ML16133A438

Issuance of Amendments Ultimate Heat Sink Temperature Increase

ML16133A438
Person / Time
Site:	Braidwood
Issue date:	07/26/2016
From:	Joel Wiebe Plant Licensing Branch III
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
Wiebe J
References
CAC MF4671, CAC MF4672
Download: ML16133A438 (38)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 26, 2016 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO)

Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

BRAIDWOOD STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE:

ULTIMATE HEAT SINK TEMPERATURE INCREASE (CAC NOS. MF4671 AND MF4672)

Dear Mr. Hanson:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 189 to Renewed Facility Operating License No. NPF-72 and Amendment No. 189 to Renewed Facility Operating License No. NPF-77 for the Braidwood Station, Unit Nos. 1 and 2, respectively. The amendments are in response to your application dated August 19, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14231A902), as supplemented by letters dated January 20, 2015 (ADAMS Accession No. ML15020A246), March 31, 2015 (ADAMS Accession No. ML15090A604), April 30, 2015 (ADAMS Accession No. ML15120A396), August 24, 2015 (ADAMS Accession No. ML15236A144), October 9, 2015 (ADAMS Accession No. ML15282A345), October 30, 2015 (ADAMS Accession No. ML15303A326), November 9, 2015 (ADAMS Accession No. ML15313A254), December 16, 2015 (ADAMS Accession No. ML15364A369), February 12, 2016 (ADAMS Accession No. ML16043A496), April 29, 2016 (ADAMS Accession No. ML16123A014), and June 16, 2016 (ADAMS Accession No. ML16169A139).

The amendments increase the ultimate heat sink temperature allowable limit in technical specification surveillance requirement 3.7.9.2 to::; 102° F.

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

Sincerely, Joel S. Wiebe, Senior Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-456 and STN 50-457

Enclosures:

1. Amendment No. 189 to NPF-72

2. Amendment No. 189 to NPF-77

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

UNITED STATES NUCLEAR REGULATORY COMMISSION

• WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-456 BRAIDWOOD STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 189 Renewed License No. NPF-72

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

A The application for amendment by Exelon Generation Company, LLC (the licensee) dated August 19, 2014, as supplemented by letters dated January 20, 2015, March 31, 2015, April 30, 2015, August 24, 2015, October 9, 2015, October 30, 2015, November 9, 2015, December 16, 2015, February 12, 2016, April 29, 2016, and June 16, 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.

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. NPF-72 is hereby amended to read as follows:

Enclosure 1

B. Hanson (2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 189 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

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

F&~? REGULATORY COMMISSION G. Edward Miller, Chilcting)

Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: July 26, 2016

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-457 BRAIDWOOD STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 189 Renewed License No. NPF-77

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

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated August 19, 2014, as supplemented by letters dated January 20, 2015, March 31, 2015, April 30, 2015, August 24, 2015, October 9, 2015, October 30, 2015, November 9, 2015, December 16, 2015, February 12, 2016, April 29, 2016, and June 16, 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.

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. NPF-77 is hereby amended to read as follows:

Enclosure 2

B. Hanson (2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 189 and the Environmental Protection Plan contained in Appendix B, both of which are attached to Renewed License No. NPF-72, dated January 27, 2016, are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

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

FOR THE 't:f.::,~;71LATORY COMMISSION G. Edward Miller, Chief (Acting)

Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: July 2 6, 2O1 6

ATTACHMENT TO LICENSE AMENDMENT NOS. 189 AND 189 RENEWED FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 DOCKET NOS. STN 50-456 AND STN 50-457 Replace the following pages of the Renewed Facility Operating Licenses 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 License NPF- 72 License NPF-77 Page 3 Page 3 License NPF- 72 License NPF-77 Page 3 Page 3 TSs TSs 3.7.9 - 1 3.7.9- 1

(2) Exelon Generation Company, 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) Exelon Generation Company, 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) Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40 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 Commission's regulations set forth in 10 CFR Chapter I and 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 The licensee is authorized to operate the facility at reactor core power levels not in excess of 3645 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 189 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No. 189 Renewed License No. NPF-72

(2) Exelon Generation Company, LLC, 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) Exelon Generation Company, LLC, 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) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. The renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and 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 The licensee is authorized to operate the facility at reactor core power levels not in excess of 3645 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 189 and the Environmental Protection Plan contained in Appendix B, both of which are attached to Renewed License No. NPF-72, dated January 27, 2016, are hereby incorporated into the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No. 189 Renewed License No. NPF-77

UHS 3.7.9 3.7 PLANT SYSTEMS 3.7.9 Ultimate Heat Sink (UHS)

LCD 3.7.9 The UHS shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. UHS inoperable. A.l Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND A.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SU RV EI LLANCE FREQUENCY SR 3.7.9.l Verify water level of UHS is ~ 590 ft Mean In accordance Sea Level (MSL). with the Surveillance Frequency Control Program SR 3.7.9.2 Verify average water temperature of UHS is In accordance S 102°F. with the Surveillance Frequency Control Program SR 3.7.9.3 Verify bottom level of UHS is s 584 ft MSL. In accordance with the Surveillance Frequency Control Program BRAIDWOOD - UNITS 1 &2 3.7.9 - 1 Amendment No. 189

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 189 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-72 AND AMENDMENT NO. 189 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-77 EXELON GENERATION COMPANY. LLC BRAIDWOOD STATION. UNITS 1 AND 2 DOCKET NOS. STN 50-456 AND STN 50-457

1.0 INTRODUCTION

By letter to the U.S. Nuclear Regulatory Commission (NRC, the Commission) dated August 19, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14231A902), as supplemented by letters dated January 20, 2015 (ADAMS Accession No. ML15020A246}, March 31, 2015 (ADAMS Accession No. ML15090A604), April 30, 2015 (ADAMS Accession No. ML15120A396}, August 24, 2015 (ADAMS Accession No. ML15236A144}, October 9, 2015 (ADAMS Accession No. ML15282A345}, October 30, 2015 (ADAMS Accession No. ML15303A326}, November 9, 2015 (ADAMS Accession No. ML15313A254), December 16, 2015 (ADAMS Accession No. ML15364A369), February 12, 2016 (ADAMS Accession No. ML16043A496}, April 29, 2016 (ADAMS Accession No. ML16123A014), and June 16, 2016 (ADAMS Accession No. ML16169A139).Exelon Generation Company, LLC (the licensee) requested changes to the technical specifications (TSs), renewed facility operating licenses, and surveillance requirements (SRs) for the Braidwood Station, Units 1 and 2 (Braidwood). The changes increase the ultimate heat sink (UHS) temperature allowable limit in technical specification surveillance requirement 3.7.9.2 from s100 degrees Fahrenheit (°F) to :5102 °F.

The supplements contained clarifying information, did not change the scope of the requested change, and did not change the NRC staff's initial proposed finding of no significant hazards consideration which was published in the Federal Register on March 31, 2015 (80 FR 17088).

The licensee's August 19, 2014, submittal requested a change to TS 3.7.9.2 to increase the UHS temperature limit to::; 102 °F for Braidwood, Units 1 and 2. Currently, the TS SR 3.7.9.2 states: "Verify average water temperature of UHS is s 100°F." If the UHS indicated temperature is> 100 °F, TS 3.7.9 Required Actions A.1 and A.2 would be entered concurrently, requiring both Braidwood, Units 1 and 2, to be placed in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Enclosure 3

The licensee stated that summer meteorological conditions in 2012 resulted in the TS UHS temperature limit being challenged. These conditions include elevated air temperatures, high humidity, and low wind speed. Specifically, July 4 through July 6, 2012, brought unprecedented hot weather and drought conditions to the northern Illinois area resulting in sustained elevated UHS temperatures. Previous revisions of the UHS design analysis used weather data from 1948 to 1976. The licensee updated this UHS design analysis to include weather data through December 31, 2012, which includes the period when the maximum indicated cooling water temperature supplied to the plant from the cooling lake was observed (i.e., July 7, 2012). In response to this new data, the licensee submitted this request to increase the TS UHS temperature limit from s 100°F to s 102°F.

Braidwood, Units 1 and 2, are Westinghouse four-loop pressurized-water reactors with large dry containments. The UHS consists of an excavated essential cooling water pond integral with the main cooling water pond. The UHS dissipates residual heat after reactor shutdown and after an accident through the essential service (SX) water system and the component cooling (CC) water system. Four SX system pumps (two per unit) take suction from the essential cooling water pond and supply cooling water to the safety-related components which include CC heat exchangers, reactor containment fan coolers (RCFCs) and chiller condensers.

2.0 REGULATORY EVALUATION

2.1 Regulatory Requirements The regulatory requirements for which the NRC staff based its review are listed below.

Appendix A to 10 CFR Part 50 provides General Design Criteria (GDC) that must be considered when developing principle design criteria for a water-cooled nuclear power plant. Section 3.1 of the Braidwood and Byron combined Updated Final Safety Analysis Report (UFSAR) discusses conformance with the GDC. The proposed amendment was evaluated against the following GDC, as incorporated into the Braidwood licensing basis through the UFSAR:

• 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," GDC 5, "Sharing of structures, systems, and components [SSCs]," requires that SSCs important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.

• Appendix A to 10 CFR 50, General Design Criteria (GDC) 13, "Instrumentation and control," requires that instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

• 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants,"

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

• 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," GDC 38, "Containment heat removal" requires that a system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any loss-of-coolant accident (LOCA) and maintain them at acceptably low levels.

• 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," GDC 44, "Cooling water," requires that a system to transfer heat from SSCs important to safety, to an UHS shall be provided. The system safety function shall be to transfer the combined heat load of these SSCs under normal operating and accident conditions.

Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

• 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants,"

GDC 50, "Containment design basis," as it relates to the containment heat removal system which shall be 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 any LOCA.

Title 10 of the Code of Federal Regulations (10 CFR), Paragraph 50.36(c)(2), states that limiting conditions for operation (LCOs) are the lowest functional capability or performance levels of equipment required for safe operation of the facility, and when an LCO is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the LCO can be met.

Paragraph 50.36(c)(3) of 10 CFR states that SRs are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

Generic Letter (GL) 96-06, "Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," dated September 30, 1996 (ADAMS Accession No. ML031110021 ), as supplemented by Supplement 1, dated November 13, 1997 (ADAMS Accession No. ML031110029), states if systems are found to be susceptible to the conditions discussed below, addressees are expected to assess the operability of affected systems and take corrective action as appropriate in accordance with the requirements stated in 10 CFR Part 50, Appendix B, and as required by the plant TS to address the issues described below:

• Cooling water systems serving the containment air coolers may be exposed to the hydrodynamic effects of water hammer during either a LOCA or a main steam line break (MSLB). These cooling water systems were not designed to withstand the hydrodynamic effects of water hammer and actions may be needed to satisfy system design and operability requirements.

• Cooling water systems serving the containment air coolers may experience two-phase flow conditions during postulated LOCA and MSLB scenarios. The heat removal assumptions for design-basis accident scenarios are based on single-phase flow conditions and actions may be needed to satisfy system design and operability requirements.

• Thermally induced over pressurization of isolated water-filled piping sections in containment could jeopardize the ability of accident-mitigating systems to perform their safety functions and could lead to a breach of containment integrity through bypass leakage. Actions may be needed to satisfy system operability requirements.

2.2 Regulatory Guidance NUREG-0693, "Analysis of Ultimate Heat Sink Cooling Ponds," dated November 1980 (ADAMS Accession No. ML12146A144).

NUREG-0800, "Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [light-water reactor] Edition." Section 6.2.2, "Containment Heat Removal Systems," Revision 5, dated March 2007 (ADAMS Accession No. ML070160661) describes the acceptable basis of the containment heat removal systems to rapidly reduce the containment pressure and temperature following a LOCA and to maintain these values at acceptably low levels.

NUREG-0800, "Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [light-water reactor] Edition." Section 9.2.5, "Ultimate Heat Sink,"

Revision 2, dated July 1981 (ADAMS Accession No. ML052350549), provides guidance for evaluating UHSs.

Regulatory Guide (RG) 1.105, Revision 3, "Setpoints for Safety-Related Instrumentation," dated December 1999 (ADAMS Accession No. ML993560062), describes a method that the NRC staff finds acceptable for use in complying with the NRC's regulations for ensuring that setpoints for safety-related instrumentation are initially within, and will remain within, the TS limits. RG 1.105 endorses Part I of Instrument Society of America Standard 67.04-1994, "Setpoints for Nuclear Safety-Related Instrumentation," subject to NRC staff clarifications. The NRC staff used this guide to determine the adequacy of the licensee's setpoint calculation methodologies and the related plant surveillance procedures.

RG 1.27, "Ultimate Heat Sink for Nuclear Power Plants," Revision 2, dated January 1976 (ADAMS Accession No. ML003739969), describes an acceptable basis to the NRC staff that may be used to implement GDC 44. The RG states that a UHS serving multiple units should be capable of providing sufficient cooling water to permit simultaneous safe shutdown and cool down of all units it serves and to maintain them in a safe shutdown condition. The RG also

states that in the event of an accident in one unit, the UHS should be able to dissipate the heat for that accident safely, to permit the concurrent safe shutdown and cool down of the remaining unit, and to maintain all units in a safe shutdown condition.

3.0 TECHNICAL EVALUATION

3.1 Instrumentation and Controls In its August 19, 2016, submittal, the licensee stated that as stated in the Braidwood Technical Specification Bases B 3.7.9, the average water temperature of the UHS is measured at the discharge of the SX pumps. The surveillance procedures require that if the temperature of any operating SX pump exceeds 97 °F, a precision temperature instrument, procured for this application, be used to verify the temperature. This is due to the larger uncertainties associated with the installed instrumentation. The submittal identifies the uncertainty of the precision thermometer as having a maximum of 0.07 °F.

In its supplement dated April 30, 2015, the licensee identified the two components of the precision temperature instrumentation as a handheld Hart Scientific Digital Thermometer (model 1521) and a Secondary Reference Thermistor Probe (Model 5610-9) and provided details on how the instruments are stored and maintained. The supplement also identified these components as being the only components in the instrument loop and they account for the entire uncertainty of 0.07 °F. However, the supplement did not identify what uncertainties were used from each device or how they factored into the overall measurement uncertainty. The NRC staff performed a regulatory audit at Braidwood to review the uncertainty calculation and the surveillance procedures for obtaining the UHS temperature. A summary of this audit is provided in ADAMS Accession No. ML15232A589.

During the audit, the NRC staff reviewed Calculation BRW-02-0107-1, Revision 0, "M&TE (Measurement & Test Equipment) Uncertainty Using Hart Scientific Thermometer for Measurement of Essential Service Water Discharge Header Temperature," to determine if the methodology was appropriately implemented and all necessary instrument uncertainties were incorporated. During the audit, staff observed the following:

• The device uncertainties used were identified and they correspond to the relevant uncertainty values identified on the latest instrument data sheets for the Hart Scientific Digital Thermometer and Secondary Reference Thermistor Probe.

• Calculation BRW-02-0107-1 uses the square root sum of the squares (SRSS) methodology to combine the random errors and then algebraically adds the non-random terms. This resulting number is combined using the SRSS with the reading error to produce the final overall uncertainty.

Based on the above, the NRC staff confirmed that the necessary instrument uncertainties for both instruments were appropriately selected and the methodology for calculating the uncertainty was appropriately applied. The NRC staff also confirmed that the licensee has considered the effect of increased UHS temperature and determined it could not adversely impact the uncertainty of the instruments.

During the audit, the NRC staff also observed the surveillance procedures that are used to obtain the UHS temperatures using the precision thermometers. The NRC staff confirmed that the procedures were clear and easy to follow for monitoring the UHS temperature using the precision instrumentation.

Based on the above, the NRC staff concludes that the precision thermometer meets the guidance in RG 1.105, Revision 3, and that GDC 13, "Instrumentation and control," is met.

Further, the NRC staff concludes the licensee adequately addressed the specific device uncertainties and overall uncertainty calculation to meet 10 CFR Section 50.36(c)(3) and demonstrates that the SR verifies that the LCO will be met.

3.1.1 Transient Analysis of the UHS The NRC staff considers that the requirements of the GDCs 5, 38, and 44, are met for the UHS if the license amendment request (LAR) adheres to the guidelines of RG 1.27. The UHS will meet the regulatory requirements and guidance if it provides sufficient cooling water to permit safe shutdown and cool down of the station for 30 days with no UHS makeup for both normal and accident conditions. The UHS must provide adequate cooling water to the SX system for all safety equipment cooled by SX which are needed for accident mitigation of one unit and simultaneous shutdown/cooldown of the other unit. The limiting design-basis-accident (OBA) is one unit with a post-LOCA and loss-of-off-site power (LOOP) with the second unit undergoing a safe non-accident shutdown. Heat that would have to be dissipated by the UHS pond following the OBA would come from containment heat removal, reactor residual heat and engineered safety features (ESF) equipment and the emergency diesel generators (EDGs).

The licensee performed a transient analysis of the SX inlet temperature (UHS return temperature) to determine the peak UHS return temperature when mitigating the OBA under the most adverse meteorological conditions. The analysis used an initial UHS temperature of 102 °F. The UHS return temperature was calculated to peak at 105.2 °F. The licensee's LAR requested that 102 °F be the maximum UHS temperature allowed by TS during Modes 1 through 4 in order not to exceed the 105.2 °F maximum temperature during a OBA. The NRC staff reviewed the licensee's assumptions, design input, methodology and conclusions and compared the licensee's proposed amendment against the above listed regulatory criteria. The NRC staff also contracted with Southwest Research Institute who performed a confirmatory analysis to determine whether the licensee's analysis predicts conservative and reliable results.

The UHS is designed to provide a heat sink for SSCs important to safety after a OBA for a minimum of 30 days without makeup. Braidwood's UHS recycles its water and relies on evaporation, conduction and convection to the environment for heat removal of the energy added to the UHS from the plant during a OBA. Transient analysis is necessary to determine peak UHS temperature, which is the temperature of the water as it enters the SX system to provide cooling to plant components. The transient analysis performed by the licensee, referred to as LAKET-PC, uses the plug flow methodology described in NUREG-0693, "Analysis of Ultimate Heat Sink Cooling Ponds," dated November 1980 (ADAMS Accession No. ML12146A144). The licensee applied similar heat transfer relationships and associated equations of NUREG-0693 to calculate the temperature transients of the UHS return temperature and UHS water inventory during the most limiting design basis event. The heat transfer relationships use various design inputs including wind speed, dry bulb temperature,

dew point temperature, solar and atmospheric radiation, atmospheric pressure, fraction of sky covered by clouds, water surface temperature, saturated vapor pressure, partial vapor pressure, relative humidity and heat input from the plant. The design inputs are taken from the most severe historical weather data that are applicable to the site. The intent in selecting the worst historical weather data is twofold. First, to determine what period of weather data would yield the highest SX inlet temperature (peak UHS return temperature) after a OBA in order to ensure the peak SX inlet temperature would not exceed the maximum allowed cooling water temperature to any of the components to be cooled by the SX. Secondly, to determine the worst weather data for a 30-day period, which causes maximum evaporation losses, in order to ensure that the UHS has sufficient volume without makeup to mitigate a OBA. The current licensing basis used weather data for the years 1948 through 1976 from Peoria and Springfield, IL which are nearby cities. The analysis provided for the proposed TS includes previously used data plus weather data from 1990 to December 31, 2012, from the Braidwood site.

The licensee's LAKET-PC model used the Ryan and Harleman relationships, referenced in NUREG-0693, for heat transfer with the environment (Ryan, P.J., and Harleman, D.R.F. 1973.

"An Analytical and Experimental Study of Transient Cooling Pond Behavior," RM. Parsons Laboratory, Massachusetts Institute of Technology, Technical Report No. 161.). The UHS pond was modeled as plug flow in discrete segments (see NURGE-0693), each segment's length being equivalent to a 3-hour travel time. In the LAKET-PC model, hot water from the plant would enter the first segment, exchange heat with the environment, and then pass to the next segment. Water does not mix between segments. The number of segments depends on the flow rate through the pond, which in turn depends on the number of SX pumps (2, 3, or 4) in operation.

According to the licensee's application:

"The computer program used to model the Braidwood UHS during the design event is ...... a one-dimensional thermal prediction model for bodies of water. The model assumes that the temperature is constant at any point along the plane perpendicular to the direction of the flow. The one dimensional model assumptions coerce the water body into an idealized rectangular channel. The movement of fluid through the one-dimensional channel is envisioned as a series of individual, distinct fluid segments. Each segment has an individual length and temperature, while the width and depth remain constant for all. The channel thus forms a queue of fluid segments, where additions are made at the inlet, and deletions are made at the outlet. Any segment that enters the channel will cause an equal amount to be expelled at the outlet. The program assumes that all segments are uniform in temperature, and each segment is allowed to react independently with the environment. The horizontal heat conduction for each segment is assumed to be negligible with respect to the heat rejection at the air-water interface and is ignored. Similarly, conductive heat loss at the water channel interface is ignored."

The licensee's one-dimensional model assumes no thermal stratification since the model depends on the temperature being constant at any point along the plane perpendicular to the direction of the flow. The plug flow model uses UHS volume and surface area. Since the UHS pond contains stagnant areas which are not part of the UHS flow path, the licensee performed a

separate calculation which determined the effective area and volume to be 82.3 percent of the design gross values. The volume and surface area of the plug flow model was set according to the effective volume and surface values.

The methodology used in the plug model is consistent with the thermal model presented in NUREG-0693, "Analysis of Ultimate Heat Sink Cooling Ponds," and the wind speed functions presented in MIT Report No. 161, "An Analytical and Experimental Study of Transient Cooling Ponds Behavior," dated January 1973, which is also referenced and cited in NUREG-0693.

Consistent with NUREG-0693, the plug flow contributing components for the net heat transfer to the UHS are:

Q =OsN + 0AN - OsR - OE - Oc + 0RJ Where:

OsN = net incident short wave solar radiation QAN = net incident long wave atmospheric radiation OsR = net rate of long wave back radiation from the lake surface OE = net rate of heat loss due to evaporation Oc = net rate of heat loss due to conduction and convection QRJ = net rate of heat rejected to the lake by the plant Other than heat rejected from the plant, the contributing components of heat transfer to/from the UHS are functions of the ambient weather conditions. Consistent with RG 1.27, the licensee should consider the worst weather conditions as determined by a recent period of record of at least 30 years of representative weather data. The licensee's analysis should account for the two worst-weather periods, i.e., the period that results in the highest peak SX temperature (UHS return temperature) of cooling water inlet to the plant and the worst 30 days of weather that result in maximum evaporation to ensure that the UHS pond has sufficient volume to perform its safety function for a minimum of 30 days without makeup.

The licensee created a file of meteorological data from three sources: (1) the on-site meteorological tower, (2) the weather station at Peoria, IL, and (3) the weather station at Springfield, IL. Braidwood is located -100 miles from Peoria, -165 miles from Springfield, and

-50 miles from the nearest large water body (i.e., Lake Michigan). The file covered the period from July 7, 1948, through December 31, 2012. On-site data were only available from January 1, 1990, through December 31, 2012, and furnished only part of the information necessary for running the LAKET-PC code; i.e., dry bulb temperature, dew point, and wind speed. Other data for that period; namely, cloud height, cloud cover, and atmospheric pressure, came from Peoria. In addition, data from Peoria was used to fill in gaps in the record of data from the on-site tower. The licensee made checks of the weather record to determine the validity of the data, and identified periods when data were missing or out of range. They used procedures such as linear interpolation between available data or substituting data from Peoria to fill in the record. They also checked the thermodynamic consistency of the data between the temperature and humidity. When the dew point from Peoria was substituted for missing dew point from the on-site tower, there was a check to make sure that dew point did not exceed dry bulb temperature. If so, the dew point was set equal to the dry bulb temperature; i.e., not greater than 100 percent relative humidity. Prior to January 1, 1990 (i.e., a period of -42 years), all

data in the file were collected from Peoria with the exception of the period January 1, 1952 to January 31, 1956 (a period of -4 years), when data were taken from Springfield, IL. Other data such as wind direction and precipitation were recorded in the data file but not used in the LAKET-PC analyses.

To account for thermal input from solar radiation the licensee described the use of a semi-physical model that used available meteorological information on cloud cover and cloud height, latitude, atmospheric pressure, dew point, time of year, atmospheric transmission factors, and surface and cloud albedo, to generate solar radiation incident on the UHS pond (Calculation AT0-0109, Revision 4, Page B7, Attachment 5 of LAR). The results of the algorithm were validated with actual measurement of solar radiation for Madison Wisconsin collected by the National Oceanic and Atmospheric Administration and found to be consistent by the licensee.

The licensee's screening for identifying worst-weather periods used the LAKET-PC computer code, starting with a fixed water temperature entering the pond, no recirculation back to the plant, and using a 3-hour travel time through the pond. The initial temperature of the pond was reset to the same initial temperature at the start of each run. Temperature was calculated using the weather data that was created from the site, Peoria and Springfield meteorological data.

For the peak temperature screening, the licensee determined the highest temperature that would occur with water at a fixed starting temperature and travel times of 24, 36, or 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> based on a flow rate provided by 4, 3 or 2 SX pumps, respectively. For weather screening purposes, the choice of starting temperature for screening was not critical, as only the time period of worst performance was being screened. A "rolling average" temperature over a 24-hour period was calculated from the results of LAKET-PC in order to define the "worst day" for peak temperature as a function of starting time of the OBA. Weather scenarios considered were:

• Worst 24-hr, 36-hr, 48-hr weather results with OBA starting at 12 AM, 3 AM, 6 AM,

• 9 AM, 12 PM, 3 PM, 6 PM, or 9 PM.

• Worst 30-day weather results with OBA starting at 12 AM, 3 AM, 6 AM, 9 AM, 12 PM, 3 PM, 6 PM, or 9 PM.

Combining the worst-time period from the screening followed by the worst 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, and also followed by the worst 30-day period.

• 30-day period leading to worst-net evaporation.

The licensee screened the available meteorological data to produce a number of meteorological data sets they expected to produce the peak UHS return temperature to the plant, and also the maximum 30-day evaporation rate. Each data set for a maximum temperature run was specified for the number of SX pumps in operation (2, 3, or 4) and one of eight OBA starting times (12 AM, 3 AM, 6 AM, 9 AM, 12 PM, 3 PM, 6 PM, or 9 PM ). The worst 24, 36 or 48-hour weather periods selected by the licensee to estimate peak return temperature were determined.

To determine the worst-weather period for the maximum 30-day water loss, the licensee performed a similar screening. The NRC staff used the METO suite of programs from NUREG-0917 to evaluate the meteorological data. The NRC staff notes that according to Figure 2 of Attachment 1 in its letter dated August 19, 2014, the licensee shows that for an initial

UHS outlet temperature at 102 °F (TS limit) from approximately 9 AM to 3 PM that the UHS outlet temperature may exceed 102 °F later that day under the most challenging meteorological conditions. Exceeding 102 °F later in the day does not affect the acceptability of the results, but would cause the licensee to enter the TS LCO. The NRC staff found that the data had a minimum of aberrant errors, none of which were a concern for the UHS study.

The licensee's analysis determined that the 3 AM start would give the highest temperature peak for the three SX pump case. The UHS temperature peaks at 105.2 °F. The three SX pump case is representative of the design basis case since an additional SX pump would start (one is already running) in the accident unit and one SX pump would be running for cooldown of the non-accident unit. The peak occurs 39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> after the OBA and exceeds 104 °F for less than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. All equipment cooled by the SX system is designed to run satisfactorily at 105.2 °F during OBA mitigation (calculations were performed at 106 °F) with the exception of the RCFCs as explained in the containment analysis described in Section 3.5. The evaluation of all other equipment cooled by the UHS at 105.2 °F is presented below in Section 3.4.

3.3 UHS Pond Hydraulics The licensee's one-dimensional model assumes no thermal stratification since the plug flow model depends on the temperature being constant at any point along the plane perpendicular to the direction of the flow. Thermal stratification, which could result mainly from the discharge of heated water into the cooler and denser ambient pond water, potentially enhances flow along the surface and reduces the travel time through the pond. The licensee defends its conclusion that the pond would not be stratified on the basis of calculations from a theoretical model for the behavior of a distinct two-layer system representing water density variation (i.e., hot water sitting on top of cool water), and no interfacial mixing except in the entrance region. This procedure is generally known as the "pond number" approach. The licensee concluded from the calculation that there would be no appreciable stratification for 2, 3, or 4 SX pumps in operation. Although, this is not necessarily non-conservative because evidence from other shallow ponds frequently show they are vertically well mixed, the NRC staff questioned the licensee's assumptions of no stratification because there is no accurate accounting of near field mixing at the surface discharge. To resolve its question, the NRC staff performed confirmatory analysis by sensitivity calculations that postulated stratification in the pond. The NRC staff's sensitivity calculations showed that stratification would not result in a return water temperature greater than that calculated by the non-stratified assumption. The analysis assumed that there was a stratified water layer floating on the top of a cooler stagnant layer. Travel time from the discharge to the intake would be half as long, and water entering the intake structure and sumps would mix with the cooler underlying water in equal proportions. The sensitivity analysis showed that, even if stratification occurred, it would not lead to degraded UHS cooling performance.

The NRC staff questioned whether the licensee's values for effective volume and surface area, which are important design input for the one dimensional plug flow model, are appropriate and conservative for determining peak UHS temperature. The staff's question centered upon the effect of the discharge configuration where the SX discharge is vertical above the water surface with estimated speed of 24 ft/sec (feet/second) for the 4 SX pump case, falling back into the pond causing extensive mixing in the near field. Probable effects include vertical and horizontal mixing, significant recirculation, and distribution of travel times. To determine whether the effective volume and surface area adequately account for this type of discharge, the NRC staff

asked the licensee to provide actual measurements of pond circulation or perform sensitivity studies over a range of effective volume and surface areas to verify conservatism in their calculation of effective volume and surface area. The licensee responded in its letter dated December 16, 2015, stating actual measurements of pond hydraulics are not available but they performed sensitivity studies with a higher and lower value of effective volume and surface area.

Since the licensee used 3-hour plug volumes in LAKET-PC, they used effective volumes based on plus and minus 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> from their base case three SX pump transit time. The results were a 76 percent and 87 percent effective volume and surface area and a 105.9 °F peak UHS inlet temperature. However, the licensee then recalculated peak UHS temperature based on a recent (October 2014) hydrographic survey, which determined a larger gross area and surface area than what was previously used. The larger gross volume returned the peak UHS return temperature to 105.2 °F for the 76 percent effective sensitivity case. The NRC staff finds this acceptable because the peak temperature of 105.2 °F used in the equipment analysis is not exceeded.

Additionally, the NRC staff questioned the licensee's use of large 3-hour plug segments when the output is expressed in fine increments (i.e., tenths of a degree). The peak UHS temperature is expressed as 105.2 °F. If the 1-hour plug segments were used, the initial plug temperature would be 5 °F higher than in the 3-hour plug size cases. The licensee responded in letters dated December 16, 2015, and February 12, 2016, stating that sensitivity studies show a peak UHS temperature of 106.05 °F when using the 1-hour plug segments. The licensee recalculated peak UHS return temperature based on the recent (October 2014) hydrographic survey, which determined a larger gross area and surface area than what was previously used.

The recalculated peak UHS temperature was 105.5 °F. The licensee stated that 105.5 °F for the 1-hr plug case is acceptable due to the impact of mixing in the near field making the 3-hour plug model acceptable at 105.2 °F. The NRC staff finds the licensee's response acceptable because the sensitivity analysis shows the licensee's overall methodology and results to be conservative.

3.4 Evaluation of Safety Equipment Cooled by the UHS/SX Systems The UHS is a heat sink for the ESF equipment cubicle coolers, ESF pump heat exchangers, EOG coolers, diesel driven auxiliary feedwater engine coolers, RCFCs, main control room (MCR) chiller condensers, CC heat exchangers which in turn cool residual heat removal (RHR) heat exchangers, chemical and volume control system and the process sampling system.

The licensee stated that all EOG coolers have margin for the new proposed limiting UHS temperature of 105.2 °F peak temperature after a OBA. The EOG vendor data sheet provided by the licensee shows an SX flow rate of 1641 gpm (gallon per miniute) at 100 °F for the design heat load. The NRC staff asked the licensee to provide actual EOG analytical and operating data (i.e. design fouling factor, as tested fouling factor, frequency of testing, tube plugging allowance, actual number of plugged tubes, design heat load, actual SX flow rate, calculated heat removal capability with design fouling factor at 105.2 °F) to substantiate the licensee's claim that the EDGs have adequate margin at the peak UHS temperature. In a letter dated December 16, 2015, the licensee provided the requested information, except they were unable to provide an as-tested fouling factor as they do not conduct performance testing on these heat exchangers, because the heat exchangers are in the GL 89-13 program and are cleaned at a proven frequency. Design calculations provided by the licensee show that the 110 percent

design heat load of the EDGs can be removed with an SX temperature of 106 °F and 54 plugged tubes per bundle using the conservative vendor supplied fouling factor value of 0.0025 and an SX flow rate of 1650 gallon per minute (gpm). The maximum number of tubes plugged in any bundle is currently 30 tubes and SX flow rate to the jacket water heat exchangers has been verified during surveillance testing of the EDGs to be greater than 1650 gpm (in the range of 1870-1920 gpm). Based on the information provided by the licensee, the NRC staff concludes the licensee's analysis to demonstrate that the EDGs will operate satisfactorily during a OBA with SX temperature at 105.2 °F to be reasonable and acceptable because the licensee's calculations show that 110 percent of the design heat load of the EDG's can be removed with a higher SX temperature of 106 °F.

The diesel-driven auxiliary feedwater pumps engines are also cooled by the SX system.

Licensee calculations show that with 106 °F SX cooling temperature, a design flow of 250 gpm and one tube plugged (currently one tube is plugged in the Unit 2 engine heat exchanger), the alarm setpoint is 1 °F higher than the calculated shell inlet temperature. Surveillance testing shows actual SX flow to the engine heat exchanger is closer to 350 gpm (100 gpm higher that the 250 gpm design flow). The increased flow beyond design flow provides adequate margin for an SX temperature of 106 °F. Based on the information provided by the licensee, the NRC staff concludes the licensee's analysis which demonstrates that the diesel driven auxiliary feedwater pumps will operate satisfactorily during a OBA with SX temperature at 105.2 °F to be reasonable and acceptable.

The MCR chiller condenser is also cooled by SX. With a revised peak SX temperature of 105.2 °F, the licensee stated that over 25 percent margin exists using reduced fouling factors from the design based on comparable heat exchangers in the GL-89-13 program. In its request for additional information (RAI) email dated March 9, 2015 (ADAMS Accession No. ML15069A004), the NRC staff asked the licensee to justify the use of reduced fouling factors.

In its RAI response dated April 30, 2015, the licensee stated that the licensee compared the unknown fouling factor of the MCR to the calculated fouling factor of the closed cooling water (CCW) heat exchangers because the MCR chillers are cleaned in a similar manner as the CCW heat exchangers in accordance with the GL 89-13 program. The calculated/actual fouling factor of the CCW heat exchangers were less than one-half the design value of the MCR chillers.

The NRC staff further questioned the licensee's comparison of fouling factors of other heat exchangers with the MCR since the other heat exchangers may not have similar SX flow aligned to them as regularly as does the MCR. The licensee, in its letter dated December 16, 2015, providing a modified evaluation with an increased level of detail. The new model allowed for variation of the refrigerant pressure in the chiller condenser, thus, altering the driving temperature difference across the heat exchanger surface (with higher chiller condenser pressures a higher saturation temperature will exist resulting in increased differential temperature). This altered calculation demonstrated that at the current worst-tube plugging level, the MCR chiller has 18 percent margin with an SX temperature peak up to 106 °F. This margin is likely greater because the analysis used the design fouling factor of 0.0015. The actual fouling factor would be smaller because in the licensee's letter dated April 30, 2015, they described the comparison of the MCR chiller fouling factor to the calculated/actual fouling factors of the component cooling water heat exchangers. The calculated/actual fouling factor was less than one-half the design value of the MCR chillers. Based on the licensee's response,

the NRC staff concludes that there is reasonable assurance that the MCR chillers will satisfactorily operate under design basis conditions with an SX temperature of 105.2 °F.

The SX system provides cooling water to various safety-related pump cubicle coolers for the following systems, the essential SX, RHR, safety injection (SI), coolant charging, spent fuel pool (SFP) cooling, diesel driven auxiliary feedwater and containment spray. Calculations performed by the licensee with an SX cooling temperature of 106 °F show greater than 1O percent margin for heat transfer requirements and margin for tube plugging limits for all coolers (except containment spray and SFP cooling which have less than 8 percent margin). All coolers are monitored consistent with the GL 89-13 program. Based on the licensee's response, the NRC staff concludes the safety-related pump cubicle coolers will satisfactorily perform their function under design basis conditions with an SX temperature of 105.2 °F because the licensee's calculations showed that at a higher SX temperature (106 °F) than the design (105.2 °F) the coolers can remove the design heat load with additional margin.

The SX system provides cooling water to various safety related pump oil coolers including pumps for the following systems; the essential SX, SI, coolant charging centrifugal, and diesel-driven auxiliary feedwater. Calculations performed by the licensee with an SX cooling temperature of 106 °F show margin for heat transfer and tube plugging requirements. These heat exchangers are cleaned and inspected per the GL 89-13 program. Based on the licensee's response, the NRC staff concludes the safety-related pump oil coolers will satisfactorily perform their function under design basis conditions with an SX temperature of 105.2 °F because the licensee's calculations showed that at a higher SX temperature (106 °F) than the design (105.2 °F) the coolers can remove the design heat load with additional margin.

The CC water heat exchangers are cooled by SX. The CCW heat exchangers were evaluated in five different scenarios including normal operation and design basis accident mitigation. The maximum allowed CC water heat exchanger outlet temperature is 105 °F during normal operation and 120 °F after initiation of RHR for a normal reactor coolant system (RCS) cooldown. In its letter dated March 31, 2015, the licensee stated that the Braidwood design basis allows CC heat exchanger outlet temperature to be 128 °F for the accident and non-accident unit during a DBA. In its letter dated February 12, 2016, the licensee stated that they evaluated the performance of the CC heat exchangers for the five scenarios using an SX temperature of 102 °F for normal operations and 106 °F for DBA conditions and found sufficient CCW heat exchanger margin in maintaining the maximum allowed CC water heat exchanger outlet temperatures stated above. Based on the licensee's response, the NRC staff concludes the CCW heat exchangers will satisfactorily function under design basis accident conditions with an SX temperature of 106 °F and will satisfactorily function under normal operating conditions with an SX temperature of 102 °F.

The licensee has performed net positive suction head calculations for the pumps ( i.e. SX and SX booster), and the motor- and diesel-driven auxiliary feedwater, that are affected by the new temperature limits of the UHS and has determined that there is sufficient margin between NPSH (net positive suction head) and NPSHR (net positive suction head required). Based on the demonstration of sufficient margin, the NRC staff finds this analysis acceptable.

Subsequent to the licensee's submittal of the LAR, the licensee addressed Westinghouse lnfoGram (IG)-14-01, dated November 5, 2014, regarding RCS metal specific heat and density,

which resulted in an increase in the energy released inside containment during a OBA. The added heat energy released results in a containment sump temperature increase by approximately 3 °F. This in turn increases the peak heat load on the RHR and CCW heat exchangers by approximately 2 percent during the time 1899 seconds through 9999 (end of analysis) seconds after the OBA. The added heat load is within the capability of the heat exchangers and the NPSH capability of the CCW pumps. The added heat would conservatively cause the licensee's peak calculated UHS return temperature to increase by 0.1 °F which has minimal effect on the equipment cooled by SX during the OBA. The small added heat load would not have any effect on the NRC staff's confirmatory analysis because the computed peak temperature is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after OBA initiation showing that the peak temperature is a result of heat transfer into the pond from the environment and not the result of heat energy from the plant.

3.5 Accident and Containment Analysis As documented in letter dated March 31, 2015, the licensee evaluated the effect of UHS temperature of 105.2 °F on the UFSAR, Chapter 15, "Accident Analyses and Transient Analyses." The licensee's analysis show that 105.2 °F is bounded by temperature assumptions for the UHS in the applicable accident analyses and transient analyses. Based on the above, the NRC staff concludes that these analyses are acceptable for the UHS TS temperature increase to 102 °F.

The licensee identified two containment integrity scenarios that could potentially be affected because UHS temperature is a design input for these analyses. The analyses identified by the licensee are: (i) "LOCA Long Term/Short Term Mass and Energy (M&E) (Containment Integrity)," and (ii) "MSLB Inside Containment/Outside Containment Mass and Energy-Dose Steam Release (Containment Integrity)."

The licensee stated that the most limiting scenario that imposes maximum heat load on the UHS is the one in which one unit is undergoing a post-design basis LOCA cooldown concurrent with a LOOP, in conjunction with the non-accident unit undergoing a normal safe shutdown.

This limiting scenario includes three sources of heat to be removed by the SX system post-LOCA: (a) containment heat removal via the RCFCs, (b) containment heat and reactor decay heat removal via the containment sumps, and ESF equipment heat loads (e.g., ESF equipment coolers and room coolers).

The licensee considered the diurnal variations of the UHS temperature and evaporation response by varying the design basis LOCA start time and analyzed the UHS performance with the proposed average water temperature 102 °F for the most limiting scenario. Based on the analysis, the licensee selected the following temperatures for the accident analysis and equipment evaluation: (a) 104 °F for analyzing the heat removal performance of the RCFCs and (b) 105.2 °F for analyzing the performance of equipment cooled by the UHS (SX cooled equipment calculations were performed at 106 °F). The licensee stated that the accident analysis temperature of 104 °F is conservative because UHS temperature remains below 104

°F for the first 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> into the event and increases to a maximum of 105.2 °F for a period less than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> from hour 36 to hour 42 post accident; and remains below 104 °F after 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br />.

Based on the above, the NRC staff finds the licensee's analysis consistent with the guidance in RG 1.27 and SRP 6.2.2.

The limiting post-OBA UHS temperature responses were determined and are used: (1) as input into the safety analysis to ensure containment responses remained within analyzed limits and (2) to evaluate performance margins of equipment cooled by the SX and CC systems.

Changes in the Current Licensing Basis To support the UHS TS SR 3.7.9.2 change from :s;100 °F to :s;102 °F, the licensee changed the design UHS water temperature to the SX system for the purposes of the containment analysis from 100 °F to 104 °F. The SX system cools the RCFCs and the CC heat exchangers. The CC system in turn provides cooling water to the RHR system heat exchangers. As stated in response to SCVB-RAl-3 (Reference 2), the licensee made the following changes via 10 CFR 50.59 when updating the containment analysis for this amendment:

a) Reduction in containment spray flow from 3285 gpm to 3113 gpm.

b) Reduction in emergency core cooling system (ECCS) flow rates.

c) Correction to the SATAN78 power shape selection option to select a chopped cosine power shape.

d) Incorporation of an NRG-approved model (WCAP-10325-P-A, "Westinghouse LOCA Mass and Energy Release Model for Containment Design March 1979 Version", May, 1983) for drift flux and break flow with inertia. The SATAN78 input model was revised to include and credit the effects for drift flux and break flow with inertia consistent with the NRC's safety evaluation report approval for WCAP-10325-P-A LOCA M&E release methodology evaluation model. The result of applying the drift flux model is a small benefit in blowdown M&E releases. In addition, the WCAP-10325-P-A methodology has been approved to incorporate improved fluid momentum flux terms, which can be applied to the break flow model, and can result in a small reduction in mass and energy release.

e) Correction of errors reported in Westinghouse Nuclear Safety Advisory Letter (NSAL)-14-2 which consists of using corrected steam generator (SG) tube material density and specific heat in the mass and energy release model.

Besides NSAL-14-2 (Reference 7), Westinghouse has issued NSALs-06-6 (Reference 5) and NSAL-11-5 (Reference 6), which are not mentioned in the LAR (Reference 1). In SCVB-RAl-2 (Reference 2), the NRC staff requested the licensee to provide the resulting changes in the following containment analyses results after correcting the M&E analysis methodology from errors reported in all NSALs: (a) containment peak pressure, (b) containment peak gas temperature for environmental equipment qualification (EEQ), (c) containment peak wall temperature, (d) containment sump peak water temperature, (e) pump NPSH and NPSHA for the pumps that draw water from the containment sump during recirculation mode of SI and CC, and (f) containment minimum pressure analysis for ECCS performance capability. In response to SCVB-RAl-2(a), (b), (c), and (d), (Reference 2), the licensee stated that all issues reported in the three NSALs were addressed in the LOCA M&E release for all of the above containment performance analyses.

In addition to the NSALs, the NRC staff became aware of a Westinghouse IG-14-1, dated November 5, 2014. This document reported that the Reference 3 M&E release analysis methodology used volumetric heat capacity for the RCS stainless steel less than the American Society for Mechanical Engineers (ASME) published value.

The NRC staff requested information to address the errors reported in the IG-14-1. In response, the licensee performed Braidwood, Units 1 and 2, specific sensitivity analyses for LOCA M&E release using the volumetric heat capacity calculated from the ASME material properties values as discussed in IG-14-1. In the analysis, the licensee used 72.6 Btu/ft3/°F as the revised bounding value of the volumetric heat capacity for the temperature range in the RCS, reactor vessel and SG secondary metal. The revised value replaces the current value of 53.2 Btu/ft3 -

°F for the volumetric heat capacity of the RCS metal. The licensee's containment analysis results before correction of errors reported in IG-14-1 and the sensitivity analysis results after correction of errors reported in IG-14-1, in response to SCVB-RAl-15(a) (Reference 12), are given below under the heading "Containment Analysis."

Containment Analyses The NRC staff evaluated the impact of an increase in the UHS TS SR 3.7.9.2 temperature limit on the following LOCA containment parameters related to containment integrity, containment heat removal, and ECCS analysis:

a) Short-term subcompartment differential pressures for subcompartment integrity, b) Long-term peak pressure for containment integrity, c) Long-term peak wall temperature for containment integrity, d) Long-term peak gas temperature for EEQ, e) Peak sump water temperature for sump heat removal, f) Pump NPSHA for the pumps that draw water from the containment sump during recirculation mode of SI and CC, and g) Minimum containment pressure for ECCS performance capability.

Short-Term M&E Release and Sub-compartment Integrity Analysis The short-term M&E release analysis is performed to evaluate the containment sub-compartment differential pressures to ensure that the sub-compartments will maintain their structural integrity during the short pressure pulse, generally less than 3 seconds, following a high energy line break. The licensee stated that the impact on the RCFCs and the RHR heat exchanger performance due to an increase in the SX and CC systems cooling water temperature takes place during the post-blowdown phase of the LOCA because these system do not have any role during the initial blowdown. Therefore, due to the short duration of the M&E release, the current licensing basis sub-compartment differential pressures are not affected. The NRC staff finds it acceptable that the current licensing basis sub-compartment differential pressure are not affected by the UHS temperature increase because the SX and CC systems are not used in the short-term and, therefore, the sub-compartment integrity will be maintained with the proposed change.

Long-Term M&E Release and Containment Pressure Response Containment M&E Release Analysis The long-term LOCA M&E release is an input to the containment integrity analysis to demonstrate that peak containment pressure and temperatures are below their design limits.

For the M&E analysis, the licensee used WCAP-10325-P-A (Reference 3), methodology,

consistent with the current licensing basis analysis, but corrected for the errors reported in NSALs 06-6, 5, 2, and IG 14-1, and determined the M&E release for approximately 106 seconds. For the containment pressure and temperature response, the licensee used COCO (Reference 4), a computer code consistent with the current licensing basis analysis. The licensee was requested to provide a comparison of the inputs and assumptions in the proposed analysis that were changed from the current analysis, and provide justification for those inputs and assumptions in which the conservatism in the proposed analysis is reduced. In its response (Reference 2), the licensee stated the following input changes resulted in an increased containment pressure response: (a) reduction in containment spray flow from 3285 gpm to 3113 gpm resulted in a containment pressure increase due to reduced heat removal; (b) reduction in ECCS flow rates resulted in a containment pressure increase due to a reduction in the condensation of steam (with the cold ECCS water), releasing more energy (from steam) into the containment; (c) correction of errors reported in Westinghouse NSAL-14-2 which consists of using corrected SG tube material density and specific heat in the M&E release model; (d) correction to the SATAN78 power shape selection option to select a chopped cosine power shape; and, (e) incorporation of an NRG-approved model (Reference 3), for drift flux and break flow with inertia. The SATAN78 input model was revised to include and credit the effects (potential benefits) for drift flux and break flow with inertia consistent with the NRC's SER approval for WCAP-10325-P-A (Reference 3), LOCA M&E release methodology evaluation model. The result of applying the drift flux model is a small benefit in blowdown M&E releases.

In addition, WCAP-10325-P-A (Reference 3), evaluation methodology has been approved to incorporate improved fluid momentum flux terms, which can be applied to the break flow model, which also can result in a small reduction in M&E release.

Containment Pressure Response The limiting calculated peak containment pressure is for the double-ended pump suction (DEPS) break. Table 1 provides the licensee's containment peak pressure results before correction of errors reported in IG-14-1 and the sensitivity analysis results, after correction of errors reported in IG-14-1 for the DEPS and double-ended hot-leg (DEHL) breaks.

Table 1: Peak Containment Pressure Results Peak Containment Pressure (psig) Containment Break Unit Before IG-14-1 After 1G-14-1 Pa (psig) Design Pressure Error Correction Error Correction (psig) 1 42.1 42.2 42.8 DEHL 2 37.7 37.8 38.4 50 1 42.1 43.8 42.8 DEPS 2 38.4 40.2 38.4 The sensitivity analysis, after correcting errors reported in IG-14-1, shows that the limiting peak containment pressure is below the containment design pressure of 50 psig (per square inch gauge). However, the limiting peak pressure exceeded the lntegrateq Leak Rate Test (ILRT)

pressure 'Pa' from 42.8 psig to 43.8 psig for Unit 1, and from 38.4 psig to 40.2 psig for Unit 2. In response to NRC staff questions, the licensee calculated the impact on the current ILRT results for both units. The result showed an increase from 0.05571 percent-weight/day to 0.056 percent-weight/day for Unit 1, and zero change from the current value of 0.1083 percent-weight/day for Unit 2 which are bounded by the permissible ILRT leakage rate limit is 0.15 percent-weight/day. Regarding the Local Leak Rate Test (LLRT), the current as-left maximum pathway leakage rate of 66.909 standard cubic feet per hour (scfh) for Unit 1 and 44.239 scfh for Unit 2 are based on test pressures of 44.98 psig for Unit 1 and 40.88 psig for Unit 2. These test pressures are greater than the sensitivity analysis peak pressures of 43.8 psig for Unit 1 and 40.2 psig for Unit 2. Therefore, increased sensitivity analysis calculated peak pressure would have no impact on the current LLRT leakage rates which are below the permissible leakage rate limit of 540.48 scfh for Unit 1, and 499.12 scfh for Unit 2.

Since the licensee performed a sensitivity analyses instead of a full revision of the containment abnormal occurrence report (AOR), the licensee has not revised the TS 'Pa' values from 42.8 psig for Unit 1, and 38.4 psig for Unit 2. As required by 10 CFR Part 50, Appendix B, Criterion Ill, "Design Control," and Criterion XVI, "Corrective Action," as well as the licensee's processes that implement Appendix B, the licensee is required to revise the TS 'Pa' as appropriate, after the AOR is revised to implement corrective actions to resolve the IG-14-1 issue. However, the licensee has demonstrated that the revised ILRT and LLRT test results based on the sensitivity analysis results of peak containment pressure of 43.8 psig for Unit 1 and 40.2 psig are bounded by the acceptable limits. Based on the demonstration that the peak containment pressures are bounded, the NRC staff concludes that the licensee's action is acceptable for this amendment.

Containment Temperature Response The limiting calculated peak containment temperature for EEQ is for the DEPS break. Table 2 provides the licensee's peak containment temperature results for environmental qualification (EQ) before correction of errors reported in IG-14-1 and the sensitivity analysis results, in response to SCVB-RAl-15(a) (Reference 12), after correction of errors reported in IG-14-1 for the DEPS and DEHL breaks.

Table 2: Peak Containment Temperature Results for EEQ Peak Containment Temperature (°F) EQ Profile Temperature Profile Break Unit Before IG-14-1 After 1G-14-1 at Time of Error Correction Error Correction Maximum Steam Temperature (°F) 1 263.50 263.62 325 DEHL 2 256.53 256.68 320 1 262.60 265.24 270 DEPS 2 256.69 259.60 270

The sensitivity analysis, after correcting errors reported in IG-14-1, shows that the limiting peak containment temperatures are bounded by the EEQ temperature profile at the time of maximum steam temperature.

Containment Sump Water Temperature Response In response to NRC staff questions (Reference 2), the licensee provided results for the sump water temperature responses for DEHL and DEPS breaks both with minimum ECCS flows. The sump water temperature response is an input to the NPSH analysis described below. Table 3 provides the licensee's peak sump water temperature results before correction of errors reported in IG-14-1 and the sensitivity analysis results, after correction of errors reported in IG-14-1 for the DEPS and DEHL breaks.

Table 3: Sump Water Temperature Results Peak Sump Water Temperature (°F)

Break Unit Before IG-14-1 Error After 1G-14-1 Error Correction Correction 1 252.01 251.99 DEHL 2 248.41 248.42 1 258.14 261.43 DEPS 2 252.90 254.86 RHR and Containment Spray (CS) Pump NPSH Analysis In response to NRC staff questions (Reference 2), the licensee provided data for the most limiting NPSH margins (NPSHA minus NPSH required NPSHR) versus sump temperature for the RHR and CS pumps during the recirculation mode of operation at SX water temperature of 104°F (see response to SCVB-RAl-14 in Reference 10). The licensee also conservatively evaluated intermediate points of RHR and CS pump suction lines for water flashing into vapor and evolving of air for acceptable performance. The NPSH margins were positive and the void fractions in the pump suction paths are sufficiently small for acceptable for pump operation.

Table 4 provides the licensee's NPSH analysis results before correction of errors reported in IG-14-1 and the sensitivity analysis results, in response to SCVB-RAl-15(a) (Reference 12), after correction of errors reported in IG-14-1 for the RHR and CS pumps.

Table 4: NPSH Results NPSH Limiting Margin Pump Before IG-14-1 Error After 1G-14-1 Error Correction Correction RHR 2.6 feet at 248.4°F 2.5 feet at 258.1°F cs 1.3 feet at 248.4°F 1.3 feet at 258.1°F The sensitivity analysis results after correction of errors reported in IG-14-1 shows positive NPSH margin for the RHR and CS pumps for both units and are, therefore, acceptable.

Minimum Containment Pressure for ECCS Performance Capability In response to NRC staff questions (Reference 2), the licensee stated that the errors reported in the three NSALs do not affect the minimum containment pressure analysis for ECCS performance capability. The NRC finds the licensee's response acceptable because the current analysis is conservative compared to the corrected analysis.

Main Steam Line Break CMSLB) Containment Analysis The licensee analyzed MSLB cases for the increased UHS temperature to the RCFCs and determined that the results for the peak containment pressure are bounded by the current design basis analysis. The limiting peak pressures for Unit 1 was calculated for a 0.9 ft 2 split break at 30 percent initial power, with a single-failure of the main steam isolation valve (MSIV) in the faulted loop, and with offsite power available. The peak pressure for Unit 1 was determined to be 34.5 psig. The limiting peak pressures for Unit 2 was calculated for a 0.83 ft 2 split break at 30 percent initial power, with a single-failure of the MSIV in the faulted loop, and with offsite power available. The peak pressure for Unit 2 was determined to be 34.4 psig. Both cases for each unit are well below the 50 psig containment design pressure.

The licensee analyzed MSLB cases for the increased UHS temperature to the RCFCs and determined that the results for the peak containment temperature are bounded by the current design basis analysis. The limiting peak temperatures were for a 1.0 ft 2 split break at 100%

initial power, with a single failure of the main steam isolation valve in the faulted loop for Unit 1, and a 1.0 ft 2 small double-ended rupture (DER) at 100% and 70% initial powers, with a single failure of the main steam isolation valve in the faulted loop for Unit 2. The limiting MSLBs assume a loss of offsite power and 0% revaporization of the condensed liquid on the containment surfaces. The peak temperature of Unit 1 was determined to be 329.0 °F which is less than the maximum Unit 1 temperature of 330.8 °F from the current design basis analyses.

The peak temperature of Unit 2 was determined to be 323.0 °F which is less than 326.3 °F from the current design basis analyses.

The licensee stated that the composite containment temperature response for Unit 1 and Unit 2 were compared to the design EEQ temperature profile. The revised temperature profile is enveloped by the existing design EEQ profile. The EEQ pressure profile uses a flat 50 psig

curve until 1200 seconds into the event and then a saturated pressure model after 1200 seconds. Therefore, the results of the analyses do not change the EEQ profile.

The NRC staff finds the licensee's evaluation for the peak containment pressure and temperature with the increased UHS temperature acceptable because the results are bounded by the current design basis analysis, and the revised EEQ temperature profile is bounded by the current EEQ profile.

Response to GL 96-06 In reference to NRC GL 96-06, "Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions", the licensee identified a concern for possible water hammer following either a LOCA or a MSLB concurrent with a LOOP during the first few minutes post-accident while the pumps and fans are restarting following the LOOP. Under this scenario, the pumps that supply cooling water to the RCFCs and fans that supply airflow will temporarily lose power. Cooling water flow will stop and boiling may occur in the RCFCs tubes, causing steam bubbles to form in the tubes within the RCFCs and expand to the attached SX piping, creating steam voids. As the SX pumps restart and the water column accelerates, the accumulated steam in the fan coolers tubes and piping will condense which could result in a water hammer when the void collapses. Hydrodynamic loads that are introduced by this water hammer event could challenge the integrity and function of the RCFCs and associated cooling water system components, as well as pose a potential challenge to containment integrity.

In response to NRC staff questions (Reference 2), the licensee stated that a Braidwood, Unit 1, specific analysis of this event was performed by modelling the affected systems using RELAP5 up to the point of void collapse and determined the resulting flow and pressures. The time histories for water hammer forces can be calculated using the NRG-approved (Reference 8), the Electric Power Research Institute (EPRI) methodology or the computer program HYTRAN. The HYTRAN computer program is part of the original licensing basis and is described in UFSAR, Appendix D, Section D.8.6 and referenced in UFSAR, Section 3.9.1.2, "Computer Programs Used in Analyses": The EPRI methodology has limitation of applicability based on voids closure velocity. The RELAP4 results for Braidwood show closure velocities above the EPRI methodology applicability criteria, therefore, licensee used HYTRAN program to calculate the time dependent forces from the void collapses. The specific Braidwood, Unit 1, RELAP5 analysis identified three cases of void collapses of interest in the SX piping. The licensee stated that loads produced in the three cases by the evaluated transient were found to be acceptable.

In response to NRC staff questions (Reference 2), the licensee stated that the Braidwood, Unit 2, RCFC piping analyses are based on the Unit 1 RCFC piping analytical models with unit specific differences reconciled during original construction. Therefore, prior to the implementation of this LAR, as part of Regulatory Commitment #1 stated in Attachment 4 of Reference 1, the licensee intends to reconcile the Braidwood, Unit 2, specific RCFC piping differences that are applicable to the GL 96-06 analysis. In SCVB-RAl-13, the NRC staff requested the licensee to explain what is implied by reconciliation of unit specific differences (if possible, provide a reference of NRC approval during initial licensing) and how the analysis for Braidwood, Unit 2, will be performed using the analysis for Unit 1. In response to NRC staff questions (Reference 9), the licensee stated:

"During original design and construction, the Braidwood Unit 2 analyses documented the review of the as-built piping configurations (piping geometry and support configuration) with respect to the Braidwood Unit 1 analyzed configuration. Variances were evaluated and reconciled. The analyses document that all piping stresses for the reconciled Braidwood Unit 2 geometry are within code allowable. The results of the Braidwood Unit 2 piping subsystem analyses confirmed compliance with applicable Codes and were reviewed as part of the original Unit 2 licensing. Use of the piping analytical models of the lead Unit (i.e., Unit 1) to the design of the replicate unit (i.e., Unit 2) is common industry practice.

The GL 96-06 RELAP model includes the entire flow path from the Ultimate Heat Sink through the Reactor Containment Fan Coolers (RCFCs) and back to the UHS. A simplified model of the flow paths parallel to the RCFCs is included in the analysis. The parallel flow paths are not modeled in detail as void formation occurs outside these parallel flow paths. Based on similarities between Unit 1 and Unit 2 Essential Service Water (SX) piping configurations (as documented in the reviews of the as-built piping subsystems), the results of the RELAP model using the Braidwood Unit 1 configuration are considered representative for both Braidwood units.

The forces on the piping that were calculated with HYTRAN were reviewed against the structural capacity of the pipe supports. The support loads and configuration were obtained from the stress analyses for the affected Braidwood Unit 1 piping subsystems.

The regulatory commitment is to validate that the structural capacity of the Unit 2 pipe support configuration is equivalent to the capacity of the Unit 1 pipe support configuration. This commitment will support the conclusion that the Unit 2 pipe supports can accommodate forces developed from void collapses."

Based on the licensee's evaluation discussed above, the NRC staff determines that the GL 96-06 issue for possible water hammer in the RCFC tubes and SX piping following a LOCA or a MSLB accident has been adequately addressed using NRC approved methodologies methods. By letter dated June 16, 2016, the licensee informed the NRC staff that it had completed the actions described by the commitment.

In summary, the licensee used NRG-approved methodology for LOCA M&E release (after correction of errors reported in NSALs-06-6, 5, and 2) and containment response analyses for determining the impact of the proposed increase in the UHS temperature. For resolving the IG-14-1 issue, the licensee performed sensitivity analysis instead of revising the AOR and demonstrated acceptable results.

The NRC staff concludes that the proposed change meets the requirements of 10 CFR Part 50 Appendix A: (1) GDC 16, because the licensee showed that the containment design conditions important to safety are not exceeded during a postulated OBA; (2) GDC 38, because the licensee showed that the containment heat removal system would reduce the containment pressure and temperature rapidly, consistent with the functioning of the UHS at the proposed temperature, and other associated systems, following design-basis accident and would maintain them at acceptable levels; (3) 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 design basis accident, and; (4)

The licensee satisfactorily addressed the GL-96-06 issue for the proposed increase in the UHS temperature. Therefore, the NRC staff finds the proposed increase in the UHS TS temperature to s102 °F acceptable.

3. 7 NRC Staff Confirmatory Analysis The NRC staff performed a confirmatory analysis based on a different model and methodology than the licensee's plug flow model. The analysis used design inputs from the licensee's data, including pond dimensions, meteorological data, pump flow rates, and heat discharged to the UHS from the plant during a OBA. The confirmatory analysis model consisted of multiple segments acting as well mixed tanks as opposed to the licensees plug flow model whose discharge plugs did not mix with cooler water upon discharge from the plant. While the licensee's plug flow enhances heat transfer between the water and the atmosphere, it tends to increase the peak temperature by not accounting for mixing of hot discharged water with cooler ambient pond water. The confirmatory model also accounted for different depths and surface areas for each segment. The NRC staff's confirmatory analysis provided base case results for the various number of SX pumps in operation. The confirmatory analysis also included sensitivity analyses by varying assumptions and/or input data to measure the effects of uncertainty in the parameters of the model. All base case model runs started with the UHS at 102 °F and pond efficiency of 80 percent. The peak return UHS temperature achieved was 104.03 °F. Peak return temperature occurred approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of elapsed time after the initiating event, OBA. This differs from the licensee's calculated peak temperature of 105.2°F approximately 39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> after the initiating event. The NRC staff's computed arrival time of approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> demonstrate that the peak temperature is a result of heat transfer into the pond from the environment and not the result of heat energy from the plant.

The NRC staff confirmatory analysis shows that the licensee's model conservatively predicts a higher return temperature than the NRC staff's confirmatory model.

The NRC staff determined that performing sensitivity evaluations was necessary to account for some uncertainty in the magnitude of mixing in the near field at the SX discharge. The pond flow in the near field at the discharge of SX to the UHS is affected by the vertical discharge of the SX effluent and the concrete pad about 10 feet below the water surface. The vertical up then down movement of discharge and the nearness to the pond's bank would have less effect on local mixing than a horizontal and underwater discharge in a large expanse of open water.

There would still be considerable local mixing, but less mixing as the flow travels toward the plant intake. The discharge would be deflected mostly parallel toward the shore and the intake.

The effect on stratification would be difficult to estimate, but the confirmatory analysis included sensitivity analyses to estimate the presence or lack of stratification and also for a larger first segment at 20 percent of total pond volume. More mixing in the near field would tend to reduce atmospheric heat transfer somewhat, but also diminish sharp rises in temperature at the pond intake. Therefore, the overall effect of more or less mixing would tend to cancel and mitigate any impact on the temperature of water entering the SX system from the UHS.

Pond efficiency is also an input with some uncertainty as it determines the effective total volume and surface area of the plug flow and mixed tank models. After discharge the downstream transport through the UHS pond would be less affected by the turbulence and mixing at the SX

discharge. The NRC staff used a finite difference model derived from the Poisson equation for steady state non-rotating flow in two dimensions to determine a pond efficiency of approximately 90 percent. The NRC staff used a conservatively small value of 80 percent in all pond performance calculations. The licensee had used an efficiency of 82.3 percent in their model.

Thus, the NRC staff concludes that the thermal efficiencies values in the staff's and the licensee's one dimensional models are conservative.

The NRC staff found the meteorological input to be accurate with acceptable uncertainty. Most meteorological data after 1990, except cloud cover and cloud ceiling (used for solar radiation estimates), used in the licensee's and NRC staff's confirmatory analysis were from Braidwood site. Peoria data was used for periods of missing data. The NRC staff compared 2010 Peoria and Braidwood data and found the differences were not significant. Dry bulb temperature was slightly lower at Braidwood, which is a slightly non-conservative difference, but dew point at Braidwood was slightly higher, which is a conservative difference. Overall the NRC staff concludes that the slight differences would not lead to higher peak temperature had Peoria data from 1990 to 2012 been used exclusively for the analyses. Confirmatory analysis model runs used the meteorological data from the hottest months (May through September) from 1948 through 2012, tabulating the peak return temperature for each run and the time to peak temperature. This was a different approach than the licensee's, which determined worst-weather periods first and used that data for their model calculations. Both the licensee's and staff's confirmatory analyses showed that the worst-cases of peak temperature all occurred between 2009 and 2012.

The NRC staff performed several sensitivity calculations postulating stratification existed in the pond and concluded stratification would not result in a return water temperature greater than that calculated by the non-stratified assumption. The analysis assumed that there was a stratified water layer floating on the top of a cooler stagnant layer. Travel time from the discharge to the intake would be half as long, and water entering the intake structure and sumps would mix with the cooler underlying water in equal proportions. More sensitivity runs were made for a 25 percent reduction in evaporative cooling, for wind speed reduction, for higher temperature at plant intake, and combinations of the aforementioned parameter variations. In addition to the above sensitivity runs, which were all performed at 102 °F, the NRC staff also ran sensitivity runs for starting temperature greater than 102 °F and 75 percent heat transfer efficiency and found that peak return temperature exceeds 105.2 °F.

Base cases and sensitivity results follow:

Results of Base Case and Sensitivity Runs - Deviations From Base-Case as Stated Description Peak T, °F Elapsed Time, Hr Base case, Ryan wind function, 30 equal segments, 104.03 8.2 wind speed reduction 0.8, 100 percent Ryan evaporative thermal efficiency (Ryan Efficiency),

2 SX pumps 3 SX pumps 104.03 8.2 4 SX pumps 104.03 8.2 Stratified layer - 3 SX pumps and half depth of pond* 103.82 7.2 First segment with 20 percent of total pond volume and 104.03 8.2 surface area Wind speed reduction factor of 0.617 104.22 8.2 75 percent Ryan efficiency 104.74 8.2 75 percent Ryan efficiency, 3 SX pumps 104.74 8.2 Wind speed reduction factor of 0.617 and 75 percent 104.90 8.2 Ryan efficiencyt

• Average of stratified and underlying water t not a probable combination with two conservative sensitivity factors, but stated for information 3.8 Summary The licensee performed transient analysis of the UHS for a OBA with an initial temperature of 102 °F and determined the peak UHS return temperature (SX inlet) to be 105.2 °F at 39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> after the event. The licensee analyzed the equipment cooled by SX at a minimum of 105.2 °F and found the equipment would operate satisfactorily to mitigate the OBA and perform shutdown and cooldown of the non-accident unit. The licensee also determined that the change in UHS temperature would not have any adverse effect on other accident analysis.

The NRC staff reviewed the licensee's assumptions, design input, methodology and results.

The NRC staff verified the applicability of the meteorological data input and concluded that it was satisfactory, but questioned some of the licensee's input data, including pond efficiency, stratification, mixing at the SX discharge and the plug flow model applicability. To provide reasonable assurance of the licensee's results, the NRC staff performed a confirmatory analysis using a different methodology with sensitivity analyses for uncertain values of input data. The confirmatory analysis calculated a peak UHS temperature slightly less than the licensee's value occurring approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the event. The sensitivity analyses showed that reasonable variations in the associated design inputs did not have adverse effects on the results. The NRC staff analyses demonstrate that the licensee's results are conservative, and therefore finds that GOCs 5, 38, and 44, are met. Therefore, the staff finds the licensee's request to change the maximum allowed UHS temperature in SR 3.7.9.2 to 102 °F is acceptable.

4.0 TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT Current TS SR 3.7.9.2 Verify average water temperature of the UHS is :::; 100°F Revised TS SR 3.7.9.2 Verify average water temperature of the UHS is :::; 102°F Based on the technical review discussed in Section 3, above, the NRC staff finds that 10 CFR 50.36(c)(2) and 10 CFR 50.36(c)(3) are met because the UHS will be operable with the UHS temperature :::; 102 °F. The NRC staff, therefore, concludes that the revised TS SR is acceptable.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Illinois State official was notified on May 11, 2016, of the proposed issuance of the amendment. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The NRC staff published an Environmental Assessment (EA) in the Federal Register on July 26, 2016 (81 FR 48851) related to this proposed action. In that EA, the NRC staff concluded that the proposed action would not have a significant effect on the quality of the human environment.

Accordingly, the NRC staff determined that an environmental impact statement was not warranted for the proposed action.

7.0 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.

8.0 REFERENCES

1. Letter from EGC to NRC dated August 19, 2014, "Request for a license Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, "Ultimate Heat Sink" (ADAMS Accession No. ML14231A902).

2. Letter from EGC to NRC dated April 30, 2015, "Response to Request for Additional Information Regarding Request for a Licensee Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, "Ultimate Heat Sink" (ADAMS Accession No. ML15120A396).

3. WCAP-10325-P-A, "Westinghouse LOCA Mass and Energy Release Model for Containment Design March 1979 Version", May 1983.

4. WCAP-8327, July 1974, "Containment Pressure Analysis Code (COCO)," WCAP-8327, July 1974 (Proprietary) (ADAMS Accession No. ML092460709).

5. NSAL-06-6, "LOCA Mass and Energy Release Analysis," June 6, 2006.

6. NSAL-11-5, 'Westinghouse LOCA Mass and Energy Release Calculation Issues," July 26, 2011, (ADAMS Accession No. ML13239A479).

7. NSAL-14-2, "Westinghouse Loss-of-Coolant Accident Mass and Energy Release Calculation Issue for Steam Generator Tube Material Properties" March 31, 2014.

8. Letter from NRC to EPRI Waterhammer Project Utility Advisory Group dated April 3, 2002, "NRC Acceptance of EPRI Report TR-113594, "Resolution of Generic Letter 96-06 Waterhammer Issues," Volumes 1 and 2 (ADAMS Accession No. ML021190358).

9. Letter from EGC to NRC dated October 9, 2015, "Response to Request for Additional Information Regarding Request for a License Amendment to Braidwood Station, Units I and 2, Technical Specification 3.7.9, "Ultimate Heat Sink," (ADAMS Accession No. ML15282A345).

10. Letter from EGC to NRC dated October 30, 2015, "Response to Request for Additional Information Regarding Request for a License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9,"Ultimate Heat Sink"," (ADAMS Accession No. ML15303A326).

11. Letter from EGC to NRC dated November 9, 2015, "Response to Request for Additional Information Regarding Request for a License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, "Ultimate Heat Sink"," (ADAMS Accession No. ML15313A301).

12. Letter from EGC to NRC dated April 29, 2016, "Response to Request for Additional Information Regarding Request for a License Amendment to Braidwood Station, Units 1 and 2, Technical Specification 3.7.9, "Ultimate Heat Sink"," (ADAMS Accession No. ML16123A014).

Principal Contributors: D. Warner G. Purciarello A. Sallman J. Voveris Date of issuance: July 2 6, 2o1 6

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

Sincerely,

!RAJ Joel S. Wiebe, Senior Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-456 and STN 50-457

Enclosures:

1. Amendment No. 189 to NPF-72

2. Amendment No. 189 to NPF-77

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