Emergency Core Cooling System Performance," Section 3.12.1. In response to SCVB-RAI-7 in Reference 2, the licensee stated that the Reference 4 methodology codes, including EPITOME, are not used in performing either Appendix K large break analyses or the best estimate large break analyses and therefore the issues identified in Reference 3 have no effect on methods used to calculate the minimum containment backpressure for the ECCS performance analysis.

According to TS Section 5.5.14b, the licensee is using the containment design pressure of 47 psig as the current value of the 10 CFR 50 Appendix J containment leakage test pressure Pa.

Since the revised analysis results remain below the design pressure, the licensee's containment leakage rate test program is not affected.

Based on the above evaluations, the NRC staff has determined that the revised analyses, after correcting the M&E release errors, and changes based on operating experience, error corrections, future margin enhancements, and incorporation of results from GSI-191 calculations are acceptable because the licensee used NRC approved methodology and conservative inputs and assumptions and provided adequate justification for changes in the inputs and assumptions from the current licensing basis analysis.

3.3 Proposed Technical Specification Bases Changes The licensee also proposed changes to the TS Bases, a licensee controlled document, to support the proposed changes to the UFSAR. These changes included changes toTS Bases Sections 3.6.4, "Containment Pressure," 3.6.5, "Containment Air Temperature," and 3.6.6, "Containment Spray System and Containment Fan Cooler Unit (FCU) System." The NRC staff reviewed the proposed TS Bases changes and concludes that they are consistent and supportive of the proposed UFSAR changes. Thus, the staff has no comments.

3.4 Conclusion The NRC staff determined and agrees with the licensee that the proposed change, after correcting the M&E release data, meets the requirements of 10 CFR Part 50 Appendix A, (1) GDC-4, because the licensee showed that SSCs important to safety are designed to accommodate the effects of and are compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents and that such SSCs are protected against dynamic effects, (2) GDC 16, because the licensee showed that the containment design conditions important to safety are not exceeded during a postulated DBA, (3) GDC 38, because the licensee showed that the containment heat removal system would reduce the containment pressure and temperature rapidly, following a design-basis accident and would maintain them at acceptable levels and (4) GDC 50, because the licensee showed that the containment heat removal system is designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from a design basis LOCA. Therefore, the proposed license amendment is 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. 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 (78 FR 19749). 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.

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) there is reasonable assurance that 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 Contributor: Ahsan Sallman, NRR Date: July 16, 2014

7.0 REFERENCES

1. Letter from Entergy to NRC dated January 28, 2013, "Proposed Technical Specification Bases Changes to Credit Four Fan Cooler Units in Containment Integrity Analysis Indian Point Unit Number 2 Docket No. 50-247 License No. DPR-26", (ADAMS Accession Number ML13042A243).

2. Letter from Entergy to NRC dated August 21, 2013, "Response to Request for Additional Information Regarding Containment Integrity Analysis (TAC NOS. MF0590 and MF0591)

Indian Point Unit Numbers 2 & 3 Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64", (ADAMS Accession Number ML13239A477).

3. Nuclear Safety Advisory Letter, NSAL-11-05, "Westinghouse LOCA Mass and Energy Release Calculation Issues," July 26, 2011, (ADAMS Accession Number ML13239A479).

4. WCAP-10325-P-A, "Westinghouse LOCA Mass and Energy Release Model for Containment Design, March 1979 Version," (ADAMS Accession Number ML080640615).

5. WCAP-8327, "Containment Pressure Analysis Code (COCO)", July, 1974, (ADAMS Accession Number ML092460709).

6. Letter from Entergy to NRC dated April 22, 2014, "Response to Request for Additional Information Regarding Containment Integrity Analysis (TAC NOS. MF0590 and MF0591)

Indian Point Unit Numbers 2 and 3 Docket Nos. 50-247 and 50-286 License Nos.

DPR-26 and DPR-64" (ADAMS Accession Number ML14121A113).

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 REGARDING REVISIONS TO THE CONTAINMENT ANALYSIS LICENSING BASIS {TAC NO. MF0590)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 276 to Facility Operating License No. DPR-26 for the Indian Point Nuclear Generating Unit No. 2 (IP2). The amendment consists of changes to the Updated Final Safety Analysis Report (UFSAR) in response to your application dated January 28, 2013, as supplemented by letters dated August 21, 2013, and April 22, 2014.

The amendment authorizes revisions to the IP2 UFSAR to credit four rather than three containment fan cooler units in the containment integrity analysis. A re-analysis of the large break loss-of-coolant accident was performed to correct methodology errors in the long-term mass and energy releases for the containment integrity analysis and crediting four containment fan cooler units for the limiting single failure is necessary to maintain the peak containment pressure within the current analysis of record.

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, IRA/

Douglas V. Pickett, 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. 276 to DPR-26

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

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