ML15342A398

Edwin I. Hatch Nuclear Plant, Unit No. 1, Issuance of Amendment Regarding Minimum Critical Power Ratio (CAC No. MF6681)

ML15342A398
Person / Time
Site:	Hatch
Issue date:	01/29/2016
From:	Martin R E Plant Licensing Branch II
To:	Pierce C R Southern Nuclear Operating Co
Martin R E
References
CAC MF6681
Download: ML15342A398 (15)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. C. R. Pierce Regulatory Affairs Director January 29, 2016 Southern Nuclear Operating

Company, Inc. Post Office Box 1295, Bin -038 Birmingham, AL 35201-1295

SUBJECT:

EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1, ISSUANCE OF AMENDMENT REGARDING MINIMUM CRITICALPOWER RATIO (CAC NO. MF6681)

Dear Mr. Pierce:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 275 to Renewed Facility Operating License DPR-57 for the Edwin I. Hatch Nuclear Plant, Unit No. 1, in response to the license amendment application dated September 1, 2015. The amendment revises the Technical Specification value of the Safety Limit Minimum Critical Power Ratio to support operation in the next fuel cycle. A copy of the related Safety Evaluation is also enclosed.

A Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Docket No. 50-321

Enclosures:

1 . Amendment No. 275 to DPR-57 2. Safety Evaluation cc w/encls:

Distribution via Listserv Sincerely, J?6 b Martin, Senior Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SOUTHERN NUCLEAR OPERATING COMPANY.

INC. GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 275 . Renewed License No. DPR-57 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit No. 1 (the facility)

Renewed Facility Operating License No. DPR-57 filed by Southern Nuclear Operating

Company, Inc. (the licensee),

acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners),

dated September 1, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 1 O CFR Chapter I; 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.

Enclosure 1 2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this 'lice_nse amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-57 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 275 are hereby incorporated in the license.

Southern Nuclear shall operate the facility in accorda.nce with the Technical Specifications and the Environmental Protection Plan. 3. This license amendment is effective as of its date of issuance and shall be implemented prior to reactor startup following the Unit 1 spring 2016 refueling outage.

Attachment:

Changes to Renewed Facility Operating License No. DPR-57 and the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Qperating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance:

January 29, 201 6 ATTACHMENT TO LICENSE AMENDMENT NO. 275 RENEWED FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 Replace the following pages of the License and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Remove Pages License DPR-57, Page 4 TSs 2.0-1 \_ Insert Pages License DPR-57, Page 4 TSs 2.0-1 for sample analysis or instrumentation calibration, or associated with radioactive apparatus or components; (6) Southern

Nuclear, pursuant to the Act and 1 O 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.

(C) This renewed 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, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional conditions specified or incorporated below: (1) Maximum Power Level Southern Nuclear is authorized to operate the facility at steady sate reactor core power levels not in excess of 2804 megawatts thermal.

(2) Technical Specifications The Technical Specifications (Appendix A) and the Environmental Plan (Appendix B), as revised through Amendment No. 275 are hereby incorporated in the renewed license.

Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. The Surveillance Requirement (SR) contained in the Technical Specifications and listed below, is not required to be performed immediately upon implementation of Amendment No. 195. The SR listed below shall be successfully demonstrated before the time and condition specified:

SR 3.8.1.18 shall be successfully demonstrated at its next regularly scheduled performance.

(3) Fire Protection Southern Nuclear shall implement and maintain in effect all provisions of the fire protection

program, which is referenced in the Updated Final Safety Analysis Report for the facility, as contained in the updated Fire Hazards Analysis and Fire Protection Program for the Edwin I. Hatch Nuclear Plant, Units 1 and 2, which was originally submitted by letter dated July 22, 1986. Southern Nuclear may make changes to the fire protection program without prior Commission approval only if the changes Renewed License No. DPR-57 Amendment No. 275

. 2.0 SAFETY (Sls) 2.1 Sls

• 2.1.1 Reactor Core Sls 2.1.1.1 With the reactor steam dome pressure

<:: 685 psig or core flow < 10% rated core flow: THERMAL POWER shall be :s; 24% ATP. 2.1.1.2 With the reactor steam dome pressure 685 psig and core flow 10% rated core flow: Sls 2.0 MCPR shall 1.09 for two recirculation loop operation 1.12 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel. 2.1.2

• Reactor Coolant System (RCS) Pressure SL Reactor steam dome pressure shall be :s; 1325 psig. 2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s: 2.2.1 ** Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods. HATCH UNIT 1 2.0-1 Amendment No. 275 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 275 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-57 SOUTHERN NUCLEAR OPERATING

COMPANY, INC. EDWIN I. HATCH NUCLEAR PLANT. UNIT NO. 1 DOCKET NO. 50-321 1.0. INTRODUCTION By license amendment request (LAR) dated September 1, 2015 (Agencywide Documents Access and Management System (ADAMS),

Accession No. ML 15252A 185), Southern Nuclear Operating

Company, Inc. (SNC, the licensee),

requested an amendment to the technical specifications

{TS) for the Edwin I. Hatch Nuclear Plant, Unit No. 1 (HNP). The proposed amendment revises TS 2.0, "Safety Limits (SLs)," by changing the safety limit minimum critical power ratio (SLMCPR) for both single and dual recirculation loop operation.

The U.S. Nuclear Regulatory Commission (NRC or Commission) staff's evaluation of the licensee's proposed changes is provided below. 2.0 REGULATORY EVALUATION Title 10 of the Code of Federal Regulations, Part 50 (1 O CFR 50), Appendix A, General Design Criterion (GDC) 10 states, in part, that the reactor core and associated

coolant, control, and protection systems shall be designed to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

The licensee's application includes the following statement regarding the GDC. The Hatch 1 construction permit was received under the 70 general design criteria discussed in "General Design Criteria for Nuclear Power Plant Construction,"

issued for comment in July 1967 and was not, therefore, developed in consideration of the 64 new general design criteria discussed in the "General Design Criteria for Nuclear Power Plants,"

effective May 21, 1971, and subsequently amended July 7, 1971. However, Criterion 6 of the design criteria on which the Hatch 1 construction permit was based is analogous to the current GDC 10. Fuel design limits can be exceeded if the fuel produces heat equal to or greater than critical power. In a boiling water reactor (BWR), heat produced by the fuel causes the water to partially vaporize in a stable process called nucleate boiling.

As the amount of heat produced by-the fuel increases, more of the water vaporizes and the vapor production changes the way the water boils. At a certain point, the efficiency of heat removal is \_ impeded by vapor production and the temperature of the fuel cladding rises disproportionately to the heat generated.

Critical power is a term used for the power at which th.e fuel departs from nucleate boiling and enters a transition to film boiling.

For BWRs, the critical power may be predicted using a correlation known as the GE (General Electric) critical quality boiling length correlation, or better known as the GEXL correlation.

Due to core wide and operational variations, the margin to boiling transition is most easily described in terms of a critical power ratio (CPR), which is defined as the rod critical power as calculated by GEXL divided by the actual rod power. The more a CPR value exceeds 1.0, the greater the margin to boiling transition is. The SLMCPR is calculated using a statistical process that takes into account operating parameters and uncertainties.

The operating limit MCPR (OLMCPR) is equal to the SLMCPR plus a CPR margin for transients.

At the OLMCPR, at least 99.9 percent of the rods avoid boiling transition during steady state operation and transients (Section 4.4, "Thermal and Hydraulic Design,"

of NUREG-0800, Revision 3, dated June 1996) caused by a single operator error or equipment malfunction.

Safety Limits are required to be included in the TS by 1 O CFR 50.36. The SLMCPR is verified on a cycle-specific basis because it is necessary to account for the core configuration-specific neutronic and thermal-hydraulic response.

3.0 TECHNICAL EVALUATION 3.1 HNP Cycle 28 Core HNP is a General Electric BWR design, or BWR/4. The licensee proposed to change the SLMCPR in TS Section 2.1.1.2 from 1.07 to 1.09 for two recirculation loop operation and from 1.09 to 1.12 for single loop operation.

This amendment supports the HNP, Unit 1, Cycle 28 core design. HNP, Unit 1, Cycle 28 core loading consists of 228 fresh Global Nuclear Fuel (GNF) 2 fuel bundles, 224 once-burnt GE14 fuel bundles, 104 twice-burnt GE14 fuel bundles, and 4 thrice-burnt or more GE14 fuel bundles.

This cycle will be the first time that the GNF2 fuel design type is loaded in HNP1 .. 3.2 Methodology GNF developed the HNP, Unit 1, Cycle 28 SLMCPR values using the following NRG-approved methodologies and uncertainties:

• NEDC-32601 P-A, Revision 0, "Methodology and Uncertainties for Safety Limit MCPR Evaluations",

(Reference

2)

• NEDC-32694P-A, Revision 0, "Power distribution Uncertainties for Safety Limit MCPR . Evaluations";

Power Distribution Methodology and Uncertainty (Reference

9)

• NEDE-24011 P-A, Revision 21, "General Electric Standard Application for Reactor Fuel" (Reference

4)

• NEDC-32505P-A, Revision 1, "A-Factor Calculation Method for GE11, GE12 and GE13 Fuel"; A-Factor Calculation Methodology (Reference

5) Plant-specific use of these methodologies must adhere to certain restrictions.

3.2.1 Methodology Restrictions Based on the review (Reference

3) of Topical Reports NEDC-32601 P-A, NEDC-32694P-A, and Amendment 25 to NEDE-24011 P-A (GESTAR II), the NRC staff identified the following restrictions for the use of these Topical Reports:

1. The TGBLA (lattice physics code) fuel rod power calculational uncertainty should be verified when applied to fuel designs not !ncluded in the benchmark comparisons of Table 3.1 of NEDC-32601 P-A, since changes in fuel design can have a significant effect on calculation accuracy.

2. The effect of the correlation of rod power calculation uncertainties should be reevaluated to insure the accuracy of A-Factor uncertainty when the methodology is applied to a new fuel lattice.

3. In view of the importance of MIP (MCPR Importance Parameter) criterion and its potential sensitivity to changes in fuel bundle designs, core loading and operating strategies, the MIP criterion should be reviewed periodically as part of the procedural review process to insure that the specific value recommended in NEDC-32601 P-A is applicable to future designs and operating strategies.

4. The 30-MONICORE bundle power calculational uncertainty should be verified when applied to fuel and core designs not included in the benchmark comparisons in Tables 3.1 and 3.2 of NEDC-32694-P.

3.2.1.1 Restrictions (1 ), (2), and (4) NEDE-24011-P-A provides a fuel design and core reload process that allows an applicant to modify fuel assembly designs without undergoing a formal NRC submittal and review, as long as they provide written notification to the NRC outlining the new design and acknowledging compliance with the requirements of NEDE-24011-P-A.

On March 14, 2007, GNF sent the NRC the aforementioned notification and generic compliance report for the GNF2 fuel assembly design (Reference 6). As part of a NRC audit related to this report, the analysis and evaluation of the GNF2 fuel design was verified to have been evaluated in accordance with the above restrictions (Reference 7). The NRC subsequently concluded in Reference 8 that upon incorporation of Amendment 33, NEDE-24011-P-A was acceptable for use with the GNF2 fuel design without any restriction.

Based on the above discussion, the NRC staff concludes that restrictions (1 ), .(2), and (4) to the plant-specific application of the NEDE-24011-P-A methodology have been addressed for the GNF2 fuel design. Therefore, use of the NEDE-24011-P-A methodology by the licensee is acceptable. 3.2.1 .2 Restriction (3) When determining SLMCPR values, power peaking and power distributions have a direct impact on which fuel bundles may be limiting with respect to boiling transition.

While the pin power peaking is incorporated by the use of A-factors, the bundle power distributions are affected by the loading pattern and rod patterns used during core operation.

GNF tracks this behavior for specific statepoints through the MIP parameter, which is proportional to the probability of boiling transition for a given rod if all bundles had the same pin power distribution.

The value allows for checking how the SLMCPR power distribution compares to previous evaluations, and how limiting the power distribution is to the nominal power distribution.

Restriction (3) of the MFN-003-99 letter,requires reviewing the MIP criterion for new fuel designs, core loading, and operating strategies (Reference 3). The NRC staff found in Section 3.4.1 of the GNF2 GESTAR II Compliance Audit Report that the GNF2 fuel design was in compliance with restriction (3) (Reference 7). In Section 3.0 of the submittal, GNF states that the SLMCPR is calculated in accordance with NEDE-24011-P-A, which has methodologies for analyzing core loading patterns and making sure there is no change in approved core design. As the energy plan, thermal margins, and reactivity margins drive the core design, the SLMCPR is calculated after the core design process is essentially complete.

With similar batch sizes (228 new GNF2 bundles in Cycle 28 and 224 new GE14 bundles in Cycle 27) and similar enrichments (4.06 weight percent in Cycle 28 and 4.1 O weight percent in Cycle 27), the NRC staff concludes that there is no significant departure from operating strategies and core loading.

Thus, the rod patterns used produce a limiting MCPR distribution that should reasonably bound the MCPR distributions that would be expected during the operation of the HNP, Unit 1, core throughout Cycle 28. In summary, the NRC staff concludes that the licensee has adequately addressed the restrictions of Topical Reports NEDC-32601 P-A, NEDC-32694P-A, Amendment 25 to NEDE-24011 P-A (GESTAR 11), and NEDC-32505P-A and that the use of these reports to evaluate the HNP, Unit 1, Cycle 28 SLMCPR is acceptable.

3.3 Major Contributors to SLMCPR Change In general, the calculated safety limit is dominated by two key parameters

1) -flatness of the core bundle-by-bundle MCPR distribution, and 2) -flatness of the bundle pin-by-pin power I A-Factor distribution.

Greater flatness in either parameter yields more rods susceptible to boiling transition and thus a higher calculated SLMCPR. With the introduction of GNF2 fuel, the bundle by bundle MCPR distribution is significantly flatter than Cycle 27 and the GEXL correlation uncertainty for the GNF2 fuel design is larger than GE14 fuel design used in the previous cycle. The combina,tion of a combined higher uncertainty and flatter distribution is sufficient to explain the increase in SLMCPR. 3.4 Departures from NRG-Approved Methodology No departures from NRG-approved methodologies were identified in the HNP, Unit 1, Cycle 28 SLMCPR calculations. 3.5 Deviations from the NRG-Approved Calculational Uncertainties 3.5.1 A-Factor The A-factor is an input into the GEXL correlation used to describe the local pin-by-pin power distribution and the fuel assembly and channel geometry on the fuel assembly critical power. The A-factor uncertainty analysis includes an allowance for power peaking modeling uncertainty, manufacturing uncertainty and channel bow uncertainty.

GNF has increased this uncertainty for all SLMCPR calculations to account for the potential impact of control blade shadow corrosion induced bow. GNF has generically increased the GEXL A-Factor uncertainty to account for an increase in channel bow due to the emerging unforeseen phenomenon called control blade shadow corrosion-induced channel bow, which is not accounted for in the channel bow uncertainty component of the approved A-Factor uncertainty (Reference 9). The HNP, Unit 1 Cycle 28 analysis shows that the expected channel bow uncertainty for HNP Unit 1 is bounded by the increase in A-factor uncertainty as technically justified in Reference

6. Thus, the NRC staff concludes that the use of the higher GEXL A-factor uncertainty described in Reference 6 adequately accounts for the expected control blade shadow corrosion induced bow. 3.5.2 Core Flow Rate and Random Effective Traversing In-core Probe (TIP) Reading GNF stated in Reference 1 O that it would expand the state points used in the determination of the SLMCPR. Consistent with that statement, GNF performs analyses at the rated core power and minimum licensed core flow point in addition to analyses at the rated core power and rated core flow point. The NRG-approved SLMCPR methodology is applied at each state p'oint that is analyzed.

For the two recirculation loop option calculations performed at 92.9% core flow, the approved uncertainty values for the core flow rate (2.5%) and the random effective TIP reading (1.2%) are conservatively adjusted by dividing them by 92.9/100.

The treatment of the core flow and random effective TIP reading uncertainties is based on a conservative assumption that the signal to noise ratio deteriorates as core flow is reduced.

The NRC staff concludes that this increase in the uncertainty should bound the original non-flow dependent uncertainties and, therefore, the NRC staff concludes that this is acceptable for HNP, Unit 1, Cycle 28. 3.5.3 Flow Area Uncertainty GNF calculated the flow area uncertainty for GNF2 and GE14 using the process described in Section 2.7 of NEDC-32601 P-A (Reference 2). The flow area uncertainty for GNF2 and GE14 fuel were conservatively bounded by a value above that found in NEDC-32601 P-A. The bounding value was used in the SLMCPR calculations.

The NRC staff concludes that the impact of flow area uncertainty is captured by the use of a bounding uncertainty value. Therefore, the proposed SLMCPR limits adequately address the uncertainties in channel flow areas for the fuel designs used in the HNP, Unit 1, Cycle 28 core. 3.5.4 Fuel Axial Power Shape Penalty A GNF report on the SLMCPR correlations indicates that they have determined that higher uncertainties and non-conservative biases in the GEXL correlations for different types of axial power shapes could potentially exist relative to the NRG-approved methodology values (Reference 11 ). The GE14 and GNF2 fuel designs are potentially affected in this manner only by Double-Hump axial power shapes. The power shape did not occur on any of the limiting bundles.

Therefore, no power shape penalties were applied to the calculated HNP, Unit 1, Cycle 28 SLMCPR values. The NRC staff concludes that the licensee adequately considered the potential for a higher SLMCPR value resulting from non-conservatisms in the GEXL correlation due to certain axial power shapes within limiting bundles.

Therefore, the use of no axial power shape penalties is acceptable.

3.6 Core Monitoring System For HNP, Unit 1, Cycle 28, the GNF 3D MONICORE System (Reference

9) will be used as the core monitoring system. The 3D MONICORE system is in widespread use throughout the GNF fueled fleet of BWRs similar to HNP. Use of a current version of 3D MONICORE provides the plant capability to perform the reactivity anomaly surveillance.

Use of 3D MONICORE has been previously evaluated and accepted by the NRC in a letter dated March 11, 1999 (Reference 3).

• Therefore, the NRC staff concludes that the use of the GNF 3D MONICORE system for HNP, Unit 1, Cycle 28 is acceptable.

3.7 Technical Evaluation Conclusion The NRC staff concludes that the licensee's proposed HNP, Unit 1, Cycle 28 SLMCPR values of 1.09 for two-recirculation-loop operation and 1.12 for single-recirculation-loop operation are acceptable for HNP, Unit 1, Cycle 28 since approved methodologies were used in accordance with staff guidelines:

The NRC staff concludes that the licensee used methods consistent with regulatory requirements and guidance identified in Section 2!.0 above. The NRC staff notes that no plant hardware or operational changes are required with this TS change. Based on the considerations discussed above, the NRC staff concludes that the proposed change is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Georgia State official was notified of the proposed issuance of the amendments.

The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 1 O CFR Part 20. The NRC staff has determined that the amendments involve 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 considerations, and there has been no public comment on the finding (80 FR 67802, November 3, 2015). 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 amendments.

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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1. Southern Nuclear Operating Company Letter to the NRC, "License Amendment Request Concerning Safety Limit Minimum Critical Power Ratio," dated September 1, 2015 (ADAMS Accession No. ML 15252A 185). 2. General Electric Nuclear Energy Licensing Topical Report NED0-32601-A, Revision 0, "Methodology and Uncertainties for Safety Limit MCPR Evaluations,"

August 1999, (ADAMS Accession No. ML 14093A216)

' 3. MFN-003-99, Letter, Frank Akstulewicz (NRC) to Glen A Watford (GE), "Acceptance for Referencing of Licensing Topical Reports NEDC-32601 P, Methodology and Uncertainties for Safety Limit MCPR Evaluations, NEDC-32694P, Power Distribution Uncertainties for Safety Limit MCPR Evaluation; and Amendment 25 to NEDE-24011-P-A on Cycle Specific Safety Limit MCPR," March 11, 1999 (TAC Nos. M97490, M99069, and M9749). 4. Global Nuclear Fuels Licensing Topical Report NEDE-24011-P-A Revision 21 "General Electric Standard Application for Reactor Fuel," May 1, 2015 (ADAMS Accession No. ML 15121A227).

5. General Electric Nuclear Energy Licensing Topical Report NEDC-32505P-A, Revision 1, "R-Factor Calculation Method for GE11, GE12, and GE13 Fuel," July 1999 (ADAMS Accession No. ML060520636).

6. Letter from GNF to NRC, FLN-2007-011, "GNF2 Advantage Generic Compliance with NEDE-24011-P-A (GESTAR II), NEDC-33270P, March 2007, and GEXL 17 Correlation for GNF2 Fuel, NEDC:-33292P, March 2007," dated March 14, 2007. (ADAMS Accession No. ML070780335)

7. Memorandum, Michelle C. Honcharik (NRC) to Stacey L. Rosenberg (NRC), "Audit Report for Global Nuclear Fuels GNF2 Advanced Fuel Assembly Design GESTAR II Compliance Audit," September 25, 2008. (ADAMS Accession No. ML081630579) 8. Final Safety Evaluation for Amendment 33 to Global Nuclear Fuel (GNF) TR NEDE-24011-P, "General Electric Standard Application for Reactor Fuel (GESTAR II)," August 30, 2010. (ADAMS Accession No. ML 102280591)

9. General Electric Nuclear Energy Licensing Topical Report NEDC-32694P-A, "Power Distribution Uncertainties for Safety Limit MCPR Evaluations,"

August 1999. (ADAMS Accession No. ML0037 40166) 10. Letter, Jason S. Post (GENE) to NRC Document control Desk with attention to Chief, Information Management Branch, et al. (NRC), "Part 21 Final Report: Non-Conservative SLMCPR,"

MFN 04-108, September 29, 2004. (ADAMS Accession No. ML042800267)

11. General Electric Nuclear Energy Licensing Topical Report NEDC-32851 P-A, Revision 5, "GEXL Correlation for GE14 Fuel," April 2011. (ADAMS Accession No. ML 111290532)

Principal Contributor:

William MacFee, NRR/DSS/SRXB Date: January 29, 2016 January 29, 2016 Mr. C. R. Pierce Regulatory Affairs Director

. Southern Nuclear Operating

Company, Inc. Post Office Box 1295, Bin -038 Birmingham, AL 35201-1295

SUBJECT:

EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1, ISSUANCE OF AMENDMENT REGARDING MINIMUM CRITICAL POWER RATIO (CAC NO. MF6681)

Dear Mr. Pierce:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 275 to Renewed Facility Operating License DPR-57 for the Edwin I. Hatch Nuclear Plant, Unit No. 1, in response to the license amendment application dated September 1, 2015. The amendment revises the Technical Specification value of the Safety Limit Minimum Critical Power Ratio to support operation in the next fuel cycle. A copy of the related Safety Evaluation is also enclosed.

A Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Docket No. 50-321

Enclosures:

Sincerely,

/RA/ Bob Martin, Senior Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

1. Amendment No. 275 to DPR-57 2. Safety Evaluation

• cc w/encls:

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