Text

Issuance of Amendments Related to the Nuclear Service Water System New Condition for Single Pond Return Header Configuration
	ML18275A278
Person / Time
Site:	Catawba
Issue date:	11/28/2018
From:	Michael Mahoney; Plant Licensing Branch II
To:	Simril T; Duke Energy Carolinas
	Mahoney M, 415-3867
References
	CAC MG0245, CAC MG0246, EPID L-2017-LLA-0297
	Download: ML18275A278 (47)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 28, 2018 Mr. Tom Simril Site Vice President Catawba Nuclear Station, Units 1 and 2 Duke Energy Carolinas, LLC 4800 Concord Road York, SC 29745

SUBJECT:

CATAWBA NUCLEAR STATION, UNITS 1 AND 2- ISSUANCE OF AMENDMENTS RELATED TO THE NUCLEAR SERVICE WATER SYSTEM NEW CONDITION FOR SINGLE POND RETURN HEADER CONFIGURATION (CAC NOS. MG0245 AND MG0246; EPID L-2017-LLA-0297)

Dear Mr. Simril:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 300 to Renewed Facility Operating License No. NPF-35 and Amendment No. 296 to Renewed Facility Operating License No. NPF-52 for the Catawba Nuclear Station, Units 1 and 2, respectively. The amendments are in response to your application dated September 14, 2017, as supplemented by letters dated May 8, August 17, September 20, October 29, and November 15, 2018.

The amendments modify Technical Specification 3.7.8, "Nuclear Service Water System (NSWS)," to add a new condition (Condition D) to allow Single Pond Return Header Operation of the NSWS with a 30-Day Completion Time.

T. Simril A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Michael Mahoney, Pro ect Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-413 and 50-414

Enclosures:

1. Amendment No. 300 to NPF-35

2. Amendment No. 296 to NPF-52

3. Safety Evaluation cc: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC NORTH CAROLINA ELECTRIC MEMBERSHIP CORPORATION DOCKET NO. 50-413 CATAWBA NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 300 Renewed License No. NPF-35

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

A. The application for amendment to the Catawba Nuclear Station, Unit 1 (the facility), Renewed Facility Operating License No. NPF-35, filed by Duke Energy Carolinas, LLC (licensee), dated September 14, 2017, as supplemented by letters dated May 8, August 17, September 20, October 29, and November 15, 2018, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) and 2.C.(7) of Renewed Facility Operating License No. NPF-35 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 300 which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications.

(7) Additional Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No. 300 are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Additional Conditions.

3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed License No. NPF-35 and Technical Specifications Date of Issuance: November 28, 2018

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC NORTH CAROLINA MUNICIPAL POWER AGENCY NO. 1 PIEDMONT MUNICIPAL POWER AGENCY DOCKET NO. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 296 Renewed License No. NPF-52

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

A. The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility), Renewed Facility Operating License No. NPF-52, filed by the Duke Energy Carolinas, LLC (the licensee), dated September 14, 2017, as supplemented by letters dated May 8, August 17, September 20, October 29, and November 15, 2018, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 2

2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) and 2.C.(7) of Renewed Facility Operating License No. NPF-35 is hereby amended to read as follows (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 296, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications.

(7) Additional Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No. 296 are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Additional Conditions.

FOR THE NUCLEAR REGULATORY COMMISSION

~-;:fa~

Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed License No. NPF-52 and Technical Specifications Date of !ssliance: Novernber 2 8 , 2 O1 8

ATTACHMENT CATAWBA NUCLEAR STATION, UNITS 1 AND 2 LICENSE AMENDMENT NO. 300 RENEWED FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO. 50-413 AND LICENSE AMENDMENT NO. 296 RENEWED FACILITY OPERATING LICENSE NO. NPF-52 DOCKET NO. 50-414 Replace the following pages of the Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert NPF-35, page 4 NPF-35, page 4 NPF-35, page 5 NPF-35, page 5 NPF-52, page 4 NPF-52, page 4 NPF-52, page 5 NPF-52, page 5 Replace the following page of the Appendix A, Technical Specifications (TS), with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert TS 3.7.8-4 TS 3.7.8-4 TS 3.7.8-5 Replace the following page of the Appendix B, Additional Conditions, with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert NPF-35, page 5 NPF-42, page 4

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 300 which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications.

(3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than December 6, 2024, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e)(4),

following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

(4) Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license.

(5) Fire Protection Program Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program that complies with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated September 25, 2013; as supplemented by letters dated January 13, 2015; January 28, 2015; February 27, 2015; March 30, 2015; April 28, 2015; July 15, 2015; August 14, 2015; September 3, 2015; December 11, 2015; January 7, 2016; March 23, 2016; June 15, 2016; August 2, 2016; September 7, 2016; and, January 26, 2017, as approved in the SE dated February 8, 2017. Except where NRG approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRG approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c),

the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

Renewed License No. NPF-35 Amendment No. 300

(6) Mitigation Strategies Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a) Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equipment and materials

4. Command and control

5. Training of response personnel (b) Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets

2. Communications

3. Minimizing fire spread

4. Procedures for implementing integrated fire response strategy

5. Identification of readily-available pre-staged equipment

6. Training on integrated fire response strategy

7. Spent fuel pool mitigation measures (c) Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to onsite responders (7) Additional Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No. 300 are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Additional Conditions.

D. The facility requires exemptions from certain requirements of Appendix J to 10 CFR Part 50, as delineated below and pursuant to evaluations contained in the referenced SER and SSERs. These include, (a) partial exemption from the requirement of paragraph III.D.2(b)(ii) of Appendix J, the testing of containment airlocks at times when the containment integrity is not required (Section 6.2.6 of the SER, and SSE Rs

3 and #4), (b) exemption from the requirement of paragraph 111.A.(d) of Appendix J, insofar as it requires the venting and draining of lines for type A tests (Section 6.2.6 of SSER #3), and (c) partial exemption from the requirements of paragraph 111.B of Appendix J, as it relates to bellows testing (Section 6.2.6 of the SER and SSER #3).

These exemptions are authorized by law, will not present an undue risk to the public health and safety, are consistent with the common defense and security, and are consistent with certain special circumstances as discussed in the referenced SER and SSE Rs. These exemptions are, therefore, hereby granted pursuant to 10 CFR 50.12. With the granting of these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.

Renewed License No. NPF-35 Amendment No. 300

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 296 which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications.

(3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than December 6, 2024, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e)(4),

following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

(4) Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license.

(5) Fire Protection Program Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program that complies with 10 CFR 50.48(a) and 10 CFR 50.48(c}, as specified in the licensee amendment request dated September 25, 2013; as supplemented by letters dated January 13, 2015; January 28, 2015; February 27, 2015; March 30, 2015; April 28, 2015; July 15, 2015; August 14, 2015; September 3, 2015; December 11, 2015; January 7, 2016; March 23, 2016; June 15, 2016; August 2, 2016; September 7, 2016; and, January 26, 2017, as approved in the SE dated February 8, 2017. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c),

the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

Renewed License No. NPF-52 Amendment No. 296

(6) Mitigation Strategies Develop and maintain strategies for addressing large fires and explosions And that include the following key areas:

(a) Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equipment and materials

4. Command and control

5. Training of response personnel (b) Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets

2. Communications

3. Minimizing fire spread

4. Procedures for implementing integrated fire response strategy

5. Identification of readily-available pre-staged equipment

6. Training on integrated fire response strategy

7. Spent fuel pool mitigation measures (c) Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to onsite responders (7) Additional Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No. 296 are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Additional Conditions.

D. The facility requires exemptions from certain requirements of Appendix J to 10 CFR Part 50, as delineated below and pursuant to evaluations contained in the referenced SER and SSER. These include, (a) partial exemption from the requirement of paragraph III.D.2(b)(ii) of Appendix J, the testing of containment airlocks at times when the containment integrity is not required (Section 6.2.6 of the SER, and SSERs #5), (b) exemption from the requirement of paragraph 111.A.(d) of Appendix J, insofar as it requires the venting and draining of lines for type A tests (Section 6.2.6 of SSER #5), and (c) partial exemption from the requirements of paragraph 111.B of Appendix J, as it relates to bellows testing (Section 6.2.6 of the SER and SSER #5). These exemptions are authorized by law, will not present an undue risk to the public health and safety, are consistent with the common defense and security, and are consistent with certain special circumstances, as discussed in the referenced SER and SSER. These exemptions are, therefore.

hereby granted pursuant to 10 CFR 50.12. With the granting of these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended. the provisions of the Act, and the rules and regulations of the Commission.

Renewed License No. NPF-52 Amendment No. 296 Corrected by letter dated September 28, 2011

NSWS 3.7.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. -----------NOTES----------- 0.1 Restore NSWS Pond 30 days

1. Entry into this return header to Condition shall only OPERABLE status.

be allowed for pre-planned activities.

2. Immediately enter Condition A of this LCO if one or more NSWS components become inoperable while in this Condition and one NSWS train remains OPERABLE.

3. Immediately enter LCO 3.0.3 if one or more NSWS components become inoperable while in this Condition and no NSWS train remains OPERABLE.

4. Entry into this Condition shall only be allowed for 60 days per 12-month period.

One NSWS Pond return header inoperable due to NSWS being aligned for single Pond return header operation.

(continued)

Catawba Units 1 and 2 3.7.8-4 Amendment Nos. 300/296

NSWS 3.7.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A, B, AND C, or D not met.

E.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3. 7. 8. 1 --------------------------------------NO TE----------------------------

1solation of NSWS flow to individual components does not render the NSWS inoperable.

Verify each NSWS manual, power operated, and In accordance with automatic valve in the flow path servicing safety related the Surveillance equipment, that is not locked, sealed, or otherwise Frequency Control secured in position, is in the correct position. Program SR 3. 7. 8. 2 ----------------------------------NO TE---------------------------------

Not required to be met for valves that are maintained in position to support NSWS single supply header operation, single Auxiliary Building discharge header operation, or single Pond return header operation.

In accordance with Verify each NSWS automatic valve in the flow path that the Surveillance is not locked, sealed, or otherwise secured in position, Frequency Control actuates to the correct position on an actual or simulated Program actuation signal.

SR 3.7.8.3 Verify each NSWS pump starts automatically on an In accordance with actual or simulated actuation signal. the Surveillance Frequency Control Program Catawba Units 1 and 2 3.7.8-5 Amendment Nos. 300/296

Amendment ImQlementation Number Additional Condition Date Prior to entry into TS 300 1. To reduce NSWS pipe stress at the 1A 3.7.8, Nuclear Service Component Cooling (KC) Heat Exchanger Water System (NSWS) piping return nozzle location, a 1/4" thick Condition D reinforcing pad will be added to the existing - One NSWS Pond reinforcing pad. return header inoperable due to

2. The plant engineering process will be utilized NSWS being aligned to develop new plant procedures and for single Pond required training to support the Single Pond return header Return Header alignment and new operator operation.

actions credited in the PRA.

Human Error Probabilities (HEPs) for the two new operator actions developed in support of Single Pond Return Header alignment LAR will be updated as needed to be consistent with the updated procedural guidance and training. Risk estimates will also be updated to include the updated HEPs.

After the HEPs are updated, it will be confirmed that the risk estimates associated with Single Pond Return Header alignment LAR are within the acceptance guidelines of RG 1.177 and RG 1.174. If the risk estimates are not within the acceptance guidelines of RG 1.177 and RG 1.174, additional risk reduction measures will be taken as needed to ensure that the acceptance guidance are met.

Renewed License No. NPF-35 Amendment No. 300 Amendment Implementation Number Additional Condition Date Prior to entry into TS 3.7.8, 296 The plant engineering process will be utilized to Nuclear Service Water develop new plant procedures and required System (NSWS) Condition D training to support the Single Pond Return - One NSWS Pond return Header alignment and new operator actions header inoperable due to credited in the PRA. NSWS being aligned for single Pond return header Human Error Probabilities (HEPs) for the two operation.

new operator actions developed in support of Single Pond Return Header alignment LAR will be updated as needed to be consistent with the updated procedural guidance and training. Risk estimates will also be updated to include the updated HEPs.

After the HEPs are updated, it will be confirmed that the risk estimates associated with this Single Pond Return Header alignment LAR are within the acceptance guidelines of RG 1.177 and RG 1.174. If the risk estimates are not within the acceptance guidelines of RG 1.177 and RG 1.174, additional risk reduction measures will be taken as needed to ensure that the acceptance guidance are met.

Renewed License No. NPF-52 Amendment No. 296 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 300 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-35 AND AMENDMENT NO. 296 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-52 DUKE ENERGY CAROLINAS. LLC CATAWBA NUCLEAR STATION. UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414

1.0 INTRODUCTION

By letter dated September 14, 2017 as supplemented by letters dated May 8, August 17, September 20, October 29, and November 15, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML172618255, ML18129A053, ML18232A245, ML18269A108, ML18303A084, and ML18319A249, respectively), Duke Energy Carolinas, LLC (the licensee), submitted a license amendment request (LAR) for the Catawba Nuclear Station, Units 1 and 2 (Catawba) to revise Technical Specification (TS) 3.7.8, "Nuclear Service Water System (NSWS)."

The supplements dated May 8, August 17, September 20, October 29, and November 15, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register (FR) on March 13, 2018 (83 FR 10914).

Specifically, the proposed amendment is a request to revise TS 3.7.8 to add a new condition (Condition D) to allow Single Pond Return Header Operation of the NSWS with a 30-Day Completion Time (CT). Single Pond Return Header Operation to the NSWS involves isolating one train of the NSWS Pond Return piping at the Auxiliary Building wall and maintaining the discharge crossover lines open between trains in the Auxiliary Building and Emergency Diesel Generator (EDG) Buildings. This provides a common safety related discharge path through the single remaining in-service Pond Return line. This alignment, Single Pond Return Header Operation, allows a Pond Return Header to be removed from service while a flow path is maintained through both trains of NSWS supplied equipment to the Standby Nuclear Service Enclosure 3

Water Pond (SNSWP). The Single Pond Return Header Operation is necessary to allow internal inspections and modifications of the NSWS Pond Return buried piping between the Auxiliary Building and the discharge to the SNSWP.

2.0 REGULATORY EVALUATION

2.1 System Descriptions and Requirements In Section 2.1, "System Design and Operation," of its letter dated September 14, 2017, the licensee provided the following description of Catawba's NSWS:

The NSWS, including Lake Wylie and the Standby Nuclear Service Water Pond (SNSWP), is the ultimate heat sink for various QA [quality assurance] Condition 1 heat loads during normal operation, design basis events (Condition II, Ill, and IV Events per [American National Standards Institute] ANSI N18.2-1973), and other design events as dictated by Catawba licensing criteria.

During normal operation, the NSWS supplies cooling water to various safety related components. While in normal operation, the maximum heat load and flow requirements on the NSWS are encountered with the Unit in Mode 5 due to decay heat removal requirements. During ANSI N18.2-1973 initiated events, the NSWS is required to support Emergency Core Cooling System (ECCS) operation by providing cooling water to various safety related components along with emergency makeup to selected QA Condition 1 Systems. The ANSI N18.2-1973 event that imposes the most stringent design basis requirement is the Condition IV Initiator of Loss of Coolant Accident (LOCA). In accordance with 10 CFR 50 Appendix A, GDC [General Design Criterion] 2, Natural Phenomena, Catawba must withstand the effects of a Safe Shutdown Earthquake (SSE) without affecting the ability of the safety systems to shut down the plant. As such, the ANSI N18.2-1973 design basis events are considered after the occurrence of an earthquake. This means that a Loss of Lake Wylie and a dual unit Loss of Offsite Power (LOOP) are assumed.

The NSWS consists of two loops (A.& B) of essential equipment and each loop is shared between Units 1 and 2. Each loop supplies two trains of NSWS, one train for Unit 1 and one train for Unit 2. One train of NSWS is sufficient to perform the safety functions of NSWS for the applicable unit, thus each unit has redundant trains of NSWS. Two bodies of water serve as the ultimate heat sink (UHS) supply to the NSWS. Lake Wylie is the normal source of NSWS and the SNSWP is the emergency source.

The licensee's letter dated July 20, 2017 discussed the automatic isolation features of the NSWS. The following information is taken from this letter. During normal operation, the NSWS loops are cross connected and discharge to Lake Wylie in a single discharge header (through 1RN843B and 1RN57 A) from NSWS cooled components other than the EDGs. Each EOG discharge is normally aligned to Lake Wylie. The NSWS discharge from the EDGs of each unit contain a manual valve that allow combining the discharge from both EDGs of the same unit to Lake Wylie or the SNSWP.

Upon loss of Lake Wylie, as detected by low level in the NSWS pump house suction pits, each EOG discharge line will shift discharge to the SNSWP, the loops A and B combined discharge header valves to Lake Wylie ( 1RN843B and 1RN57 A) will close, the loops A and B discharge

header cross-connect valves (1 RN53B and 1RN54A) will close, the loops A and B discharge valves ( 1RN58B and 1RN63A) to the SNSWP will open to allow discharge to the SNSWP and the NSWS supply header crossover valves will open.

A containment high-high pressure signal (SP) from either unit will open the loops A and B discharge valves (1RN58B and 1RN63A) to the SNSWP and close the NSWS supply header crossover valves for that unit. Thus, during a loss-of-coolant-accident (LOCA) design basis accident (DBA) with a loss of Lake Wylie and LOOP, each unit has two independent and redundant supply and return trains of NSWS. Each train is capable of performing the NSWS safety functions for that Unit, which satisfies the redundancy aspect of Criterion 44 "Cooling water," Appendix A of Title 10 of the Code of Federal Regulations ( 10 CFR) Part 50 (stating "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.").

Catawba TS Limiting Condition for Operation (LCO) 3.7.8 requires two NSWS trains to be OPERABLE. The TS Bases, which are a summary statement of the reasons for such specifications, state that two NSWS trains are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post-accident heat loads, assuming that the worst case single active failure occurs coincident with LOOP. Per TS 3.7.8 Required Action A.1, if one NSWS train is inoperable during operational MODE 1 (Power Operation), 2 (Startup), 3 (Hot Standby), or 4 (Hot Shutdown with all reactor vessel head bolts fully tensioned), then the NSWS train must be restored to operable within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (i.e., 72-hour CT}, and certain other additional actions must be taken. Per the TS 3.7.8 Bases, the 72-hour CT is based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this time period. If the system is not restored, then per TS

3. 7 .8 Condition D, Actions D.1 and D.2, the licensee must place the reactor in MODE 3 in six hours, and MODE 5 (Cold Shutdown with all reactor vessel head closure bolts fully tensioned) in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Updated Final Safety Analysis Report (UFSAR) Section 9.2.1.3 states that the NSWS is designed to supply the cooling water requirements of a simultaneous LOCA on one unit and cooldown to the other unit assuming a single failure anywhere in the system, LOOP, and loss of Lake Wylie.

2.2 Licensee's Proposed Changes The licensee proposed to revise TS 3.7.8 as follows:

Current Condition D, states:

Required Action and associated Completion Time of Condition A, B, or C not met.

The licensee proposed to rename Condition D as Condition E stating:

Required Action and associated Completion Time of Condition A, B, C, or D not met.

The licensee proposed to add a new Condition D stating:

CONDITION REQUIRED ACTION COMPLETION TIME D. ----------N()TES------- D. 1 Restore NSWS Pond 30 days

1. Entry into this return header to Condition shall only OPERABLE status.

be allowed for pre-planned activities.

2. Immediately enter Condition A of this LCO if one of more NSWS components becoming inoperable while in this Condition and one NSWS train remains OPERABLE.

3. Immediately enter LC() 3.0.3 if one or more NSWS components become inoperable while in this Condition and no NSWS train remain OPERABLE.

4. Entry into this Condition shall only be allowed for 60 days per 12-month period.

One NSWS Pond return header inoperable due to NSWS being aligned for single Pond return header operation.

Current Note for supporting requirement (SR) 3.7.8.2, states:

Not required to be met for valves that are maintained in position to support NSWS single supply or discharge header operation.

The Note for SR 3.7.8.2 will be revised to state:

Not required to be met for valves that are maintained in position to support NSWS single supply, single Auxiliary Building, discharge header operation, or single Pond return header operation.

2.3 Applicable Regulatory Requirements The NRC staff considered the following regulatory requirements, licensing and design basis information, and guidance during its review of the proposed changes.

Paragraph 10 CFR 50.36(c)(3), "Surveillance requirements," states:

Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

Paragraph 10 CFR 50.40 states that in determining that an operating license will be issued, the Commission will be guided by, among other things, this consideration:

(a) ... the processes to be performed, the operating procedures, the facility and equipment, the use of the facility, and other technical specifications, or the proposals, in regard to any of the foregoing collectively provide reasonable assurance that the applicant will comply with the regulations in this chapter, including the regulations in part 20 of this chapter, and that the health and safety of the public will not be endangered.

Paragraph 10 CFR 50.57(a) that an operating license may be issued upon finding, among other things, that:

(3) There is reasonable assurance (i) that the activities authorized by the operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the regulations in this chapter.

UFSAR Section 3.1 "Conformance with General Design Criteria," states, in part:

CRITERION 44- COOLING WATER-A system to transfer heat from structures, systems, and components important to safety, to an ultimate heat sink shall be provided. The system safety function shall be to transfer the combined heat load of these structures, systems, and components 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.

UFSAR Section 9.2.1.1, "Design Bases," states, in part:

The Nuclear Service Water System (RN) provides essential auxiliary support functions to Engineered Safety Features of the station. The system is designed to supply cooling water to various heat loads in both the safety and non-safety portions of each unit. Provisions are made to ensure a continuous flow of cooling water to those systems and components necessary for plant safety during normal operation and under accident conditions. Sufficient redundancy of piping and components is provided to ensure that cooling is maintained to essential loads at all times.

The Nuclear Service Water System is designed to withstand a safe shutdown earthquake and to prevent any single failure from limiting the ability for the engineered safety features to perform their safety functions. Sufficient pump capacity is included to provide the cooling water to shutdown each unit and the valves are arranged in such a way that loss of one train does not jeopardize the entire system.

The following are applicable regulatory guidance and policy documents for evaluating the risk impact of the proposed TS change:

Regulatory Guide (RG) 1.174, Revision 2 (May, 2011, ADAMS Accession No. ML100910006), "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis." describes a risk-informed approach, acceptable to the NRC, for assessing the nature and impact of proposed licensing basis changes by considering engineering issues and applying risk insights.

Regulatory Guide 1.200, Revision 2 (ADAMS Accession No. ML090410014), "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities." describes an acceptable approach for determining whether the quality of the probabilistic risk assessment (PRA) models, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA models can be used in regulatory decision making for light-water reactors.

Regulatory Issue Summary (RIS) 2007-06 (March 2007), "Regulatory Guide 1.200 Implementation." describes how the NRC will implement its technical adequacy review of plant-specific PRAs used to support risk-informed licensing actions after the issuance of RG 1.200.

Regulatory Guide 1.177, Revision 1 (May, 2011, ADAMS Accession No. ML100910008),

"An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications."," describes a risk-informed approach, acceptable to the NRC, for assessing proposed permanent TS changes in CTs. In addition, this RG provides risk acceptance guidelines for evaluating the results of such assessments. In implementing

risk-informed decisionmaking, TS changes are expected to meet the following five key principles outlined in RG 1.177:

1. The proposed change meets the current regulations unless it is explicitly related to a requested exemption.

2. The proposed change is consistent with the defense-in-depth philosophy.

3. The proposed change maintains sufficient safety margins.

4. When the proposed change result in an increase in core damage frequency (CDF), or risk, the increases should be small and consistent with the intent of the Commission's Safety Goal Policy Statement.

5. The impact of the proposed change should be monitored using performance measurement strategies.

2.4 License Condition The licensee proposed license conditions in its October 29, and November 15, 2018, letters.

The license conditions, which will be added to Appendix B, Additional Conditions, of the Catawba, Units 1 and 2, Facility Operating Licenses, NPF-35 and NPF-52.

Catawba, Unit 1, Facility Operating License, NPF-35 Amendment Additional Condition lmQlementation Date Number Prior to entry into TS 3.7.8, 300 1. To reduce NSWS pipe stress at the 1A Nuclear Service Water Component Cooling (KC) Heat Exchanger System (NSWS) Condition D piping return nozzle location, a 1/4" thick - One NSWS Pond return reinforcing pad will be added to the existing header inoperable due to reinforcing pad. NSWS being aligned for single Pond return header

2. The plant engineering process will be utilized operation.

to develop new plant procedures and required training to support the Single Pond Return Header alignment and new operator actions credited in the PRA.

Human Error Probabilities (HEPs) for the two new operator actions developed in support of Single Pond Return Header alignment LAR will be updated as needed to be consistent with the updated procedural guidance and training. Risk estimates will also be updated to include the updated HEPs.

After the HEPs are updated, it will be confirmed that the risk estimates associated with Single Pond Return Header alignment LAR are within the acceptance guidelines of RG 1.177 and RG 1.174. If the risk estimates are not within the acceptance guidelines of RG 1.177 and RG 1.174,

additional risk reduction measures will be taken as needed to ensure that the acceptance guidance are met.

Catawba, Unit 2, Facility Operating License, NPF-52 Amendment Additional Condition lm12lementation Date Number Prior to entry into TS 3.7.8, 296 The plant engineering process will be utilized to Nuclear Service Water develop new plant procedures and required System (NSWS) Condition D training to support the Single Pond Return - One NSWS Pond return Header alignment and new operator actions header inoperable due to credited in the PRA. NSWS being aligned for single Pond return header Humar.i Error Probabilities (HEPs) for the two new operation.

operator actions developed in support of Single Pond Return Header alignment LAR will be updated as needed to be consistent with the updated procedural guidance and training. Risk estimates will also be updated to include the updated HEPs.

After the HEPs are updated, it will be confirmed that the risk estimates associated with this Single Pond Return Header alignment LAR are within the acceptance guidelines of RG 1.177 and RG 1.174. If the risk estimates are not within the acceptance guidelines of RG 1.177 and RG 1.174, additional risk reduction measures will be taken as needed to ensure that the acceptance guidance are met.

Section 3.1.1 and 3.3.1.1.2 of this safety evaluation provides the NRC staff's technical review of the proposed license conditions.

3.0 TECHNICAL EVALUATION

3.1 NSWS Single Pond Return Header 012eration On page 6 of its letter dated September 14, 2018 letter, the licensee described the proposed NSWS Single Pond Return Header Operation as follows:

The proposed NSWS Single Pond Return Header Operation will involve isolating one train of the NSWS Pond Return piping at the Auxiliary Building wall. .. and each EOG Building .... [The discharge crossover lines between trains in the Auxiliary Building and EOG Buildings will be maintained open.] This [lineup] will provide a common safety related discharge path through the single shared in-service NSWS Pond Return line. This proposed alignment will allow a Pond Return Header to be removed from service while a shared discharge flow path is maintained for all essential NSWS Supplied equipment to the SNSWP. While in this alignment, the NSWS will be capable of supplying required essential equipment (both trains of both units) with the design basis cooling water flows to support accident mitigation on one unit and the cool down loads of the other unit.

Additionally, in this proposed alignment, the NSWS will be pre-aligned to the SNSWP, which is the assured source of cooling water for the NSWS. This removes the risk of an active failure where an automatic valve fails to reposition on a loss of Lake Wylie.

Specifically:

1. One of the two "in series" Auxiliary Building Lake Return Isolation valves

( 1RN57A or 1RN843B) will be open to increase the reliability of swapping discharge to Lake Wylie, if the alternate discharge path is needed.

2. The NSWS Return Header Crossover Valves 1RN53B and 1RN54A are open with power removed, and therefore will not auto-close on low-low NSWS suction pit level or Transfer to the Auxiliary Shutdown Panel.

3. The automatic valves in the safety flow path for alignment to the SNSWP are open with power removed. For work and inspections on the "A" Train Pond Return Header, valves 1RN58B, 2RN848B, and 1RN848B are opened with power removed. Similarly, for work and inspections on the "B" Train Pond Return Header, valves 1RN63A, 2RN846A, and 1RN846A are opened with power removed.

4. The NSWS suction supply is aligned to the SNSWP.

5. The Unit 1 and Unit 2 NSWS Non-Essential headers are isolated. Isolation valves are closed with power removed.

6. Power remains on the closed Motor Operated valves, which isolate the Lake Wylie Return flow paths. This allows a rapid re-establishment of discharge flow if the alternate path is needed. These are valves 1(2)RN849B, 1(2)RN847A, and either 1RN843B or 1RN57 A.

7. The four manually operated NSWS Return Header Crossover Valves 1(2)RNP08 and 1(2)RNP09, in both units Diesel Generator Buildings, will be locked open.

General Comments:

1. The NSWS cannot be aligned in Single Pond Return Header Operation if the NSWS is already in the NSWS Single Auxiliary Building Discharge Header alignment or the NSWS Single Supply Header alignment. These configurations are described in CNS TSs 3.7.8 and the associated TS Bases. The combination of any two of these alignments has not been analyzed.

2. While the NSWS is aligned in Single Pond Return Header Operation, Unit 1 and Unit 2 are in a "TS Action Statement" for the affected NSWS Pond return header (the isolated Pond Return Header).

3. It is intended that NSWS Single Pond Return Header Operation be utilized with both Units in Mode 1. There is no specific mode requirement for use of this alignment.

4. This requested condition will be entered for preplanned maintenance and inspections only. It is anticipated that entry into the condition should not be required more often than once per year, per train.

The NRC staff determined that the system lineup for the Single Pond Return Header Operation would complicate the actions necessary to restore or maintain the NSWS safety functions for the following unlikely events:

spurious motor operated valve (MOV) closure or failure in an Auxiliary Building discharge line or combined EDG discharge to the SNSWP.

NSWS piping breaks or leaks.

The actions to restore or maintain NSWS safety functions are discussed in the Defense-In-Depth evaluation below.

The NRC staff used RG 1.177 in performing a detailed review of the licensee's request and compared the request against applicable regulatory criteria. Section 3.1 of this safety evaluation specifically addresses the Traditional Engineering Considerations discussed in Section 2.2 of RG 1.177. In completing this evaluation, the NRC staff considered the information that was provided by the licensee's in its letter dated September 14, 2017, as supplemented by letters dated May 8, October 29, and November 15, 2018.

3.1.1 Defense-In-Depth Evaluation Section 2.2.1 of RG 1.177 discusses defense-in-depth, and provides a number of elements that can be used as guidelines for assessing defense-in-depth, while noting that other equivalent acceptance guidelines may also be used. Consistency with the defense-in-depth philosophy is maintained under the following circumstances:

A reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation is preserved to the extent that such balance is needed to meet the acceptance criteria of the specific design-basis accidents and transients.

Over-reliance on programmatic activities as compensatory measures associated with the change in the licensing basis is avoided.

System redundancy, independence, and diversity are maintained commensurate with the expected frequency and consequences of challenges to the system.

Defenses against potential common-cause failures (CCFs) are maintained and the potential for introduction of new CCF mechanisms is assessed

Independence of physical barriers is not degraded.

Defenses against human errors are maintained.

The intent of the plant's design criteria is maintained.

The NRC staff's assessment:

A reasonable balance among prevention of core damage, prevention of containment failure and consequence mitigation is preserved.

In the proposed lineup in new Condition D, one NSWS Pond return header is isolated at the Auxiliary Building wall and in both EDG Buildings making the return header inoperable and preventing its discharge to the SNSWP. The discharge crossover lines are maintained open

between trains in the Auxiliary Building through 1RN53B (RN is Service Water System) and 1RN54A and the EOG Buildings through 1/2RN848B and 1/2RN846A. This lineup provides a common safety related discharge path through the single remaining in-service Pond Return Header. This alignment allows the other NSWS Pond Return Header to be removed from service while a flow path is maintained through both trains of NSWS supplied equipment to the SNSWP.

The NSWS A and B Trains for both units, with all components operable other than the isolated NSWS Single Pond Return Header, will still provide cooling to its respective equipment that are important to safety, allowing the equipment to perform their safety functions through use of the common discharge header. In the event of a OBA in one unit, the NSWS, consisting of trains 1A, 1B, 2A, and 2B, has sufficient NSWS flow to mitigate the accident in one unit and shutdown and cooldown the non-accident unit assuming a single failure anywhere other than the common discharge line. This alignment is vulnerable to very unlikely component failures that are described in the discussion below on redundancy, independence and diversity. This lineup will be allowed only for 60 days per calendar year for pre-planned maintenance only. With all safety related NSWS safety related loads fully functional in this lineup, the ability to prevent core damage and prevent containment failure is preserved during the limited allowed maintenance period of 60 days per calendar year.

The proposed LAR does not affect liquid or gaseous radioactive effluents or filtration systems; therefore, radiological mitigation is not affected.

Based on the above described NSWS capability and the absence of any impact on radiological mitigation factors, the NRC staff considers that a reasonable balance among prevention of core damage, prevention of containment failure and consequence mitigation is preserved during the single discharge header alignment during a 30-day CT.

Over-reliance on programmatic activities to compensate for weaknesses in plant design is avoided.

The licensee is not relying on programs to compensate for the single NSWS Pond return header lineup. Programmatic activities to be used in accomplishing the proposed maintenance lineup include additional training on the single NSWS Pond return header alignment in addition to the normal activities for pre-planned maintenance. The NRC staff does not consider this to be an over-reliance on programmatic activities.

System redundancy, independence, and diversity are maintained commensurate with the expected frequency of challenges to the system.

System redundancy, independence, and diversity are affected when in new Condition 0, but, as described below, they are maintained commensurate with the expected frequency of challenges to the system. Although the NSWS A and B loops have a common discharge header to the SNSWP, the NSWS will be capable of supplying required essential equipment (both NSWS trains of both units) with design basis cooling water flows to support accident mitigation on one unit and the shutdown and cooldown loads of the other unit.

In new Condition 0, the discharge of NSWS from both EOGs of a unit (either 1/2RN848B or 1/2RN846A) and the discharge of NSWS from each unit's essential safety systems (ESS) trains (either 1RN58B or 1RN63A) would be vulnerable to a spurious closure of a motor operated valve (MOV). However, the vulnerable MOVs will be opened with power removed, eliminating

the possibility of a spurious active failure. Furthermore, if an alternate discharge path is needed, MOVs 1/2RN849B, 1/2RN846A, 1RN57A can be repositioned from the control room to divert NSWS discharge to Lake Wylie instead of to the NSWS Pond, thereby restoring NSWS flow.

Another vulnerability would be the spurious closing of either 1RN3A or 1RN4B which supply the two NSWS pump suction pits. In a letter dated March 9, 2018 (ML18068A505), the NRC staff asked the licensee to describe necessary actions to either prevent a spurious closing of 1 RN3A/4B or bring both units to safe shutdown after a OBA with a spurious closing of 1RN3A or 1RN4B, when in the NSWS Single Pond Return Header lineup. By letter dated May 8, 2018 (ML18129A053), the licensee stated that the spurious operation, or closing, of 1 RN3A or 1 RN4B will be prevented by opening each valve during the alignment of the NSWS to the SNSWP and removing power from the valve's motor operator. The NRC staff concludes these proposed actions are acceptable.

Another vulnerability is the cross connection of the A and B ESS headers by the NSWS header crossover valves (1 RN53B and 1RN54A). These valves will be opened with power removed.

These crossover valves are open to allow the A loop and B loop ESS headers to discharge through the single NSWS Single Pond Return Header. Valves 1RN53B and 1RN54B would normally auto-close on a RN low-low suction pit level, but with 1RN3A and 1RN4B open with power removed, an RN low pit level is very unlikely, therefore, the auto close feature would not be needed. But, with these valves open and power removed, the valves cannot be shut to isolate the headers in response to a leak or break. Leaks and breaks in piping are considered passive failures and are not assumed subsequent to a OBA (i.e., a leak or break concurrent with a OBA is highly unlikely and is not considered a credible event). The NRC staff notes, however, that a leak or break could be an initiating event.

Considering leaks and breaks as initiating events in the Single NSWS Pond Return Header alignment, redundancy of NSWS Pond return headers is not available in this alignment. If the NSWS Pond Return headers were available, any NSWS pipe break could be mitigated by simply isolating the crossover valves and relying upon the redundant header. However, leaks and breaks are very unlikely considering that the licensee has stated that the NSWS piping in the Auxiliary Building is robustly designed and supported. Moreover, for all locations where leaks are required to be postulated, leaks can be mitigated by isolation of affected piping and re-alignment to Lake Wylie could be attained with continued operation of NSWS. On page 28 of its letter dated September 14, 2017, the licensee further stated:

For all NSWS leaks, Station Abnormal Procedures will direct Operators to troubleshoot, determine the leak location, and perform re-alignment to isolate the leak and ensure adequate NSWS flow. NSWS Single Pond Return Header Operation for each train is such that motor operated valves will be used to perform the majority of the isolations and re-alignment to the SNSWP. A few manual isolation valves are part of the isolation and are in accessible areas of the Auxiliary Building or EOG Buildings. In addition, all identified motor operated valves have local hand wheels on the actuators and can be locally (manually) repositioned in the event of a loss of actuator power.

While the response to pipe leaks while in Single NSWS Pond Return Header Operation is not as simple as isolation of crossovers and switching to the redundant train, it is reasonable to expect resolution in the assumed 30 minutes per the Catawba Pipe Rupture Specification. Postulated leaks while in NSWS

Single Pond Return are bounded by the existing case of an 1898 GPM leak on the 42" NSWS Header.

On page 17 of its letter dated September 14, 2017, the licensee also stated:

No single active failure can be postulated that result in a leak or a significant diversion of flow that would affect the ability of the essential headers to provide required flow to essential components.

The licensee will install a X inch reinforcing pad at the 1A Component Cooling Heat Exchanger per plant modification prior to entry into new Condition D. The X inch reinforcing pad must be installed prior to entering NSWS Single Pond Return Header Operation. This will ensure that for all NSWS sections where pipe ruptures are postulated, leaks can be isolated with the NSWS continuing to operate with adequate equipment to support shutdown of both units. Installation of the X inch pad at the 1A Component Cooling Heat Exchanger is a one-time permanent modification. A License Condition is being added (see Section 2.4 of this safety evaluation), to ensure the installation of this pad is completed prior to any entrance of the new Condition D.

The licensee modeled 16 cases considering either the NSWS Pond Return Header A or B out of service, a LOCA in either unit with LOOP in both units, and a loss of either EOG or the corresponding NSWS pump as the single failure. In all cases, the results show the NSWS system is capable of providing adequate flow and pressure for all design loads. The licensee will perform flow balance testing prior to entry into the Single NSWS Pond Return Header Operation to ensure the system will operate as predicted.

In conclusion, for a passive failure external leak or break in the Auxiliary Building piping while in Single NSWS Pond Return Header Operation, the NSWS is considered to still be capable of meeting its design requirements even though train separation does not occur. In the existing dual discharge header design, a leak on one train can be isolated such that flow to essential components on the other train is not affected. It also allows the leak to be subsequently stopped by shutdown of pumps in the faulted train. However, in NSWS Single Pond Return Header Operation, as discussed above, passive failure leaks will not divert enough flow to starve essential equipment of needed flow. In addition, the amount of leakage postulated can be tolerated on a long-term basis without affecting the function of the NSWS or the ultimate heat sink.

Finally, the NRC staff notes that, the licensee has taken adequate compensatory measures, such as pre-alignment of MOVs and removing MOV power as appropriate, to maintain redundant functionality of NSWS for both units. Passive failures such as leaks and breaks are unlikely and can be mitigated. Therefore, the NRC staff considers that the single common discharge header lineup as proposed by the licensee has sufficient redundancy, independence and diversity commensurate with the expected frequency and consequences of challenges to the system.

Defenses against potential common-cause failures (CCFs) are maintained and the potential for introduction of new CCF mechanisms is assessed.

Possible CCFs that could disable both trains of NSWS to either or both units are: (1) breakage of the common discharge line, (2) an active failure of an MOV, (3) external leakage, (4) internal blockage, such as a stem to disc failure (5) operator error (shutting an MOV or manual valve).

Pipe breakage or inadvertent closure of one MOV can cause loss of both EDGs of a unit or loss of both ESS trains. As noted previously, breakage of the common discharge line is a passive failure and is not considered credible as a single failure subsequent to a DBA, but could be an initiating event. As an initiating event, the Catawba Pipe Rupture Analysis Criteria Specification has a bounding case of 1898 gallons per minutes (gpm) leak on a 42 inch pipe. Station procedures direct operators to isolate the leak and realign the system to ensure adequate NSWS flow. The licensee has stated that for all locations where leaks are required to be postulated, leaks can be mitigated by isolation of affected piping and re-alignment of suction and discharge sources to Lake Wylie, if needed, with continued operation of the NSWS.

The licensee states that spurious or inadvertent closure of an MOV will be averted by pre-aligning applicable MOVs open and removing power. Additionally, plant operators have the capability from the control room to realign the NSWS discharge to Lake Wylie.

Credible external leakage is limited to flange or packing leaks which can be mitigated by Nuclear Safety Related sump pumps. In NSWS Single Pond Return Header Operation, passive failure leaks will not divert enough flow to starve essential equipment of its needed flow. In addition, the amount of leakage can be tolerated on a long-term basis without affecting the function of the NSWS or ultimate heat sink.

Flow blockage by failure of a MOV could cause loss or partial loss of NSWS. The licensee will open and remove power from the crossover MOVs ( 1RN53B and 1RN54A) and SNSWP isolation valve ( 1RN58B or 1RN63A as applicable). Removing power from these MOVs will eliminate spurious operation and operator error in unintended positioning of these MOVs.

Disc/stem failure is considered a passive failure and not credible as a single failure after a DBA, but it could be an initiating event. Mitigation of this type of failure is re-alignment of the NSWS system to an alternate heat sink.

Based on the above described NSWS capability and protection from CCFs as described above, the NRC staff concludes that defenses against CCF potential are maintained during the single discharge header alignment for a 30 day CT, limited to 60 days per calendar year.

Independence of physical barriers is not degraded.

The proposed TS change does not affect the fuel cladding, reactor coolant pressure boundary, or the containment. Thus, the independence of these barriers is not affected by the TS amendment.

Defenses against human errors are maintained.

Possible human errors would include incorrect operation of valves. However, the licensee stated that MOVs in the discharge headers that either isolate the work area or are in line to the common discharge header will be prepositioned and have power removed. The licensee will revise procedures and train personnel for the evolution. Therefore, defenses against human errors are maintained.

The intent of the plant's design criteria is maintained.

The licensee is not proposing to modify the plant design criteria. The licensee requests modification to how the plant is operated for a period of maximum 60 days per calendar year and only for specific pre-planned maintenance work.

In Section 2.1 of its letter dated September 14, 1017 the licensee states, part:

During normal operation, the NSWS supplies cooling water to various safety related components. While in normal operation, the maximum heat load and flow requirements on the NSWS are encountered with the Unit in Mode 5 due to decay heat removal requirements. During ANSI N18.2-1973 initiated events, the NSWS is required to support Emergency Core Cooling System (ECCS) operation by providing cooling water to various safety related components along with emergency makeup to selected QA Condition 1 Systems. The ANSI N18.2-1973 event that imposes the most stringent design basis requirement is the Condition IV Initiator of Loss of Coolant Accident (LOCA).

The NRC staff notes any single failure in the common discharge line would prevent the cooling system from meeting its design function. However, as described above, the credible failures in the common discharge line are compensated by prepositioning valves, locking out power, using the safety tag system and training operators.

The licensee has performed calculations that show that when the NSWS is aligned for NSWS Single Pond Return Header Operation, there is adequate NSWS flow to all design loads assuming a LOCA in either unit and LOOP in both units, with a worst case single failure of any EOG and its corresponding NSWS pump.

Therefore, the NRC concludes the intent of plant design is maintained and is acceptable for the limited period of 60 days per calendar year and only for the maintenance activities described above.

3.1.2 Safety Margins The extended CT is not in conflict with Codes and Standards approved for use by the NRC relevant to the NSWS. In the proposed Single Pond Return Header lineup and assuming no additional failures, all NSWS cooling loads receive full design flow, and thus, the NSWS is fully ready to perform its safety functions and satisfy the acceptance criteria for all applicable safety analysis as specified in the UFSAR.

3.1.3 Surveillance Requirement 3. 7 .8.2 SR 3.7.8.2 requires verification that each NSWS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

However, when in single supply, or single Auxiliary Building discharge header operation, or single pond return header operation, certain automatic valves in the system are intentionally secured in position and will not automatically reposition in response to an actuation signal.

Since these NSWS lineups involve securing these valves in position when in the single supply, or single Auxiliary Building discharge header, or single pond return header lineup, SR 3.7.8.2 need not be performed for those normally automatically operated valves when in these NSWS lineups.

3.1.4 Summary of RG 1.177 Considerations System redundancy, independence, and diversity are maintained commensurate with the expected frequency and consequences of challenges to the system. Although complete discharge header redundancy is not maintained during the limited CT extension, system redundancy, independence, and diversity are maintained commensurate with the expected frequency and consequences of challenges to the NSWS system because the likelihood of a failure of the remaining discharge header is very low. Additionally, the licensee has identified satisfactory compensatory measures that would restore NSWS flow if the unlikely failure of the remaining discharge header occurred. The licensee has also pre-aligned the NSWS system by pre-positioning and removing power to various MOVs to reduce the likelihood of failures that would challenge the NSWS safety functions. In the single discharge header lineup, flow calculations show adequate NSWS flow to all design loads assuming a OBA in either unit with a worst case single failure of one EDG and its corresponding NSWS pump. Other vulnerabilities include active failure of an MOV, pipe leaks, and pipe rupture. These vulnerabilities are mitigated by pre-alignment, removing power to certain MOVs and relying on operator action to isolate leaks or ruptures while maintaining adequate NSWS flow and shifting to Lake Wylie as a source of NSWS.

Entry in the NSWS Single Pond Return Header Operation is restricted to and contingent upon:

a. Having both trains of EDGs, NSWS Pumps, and NSWS supplied equipment on both units operable.

b. Limited to 60 days per calendar year and only for specific pre-planned maintenance as identified in the LAR.

c. Installation of a~ inch thick reinforcing pad at the 1A Component Cooling (KC) head exchanger, as described in the licensee's letters dated September 14, 2017 and October 29, 2018. A License Condition is added to address this issue (see section 2.4 and 3.1.1 of this safety evaluation for further information).

d. NSWS Flow Balance testing performed immediately prior to each entry into TS Condition D. This will ensure the NSWS is capable of providing adequate cooling water flow to support LOCA loads on one unit, concurrent with the shutdown loads of the other unit, while assuming the most limiting single failure, which is loss of one EDG and its associated NSWS Pump.

e. Not being in either NSWS Single Supply Header alignment (Condition B) or NSWS Single Auxiliary Building Discharge Header alignment (Condition C).

The NSWS Single Pond Return Header Operation allows internal inspections and modifications of the NSWS Pond Return buried piping between the Auxiliary Building and the discharge to the SNSWP. The amount of time required to perform current internal inspections and planned modifications will take in excess of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , which is the current TS CT for the NSWS described in TS 3.7.8. The licensee's proposed 30-day CT for NSWS Single Pond Return Header Operation is supported by probabilistic risk assessment (PRA), as described in Section 3.3 of this safety evaluation.

3.2 Summary of NSWS Single Pond Return Header Operation The NRC staff has reviewed the plant system and requirements for NSWS Single Pond Return Header Operation, as described in Section 2.0 of this safety evaluation for specific pre-planned maintenance for an extended completion time up to 30 days (two entries per calendar year, for a total of 60 days). Based on the above, the NRC staff finds the proposed changes are acceptable and that they meets the requirements of 10 CFR 50.40(a), 10 CFR 50.57(a)(3), and 10 CFR 50.35(c)(3), and considerations of RG 1.177.

As discussed in Section 3.3 of this safety evaluation, the requested 30-day CT for NSWS Single Pond Return Header Operation is also supported by the licensee's probabilistic risk assessment.

3.3 Risk-Informed Considerations 3.3.1 Key Principle 4: Change in Risk Consistent with the Commission's Safety Goal Policy Statement Regulatory Guide 1.177 outlines a three-tiered approach for evaluating the risk associated with a proposed change to a TS CT:

Tier 1 assesses the risk impact of the proposed change in accordance with acceptance guidelines consistent with the Commission's Safety Goal Policy Statement, as documented in RG 1.177. The Tier 1 assessment evaluates the impact of the proposed change on operational plant risk as represented by the change in core damage frequency (flCDF) and the change in large early release frequency (flLERF). In addition to operational plant risk, the Tier 1 assessment evaluates plant risk while equipment covered by the CT change is out-of-service, as represented by the incremental conditional core damage probability (ICCDP) and the incremental conditional large early release probability (ICLERP). The Tier 1 assessment also addresses the quality or technical adequacy of the licensee's plant-specific PRA model used to assess the changes in risk.

Tier 2 identifies and evaluates any potential risk-significant plant configurations that could result if any equipment, in addition to that associated with the proposed license amendment, is taken out-of-service simultaneously, or if other risk-significant operational factors, such as concurrent system or equipment testing, are involved. The purpose of this evaluation is to ensure that there are appropriate restrictions on dominant risk-significant equipment configurations associated with the proposed change.

Tier 3 addresses the licensee's overall configuration risk management program (CRMP) to ensure that the licensee has established adequate programs and procedures for identifying risk-significant plant configurations resulting from maintenance or other operational activities, and that appropriate compensatory measures are taken to avoid risk-significant configurations that may not have been considered in the Tier 2 evaluation. Compared with Tier 2, Tier 3 provides additional coverage to ensure that the licensee identifies, in a timely manner, any potentially risk-significant equipment outage configurations, and that the licensee evaluates appropriately, the risk impact of out-of-service equipment prior to performing any maintenance activity over extended periods of plant operation.

3.3.1.1 Tier 1: PRA Quality and Insights In accordance with Tier 1 of the three-tiered approach outlined in RG 1.177, the licensee should evaluate the change in plant risk resulting from the proposed TS change as represented by the

~CDF, ~LERF, ICCDP, and ICLERP. To support this evaluation, two aspects should be considered: (1) the quality or technical adequacy of the PRA, and (2) the PRA insights and findings. The licensee should demonstrate that its PRA is acceptable for assessing the proposed TS change and identify the impact of the TS change on plant risk.

3.3.1.1.1 PRA Quality Regulatory Guide 1.174 states, "[t]he scope, level of detail, and technical adequacy of the PRA are to be commensurate with the application for which it is intended and the role the PRA results play in the integrated decision process." The technical adequacy of the PRA must be compatible with the safety implications of the proposed TS change and the role that the PRA plays in justifying that change. That is, the more the potential change in risk or the greater the uncertainty in that risk from the proposed TS change, or both, the more rigor that must go into ensuring the technical adequacy of the PRA. This applies to Tier 1, and it also applies to Tier 2 and Tier 3 to the extent that a PRA model is used.

Regulatory Guide 1.200 describes an acceptable approach for determining whether the technical adequacy of the PRA, in total or the parts that are used to support an application, is sufficient for use in regulatory decision making for light-water reactors. Regulatory Guide 1.200, Revision 2, endorses, with clarifications and qualifications, the use of the American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) PRA standard ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications." The ASME/ANS PRA standard is the industry consensus standard for PRAs for internal events, internal flooding, fires, and other external events (i.e., seismic, external flooding, high winds, and so on).

The ASME/ANS PRA standard provides technical supporting requirements in terms of three Capability Categories (CCs). Regulatory Position 2.1 of RG 1.200 states that if a licensee demonstrates (by, e.g., using a peer review and self-assessment in accordance with Regulatory Position 2.2 of RG 1.200) that the parts of a PRA that are used to support an application comply with the ASME/ANS PRA standard, when supplemented to account for the NRC regulatory positions contained in Appendix A, the NRC would consider the PRA to be adequate to support the applicable risk-informed regulatory application.

Regulatory Position 4.2 of RG 1.200 states, in part, that the application should discuss the resolution of the peer review's facts and observations (F&Os) that are applicable to the parts of the PRA that support the application. Appendix X to Nuclear Energy Institute (NEI) 05-04/07-12/12-[13], "Close-Out of Facts and Observations (F&Os)" (NEI Appendix X) (ADAMS Accession No. ML17086A431 ), as accepted with conditions by NRC letter dated May 3, 2017 (ADAMS Accession No. ML17079A427), provides guidance for closing F&Os. The NEI Appendix X states in part, "[o]nce an F&O is closed out, the utility is not required to present and explain them in peer reviews, NRC submittals, or other requests excluding NRC audits." The May 3, 2017 letter also states in part, "[t]he NRC also intends to periodically conduct audits of a licensee's implementation of the Appendix X F&O closure process, as well as review a sampling of the final independent assessment team reports."

Scope of the PRA Section 2.3.2 of RG 1.177 states that:

As a minimum, the evaluations of CDF and LERF should be performed [by the licensee] to support any risk-informed changes to TS. The scope of the analysis should include all hazard groups (i.e., internal events, internal flooding, fires, seismic events, high winds, and other external hazards).

Section 2.3.1 of RG 1.174 states in part that a qualitative treatment of the missing modes and hazard groups may be sufficient when the licensee can demonstrate that those risk contributions would not affect the decision.

In its letter dated September 14, 2017, as supplemented by letters dated August 17 and September 20, 2018, the licensee performed a quantitative evaluation of the change in risk (i.e.,

6CDF, 6LERF, ICCDP, and ICLERP) resulting from the proposed TS change for at-power internal events, internal flooding, fire, and high winds. The licensee provided a qualitative evaluation for seismic hazards and other external events demonstrating that those risk contributions do not impact the conclusions of the LAR. LAR Section 3.2.5, "PRA Model Configuration and Control Program," discusses the license's PRA configuration and control program for maintaining and updating the PRA and confirmed that the PRA models (i.e., internal events, internal flooding, fires, and high winds) and associated risk assessments used to support the LAR sufficiently represents the as-built, as-operated plant.

Based on the above, the NRC staff finds that, when compared to the guidance contained in RGs 1.174, 1.177, and 1.200, the licensee's risk assessment is of sufficient scope for use in this specific risk-informed application.

Internal Events PRA (excluding LERF)

The quality of the licensee's PRA must be commensurate with the safety significance of the proposed TS change and the role the PRA plays in justifying the change. That is, the higher change in risk or the greater the uncertainty in that risk from the requested TS change, or both, the more rigor that must go into ensuring the quality of the PRA.

RG 1.200 provides regulatory guidance for assessing the technical adequacy of a PRA. The current revision (Revision 2) of this RG endorses (with comments and qualifications) the use of the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS)

RA-Sa-2009, "Addenda to ASME RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," NEI 00-02, "PRA Peer Review Process Guidelines," and NEI 05-04, "Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard."

The licensee has performed an assessment of the PRA models used to support the license amendment using the guidance of RG 1.200 to assure that the PRA models are capable of determining the change in risk due to changes to the completion time of the NSWS, using plant-specific data and models. Capability Category II of the endorsed PRA standard is the target capability level for supporting requirements for the internal events PRA for this application. Any identified deficiencies to those requirements are assessed further to determine any impacts to proposed changes to the plant, including the use of sensitivity studies where appropriate.

Section 3.2.2, "PRA Quality/Technical Adequacy," of the licensees September 14, 2017, letter, as supplemented by the licensee's response to request for additional information (RAI) 08 in letter dated August 17, 2018 (ADAMS Accession No. ML18232A245), addresses technical adequacy of the non-LERF portion of the Catawba internal events PRA (IEPRA (non-LERF)).

The IEPRA (non-LERF) received a full-scope peer review in December 2015 using the process defined in NEI 05-04, "Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard, Revision 2," (ADAMS Accession No. ML083430462). This peer review was against the internal events PRA requirements (excluding LERF) of the ASME/ANS RA-Sa-2009 PRA standard and RG 1.200, Revision 2, with consideration for identified changes from Addendum A (ASME/ANS RA-Sa-2009) to Addendum B (ASME/ANS RA-Sb-2013) of the PRA standard. NRC does not endorse ASME/ANS RA-Sb-2013; therefore, the licensee performed a gap assessment and identified no gaps between the internal events peer review (excluding LERF) and the requirements of RG 1.200, Revision 2. A number of IEPRA F&Os were generated from the 2015 peer review.

An F&O closure review of the IEPRA (non-LERF) F&Os was conducted by independent assessment in August 2017 in accordance with NEI Appendix X to NEI 05-04, as accepted by NRC letter dated May 3, 2017. The scope of this F&O closure review included all finding-level F&Os associated with the IEPRA (non-LERF), including those finding-level F&Os associated with SRs that were met at CC II. Consistent with NEI Appendix X, as accepted by NRC letter dated May 3, 2017, the independent assessment team was provided with written justification regarding whether each F&O resolution was PRA maintenance or PRA upgrade as defined in ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2. The licensee also described how the F&O closure review team met the appropriate qualifications and criteria in Section X.1.3 of NEI Appendix X. As a result, Catawba "closed out" a number of IEPRA (non-LERF) F&Os that were not included in the submittal. One F&O resolution was considered a PRA upgrade (related to SRs HR-A1 through HR-C3 of the PRA standard), and a focused-scope peer review was performed in September 2017 in accordance with RG 1.200, Revision 2, to close this F&O and show that the associated SRs are fully met at CC 11/111, with no new F&Os.

A NRC audit team conducted a regulatory audit from May 8 - 10, 2018 at the Duke Energy Offices in Charlotte, NC (ADAMS Accession Number ML18249A046). The NRC staff performed a detailed review of the Catawba F&O closure report for IEPRA (non-LERF). NRC staff activities during the audit included: (1) confirming that the independent assessment team's basis for closure of each F&O included an assessment of whether the F&O resolution is a PRA maintenance or PRA upgrade; (2) verifying that each F&O resolution was met at CC II, or met if there was no separate CC II; and (3) the independent assessment team closed-out F&Os that were appropriately documented and incorporated into the PRA model. Based on confirmation that the F&O closure review is consistent with the NRC accepted process, the NRC staff concludes that the licensee has adequately implemented the F&O closure process for the IEPRA (non-LERF). .1, "PRA Peer Review Findings and Resolutions, Internal Events, CDF," of the licensees September 14, 2017, letter, as supplemented by response to RAI 06.b in letter dated August 17, 2018, provides eight (8) internal events (excluding LERF) finding-level F&Os that remain open after the December 2015 peer review and the August 2017 F&O closure review, and the licensee's disposition to these F&Os. Each F&O was dispositioned by either providing a description of how the F&O was resolved or providing an assessment of the impact of resolution of the F&O on the results for the LAR for the proposed TS change. The NRC staff evaluated each F&O and the licensee's disposition in Attachment 4.1 of the licensee's*

September 14, 2017, letter, to determine whether the F&O had any significant impact for the

application. The NRC staff finds that all F&Os were properly assessed and dispositioned to support the IEPRA (non-LERF) technical adequacy for the proposed TS change. Finally, the LAR risk results (i.e., ~CDF, ~LERF, ICCDP, and ICLERP) meet the RG 1.174 and RG 1.177 risk acceptance guidelines by a large margin. This provides additional confidence that any uncertainties associated with the Catawba IEPRA would not change the conclusions of this assessment.

Based on the above, the technical acceptability of the PRA for internal events (excluding LERF),

as described by the licensee, is sufficient for use in supporting this specific risk-informed application.

Internal Flooding PRA and LERF portion of Internal Events PRA Section 3.2.2 of the licensee's September 14, 2017, letter, as supplemented by response to RAls 08 and 1O.a, in letter dated August 17, 2018, addresses technical adequacy of the Catawba internal flooding PRA (FLPRA) and LERF portion of the IEPRA (IEPRA (LERF)).

These PRAs received a full-scope peer review in 2012 against the associated PRA requirements of the ASME/ANS RA-Sa-2009 PRA standard, as endorsed by RG 1.200, Revision 2. A number of F&Os were generated from the 2012 peer review. The Catawba FLPRA is based on Unit 1 only. The licensee explained how the Unit 1 model is applicable to Unit 2 due to nearly identical systems, structures, and components (SSCs), design and operation, and similar spatial configuration. Appropriate modeling changes were made in consideration of the differences between the units, with Unit 1 results bounding that of Unit 2.

Section 3.2.2 of the licensee September 14, 2017, letter, as supplemented by response to RAI 05 in letter dated August 17, 2018, discusses the closure of the FLPRA and IEPRA (LERF)

F&Os. In December 2015, an independent review of the FLPRA and IEPRA (LERF) F&Os was performed to determine whether the F&Os were resolved and the corresponding SRs are met at CC II or greater. However, this independent review was performed prior to NRC acceptance of NEI Appendix X on May 3, 2017. Upon NRC acceptance of NEI Appendix X, the licensee identified a deficiency between the 2015 independent review of FLPRA and IEPRA (LERF) F&O resolutions and NEI Appendix X, as accepted by NRC letter dated May 3, 2017, where the 2015 independent review did not identified whether the F&O resolutions were PRA maintenance or PRA upgrades. To resolve this deficiency and meet the NEI Appendix X requirements, the same individuals who performed the 2015 independent review performed a second independent review in July 2017. The 2017 independent review included an assessment of whether each F&O resolution constituted a PRA maintenance or PRA upgrade, and an assessment of how each requirement of NEI Appendix X, as accepted by NRC letter dated May 3, 2017, was met by the combined independent reviews in 2015 and 2017. The scope of the 2015 and 2017 independent reviews included all finding-level F&Os associated with the FLPRA and IEPRA (LERF), including those finding-level F&Os associated with SRs that were met at CC II.

Consistent with NEI Appendix X, as accepted by NRC letter dated May 3, 2017, the 2017 independent review team was provided with written justification regarding whether each F&O resolution was PRA maintenance or PRA upgrade as defined in ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2. The licensee also described how the independent review team met the appropriate qualifications and criteria in Section X.1.3 of NEI Appendix X. As a result, Catawba "closed out" a number of FLPRA and IEPRA (LERF) F&Os that were not included in the submittal. During the May 8 - 10 audit, the NRC staff performed a detailed review of the Catawba F&O closure reports for FLPRA and IEPRA (LERF). Based on confirmation that the combined independent reviews in 2015 and 2017 is consistent with the

NRC accepted process, the NRC staff concludes that the licensee has adequately implemented the F&O closure process for the FLPRA and IEPRA (LERF). .2, "PRA Peer Review Findings and Resolutions, Internal Events, LERF," of the licensees September 14, 2017, letter, provides five finding-level F&Os for IEPRA (LERF) that remain open after the combined independent reviews in 2015 and 2017, and the licensee's disposition to these F&Os. Each F&O was dispositioned by either providing a description of how the F&O was resolved or providing an assessment of the impact of resolution of the F&O for the proposed TS change. The NRC staff evaluated each F&O and the licensee's disposition, to determine whether the F&O had any significant impact for the application. The NRC staff finds that all F&Os were properly assessed and dispositioned to support the IEPRA (LERF) technical adequacy for the proposed TS change. No finding-level F&Os for internal flooding remain open after the combined independent reviews in 2015 and 2017. Finally, the LAR risk results (i.e., ~CDF, ~LERF, ICCDP, and ICLERP) meet the RG 1.174 and RG 1.177 risk acceptance guidelines by a large margin. This provides additional confidence that any uncertainties associated with the Catawba FLPRA and IEPRA (LERF) would not change the conclusions of this assessment.

Based on the above, the technical acceptability of the PRA for internal flooding and the LERF portion of internal events, as described by the licensee, is sufficient for use in supporting this specific risk-informed application.

Fire PRA The Catawba internal fire PRA (FPRA) address both CDF and LERF. The licensee used RG 1.200, Revision 2, to address the technical adequacy of the FPRA to assure the PRA is capable of accurately characterizing the risk impact from internal fire associated with the TS CT change for NSWS. Consistent with the guidance of RG 1.177 and Revision 2 of RG 1.200, capability category II of ASME/ANS RA-Sa-2009 was applied as the standard, and any identified deficiencies to those requirements (i.e., if either a supporting requirement was not met or met at CCI) were assessed further to determine any impacts to the risk evaluation.

Section 3.2.2 of the licensees September 14, 2017, letter, as supplemented by response to RAls 07 and 1O.a, in letter dated August 17, 2018, addresses technical adequacy of the Catawba fire PRA (FPRA). The Catawba FPRA is composed of site-specific models for Unit 1 and Unit 2, and received a full-scope peer review in July 2010 using the process defined in 'NEI 07-12, "Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, Revision O." This peer review was against the associated PRA requirements of the ASME/ANS RA-Sa-2009 PRA standard, as endorsed by RG 1.200, Revision 2. Attachment 4.5, "PRA Peer Review Findings and Resolutions, Fire," of the LAR provides twenty (20) finding-level F&Os for the FPRA that remain open after the 2010 peer review. Each F&O was dispositioned by either providing a description of how the F&O was resolved or providing an assessment of the impact of resolution of the F&O on the results for the LAR for the proposed TS change. The NRC staff evaluated each F&O and the licensee's disposition in LAR Attachment 4.5, as supplemented, to determine whether the F&O had any significant impact for the application. The NRC staff finds that all F&Os were properly assessed and dispositioned to support the FPRA technical adequacy for the proposed TS change.

In addition, the NRC staff reviewed the safety evaluation associated with the licensee's transition to National Fire Protection Association Standard 805 (NFPA 805), dated February 8, 2017 (ADAMS Accession No. ML16137A308), to identify any issues related to the technical

adequacy of the FPRA that could impact the LAR for the proposed TS change. The Catawba NFPA 805 safety evaluation concluded that the licensee's FPRA approach, methods, and data are acceptable based on: (1) the PRA models (i.e., internal events and fire) adequately represent the current as-built, as-operated plant configuration; (2) the PRA models conform to the applicable industry PRA standards for internal events and fires, considering the acceptable disposition of the peer review and NRC staff review findings which have been incorporated into the FPRA model; and (3) the fire modeling used to support the development of the FPRA has been confirmed as appropriate and acceptable. Finally, the LAR risk results (i.e., ~CDF,

~LERF, ICCDP, and ICLERP) meet the RG 1.174 and RG 1.177 risk acceptance guidelines by a large margin. This provides additional confidence that any uncertainties associated with the Catawba FPRA would not change the conclusions of this assessment.

Based on the discussion above, the technical acceptability of the FPRA, as described by the licensee, is sufficient for use in supporting this specific risk-informed application.

High Winds PRA The Catawba High Winds PRA (HWPRA) address both CDF and LERF. The licensee used RG 1.200, Revision 2, to address the technical adequacy of the HWPRA to assure the PRA is capable of accurately characterizing the risk impact from high winds associated with the TS CT change for NSWS. Consistent with the guidance of RG 1.177 and Revision 2 of RG 1.200, capability category II of ASME/ANS RA-Sa-2009 was applied as the standard, and any identified deficiencies to those requirements (i.e., if either a supporting requirement was not met or met at CCI) were assessed further to determine any impacts to the risk evaluation.

Section 3.2.2 of the licensees September 14, 2017, letter, as supplemented by response to RAls 07, 08, and 10.a, in letter dated August 17, 2018, addresses technical adequacy of the Catawba high winds PRA (HWPRA). The Catawba HWPRA is composed of site-specific models for Unit 1 and Unit 2, and incorporates the latest updates of the internal event PRA. The HWPRA received a full-scope peer review in August 2013 against the associated PRA requirements of the ASME/ANS RA-Sa-2009 PRA standard, as endorsed by RG 1.200, Revision 2. Attachment 4.4, "PRA Peer Review Findings and Resolutions, High Winds," of the LAR, as supplemented by response to RAI 06.a in letter dated August 17, 2018, provides five (5) finding-level F&Os for the HWPRA that remain open after the 2013 peer review. Each F&O was dispositioned by either providing a description of how the F&O was resolved or providing an assessment of the impact of resolution of the F&O on the results for the LAR for the proposed TS change. The NRC staff evaluated each F&O and the licensee's disposition in LAR .4, as supplemented, to determine whether the F&O had any significant impact for the application. The NRC staff finds that all F&Os were properly assessed and dispositioned to support the HWPRA technical adequacy for the proposed TS change. Finally, the LAR risk results (i.e., ~CDF, ~LERF, ICCDP, and ICLERP) meet the RG 1.174 and RG 1.177 risk acceptance guidelines by a large margin. This provides additional confidence that any uncertainties associated with the Catawba HWPRA would not change the conclusions of this assessment.

Based on the above, the technical acceptability of the HWPRA, as described by the licensee, is sufficient for use in supporting this specific risk-informed application.

Seismic Hazards and Other External Events Regulatory Position 2.3.2 of RG 1.177 states that the scope of the analysis should include all hazard groups (i.e., internal events, internal flooding, internal fires, seismic events, high winds, and other external hazards) unless it can be shown that the contribution from specific hazard groups does not affect the decision.

Section 3.2.4.6, "Seismic Risk," of the licensees September 14, 2017, letter, discusses the seismic risk. Structures such as NSWS pump structure as well as the SNSWP intake and discharge structure were screened out from the seismic PRA analysis that had been performed for the Catawba Individual Plant Examination of External Events (IPEEE) submittal. The licensee concluded that the consideration of risk from seismic events "while the plant is in the SNSWP single return header configuration is not a significant factor for this assessment."

However, it was not clear to the NRC staff whether the licensee's qualitative assessment has considered the impact of seismic failures for all SSCs that could affect the change in risk for this application. As indicated in the licensee's response to RAI 16 in the letter dated August 17, 2018, the licensee provided the median fragilities for the NSWS equipment considered for this application, which met or exceeded the screening values for median fragility of SSCs as part of the IPEEE submittal. The licensee also explained that the EDGs, the power sources required to power the NSWS pumps and associated valves, are considered rugged; however, some of the support components have lower fragilities and were included in the IPEEE analysis. The licensee further stated that these EDGs become the predominant SSCs of interest during a seismic-induced LOOP event. The licensee stated this would be true regardless of whether the IPEEE plant hazard analysis or the updated plant hazard analysis is used to perform a seismic risk evaluation, and which hazard analysis is utilized would not impact this application. The NRC staff finds that the screened-out SSCs are more rugged than the EDG support components (as shown in the fragilities provided in the RAI response), and, therefore, EDG failures are the dominant failures. Thus, the NRC staff concludes that the risk associated with the aforementioned SSCs does not impact the decision on the proposed TS change.

As indicated in the licensee's response to RAI 09 in the letter dated August 17, 2018, external hazards were evaluated by the licensee to determine if they impacted the application. All other external hazards were screened out from applicability through a plant-specific evaluation performed in accordance with Generic Letter (GL) 88-20, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f)," GL 88-20, Supplement 4, dated June 28, 1991. Additionally, as part of the licensee's Fukushima Near-Term Task Force {NTTF), Recommendation 2.1 response (ADAMS ML14077A054), the licensee provided its response to external flooding. Additionally, the NRC staff finds that the licensee followed RG 1.177 by performing quantitative or qualitative bounding analyses of hazards not addressed using PRA and determining that those hazards do not impact this application.

PRA Quality Conclusions Based on the above, the NRC staff concludes that: (1) the licensee's risk assessment is of sufficient scope for use in this risk-informed application; (2) the Catawba PRA models (i.e.,

internal events, internal flooding, fire, and high winds PRAs) meet the guidance in Regulatory Guide 1.200, Revision 2, and are, therefore, acceptable to adequately predict the change in CDF and LERF for use in this risk-informed application; and (3) the hazards not addressed using PRA (i.e., seismic and other external hazards) were determined not to impact this application. In addition, the LAR risk results (i.e., ~CDF, ~LERF, ICCDP, and ICLERP) meet

the RG 1.174 and RG 1.177 risk acceptance guidelines by a large margin. This provides additional confidence that any uncertainties associated with the Catawba PRAs would not change the conclusions of this assessment.

3.3.1.1.2 PRA Results and Insights The licensee evaluated the risk impact of the proposed TS change using the internal events, internal flooding, internal fire, and high winds PRA models. This risk evaluation is specific to the 30-day CT for the Catawba NSWS Single Pond Return Header Operation with all relevant configurations represented in the PRA models, including the following considerations:

While there is no specific plant mode requirement for use of the SNSWP alignment, the risk evaluation conservatively assumes at-power operation.

Condition D of this TS Condition specifies the time limitation of 60 days per 12-month period for Single Pond Return Header Operation. Therefore, the risk evaluation assumed one TS entry per header train during a year (i.e., two entries into the TS per year).

As indicated in the licensee's response to RAI 10.b in the letter dated August 17, 2018, with the exception of the EDGs and NSWS SSCs, the risk evaluation assumed average test and maintenance unavailabilities for SSCs modeled in the PRA, consistent with RG 1.177.

Entry into this TS Condition prohibits performing scheduled or planned maintenance that renders the EDGs and NSWS SSCs unavailable on either train of NSWS for either unit is prohibited. For this reason, the corresponding test and maintenance events for these SSCs were set to O (i.e., not unavailable due to test and maintenance) in the risk evaluation.

Entry into this TS Condition requires NSWS to be pre-aligned to the SNSWP prior to entering the Single Pond Return Header Condition. This means the normal supply and discharge flow paths from/ to Lake Wylie will be isolated and normal flow will be provided via the SNSWP supply and return headers. Placing the NSWS in this configuration creates a more 'passive' condition since potential failures due to an automatic swap from Lake Wylie to the SNSWP on low lake level are eliminated.

Valve alignment and control (e.g., valve locked open, power removed from valve and tagged to prevent local operation) assumed in the risk evaluation is consistent with that required for entry into this TS Condition. For FPRA modeling purposes, a motor-operated valve (MOV) with power removed represents a removal of an ignition source.

The use of FLEX equipment and Emergency Supplemental Power Source (ESPS) was not credited in the risk evaluation. This is a conservative assumption.

Entry into this TS Condition is only allowed for pre-planned activities, therefore, the increased potential for a common cause failure of NSWS is minimal during the 30-day CT.

As discussed in LAR Section 3.2.3.2, "General Assumptions and Model Changes," and supplemented by the licensee's response to RAI 11 in its letter dated September 20, 2018, the fire risk evaluation credited two recovery actions new to the plant to address NSWS flow diversion caused by spurious operation of MOVs due to fire. The human error probabilities (HEPs) associated with these recovery actions were based on a screening value of 1.00E-02, because procedural guidance and training have not yet been finalized. The licensee performed a detailed human reliability analysis of the two new recovery actions consistent with the guidance in NUREG-1921, "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines - Final Report" (July 2012) (ADAMS Accession No. ML12216A104), to demonstrate the recovery actions are feasible and the associated HEP screening values are conservative. The licensee also discussed the cues or indications that operators will use to initiate these recovery actions and the timeline to perform these actions. In response to RAI 11, the licensee proposed a license condition (see section 2.4 of this SE) that, prior to implementing the NSWS Single Pond Return Header Operation with a 30-day CT, the plant engineering process will be utilized to develop new plant procedures and required training to support the identified alignment by the two new recovery actions credited in the PRA. The HEPs for these recovery actions will be revised, as needed, to be consistent with the procedural guidance and training. Risk estimates will also be updated to include the updated HEPs and confirm that the risk acceptance guidelines of RG 1.177 and RG 1.174 are met. If the risk estimates are not met, additional risk reduction measures will be taken as needed to ensure that the acceptance guidance are met. Based on the discussion above, the NRC staff concludes the licensee's process utilized to develop new plant procedures and required training to support the identified alignment by the two new recovery actions, is acceptable.

Risk evaluation results for Catawba's NSWS Single Pond Return Header Operation with a proposed 30-day CT are presented in the table below and compared to the risk acceptance guidelines of RG 1.174 and 1.177. The licensee's calculated risk values in this table are well below the RG 1.174 and RG 1.177 risk acceptance guidelines for change in risk (i.e., f1CDF and f1LERF) and incremental increase in risk (i.e., ICCDP and ICLERP), and are, therefore, acceptable. The ICCDP and ICLERP values indicate a slight risk reduction relative to the base case risk due to: (1) valve alignment and control (e.g., valve locked open, power removed from valve and tagged to prevent local operation) prior to entering the NSWS Single Pond Return Header Operation; (2) TS that prohibits performing scheduled maintenance that renders the EDGs and NSWS SSCs unavailable; and (3) the two new recovery actions to address NSWS flow diversion.

30-day unavailability of the NSWS Single Pond Return Header Operation Catawba, Unit 1 Risk Metric Acceptance Guideline PRA Results

~CDF RG 1.174, Figure 4 (Region II or Ill) 5.22E-07 (Region Ill)

~LERF RG 1.174, Figure 5 (Region II or Ill) 3.58E-08 (Region Ill)

ICCDP < 1.0E-6 -4.82E-07 ICLERP < 1.0E-7 -9.53E-8 Catawba, Unit 2 Risk Metric Acceptance Guideline PRA Results

~CDF RG 1.174, Figure 4 (Region II or Ill) 6.20E-07 (Region Ill)

~LERF RG 1.174, Figure 5 (Region II or Ill) 5.33E-08 (Region Ill)

ICCDP < 1.0E-6 -4.46E-07 ICLERP < 1.0E-7 -8.88E-08 3.3.1.1.3 Sensitivity and Uncertainty Analyses Regulatory Position 2.3.5 of RG 1.177 states that the risk resulting from TS CT changes is relatively insensitive to uncertainties, because uncertainties associated with CT changes tend to similarly affect the base case and the change case.

In response to RAI 14, in its letter dated August 17, 2018, the licensee described the approach used to identify and characterize key assumptions and key sources of uncertainty consistent with NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making" (ADAMS Accession No. ML090970525). The licensee addressed parameter, modeling, and completeness uncertainties for the PRAs used in the risk evaluation. None of the identified assumptions or sources of uncertainty were found to be key (or significant) to the application. Also, based on the large margin between the risk evaluation results and the risk acceptance guidelines of RG 1.174 and 1.177 and the compensatory actions discussed in LAR Section 3.2.7, "Compensatory Actions," which the licensee committed to implement, the NRC staff concludes that PRA model uncertainties are not sufficient to change the conclusions in the licensees letter dated September 14, 2017, as supplemented by its letter dated August 17, 2018.

3.3.1.2 Tier 2: Avoidance of Risk-Significant Plant Configurations The avoidance of risk-significant plant configurations limits potentially high-risk configurations that could exist if equipment, in addition to that associated with the proposed TS change, is simultaneously removed from service. Risk-significant operational factors, such as concurrent system or equipment testing, should also be avoided. Therefore, the guidance in RG 1.177 provides that a licensee's Tier 2 evaluation should ensure that appropriate restrictions are placed on dominant risk-significant configurations relevant to the proposed TS change.

In Section 3.2.6, "Tier 2 Component Evaluation," of the licensee's September 14, 2017, letter, as supplemented by its response to RAI 12 in the letter dated August 17, 2018, the licensee identifies the SSCs that are significant to risk during the 30-day CT for NSWS Single Pond Return Header Operation. These SSCs were identified based on configuration-specific risk insights provided by the Catawba IEPRA, FLPRA, FPRA, and HWPRA, and as part of the

Catawba CRMP. The mechanisms used to ensure that appropriate restrictions are placed on dominant risk-significant configurations during the 30-day CT include: (1) Technical Specifications , (2) cycle schedule (i.e., testing and maintenance of plant systems are grouped in a rotating cycle of Work Weeks based on TS, PRA, and resource loading), (3) protected equipment schemes, and (4) an Electronic Risk Assessment Tool (ERAT) that calculates the CDF and LERF for equipment out of service and requires the implementation of risk management actions to reduce risk when risk-significant configurations are entered. The TS prohibits performing scheduled or planned maintenance that renders certain SSCs or combinations of SSCs unavailable. The risk assessment assumptions regarding SSC availability during the 30-day CT is consistent with the TS (e.g., during the 30-day CT, planned or discretionary maintenance that renders the EDGs and NSWS SSCs unavailable on either train of NSWS of either unit is prohibited).

Based on the above, the NRC staff finds that the licensee performed its Tier 2 risk evaluation in accordance with the guidance specified in RG 1.177 and is acceptable for use in this specific risk-informed application 3.3.1.3 Tier 3: Risk-Informed Configuration Risk Management Regulatory Guide 1.177 states that Tier 3 is the establishment of an overall CRMP to ensure that other potentially lower probability, but nonetheless risk-significant, configurations resulting from maintenance and other operational activities are identified and managed. Because the Maintenance Rule, as codified in 10 CFR 50.65(a)(4), requires licensees to assess and manage the potential increase in risk that may result from activities such as surveillance testing, and corrective and preventive maintenance, a licensee may use its existing Maintenance Rule program to satisfy Tier 3.

In the response to RAI 13, in its letter dated August 17, 2018, the licensee discusses how Tier 3 is met during the 30-day CT through the licensee's compliance with 10 CFR 50.65(a)(4). The licensee's CRMP meets the requirements of 10 CFR 50.65(a)(4) and minimizes plant risk through a blended approach of quantitative and qualitative assessments using ERAT. Plant risk is analyzed via a "look-ahead" of plant configurations over a specified period of time. Prior to entering the 30-day CT, the plant schedule is reviewed to identify and correct any significant potential risk impacts that may occur during the CT. During the CT, risk will be monitored and any emergent risk configurations will be addressed appropriately. The licensee's CRMP requires the implementation of risk management actions to help reduce risk when risk-significant configurations are entered. Thus, plant risk will be effectively managed prior to and during the 30-day CT.

Based on the above, the NRC staff finds that the licensee's Tier 3 CRMP is in accordance with the guidance specified in RG 1.177 and is acceptable for use in this specific risk-informed application.

3.3.1.4 Key Principle 4 Conclusions The licensee has demonstrated that the scope, level of detail, and technical acceptability of its PRA models are sufficient to support the proposed 30-day CT for NSWS Single Pond Return Header Operation. The risk metrics (i.e., ~CDF, ~LERF, ICCDP, and ICLERP) used to support the LAR meet the risk acceptance guidelines in RG 1.174 and RG 1.177. The NRC staff finds that the licensee has followed the three-tiered approach outlined in RG 1.177 to evaluate the

risk associated with the proposed TS change, and, therefore, the proposed change sufficiently addresses the fourth key principle of RG 1.177.

3.3.2 Key Principle 5: Performance Measurement Strategies - Implementation and Monitoring Program Regulatory Guide 1.177 states that a licensee is to use a three-tiered approach in implementing a proposed TS change such as the change proposed in this LAR. Application of the three-tiered approach is in keeping with the fundamental principle that the proposed change is consistent with the defense-in-depth philosophy. Application of the three-tiered approach provides assurance that defense-in-depth will not be significantly impacted by the proposed change.

Furthermore, RG 1.177 states that, to ensure that an extension of a TS CT does not degrade operational safety over time, the licensee should ensure, as part of its Maintenance Rule program (10 CFR 50.65), that when equipment does not meet its performance criteria, the evaluation required under the Maintenance Rule includes prior related TS changes in its scope.

Based on the NRC staff's review of the amendment request, as supplemented, the NRC staff finds that the licensee provided an acceptable evaluation of the proposed TS change against the three-tiered approach. In addition, the NSWS is monitored under the Catawba maintenance rule (MR) program. If the established MR reliability or availability performance criteria for the NSWS is exceeded, the performance is are evaluated for 10 CFR 50.65(a)(1) actions, which requires increased management attention and goal setting in order to restore the performance to an acceptable level. The NRC staff concludes that the implementation of the licensee's MR for NSWS sufficiently addresses the fifth key principle of RG 1.177.

3.4 Risk-Informed Considerations Summary The NRC staff finds that the risk impact of the proposed TS change as estimated by the change in risk (i.e., flCDF and flLERF) and incremental increase in risk (i.e., ICCDP and ICLERP) is consistent with the risk acceptance guidelines specified in RG 1.174 and RG 1.177. The licensee's methodology for assessing the risk impact utilized PRA models that are of sufficient scope and technically acceptable. The hazards which were not addressed using PRA (i.e.,

seismic and other external hazards) were determined not to impact this application. The NRC staff concludes that the licensee has followed the three-tiered approach in RG 1.177 and meet Key Principles 4 and 5 outlined in RG 1.177.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the NRC staff notified the South Carolina State official of the proposed issuance of the amendments on October 10, 2018. The State official confirmed on October 19, 2018, that the State of South Carolina had no comments.

5.0 PUBLIC COMMENTS On March 13, 2018 the NRC staff published a "Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing," in the Federal Register, regarding the license amendment request (83 FR 10914). In accordance with the requirements in 10 CFR 50.91, the notice provided a 30-day period for public comment on the proposed no significant hazards consideration determination. One comment from a member of the public was received,

however it was not related to the no significant hazards consideration determination nor the LAR. The comment can be found at www.regulations.gov, reference NRC-2018-0045-0001.

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on this finding (83 FR 10914; March 13, 2018). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22( c )(9). Pursuant to 10 CFR 51.22(b ), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

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.

Principal Contributors: G. Purciarello, NRR T. Hilsmeier, NRR T. Tjader, NRR Date: November 28, 2018

ML18275A278 *b memorandum +b E-mail OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DSS/SC PB/BC*

NAME MMahoney KGoldstein SAnderson DATE 10/19/18 10/11/18 & 11/27/18 10/2/18 OFFICE NRR/DRA/APLA/BC* NRR/DSS/STSB/BC* OGC-NLO+

NAME SRosenburg VCusamano DRoth DATE 9/26/18 10/31/18 11/27/18 OFFICE NRR/DORL/LPL2-1/BC NRR/DORL/LPL2-1/PM NAME MMarkley MMahoney DATE 11/28/18 11/28/18

v t e Letter Sequence Approval
CAC:MG0245, Control Room Habitability (Approved, Closed) CAC:MG0246, Control Room Habitability (Approved, Closed) EPID:L-2017-LLA-0297, Control Room Habitability (Approved, Closed)
Initiation Request Acceptance Supplement, Supplement, Supplement Administration Questions 9 March 2018 Answers 8 May 2018 17 August 2018 Results Approval Other: ML18117A187
MONTHYEAR2018-11-28 [Table View]

ML18275A278

Text

SUBJECT:

Dear Mr. Simril:

Enclosures:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

6.0 ENVIRONMENTAL CONSIDERATION

7.0 CONCLUSION

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