ML16249A001

Issuance of Amendments Small Break Loss of Coolant Accident Reanalysis

ML16249A001
Person / Time
Site:	Millstone
Issue date:	09/30/2016
From:	Richard Guzman Plant Licensing Branch 1
To:	Heacock D Dominion Nuclear Connecticut
Guzman R
References
CAC MF6700
Download: ML16249A001 (23)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 30, 2016 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

MILLSTONE POWER STATION, UNIT NO. 2 - ISSUANCE OF AMENDMENT RE: SMALL BREAK LOSS OF COOLANT ACCIDENT REANALYSIS (CAC NO. MF6700)

Dear Mr. Heacock:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 329 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2 (MPS2), in response to your application dated September 1, 2015, as supplemented on March 24, 2016.

The amendment revises the MPS2 Technical Specifications (TSs) to add the evaluation model EMF-2328(P)(A), Supplement 1, "PWR [pressurized water reactor] Small Break LOCA [loss-of coolant accident] Evaluation Model S-RELAP5 Based," and EMF-92-116(P)(A), Supplement 1, "Generic Mechanical Design Criteria for PWR Fuel Designs," to the TS Section 6.9.1.8.b list of analytical methods used to determine core operating limits as a result of reanalyzing the small break loss-of-coolant accident.

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

Sincerely, fJ!na~

Richard V. Guzman, Sr. Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-336

Enclosures:

1. Amendment No. 329 to DPR-65

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

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DOMINION NUCLEAR CONNECTICUT. INC.

DOCKET NO. 50-336 MILLSTONE POWER STATION. UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 329 Renewed License No. DPR-65

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

A. The application for amendment by Dominion Nuclear Connecticut, Inc. (the licensee) dated September 1, 2015, as supplemented by letter dated March 24, 2016, 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.

Enclosure 1

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 Renewed Facility Operating License No. DPR-65 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 329, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of issuance, and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Travis L. Tate, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. DPR-65 Date of Issuance: September 30, 2016

MILLSTONE POWER STATION. UNIT NO. 2 ATTACHMENT TO LICENSE AMENDMENT NO. 329 RENEWED FACILITY OPERATING LICENSE NO. DPR-65 DOCKET NO. 50-336 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert 3 3 Replace the following pages of the Appendix A Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 6-18a 6-18a 6-19 6-19

ADMINISTRATIVE CONTROLS CORE OPER!\TING LIMITS REPORT 6.9.1.8 a. Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle.

3/4.1.1.1 SHUTDOWN MARGIN (SDM) 3/4.1.1.4 Moderator Temperature Coefficient 3/4.1.3.6 Regulating CEA Insertion Limits 3/4.2.1 Linear Heat Rate 3/4.2.3 TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTOR-FTr 3/4.2.6 DNB Margin

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1) EMF-96-029(P)(A) Volumes 1and2, "Reactor Analysis System for PWRs Volume 1 - Methodology Description, Volume 2 -Benchmarking Results,"

Siemens Power Corporation.

2) ANF-84-73 Appendix B (P)(A), "Advanced Nuclear Fuels Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," Advanced Nuclear Fuels.

3) XN-NF-82-2l(P)(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company.

4) XN-75-32(P)(A) Supplements 1 through 4, "Computational Procedure for Evaluating Fuel Rod Bowing," Exxon Nuclear Company.

5) EMF-2328(P)(A) and Supplement 1, "PWR Small Break LOCA Evaluation Model S-RELAP5 Based."

6) EMF-2087(P)(A), "SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCA Applications," Siemens Power Corporation.

7) XN-NF-78-44(NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company.

MILLSTONE - UNIT 2 6-18a Amendment No.~' 93-, -l-G4, +H, +-1-§.,

+l-9, ~.~,-148,--l-6:3,-l-69,~,

~' ~' ~. ~' +/-&+, ~' 3+/--G, 329

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (CONT.)

8) XN-NF-621(P)(A), "Exxon Nuclear DNB Correlation for PWR Fuel Designs," Exxon Nuclear Company.

9) XN-NF-82-06(P)(A), and Supplements 2, 4 and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup," Exxon Nuclear Company.

10) ANF-88-133(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWd/MTU," Advanced Nuclear Fuels Corporation.

11) XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," Exxon Nuclear Company.

12) ANF-89-151(P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis ofNon-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation.

13) EMF-1961(P)(A), "Statistical Setpoint/Transient Methodology for Combustion Engineering Type Reactors," Siemens Power Corporation.

14) EMF-2310(P)(A), "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," Framatome ANP.

15) EMF-92-153(P)(A) and Supplement 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Power Corporation.

16) EMF-92-116(P)(A) and Supplement 1, "Generic Mechanical Design Criteria for PWR Fuel Designs."

17) BAW-10240(P)(A) Revision 0, "Incorporation of MS' Properties in Framatome ANP Approved Methods, May 2004.

c. The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met.

d. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

MILLSTONE - UNIT 2 6-19 Amendment No.--148,-l-63,-+/-+/-8,--+/-§-0

~' +/-8-1-, +/-9-1-, ;t-9.§., ~' ~'

329

Connecticut, in accordance with the procedures and limitations set forth in this renewed operating license; (2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) 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 and equipment calibration or associated with radioactive apparatus or components; (5) 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.

C. This renewed operating 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, Section 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and 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 The licensee is authorized to operate the facility at steady-state reactor core power levels not in excess of 2700 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 329 are hereby incorporated in the renewed license.

The licensee shall operate the facility in accordance with the Technical Specifications.

Renewed License No. DPR-65 Amendment No. 329

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 329 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-65 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION, UNIT NO. 2 DOCKET NO. 50-336

1.0 INTRODUCTION

By application dated September 1, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15253A205 - Reference 1), as supplemented by letter dated March 24, 2016 (ADAMS Accession No. ML16096A388 - Reference 2), Dominion Nuclear Connecticut, Inc. (DNC, the licensee) requested changes to the technical specification (TS) Section 6.9.1.8.b, "Core Operating Limits Report." The proposed change would add the evaluation model from the topical report EMF-2328(P)(A), Supplement 1, "PWR [pressurized water reactor] Small Break LOCA [loss-of coolant accident] Evaluation Model S-RELAP5 Based," (Reference 6) and topical report EMF-92-116(P)(A), Supplement 1, "Generic Mechanical Design Criteria for PWR Fuel Designs," (Reference 9) to the Millstone Power Station, Unit No. 2 (MPS2) TS Section 6.9.1.8.b list of analytical methods to determine core operating limits. The licensee proposed the change as a result of reanalyzing the small break loss-of-coolant accident (SBLOCA).

The supplemental letter dated March 24, 2016, 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 determination as published in the Federal Register (FR) on December 8, 2015 (80 FR 76318).

2.0 REGULATORY EVALUATION

2.1 Background The current SBLOCA analysis of record (AOR) for MPS2 is based on the NRG-approved EMF-2328(P)(A) (Reference 3) evaluation model (EM). Subsequent to the implementation of this EM, a 30-day report for emergency core cooling system (ECCS) model changes pursuant to the requirements of Title 10 to the Code of Federal Regulations (10 CFR) Section 50.46 (Reference 4) was submitted due to errors identified in the EM that had a >50 °F impact on Enclosure 2

peak cladding temperature (PCT). By letter dated November 1, 2012 (Reference 5), the licensee stated its intent to submit an SBLOCA reanalysis using EMF-2328(P)(A) Revision 0, Supplement 1 (Reference 6) within a year of the NRC staff's approval of EMF-2328 (P)(A)

Supplement 1. The changes to the SBLOCA EM found in Reference 6 were intended to improve the rigor and completeness of the original methodology, while also addressing several staff issues raised regarding the AREVA SBLOCA methodology over the last several years.

EMF-2328 (P)(A), Supplement 1 was generically approved for use in September 2015 (Reference 7). This LAR is applicable to the MPS2 SBLOCA EM reanalysis using Supplement 1.

The NRC staff notes that subsequent to the NRC approval of EMF-2328(P)(A) Supplement 1, the final proprietary and non-proprietary versions of EMF-2328 (P)(A), Supplement 1 (P)(A) were published by AREVA letter dated December 15, 2015 (Reference 8). There were no significant changes between EMF-2328(P)(A) Supplement 1 and EMF-2328(P)(A), Supplement 1 (P)(A) that impacted the staff's review.

Additionally, the SBLOCA EM (Reference 3 and Reference 6), makes use of fuel conditions from RODEX2 code. RODEX2 was updated to include correction factors to account for fuel pellet thermal conductivity degradation (TCD) with increased burnup. These correction factors are documented in Supplement 1 to EMF-92-116(P)(A) (Reference 9), which was generically approved for use in February of 2015. The RODEX2 code is also used in the evaluation of fuel mechanical design criteria. Supplement 1 to EMF-92-116(P)(A) provides the applicable correction factors for these criteria.

The SBLOCA reanalysis, was completed using the AREVA Standard CE14 HTP fuel assembly which includes M5 fuel rod cladding. An exemption from 10 CFR 50.46 and Appendix K to allow the use of the M5 alloy for fuel rod cladding was approved in letter dated May 12, 2015 (Reference 10).

In addition to the analysis completed with the M5 fuel rod cladding, an SBLOCA evaluation of the AREVA fuel product with Zirc-4 cladding, which is the current fuel product used at MPS2, is provided in the LAR. This SBLOCA reanalysis is expected to be implemented prior to the implementation of the AREVA M5 fuel product.

2.2 Proposed Changes The licensee proposed to revise TS 6.9.1.8.b, "Core Operating Limits Report, to add updated AREVA SBLOCA methodology (Reference 6), and the updated AREVA topical report on Generic Mechanical Design Criteria of PWR Fuel Designs (Reference 9) to the list of analytical methods used to determine the core operating limits for MPS2. Additionally, the licensee proposed to delete the company names that are currently found in the documents for clarity since the company name changed between the issuance of the original documents and Supplement 1 for both documents.

Specifically, the licensee has proposed to make the following changes to TS 6.9.1.8.b:

Change from:

5) EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model S-RELAP5 Based,"

Framatome ANP.

16) EMF-92-116(P)(A) Revision 0, "Generic Mechanical Design Criteria for PWR Fuel Designs," Siemens Power Corporation.

Change to:

5) EMF-2328(P)(A) and Supplement 1, "PWR Small Break LOCA Evaluation Model S-RELAP5 Based."

16) EMF-92-116(P)(A) and Supplement 1, "Generic Mechanical Design Criteria for PWR Fuel Design Criteria for PWR Fuel Designs."

2.3 Applicable Regulatory Requirements The regulatory requirements that the staff considered in its review of the proposed license amendment include the following:

• 10 CFR 50.36, "Technical Specifications," insofar as it requires TSs that include limiting conditions of operations, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility, which are, in part, established for a process variable, design feature, or operating restriction that is an initial condition of a design basis accident analysis that either assumes failure of or presents a challenge to the integrity of a fission product barrier.

• Generic Letter (GL) 88-16 (Reference 16), "Guidance for Technical Specifications Changes for Cycle-Specific Parameter Limits," insofar as it provides guidance for modifying TSs to remove cycle-specific parameter limits from the TSs to a licensee controlled COLA.

• 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," insofar as it requires that each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding must be provided with an ECCS that must be designed so that its calculated cooling performance following a postulated LOCA conforms to the criteria set forth in 10 CFR 50.46(b) including PCT, maximum cladding oxidation, maximum hydrogen generation, and coolable geometry. 1 1

Note that the 10 CFR 50.46(b)(5) criterion, long term core cooling, is not addressed in this application. An exemption from 10 CFR 50.46 and 10 CFR Appendix K was granted for MPS2 on May 12, 2015, to allow the use of the M5 alloy for fuel cladding (Reference 10).

• 10 CFR 50, Appendix A, GDC-35, Emergency core cooling, insofar as it requires that a system to provide abundant emergency core cooling be provided to transfer heat from the reactor core following any LOCA at a rate so that fuel clad damage that could interfere with continued effective core cooling will be prevented.

• 10 CFR 50, Appendix K, insofar as it established required and acceptable features of evaluation models for heat removal by the ECCS after the blowdown phase of a LOCA.

• 10 CFR 50, Appendix A, GDC-10, Reactor design, insofar as it requires that the reactor core and associated coolant, control, and protection systems be designed with appropriate margin to ensure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

Additionally, the staff used Standard Review Plan (SRP) 15.6.5, "Loss-of-Coolant Resulting from Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary" as guidance for reviewing the application.

3.0 TECHNICAL EVALUATION

The NRC staff used the applicable sections of SRP 15.6.5 to complete the review. The staff's evaluation involved the:

• Review of the implementation of the updated AREVA SBLOCA EM (Reference 3 and Reference 6).

• Review of the implementation of the updated AREVA topical report on Generic Mechanical Design Criteria for PWR Fuel Designs (Reference 9).

• Review of the SBLOCA evaluation of Zirc-4 Fuel.

3.1 Implementation of Supplement 1 to EMF-2328 Supplement 1 to EMF-2328 (Reference 6) is used in combination with EMF-2328 (Reference 3) for this MPS2 SBLOCA analysis. This is the first application of Supplement 1 to EMF-2328 which provides additional modeling information regarding the following eight areas:

1. Spectrum of break sizes

2. Core bypass flow paths in the reactor vessel model

3. Reactivity feedback model

4. Delayed reactor coolant pump (RCP) trip sensitivity study

5. Maximum accumulator/safety injection tank (SIT) temperature sensitivity study

6. Loop seal clearing (crossover leg modeling)

7. Break in attached piping sensitivity study

8. Core nodalization model

The application was reviewed to ensure appropriate implementation of the evaluation model with a principal focus on the significant changes found in Supplement 1 to EMF-2328(P)(A) which are discussed in the staff's evaluation below. For the eight modeling areas noted above, the staff reviewed the licensee's application of each for consistency with the approved methodology of Supplement 1 to EMF-2328. The staff's review identified the need for additional information in each of the areas, except core bypass flow paths.

In RAl-1 (Reference 11), the NRC staff requested the licensee identify which COLR parameters are governed by EMF-2328(P)(A) and EMF-2328(P)(A), Supplement 1 methods. The licensee responded by letter dated March 24, 2016 (Reference 2), and stated that the Supplement 1 methods establish TS 3.1.3.6, "Regulating CEA Insertion Limits", TS 3.2.1, "Linear Heat Rate",

and TS 3.2.3, "Total Unrodded Integrated Radial Peaking Factor". The staff reviewed the methodology and the proposed TSs as relocated to the COLR, and concludes that this methodology is appropriate for use to establish the limits for these TSs.

Additionally, in RAl-11 (Reference 11), the staff asked the licensee whether the ECCS modeling errors contained in the most recent MPS2 SBLOCA 50.46 PCT rackup sheet (Reference 13) were accounted for in their analysis. The licensee stated in the March 24, 2016, supplement (Reference 2) that its submittal was completed using the most current version of S-RELAP5, which would take into account all reportable errors. Therefore, the staff concludes that an appropriate version of S-RELAP5 was used in the analysis.

Updated Modeling and Initial Conditions The staff reviewed the application to ensure that the plant model was updated in accordance with Supplement 1 of EMF-2328(P)(A) and the appropriate initial conditions were used. The staff questioned the modeling of the core nodalization (RAl-2 of Reference 11) to ensure that the appropriate number of nodes were modeled, as well as the reactivity feedback modeling (RAl-3 of Reference 11 ), to ensure that the appropriate Moderator Temperature Coefficient (MTC) value was incorporated into the analysis. The licensee stated in its RAI response (Reference 2) that the active core was nodalized following the methodology in Supplement 1 of EMF-2328(P)(A) and that maximum plausible value of the MTC at full power was used as a basis for the moderator reactivity feedback as prescribed by the methodology. Based on its review of these aspects of the model, the staff determined the licensee followed the methodology appropriately. Therefore, the staff concludes that the licensee correctly implemented the updated plant model in accordance with Supplement 1 of EMF-2328(P)(A).

The staff reviewed the initial conditions used in the analysis to ensure that the values were appropriate for the SBLOCA analysis and consistent with the current licensing basis. The current licensed thermal power is 2700 MWt (2754 MWt including 2% uncertainty), a peak linear heat rate (PLHR) of 15.1 kW/ft, and a nominal radial peaking of 1.69 (1.827 including uncertainty). In this application, the thermal power, PLHR, and the nominal radial peaking factor are unchanged.

The radial peaking factor including uncertainties has changed from the value used for the SBLOCA analysis found in the current final safety analysis report (FSAR). The nominal value was unchanged from the FSAR; however, the uncertainty was increased. The staff asked the licensee to provide additional information regarding the change in uncertainty in RAl-4

(Reference 11 ). The licensee stated in its RAI response (Reference 2) that the difference was caused by the rodded augmentation factor which accounts for the effects of rod insertion on the total radial peaking factor. This value was recommended by AREVA to bound future plant reloads. Assuming an increase in the radial peaking factor in the SBLOCA analysis is conservative, the NRC staff concludes that this is an acceptable change.

Additionally, relative to the values used in the current FSAR SBLOCA analysis, the SIT fluid temperature, steam generator (SG) secondary side pressure, the auxiliary feedwater (AFW) temperature, and AFW flow have changed. The staff questioned the differences in these parameters in RAl-4 and RAl-6 (Reference 11) and the licensee provided the following discussion in its RAI response (Reference 2):

• The SIT temperature change is consistent with the new guidelines provided in Section 6 of EMF-2328(P)(A), Supplement 1 (Reference 6). The staff reviewed the guidelines in Reference 6 and the SIT temperatures used in the LAR and determined that the analysis is consistent with the methodology.

• The SG secondary side pressure was different due to the reported values in the FSAR and LAR (FSAR reported the steady-state pressure calculated by S-RELAP5 and the LAR reported the target pressure used in the steady state analysis to benchmark SG plant conditions with no tube plugging). Additionally, the licensee stated that the calculated S-RELAP5 SG secondary side pressure is slightly different between the LAR and FSAR. The staff determined that the licensee's justification is reasonable since small differences between the steady state SG secondary side pressure will have a minimal impact on the analysis results given that after the break is simulated, the turbine will be isolated from the SG and the secondary side pressure will quickly increase to the main steam safety valve setpoint. Therefore, the NRC staff finds this change is acceptable.

• The AFW temperature is different because the FSAR analysis used a maximum value and the LAR used a nominal value. The licensee stated that the AFW is a second order effect and the use of a nominal temperature is suggested in SBLOCA guidelines for Reference 3. The SG secondary side is important during pump coast down and the natural circulation phase of the SBLOCA transient in that it acts as a heat sink to remove core decay heat. Primary to secondary side heat transfer is important for natural circulation to occur. Primary to secondary heat flux is driven by the temperature gradient and the effective heat transfer coefficient from the primary to secondary side. While the temperature gradient could potentially be impacted by the change in AFW temperature, its impact is expected to be small. More importantly, there is enough SG tube coverage (i.e. secondary side water level) such that a breakdown of natural circulation during this phase of the transient would not hinder primary to secondary heat transfer. This breakdown was not observed in the analysis results. Given that the AFW temperature change is expected to have a small impact on primary to secondary side heat transfer, and SG tube water coverage has a larger impact, the NRC finds that modeling the AFW nominal temperature is acceptable.

• The AFW flow modeling appeared to be different between Figure 4-13 in the LAR and Figure 14.6.5.2-13 of the FSAR. The licensee stated that this was due to different

modeling techniques used in the FSAR and LAR analysis. In the FSAR analysis, the licensee used a variable flow rate dependent on pressure. In the LAR, a minimum flow was used. This minimum AFW flow bounds the minimum AFW flow requirements for the plant. For an SBLOCA analysis, primary to secondary side SG heat transfer is important during the pump coast down and natural circulation portion of the transient. As discussed in the previous bullet, if the secondary side SG tubes are significantly uncovered, the effective primary to secondary heat transfer coefficient will be significantly decreased and could hinder heat removal capabilities of the SGs. The NRC staff determined the licensee's assumption of minimum AFW flow is conservative, and therefore, the staff finds the proposed modeling technique used in the LAR is acceptable.

Additionally, in RAl-5 (Reference 11 ), the staff asked for clarification of the legend in Figure 3-4, "Axial Power Distribution Comparison," in the LAR. The licensee stated in its response (Reference 2) that the solid line represented the step axial adjusted to FQ/FflH limit and that the dashed line was the user input step axial. Additionally, the staff questioned in RAl-5 if the axial power distribution used in the FSAR had changed relative to that used in the current licensing basis. The licensee stated that there are slight differences between the two due to differences in fuel cycle designs, but the axial power distributions peak at the same elevation and they were created following the same guidelines to satisfy the TS limit. Given this, the NRC staff concludes that an appropriate axial power distribution was used in the LAR.

Spectrum of Break Sizes Supplement 1 to EMF-2328(P)(A) requires a break spectrum that has a wide enough range to establish a clear trend in the PCT and to identify the limiting break size, from the smallest break that exceeds the capacity of the makeup system up to, and including, 10% of the cold leg area break. Additionally, it requires that a finer break size resolution be evaluated about the limiting break. The analyzed spectrum of breaks considered in this analysis was from a 2.0 inch diameter break, up to and including a 9.49 inch diameter break (equivalent to 10% of the cold leg area). The limiting break size was determined by the licensee to be a 3. 78 inch break. The staff determined that the appropriate refinement was used and the limiting break was identified as the licensee correctly followed the methodology of Supplement 1 to EMF-2328(P)(A).

Therefore, the staff finds that the licensee's proposed break size spectrum is acceptable.

Additionally, EMF-2328(P)(A) Supplement 1 requires that the largest small break that depressurizes to a pressure just above the SIT actuation pressure be included in the break spectrum. The staff determined this break was included in the spectrum of breaks evaluated, using a finer break resolution, which is consistent with the methodology of Supplement 1 to EMF-2328(P)(A).

The staff requested in RAl-7 (Reference 11) the plots for the transition break (3. 79-inch diameter break) case. The staff reviewed the six plots submitted by the licensee and determined that the plant model behaved as expected, consistent with SBLOCA phenomena.

Most notably, from the review of the Loop Seal Void Fraction Plot (Figure 2 in the licensee's RAI response), the staff observed that multiple loop seals clear (e.g., multiple loop seals are voided).

Since multiple loops clear, there is less loop resistance to the break which provides improved venting capability. This results in a higher mixture level (of liquid/vapor) in the core (see Figure

5 in the RAI response and Figure 4-21 in the LAR), which leads to a lower PCT. Based upon these observations, the staff concluded that the system response for this transition break size and model configuration behaves as expected.

Delayed Reactor Coolant Pump Trip The licensee performed evaluations for a spectrum of hot and cold leg breaks with delayed RCP trip time to demonstrate compliance with 10 CFR 50.46. Continued operation of the RCPs following an SBLOCA could reduce the effectiveness of the ECCS for certain break sizes. The licensee's RCP trip study was performed following the guidelines in GL 86-06, "Implementation of TMI [Three Mile Island] Action Item 11.K.3.5, "Automatic Trip of Reactor Coolant Pumps,"

(Reference 17). GL 86-06 provides guidance for demonstrating compliance with the 10 CFR 50.46 criteria and, additionally, outlines delayed RCP trip sensitivity studies to estimate the maximum delay time for operator action to satisfy the 10 CFR 50.46 ECCS criteria. Since GL 86-06 provides a conservative approach for modeling delayed RCP trips, the NRC staff finds the licensee's approach acceptable.

To demonstrate compliance with the 10 CFR 50.46 criteria and to support the RCP trip criteria, cold and hot leg break RCP trip sensitivity studies were completed by the licensee. These sensitivity studies were completed for a spectrum of breaks from 2.0 inches to 9.49 inches in diameter, and included almost the same break spectrum (all except for the 3.9-inch break),

biasing of loops seals and transition break size as in the base break spectrum for both hot and cold leg breaks; however, the RCP trip was delayed until two minutes after loss of subcooling margin in the cold leg pump suction. The licensee provided the PCT results in an RAI response (Reference 2). The NRC staff reviewed these results and determined the hot leg breaks had a stable trend that peaked at the 5-inch break (PCT of 1575°F). For the cold leg results, the PCT peak was the 4.02-inch break (PCT of 1644°F). For break sizes greater that 4.02-inches, the PCT results showed a consistent downward trend.

For break sizes smaller than 4.02-inches, there was a large PCT jump (230°F) between the limiting 4.02-inch break and the next smallest break (3.79-in break). Since the jump was large, the NRC staff could not definitively determine that the limiting PCT break size for this study had been identified. During a clarification call on April 18, 2016, the licensee indicated that they would provide a 3.9-inch break case result to demonstrate that the limiting PCT had been identified. In an e-mail dated April 27, 2016 (Rererence 12), the licensee provided the 3.9-inch case which resulted in a PCT of 1606°F. Since this result is less than the PCT for the limiting break and was a reasonable step in break size from the limiting case, the staff determined that the licensee has appropriately identified the limiting cold leg break for the RCP trip sensitivity studies. Therefore, the staff finds that licensee's RCP trip sensitivity studies are acceptable because there is two minutes to trip the RCPs after losing subcooling margin, and the results are bounded by the base case limiting break spectrum results.

An evaluation was completed by the licensee using the same model as the RCP studies to estimate the maximum delay time available for an RCP trip in order to meet the 10 CFR 50.46 acceptance criteria. This evaluation provided a more realistic response, consistent with the

approach outlined in GL 86-06. 2 For the licensee's study, the 3.0-inch, 4.0-inch, and 5.0-inch break sizes were analyzed. GL 86-06 requires that the limiting break from the cold and hot leg break RCP trip sensitivity studies be used in this evaluation. A 4.0-inch break was used as opposed to the limiting 4.02-inch break case from the RCP delay study. This is a very small change in break size, thus the results would be very similar; thus the NRC staff finds using a 4.0-inch break case for this study acceptable.

The evaluation used best-estimate assumptions: (1) the decay heat multiplier was reduced from 1.2 to 1.0, and (2) the critical break flow model was changed from the Moody model to the homogeneous relaxation model. The NRC staff agrees with these parameters being relaxed in order to estimate the maximum delay time for the operators to trip the RCPs since the relaxation of these parameters would be expected to provide a result that more closely resembles the realistic plant response. The results showed that there was an estimated 15 minutes and 10 minutes, for the limiting cold and hot leg break respectively, after loss of subcooling for the operators to trip the RCPs and to still meet the 10 CFR 50.46 ECCS acceptance criteria.

The NRC questioned in RAl-8 (Reference 11) how this study supports the plant's Emergency Operator Procedures (EOPs). The licensee stated in the RAI response (Reference 2) that the operators are directed to stop all RCPs if the pressurizer pressure drops to less than the minimum RCP net positive suction head (NPSH) limit which is based on a conservative pressurizer pressure as a function of RCS cold leg temperature. In the evaluation, the time for the operators to trip the RCPs begins on loss of RCS cold leg subcooling. Following an SBLOCA, the time to reach cold leg saturation conditions (i.e., when the RCP trip is simulated in this evaluation) is greater than the time to loss of NPSH (i.e., when operators are directed to stop all RCPs). Thus, existing plant procedures require sooner operator action to stop the RCPs compared to what was assumed in the analysis which is conservative. The staff reviewed the RAI response and the LAR, and concludes that the operator action time determined in the RCP trip evaluation is conservative.

Attached Piping Break Sensitivity Study The licensee has analyzed a doubled ended guillotine (DEG) break of a SIT line. The SIT lines deliver safety injection from the associated SIT, low pressure safety injection pumps, and high pressure safety injection pumps. An SIT line with a DEG break provides less resistance to safety injection flow compared to the other intact lines; thus, more flow would be diverted to the broken line compared to the intact lines which could reduce the ECCS flow to the RCS.

Therefore, in order to assure acceptable ECCS performance, the licensee performed a sensitivity study.

The results of the sensitivity study is a calculated PCT of 1239°F which is lower than the break spectrum analysis and meets the 10 CFR 50.46 ECCS criteria.

2 Note that the licensee's study was completed to demonstrate that there was margin for the operators to trip the RCPs assuming a more realistic model is applied. Also note that 10 CFR 50.46 acceptance criteria calculated from this sensitivity study is not used as the licensing basis for MPS2.

The NRC staff reviewed the applicable Figures in Chapter 6 of the FSAR, and concludes that a DEG SIT line break is appropriate for the attached piping break sensitivity study since modeling the DEG break in this line is conservative.

Safety Injection Temperature Sensitivity Study The licensee completed a safety injection temperature sensitivity study following the direction provided in EMF-2328(P)(A) Revision 0, Supplement 1 (Reference 6). The NRC reviewed the applicable sections of the FSAR to confirm safety injection fluid temperatures used in the base break spectrum analysis and the sensitivity study were consistent with plant operation and concludes that the appropriate temperatures were used. The NRC staff reviewed the analyses completed in the sensitivity study and concludes that the approach taken by the licensee for its sensitivity study was adequate and consistent with EMF-2328(P)(A) Revision 0, Supplement 1.

The NRC staff reviewed the results from this safety injection temperature sensitivity study and the base break spectrum analysis and concludes that the results support the conclusion that the analysis demonstrates 10 CFR 50.46 acceptance criteria compliance.

Results In Attachment 3 of the LAR, the licensee provided results for the MPS2 SBLOCA analysis. The calculated PCT, the maximum local oxidation, and the maximum core wide oxidation were from the 3. 78-inch cold leg break from the cold leg break spectrum study and are summarized in the table below.

Parameter SBLOCA Analysis Results 10 CFR 50.46 Limit PCT 1707°F 2200°F (10 CFR 50.46(b)(1))

Maximum Local Oxidation 5.8% (includes pre-transient 17.0% (10 CFR 50.46(b)(2))

oxidation of 2.3%)

Maximum Total Core-Wide 0.04% 1.0% (10 CFR 50.46(b)(3))

Oxidation The NRC staff finds these results are within the limits of 10 CFR 50.46, and are therefore, acceptable.

3.2 Zirc-4 Fuel Evaluation The licensee plans to implement the updated SBLOCA EM (Reference 3 and Reference 6) prior to loading of the M5 fuel product and, as such, an evaluation of the current Zirc-4 fuel product was completed. The licensee completed an evaluation of several break sizes surrounding the limiting break using the same model as the M5 model with only a change to the fuel. The results of the evaluation showed that the same break size as the M5fuel analysis was limiting with a +4°F temperature difference for Zirc-4. Additionally, the staff requested in RAl-9 (Reference 11) that the break sizes evaluated and their associated PCTs be provided.

The licensee in its March 24, 2016, response (Reference 2) provided information which showed the same general trend of PCT vs. break size for both fuel types. The licensee proposed the

+4°F temperature difference will be applied as a PCT penalty, which will be implemented until the Zirc-4 fuel is no longer limiting. The licensee will evaluate the need for the Zirc-4 penalty for each reload core; and once the Zirc-4 is determined to be no longer limiting, it will be removed

and evaluated in accordance with the requirements of 10 CFR 50.46. Therefore, the NRC staff finds that this evaluation is adequate for Zirc-4 fuel and that the licensee's approach to manage the PCT penalty is acceptable.

Additionally, the NRC staff evaluated the maximum local oxidation for Zirc-4 fuel. The staff used the Baker-Just correlation in a conservative hand confirmation calculation to evaluate the maximum local oxidation for Zirc-4 fuel. The staff found the conservative hand calculated results indicated the maximum local oxidation met the 10 CFR 50.46 acceptance criteria. Given that the Zirc-4 fuel is the outgoing fuel, its oxidation will become less significant once M5 is introduced. The staff also determined that the licensee's current AOR in the FSAR was completed with Zirc-4 fuel and demonstrates compliance with 10 CFR 50.46. Therefore, based on the above findings, the staff concludes that the Zirc-4 fuel in this analysis meets the 10 CFR 50.46 acceptance criteria.

3.3 Implementation of Supplement 1 to EMF-92-116 Supplement 1 to EMF-92-116 (Reference 9) was generically approved for use in performing fuel thermo-mechanical calculations for PWRs using U02 and urania-gadolinia fuel rods with Zirc-4 or M5 cladding materials in February 2015. Supplement 1 to EMF-92-116 (Reference 9) provides correction factors to be applied to the RODEX2 predictions to account for TCD. NRC Information Notice 2009-23, "Nuclear Fuel Thermal Conductivity Degradation," (Reference 14) describes an issue concerning the ability of legacy thermal-mechanical fuel modeling codes to predict the exposure-dependent degradation of fuel thermal conductivity accurately. A safety concern with TCD in a LOCA would be that fuel temperatures modeled incorrectly would affect the initial stored energy, causing the LOCA evaluation model to predict erroneous PCTs. The staff questioned in RAl-10 (Reference 11) if the licensee included the TCD correction factors in the SBLOCA analysis. The licensee stated in its RAI response dated March 24, 2016 (Reference 2) that, for SBLOCA, initial fuel rod temperatures have no impact on the analysis results. The licensee stated that initial stored energy during an SBLOCA is removed via the SGs and containment during the pump coastdown, and no cladding temperature excursion would be expected during this phase of the LOCA. The SBLOCA transient is dependent on decay heat, and has little, if any, relationship to the initial stored energy. The licensee also stated that TCD can impact the prediction of rod internal pressure which can impact the limiting fuel power histories selected for the SBLOCA analysis. The staff determined that the licensee's approach is consistent with past precedent (Reference 15), and the licensee has appropriately accounted for TCD correction factor applicability in the SBLOCA analysis. Therefore, the staff finds that the licensee's technical justification for addressing TCD for the SBLOCA analysis is acceptable.

Additionally, as described in EMF-92-116(P)(A) Supplement 1 (P)(A) (Reference 9), RODEX2 is used to evaluate various mechanical fuel design criteria. EMF-92-116(P)(A) Supplement 1 (P)(A) (Reference 9) is appropriate for Zirc-4 or M5 cladding as it assesses the impact that TCD has on the fuel design criteria and provides the applicable correction factors to the RODEX2 results. Since this application is limited to Zirc-4 and M5 cladding, the NRC staff finds the implementation of EMF-92-116(P)(A) Supplement 1 (P)(A) (Reference 9) to evaluate the mechanical fuel design criteria outlined in the LAR acceptable.

The staff asked the licensee in RAl-1 (Reference 11) to identify which COLR parameters are governed by this methodology. The licensee responded in the supplement dated March 24, 2016 (Reference 2), stating that the methodology of EMF-2328(P)(A) including Supplement 1 to EMF-2328(P)(A) is used to establish limits for the COLR parameters in TS 3.1.3.6, "Regulating CEA Insertion Limits," TS 3.2.1, "Linear Heat Rate," and TS 3.2.3, "Total Unrodded Integrated Radial Peaking Factor to demonstrate acceptable SBLOCA results. The staff reviewed the methodology and concludes it is appropriate to establish the COLR parameters (in TS 3.1.3.6, TS 3.2.1, and TS 3.2.3) since the linear heat rate and radial peaking factor are used in the analysis.

3.4 Applicable Regulatory Requirement Compliance The licensee proposed to revise TS 6.9.1.8.b, "Core Operating Limits Report" to add updated SBLOCA methodology in EMF-2328(P)(A) (Reference 6) and an updated AREVA topical report on Generic Mechanical Design Criteria for PWR Fuel Designs, EMF-92-116(P)(A), Supplement 1 (Reference 9). TS 6.9.1.8.b lists the approved methodologies used to establish the parameters found in the COLR. The COLR is a licensee controlled document which contains various cycle-specific parameters that are relocated from the limiting conditions of operation (LCO) in the TS and are established using NRG-approved methodologies. The licensee has used GL 88-16 guidance (Reference 16) by including these methodologies in TS 6.9.1.8.b. In accordance with10 CFR 50.36, the licensee is required to include LCOs in the TSs to establish limiting conditions for safe operation. The licensee proposed to use EMF-2328(P)(A),

Supplement 1 and EMF-92-116(P)(A), Supplement 1 (P)(A) to determine TSs 3.1.3.6, 3.2.1, and 3.2.3 limits in the COLR, and the staff has determined that these methodologies are acceptable. Therefore, the LCO requirement in 10 CFR 50.36 is met.

The MPS2 SBLOCA calculated results demonstrate that the ECCS is designed to meet the 10 CFR 50.46 acceptance criteria, and additionally, demonstrate that the ECCS is designed to meet the 10 CFR 50, Appendix A, GDC-35 design criteria. The staff determined that the analysis was correctly implemented using an NRG-approved SBLOCA evaluation model which is based on the requirements in 10 CFR 50, Appendix K.

EMF-92-116(P)(A), Supplement 1 (P)(A) is implemented to support the evaluation of fuel mechanical design criteria by accounting for the impacts of TCD. EMF-92-116(P)(A),

Supplement 1 (P)(A) has been approved by the staff for use of the Zirc-4 and M5 fuel products.

The evaluation of fuel mechanical design criteria ensures that fuel design limits are not exceeded during any condition of normal operation including the effects of anticipated operation occurrences. Therefore, the staff concludes that the implementation of EMF-92-116(P)(A),

Supplement 1 (P)(A) supports compliance with the 10 CFR 50, Appendix A, GDC-10 criteria.

3.5 NRC Staff Conclusion

The NRC staff reviewed the licensee's application to ensure that: (1) there is a reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) activities proposed 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 the health and safety of the public. The NRC staff evaluated the licensee's proposed changes to TS 6.9.1.8.b, "Core Operating Limits Report" to support

reanalyzing the SBLOCA. Based on the considerations discussed above, the NRC staff concludes that the proposed changes to TS 6.9.1.8.b are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Connecticut State official was notified on August 30, 2016, 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 no significant change in the types or significant increase in the amounts 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 (80 FR 76318). 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.

7.0 REFERENCES

1. Mark D. Sartain, letter to U.S. Nuclear Regulatory Commission, "Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 2, Proposed License Amendment Request - Small Break Loss of Coolant Accident Reanalysis," September 1, 2015 (ADAMS Accession No. ML15253A205).

2. Mark D. Sartain, Supplemental letter to U.S. Nuclear Regulatory Commission, "Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 2, Response to Request for Additional Information for Proposed License Amendment Request Regarding Small Break Loss of Coolant Accident Reanalysis (CAC NO. MF6700),"

March 24, 2016 (ADAMS Accession No. ML16096A388).

3. EMF-2328(P)(A) Revision 0, "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," March 2001 (ADAMS Package Accession No. ML011410426).

4. J. Alan Price, letter to U.S. Nuclear Regulatory Commission, "Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 2, 30-Day Report for Emergency Core Cooling System Model Changes Pursuant to the Requirements of 10 CFR 50.46,"

January 25, 2012 (ADAMS Accession No. ML12031A147).

5. J. Alan Price, letter to U.S. Nuclear Regulatory Commission, "Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 2, Response to Request for Additional Information Regarding the 30-Day Report for Emergency Core Cooling System Model Changes (TAC No. ME7881)," November 1, 2012 (ADAMS Accession No. ML12311A029)

6. EMF-2328(P)(A) Revision 0, Supplement 1, "PWR Small Break LOCA Evaluation Model S-RELAP5 Based, March 2012. (ADAMS Package Accession No. ML120650754).

7. "Final Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report EMF-2328(P)(A), Revision 0, Supplement 1, Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," September 1, 2015 (ADAMS Accession No. ML 1521 OA257).

8. EMF-2328 (P)(A), Revision 0, Supplement 1 (P)(A), "PWR Small Break LOCA Evaluation Model S-RELAP5 Based, December 2015 (ADAMS Package Accession No. ML15352A300).

9. EMF-92-116 (P)(A), Revision 0, Supplement 1 (P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs," February 2015 (ADAMS Package Accession No. ML15148A156).

10. Richard Guzman letter to Dominion Nuclear Connecticut, In. "Millstone Power Station, Unit No.2 - Exemption from 10 CFR 50.46 and Appendix K to Allow Use of M5' Allow for Fuel cladding (TAC No. MF3917)," May 12, 2015. (ADAMS Accession No. ML15093A497).

11. R. Guzman, NRC Email to M. Whitlock, Dominion Nuclear Connecticut, Inc., Millstone Power Station, Units 2, "MPS2 LAR Small Break Loss of Coolant Accident Re-analysis - Request for Additional Information (MF6700)," February 16, 2016 (ADAMS Accession No. ML16047A171).

12. Wanda D. Craft (Dominion Nuclear Connecticut, Inc.) e-mail to Richard Guzman (NRC), "Millstone Power Station, Unit 2, Re: Supplemental RAI Clarification Information - SBLOCA LAR (CAC No. MF6700)," April 27, 2016. (ADAMS Accession No. ML16122A001 ).

13. Mark D. Sartain, letter to U.S. Nuclear Regulatory Commission, "Dominion Nuclear Connecticut, Inc. Virginia Electric and Power Company Millstone Power Station Units 2 and 3, North Anna Power Station Units 1 and 2 Surry Power Station Units 1 and 2, 2014 Annual Report of Emergency Core Cooling System (ECCS) Model Changes

Pursuant to the Requirements of 10 CFR 50.46," June 30, 2015 (ADAMS Accession No. ML15188A346).

14. NRC Information Notice 2009-23, "Nuclear Fuel Thermal Conductivity Degradation,"

October 8, 2009 (ADAMS Accession No. ML091550527).

15. Christopher R. Costanzo, letter to U.S. Nuclear Regulatory Commission, "(RAI Response to SRXB-RAl-1 and SNPB RAl-2 thru RAl-20 for the Technical Specification LAR and Exemption Request Regarding the Transitioning to AREVA Fuel, (St Lucie Unit 2)," October 2, 2015 (ADAMS Accession No. ML15279A222).

16. NRC Generic Letter (GL) 88-16, "Guidance for Technical Specifications Changes for Cycle-Specific Parameter Limits," October 04, 1988 (ADAMS Accession No. ML031130447).

17. NRC Generic Letter (GL) 86-06, "Implementation of TMI Action Item 11.K.3.5, "Automatic Trip of Reactor Coolant Pumps," May 29, 1986 (ADAMS Accession No. ML031150282).

Principal Contributor: Joshua M. Borromeo Date: September 30, 2016

September 30, 2016 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

MILLSTONE POWER STATION, UNIT NO. 2 - ISSUANCE OF AMENDMENT RE: SMALL BREAK LOSS OF COOLANT ACCIDENT REANALYSIS (CAC NO. MF6700)

Dear Mr. Heacock:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 329 to Renewed Facility Operating License No. DPR-65 for the Millstone Power Station, Unit No. 2 (MPS2), in response to your application dated September 1, 2015, as supplemented on March 24, 2016.

The amendment revises the MPS2 Technical Specifications (TSs) to add the evaluation model EMF-2328(P)(A), Supplement 1, "PWR [pressurized water reactor] Small Break LOCA [loss-of coolant accident] Evaluation Model S-RELAP5 Based," and EMF-92-116(P)(A), Supplement 1, "Generic Mechanical Design Criteria for PWR Fuel Designs," to the TS Section 6.9.1.8.b list of analytical methods used to determine core operating limits as a result of reanalyzing the small break loss-of-coolant accident.

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

Sincerely,

/RA/

Richard V. Guzman, Sr. Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-336

Enclosures:

1. Amendment No. 329 to DPR-65

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

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